RA-23-0049, Enclosure 1 - Shearon Harris Nuclear Power Plant, Unit 1, Updated Final Safety Analysis Report, Amendment 65 - Redacted Version (Publicly Available Information) (2024)

[[:#Wiki_filter:U.S. Nuclear Regulatory Commission RA-23-0049 Enclosure 1 Shearon Harris Nuclear Power Plant Updated Final Safety Analysis Report, Amendment 65 -Redacted Version (Publicly Available Information)The Attachment to this enclosure contains only the documents incorporated by reference not required for redaction.

Updated Final Safety Analysis Report (FSAR)Shearon Harris Nuclear Power Plant Amendment 65

SHNPP FSAR Page 1 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.List of Effective Pages 86 65 Chapter 1 - Introduction and General Description of Plant Table of Contents 2 65 Chapter 1 70 65 Tables - List of Tables Table 1.1.1-1 15 64 Table 1.1.1-2 5 61 Table 1.1.1-3 1 63 Table 1.3.1-1 11 65 Table 1.3.2-1 2 65 Table 1.6-1 12 63 Table 1.6-2 1 61 Table 1.6-3 16 62 Table 1.6-4 1 64 Table 1.8-1 2 65 Figures - List of Figures Figure 1.1.1-1 1 61 Figure 1.2.2-2 1 61 Figure 1.2.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-4 (Deleted in Amendment 15) --- N/A Figure 1.2.2-5 (Deleted by Amendment 10) --- N/A Figure 1.2.2-6 (Deleted by Amendment 10) --- N/A Figure 1.2.2-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-8 (Deleted by Amendment 15) --- N/A Figure 1.2.2-9 (Deleted by Amendment 10) --- N/A Figure 1.2.2-10 (Deleted by Amendment 10) --- N/A Figure 1.2.2-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-12 (Deleted by Amendment 15) --- N/A Figure 1.2.2-13 (Deleted by Amendment 10) --- N/A Figure 1.2.2-14 (Deleted by Amendment 10) --- N/A Figure 1.2.2-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-16 (Deleted by Amendment 15) --- N/A Figure 1.2.2-17 (Deleted by Amendment 10) --- N/A

SHNPP FSAR Page 2 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 1.2.2-18 (Deleted by Amendment 10) --- N/A Figure 1.2.2-19 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-20 (Deleted by Amendment 15) --- N/A Figure 1.2.2-21 (Deleted by Amendment 10) --- N/A Figure 1.2.2-22 (Deleted by Amendment 10) --- N/A Figure 1.2.2-23 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-24 (Deleted by Amendment 15) --- N/A Figure 1.2.2-25 (Deleted by Amendment 10) --- N/A Figure 1.2.2-26 (Deleted by Amendment 10) --- N/A Figure 1.2.2-27 (Deleted by Amendment 55) --- N/A Figure 1.2.2-28 (Deleted by Amendment 15) --- N/A Figure 1.2.2-29 (Deleted by Amendment 10) --- N/A Figure 1.2.2-30 (Deleted by Amendment 10) --- N/A Figure 1.2.2-31 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 55 Figure 1.2.2-32 (Deleted by Amendment 15) --- N/A Figure 1.2.2-33 (Deleted by Amendment 10) --- N/A Figure 1.2.2-34 (Deleted by Amendment 10) --- N/A Figure 1.2.2-35 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-36 (Deleted by Amendment 15) --- N/A Figure 1.2.2-37 (Deleted by Amendment 10) --- N/A Figure 1.2.2-38 (Deleted by Amendment 10) --- N/A Figure 1.2.2-39 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-40 (Deleted by Amendment 15) --- N/A Figure 1.2.2-41 (Deleted by Amendment 10) --- N/A Figure 1.2.2-42 (Deleted by Amendment 10) --- N/A Figure 1.2.2-43 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-44 (Deleted by Amendment 15) --- N/A Figure 1.2.2-45 (Deleted by Amendment 10) --- N/A Figure 1.2.2-46 (Deleted by Amendment 10) --- N/A Figure 1.2.2-47 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-48 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-49 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-50 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-51 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-52 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-53 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 3 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 1.2.2-54 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-55 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-56 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-57 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-58 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-59 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-59A (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-60 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-61 (Deleted by Amendment 15) --- N/A Figure 1.2.2-62 (Deleted by Amendment 10) --- N/A Figure 1.2.2-63 (Deleted by Amendment 10) --- N/A Figure 1.2.2-64 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-65 (Deleted by Amendment 15) --- N/A Figure 1.2.2-66 (Deleted by Amendment 10) --- N/A Figure 1.2.2-67 (Deleted by Amendment 10) --- N/A Figure 1.2.2-68 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-69 (Deleted by Amendment 15) --- N/A Figure 1.2.2-70 (Deleted by Amendment 10) --- N/A Figure 1.2.2-71 (Deleted by Amendment 10) --- N/A Figure 1.2.2-72 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-73 (Deleted by Amendment 15) --- N/A Figure 1.2.2-74 (Deleted by Amendment 10) --- N/A Figure 1.2.2-75 (Deleted by Amendment 10) --- N/A Figure 1.2.2-76 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-77 (Deleted by Amendment 15) --- N/A Figure 1.2.2-78 (Deleted by Amendment 10) --- N/A Figure 1.2.2-79 (Deleted by Amendment 10) --- N/A Figure 1.2.2-80 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-81 (Deleted by Amendment 15) --- N/A Figure 1.2.2-82 (Deleted by Amendment 10) --- N/A Figure 1.2.2-83 (Deleted by Amendment 10) --- N/A Figure 1.2.2-84 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-85 (Deleted by Amendment 10) --- N/A Figure 1.2.2-86 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.2.2-87 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 1.5.2-1 (Deleted by Amendment 48) --- N/A

SHNPP FSAR Page 4 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Chapter 2 - Site Characteristics Table of Contents 10 65 Chapter 2 369 65 Tables - List of Tables Table 2.1.2-1 1 61 Table 2.1.3-1 (Deleted by Amendment 59) --- N/A Table 2.1-3-2 (Deleted by Amendment 59) --- N/A Table 2.1.3-3 (Deleted by Amendment 59) --- N/A Table 2.1.3-4 (Deleted by Amendment 59) --- N/A Table 2.1.3-5 (Deleted by Amendment 59) --- N/A Table 2.1.3-6 1 61 Table 2.2.2-1 1 61 Table 2.2.2-2 1 61 Table 2.2.3-1 1 61 Table 2.2.3-2 1 61 Table 2.2.3-3 1 61 Table 2.2.3-4 2 65 Table 2.2.3-5 1 65 Table 2.3.1-1 1 61 Table 2.3.1-2 1 61 Table 2.3.1-3 1 61 Table 2.3.1-4 2 61 Table 2.3.1-5 1 61 Table 2.3.1-6 1 61 Table 2.3.1-7 1 61 Table 2.3.2-1 7 61 Table 2.3.2-2 17 61 Table 2.3.2-3 1 61 Table 2.3.2-4 1 61 Table 2.3.2-5 1 61 Table 2.3.2-6 1 61 Table 2.3.2-7 1 61 Table 2.3.2-8 1 61 Table 2.3.2-9 1 61 Table 2.3.2-10 1 61

SHNPP FSAR Page 5 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.3.2-11 1 61 Table 2.3.2-12 1 61 Table 2.3.2-13 1 61 Table 2.3.2-14 1 61 Table 2.3.2-15 1 61 Table 2.3.2-16 1 61 Table 2.3.2-17 1 61 Table 2.3.2-18 1 61 Table 2.3.2-19 1 61 Table 2.3.2-20 1 61 Table 2.3.2-21 1 61 Table 2.3.2-22 1 61 Table 2.3.2-23 1 61 Table 2.3.2-24 1 61 Table 2.3.2-25 9 61 Table 2.3.2-26 9 61 Table 2.3.3-1 (Deleted by Amendment 51) --- N/A Table 2.3.3-2 (Deleted by Amendment 51) --- N/A Table 2.3.3-3 1 61 Table 2.3.3-4 1 61 Table 2.3.3-5 (Deleted by Amendment 51) --- N/A Table 2.3.3-5a (Deleted by Amendment 51) --- N/A Table 2.3.3-5b (Deleted by Amendment 51) --- N/A Table 2.3.3-6 8 61 Table 2.3.3-7 8 61 Table 2.3.3-8 8 61 Table 2.3.3-9 8 61 Table 2.3.3-10 8 61 Table 2.3.3-11 8 61 Table 2.3.3-12 8 61 Table 2.3.3-13 8 61 Table 2.3.3-14 8 61 Table 2.3.3-15 8 61 Table 2.3.3-16 95 61 Table 2.3.3-17 97 61 Table 2.3.3-18 1 61

SHNPP FSAR Page 6 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.3.4-1 1 61 Table 2.3.4-2 1 61 Table 2.3.4-3 1 61 Table 2.3.4-4 1 61 Table 2.3.4-5 1 61 Table 2.3.5-1 1 61 Table 2.3.5-2 1 61 Table 2.3.5-3 1 61 Table 2.3.5-4 2 61 Table 2.3.5-5 2 61 Table 2.3.5-6 2 61 Table 2.3.5-7 2 61 Table 2.4.1-1 2 61 Table 2.4.1-2 1 61 Table 2.4.1-3 2 61 Table 2.4.1-4 1 61 Table 2.4.1-5 1 61 Table 2.4.1-6 1 61 Table 2.4.2-1 1 61 Table 2.4.2-2 1 61 Table 2.4.2-3 1 61 Table 2.4.2-4 1 61 Table 2.4.3-1 1 61 Table 2.4.3-2 1 61 Table 2.4.3-3 2 61 Table 2.4.5-1 1 61 Table 2.4.5-2 1 61 Table 2.4.11-1 (Deleted by Amendment 15) --- N/A Table 2.4.11-1a (Deleted by Amendment 15) --- N/A Table 2.4.11-2 1 61 Table 2.4.11-3 1 61 Table 2.4.11-4 (Deleted by Amendment 15) --- N/A Table 2.4.11-5 (Deleted by Amendment 15) --- N/A Table 2.4.11-6 (Deleted by Amendment 15) --- N/A Table 2.4.11-7 (Deleted by Amendment 15) --- N/A Table 2.4.11-8 (Deleted by Amendment 15) --- N/A

SHNPP FSAR Page 7 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.4.11-9 (Deleted by Amendment 15) --- N/A Table 2.4.11-10 1 61 Table 2.4.11-11 (Deleted by Amendment 15) --- N/A Table 2.4.11-12 1 61 Table 2.4.11-13 (Deleted by Amendment 15) --- N/A Table 2.4.11-14 (Deleted by Amendment 46) --- N/A Table 2.4.11-15 (Deleted by Amendment 46) --- N/A Table 2.4.11-16 (Deleted by Amendment 15) --- N/A Table 2.4.11-17 (Deleted by Amendment 15) --- N/A Table 2.4.11-18 1 61 Table 2.4.11-19 1 61 Table 2.4.11-20 1 61 Table 2.4.12-1 1 61 Table 2.4.13-1 2 61 Table 2.4.13-2 2 61 Table 2.4.13-3 1 61 Table 2.4.13-4 1 61 Table 2.4.13-5 1 61 Table 2.4.13-6 1 61 Table 2.4.13-7 1 61 Table 2.4.13-8 1 61 Table 2.5.2-1 13 61 Table 2.5.2-2 1 61 Table 2.5.2-3 1 61 Table 2.5.2-4 1 61 Table 2.5.4-1 1 61 Table 2.5.4-2 2 61 Table 2.5.4-3 1 61 Table 2.5.4-4 1 61 Table 2.5.4-5 1 61 Table 2.5.4-6 1 61 Table 2.5.4-7 1 61 Table 2.5.4-8 1 61 Table 2.5.4-9 1 61 Table 2.5.6-1 1 61 Table 2.5.6-2 1 61

SHNPP FSAR Page 8 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.5.6-3 1 61 Table 2.5B-1 1 61 Table 2.5B-2 1 61 Table 2.5B-3 1 61 Table 2.5B-4 1 61 Table 2.5C.2-1 1 61 Table 2.5C.2-2 2 61 Table 2.5C.2-3 1 61 Table 2.5C.2-4 1 61 Table 2.5C.2-5 1 61 Table 2.5C.2-6 1 61 Table 2.5C.2-7 1 61 Table 2.5C.2-8 1 61 Table 2.5C.2-9 1 61 Table 2.5C.2-10 1 61 Table 2.5C.2-11 1 61 Table 2.5C.2-12 1 61 Table 2.5C.2-13 1 61 Table 2.5C.2-14 1 61 Table 2.5C.2-15 1 61 Table 2.5C.2-16 1 61 Table 2.5C.2-17 1 61 Table 2.5C.2-18 1 61 Table 2.5C.2-19 1 61 Table 2.5C.2-20 1 61 Table 2.5C.2-21 1 61 Table 2.5C.2-22 1 61 Table 2.5C.2-23 1 61 Table 2.5C.2-24 1 61 Table 2.5C.2-25 1 61 Table 2.5C.2-26 1 61 Table 2.5C.3-1 1 61 Table 2.5C.3-2 1 61 Table 2.5C.3-3 1 61 Table 2.5D-1 1 61 Table 2.5D-2 1 61

SHNPP FSAR Page 9 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.5D-3 1 61 Table 2.5D-4 1 61 Table 2.5D-5 1 61 Table 2.5D-6 1 61 Table 2.5D-7 1 61 Table 2.5D-8 1 61 Table 2.5D-9 1 61 Table 2.5D-10 1 61 Table 2.5D-11 1 61 Table 2.5D-12 1 61 Table 2.5D-13 1 61 Table 2.5D-14 1 61 Table 2.5D-15 1 61 Table 2.5D-16 1 61 Table 2.5D-17 1 61 Table 2.5D-18 1 61 Table 2.5D-19 1 61 Table 2.5D-20 1 61 Table 2.5D-21 1 61 Table 2.5D-22 1 61 Table 2.5D-23 1 61 Table 2.5D-24 1 61 Table 2.5D-25 1 61 Table 2.5D-26 1 61 Table 2.5D-27 1 61 Table 2.5D-28 1 61 Table 2.5D-29 1 61 Table 2.5D-30 1 61 Table 2.5D-31 1 61 Table 2.5D-32 1 61 Table 2.5D-33 1 61 Table 2.5D-34 1 61 Table 2.5D-35 1 61 Table 2.5D-36 1 61 Table 2.5D-37 1 61 Table 2.5D-38 1 61

SHNPP FSAR Page 10 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 2.5D-39 1 61 Table 2.5D-40 1 61 Table 2.5D-41 1 61 Table 2.5D-42 1 61 Table 2.5D-43 1 61 Table 2.5D-44 1 61 Table 2.5D-45 1 61 Table 2.5D-46 1 61 Table 2.5D-47 1 61 Table 2.5D-48 1 61 Table 2.5D-49 1 61 Figures - List of Figures Figure 2.1.1-1 1 61 Figure 2.1.2-1 1 61 Figure 2.1.3-1 1 61 Figure 2.1.3-2 1 61 Figure 2.1.3-3 1 61 Figure 2.1.3-4 1 61 Figure 2.1.3-5 1 61 Figure 2.1.3-6 1 61 Figure 2.2.2-1 1 61 Figure 2.2.3-1 1 61 Figure 2.2.3-2 1 61 Figure 2.2.3-3 1 61 Figure 2.2.3-4 1 61 Figure 2.2.3-5 1 61 Figure 2.2.3-6 1 61 Figure 2.2.3-7 1 61 Figure 2.2.3-8 1 61 Figure 2.3.1-1 1 61 Figure 2.3.1-2 1 61 Figure 2.3.2-1 1 61 Figure 2.3.2-2 1 61 Figure 2.3.2-3 1 61 Figure 2.3.2-4 1 61 Figure 2.3.2-5 1 61

SHNPP FSAR Page 11 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.3.2-6 1 61 Figure 2.3.2-7 1 61 Figure 2.3.2-8 1 61 Figure 2.3.2-9 1 61 Figure 2.3.2-10 1 61 Figure 2.3.2-11 1 61 Figure 2.3.2-12 1 61 Figure 2.3.2-13 1 61 Figure 2.3.2-14 1 61 Figure 2.3.2-15 1 61 Figure 2.3.2-16 1 61 Figure 2.3.2-17 1 61 Figure 2.3.2-18 1 61 Figure 2.3.2-19 1 61 Figure 2.3.3-1 1 61 Figure 2.3.4-1 1 61 Figure 2.3.4-2 1 61 Figure 2.3.4-3 1 61 Figure 2.4.1-1 1 61 Figure 2.4.1-2 1 61 Figure 2.4.1-3 1 61 Figure 2.4.1-4 1 61 Figure 2.4.1-5 1 61 Figure 2.4.1-6 1 61 Figure 2.4.1-7 1 61 Figure 2.4.1-8 1 61 Figure 2.4.2-1 1 61 Figure 2.4.2-2 1 61 Figure 2.4.2-3 1 61 Figure 2.4.2-4 1 61 Figure 2.4.2-5 1 61 Figure 2.4.3-1 1 61 Figure 2.4.3-2 1 61 Figure 2.4.3-3 1 61 Figure 2.4.3-4 1 61 Figure 2.4.3-5 1 61

SHNPP FSAR Page 12 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.4.3-6 1 61 Figure 2.4.3-7 1 61 Figure 2.4.3-8 1 61 Figure 2.4.3-9 1 61 Figure 2.4.3-10 1 61 Figure 2.4.3-11 1 61 Figure 2.4.3-12 (Deleted by Amendment 10) --- N/A Figure 2.4.5-1 1 61 Figure 2.4.8-1 1 61 Figure 2.4.11-1 (Deleted by Amendment 15) --- N/A Figure 2.4.11-2 1 61 Figure 2.4.11-3 1 61 Figure 2.4.11-4 1 61 Figure 2.4.11-5 1 61 Figure 2.4.11-6 (Deleted by Amendment 15) --- N/A Figure 2.4.11-7 (Deleted by Amendment 15) --- N/A Figure 2.4.11-8 (Deleted by Amendment 15) --- N/A Figure 2.4.11-9 1 61 Figure 2.4.13-1 1 61 Figure 2.4.13-2 1 61 Figure 2.4.13-3 1 61 Figure 2.4.13-4 1 61 Figure 2.5.1-1 1 61 Figure 2.5.1-2 1 61 Figure 2.5.1-3 1 61 Figure 2.5.1-3a 1 61 Figure 2.5.1-4 1 61 Figure 2.5.1-5 1 61 Figure 2.5.1-6 1 61 Figure 2.5.1-7 1 61 Figure 2.5.1-8 1 61 Figure 2.5.1-9 1 61 Figure 2.5.1-10 1 61 Figure 2.5.1-11 1 61 Figure 2.5.1-12 1 61 Figure 2.5.1-13 1 61

SHNPP FSAR Page 13 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.1-14 1 61 Figure 2.5.1-15 1 61 Figure 2.5.1-16 1 61 Figure 2.5.2-1 1 61 Figure 2.5.2-1a 1 61 Figure 2.5.2-2 1 61 Figure 2.5.2-3 1 61 Figure 2.5.2-4 1 61 Figure 2.5.2-5 1 61 Figure 2.5.2-6 1 61 Figure 2.5.2-7 1 61 Figure 2.5.2-8 1 61 Figure 2.5.2-9 1 61 Figure 2.5.2-10 1 61 Figure 2.5.2-11 1 61 Figure 2.5.2-12 1 61 Figure 2.5.2-13 1 61 Figure 2.5.2-14 1 61 Figure 2.5.2-15 1 61 Figure 2.5.2-16 1 61 Figure 2.5.2-17 1 61 Figure 2.5.2-18 1 61 Figure 2.5.3-1 1 61 Figure 2.5.3-2 1 61 Figure 2.5.3-3 1 61 Figure 2.5.3-4 1 61 Figure 2.5.3-5 1 61 Figure 2.5.3-6 1 61 Figure 2.5.3-7 1 61 Figure 2.5.3-8 1 61 Figure 2.5.3-9 1 61 Figure 2.5.3-10 1 61 Figure 2.5.3-11 1 61 Figure 2.5.3-12 1 61 Figure 2.5.3-13 1 61 Figure 2.5.3-14 1 61

SHNPP FSAR Page 14 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.3-15 1 61 Figure 2.5.3-16 1 61 Figure 2.5.3-17 1 61 Figure 2.5.3-18 1 61 Figure 2.5.4-1 1 61 Figure 2.5.4-2 1 61 Figure 2.5.4-3 1 61 Figure 2.5.4-4 1 61 Figure 2.5.4-5 1 61 Figure 2.5.4-6 1 61 Figure 2.5.4-7 1 61 Figure 2.5.4-8 1 61 Figure 2.5.4-9 1 61 Figure 2.5.4-10 1 61 Figure 2.5.4-11 1 61 Figure 2.5.4-12 1 61 Figure 2.5.4-13 1 61 Figure 2.5.4-14 1 61 Figure 2.5.4-15 1 61 Figure 2.5.4-16 1 61 Figure 2.5.4-17 1 61 Figure 2.5.4-18 1 61 Figure 2.5.4-19 1 61 Figure 2.5.4-20 1 61 Figure 2.5.4-21 1 61 Figure 2.5.4-22 1 61 Figure 2.5.4-23 1 61 Figure 2.5.4-24 1 61 Figure 2.5.4-25 1 61 Figure 2.5.4-26 1 61 Figure 2.5.4-27 1 61 Figure 2.5.4-28 1 61 Figure 2.5.4-29 1 61 Figure 2.5.4-30 1 61 Figure 2.5.4-31 1 61 Figure 2.5.4-32 1 61

SHNPP FSAR Page 15 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.4-33 1 61 Figure 2.5.4-34 1 61 Figure 2.5.4-35 1 61 Figure 2.5.4-36 1 61 Figure 2.5.4-37 1 61 Figure 2.5.4-38 1 61 Figure 2.5.4-39 1 61 Figure 2.5.4-40 1 61 Figure 2.5.4-41 1 61 Figure 2.5.4-42 1 61 Figure 2.5.4-43 1 61 Figure 2.5.4-44 1 61 Figure 2.5.4-45 1 61 Figure 2.5.4-46 1 61 Figure 2.5.4-47 1 61 Figure 2.5.4-48 1 61 Figure 2.5.4-49 1 61 Figure 2.5.4-50 1 61 Figure 2.5.4-51 1 61 Figure 2.5.4-52 1 61 Figure 2.5.4-53 1 61 Figure 2.5.4-54 1 61 Figure 2.5.4-55 1 61 Figure 2.5.4-56 1 61 Figure 2.5.4-57 1 61 Figure 2.5.4-58 1 61 Figure 2.5.4-59 1 61 Figure 2.5.4-60 1 61 Figure 2.5.4-61 1 61 Figure 2.5.4-62 1 61 Figure 2.5.4-63 1 61 Figure 2.5.4-64 1 61 Figure 2.5.4-65 1 61 Figure 2.5.4-66 1 61 Figure 2.5.4-67 1 61 Figure 2.5.4-68 1 61

SHNPP FSAR Page 16 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.4-69 1 61 Figure 2.5.4-70 1 61 Figure 2.5.4-71 1 61 Figure 2.5.4-72 1 61 Figure 2.5.4-73 1 61 Figure 2.5.4-74 1 61 Figure 2.5.4-75 1 61 Figure 2.5.4-76 1 61 Figure 2.5.4-77 1 61 Figure 2.5.4-78 1 61 Figure 2.5.4-79 1 61 Figure 2.5.4-80 1 61 Figure 2.5.4-81 1 61 Figure 2.5.4-82 1 61 Figure 2.5.4-83 1 61 Figure 2.5.4-84 1 61 Figure 2.5.4-85 1 61 Figure 2.5.4-86 1 61 Figure 2.5.4-87 1 61 Figure 2.5.4-88 1 61 Figure 2.5.4-89 1 61 Figure 2.5.4-90 1 61 Figure 2.5.4-91 1 61 Figure 2.5.4-92 1 61 Figure 2.5.4-93 1 61 Figure 2.5.4-94 1 61 Figure 2.5.4-95 1 61 Figure 2.5.4-96 1 61 Figure 2.5.4-97 1 61 Figure 2.5.4-98 1 61 Figure 2.5.4-99 (Deleted by Amendment 21) --- N/A Figure 2.5.4-100 (Deleted by Amendment 21) --- N/A Figure 2.5.4-101 (Deleted by Amendment 21) --- N/A Figure 2.5.4-102 (Deleted by Amendment 21) --- N/A Figure 2.5.4-103 (Deleted by Amendment 21) --- N/A Figure 2.5.4-104 (Deleted by Amendment 21) --- N/A

SHNPP FSAR Page 17 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.4-105 (Deleted by Amendment 21) --- N/A Figure 2.5.4-106 1 61 Figure 2.5.4-107 1 61 Figure 2.5.4-108 1 61 Figure 2.5.4-109 1 61 Figure 2.5.4-110 1 61 Figure 2.5.4-111 1 61 Figure 2.5.4-112 1 61 Figure 2.5.4-113 1 61 Figure 2.5.4-114 (Deleted by Amendment 21) --- N/A Figure 2.5.4-115 (Deleted by Amendment 21) --- N/A Figure 2.5.4-116 (Deleted by Amendment 21) --- N/A Figure 2.5.4-117 (Deleted by Amendment 21) --- N/A Figure 2.5.4-118 (Deleted by Amendment 21) --- N/A Figure 2.5.4-119 (Deleted by Amendment 21) --- N/A Figure 2.5.4-120 1 61 Figure 2.5.4-121 1 61 Figure 2.5.4-122 1 61 Figure 2.5.4-123 1 61 Figure 2.5.4-124 1 61 Figure 2.5.4-125 1 61 Figure 2.5.4-126 1 61 Figure 2.5.4-127 1 61 Figure 2.5.4-128 1 61 Figure 2.5.4-129 1 61 Figure 2.5.4-130 1 61 Figure 2.5.4-131 1 61 Figure 2.5.6-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-9 1 61

SHNPP FSAR Page 18 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 2.5.6-10 1 61 Figure 2.5.6-11 1 61 Figure 2.5.6-12 1 61 Figure 2.5.6-13 1 61 Figure 2.5.6-14 1 61 Figure 2.5.6-15 1 61 Figure 2.5.6-16 1 61 Figure 2.5.6-17 1 61 Figure 2.5.6-18 1 61 Figure 2.5.6-19 1 61 Figure 2.5.6-20 1 61 Figure 2.5.6-21 1 61 Figure 2.5.6-22 1 61 Figure 2.5.6-23 1 61 Figure 2.5.6-24 1 61 Figure 2.5.6-25 1 61 Figure 2.5.6-26 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5.6-27 1 61 Figure 2.5.6-28 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 2.5K-1 1 61 Figure 2.5K-2 1 61 Figure 2.5K-3 1 61 Figure 2.5K-4 1 61 Figure 2.5K-5 1 61 Figure 2.5K-6 1 61 Figure 2.5K-7 1 61 Figure 2.5K-8 1 61 Figure 2.5K-9 1 61 Chapter 3 - Design of Structures, Components, Equipment and Systems Table of Contents 14 65 Chapter 3 510 65 Tables - List of Tables Table 3.2.1-1 27 65 Table 3.3.0-1 1 61 Table 3.3.1-1 1 61

SHNPP FSAR Page 19 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 3.3.1-2 1 61 Table 3.4.1-1 1 61 Table 3.5.1-1 8 65 Table 3.5.1-2 2 63 Table 3.5.1-2a 2 63 Table 3.5.1-3 1 61 Table 3.5.1-4 1 63 Table 3.5.1-5 1 61 Table 3.5.1-6 1 61 Table 3.5.1-7 1 61 Table 3.5.1-8 (Deleted by Amendment 62) --- N/A Table 3.5.1-9 (Deleted by Amendment 62) --- N/A Table 3.5.1-10 (Deleted by Amendment 62) --- N/A Table 3.5.1-11 (Deleted by Amendment 62) --- N/A Table 3.5.1-12 (Deleted by Amendment 62) --- N/A Table 3.5.1-13 (Deleted by Amendment 62) --- N/A Table 3.5.1-14 (Deleted by Amendment 62) --- N/A Table 3.5.1-15 (Deleted by Amendment 58) --- N/A Table 3.5.1-16 (Deleted by Amendment 58) --- N/A Table 3.5.1-17 1 61 Table 3.5.2-1 2 61 Table 3.5.3-1 1 65 Table 3.5.3-2 1 65 Table 3.5.3-3 1 61 Table 3.6.2-1 1 61 Table 3.6A-1 1 61 Table 3.6A-2 1 61 Table 3.6A-3 1 61 Table 3.6A-4 1 63 Table 3.6A-5 2 61 Table 3.6A-6 1 61 Table 3.6A-7 1 61 Table 3.6A-8 1 61 Table 3.6A-9 1 61 Table 3.6A-10 1 61 Table 3.6A-11 2 61

SHNPP FSAR Page 20 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 3.6A-12 1 61 Table 3.6A-13 1 63 Table 3.6A-14 1 61 Table 3.6A-15 4 65 Table 3.6A-16 1 61 Table 3.6A-17.1 1 61 Table 3.6A-17.2 1 61 Table 3.6A-17.3 1 61 Table 3.6A-18 3 65 Table 3.6A-19 1 61 Table 3.6A-20 1 61 Table 3.6A-21 1 65 Table 3.7.1-1 1 61 Table 3.7.1-2 1 61 Table 3.7.2-1 1 65 Table 3.7.2-2 1 61 Table 3.7.2-3 1 61 Table 3.7.2-4 1 61 Table 3.7.2-5 1 61 Table 3.7.2-6 1 61 Table 3.7.2-7 3 61 Table 3.7.2-8 3 61 Table 3.7.2-9 2 61 Table 3.7.2-10 3 61 Table 3.7.2-11 2 61 Table 3.7.2-12 2 61 Table 3.7.2-13 4 61 Table 3.7.2-14 1 61 Table 3.7.2-15 3 61 Table 3.7.2-16 1 61 Table 3.7.2-17 1 61 Table 3.7.2-18 1 61 Table 3.7.3-1 1 61 Table 3.7.3-2 2 61 Table 3.7.3-2A 1 61 Table 3.7.3-2B 1 61

SHNPP FSAR Page 21 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 3.7.3-2C 1 61 Table 3.7.3-2D 1 61 Table 3.7.4-1 1 64 Table 3.8.1-1 1 61 Table 3.8.1-2 1 61 Table 3.8.1-3 1 61 Table 3.8.1-4 1 65 Table 3.8.1-5 1 61 Table 3.8.1-6 1 61 Table 3.8.1-7 1 65 Table 3.8.1-8 1 61 Table 3.8.1-9 1 61 Table 3.8.1-10 1 61 Table 3.8.1-11 (Deleted by Amendment 27) --- N/A Table 3.8.1-12 1 61 Table 3.8.1-13 1 61 Table 3.8.1-14 1 61 Table 3.8.1-15 1 61 Table 3.8.1-16 1 61 Table 3.8.2-1 1 61 Table 3.8.2-2 1 61 Table 3.8.2-3 1 61 Table 3.8.2-4 1 61 Table 3.8.2-5 1 61 Table 3.8.2-6 1 61 Table 3.8.3-1 1 61 Table 3.8.4-1 1 61 Table 3.8.4-2 2 61 Table 3.8.4-3 1 61 Table 3.8A-1 1 61 Table 3.9.1-1 2 61 Table 3.9.1-2 1 61 Table 3.9.1-2a 1 61 Table 3.9.1-3 1 61 Table 3.9.1-4 1 61 Table 3.9.1-5 1 61

SHNPP FSAR Page 22 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 3.9.2-1 8 63 Table 3.9.3-1 1 61 Table 3.9.3-2 1 65 Table 3.9.3-3 1 65 Table 3.9.3-4 1 65 Table 3.9.3-5 1 65 Table 3.9.3-6 1 61 Table 3.9.3-7 1 61 Table 3.9.3-7a 1 61 Table 3.9.3-8 1 65 Table 3.9.3-9 1 61 Table 3.9.3-10 1 61 Table 3.9.3-11 1 61 Table 3.9.3-12 1 61 Table 3.9.3-13 7 61 Table 3.9.3-14 22 65 Table 3.9.3-15 1 61 Table 3.9.3-16 (Deleted by Amendment 56) --- N/A Table 3.9.5-1 1 61 Table 3.9C.12-1 1 61 Table 3.9C.12-2 1 61 Table 3.9C.14-1 1 61 Table 3.9C.14-2 1 61 Table 3.9C.14-3 1 61 Table 3.9C.14-4 1 61 Table 3.9C.14-5 1 61 Table 3.9C.14-6 1 61 Table 3.9C.14-7 1 61 Table 3.9C.14-8 1 61 Table 3.9C.14-9 1 61 Table 3.9C.14-10 1 61 Table 3.9C.14-11 1 61 Table 3.9C.14-12 1 61 Table 3.9C.14-13 1 61 Table 3.9C.14-14 1 61 Table 3.9C.14-15 1 61

SHNPP FSAR Page 23 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 3.9C.14-16 1 61 Table 3.9C.14-17 1 61 Table 3.9C.14-18 1 61 Table 3.9C.14-19 1 61 Table 3.9C.14-20 1 61 Table 3.9C.15-1 1 65 Table 3.10.1-1 4 61 Table 3.10.1-2 4 61 Table 3.11.0-1 2 62 Table 3.11.0-2 2 61 Table 3.11.0-3 1 61 Table 3.11.1-1 3 61 Table 3.11C-1 1 61 Table 3.11E-1 1 61 Table 3.11E-2 1 61 Figures - List of Figures Figure 3.3.1-1 1 61 Figure 3.3.1-2 1 61 Figure 3.3.2-1 1 61 Figure 3.3.2-2 1 61 Figure 3.3.2-3 1 61 Figure 3.3.2-4 1 61 Figure 3.3.2-5 1 61 Figure 3.4.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.4.1-2 1 61 Figure 3.4.1-3 1 61 Figure 3.4.1-4 1 61 Figure 3.4.1-5 (Deleted by Amendment 15) --- N/A Figure 3.4.1-6 1 61 Figure 3.4.1-7 1 61 Figure 3.4.1-8 1 61 Figure 3.4.1-9 1 61 Figure 3.4.1-10 2 61 Figure 3.4.1-11 2 61 Figure 3.4.1-12 2 61 Figure 3.4.1-13 1 61

SHNPP FSAR Page 24 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.4.1-14 1 61 Figure 3.4.1-15 1 61 Figure 3.4.1-16 1 61 Figure 3.5.1-1 1 61 Figure 3.5.1-2 (Deleted by Amendment 15) --- N/A Figure 3.5.1-3 1 61 Figure 3.5.1-4 (Deleted by Amendment 62) --- N/A Figure 3.5.1-4a (Deleted by Amendment 62) --- N/A Figure 3.5.1-5 (Deleted by Amendment 62) --- N/A Figure 3.6.1-1 1 61 Figure 3.6.2-1 1 61 Figure 3.6.2-2 1 61 Figure 3.6.2-3 1 61 Figure 3.6.2-4 1 61 Figure 3.6.2-5 1 61 Figure 3.6.2-6 1 61 Figure 3.6.2-7 1 61 Figure 3.6.2-8 1 61 Figure 3.6.2-9 (Deleted by Amendment 51) --- N/A Figure 3.6A-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-1-CALC 3 61 Figure 3.6A-1-PLOT-A 2 61 Figure 3.6A-1-PLOT-B 2 61 Figure 3.6A-1-PLOT-C 2 61 Figure 3.6A-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-3 (Deleted by Amendment 27) --- N/A Figure 3.6A-4 (Deleted by Amendment 27) --- N/A Figure 3.6A-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-5-CALC 3 61 Figure 3.6A-5-PLOT-A 3 61 Figure 3.6A-5-PLOT-B 3 61 Figure 3.6A-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-6-CALC 1 61 Figure 3.6A-6-PLOT-A 3 61 Figure 3.6A-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-7-PLOT-C 2 61

SHNPP FSAR Page 25 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.6A-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-8.1-CALC 3 61 Figure 3.6A-8.1-PLOT-A 1 61 Figure 3.6A-8.1-PLOT-B 1 61 Figure 3.6A-8.1-PLOT-C 2 61 Figure 3.6A-8.2-CALC 2 61 Figure 3.6A-8.2-PLOT-A 3 61 Figure 3.6A-8.2-PLOT-B 3 61 Figure 3.6A-9 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-9-CALC 3 61 Figure 3.6A-9-PLOT-A 3 61 Figure 3.6A-9-PLOT-B 2 61 Figure 3.6A-9-PLOT-C 2 61 Figure 3.6A-10 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-10-CALC 3 61 Figure 3.6A-10-PLOT-A 1 61 Figure 3.6A-10-PLOT-B 1 61 Figure 3.6A-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-12 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-12-CALC 7 63 Figure 3.6A-12-PLOT-A 4 61 Figure 3.6A-12-PLOT-B 2 61 Figure 3.6A-12-PLOT-C 1 63 Figure 3.6A-12-PLOT-D 2 63 Figure 3.6A-12-PLOT-E 2 61 Figure 3.6A-12-PLOT-F 1 63 Figure 3.6A-12-PLOT-G 1 63 Figure 3.6A-13 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-14-CALC 3 61 Figure 3.6A-15 1 61 Figure 3.6A-15-CALC 3 61 Figure 3.6A-16 (Deleted by Amendment 45) --- N/A Figure 3.6A-16-CALC (Deleted by Amendment 45) --- N/A Figure 3.6A-17 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-17-CALC 1 61

SHNPP FSAR Page 26 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.6A-17-PLOT-A 3 61 Figure 3.6A-18 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-18-CALC 1 61 Figure 3.6A-18-PLOT-A 2 61 Figure 3.6A-19 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-19-CALC (Deleted by Amendment 27) --- N/A Figure 3.6A-19-PLOT-A (Deleted by Amendment 27) --- N/A Figure 3.6A-20 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-20-CALC (Deleted by Amendment 49) --- N/A Figure 3.6A-20-PLOT-A (Deleted by Amendment 51) --- N/A Figure 3.6A-20-PLOT-B (Deleted by Amendment 51) --- N/A Figure 3.6A-20-PLOT-C (Deleted by Amendment 51) --- N/A Figure 3.6A-20.1-CALC 3 61 Figure 3.6A-20.2 CALC 1 61 Figure 3.6A-20.3 CALC 1 61 Figure 3.6A-21 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-21-CALC 2 61 Figure 3.6A-22 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-23 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-23-CALC 1 61 Figure 3.6A-24 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-24-CALC 3 61 Figure 3.6A-24-PLOT-A 1 61 Figure 3.6A-24-PLOT-B 1 61 Figure 3.6A-24-PLOT-C 1 61 Figure 3.6A-25 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-25-CALC 3 63 Figure 3.6A-25-PLOT-A 1 63 Figure 3.6A-25-PLOT-B 1 63 Figure 3.6A-25-PLOT-C 1 63 Figure 3.6A-26 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-26-CALC 1 61 Figure 3.6A-26-PLOT-A 3 61 Figure 3.6A-27 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.6A-28 (Deleted by Amendment 27) --- N/A Figure 3.6A-29 1 61

SHNPP FSAR Page 27 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.6A-30 1 61 Figure 3.6A-31 1 61 Figure 3.6A-32 1 61 Figure 3.6A-32-PLOT-A 3 61 Figure 3.6A-32.1 1 61 Figure 3.6A-32.2 1 63 Figure 3.6A-32.2-CALC 1 61 Figure 3.6A-32.2-PLOT-A 1 61 Figure 3.6A-33 1 61 Figure 3.6A-33.1 1 61 Figure 3.6A-34 1 61 Figure 3.6A-34a 1 61 Figure 3.6A-34b 1 61 Figure 3.6A-35 1 61 Figure 3.6A-36 (Deleted by Amendment 39) --- N/A Figure 3.6A-37 (Deleted by Amendment 39) --- N/A Figure 3.6A-38 (Deleted by Amendment 39) --- N/A Figure 3.6A-39 (Deleted by Amendment 39) --- N/A Figure 3.6A-40 1 61 Figure 3.6A-41 1 61 Figure 3.6A-42 1 61 Figure 3.7.1-1 1 61 Figure 3.7.1-2 1 61 Figure 3.7.1-3 1 61 Figure 3.7.1-4 1 61 Figure 3.7.1-5 1 61 Figure 3.7.1-6 1 61 Figure 3.7.1-7 1 61 Figure 3.7.1-8 1 61 Figure 3.7.1-9 1 61 Figure 3.7.1-10 1 61 Figure 3.7.1-11 1 61 Figure 3.7.1-12 1 61 Figure 3.7.1-13 1 61 Figure 3.7.1-14 1 61 Figure 3.7.1-15 1 61

SHNPP FSAR Page 28 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.7.1-16 1 61 Figure 3.7.1-17 1 61 Figure 3.7.1-18 1 61 Figure 3.7.1-19 1 61 Figure 3.7.1-20 1 61 Figure 3.7.1-21 1 61 Figure 3.7.1-22 1 61 Figure 3.7.1-23 1 61 Figure 3.7.1-24 1 61 Figure 3.7.1-25 1 61 Figure 3.7.1-26 1 61 Figure 3.7.1-27 1 61 Figure 3.7.1-28 1 61 Figure 3.7.1-29 1 61 Figure 3.7.1-30 1 61 Figure 3.7.1-31 1 61 Figure 3.7.1-32 1 61 Figure 3.7.1-33 1 61 Figure 3.7.1-34 1 61 Figure 3.7.1-35 1 61 Figure 3.7.1-36 1 61 Figure 3.7.1-37 1 61 Figure 3.7.1-38 1 61 Figure 3.7.1-39 1 61 Figure 3.7.1-40 1 61 Figure 3.7.2-1 1 61 Figure 3.7.2-2 1 61 Figure 3.7.2-3 1 61 Figure 3.7.2-4 1 61 Figure 3.7.2-5 1 61 Figure 3.7.2-6 1 61 Figure 3.7.2-7 1 61 Figure 3.7.2-8 1 61 Figure 3.7.2-9 1 61 Figure 3.7.2-10 1 61 Figure 3.7.2-11 1 61

SHNPP FSAR Page 29 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.7.2-12 1 61 Figure 3.7.2-13 1 61 Figure 3.7.2-14 1 61 Figure 3.7.2-15 1 61 Figure 3.7.2-16 1 61 Figure 3.7.3-1 (Deleted by Amendment 9) --- N/A Figure 3.7.3-2 1 61 Figure 3.7.3-3 1 61 Figure 3.7.3-4 1 61 Figure 3.7.3-5 1 61 Figure 3.7.3-6 1 61 Figure 3.7.3-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.7.3-8 1 61 Figure 3.7.3-9 1 61 Figure 3.7.3-10 1 61 Figure 3.7.3-11 1 61 Figure 3.7.3-12 1 61 Figure 3.8.1-1 1 61 Figure 3.8.1-2 1 61 Figure 3.8.1-3 1 61 Figure 3.8.1-4 1 61 Figure 3.8.1-5 1 61 Figure 3.8.1-6 1 61 Figure 3.8.1-7 1 61 Figure 3.8.1-8 1 61 Figure 3.8.1-9 1 61 Figure 3.8.1-10 1 61 Figure 3.8.1-11 1 61 Figure 3.8.1-12 1 61 Figure 3.8.1-13 1 61 Figure 3.8.1-14 1 63 Figure 3.8.1-15 1 61 Figure 3.8.1-16 1 61 Figure 3.8.1-17 1 61 Figure 3.8.1-18 1 61 Figure 3.8.1-19 1 61

SHNPP FSAR Page 30 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.8.1-20 1 61 Figure 3.8.1-21 1 61 Figure 3.8.1-22 1 61 Figure 3.8.1-23 1 61 Figure 3.8.1-24 1 61 Figure 3.8.1-25 1 61 Figure 3.8.1-26 1 61 Figure 3.8.1-27 1 61 Figure 3.8.1-28 1 61 Figure 3.8.1-29 1 61 Figure 3.8.1-30 1 61 Figure 3.8.1-31 1 61 Figure 3.8.1-32 1 61 Figure 3.8.1-33 1 61 Figure 3.8.1-34 1 61 Figure 3.8.1-35 1 61 Figure 3.8.1-36 1 61 Figure 3.8.1-37 1 61 Figure 3.8.1-38 1 61 Figure 3.8.1-39 1 61 Figure 3.8.1-40 1 61 Figure 3.8.1-41 1 61 Figure 3.8.1-42 1 61 Figure 3.8.1-43 1 61 Figure 3.8.1-44 1 61 Figure 3.8.1-45 1 61 Figure 3.8.1-46 1 61 Figure 3.8.1-47 1 61 Figure 3.8.1-48 1 61 Figure 3.8.1-49 1 61 Figure 3.8.1-50 1 61 Figure 3.8.2-1 1 61 Figure 3.8.2-2 1 61 Figure 3.8.2-3 1 61 Figure 3.8.2-4 1 61 Figure 3.8.2-5 1 61

SHNPP FSAR Page 31 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.8.2-6 1 61 Figure 3.8.2-7 1 61 Figure 3.8.2-8 (Deleted by Amendment 36) --- N/A Figure 3.8.2-9 1 61 Figure 3.8.2-10 1 61 Figure 3.8.2-11 1 61 Figure 3.8.2-12 (Deleted by Amendment 36) --- N/A Figure 3.8.2-13 (Deleted by Amendment 36) --- N/A Figure 3.8.2-14 1 61 Figure 3.8.2-15 1 61 Figure 3.8.2-16 1 61 Figure 3.8.2-17 1 61 Figure 3.8.2-18 1 61 Figure 3.8.2-19 1 61 Figure 3.8.3-1 1 61 Figure 3.8.3-2 1 61 Figure 3.8.3-3 1 61 Figure 3.8.3-4 1 61 Figure 3.8.3-5 1 61 Figure 3.8.3-6 1 61 Figure 3.8.3-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.3-8 1 61 Figure 3.8.3-9 1 61 Figure 3.8.3-10 1 61 Figure 3.8.3-11 (Deleted by Amendment 40) --- N/A Figure 3.8.3-12 (Deleted by Amendment 40) --- N/A Figure 3.8.3-13 1 61 Figure 3.8.3-14 1 61 Figure 3.8.3-15 1 61 Figure 3.8.3-16 1 61 Figure 3.8.4-1 1 61 Figure 3.8.4-2 1 61 Figure 3.8.4-3 1 61 Figure 3.8.4-4 1 61 Figure 3.8.4-5 1 61 Figure 3.8.4-6 1 61

SHNPP FSAR Page 32 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.8.4-7 1 61 Figure 3.8.4-8 1 61 Figure 3.8.4-9 1 61 Figure 3.8.4-10 1 61 Figure 3.8.4-11 1 61 Figure 3.8.4-12 1 61 Figure 3.8.4-13 1 61 Figure 3.8.4-14 1 61 Figure 3.8.4-15 1 61 Figure 3.8.4-16 1 61 Figure 3.8.4-17 1 61 Figure 3.8.4-18 1 61 Figure 3.8.4-19 1 61 Figure 3.8.4-20 1 61 Figure 3.8.4-21 1 61 Figure 3.8.4-22 1 61 Figure 3.8.4-22a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-23 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-24 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-25 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-26 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-27 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-28 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-29 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-30 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-31 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-32 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-33 1 61 Figure 3.8.4-34 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-35 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-36 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-37 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.4-38 1 61 Figure 3.8.4-39 1 61 Figure 3.8.4-40 1 61 Figure 3.8.4-41 1 61

SHNPP FSAR Page 33 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.8.4-42 1 61 Figure 3.8.4-43 (Deleted by Amendment 27) --- N/A Figure 3.8.4-44 (Deleted by Amendment 27) --- N/A Figure 3.8.4-45 1 61 Figure 3.8.5-1 1 61 Figure 3.8.5-2 1 61 Figure 3.8.5-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.5-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.8.5-5 1 61 Figure 3.8.5-6 1 61 Figure 3.8.5-7 1 61 Figure 3.9.1-1 1 61 Figure 3.9.1-2 1 61 Figure 3.9.1-3 (Deleted by Amendment 51) --- N/A Figure 3.9.1-4 (Deleted by Amendment 51) --- N/A Figure 3.9.2-1 1 61 Figure 3.9.3-1 (Deleted by Amendment 27) --- N/A Figure 3.9.4-1 1 63 Figure 3.9.4-2 1 63 Figure 3.9.4-3 1 61 Figure 3.9.4-4 1 61 Figure 3.9.5-1 1 61 Figure 3.9.5-2 1 61 Figure 3.9.5-3 1 61 Figure 3.10.1-1 1 61 Figure 3.11.1-1 (Deleted by Amendment 40) --- N/A Figure 3.11.1-2 (Deleted by Amendment 56) --- N/A Figure 3.11.4-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.4-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.4-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.4-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.6-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.6-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.6-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11.7-1 (Deleted by Amendment 55) --- N/A Figure 3.11.7-2 (Deleted by Amendment 55) --- N/A

SHNPP FSAR Page 34 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.11.7-3 (Deleted by Amendment 55) --- N/A Figure 3.11.7-4 (Deleted by Amendment 55) --- N/A Figure 3.11B-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-7 (Deleted by Amendment 15) --- N/A Figure 3.11B-8 (Deleted by Amendment 15) --- N/A Figure 3.11B-9 (Deleted by Amendment 15) --- N/A Figure 3.11B-10 (Deleted by Amendment 15) --- N/A Figure 3.11B-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-12 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-13 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-16 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-17 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-18 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-19 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-20 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-21 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-22 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-23 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-24 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-25 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-26 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-27 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-28 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11B-29 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 3.11C-1 1 61 Figure 3.11C-2 1 61 Figure 3.11C-3 1 61 Figure 3.11E-1 1 61 Figure 3.11E-2 1 61

SHNPP FSAR Page 35 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 3.11E-3 1 61 Figure 3.11E-4 1 61 Figure 3.11E-5 1 61 Chapter 4 - Reactor Table of Contents 4 65 Chapter 4 99 65 Tables - List of Tables Table 4.1.1-1 2 61 Table 4.1.1-2 1 63 Table 4.2.2-1 2 61 Table 4.3.2-1 1 64 Table 4.3.2-2 2 64 Table 4.3.2-3 1 64 Table 4.3.2-4 (Deleted by Amendment 40) --- N/A Table 4.3.2-5 1 61 Table 4.3.2-6 1 61 Table 4.3.2-7 1 64 Table 4.3.2-8 1 61 Table 4.3.2-9 1 61 Table 4.3.2-10 1 61 Table 4.3.2-11 1 61 Table 4.4.3-1 1 64 Table 4.4.3-2 1 64 Table 4.4.3-3 (Deleted by Amendment 64) -- N/A Table 4.4.4-1 2 65 Table 4.4.4-2 1 65 Table 4.5.1-1 1 61 Figures - List of Figures Figure 4.2.2-1A 1 61 Figure 4.2.2-1B 1 61 Figure 4.2.2-1C 1 61 Figure 4.2.2-1D 1 61 Figure 4.2.2-2A 1 61 Figure 4.2.2-2B 1 61 Figure 4.2.2-2C 1 64

SHNPP FSAR Page 36 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 4.2.2-3A 1 61 Figure 4.2.2-3B 1 61 Figure 4.2.2-3C 1 61 Figure 4.2.2-4 1 61 Figure 4.2.2-5A 1 61 Figure 4.2.2-5B 1 61 Figure 4.2.2-6 1 61 Figure 4.2.2-7 1 61 Figure 4.2.2-8 1 61 Figure 4.2.2-9 1 61 Figure 4.2.2-9a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 4.2.2-10 1 61 Figure 4.2.2-11 (Deleted by Amendment 45) --- N/A Figure 4.2.2-12 (Deleted by Amendment 45) --- N/A Figure 4.2.2-13 (Deleted by Amendment 41) --- N/A Figure 4.2.2-14 (Deleted by Amendment 45) --- N/A Figure 4.2.2-15 1 61 Figure 4.3.2-1 (Deleted by Amendment 43) --- N/A Figure 4.3.2-1A 1 61 Figure 4.3.2-1b 1 61 Figure 4.3.2-2 1 61 Figure 4.3.2-3 (Deleted by Amendment 41) --- N/A Figure 4.3.2-4a 1 61 Figure 4.3.2-4b (Deleted by Amendment 45) --- N/A Figure 4.3.2-5 (Deleted by Amendment 45) --- N/A Figure 4.3.2-6 1 61 Figure 4.3.2-7 1 61 Figure 4.3.2-8 1 61 Figure 4.3.2-9 1 61 Figure 4.3.2-10 1 61 Figure 4.3.2-11 1 61 Figure 4.3.2-12 1 61 Figure 4.3.2-13 1 61 Figure 4.3.2-14 1 61 Figure 4.3.2-15 1 61 Figure 4.3.2-16 1 61

SHNPP FSAR Page 37 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 4.3.2-17 1 61 Figure 4.3.2-18 (Deleted by Amendment 45) --- N/A Figure 4.3.2-19 (Deleted by Amendment 45) --- N/A Figure 4.3.2-20 (Deleted by Amendment 45) --- N/A Figure 4.3.2-21 1 64 Figure 4.3.2-22 (Deleted by Amendment 45) --- N/A Figure 4.3.2-23 (Deleted by Amendment 45) --- N/A Figure 4.3.2-24 (Deleted by Amendment 45) --- N/A Figure 4.3.2-25 (Deleted by Amendment 45) --- N/A Figure 4.3.2-26 (Deleted by Amendment 45) --- N/A Figure 4.3.2-27 (Deleted by Amendment 45) --- N/A Figure 4.3.2-28 1 61 Figure 4.3.2-29 1 61 Figure 4.3.2-30 1 61 Figure 4.3.2-31 1 61 Figure 4.3.2-32 1 61 Figure 4.3.2-33 1 61 Figure 4.3.2-34 1 61 Figure 4.3.2-35 1 61 Figure 4.3.2-36 1 61 Figure 4.3.2-37 1 61 Figure 4.3.2-38 1 63 Figure 4.3.2-39 1 63 Figure 4.3.2-40 1 61 Figure 4.3.2-41 1 61 Figure 4.3.2-42 1 63 Figure 4.3.2-43 1 63 Figure 4.4.3-1 (Deleted by Amendment 64) -- N/A Figure 4.4.3-2 (Deleted by Amendment 64) -- N/A Figure 4.4.3-3 (Deleted by Amendment 64) -- N/A Figure 4.4.4-1 1 61 Figure 4.4.4-2 1 61 Chapter 5 - Reactor Coolant System and Connected Systems Table of Contents 5 65 Chapter 5 113 65

SHNPP FSAR Page 38 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Tables - List of Tables Table 5.1.0-1 2 61 Table 5.2.1-1 1 63 Table 5.2.1-2 1 61 Table 5.2.2-1 1 61 Table 5.2.3-1 3 63 Table 5.2.3-2 1 63 Table 5.2.3-3 (Deleted by Amendment 41) --- N/A Table 5.3.1-1 1 63 Table 5.3.1-2 1 63 Table 5.3.1-3 (Deleted by Amendment 15) --- N/A Table 5.3.1-6 1 61 Table 5.3.1-7 1 64 Table 5.3.1-8 1 64 Table 5.3.1-9 (Deleted by Amendment 15) --- N/A Table 5.3.1-10 (Deleted by Amendment 15) --- N/A Table 5.3.1-11 (Deleted by Amendment 15) --- N/A Table 5.3.1-12 (Deleted by Amendment 3) --- N/A Table 5.3.1-13 (Deleted by Amendment 3) --- N/A Table 5.3.1-14 (Deleted by Amendment 3) --- N/A Table 5.3.1-15 (Deleted by Amendment 3) --- N/A Table 5.3.1-16 (Deleted by Amendment 3) --- N/A Table 5.3.1-17 (Deleted by Amendment 3) --- N/A Table 5.3.1-18 1 61 Table 5.3.1-19 (Deleted by Amendment 15) --- N/A Table 5.3.1-20 (Deleted by Amendment 3) --- N/A Table 5.3.1-21 (Deleted by Amendment 3) --- N/A Table 5.3.1-22 1 64 Table 5.3.3-1 1 63 Table 5.4.1-1 2 65 Table 5.4.1-2 1 61 Table 5.4.2-1 1 61 Table 5.4.2-2 1 65 Table 5.4.3-1 1 61 Table 5.4.3-2 1 61 Table 5.4.7-1 1 61

SHNPP FSAR Page 39 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 5.4.7-1A 1 61 Table 5.4.7-2 1 61 Table 5.4.7-3 5 61 Table 5.4.10-1 1 61 Table 5.4.10-2 1 61 Table 5.4.10-3 1 61 Table 5.4.11-1 1 61 Table 5.4.11-2 1 61 Table 5.4.12-1 1 61 Table 5.4.12-2 1 61 Table 5.4.12-3 2 61 Table 5.4.13-1 1 61 Figures - List of Figures Figure 5.1.1-1 3 61 Figure 5.1.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 5.1.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 5.1.2-3 (Deleted by Amendment 48) --- N/A Figure 5.1.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 5.3.1-1 1 61 Figure 5.3.1-2 (Deleted by Amendment 15) --- N/A Figure 5.3.1-3 1 61 Figure 5.3.3-1 1 63 Figure 5.4.1-1 1 61 Figure 5.4.1-2 1 61 Figure 5.4.2-1 1 61 Figure 5.4.7-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 5.4.7-2 5 61 Figure 5.4.7-3 1 61 Figure 5.4.7-4 1 61 Figure 5.4.10-1 1 61 Figure 5.4.11-1 1 61 Figure 5.4.12-1 1 61 Figure 5.4.13-1 1 61 Figure 5.4.13-2 1 61 Figure 5.4.13-3 1 61 Figure 5.4.13-4 1 61

SHNPP FSAR Page 40 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 5.4.14-1 1 61 Figure 5.4.14-2 1 61 Figure 5.4.14-3 1 61 Figure 5.4.14-4 1 61 Figure 5.4.14-5 (Deleted by Amendment 40) --- N/A Figure 5.4.14-6 (Deleted by Amendment 40) --- N/A Figure 5.4.14-7 (Deleted by Amendment 40) --- N/A Chapter 6 - Engineered Safety Features Table of Contents 5 65 Chapter 6 158 65 Tables - List of Tables Table 6.1.1-1 2 62 Table 6.1.1-2 1 63 Table 6.1.2-1 1 61 Table 6.1.2-2 2 65 Table 6.2.1-1 1 61 Table 6.2.1-2 1 63 Table 6.2.1-3 1 61 Table 6.2.1-4 1 63 Table 6.2.1-5 1 63 Table 6.2.1-6 2 63 Table 6.2.1-7 4 62 Table 6.2.1-8 2 62 Table 6.2.1-9 2 65 Table 6.2.1-10 (Deleted by Amendment 51) --- N/A Table 6.2.1-11 1 61 Table 6.2.1-12 5 61 Table 6.2.1-13 8 61 Table 6.2.1-14 5 61 Table 6.2.1-15 5 61 Table 6.2.1-16 4 61 Table 6.2.1-17 1 61 Table 6.2.1-17a 1 61 Table 6.2.1-18 1 61 Table 6.2.1-18a 1 61

SHNPP FSAR Page 41 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 6.2.1-19 1 61 Table 6.2.1-20 2 61 Table 6.2.1-20A 1 61 Table 6.2.1-20B 1 61 Table 6.2.1-21 1 61 Table 6.2.1-22 2 61 Table 6.2.1-23 1 61 Table 6.2.1-24 3 61 Table 6.2.1-25 1 61 Table 6.2.1-26 3 61 Table 6.2.1-27 1 61 Table 6.2.1-28 1 61 Table 6.2.1-29a 2 61 Table 6.2.1-29b 4 65 Table 6.2.1-30 (Deleted by Amendment 51) --- N/A Table 6.2.1-31 (Deleted by Amendment 51) --- N/A Table 6.2.1-32 (Deleted by Amendment 51) --- N/A Table 6.2.1-33 4 63 Table 6.2.1-34 (Deleted by Amendment 51) --- N/A Table 6.2.1-35 4 61 Table 6.2.1-36 5 63 Table 6.2.1-37 (Deleted by Amendment 51) --- N/A Table 6.2.1-38 (Deleted by Amendment 51) --- N/A Table 6.2.1-39 (Deleted by Amendment 51) --- N/A Table 6.2.1.40 4 61 Table 6.2.1-41 4 63 Table 6.2.1-42 (Deleted by Amendment 51) --- N/A Table 6.2.1-43 1 61 Table 6.2.1-44 1 63 Table 6.2.1-45 (Deleted by Amendment 51) --- N/A Table 6.2.1-46 (Deleted by Amendment 51) --- N/A Table 6.2.1-47 1 63 Table 6.2.1-48 (Deleted by Amendment 51) --- N/A Table 6.2.1-49 1 61 Table 6.2.1-50 (Deleted by Amendment 63) --- N/A Table 6.2.1-51 1 61

SHNPP FSAR Page 42 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 6.2.1-52 1 63 Table 6.2.1-53 (Deleted by Amendment 51) --- N/A Table 6.2.1-54 (Deleted by Amendment 51) --- N/A Table 6.2.1-55 1 65 Table 6.2.1-56 (Deleted by Amendment 51) --- N/A Table 6.2.1-57 (Deleted by Amendment 51) --- N/A Table 6.2.1-58A 1 61 Table 6.2.1-58B 1 61 Table 6.2.1-59 (Deleted by Amendment 46) --- N/A Table 6.2.1-60 (Deleted by Amendment 46) --- N/A Table 6.2.1-61 (Deleted by Amendment 46) --- N/A Table 6.2.1-62 1 61 Table 6.2.1-63 2 61 Table 6.2.1-64 1 61 Table 6.2.1-65 1 65 Table 6.2.1-66 3 61 Table 6.2.2-1 3 64 Table 6.2.2-2 (Deleted by Amendment 48) --- N/A Table 6.2.2-3 1 61 Table 6.2.2-4 1 61 Table 6.2.2-5 1 61 Table 6.2.2-6 1 61 Table 6.2.2-7 1 61 Table 6.2.2-8 1 62 Table 6.2.2-9 1 61 Table 6.2.2-10 (Deleted by Amendment 43) --- N/A Table 6.2.2-11 (Deleted by Amendment 43) --- N/A Table 6.2.4-1 (includes index, abbreviations, definitions and notes pages) 33 65 Table 6.2.4-2 4 65 Table 6.2.5-1 (Deleted by Amendment 62) --- N/A Table 6.2.5-2 2 61 Table 6.2.5-3 1 61 Table 6.2.5-3a 1 61 Table 6.2.5-4 1 63 Table 6.2.5-5 1 61 Table 6.2.5-6 1 65

SHNPP FSAR Page 43 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 6.2.5-7 1 62 Table 6.2B-1 (Deleted by Amendment 51) --- N/A Table 6.2B-2 (Deleted by Amendment 51) --- N/A Table 6.3.1-1 9 65 Table 6.3.2-1 2 65 Table 6.3.2-2 1 61 Table 6.3.2-3 2 61 Table 6.3.2-4 1 61 Table 6.3.2-5 1 61 Table 6.3.2-6 2 61 Table 6.3.2-7 1 61 Table 6.3.2-8 1 61 Table 6.3.2-9 2 62 Table 6.3.2-10 1 64 Table 6.4.2-1 1 61 Table 6.4.2-2 1 61 Table 6.4.4-1 1 61 Table 6.5.1-1 2 61 Table 6.5.1-2 2 63 Table 6.5.1-3 2 61 Table 6.5.1-4 1 65 Table 6.5.1-5 1 65 Table 6.5.2-1 1 61 Table 6.5.3-1 1 62 Figures - List of Figures Figure 6.2.1-1 1 63 Figure 6.2.1-1a 1 63 Figure 6.2.1-2 1 63 Figure 6.2.1-3 1 63 Figure 6.2.1-4 1 63 Figure 6.2.1-5a 1 63 Figure 6.2.1-5b 1 63 Figure 6.2.1-6 (Deleted by Amendment 51) --- N/A Figure 6.2.1-6a 1 63 Figure 6.2.1-6b 1 63 Figure 6.2.1-7 (Deleted by Amendment 51) --- N/A

SHNPP FSAR Page 44 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-8 (Deleted by Amendment 51) --- N/A Figure 6.2.1-9 1 61 Figure 6.2.1-10 (Deleted by Amendment 51) --- N/A Figure 6.2.1-10a 1 61 Figure 6.2.1-10b 1 61 Figure 6.2.1-11 1 61 Figure 6.2.1-12 1 61 Figure 6.2.1-13 1 61 Figure 6.2.1-14 1 61 Figure 6.2.1-15 1 61 Figure 6.2.1-16 1 61 Figure 6.2.1-17 1 61 Figure 6.2.1-18 1 61 Figure 6.2.1-19 1 61 Figure 6.2.1-20 1 61 Figure 6.2.1-21 1 61 Figure 6.2.1-22 1 61 Figure 6.2.1-23 1 61 Figure 6.2.1-24 1 61 Figure 6.2.1-25 1 61 Figure 6.2.1-26 1 61 Figure 6.2.1-27 1 61 Figure 6.2.1-28a 1 61 Figure 6.2.1-28b 1 61 Figure 6.2.1-28c 1 61 Figure 6.2.1-28d 1 61 Figure 6.2.1-28e 1 61 Figure 6.2.1-28f 1 61 Figure 6.2.1-28g 1 61 Figure 6.2.1-28h 1 61 Figure 6.2.1-28i 1 61 Figure 6.2.1-29 1 61 Figure 6.2.1-30 1 61 Figure 6.2.1-31 1 61 Figure 6.2.1-32 1 61 Figure 6.2.1-33 1 61

SHNPP FSAR Page 45 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-34 1 61 Figure 6.2.1-35 1 61 Figure 6.2.1-36 1 61 Figure 6.2.1-37 1 61 Figure 6.2.1-38 1 61 Figure 6.2.1-39 1 61 Figure 6.2.1-40 1 61 Figure 6.2.1-41 1 61 Figure 6.2.1-42 1 61 Figure 6.2.1-43 1 61 Figure 6.2.1-44 1 61 Figure 6.2.1-45 1 61 Figure 6.2.1-46 1 61 Figure 6.2.1-47 1 61 Figure 6.2.1-48 1 61 Figure 6.2.1-49 1 61 Figure 6.2.1-50 1 61 Figure 6.2.1-51 1 61 Figure 6.2.1-52a 1 61 Figure 6.2.1-52b 1 61 Figure 6.2.1-52c 1 61 Figure 6.2.1-52d 1 61 Figure 6.2.1-52e 1 61 Figure 6.2.1-52f 1 61 Figure 6.2.1-52g 1 61 Figure 6.2.1-52h 1 61 Figure 6.2.1-52i 1 61 Figure 6.2.1-53 1 61 Figure 6.2.1-54 1 61 Figure 6.2.1-55 1 61 Figure 6.2.1-56 1 61 Figure 6.2.1-57 1 61 Figure 6.2.1-58 1 61 Figure 6.2.1-59 1 61 Figure 6.2.1-60 1 61 Figure 6.2.1-61 1 61

SHNPP FSAR Page 46 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-62 1 61 Figure 6.2.1-63 1 61 Figure 6.2.1-64 1 61 Figure 6.2.1-65 1 61 Figure 6.2.1-66 1 61 Figure 6.2.1-67 1 61 Figure 6.2.1-68 1 61 Figure 6.2.1-69 1 61 Figure 6.2.1-70 1 61 Figure 6.2.1-71 1 61 Figure 6.2.1-72 1 61 Figure 6.2.1-73 1 61 Figure 6.2.1-74 1 61 Figure 6.2.1-75 1 61 Figure 6.2.1-76 1 61 Figure 6.2.1-77 1 61 Figure 6.2.1-78 1 61 Figure 6.2.1-79 1 61 Figure 6.2.1-80 1 61 Figure 6.2.1-81 1 61 Figure 6.2.1-82 1 61 Figure 6.2.1-83 1 61 Figure 6.2.1-84 1 61 Figure 6.2.1-85 1 61 Figure 6.2.1-86 1 61 Figure 6.2.1-87 1 61 Figure 6.2.1-88 1 61 Figure 6.2.1-89 1 61 Figure 6.2.1-90 1 61 Figure 6.2.1-91 1 61 Figure 6.2.1-92 1 61 Figure 6.2.1-93 1 61 Figure 6.2.1-94 1 61 Figure 6.2.1-95 1 61 Figure 6.2.1-96 1 61 Figure 6.2.1-97 1 61

SHNPP FSAR Page 47 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-98 1 61 Figure 6.2.1-99 1 61 Figure 6.2.1-100 1 61 Figure 6.2.1-101 1 61 Figure 6.2.1-102 1 61 Figure 6.2.1-103 1 61 Figure 6.2.1-104 1 61 Figure 6.2.1-105 1 61 Figure 6.2.1-106 1 61 Figure 6.2.1-107 1 61 Figure 6.2.1-108 1 61 Figure 6.2.1-109 1 61 Figure 6.2.1-110 1 61 Figure 6.2.1-111 1 61 Figure 6.2.1-112 1 61 Figure 6.2.1-113 1 61 Figure 6.2.1-114 1 61 Figure 6.2.1-115 1 61 Figure 6.2.1-116 1 61 Figure 6.2.1-117 1 61 Figure 6.2.1-118 1 61 Figure 6.2.1-119 1 61 Figure 6.2.1-120 1 61 Figure 6.2.1-121 1 61 Figure 6.2.1-122 1 61 Figure 6.2.1-123 1 61 Figure 6.2.1-124 1 61 Figure 6.2.1-125 1 61 Figure 6.2.1-126 1 61 Figure 6.2.1-127 1 61 Figure 6.2.1-128 1 61 Figure 6.2.1-129 1 61 Figure 6.2.1-130 1 61 Figure 6.2.1-131 1 61 Figure 6.2.1-132 1 61 Figure 6.2.1-133 1 61

SHNPP FSAR Page 48 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-134 1 61 Figure 6.2.1-135 1 61 Figure 6.2.1-136 1 61 Figure 6.2.1-137 1 61 Figure 6.2.1-138 1 61 Figure 6.2.1-139 1 61 Figure 6.2.1-140 1 61 Figure 6.2.1-141 1 61 Figure 6.2.1-142 1 61 Figure 6.2.1-143 1 61 Figure 6.2.1-144 1 61 Figure 6.2.1-145 1 61 Figure 6.2.1-146 1 61 Figure 6.2.1-147 1 61 Figure 6.2.1-148 1 61 Figure 6.2.1-149 1 61 Figure 6.2.1-150 1 61 Figure 6.2.1-151 1 61 Figure 6.2.1-152 1 61 Figure 6.2.1-153 1 61 Figure 6.2.1-154 1 61 Figure 6.2.1-155 1 61 Figure 6.2.1-156 1 61 Figure 6.2.1-157 1 61 Figure 6.2.1-158 1 61 Figure 6.2.1-159 1 61 Figure 6.2.1-160 1 61 Figure 6.2.1-161 1 61 Figure 6.2.1-162 1 61 Figure 6.2.1-163 1 61 Figure 6.2.1-164 1 61 Figure 6.2.1-165 1 61 Figure 6.2.1-166 1 61 Figure 6.2.1-167 1 61 Figure 6.2.1-168 1 61 Figure 6.2.1-169 1 61

SHNPP FSAR Page 49 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-170 1 61 Figure 6.2.1-171 1 61 Figure 6.2.1-172 1 61 Figure 6.2.1-173 1 61 Figure 6.2.1-174 1 61 Figure 6.2.1-175 1 61 Figure 6.2.1-176 1 61 Figure 6.2.1-177 1 61 Figure 6.2.1-178 1 61 Figure 6.2.1-179 1 61 Figure 6.2.1-180 1 61 Figure 6.2.1-181 1 61 Figure 6.2.1-182 1 61 Figure 6.2.1-183 1 61 Figure 6.2.1-184 1 61 Figure 6.2.1-185 1 61 Figure 6.2.1-186 1 61 Figure 6.2.1-187 1 61 Figure 6.2.1-188 1 61 Figure 6.2.1-189 1 61 Figure 6.2.1-190 1 61 Figure 6.2.1-191 1 61 Figure 6.2.1-192 1 61 Figure 6.2.1-193 1 61 Figure 6.2.1-194 1 61 Figure 6.2.1-195 1 61 Figure 6.2.1-196 1 61 Figure 6.2.1-197 1 61 Figure 6.2.1-198 1 61 Figure 6.2.1-199 1 61 Figure 6.2.1-200 1 61 Figure 6.2.1-201 1 61 Figure 6.2.1-202 1 61 Figure 6.2.1-203 1 61 Figure 6.2.1-204 1 61 Figure 6.2.1-205 1 61

SHNPP FSAR Page 50 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-206 1 61 Figure 6.2.1-207 1 61 Figure 6.2.1-208 1 61 Figure 6.2.1-209 1 61 Figure 6.2.1-210 1 61 Figure 6.2.1-211 1 61 Figure 6.2.1-212 1 61 Figure 6.2.1-213 1 61 Figure 6.2.1-214 1 61 Figure 6.2.1-215 1 61 Figure 6.2.1-216 1 61 Figure 6.2.1-217 1 61 Figure 6.2.1-218 1 61 Figure 6.2.1-219 1 61 Figure 6.2.1-220 1 61 Figure 6.2.1-221 1 61 Figure 6.2.1-222 1 61 Figure 6.2.1-223 1 61 Figure 6.2.1-224 1 61 Figure 6.2.1-225 1 61 Figure 6.2.1-226 1 61 Figure 6.2.1-227 1 61 Figure 6.2.1-228 1 61 Figure 6.2.1-229 1 61 Figure 6.2.1-230 1 61 Figure 6.2.1-231 1 61 Figure 6.2.1-232 1 61 Figure 6.2.1-233 1 61 Figure 6.2.1-234 1 61 Figure 6.2.1-235 1 61 Figure 6.2.1-236 1 61 Figure 6.2.1-237 1 61 Figure 6.2.1-238 1 61 Figure 6.2.1-239 1 61 Figure 6.2.1-240 1 61 Figure 6.2.1-241 1 61

SHNPP FSAR Page 51 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-242 1 61 Figure 6.2.1-243 1 61 Figure 6.2.1-244 1 61 Figure 6.2.1-245 1 61 Figure 6.2.1-246 1 61 Figure 6.2.1-247 1 61 Figure 6.2.1-248 1 61 Figure 6.2.1-249 1 61 Figure 6.2.1-250 1 61 Figure 6.2.1-251 1 61 Figure 6.2.1-252 1 61 Figure 6.2.1-253 1 61 Figure 6.2.1-254 1 61 Figure 6.2.1-255 1 61 Figure 6.2.1-256 1 61 Figure 6.2.1-257 1 61 Figure 6.2.1-258 1 61 Figure 6.2.1-259 1 61 Figure 6.2.1-260 1 61 Figure 6.2.1-261 1 61 Figure 6.2.1-262 1 61 Figure 6.2.1-263 1 61 Figure 6.2.1-264 1 61 Figure 6.2.1-265 1 61 Figure 6.2.1-266 1 61 Figure 6.2.1-267 1 61 Figure 6.2.1-268 1 61 Figure 6.2.1-269 1 61 Figure 6.2.1-270 1 61 Figure 6.2.1-271 1 61 Figure 6.2.1-272 1 61 Figure 6.2.1-273 1 61 Figure 6.2.1-274 1 61 Figure 6.2.1-275 1 61 Figure 6.2.1-276 1 61 Figure 6.2.1-277 1 61

SHNPP FSAR Page 52 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.1-278 1 61 Figure 6.2.1-279 1 0 Figure 6.2.1-280 1 61 Figure 6.2.1-281 1 61 Figure 6.2.1-282 1 61 Figure 6.2.1-283 1 61 Figure 6.2.1-284 1 61 Figure 6.2.1-285 1 61 Figure 6.2.1-286 1 61 Figure 6.2.1-287 1 61 Figure 6.2.1-288 1 61 Figure 6.2.1-289 1 61 Figure 6.2.1-290 1 61 Figure 6.2.1-291 1 61 Figure 6.2.1-292 1 61 Figure 6.2.1-293 1 61 Figure 6.2.1-294 1 61 Figure 6.2.1-295 1 61 Figure 6.2.1-296 1 61 Figure 6.2.1-297 1 61 Figure 6.2.1-298 1 61 Figure 6.2.1-299 1 61 Figure 6.2.1-300 1 61 Figure 6.2.1-301 1 61 Figure 6.2.1-302 (Deleted by Amendment 46) --- N/A Figure 6.2.1-303 1 61 Figure 6.2.1-304 (Deleted by Amendment 46) --- N/A Figure 6.2.1-305 (Deleted by Amendment 46) --- N/A Figure 6.2.1-306 1 61 Figure 6.2.1-307 1 61 Figure 6.2.1-308 1 61 Figure 6.2.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-4 1 61 Figure 6.2.2-5 (Deleted by Amendment 48) --- N/A

SHNPP FSAR Page 53 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.2.2-6 1 61 Figure 6.2.2-7 1 61 Figure 6.2.2-8 1 61 Figure 6.2.2-9 1 61 Figure 6.2.2-10 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-12 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-13 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-16 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-17 1 61 Figure 6.2.2-18 1 61 Figure 6.2.2-19 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.2.2-20 1 61 Figure 6.2.5-1 (Deleted by Amendment 62) --- N/A Figure 6.2.5-2 (Deleted by Amendment 62) --- N/A Figure 6.2.5-3 (Deleted by Amendment 62) --- N/A Figure 6.2.5-4 (Deleted by Amendment 62) --- N/A Figure 6.2.5-5 (Deleted by Amendment 58) --- N/A Figure 6.2.5-6 (Deleted by Amendment 62) --- N/A Figure 6.2.5-7 1 61 Figure 6.2A-1 1 61 Figure 6.2A-2 1 61 Figure 6.3.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.3.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.3.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 6.3.2-4 1 61 Figure 6.3.2-5 1 61 Figure 6.3.2-6 1 61 Notes to Figure 6.3.2-4 through 6.3.2-6 10 61 Figure 6.3.2-7 (Deleted by Amendment 27) --- N/A Figure 6.3.2-8 1 61 Figure 6.3.2-9 1 61 Figure 6.4.2-1 (Deleted by Amendment 15) --- N/A Figure 6.5.2-1 (Deleted by Amendment 51) --- N/A

SHNPP FSAR Page 54 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 6.5.2-2 1 61 Figure 6.5.2-3 1 63 Figure 6.5.2-4 (Deleted by Amendment 27) --- N/A Figure 6.5.2-5 (Deleted by Amendment 27) --- N/A Figure 6.5.2-6 (Deleted by Amendment 27) --- N/A Chapter 7 - Instrumentation and Controls Table of Contents 5 65 Chapter 7 188 65 Tables - List of Tables Table 7.1.0-1 5 61 Table 7.1.1-1 3 63 Table 7.1.1-2 1 61 Table 7.2.1-1 2 65 Table 7.2.1-2 1 65 Table 7.2.1-3 1 63 Table 7.2.2-1 (Deleted by Amendment 48) --- N/A Table 7.3.1-1 2 61 Table 7.3.1-2 1 61 Table 7.3.1-3 2 61 Table 7.3.1-4 1 61 Table 7.3.1-5 6 64 Table 7.3.1-6 1 61 Table 7.3.1-7 3 61 Table 7.3.1-8 1 61 Table 7.3.1-9 1 61 Table 7.3.1-10 1 61 Table 7.3.1-11 1 61 Table 7.3.1-12 1 61 Table 7.3.2-1 1 61 Table 7.4.1-1 4 61 Table 7.4.1-2 8 61 Table 7.5.1-1 1 65 Table 7.5.1-2 1 65 Table 7.5.1-3 2 61 Table 7.5.1-4 1 61

SHNPP FSAR Page 55 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 7.5.1-5 1 65 Table 7.5.1-6 1 65 Table 7.5.1-7 1 65 Table 7.5.1-8 1 61 Table 7.5.1-9 3 65 Table 7.5.1-10 1 65 Table 7.5.1-11 3 65 Table 7.5.1-12 1 61 Table 7.5.1-13 1 65 Table 7.5.1-14 2 65 Table 7.5.1-15 1 61 Table 7.5.1-16 1 61 Table 7.7.1-1 1 61 Table 7.7.1-2 (Deleted by Amendment 40) --- N/A Figures - List of Figures Figure 7.1.1-1 1 61 Figure 7.2.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.2.1-2 1 61 Figure 7.2.1-3 1 61 Figure 7.3.1-1 2 61 Figure 7.3.1-2 2 61 Figure 7.3.1-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-9 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-9a (Deleted by Amendment 48) --- N/A Figure 7.3.1-9b (Deleted by Amendment 48) --- N/A Figure 7.3.1-10 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-12 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-13 4 61 Figure 7.3.1-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 56 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 7.3.1-15a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-16 2 61 Figure 7.3.1-17 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-18 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-19 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-20 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-21 1 61 Figure 7.3.1-22 3 61 Figure 7.3.1-23 2 61 Figure 7.3.1-24 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-25 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-26 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-27 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.3.1-28 (Deleted by Amendment 48) --- N/A Figure 7.3.2-1 1 61 Figure 7.3.2-2 1 61 Figure 7.4.1-1 1 61 Figure 7.4.1-2 1 61 Figure 7.4.1-3 1 61 Figure 7.4.1-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.4.1-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.4.1-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.4.1-7 1 61 Figure 7.4.1-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-1 1 61 Figure 7.5.1-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-3 1 61 Figure 7.5.1-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-8 1 61 Figure 7.5.1-9 1 61 Figure 7.5.1-10 1 61 Figure 7.5.1-11 1 61 Figure 7.5.1-12 1 61

SHNPP FSAR Page 57 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 7.5.1-13 1 61 Figure 7.5.1-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.5.1-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-1 1 61 Figure 7.6.1-2 1 61 Figure 7.6.1-3 1 61 Figure 7.6.1-4 1 61 Figure 7.6.1-5 1 61 Figure 7.6.1-6 1 61 Figure 7.6.1-7 1 61 Figure 7.6.1-8 1 61 Figure 7.6.1-9 1 61 Figure 7.6.1-10 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-11 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-12 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-13 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-14 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.6.1-15 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 7.7.1-1 1 61 Figure 7.7.1-2 1 61 Figure 7.7.1-3 1 61 Figure 7.7.1-4 1 61 Figure 7.7.1-5 1 61 Figure 7.7.1-6 1 61 Figure 7.7.1-7 (Deleted by Amendment 27) --- N/A Figure 7.7.1-8 1 61 Figure 7.7.1-9 1 61 Figure 7.7.2-1 1 61 Figure 7.7.2-2 1 61 Chapter 8 - Electric Power Table of Contents 1 65 Chapter 8 96 65 Tables - List of Tables Table 8.1.3-1 4 65 Table 8.2.1-1 1 63

SHNPP FSAR Page 58 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 8.3.1-1 6 65 Table 8.3.1-2a 2 61 Table 8.3.1-2b 2 61 Table 8.3.1-2c 5 65 Table 8.3.1-3 1 61 Table 8.3.1-4 1 61 Table 8.3.1-5 1 61 Table 8.3.1-6 1 61 Table 8.3.1-7 1 61 Table 8.3.1-8 1 61 Table 8.3.1-9 4 61 Table 8.3.1-10 1 61 Table 8.3.2-1 1 61 Table 8.3.2-2 (Deleted by Amendment 46) --- N/A Table 8.3.2-3 (Deleted by Amendment 46) --- N/A Table 8.3.2-4 (Deleted by Amendment 46) --- N/A Table 8.3.2-5 1 61 Figures - List of Figures Figure 8.1.1-1 1 61 Figure 8.1.3-1 1 61 Figure 8.1.3-2 1 61 Figure 8.1.3-3 1 62 Figure 8.2.1-1 1 63 Figure 8.2.1-2 1 63 Figure 8.2.1-3 1 61 Figure 8.2.1-4 1 61 Figure 8.2.1-5 1 63 Figure 8.2.1-6 1 63 Figure 8.2.1-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 8.2.1-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 8.2.1-9 1 61 Figure 8.2.2-1 1 63 Figure 8.2.2-2 1 63 Figure 8.2.2-3 2 61 Figure 8.2.2-4 1 63 Figure 8.2.2-5 1 61

SHNPP FSAR Page 59 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 8.2.2-6 1 63 Figure 8.2.2-7 1 63 Figure 8.2.2-8 1 61 Figure 8.2.2-9 (Deleted by Amendment 60) --- N/A Figure 8.3.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 8.3.1-2 1 61 Figure 8.3.1-3 1 61 Figure 8.3.1-4 1 61 Figure 8.3.1-5 1 61 Figure 8.3.1-6 1 61 Figure 8.3.1-7 1 61 Figure 8.3.1-8 1 61 Figure 8.3.1-9 1 61 Figure 8.3.1-10 2 61 Figure 8.3.1-11 1 61 Figure 8.3.1-12 1 61 Figure 8.3.1-13 1 61 Chapter 9 - Auxiliary Systems Table of Contents 8 65 Chapter 9 267 65 Tables - List of Tables Table 9.1.1-1 (Deleted by Amendment 43) --- N/A Table 9.1.2-1 1 61 Table 9.1.2-2 1 63 Table 9.1.3-1A (Deleted by Amendment 48) --- N/A Table 9.1.3-1B (Deleted by Amendment 48) --- N/A Table 9.1.3-1C (Deleted by Amendment 48) --- N/A Table 9.1.3-2 5 64 Table 9.1.3-3 (Deleted by Amendment 43) --- N/A Table 9.1.4-1 9 61 Table 9.2.1-1 2 63 Table 9.2.1-2 1 63 Table 9.2.1-3 (Deleted by Amendment 51) --- N/A Table 9.2.1-4 1 63 Table 9.2.1-5 (Deleted by Amendment 63) --- N/A

SHNPP FSAR Page 60 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 9.2.1-6 1 62 Table 9.2.1-7 1 61 Table 9.2.1-8 (Deleted by Amendment 51) --- N/A Table 9.2.1-9 (Deleted by Amendment 51) --- N/A Table 9.2.1-10 (Deleted by Amendment 63) --- N/A Table 9.2.1-11 (Deleted by Amendment 63) --- N/A Table 9.2.1-12 1 63 Table 9.2.1-13 1 61 Table 9.2.1-14 1 61 Table 9.2.2-1 1 61 Table 9.2.2-2 (Deleted by Amendment 44) --- N/A Table 9.2.2-3 1 61 Table 9.2.2-4 4 65 Table 9.2.3-1 (Deleted by Amendment 27) --- N/A Table 9.2.3-2 (Deleted by Amendment 27) --- N/A Table 9.2.3-3 3 61 Table 9.2.3-4 1 61 Table 9.2.6-1 1 61 Table 9.2.8-1 2 61 Table 9.2.8-2 2 61 Table 9.2.9-1 1 61 Table 9.2.10-1 1 61 Table 9.2.10-2 (Deleted by Amendment 44) --- N/A Table 9.2.10-3 1 61 Table 9.3.1-1 1 61 Table 9.3.1-2 1 61 Table 9.3.2-1 1 61 Table 9.3.2-1A 1 61 Table 9.3.2-2 1 65 Table 9.3.2-2A 1 62 Table 9.3.2-2B 1 62 Table 9.3.2-3 (Deleted by Amendment 48) --- N/A Table 9.3.4-1 1 61 Table 9.3.4-2 6 61 Table 9.3.4-3 2 61 Table 9.3.4-4 12 65

SHNPP FSAR Page 61 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 9.4.0-1 3 63 Table 9.4.0-2 5 61 Table 9.4.1-1 3 61 Table 9.4.1-2 3 61 Table 9.4.1-3 (Deleted by Amendment 15) --- N/A Table 9.4.1-4 2 61 Table 9.4.2-1 4 61 Table 9.4.2-2 1 61 Table 9.4.3-1 3 61 Table 9.4.3-2 3 61 Table 9.4.3-3 1 61 Table 9.4.3-4 1 65 Table 9.4.3-5 2 65 Table 9.4.3-6 2 65 Table 9.4.3-7 1 61 Table 9.4.4-1 2 61 Table 9.4.4-2 1 61 Table 9.4.4-3 1 61 Table 9.4.4-4 (Deleted by Amendment 62) --- N/A Table 9.4.4-5 1 61 Table 9.4.4-6 1 61 Table 9.4.5-1 3 61 Table 9.4.5-2 1 61 Table 9.4.5-3 2 61 Table 9.4.5-4 2 61 Table 9.4.5-5 1 61 Table 9.4.5-6 2 64 Table 9.4.5-7 2 61 Table 9.4.5-8 1 61 Table 9.4.5-9 2 61 Table 9.4.5-10 1 61 Table 9.4.7-1 1 61 Table 9.4.7-2 1 61 Table 9.4.7-3 2 61 Table 9.4.7-4 1 61 Table 9.4.8-1 1 61

SHNPP FSAR Page 62 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 9.4.9-1 3 61 Table 9.4.9-2 1 61 Table 9.4.10-1 2 61 Table 9.5.1-1 1 61 Table 9.5.2-1 1 61 Table 9.5.2-2 1 64 Table 9.5.3-1 2 61 Table 9.5.4-1 1 65 Table 9.5.4-2 1 61 Table 9.5.5-1 1 61 Table 9.5.6-1 2 61 Table 9.5.7-1 2 61 Table 9.5.7-2 2 65 Figures - List of Figures Figure 9.1.1-1 1 61 Figure 9.1.1-2 1 61 Figure 9.1.2-1 1 61 Figure 9.1.2-2 1 61 Figure 9.1.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.1.3-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.1.3-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.1.3-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.1.4-1 1 61 Figure 9.1.4-2 1 61 Figure 9.1.4-3 1 61 Figure 9.1.4-4 1 61 Figure 9.1.4-4a 1 61 Figure 9.1.4-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.1.4-6 1 61 Figure 9.1.4-7 1 61 Figure 9.1.4-8 1 61 Figure 9.1.4-9 1 61 Figure 9.1.4-10 1 61 Figure 9.1.4-11 1 61 Figure 9.1.4-12 1 61 Figure 9.1.4-13 1 61

SHNPP FSAR Page 63 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 9.2.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.1-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.2-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.2-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.3-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.3-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.4-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.6-1 1 61 Figure 9.2.8-1 1 61 Figure 9.2.8-2 1 63 Figure 9.2.8-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.9-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.9-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.10-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.2.10-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.1-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.1-2a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.1-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.1-3a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-1a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-1b (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-2a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-2b (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.3-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.3-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.3-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.3-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 64 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 9.3.3-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-5 1 61 Figure 9.3.4-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.4-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.3.6-1 2 61 Figure 9.4.0-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.0-2 1 61 Figure 9.4.1-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.2-2 1 61 Figure 9.4.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.3-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.4-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.5-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.5-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.9-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.4.9-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.5.2-1 1 61 Figure 9.5.2-2 1 61 Figure 9.5.2-3 1 61 Figure 9.5.2-4 (Deleted by Amendment 10) --- N/A Figure 9.5.2-5 1 61 Figure 9.5.2-6 1 61 Figure 9.5.3-1 1 61 Figure 9.5.3-2 1 61 Figure 9.5.4-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 65 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 9.5.4-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.5.5-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.5.5-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.5.6-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 9.5.7-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figures 9.5A-1 through Figure 9.5A-41 (Deleted by Amendment 53) --- N/A Chapter 10 - Steam and Power Conversion System Table of Contents 4 65 Chapter 10 82 65 Tables - List of Tables Table 10.1.0-1 3 62 Table 10.2.1-1 1 62 Table 10.3.1-1 2 61 Table 10.3.2-1 1 61 Table 10.4.1-1 1 61 Table 10.4.1-2 1 61 Table 10.4.4-1 1 61 Table 10.4.4-2 1 61 Table 10.4.5-1 1 61 Table 10.4.5-2 1 61 Table 10.4.5-3 1 61 Table 10.4.6-1 (Deleted by Amendment 26) --- N/A Table 10.4.6-2 1 61 Table 10.4.6-3 1 61 Table 10.4.7-1 1 61 Table 10.4.7-2 1 61 Table 10.4.7-3 1 61 Table 10.4.7-4 1 61 Table 10.4.8-1 1 61 Table 10.4.9-1 1 61 Table 10.4.9-2 1 61 Table 10.4.9-3 5 61 Table 10.4.9A-1 4 61 Table 10.4.9A-2 15 63 Table 10.4.9A-3 7 61

SHNPP FSAR Page 66 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 10.4.9A-4 1 61 Table 10.4.9A-5 9 61 Table 10.4.9A-6 7 61 Table 10.4.9A-7 4 61 Table 10.4.9A-8 1 61 Table 10.4.9A-9 2 61 Table 10.4.9A-10 3 61 Table 10.4.9A-11 2 61 Table 10.4.9B-1 1 61 Table 10.4.9B-2 2 65 Table 10.4.9B-3 1 61 Table 10.4.9B-4 3 61 Figures - List of Figures Figure 10.1.0-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-3a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-6a (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.1.0-7 1 62 Figure 10.1.0-8 (Deleted by Amendment 51) --- N/A Figure 10.1.0-9 (Deleted by Amendment 62) --- N/A Figure 10.2.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-4 1 61 Figure 10.2.2-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-9 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.2.2-10 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) 1 62 Figure 10.4.5-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.5-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 67 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 10.4.6-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.6-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.6-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.7-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.7-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.7-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.7-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 10.4.9B-1 1 65 Figure 10.4.9B-2 1 63 Chapter 11 - Radioactive Waste Management Table of Contents 3 65 Chapter 11 70 65 Tables - List of Tables Table 11.1.1-1 3 61 Table 11.1.1-2 1 61 Table 11.1.1-3 1 61 Table 11.1.1-4 1 61 Table 11.1.1-5 2 61 Table 11.1.2-1 2 61 Table 11.1.2-2 1 61 Table 11.1.4-1 1 61 Table 11.1.6-1 1 61 Table 11.1.7-1 1 61 Table 11.1.8-1 1 61 Table 11.1.9-1 1 61 Table 11.2.1-1 1 61 Table 11.2.1-2 1 61 Table 11.2.1-3 1 61 Table 11.2.1-4 1 61 Table 11.2.1-5 1 61 Table 11.2.1-6 1 61 Table 11.2.1-7 4 65 Table 11.2.1-8 (Deleted by Amendment 43) --- N/A Table 11.2.1-9 (Deleted by Amendment 43) --- N/A Table 11.2.1-10 (Deleted by Amendment 43) --- N/A

SHNPP FSAR Page 68 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 11.2.1-11 (Deleted by Amendment 43) --- N/A Table 11.2.1-12 (Deleted by Amendment 43) --- N/A Table 11.2.1-13 (Deleted by Amendment 43) --- N/A Table 11.2.1-14 (Deleted by Amendment 43) --- N/A Table 11.2.1-15 (Deleted by Amendment 43) --- N/A Table 11.2.3-1 1 61 Table 11.2.3-2 1 61 Table 11.2.3-3 1 61 Table 11.2.3-4 1 61 Table 11.2.3-5 1 61 Table 11.3.2-1 1 61 Table 11.3.2-2 1 61 Table 11.3.2-3 1 61 Table 11.3.2-4 (Deleted by Amendment 46) --- N/A Table 11.3.2-5 (Deleted by Amendment 46) --- N/A Table 11.3.2-6 1 65 Table 11.3.2-7 2 61 Table 11.3.2-8 1 61 Table 11.3.3-1 1 61 Table 11.3.3-2 1 61 Table 11.3.3-3 1 61 Table 11.3.3-4 1 61 Table 11.3.3-5 1 61 Table 11.4.1-1 1 61 Table 11.4.1-1a 1 61 Table 11.4.1-2 1 61 Table 11.4.1-3 2 61 Table 11.4.1-4 2 61 Table 11.4.1-5 2 61 Table 11.4.1-6 3 65 Table 11.4.1-7 4 65 Table 11.4.1-8 1 61 Table 11.4.1-9 1 61 Table 11.4.2-1 1 61 Table 11.4.2-1a 1 65 Table 11.4.2-2 2 61

SHNPP FSAR Page 69 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 11.4.2-3 2 61 Table 11.4.2-3a 1 65 Table 11.4.2-3b 1 65 Table 11.4.2-4 1 61 Table 11.4.2-5 (Deleted by Amendment 43) --- N/A Table 11.4.2-6 (Deleted by Amendment 43) --- N/A Table 11.4.2-7 (Deleted by Amendment 43) --- N/A Table 11.4.2-8 (Deleted by Amendment 43) --- N/A Table 11.4.2-9 (Deleted by Amendment 43) --- N/A Table 11.5.1-1 1 61 Table 11.5.2-1 1 61 Table 11.5.2-2 2 64 Figures - List of Figures Figure 11.2.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.2.2-9 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.3.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.3.2-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.3.2-3 1 61 Figure 11.3.2-4 1 61 Figure 11.3.2-5 1 61 Figure 11.3.2-6 1 61 Figure 11.3.2-7 1 62 Figure 11.3.2-8 1 62 Figure 11.4.2-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) 61 Figure 11.4.2-2 (Deleted by Amendment 15) --- N/A Figure 11.4.2-3 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.4.2-4 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.4.2-5 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.4.2-6 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61

SHNPP FSAR Page 70 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 11.4.2-7 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.4.2-8 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.4.2-9 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 11.5.2-1 (Deleted by Amendment 64) --- N/A Chapter 12 - Radiation Protection Table of Contents 4 65 Chapter 12 103 65 Tables - List of Tables Table 12.2.1-1 1 61 Table 12.2.1-2 1 61 Table 12.2.1-3 1 61 Table 12.2.1-4 1 61 Table 12.2.1-5 1 61 Table 12.2.1-6 1 61 Table 12.2.1-7 1 61 Table 12.2.1-8 1 61 Table 12.2.1-9 1 61 Table 12.2.1-10 1 61 Table 12.2.1-11 1 61 Table 12.2.1-12 1 61 Table 12.2.1-13 1 61 Table 12.2.1-14 1 61 Table 12.2.1-15 1 61 Table 12.2.1-16 1 61 Table 12.2.1-17 1 61 Table 12.2.1-18 1 61 Table 12.2.1-19 1 61 Table 12.2.1-20 1 61 Table 12.2.1-21 1 61 Table 12.2.1-22 1 61 Table 12.2.1-23 1 61 Table 12.2.1-24 1 61 Table 12.2.1-25 1 61 Table 12.2.1-26 1 61 Table 12.2.1-27 1 61

SHNPP FSAR Page 71 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 12.2.2-1 1 61 Table 12.2.2-2 1 61 Table 12.2.2-3 1 65 Table 12.2.2-4 1 61 Table 12.3.1-1 1 61 Table 12.3.1-2 1 61 Table 12.3.1-3 1 61 Table 12.3.2-1 1 61 Table 12.3.2-2 1 61 Table 12.3.2-3 1 65 Table 12.3.2-4 (Deleted by Amendment 48) --- N/A Table 12.3.2-5 (Deleted by Amendment 48) --- N/A Table 12.3.2-6 (Deleted by Amendment 48) --- N/A Table 12.3.2-7 (Deleted by Amendment 48) --- N/A Table 12.3.2-8 (Deleted by Amendment 48) --- N/A Table 12.3.2-9 1 61 Table 12.3.4-1 3 65 Table 12.3.4-2 1 64 Table 12.4.1-1 1 61 Table 12.4.2-1 (Deleted by Amendment 48) --- N/A Table 12.4.2-2 (Deleted by Amendment 48) --- N/A Table 12.4.2-3 (Deleted by Amendment 48) --- N/A Table 12.4.2-4 (Deleted by Amendment 48) --- N/A Table 12.4.2-5 (Deleted by Amendment 48) --- N/A Table 12.4.2-6 (Deleted by Amendment 48) --- N/A Table 12.4.2-7 (Deleted by Amendment 48) --- N/A Table 12.4.2-8 (Deleted by Amendment 48) --- N/A Table 12.4.2-9 (Deleted by Amendment 48) --- N/A Table 12.4.2-10 (Deleted by Amendment 48) --- N/A Table 12.4.2-11 (Deleted by Amendment 48) --- N/A Table 12.4.2-12 (Deleted by Amendment 48) --- N/A Table 12.4.2-13 1 61 Table 12.5.2-1 (Deleted by Amendment 48) --- N/A Figures - List of Figures Figure 12.1.1-1 1 61 Figure 12.2.1-1 1 61

SHNPP FSAR Page 72 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 12.2.1-2 1 61 Figure 12.2.1-3 1 61 Figure 12.3.2-1 1 61 Figure 12.3.2-2 1 61 Figure 12.3.2-3 1 61 Figure 12.3.2-4 1 61 Figure 12.3.2-5 1 61 Figure 12.3.2-6 1 61 Figure 12.3.2-7 1 61 Figure 12.3.2-8 1 61 Figure 12.3.2-9 1 61 Figure 12.3.2-10 1 61 Figure 12.3.2-11 1 61 Figure 12.3.2-12 1 61 Figure 12.3.2-13 1 61 Figure 12.3.2-14 1 61 Figure 12.3.2-15 1 61 Figure 12.3.2-16 1 61 Figure 12.3.2-17 1 61 Figure 12.3.2-18 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 12.3.3-1 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 12.3.3-2 (Refer to FSAR Table 1.6-3 for Design Document Incorporated by Reference) --- 61 Figure 12.3A-1 1 61 Figure 12.3A-2 1 61 Figure 12.3A-3 1 61 Figure 12.3A-4 1 61 Figure 12.3A-5 1 61 Figure 12.3A-6 1 61 Figure 12.3A-7 1 61 Figure 12.3A-8 1 61 Figure 12.3A-9 1 61 Figure 12.3A-10 1 61 Figure 12.3A-11 1 61 Figure 12.3A-12 1 61 Figure 12.3A-13 1 61 Figure 12.3A-14 1 61

SHNPP FSAR Page 73 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 12.3A-15 1 61 Figure 12.3A-16 1 61 Figure 12.3A-17 1 61 Figure 12.3A-18 1 61 Figure 12.3A-19 1 61 Figure 12.3A-20 1 61 Figure 12.3A-21 1 61 Chapter 13 - Conduct of Operations Table of Contents 2 65 Chapter 13 32 65 Tables - List of Tables Table 13.1.2-1 (Deleted by Amendment 48) --- N/A Table 13.7-1 1 65 Figures - List of Figures Figure 13.1.1-1 1 64 Figure 13.1.2-1 1 64 Figure 13.1.2-2 (Deleted by Amendment 59) --- N/A Figure 13.4.2-1 (Deleted by Amendment 46) --- N/A Figure 13.5.1-1 1 61 Chapter 14 - Initial Test Program Table of Contents 2 65 Chapter 14 120 65 Tables - List of Tables There are no tables for Chapter 14 N/A N/A Figures - List of Figures Figure 14.2.11-1 1 61 Figure 14.2.11-2 1 61 Chapter 15 - Accident Analysis Table of Contents 8 65 Chapter 15 147 65 Tables - List of Tables Table 15.0.1-1 2 65 Table 15.0.3-1 (Deleted by Amendment 48) --- N/A Table 15.0.3-2 (Deleted by Amendment 48) --- N/A

SHNPP FSAR Page 74 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 15.0.3-3 (Deleted by Amendment 48) --- N/A Table 15.0.3-4 (Deleted by Amendment 48) --- N/A Table 15.0.3-5 2 63 Table 15.0.3-6 (Deleted by Amendment 63) --- N/A Table 15.0.6-1 (Deleted by Amendment 48) --- N/A Table 15.0.6-2 2 63 Table 15.0.7-1 (Deleted by Amendment 48) --- N/A Table 15.0.8-1 3 64 Table 15.0.9-1 1 64 Table 15.0.9-2 1 61 Table 15.0.9-3 1 61 Table 15.0.9-4 1 61 Table 15.0.9-5 1 61 Table 15.0.9-6 1 61 Table 15.0.9-7 1 61 Table 15.0.13-1 1 65 Table 15.0.13-2 1 64 Table 15.1.2-1 1 64 Table 15.1.2-2 1 64 Table 15.1.2-3 1 64 Table 15.1.2-4 1 64 Table 15.1.2-5 (Deleted by Amendment 51) --- N/A Table 15.1.3-1 1 63 Table 15.1.3-2 1 63 Table 15.1.3-3 1 63 Table 15.1.3-4 1 63 Table 15.1.4-1 (Deleted by Amendment 48) --- N/A Table 15.1.5-1 1 61 Table 15.1.5-2 (Deleted by Amendment 48) --- N/A Table 15.1.5-3 1 64 Table 15.1.5-4 (Deleted by Amendment 51) --- N/A Table 15.1.5-5 1 63 Table 15.1.5-6 1 63 Table 15.1.5-7 1 61 Table 15.2.3-1 1 63 Table 15.2.3-2 1 63

SHNPP FSAR Page 75 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 15.2.3-3 1 63 Table 15.2.3-4 1 63 Table 15.2.3-5 1 63 Table 15.2.3-6 1 63 Table 15.2.6-1 (Deleted by Amendment 65) --- N/A Table 15.2.6-2 (Deleted by Amendment 65) --- N/A Table 15.2.6-3 (Deleted by Amendment 65) --- N/A Table 15.2.6-4 (Deleted by Amendment 65) --- N/A Table 15.2.6-5 1 61 Table 15.2.6-6 1 63 Table 15.2.7-1 1 65 Table 15.2.7-2 1 65 Table 15.2.7-3 1 65 Table 15.2.7-4 1 65 Table 15.2.7-5 1 65 Table 15.2.7-6 1 65 Table 15.2.8-1 1 63 Table 15.2.8-2 1 63 Table 15.2.8-3 1 63 Table 15.2.8-4 1 63 Table 15.2.8-5 1 63 Table 15.2.8-6 1 61 Table 15.2.8-7 1 63 Table 15.2.8-8 1 63 Table 15.3.1-1 (Deleted by Amendment 48) --- N/A Table 15.3.2-1 1 63 Table 15.3.2-2 1 63 Table 15.3.2-3 1 63 Table 15.3.2-4 1 63 Table 15.3.3-1 1 63 Table 15.3.3-2 1 63 Table 15.3.3-3 1 63 Table 15.3.3-4 1 64 Table 15.3.3-5 1 64 Table 15.3.3-6 1 64 Table 15.3.3-7 1 64

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SHNPP FSAR Page 78 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 15.6.5-14 1 61 Table 15.6.5-15 2 64 Table 15.6.5-16 1 64 Table 15.6.5-17 (Deleted by Amendment 63) --- N/A Table 15.6.5-18 (Deleted by Amendment 63) --- N/A Table 15.7.1-1 1 61 Table 15.7.1-2 1 63 Table 15.7.2-2 (Deleted by Amendment 51) --- N/A Table 15.7.2-3 (Deleted by Amendment 51) --- N/A Table 15.7.4-1 1 64 Table 15.7.4-2 1 64 Table 15.7.4-3 1 64 Table 15.7.4-4 1 64 Table 15.7.4-5 (Deleted by Amendment 49) --- N/A Table 15.7.4-6 (Deleted by Amendment 49) --- N/A Table 15.7.4-7 (Deleted by Amendment 49) --- N/A Table 15.7.4-8 (Deleted by Amendment 51) --- N/A Table 15.7.4.9 (Deleted by Amendment 51) --- N/A Table 15.7.5-1 1 61 Figures - List of Figures Figure 15.0.5-4 (Deleted by Amendment 63) --- N/A Figure 15.0.5-5 (Deleted by Amendment 63) --- N/A Figure 15.0.5-6 1 63 Figure 15.0A.1-1 1 61 Figure 15.1.2-1 1 64 Figure 15.1.2-2 1 64 Figure 15.1.2-3 1 64 Figure 15.1.2-4 1 64 Figure 15.1.2-5 1 64 Figure 15.1.2-6 (Deleted by Amendment 51) --- N/A Figure 15.1.2-7 (Deleted by Amendment 51) --- N/A Figure 15.1.2-8 (Deleted by Amendment 51) --- N/A Figure 15.1.2-9 (Deleted by Amendment 51) --- N/A Figure 15.1.2-10 (Deleted by Amendment 51) --- N/A Figure 15.1.3-1 1 63 Figure 15.1.3-2 1 63

SHNPP FSAR Page 79 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 15.1.3-3 1 63 Figure 15.1.3-4 1 63 Figure 15.1.4-1 (Deleted by Amendment 48) --- N/A Figure 15.1.4-2 (Deleted by Amendment 48) --- N/A Figure 15.1.4-3 (Deleted by Amendment 48) --- N/A Figure 15.1.4-4 (Deleted by Amendment 48) --- N/A Figure 15.1.4-5 (Deleted by Amendment 48) --- N/A Figure 15.1.5-1 1 63 Figure 15.1.5-2 1 63 Figure 15.1.5-3 1 63 Figure 15.1.5-4 1 63 Figure 15.1.5-5 1 63 Figure 15.1.5-6 1 63 Figure 15.2.3-1 1 63 Figure 15.2.3-2 1 63 Figure 15.2.3-3 1 63 Figure 15.2.3-4 1 63 Figure 15.2.3-5 1 63 Figure 15.2.3-6 1 63 Figure 15.2.3-7 1 63 Figure 15.2.3-8 (Deleted by Amendment 51) --- N/A Figure 15.2.3-9 1 63 Figure 15.2.3-10 1 63 Figure 15.2.3-11 1 63 Figure 15.2.3-12 1 63 Figure 15.2.6-1 1 62 Figure 15.2.6-2 1 62 Figure 15.2.6-3 1 62 Figure 15.2.6-4 1 62 Figure 15.2.6-5 1 62 Figure 15.2.6-6 (Deleted by Amendment 51) --- N/A Figure 15.2.7-1 1 65 Figure 15.2.7-2 1 65 Figure 15.2.7-3 1 65 Figure 15.2.7-4 1 65 Figure 15.2.7-5 1 65

SHNPP FSAR Page 80 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 15.2.7-6 1 65 Figure 15.2.7-7 1 65 Figure 15.2.7-8 1 65 Figure 15.2.7-9 1 65 Figure 15.2.7-10 1 65 Figure 15.2.7-11 1 65 Figure 15.2.7-12 1 65 Figure 15.2.7-13 1 65 Figure 15.2.7-14 1 65 Figure 15.2.7-15 1 65 Figure 15.2.7-16 1 65 Figure 15.2.8-1 1 63 Figure 15.2.8-2 1 63 Figure 15.2.8-3 1 63 Figure 15.2.8-4 1 63 Figure 15.2.8-5 1 63 Figure 15.2.8-6 1 63 Figure 15.2.8-7 1 63 Figure 15.2.8-8 1 63 Figure 15.2.8-9 1 63 Figure 15.2.8-10 (Deleted by Amendment 63) --- N/A Figure 15.2.8-11 1 63 Figure 15.2.8-12 1 63 Figure 15.2.8-13 1 63 Figure 15.2.8-14 1 63 Figure 15.2.8-15 1 63 Figure 15.2.8-16 1 63 Figure 15.2.8-17 1 63 Figure 15.2.8-18 1 63 Figure 15.2.8-19 1 63 Figure 15.2.8-20 (Deleted by Amendment 63) --- N/A Figure 15.2.8-21 1 63 Figure 15.2.8-22 1 63 Figure 15.2.8-23 1 63 Figure 15.2.8-24 1 63 Figure 15.2.8-25 1 63

SHNPP FSAR Page 81 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 15.2.8-26 1 63 Figure 15.2.8-27 1 63 Figure 15.2.8-28 1 63 Figure 15.2.8-29 1 63 Figure 15.2.8-30 1 63 Figure 15.2.8-31 1 63 Figure 15.2.8-32 1 63 Figure 15.3.1-2 (Deleted by Amendment 51) --- N/A Figure 15.3.1-3 (Deleted by Amendment 51) --- N/A Figure 15.3.2-1 1 63 Figure 15.3.2-2 1 63 Figure 15.3.2-3 1 63 Figure 15.3.2-4 1 63 Figure 15.3.2-5 1 63 Figure 15.3.2-6 1 63 Figure 15.3.2-7 1 63 Figure 15.3.3-1 1 64 Figure 15.3.3-2 1 64 Figure 15.3.3-3 1 64 Figure 15.3.3-4 1 64 Figure 15.3.3-5 1 64 Figure 15.3.3-6 1 64 Figure 15.3.3-7 1 64 Figure 15.3.3-8 1 64 Figure 15.3.3-9 1 64 Figure 15.3.3-10 1 64 Figure 15.3.3-11 1 64 Figure 15.3.3-12 1 64 Figure 15.3.3-13 1 64 Figure 15.4.1-1 1 63 Figure 15.4.1-2 1 63 Figure 15.4.1-3 1 63 Figure 15.4.1-4 1 63 Figure 15.4.1-5 (Deleted by Amendment 51) --- N/A Figure 15.4.1-6 (Deleted by Amendment 51) --- N/A Figure 15.4.2-1 1 63

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SHNPP FSAR Page 84 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Figure 15.6.3-1 1 61 Figure 15.6.3-2 1 61 Figure 15.6.3-3 1 61 Figure 15.6.3-4 1 61 Figure 15.6.3-5 1 61 Figure 15.6.3-6 1 61 Figure 15.6.3-7 1 61 Figure 15.6.3-8 1 61 Figure 15.6.3-9 1 61 Figure 15.6.3-10 1 61 Figure 15.6.3-11 1 61 Figure 15.6.3-12 1 61 Figure 15.6.3-13 1 61 Figure 15.6.3-14 1 61 Figure 15.6.3-15 1 61 Figure 15.6.3-16 1 61 Figure 15.6.3-17 1 61 Figure 15.6.3-18 1 61 Figure 15.6.3-19 1 61 Figure 15.6.3-20 (Deleted by Amendment 51) --- N/A Figure 15.6.3-21 (Deleted by Amendment 51) --- N/A Figure 15.6.3-22 1 61 Figure 15.6.3-23 1 61 Figure 15.6.3-24 1 61 Figure 15.6.5-1 1 64 Figure 15.6.5-2 (Deleted by Amendment 64) --- N/A Figure 15.6.5-3 1 64 Figure 15.6.5-4 (Deleted by Amendment 64) --- N/A Figure 15.6.5-5 (Deleted by Amendment 64) --- N/A Figure 15.6.5-6 1 64 Figure 15.6.5-7 1 64 Figure 15.6.5-8 1 64 Figure 15.6.5-9 (Deleted by Amendment 64) --- N/A Figure 15.6.5-10 (Deleted by Amendment 64) --- N/A Figure 15.6.5-11 (Deleted by Amendment 64) --- N/A Figure 15.6.5-12 (Deleted by Amendment 64) --- N/A

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SHNPP FSAR Page 86 of 86 LIST OF EFFECTIVE PAGES Total Amendment Title Pages No.Table 16.3-3 1 61 Table 16.3-4 1 61 Table 16.3-5 1 61 Table 16.3-6 1 61 Table 16.3-7 1 61 Table 16.3-8 1 61 Table 16.3-9 1 61 Figures - List of Figures Figure 16.3-1 (Deleted by Amendment 56) --- N/A Chapter 17 - Quality Assurance Table of Contents 1 65 Chapter 17 1 65 Tables - List of Tables There are no tables for Chapter 17 N/A N/A Figures - List of Figures There are no figures for Chapter 17 N/A N/A Chapter 18 - Final Safety Analysis Report Supplement for License Renewal Table of Contents 3 65 Chapter 18 34 65 Tables - List of Tables There are no tables for Chapter 18 N/A N/A Figures - List of Figures There are no figures for Chapter 18 N/A N/A Appendices TMI Appendix 16 65 Appendix A - Index of NRC Question Responses and FSAR Location 19 65 Appendix B - DSER Open Items Index 35 65

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT ................................... 1

1.1 INTRODUCTION

.......................................................................................................... 1 1.1.1 GENERAL INFORMATION ..................................................................................... 1 1.1.2 STATION LOCATION ............................................................................................. 1 1.1.3 NUCLEAR STEAM SUPPLIER ............................................................................... 1 1.1.4 CONTAINMENT ...................................................................................................... 2 1.1.5 CORE THERMAL POWER ..................................................................................... 2 1.1.6 SCHEDULE ............................................................................................................. 2 1.2 GENERAL PLANT DESCRIPTION ................................................................................ 2 1.2.1 PRINCIPAL SITE CHARACTERISTICS ................................................................. 3 1.2.1.1 Location and Population.............................................................................. 3 1.2.1.2 Hydrology .................................................................................................... 3 1.2.1.3 Meteorology ................................................................................................ 4 1.2.1.4 Geology and Seismology ............................................................................ 4 1.2.1.5 Nearby Industry and Comments ................................................................. 5 1.2.2 CONCISE PLANT DESCRIPTION .......................................................................... 5 1.2.2.1 Principal Structures ..................................................................................... 5 1.2.2.2 Nuclear Steam Supply System ................................................................... 5 1.2.2.3 Engineered Safety Features ....................................................................... 6 1.2.2.4 Instrumentation and Control Systems ......................................................... 7 1.2.2.5 Electrical System....................................................................................... 10 1.2.2.6 Steam and Power Conversion System ..................................................... 10 1.2.2.7 Nuclear Fuel Handling and Storage Systems ........................................... 11 1.2.2.8 Cooling Water and Other Auxiliary Systems ............................................. 11 1.2.2.9 Waste Processing System ........................................................................ 14 1.2.2.10 Radiation Monitoring System .................................................................... 15 1.3 COMPARISON TABLES .............................................................................................. 15 1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS .......................................... 15 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION......................... 15 Amendment 65 Page i of ii

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS .............................................. 15 1.4.1 OWNERS .............................................................................................................. 15 1.4.2 APPLICANTS ........................................................................................................ 15 1.4.3 ARCHITECT/ENGINEER ...................................................................................... 16 1.4.4 CONTRACTORS ................................................................................................... 16 1.4.5 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) SUPPLIER ................................. 17 1.4.6 CONSULTANTS .................................................................................................... 17 1.4.6.1 Dames & Moore ........................................................................................ 17 1.4.6.2 Dr. Jasper L. Stuckey ................................................................................ 17 1.4.6.3 Dr. B. J. Copeland ..................................................................................... 17 1.4.6.4 Dr. Joffre L. Coe, Director, Research Laboratories of Anthropology, University of North Carolina ...................................................................... 17 1.4.6.5 Aquatic Control, Inc. .................................................................................. 17 1.4.6.6 Mr. William Beck, Florida A&M University ................................................. 17 1.4.6.7 Dr. Samuel Mozley, North Carolina State University ................................ 17 1.4.6.8 Pickard, Lowe & Garrick, Inc. .................................................................... 17 1.4.6.9 Research Triangle Institute ....................................................................... 17 1.4.6.10 H.A.F.A. International, Inc. ........................................................................ 18 1.4.6.11 TERA Corporation ..................................................................................... 18 1.4.6.12 Law Engineering Testing Co. .................................................................... 18 1.4.6.13 Nello L. Teer Co. ....................................................................................... 18 1.4.6.14 Southwest Research Institute ................................................................... 18 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION .............................. 18 1.6 MATERIAL INCORPORATED BY REFERENCE ......................................................... 18 1.7 DRAWINGS AND OTHER DETAILED INFORMATION ............................................... 19

REFERENCES:

SECTION 1.7 ............................................................................................ 19 1.8 CONFORMANCE TO NRC REGULATORY GUIDES .................................................. 19

REFERENCES:

SECTION 1.8 ........................................................................................... 69 Amendment 65 Page ii of ii

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1.1.1 GENERAL INFORMATION This Final Safety Analysis Report (FSAR) is submitted in support of the Carolina Power & Light Company's (CP&L) application for a Class 103 facility operating license for the Shearon Harris Nuclear Power Plant (SHNPP). This FSAR has been organized in accordance with the guidelines contained in Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants" (Revision 3 dated November 1978), and the regulations of the NRC set forth in 10 CFR 50.On July 2, 2012, Duke Energy Corporation completed closure of a corporate merger with Progress Energy (i.e., Carolina Power & Light Company). Subsequently, on October 21, 2013, the NRC approved changing the licensee name from Carolina Power & Light Company to Duke Energy Progress, Inc. Until all sections of the FSAR are updated to incorporate this name change, where Carolina Power & Light Company (CP&L) or Progress Energy is used, except in an historical context, they shall mean Duke Energy Progress, Inc.Lists of acronyms, abbreviations and names of major buildings and structures used throughout this FSAR are given in Tables 1.1.1-1, 1.1.1-2, and 1.1.1-3, respectively. Figures 1.1.1-1 and 1.1.1-2 provide flow diagram symbols while Figure 1.1.1-1a provides piping and instrumentation symbols used on engineering drawings throughout the FSAR.Duke Energy Progress, Inc. owns the plant. Duke Energy Progress, Inc. has the overall responsibility to ensure that it is designed, constructed, and operated without undue risk to the health and safety of the public. Ebasco Services, Incorporated is the architect/engineer responsible for the design, engineering, and equipment and material procurement for SHNPP.This includes all plant structures, systems, and components except for those provided by Westinghouse Electric Corporation, the Nuclear Steam Supply System (NSSS) Supplier. Daniel Construction Company, Inc., as the constructor, performed the major part of the plant construction. Selected portions of the work, however, were performed by other contractors under direct supervision of CP&L.1.1.2 STATION LOCATION The SHNPP site is located in the extreme southwest corner of Wake County, North Carolina, and the southeast corner of Chatham County, North Carolina. The city of Raleigh, North Carolina, is approximately 16 miles northeast and the city of Sanford, North Carolina, is about 15 miles southwest.1.1.3 NUCLEAR STEAM SUPPLIER The Nuclear Steam Supply System (NSSS) for the Unit is a pressurized water reactor (PWR) consisting of three closed reactor coolant loops connected in parallel to the reactor vessel, each containing a reactor coolant pump and a steam generator. An electrically heated pressurizer is connected to the "hot" leg of one of the loops. The NSSS, along with the design and fabrication of the initial fuel core, is supplied by Westinghouse Electric Corporation.Amendment 65 Page 1 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.1.4 CONTAINMENT The Containment is a steel lined reinforced concrete structure in the form of a vertical right cylinder with a hemispherical dome and a flat base with a recess beneath the reactor vessel.The Containment is designed by Ebasco Services Incorporated, architect/engineer for SHNPP.1.1.5 CORE THERMAL POWER The Unit is licensed for a core thermal power output of 2948 megawatts thermal (Mwt). The total unit thermal output is approximately 2960.4 Mwt, which includes 12.4 Mwt from the reactor coolant pumps. The thermal output corresponds to an electrical output of approximately 930 megawatts electric (Mwe) net or 985 Mwe gross. All safety systems, including containment and engineered safety features, have been analyzed for operation at a core thermal power of up to 2958 MWt.The originally licensed reactor thermal power was 2775 Mwt. A licensing request was granted in the year 2001 to increase reactor power to 2900 Mwt. This was accomplished in conjunction with replacing the steam generators in Refueling Outage number 10 (RFO-10), in the fall/winter of 2001. The effects of these major changes to the plant are seen throughout the FSAR and are sometimes noted with SGR for Steam Generator Replacement Project and PUR for Power Uprate Project.In the spring of 2012, (RFO-17) implementation of a Measurement Uncertainty Recapture -Power Uprate (MUR-PU) was implemented by a licensing request allowing changes to the 10 CFR 50, Appendix K power measurement uncertainty requirements. This change reduced the required power uncertainty margin from 2% (measuring feedwater flow with the venturis) to the uncertainty associated with measuring feedwater flow using the Caldon Leading Edge Flow Meters (LEFMs) installed during RFO-16, in the fall of 2010. The more accurate LEFMs reduced the core power measurement uncertainty from 2.0% to 0.34%, allowing an increase in reactor core power of 1.66% from 2900 MWt to 2948 MWt. The effects of this change to the plant are seen throughout the FSAR and are sometimes noted with MUR-PU. Unless otherwise noted, the values and other information contained in the FSAR are based on the plant configuration after MUR-PU. The MUR-PU is defined as an increase in core thermal power from 2900 MWt (post SGR/PUR) to a (current) core power of 2948 MWt. These are nominal values without uncertainties.1.1.6 SCHEDULE The construction schedule for SHNPP is based on a commercial operation date in the fourth quarter of 1986. This schedule requires that an operating license be issued in time for fuel loading by June 1986.1.2 GENERAL PLANT DESCRIPTION The Shearon Harris Nuclear Power Plant (SHNPP) is designed to function as an electric generating station, and generate electric power utilizing a pressurized water reactor (PWR) and a closed regenerative cycle steam turbine-generator. The inherent design of the pressurized water, closed-cycle reactor minimizes the quantities of fission products released to the atmosphere. Four barriers exist between the fission product accumulation and the environment.These are the uranium dioxide fuel matrix, the fuel cladding, the reactor vessel and reactor Amendment 65 Page 2 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 coolant loops, and the Containment. The consequences of a breach of the fuel cladding are greatly reduced by the ability of the uranium dioxide lattice to retain fission products. Escape of fission products through fuel cladding defects would be contained within the reactor pressure vessel, reactor coolant loops and associated auxiliary systems. Breach of these systems or equipment would release the fission products to the Containment where they would be contained. The Containment is designed to adequately contain these fission products under the most severe accident conditions, as analyzed in Chapter 15.Several engineered safety features have been incorporated into the plant design to reduce the consequences of a loss-of-coolant-accident (LOCA). These safety features include an Emergency Core Cooling System (ECCS). This system automatically delivers borated water to the reactor vessel for cooling the core under high and low reactor coolant pressure conditions.The ECCS also serves to effectively insert negative reactivity into the core in the form of borated water during Unit cooldown following a steam line break or an accidental steam release.Additional safety features include the Containment Spray System (CSS) which serves to remove thermal energy from the Containment in the event of a LOCA. The concentration of post LOCA airborne iodine fission products within the Containment will also be reduced by the CSS.1.2.1 PRINCIPAL SITE CHARACTERISTICS 1.2.1.1 Location and Population The site occupies approximately 10,723 acres of land in southwest Wake County and southeast Chatham County, North Carolina. The environment is rural and primarily devoted to farming and dairying. Local industrial activity is centered in an area west-southwest from the plant.Another major center of industrial and research activity is located to the north-northwest. The exclusion area is shown on Figure 2.1.2-1. The shortest distance to the exclusion boundary is 6640 ft. in the northwest direction; the longest distance is 7200 ft. in the south direction. The area within the exclusion boundary is directly controlled by DEP. The three mile low population zone was populated by 534 persons as of 1970. The nearest population center as defined in 10 CFR 100 is Cary at a distance of 10 miles to the city's nearest boundary. The fifty-mile radius population was approximately 1,300,000 people (1980 population). Population projections for the period of operation of the plant are located in Section 2.1.3. See Section 2.1 for a more detailed discussion of geography and demography.1.2.1.2 Hydrology The plant site is located at the confluence of Buckhorn and Whiteoak Creeks, just north of the Cape Fear River. The power block area is located between Tom Jack and Thomas Creeks.Figure 2.1.1-1 shows a plan of the site development.The principal water source for the plant is the Main Reservoir which is formed by an impoundment of Buckhorn Creek just below its confluence with Whiteoak Creek. The project design also includes an adjoining and independent Auxiliary Reservoir for emergency cooling purposes. See Section 2.4 for a more detailed discussion of hydrology.Amendment 65 Page 3 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.2.1.3 Meteorology The meteorological conditions of the plant site are those typical of the transition zone delineating the Coastal Plain and Piedmont climatological classifications. Climatology of North Carolina is largely dependent upon elevation above sea level and distance from the Atlantic Ocean. The inland location of the plant site (115 miles from the Atlantic Ocean) modifies the effects of coastal storms, both tropical and extratropical, so that they are reduced in intensities to levels which are generally no greater than those produced by regional heavy thunderstorms.The severity of continental air mass systems approaching from the northwest is modified by the Appalachian Mountain range which acts as a protective barrier. The area is not on a usual path of either continental or coastal cyclonic storm centers, although frontal passages are frequent.The major weather influence in the region is the predominance of the subtropical belt of high pressure.Influence of the Atlantic Ocean is reflected in generally high moisture content of the air masses usually over the region. The thermal effects of the ocean coupled with the barrier formed by the mountain range, result in a high frequency of occurrence of northeasterly winds in the fall and of southwesterly winds in the spring. The predominant annual wind direction at the plant site occurs from the southwesterly sectors; however, the bimodal wind direction characteristics of the region are evident from the onsite data. See Section 2.3 for a more detailed discussion of meteorology.1.2.1.4 Geology and Seismology The region surrounding the site is generally characterized by a gently rolling topography resulting from extensive weathering and erosion of the underlying bedrock. The site is located in the southeastern part of the Durham Basin, which is in the northern part of the Deep River Triassic Basin. Sediments that underlie much of the southeastern portion of the Durham Basin were placed as alluvial fans and stream channels and flood plain deposits. Below an occasional thin layer of alluvial sand and/or clay, there are from 0 to 15 ft. of residual soil. The depth of weathering below this to sound rock generally varies from about 0 to 15 ft. depending on the type of underlying rock. The foundations have been placed on sound rock.A small fault was discovered during excavation for the Waste Processing Building. The studies performed showed that this fault is not a capable fault, as documented in the Shearon Harris Fault Investigation Report submitted to the NRC in 1975.The nearest known fault outside the site is one lying just west of Merry Oaks about three miles to the southwest of the site.Test borings showed nothing that would indicate the development of faults, joints, slickensides or other structural weakness since the late Triassic and early Jurassic time.The site ground accelerations for the Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE) are 0.150 and 0.075g, respectively. See Section 2.5 for a more detailed discussion of geology and seismology.Amendment 65 Page 4 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.2.1.5 Nearby Industry and Commerce Industrial activity in the region surrounding the plant site is not intensive and is concentrated to the north-northwest. See Section 2.2 for more information.1.2.2 CONCISE PLANT DESCRIPTION 1.2.2.1 Principal Structures Major plant structures include the Containment Building; Reactor Auxiliary Building, which contains the Control Room; Turbine Building; Waste Processing Building; Diesel Generator Building; a Service Building; Fuel Handling Building; Tank Building; and Cooling Tower. Figure 1.2.2-1 indicates the site plan for the Shearon Harris Nuclear Power Plant. Figures 1.2.2-3 through 1.2.2-87 indicate the general arrangements of plant structures and major equipment.The seismic criteria used to design the structures and equipment in the plant are described in Sections 3.7, 3.8, and 3.10. The maximum horizontal ground acceleration for the operating condition is 0.075g (Operating Basis Earthquake). However, the design ensures that no undue risk to public health and safety results from a horizontal ground acceleration of 0.15g (Safe Shutdown Earthquake).1.2.2.2 Nuclear Steam Supply System The Nuclear Steam Supply System consists of a pressurized water reactor, a Reactor Coolant System (RCS), and associated auxiliary systems. The Reactor Coolant System is arranged as three closed reactor coolant loops connected in parallel to the reactor vessel, each containing a reactor coolant pump and steam generator. An electrically heated pressurizer is connected to the hot leg of one reactor coolant loop.The reactor core is composed of uranium dioxide pellets enclosed in pressurized Zirconium alloy or M5 tubes with welded end plugs. The tubes are supported in assemblies by a spring clip grid structure. The mechanical control rods consist of clusters of stainless steel clad silver-indium- cadmium or Hafnium absorber rods and Zircaloy or Q12 guide tubes located within the fuel assembly.Fuel rod cladding is designed to maintain cladding integrity throughout fuel life. Fission gas released within the rods and other factors affecting design life are considered for the maximum expected exposure. The reactor and control systems are designed so that any xenon transients will be adequately damped. Assuming the chemical shim requirements are met (i.e., proper concentration of boric acid in the coolant), the rod cluster control assemblies (RCCA) are capable of holding the core subcritical at hot zero power condition with margin following a trip, even with the most reactive RCCA stuck in the fully withdrawn position.The reactor, in conjunction with its protective systems, is designed to safely accommodate the anticipated operational occurrences. See Chapter 4 for further information. The reactor vessel and reactor internals contain and support the fuel and control rods. The reactor vessel is cylindrical with a hemispherical head and is clad internally with stainless steel. The pressurizer is a vertical cylindrical pressure vessel with a hemispherical head and is equipped with electrical heaters and spray nozzles for system pressure control.Amendment 65 Page 5 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The steam generators are vertical U-Tube type heat exchangers utilizing Inconel tubes. Internal moisture separating equipment reduces the moisture carryover of the steam at the outlet nozzle to 0.10 weight percent or less under maximum design load and ramp conditions defined in Section 5.4.2.The reactor coolant pumps are vertical, single stage, centrifugal pumps equipped with controlled leakage shaft seals.Auxiliary systems are provided to charge the RCS and to add makeup water, purify reactor coolant water, provide chemicals for corrosion inhibition and reactor control, remove residual heat when the reactor is shut down, provide for emergency safety injection, vent and drain the RCS, and provide sampling of reactor coolant water.The RCS is designed and constructed to maintain its integrity throughout the plant life.Appropriate means for testing and inspection are provided. The reactor for the Shearon Harris Nuclear Power Plant is designed with provisions to permit plutonium recycle. The provisions included in the design consist of a control rod pattern which is based on current technology, and additional hardware such as guide tubes, vessel head adapters, control rod drive mechanisms, control rods, control rod position indicators, switches, required power supplies, and control room indications. See Chapter 5 for further information.1.2.2.3 Engineered Safety Features The Engineered Safety Features (ESF) provided have sufficient redundancy of components and power sources such that under the conditions of a loss-of-coolant-accident (LOCA) they can maintain the integrity of the Containment and ensure that the limits of 10 CFR 50.67 are not exceeded even when operating with partial effectiveness. The main engineered safety features are the Emergency Core Cooling System (ECCS), the Containment Building, the Containment Spray System (CSS), the Containment Cooling System (CCS), Combustible Gas Control System and Control Room Habitability System. The functions they serve are summarized below, and other ESF systems are listed:a) The ECCS injects borated water into the RCS. The ECCS limits damage to the core and limits the energy and fission products released into the Containment following a loss-of-coolant-accident. All components necessary for the proper operation of the engineered safety features are operable from the Control Room.b) The steel-lined reinforced concrete Containment Building, with its associated Containment Isolation System (CIS), provides a reliable barrier against the escape of fission products under various environmental conditions following a LOCA. Containment isolation is initiated by several process variables as discussed in Sections 6.2.4 and 7.3.c) The Containment Spray System provides a spray of cool water containing sodium hydroxide to reduce containment pressure and remove the volatile iodine from the containment atmosphere following a LOCA.d) The Containment Cooling System consists of four cooling units which provide a redundant means of reducing the containment pressure following a LOCA.Amendment 65 Page 6 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 e) The Combustible Gas Control System maintains a safe post-LOCA hydrogen concentration within the Containment.f) The habitability systems insure control room habitability following a LOCA.See Chapter 6 for further information.1.2.2.4 Instrumentation and Control Systems Instrumentation and Control Systems provide the reactor operators with required information and control capability to operate in a safe and efficient manner. Where safety functions are involved, logic circuitry and actuators are provided to initiate equipment actions without operator assistance.1.2.2.4.1 Controls The Reactor Control System is used for startup and shutdown of the reactor and for adjustment of the reactor power in response to turbine load demand. The Nuclear Steam Supply System (NSSS) is capable of accommodating 10 percent of full power step changes in plant load and 5 percent of full power per minute ramp changes over the range from 15 percent up to 100 percent full power, without reactor trip.Overall reactivity control is achieved by the combination of chemical shim and rod cluster control assemblies (RCCA). Long-term regulation of core reactivity is accomplished by adjusting the concentration of boric acid in the reactor coolant. Short-term reactivity control for power change is accomplished by moving the RCCA.The reactor automatic control system is designed to maintain a programmed average temperature in the reactor coolant during steady state operation and to ensure that plant conditions do not reach trip settings as the result of a transient caused by a design load change.The function of the Reactor Control System is to provide automatic control of the RCCA during power operation of the reactor. The system uses input signals including neutron flux, coolant temperature, and turbine load. The Chemical and Volume Control System (Chapter 9) supplements the Reactor Control System by the addition and removal of varying amounts of boric acid solution.When the reactor is critical, the best indication of reactivity status in the core is the position of the rod control bank in relation to power and average coolant temperature. The direct relationship between control rod position and power is the relationship which establishes the lower insertion limit calculated by the rod insertion limit monitor. There are two alarm setpoints to alert the operator to take corrective action in the event a rod control bank approaches or reaches its lower insertion limit.Any unexpected change in the position of the control banks under automatic control, or a change in reactor coolant temperature under manual control provides a direct and immediate indication of a change in the reactivity status of the reactor. In addition, periodic samples are taken for determination of the reactor coolant boron concentration. The variation in concentration during core life provides a further check on the reactivity status of the reactor Amendment 65 Page 7 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 including core depletion. The provisions for monitoring the primary coolant boron concentration are discussed in Chapter 9.The Reactor Control System is designed to enable the reactor to follow load changes automatically when the output is above approximately 15 percent of rated power. Control rod positioning may be performed automatically when plant output is above this value, and manually at any time. The automatic control is limited by the control rod automatic withdrawal stopping at C-11 interlock and above.Following a reactor and turbine trip, residual heat stored in the reactor coolant is removed without actuating the steam generator safety valves by venting steam to atmosphere or bypassing steam to the condenser with power operated valves and addition of main or auxiliary feedwater. Reactor Coolant System temperature is reduced in this way to the no load, hot standby, or cold shutdown condition.1.2.2.4.2 Instrumentation The primary function of nuclear instrumentation is to safeguard the reactor by monitoring the neutron flux and generating appropriate trips and alarms for various phases of reactor operating and shutdown conditions. It also provides a secondary control function by indicating reactor status during start-up and power operation. The Nuclear Instrumentation System (NIS) uses information from three separate types of instrumentation channels to provide three discrete protection levels. Each range of instrumentation ("source", "intermediate", and "power")provides the overpower reactor trip protection required during operation in that range. The overlap of instrument ranges provides reliable continuous protection beginning with source level through the intermediate and low power level. As the reactor power increases, the overpower protection level is increased according to plant procedures after satisfactory higher range instrumentation operation is obtained. Automatic reset to more restrictive trip protection is provided when reducing power.Various types of neutron detectors, with appropriate solid-state electronic circuitry, are used to monitor the neutron flux from a completely shutdown condition to 120 percent of full power. The power range channels are capable of recording overpower excursions up to 200 percent of full power. The neutron flux covers a wide range between these extremes. Therefore, monitoring with several ranges of instrumentation is necessary. The lowest range ("source" range) covers six decades of neutron flux. The next range ("intermediate" range) covers eight decades.Detectors and instrumentation are chosen to provide overlap between the higher portion of the source range and the lower portion of the intermediate range. The highest range of instrumentation ("power" range) covers approximately 2.2 decades of the total instrumentation range. This is a linear range that overlaps with the higher portion of the intermediate range.The system described above provides control room indication and recording of signals proportional to reactor neutron flux during core loading, shutdown, start-up, and power operation, as well as during subsequent refueling. Start-up rate indication for the source and intermediate range channels is provided at the main control board. Reactor trip and rod stop control and alarm signals are transmitted to the Reactor Control and Reactor Protection Systems (RPS) automatic plant control.The engineered safety features instrumentation measures temperatures, pressures, flows and levels in the RCS, Steam and Power Conversion System, Containment, and auxiliary systems Amendment 65 Page 8 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 to actuate and monitor the operation of the engineered safety features equipment. Process variables required on a continuous basis for the start-up, operation, and shutdown of the engineered safety features systems are indicated, recorded and controlled from the Control Room. The quantity and type of process instrumentation ensure safe and orderly operation of all systems and processes over the full operating range of the plant.The Engineered Safety Features Instrumentation System actuates the Safety Injection System, Phase A and Phase B Containment Isolation, CSS, the diesel generators and vital auxiliary systems.In general, the loss of instrumentation power to the sensors, instruments, or logic devices in the engineered safety features instrumentation places that channel in the trip mode. Exceptions which require instrument power for actuation are described in FSAR Section 7.3.2.In order to preclude unsafe conditions for plant equipment or personnel, the Reactor Protection System (RPS) is provided. The RPS consists of sensors, calculators, logic and other equipment necessary to monitor selected nuclear steam supply system (NSSS) conditions and to affect reliable and rapid reactor shutdown (reactor trip) if any or a combination of the monitored conditions approach specified safety system settings. The RPS's functions are to protect the core fuel design limits and reactor coolant pressure boundary for anticipated operational occurrences and also to provide assistance in limiting conditions for certain accidents. The RPS is independent of the Reactor Control System, although the control system is dependent upon some signals derived from the protection system through isolation amplifiers. The Reactor Protection System may be used during a normal plant shutdown by inserting a manual reactor trip after the turbine is off line, if desired.Protection and operational reliability are achieved by providing redundant instrumentation channels for each protective function. These redundant channels are electrically isolated and physically separated. The channel design incorporates separate sensors, separate power supplies, separate rack and panel mounted equipment, and separate logic devices for the actuation of the protective function. For protective functions where two-out-of-three or two-out-of-four coincident actuation is provided, a single channel failure will not impair the protective function.The RPS is designed so that loss of voltage, the most probable mode of failure, in each channel or logic train results in a signal calling for a trip. The protection system design combines redundant sensors and channel independence with coincident trip philosophy so that a safe and reliable system is provided in which a single failure will not violate reactor protection criteria.The design philosophy for Reactor Protection and Reactor Control Systems is to make use, for both protection and control functions, of a variety of measurements. The protection and control systems are separate and identifiable. The RPS continuously monitors system variables by different means, demonstrating protection system diversity. The extent of RPS diversity has been evaluated for a wide variety of postulated accidents. Generally, two or more diverse protective functions would terminate an accident before unacceptable consequences could occur.In the RPS two reactor trip breakers are actuated by two separate logic matrices which interrupt power to the control rod drive mechanism. The breaker main contacts are connected in series Amendment 65 Page 9 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 with the power supply so that opening either breaker interrupts power to all full length control rod drive mechanisms, permitting the rods to free fall into the core.Further details on redundancy are provided through the description of the respective systems covered by the various subsections in this chapter. The power supply for the protection system is discussed in Chapter 8.1.2.2.5 Electrical System The SHNPP turbine generator provides power at nominal 22kV and is directly connected to a main transformer bank which steps up the voltage to the transmission lines, rated 230kV nominal.Offsite power is provided to two half-capacity start-up transformers, at 230kV. These start-up transformers have the capability to supply the required safe shutdown and engineered safety features loads for the Unit. The plant auxiliary power is supplied at 6.9kV, 480V and 120V AC.Redundant sources of onsite power are provided by two diesel generators, either of which is capable of supplying sufficient engineered safety features (ESF) loads to ensure safe shutdown and to maintain the Unit in a safe condition in the event of a complete loss of offsite power.The ESF redundant systems have been electrically and physically designed and segregated so that a single electrical fault or a single credible event will not cause loss of power to both sets of redundant essential electrical components. See Chapter 8 for further details.1.2.2.6 Steam and Power Conversion System The Steam and Power Conversion Systems transform the thermal output of the reactor into electrical power via a turbine-generator. This is accomplished by the transfer of heat from the primary coolant loop to a secondary coolant loop through the steam generators. In the secondary loop, feedwater enters the steam generators and is heated to produce steam. This steam drives the turbine generator, is condensed by rejecting heat to the Circulating Water System, and is recirculated to the Feedwater System. The secondary coolant loop provides an additional barrier to the release of radioactivity to the environment.The SHNPP has a Westinghouse (now Siemens Energy, Inc.) turbine-generator, rated at approximately 1039.4 MWe, which converts the potential energy of the steam into electrical energy. The steam supply path has the necessary flexibility, relief, and isolation valves to ensure integrity and safety.The turbine is a three-element, tandem-compound, four flow-exhaust, 1800 rpm unit with moisture separation and single stage reheat between the high-pressure and low-pressure elements. The AC generator is directly connected to the turbine generator shaft.The Steam Dump System provides the capability to sustain sudden large load decreases up to and including full load loss down to auxiliary loads, concurrent with the loss of external auxiliary power.The Feedwater and Condensate System is a closed system that deaerates the condensate and pumps it from the condenser hotwell through the feedwater heaters to the steam generators. In Amendment 65 Page 10 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 the event of the loss of normal feedwater from any cause the safety related Auxiliary Feedwater System will provide water to the steam generators. The safety related Auxiliary Feedwater System includes two 100 percent capacity motor driven pumps and one 200 percent capacity steam turbine driven pump. See Section 10.4 for additional information.1.2.2.7 Nuclear Fuel Handling and Storage Systems The Fuel Handling and Storage System provides for the safe handling of fuel assemblies and control element assemblies and for the required assembly, disassembly, and storage of the reactor vessel head and internals. The Nuclear Fuel Storage System is designed to store new fuel, and spent fuel produced at the SHNPP, H. B. Robinson Steam Electric Plant, and Brunswick Steam Electric Plant in the fuel pools.The fuel storage system is designed such that the integrity of the fuel is maintained under normal and abnormal conditions. The reactor is refueled with equipment designed to handle spent fuel under water from the time it leaves the reactor vessel until it is placed in a fuel storage pool. The spent fuel is then placed in a cask for shipment from the site.Underwater transfer of spent fuel provides an optically transparent radiation shield, as well as a reliable source of coolant for removal of decay heat.The fuel handling structure may be generally divided in two areas: the refueling cavity which is flooded only during shutdown for refueling; and the spent fuel pools and fuel transfer canal system, which are kept full of water and are always accessible to operating personnel. The refueling cavity and the fuel transfer canal are connected by the fuel transfer tube through which an underwater conveyor transfers the fuel.The manipulator crane, fuel transfer tube, manual tools and spent fuel bridge crane facilitate the transfer of fuel from the refueling cavity to the spent fuel racks.The Spent Fuel Pool Cooling and Cleanup System removes decay heat and impurities from the fuel pools.New fuel assemblies may be stored in any applicable fuel pool. New fuel is delivered to the reactor by lowering it into the appropriate spent fuel pool and taking it through the fuel transfer system. See Section 9.1 for further information.1.2.2.8 Cooling Water and Other Auxiliary Systems 1.2.2.8.1 Circulating and Service Water System The Circulating Water System provides the main condenser with a continuous supply of cooling water for removing the heat rejected by the main turbine. Three 33-1/3 percent capacity circulating water pumps, sized for the maximum heat rejection and the required system head, take suction from the cooling tower basin and deliver water to the condenser inlet waterboxes through two large reinforced concrete pipes. After leaving the condenser, the heated circulating water returns to the cooling tower. In addition to circulating water, service water for cooling of auxiliary equipment in the secondary portions is provided during normal operation from the cooling tower basin by means of service water pumps located in a separate intake pump structure. The service water is returned to the Circulating Water System at the outlet from the Amendment 65 Page 11 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 condenser, for cooling by the Cooling Tower. The Unit has a Service Water System designed to provide redundant cooling water to those components necessary for safety either during normal operation or under accident conditions.1.2.2.8.2 Component Cooling Water System The Component Cooling Water System (CCWS) is an intermediate cooling water system serving components and systems important to the safety of the plant. The CCWS is designed to meet all assigned plant component cooling loads during normal operation, assuming the highest possible service water temperature (95°F). At that temperature, there are no limitations placed on normal plant operation.Component cooling water transfers heat from the various components to the component cooling heat exchangers which are cooled by the Service Water System (SWS). Since the heat is transferred from the component cooling water to the service water, the CCWS serves as an intermediate system between various auxiliary systems and the SWS. This double barrier arrangement reduces the probability of leakage of potentially radioactive effluent into the Service Water System and insures that any leakage of the radioactive fluid from the components being cooled is contained within the plant.Because the CCWS functions continuously during the life of the plant, contamination of this system by the SWS, which is treated with chlorine, is avoided by operating the CCWS at a higher pressure than the SWS.The function of the CCWS of acting as a barrier against leakage of radioactive RCS coolant to the environment is assured by radiation monitors on the CCWS pump inlet lines, water level indication on the surge tank, and by the ability to valve off leaking components. Since the CCWS is one of the Engineered Safety Features and is vital during recovery from an accident, redundancy requirements are included in the system design. In consideration of single failure criteria, the CCWS consists of two separate flow paths for all engineered safeguard functions.See Section 9.2 for a discussion of the CCWS.1.2.2.8.3 Chemical and Volume Control System The Chemical and Volume Control System (CVCS) is designed to (1) adjust the concentration of chemical neutron absorber in the reactor coolant for reactivity control, (2) maintain the proper water inventory in the Reactor Coolant System (RCS), (3) provide the required seal water flow for the reactor coolant pump shaft seals, (4) provide high pressure flow to the Emergency Core Cooling System (ECCS) (this function is described in Section 6.2.), (5) maintain proper concentration of corrosion inhibiting chemicals in the reactor coolant, and (6) reduce the coolant inventory of corrosion products and fission products.During normal operation, this system also provides for introduction of the following chemicals:a) Hydrogen to the volume control tank, b) Nitrogen as required for purging the volume control tank, c) Hydrazine and lithium hydroxide as required via the chemical mixing tank to the suction of the charging pumps.Amendment 65 Page 12 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The Unit has one boric acid tank and auxiliary equipment.In addition to the reactivity control achieved by the rod cluster control assemblies, reactivity control is provided by the CVCS which regulates the concentration of boric acid solution neutron absorber in the RCS. The system is designed to prevent, under system malfunction, uncontrolled or inadvertent reactivity changes which might cause operation in excess of design limits. For further information, see Chapter 9.1.2.2.8.4 Boron Thermal Regeneration System The Boron Thermal Regeneration System accepts borated water letdown from the Reactor Coolant System and returns it, with boron added or deleted as required to accomplish reactor coolant boron concentration changes for load follow. The boration and dilution rates made possible by the system are adequate to handle xenon transients resulting from the design load cycle. See Chapter 9 for further information.1.2.2.8.5 Boron Recycle System The Boron Recycle System receives and processes reactor coolant effluent for reclamation of the boron and purified water.1.2.2.8.6 Residual Heat Removal System The Residual Heat Removal System (RHRS) is designed to remove residual and sensible heat from the core and to reduce the temperature of the Reactor Coolant System during the second phase of Unit cooldown. The RHRS is placed in operation approximately four hours after reactor shutdown depending on reactor coolant temperature. As secondary functions, the RHRS is used for low head safety injection and for transfer of refueling water between the refueling water storage tank and the refueling cavity at the beginning and end of refueling operations. See Chapter 9 for further information.1.2.2.8.7 Primary Sampling System (Nuclear)The Primary Sampling System collects samples of the fluids in the Reactor Coolant System and auxiliary systems for analysis. The system consists of two sampling panels and they are operated from two sampling rooms. Chemical and radiochemical analyses are performed on the samples, and the results are used to regulate boron concentration adjustments, monitor fuel rod integrity, evaluate ion exchanger and filter performance, specify chemical additions to the various systems, and maintain the proper hydrogen overpressure on the volume control tank.For details, see Chapter 9.1.2.2.8.8 Fire Protection System The Fire Protection System provides fire prevention through the control, separation and guarding of sources of ignition; fire limitation by means of fire cutoffs and barriers; fire detection in areas containing safety related equipment or areas of high combustible loading; fire extinguishment by means of installed facilities commensurate with the fire hazard presented.The fire extinguishing function is performed by an automatic sprinkler system, with backup by manual suppression systems. See Chapter 9 for further information.Amendment 65 Page 13 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.2.2.8.9 Compressed Air System The Compressed Air System provides dry, oil-free compressed air to the instrument air and service air systems. This air is used to operate pneumatic instruments, controls, isolation valves and power relief valves. The system also supplies air for normal maintenance work. For details see Chapter 9.1.2.2.8.10 Plant Ventilation System The plant ventilation system provides suitable thermal environment and air quality for personnel comfort, health and safety, and proper equipment operation and integrity. The subsystems that make up the ventilation system are: 1) Control Room Ventilation System, 2) Reactor Auxiliary Building Ventilation System, 3) Containment Ventilation System, 4) Fuel Handling Building Ventilation System, 5) Essential Services Chilled Water System, 6) Waste Processing Building Ventilation System, 7) Turbine Building Ventilation System, and 8) Diesel Generator Building Ventilation System.1.2.2.8.11 Demineralized Water System The Demineralized Water System supplies deaerated, treated, demineralized water for various uses throughout the plant. Among these are reactor makeup storage tank, condensate storage tank, refueling water storage tank, radwaste equipment, radiation elements, and the fuel cask decontamination facility. Demineralized water is also used for all-purpose flushing and cleaning of tools, fixtures, etc. which may have become contaminated. (See Chapter 9 for further information.)1.2.2.9 Waste Processing System The Waste Processing System is designed to collect, monitor and process all liquid, gaseous and solid radioactive wastes originating from the operation of the plant. The principal design objective is to ensure that the release of radioactive material both in the plant and to the environs does not exceed the limits set forth in 10 CFR 20 and Appendix I of 10 CFR 50. In addition, the Waste Processing System is designed to provide for "as low as reasonably achievable (ALARA)," radioactive release to the environment. The Waste Processing System consists of three major subsystems: Liquid Waste Processing System, Gaseous Waste Processing System, and Solid Waste Processing System.Liquid and gaseous waste handling at the plant site is based on achieving the lowest practicable radioactive release to the environment using "state of the art technology". Sources of radioactive gaseous waste are segregated into those which are either processed and discharged or processed and held within the plant. Liquid wastes are segregated into those which are 1) processed and recycled, 2) processed and either recycled or discharged, or 3) collected with the capability of processing before discharge should treatment be required. The concept of segregation is further carried out in waste handling such that each type of waste can be handled in specifically designed subsystems. This design philosophy results in minimizing both the quantity of contaminated effluents to be processed and subsequently released, as well as the residual radioactivity contained in these effluents. The solid waste is handled by the Solid Waste Processing System, which is designed to collect, package, store and ship offsite all the solid radioactive waste resulting from plant operation. See Chapter 11 for further information.Amendment 65 Page 14 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.2.2.10 Radiation Monitoring System The Radiation Monitoring System (RMS) is a group of independent radiation monitors used to measure the levels of radioactivity within various process streams, ventilation ducts, and plant general areas. The monitors are designed to provide information on the radiation levels to the plant Control Room as well as locally at the monitors. The monitors initiate alarm signals when predetermined setpoints are exceeded and initiate various control functions as required when selected alarms are initiated. The RMS continuously records the levels of radiation in the various streams and areas to indicate trends of increasing radiation so as to achieve ALARA radiation releases and personnel exposures. The RMS has been categorized in accordance with 10 CFR 50.69, refer to Section 13.7.1.3 COMPARISON TABLES 1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS Table 1.3.1-1 presents a design comparison of the Shearon Harris Nuclear Power Plant with the Beaver Valley Station and the North Anna Station.These three facilities have the same Nuclear Steam System Supplier (Westinghouse), the same NSSS design (three steam generators), and approximately the same power output. All values presented in the table are characteristic of the first operating cycle for each plant.1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION Table 1.3.2-1 presents a comparison of final and preliminary information for the Shearon Harris Nuclear Power Plant.Table 1.3.2-1 is presented for historical information only.1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 OWNERS Carolina Power & Light Company 1.4.2 APPLICANTS Carolina Power & Light Company and North Carolina Eastern Municipal Power Agency are the applicants for the operating licenses (the North Carolina Eastern Municipal Power Agency ownership was transferred to Duke Energy Progress, Inc. on July 31, 2015). Carolina Power &Light Company (now Duke Energy Progress, Inc.) is responsible for the design, construction, and operation of the Shearon Harris Nuclear Power Plant. Carolina Power & Light Company has been active in the nuclear power field since 1956, when the Company joined with three neighboring utilities to form the Carolinas-Virginia Nuclear Power Association to build and operate a nuclear steam generating plant at Parr, South Carolina. Since then, Carolina Power& Light Company has placed three nuclear units in commercial operation--the H. B. Robinson Plant at Hartsville, South Carolina, in 1971; the Brunswick Steam Electric Plant Unit 2 at Southport, North Carolina, in 1975; and the Brunswick Steam Electric Plant Unit 1 in 1977.Carolina Power & Light Company has engaged the contractors noted below to perform Amendment 65 Page 15 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 engineering and consultant services and provide equipment for Shearon Harris Nuclear Power Plant. Carolina Power & Light Company has participated in the plant design, has reviewed the plant design, and has overall responsibility for SHNPP.1.4.3 ARCHITECT/ENGINEER Ebasco Services Incorporated (Ebasco) is the architect-engineer responsible to Carolina Power& Light Company for the design, engineering, and equipment and material procurement of SHNPP. This includes all plant structures, systems, and components other than those provided by the Nuclear Steam Supply System (NSSS) supplier, except for any promotional or other nonplant-oriented structures which will be done by outside contractors.Ebasco has used the services of:1.4.3.1 S. Seroho Associates (Waste Processing) 1.4.3.2 Franklin Research Institute (Applied Physics) 1.4.3.3 EDS Nuclear (Stress Analysis) 1.4.3.4 Dr. W. C. Pitman (Seismology) 1.4.3.5 Dr. A. L. Odom (Radiometric Age Determinations) 1.4.3.6 Dr. P. C. Ragland (Geochemistry) 1.4.3.7 Dr. S. B. Weed (Clay Mineralogy) 1.4.3.8 Dr. J. B. Butler (Petrography) 1.4.3.9 Dr. D. F. Schutz (Radiometric Age Determinations) 1.4.3.10 Dr. J. de Boer (Paleomagnetic Studies) 1.4.3.11 Dr. G. A. Kiersch (Geologic Report Reviewer) 1.4.3.12 Lehigh University (See Section 3.8) 1.4.4 CONTRACTORS Carolina Power & Light Company has selected the services of Daniel Construction Company to construct the Shearon Harris Nuclear Power Plant. Carolina Power & Light Company is performing ASME nuclear code construction under Carolina Power & Light Company's quality assurance program which was evaluated and accepted by the ASME Survey team. The constructor, Daniel International Corporation, is working under direct supervision and technical control of Carolina Power & Light Company management personnel at the site. The responsibility for construction activities of this nuclear plant is that of Carolina Power & Light Company.Amendment 65 Page 16 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.4.5 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) SUPPLIER The NSSS, along with the design and fabrication of the initial fuel, is supplied by Westinghouse Electric Corporation.1.4.6 CONSULTANTS 1.4.6.1 Dames & Moore This firm has been retained to work in conjunction with Ebasco as consultants for seismic studies for the plant.1.4.6.2 Dr. Jasper L. Stuckey Dr. Stuckey has provided assistance in the area of geology.1.4.6.3 Dr. B. J. Copeland Dr. Copeland provided a report entitled "Ecological Report for CP&L on White Oak Creek Site" which briefly summarized the terrestrial and aquatic ecology of the White Oak basin.1.4.6.4 Dr. Joffre L. Coe, Director, Research Laboratories of Anthropology, University of North Carolina The Research Laboratories of Anthropology performed archaeological (historic and prehistoric) surveys of the site.1.4.6.5 Aquatic Control, Inc.Aquatic Control, Inc. provided consulting services in the fields of aquatic and terrestrial ecology by conducting and reporting on-site baseline surveys of the biota.1.4.6.6 Mr. William Beck, Florida A&M University Mr. Beck provided services for verification of benthic specimen identifications.1.4.6.7 Dr. Samuel Mozley, North Carolina State University Dr. Mozley provided services for verification of benthic specimen identifications.1.4.6.8 Pickard, Lowe & Garrick, Inc.This firm provided assistance as a nuclear consultant.1.4.6.9 Research Triangle Institute The Research Triangle Institute provided consulting services in the fields of meteorology and demography.Amendment 65 Page 17 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.4.6.10 H.A.F.A. International, Inc.This company provided consulting engineering services by reviewing the ASME Code Class 1, 2, and 3 boundary designations and reviewing preservice and inservice inspection requirements for each system, reviewing the pump and valve test program as well as the Type C testing program required by 10 CFR 50 Appendix J.1.4.6.11 TERA Corporation TERA Corporation provided consulting services involving the preparation and review of various subsections of the Radiation Protection section of the SHNPP FSAR.1.4.6.12 Law Engineering Testing Co.Law Engineering conducted geotechnical testing.1.4.6.13 Nello L. Teer Co.Nello L. Teer Co. performed excavation work at plant site.1.4.6.14 Southwest Research Institute This firm has provided review of the SHNPP for access to perform Preservice/Inservice Inspection.1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION Deleted by Amendment No. 48 The remainder of Section 1.5 was deleted by Amendment No. 48 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 lists topical reports which provide information additional to that provided in this FSAR and have been filed separately with the Nuclear Regulatory Commission (NRC) in support of this and similar applications.A legend to the review status code letters follows:A - NRC review complete; NRC acceptance letter issued.AE - NRC accepted as part of the Westinghouse emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.B- Submitted to NRC as background information; not undergoing formal NRC review.O- On file with NRC; older generation report with current validity; not actively under formal NRC review.Amendment 65 Page 18 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 U- Actively under formal NRC review.N- Not applicable; i.e., open literature, etc.Table 1.6-2 lists other reports which provide information additional to that provided in this FSAR and have been filed separately with Nuclear Regulatory Commission (NRC) in support of this application.Table 1.6-3 provides a list of Design Documents which are incorporated by reference as part of the FSAR. These Design Documents are to be used in lieu of the cross-referenced FSAR Figures and Tables which have been removed from the FSAR.Table 1.6-4 lists plant procedures, programs, or manuals which are incorporated by reference into the FSAR.1.7 DRAWINGS AND OTHER DETAILED INFORMATION This section is not applicable to updated FSAR (Reference 1.7.0-1).The remainder of Section 1.7 was deleted by Amendment No. 48

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SECTION 1.7 1.7.0-1 Generic Letter 81-06 issued by Darrell G. Eisenhut regarding Periodic Updating of Final Safety Analysis Reports (FSARs), December 15, 1980.1.8 CONFORMANCE TO NRC REGULATORY GUIDES This section describes the extent to which the SHNPP project complies with all applicable NRC regulatory guides. All regulatory guides which by virtue of the implementation section of the guide itself are applicable to the project, or have been designated as Category 2, 3, or 4 by the NRC's Regulatory Requirements Review Committee have, as a minimum, been addressed.Whenever the requirements of the technical specifications conflict with the requirements of regulatory guides and codes and standards, the requirements of the technical specifications shall govern.Specific applicability of referenced standards (i.e., standards other than the primary one endorsed by the specific regulatory guide) are noted in that portion of this section where the regulatory guide has endorsed it as the primary standard. The extent to which a standard applies to a particular situation will be determined by responsible plant management as evidenced through concurrence/approval of governing procedures or other documents.Regulatory Guide 1.1 NET POSITIVE SUCTION HEAD (NPSH) FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS (REV. 0)The SHNPP Project complies with Regulatory Guide 1.1.FSAR

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Sections 6.2.2, 6.3.2, and 6.5.2.Amendment 65 Page 19 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.2 THERMAL SHOCK TO REACTOR PRESSURE VESSELS (REV. 0)This Regulatory Guide has been withdrawn by the NRC. The following is historical information.Refer to Section 5.3.1.8 for current information regarding Pressurized Thermal Shock.The SHNPP project complies with this guide as described below:Westinghouse follows all the recommendations of Regulatory Guide 1.2. Regulatory Position C.1 is followed by Westinghouse's own analytical and experimental programs as well as by participation in the heavy section steel technology (HSST) programs at Oak Ridge National Laboratory.Analytical techniques have been developed by Westinghouse to perform fracture evaluations of reactor vessels under thermal shock loadings.Under the HSST program a number of six in. thick 39 in. outside diameter steel pressure vessels containing carefully prepared and sharpened surface cracks are being tested. Test conditions include both hydraulic internal pressure loadings and thermal shock loadings. The objective of this program is to validate analytical fracture mechanics techniques and demonstrate quantitatively the margin of safety inherent in reactor pressure vessels.A number of vessels have been tested under hydraulic pressure loadings, and results have confirmed the validity of fracture analysis techniques. The results and implications of the hydraulic pressure tests are summarized in Oak Ridge National Laboratory report ORNL-TM-5090.Three thermal shock experiments have been completed and are now being evaluated.Preliminary information indicates that the analytical techniques do agree favorably with experimental results. Westinghouse is continuing to obtain fracture toughness data for reactor pressure vessel steels through internally funded programs as well as HSST sponsored work.Fracture toughness testing of irradiated compact tension fracture toughness specimens has been completed. The complete post-irradiation data on 0.394 in., 2 in., and 4 in. thick specimens are now available from the HSST program. Both static and dynamic post-irradiation fracture toughness data have been obtained. Evaluation of the data obtained to date on material irradiated to fluences between 2.2 and 4.5 x 1019 n/cm2 indicates that the reference toughness curve as contained in the American Society of Mechanical Engineers (ASME) Code, Section III, remains a conservative lower bound for toughness values for pressure vessel steels.Regulatory Guide 1.3 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT ACCIDENT FOR BOILING WATER REACTORS (REV 2)This guide is not applicable to SHNPP.Amendment 65 Page 20 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.4 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOCA FOR PWR'S (REV. 2)The LOCA dose analysis for the SHNPP project is not based on Regulatory Guide 1.4. The Alternate Source Term methodology is being used following SGR/PUR, therefore the guidance of Regulatory Guide 1.183 (Reference 1.8-16) was followed.FSAR

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Sections 15.0A and 15.6.5.Regulatory Guide 1.5 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS (REV 0)This guide is not applicable to SHNPP.Regulatory Guide 1.6 INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE)POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS (REV. 0)The SHNPP project complies with Regulatory Guide 1.6 as described in FSAR Section 8.3.1.2.3.Regulatory Guide 1.7 CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT (REV. 3)The SHNPP project complies with Regulatory Guide 1.7 with the clarification that the hydrogen analyzers are powered via associated circuits from 1E power sources, classified per Reg Guide 1.97 Rev. 3 as Category 3, and environmentally qualified to function after beyond design-basis accidents using the methodology of IEEE-323-1974.FSAR

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6.2.5, 7.3.1.6.2.e.Regulatory Guide 1.8 QUALIFICATION AND TRAINING OF PERSONNEL FOR NUCLEAR POWER PLANTS (REVISION 4)SHNPP has adopted ANSI/ANS 3.1-2014, which is endorsed by Regulatory Guide 1.8, as the standard to be used for the selection, qualification and training of site personnel.Individuals filling positions who met the requirements of the previous standard(s) or requirements at the time of implementation of this standard and Regulatory Guide can be considered to meet any more restrictive aspects of the requirements of this standard for that position without further review and documentation.Regulatory Guide 1.9 SELECTION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES (REV. 2)The SHNPP project complies with Regulatory Guide 1.9, as presented in FSAR Sections 8.3.1.2.4 and 14.2.12.1.16.Amendment 65 Page 21 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.10 MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES (REV. 1)The SHNPP project complies with Regulatory Guide 1.10 with the following clarification:In regard to regulatory position C.3.(b), the design and as-built drawings for the SHNPP project indicate the relative locations of all cadweld splices that are made either to meet design requirements or as replacements for cadweld repairs or test splices. It is not necessary to dimensionally indicate the exact cadweld splice location on the drawings. The above documents will be maintained for the entire life of the plant.FSAR

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Section 3.8.1.6, Appendix 3.8A.Regulatory Guide 1.11 INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT (3-10-71) AND SUPPLEMENT TO SAFETY GUIDE II, 2-17-72 The SHNPP project meets the intent of Safety Guide 1.11 as follows:There are four instrument lines which penetrate the Containment that are designed to Safety Guide 1.11. The lines are associated with pressure transmitters whose signals can initiate safety injection and containment isolation; and they are the only containment pressure-measuring signals available to initiate containment spray. They do not have automatic isolation valves since these instruments must be operable during a postulated accident to initiate containment spray. These instrument lines are connected to the containment atmosphere by a filled and sealed hydraulic transmission system similar to a sealed pressurizer water reference leg. This arrangement, together with the pressure sensors external to the Containment, forms a double barrier and is otherwise in agreement with Safety Guide 1.11.Should a leak occur outside Containment, the sealed bellows inside Containment, which is designed to withstand full containment design pressure, will prevent the escape of containment atmosphere. Should a leak occur inside Containment, the diaphragm in the transmitter, which is designed to withstand full containment design pressure, will prevent any escape from Containment. This arrangement provides automatic double barrier isolation without operator action and without sacrificing any reliability with regard to its safeguards functions (i.e., no valves to be inadvertently closed). Both the bellows and tubing inside Containment and the transmitter and tubing outside Containment are enclosed by protective shielding. Because of this sealed fluid filled bellows system, a postulated severance of the line during either normal operation or accident conditions will not result in any release from the Containment.FSAR

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Section 6.2.4.Regulatory Guide 1.12 INSTRUMENTATION FOR EARTHQUAKES (REV. 2).The SHNPP project complies with Regulatory Guide 1.12. Interpretations and clarifications are described in FSAR Section 3.7.4. Regulatory Guide 1.12 Revision 2 was issued for use in March 1997. This regulatory guide provides licensees and applicants with new guidance that the staff of the NRC considers acceptable for use in selecting nuclear power plant instrumentation for earthquakes. The Harris Nuclear Plant adopts the guidance provided in Regulatory Guide 1.12 Revision 2 for the seismic instrumentation modifications occurring after April 2021.Amendment 65 Page 22 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FSAR

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3.7.4.Regulatory Guide 1.13 SPENT FUEL STORAGE FACILITY DESIGN BASIS (REV 1)The SHNPP project complies with Regulatory Guide 1.13, based on the understanding that automatic actuation of the Fuel Handling Building HVAC System satisfies the requirement in regulatory position C.7 for automatic actuation of the filtration system and that automatic actuation of the fuel pool purification system is not required.FSAR

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9.1.3., 9.1.4.Regulatory Guide 1.14 REACTOR COOLANT PUMP FLYWHEEL INTEGRITY (REV. 1)The SHNPP project follows the recommendations of this guide with the following exceptions:a) Post-spin inspection - Westinghouse has shown in Reference 1.8-1 that the flywheel would not fail at 290 percent of normal speed for a flywheel flaw of 1.15 inches or less in length. Results for a double-ended guillotine break at the pump discharge with full separation of pipe ends assumed, show the maximum overspeed to be less than 110 percent of normal speed. The maximum overspeed was calculated in Reference 1.8-1 to be about 280 percent of normal speed for the same postulated break, and an assumed instantaneous loss of power to the reactor coolant pump. In comparison with the overspeed presented above, the flywheel is tested at 125 percent of normal speed. Thus, the flywheel could withstand a speed up to 2.3 times greater than the flywheel spin test speed of 125 percent provided that no flaws greater than 1.15 inches are present. If the maximum speed were 125 percent of normal speed or less, the critical flaw size for failure would exceed 6 in. in length.Nondestructive tests and critical dimension examinations are all performed before the spin test. The inspection methods employed (described in Reference 1.8-1) provide assurance that flaws significantly smaller than the critical flaw size of 1.15 in.for 290 percent of normal speed would be detected. Flaws in the flywheel will be recorded in the pre-spin inspection program (see Reference 1.8-1). Flaw growth attributable to the spin test (i.e., from a single reversal of stress, up to speed and back), under the most adverse conditions, is about three orders of magnitude smaller than that which nondestructive inspection techniques are capable of detecting. For these reasons, Westinghouse performs no post-spin inspections and believes that pre-spin test inspections are adequate.b) Interference for stress and excessive deformation - Much of Revision 1 to Regulatory Guide 1.14 deals with stresses in the flywheel resulting from the interference fit between the flywheel and the shaft. Because the Westinghouse design specifies a light interference fit between the flywheel and the shaft, at zero speed, the loop stresses and radial stresses at the flywheel bore are negligible. Centering of the flywheel relative to the shaft is accomplished by means of keys and/or centering devices attached to the shaft, and at normal speed, the flywheel is not in contact with the shaft in the sense intended by Revision 1. Hence, the definition of "Excessive Deformation," as defined in Revision 1 of Regulatory Guide 1.14, is not applicable to the Westinghouse design since the enlargement of the bore and subsequent partial separation of the flywheel from the shaft does not cause unbalance of the flywheel.Amendment 65 Page 23 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Extensive Westinghouse experience with reactor coolant pump flywheels installed in this fashion has verified the adequacy of the design.Westinghouse's position is that combined primary stress levels, as defined in Revision 0 of Safety Guide 14 (Regulatory Positions C.2.a and C.2.c) are both conservative and proven and that no changes to these stress levels are necessary. Westinghouse designs to these stress limits and thus, does not have permanent distortion of the flywheel bore at normal or spin test conditions.c) Discussion B, cross rolling ratio of 1 to 3 - Westinghouse's position is that specification of a cross rolling ratio is unnecessary since past evaluations have shown that ASME SA-533, Grade B, Class 1 materials produced without this requirement have suitable toughness for typical flywheel applications. Proper material selection and specification of minimum material properties in the transverse direction adequately ensure flywheel integrity. An attempt to gain isotropy in the flywheel material by means of cross rolling is unnecessary since adequate margins of safety are provided by both flywheel material selection (ASME SA-533, Grade B, Class 1) and by specifying minimum yield and tensile levels and toughness test values taken in the direction perpendicular to the maximum working direction of the material.d) Regulatory position C.1.a, relative to vacuum-melting and degassing process or the electroslag process - The requirements for vacuum melting and degassing process or the electroslag process are not essential in meeting the balance of the regulatory position nor do they, in themselves, ensure compliance with the overall regulatory position. The initial Safety Guide 14 (10/27/71) stated that the "flywheel material should be produced by a process that minimized flaws in the material and improves its fracture toughness properties." This is accomplished by using ASME SA-533 material including vacuum treatment.e) Regulatory position C.2.b - Westinghouse suggests that this paragraph be reworded as follows in order to remove the ambiguity of reference to an undefined overspeed transient.

 "Design speed should be 125 percent of normal speed or the speed to which the pump motor might be electrically driven by station turbine generator during anticipated transients, whichever is greater. Normal speed is defined as the synchronous speed of the alternating current drive motor at 60 hertz."

f) In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.FSAR

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5.4.1.Regulatory Guide 1.15 TESTING OF REINFORCING BARS FOR CATEGORY I STRUCTURES (REV. 1)The SHNPP project complies with this guide.Amendment 65 Page 24 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FSAR

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Section 3.8.1 Regulatory Guide 1.16 REPORTING OF OPERATING INFORMATION-APPENDIX A TECHNICAL SPECIFICATIONS (REV 4)The SHNPP project complies with this Regulatory Guide except as noted below.In lieu of positions C.1 and C.2 of the Regulatory Guide, those reports indicated in the guide shall be reported as per the plant Technical Specifications.In lieu of position C.1.c of the Regulatory Guide, SHNPP will provide to the NRC the operating data described in Generic Letter 97-02 "Revised Contents of the Monthly Operating Report" via an industry database (e.g., the Consolidated Data Entry (CDE) program) by the end of the month following each calendar quarter.Regulatory Guide 1.17 PROTECTION OF NUCLEAR POWER PLANTS FROM INDUSTRIAL SABOTAGE (REV. 1)SHNPP does not commit to Regulatory Guide 1.17. The SHNPP Security Plan, which has been submitted separately, addresses protection from radiological sabotage per 10 CFR 73.FSAR

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Section 13.6.Regulatory Guide 1.18 STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENT (REV. 1)The SHNPP project complies with this guide with the following exceptions:The SHNPP Containment is a non-prototype reinforced concrete structure as discussed in FSAR Section 3.8.1.7.1. The regulatory positions pertaining to prototype containments are not applicable.FSAR

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Section 3.8.1.7.1.Regulatory Guide 1.19 NONDESTRUCTIVE TESTING OF PRIMARY CONTAINMENT LINER WELDS (REV. 1)The SHNPP project complies with Regulatory Guide 1.19 with the following clarifications and exceptions:a) Regulatory position C.1.b: The SHNPP project uses magnetic-particle or liquid penetrant testing for examining the liner seam welds where radiographic testing is not feasible or where the weld is located in areas which are not accessible after construction. It is felt that liquid penetrant testing, being a surface examination method, is more suited than ultrasonic testing for detection of possible leak prone discontinuities.b) Regulatory position C.1.c: The SHNPP project does not require hourly checking of the soap solution for adequacy of bubble formation properties because this is not a requirement of the ASME Code, Section III, Division 2, Subsection CC.Amendment 65 Page 25 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 c) Regulatory positions 7.a, 7.b, 8.a, and 8.b: The SHNPP project complies with the requirements of the ASME Code, Section III, Division 2, Subsection CC as further discussed in Section 3.8.1 and Appendix 3.8A. No spot radiography was performed prior to April 29, 1977.d) Whenever there is a conflict between this regulatory guide and the ASME Code Section III Division 2, Winter 75 addenda the requirements of the ASME Code with the clarifications shown in Appendix 3.8A will prevail.FSAR

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Section 3.8.1.Regulatory Guide 1.20 COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM FOR REACTOR INTERNALS DURING PREOPERATIONAL AND INITIAL STARTUP TESTING (REV 2)The Westinghouse position regarding this guide is as described in the FSAR Section 3.9.2.4.FSAR

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Section 3.9.2.Regulatory Guide 1.21 MEASURING, EVALUATING AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT WATER COOLED NUCLEAR POWER PLANTS (REV. 1)The SHNPP project complies with this guide.FSAR

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Section 11.5.1.Regulatory Guide 1.22 PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS (REV 0)The SHNPP project complies with this guide as described in Section 7.1.2.5 and 7.3.2.FSAR

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Sections 7.1.2 and 7.3.2.Regulatory Guide 1.23 ONSITE METEOROLOGICAL PROGRAMS (REV. 0)The SHNPP project complies with this guide.FSAR

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Section 2.3.3.Regulatory Guide 1.24 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED WATER REACTOR RADIOACTIVE GAS STORAGE TANK FAILURE (REV 0)The SHNPP project complies with the intent of Regulatory Guide 1.24. Methodology and assumptions differences are:

1) Dispersion factors are determined based on Regulatory Guide 1.145 methodology, as described in FSAR Section 2.3.4.

Amendment 65 Page 26 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1

2) Instead of calculating "whole body doses," Total Effective Dose Equivalent (TEDE) doses were determined during the SGR/PUR effort, as described in Regulatory Guide 1.183 (Reference 1.8-16).

FSAR

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15.7.1 Regulatory Guide 1.25 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS (REV 0)The Fuel Handling Accident dose for the SHNPP project is not based on Regulatory Guide 1.25.The Alternate Source Term Methodology is being used following SGR/PUR, and therefore the guidance from Regulatory Guide 1.183 (Reference 1.8-16) was followed.FSAR

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15.7.4.Regulatory Guide 1.26 QUALITY GROUP CLASSIFICATIONS AND STANDARDS FOR WATER-, STEAM-, AND RADIOACTIVE WASTE CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS (REV. 3)Systems and components are classified Safety-Related in accordance with ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", and ANSI N18.2a-1975, "Revision and Addendum to Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."These categories are equivalent to the Quality Group categories of the Regulatory Guide.FSAR Reference 3.2.2.Regulatory Guide 1.27 ULTIMATE HEAT SINK FOR NUCLEAR POWER PLANTS (REV 2)The SHNPP project complies with this guide with the following clarifications:Regulatory Position C.1.a-Section 2.4.11.1 indicates the assumptions used to calculate the maximum drawdown in the ultimate heat sink water source. Table 2.4.11-3 tabulates the meteorological conditions utilized to compute the drawdown for the Auxiliary Reservoir and for the Main Reservoir. As indicated in Calc SW-0085, there is considerable additional water availability following the calculated periods of drawdown. Carolina Power & Light Company believes that utilization of the meteorological conditions specified in Regulatory Position C.1.a during the calculated periods of drawdown would not jeopardize ultimate heat sink availability.Regulatory Position C.1.b-Section 2.4.11.7 indicates the assumptions used to calculate the maximum service water system (SWS) inlet temperature during the auxiliary reservoir ultimate heat sink operations. The Auxiliary Reservoir is the critical water source for determining this parameter during a critical period. The meteorological conditions that maximize the service water temperature are high solar heating, high ambient air temperature, high relative humidity, and low wind speed. The worst meteorology for one day occurred on June 27, 1952, and the Amendment 65 Page 27 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 worst month occurred between July 18 and August 15, 1949. The average values of these meteorological parameters for 1-day to 30-day periods were calculated based on these days to estimate the maximum service water system inlet temperature. The limiting analysis calculated a maximum pre-accident reservoir temperature of 94.2°F using a composite 10-day period including the worst 9-day consecutive meteorology (7/22/49 - 7/30/49) plus the worst 1-day (6/27/52). The auxiliary Reservoir initial water temperature of 82.2°F is assumed in the analysis based on the July reservoir equilibrium water temperature for the normal meteorological conditions. The analysis also found that a pre-accident reservoir temperature of 94°F would result in a final, 30-day, post-LOCA temperature of 95.33°F, which is just slightly above the ESW design-basis temperature of 95°F. This is acceptable because the reservoir analysis does not account for thermal stratification which would result in a pre-accident temperature below 94°F and a post-accident temperature below 95°F. (Reference Calc SW-0085.)FSAR

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Section 2.4, 9.2.5.Regulatory Guide 1.28 QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION)Conformance with Regulatory Guide 1.28 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3).Regulatory Guide 1.29 SEISMIC DESIGN CLASSIFICATION (REV 3)The SHNPP project complies with this guide with the following clarification: Westinghouse classifies each component important to safety as Safety Class 1, 2, or 3 and these classes are qualified to remain functional in the event of the safe shutdown earthquake, except where exempted by meeting all of the below requirements. Portions of systems required to perform the same safety function as required of a safety class component which is part of that system shall be likewise qualified or granted exemption. Conditions to be met for exemption are:a) Failure would not directly cause an ANS Condition III or IV event (as defined in ANSI N18.2-1973),b) There is no safety function to mitigate, nor could failure prevent mitigation of, the consequences of an ANS Condition III or IV event, c) Failure during or following any ANS Condition II event would result in consequences no more severe than allowed for an ANS Condition III event, and d) Routine post-seismic procedures would disclose loss of the safety function.For items classified as Seismic Category I, only the pertinent requirements of 10 CFR 50 Appendix B apply, lineage traceability of materials is not required and certificates of compliance in lieu of certified material test reports are considered acceptable.FSAR Reference Sections 3.2, 7.6.2.2.e, and 8.3.1.Amendment 65 Page 28 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.30 QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION AND TESTING OF INSTRUMENTATION AND ELECTRIC EQUIPMENT Conformance with Regulatory Guide 1.30 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.31 CONTROL FERRITE CONTENT IN STAINLESS STEEL WELD MATERIAL (REV. 2)The SHNPP meets the intent of this guide as follows:NSSS - The Westinghouse position concerning the control of delta ferrite in stainless steel welding is discussed in Section 5.2.3. The Westinghouse production weld verification program, as described in Reference 1.8-2, was approved as a satisfactory substitute for conformance with the NRC Interim Position on Regulatory Guide 1.31 (April, 1974). The results of the verification program have been summarized and documented in Reference 1.8-3.Balance of Plant - The extent of compliance with this guide described in Section 10.3.6.2 is applicable to all balance of plant austenitic stainless steel components.Field Work - CP&L will comply with revision 3 in lieu of revision 2 of this guide.FSAR

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Sections 5.2.3, and 10.3.6.Regulatory Guide 1.32 CRITERIA FOR SAFETY-RELATED ELECTRIC POWER SYSTEMS FOR NUCLEAR POWER PLANTS (REV. 2)The SHNPP project complies with the design requirements of this guide except as noted in Section 8.3.2.2.1.3 regarding the Class IE DC Power System. Engineered Safety Features System initiation applications and the instrumentation and control power supply system analysis are discussed in Sections 7.3.2.2 and 7.6.2.3, respectively.FSAR

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Sections 7.3.2.2, 7.6.2.3, 8.3.1.2, and 8.3.2.2.Regulatory Guide 1.33 QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION)Conformance with Regulatory Guide 1.33 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.34 CONTROL OF ELECTROSLAG WELD PROPERTIES (REV. 0)The SHNPP project complies with this guide as described below:NSSS - Where electroslag welding is used in fabricating nuclear plant components, the Westinghouse procurement practice requires vendors to follow the recommendations of Regulatory Guide 1.34.Amendment 65 Page 29 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Field Work and Balance of Plant items - This guide is not applicable since electroslag welding will not be performed on low alloy steels and austenitic stainless steels.Regulatory Guide 1.35 INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES (REV 2)This guide is not applicable to the SHNPP project.Regulatory Guide 1.36 NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL (REV 0)The SHNPP project complies with this guide as described below:NSSS The Westinghouse practice meets the recommendations of Regulatory Guide 1.36 and is more stringent in several respects as discussed below.The tests for qualification specified by this Regulatory Guide (ASTM C692-71 or RDT M12-1T) allow use of the tested insulation materials if no more than one of the metallic test samples crack. Westinghouse rejects the tested insulation material if any of the test samples crack.Fiberglass insulation procured after 1977 may be tested in accordance with ASTM C692-77 which endorses a more stringent 4 out of 4 acceptable sample criteria, a position that has been accepted by the NRC (Reference 1.8-17)The Westinghouse procedure is more specific than the procedures suggested by this Regulatory Guide, in that the Westinghouse specification requires determination of leachable chloride and fluoride ions from a sample of the insulating material. The procedures in this Regulatory Guide, ASTM D512 and ASTM D1179, do not differentiate between leachable and unleachable halogen ions.In addition, Westinghouse experience indicates that only one of the three methods allowed under ASTM D512 and ASTM D1179 for chloride and fluoride analysis is sufficiently accurate for reactor applications. This is the "referee" method, which is used by Westinghouse.Balance of Plant All balance of plant thermal insulation meets the recommendations of this guide.Fiberglass insulation procured after 1977 may be tested in accordance with ASTM C692-77 which endorses a more stringent 4 out of 4 acceptable sample criteria, a position that has been accepted by the NRC (Reference 1.8-17).FSAR

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Section 5.2.3.Amendment 65 Page 30 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.37 QUALITY ASSURANCE REQUIREMENTS FOR CLEANING FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.37 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.38 QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.38 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.39 HOUSEKEEPING REQUIREMENTS FOR WATER COOLED NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.39 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.40 QUALIFICATION TESTS FOR CONTINUOUS DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS (REV 0)The SHNPP project will comply with this guide for Class 1 continuous duty motors located inside Containment.FSAR

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Sections 3.10, 3.11, and 8.3.1.Regulatory Guide 1.41 PREOPERATIONAL TESTING OF REDUNDANT ON-SITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS (REV 0)The SHNPP project will comply with this guide.FSAR

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Sections 8.3.1, and 14.2.Regulatory Guide 1.42 INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT-WATER COOLED NUCLEAR POWER REACTORS (REV 1)This guide was withdrawn by the NRC on March 18, 1976.Regulatory Guide 1.43, CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS (REV. 0)The SHNPP project complies with this guide as described below:Amendment 65 Page 31 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 NSSS - Westinghouse practices achieve the same purpose as Regulatory Guide 1.43 by requiring qualification of any high heat input process, such as the submerged-arc wide-strip welding process and the submerged-arc 6-wire process used on ASME SA-508, Class 2, material, with a performance test as described in Regulatory Position 2 of the guide. No qualifications are required by the regulatory guide for ASME SA-533 material and equivalent chemistry for forging grade ASME SA-508, Class 3, material.The fabricator monitors and records the weld parameters to verify agreement with the parameters established by the procedure qualification as stated in Regulatory Position C.3.Field Work and Balance of Plant - Cladding of low alloy steels for safety-related components is not done.Regulatory Guide 1.44 CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL (REV. 0)NSSS - The Westinghouse position on Regulatory Guide 1.44 is discussed in part in Section 5.2.Westinghouse compliance with the separate positions of this regulatory guide are as follows:The use of processing, packaging and shipping controls, and preoperational cleaning to preclude adverse effects of exposure to contaminants on all stainless steel materials are in accordance with Regulatory Position C.1.Austenitic stainless steel materials are utilized in the final heat treated conditions required by the respective ASME Code, Section II, material specification for the particular type or grade of alloy in accordance with Regulatory Position C.2.The Westinghouse position concerning material inspection programs and Regulatory Position C.3 is discussed in Section 5.2.3.4.Westinghouse meets the intent of Regulatory Position C.4 in the manner discussed in detailed in Section 5.2. Exception (b) to Regulatory Position C.4 is covered in the discussion of delta ferrite in Section 5.2.Westinghouse practices are in agreement with Regulatory Position C.5 in the manner discussed in Section 5.2. Exception (a) to Regulatory Postion C.5 is covered in the discussion of delta ferrite in Section 5.2.Westinghouse practices are in agreement with Regulatory Position C.6 in the manner discussed in Section 5.2.Balance of Plant - The extent of compliance described in Section 10.3.6.2 is applicable to all balance of plant austenitic stainless steel components.Field Work - CP&L will comply with this guide.FSAR

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Sections 5.2, and 10.3.6.Amendment 65 Page 32 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.45 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS (REV 0)The SHNPP project complies with the intent of this guide as described in Section 5.2.5, except as described below.FSAR

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Section 5.2.5.Regulatory Guide 1.45 requires that all leakage detector systems be able to respond to a 1 gpm, or its equivalent, leakage increase in 1 hour or less. It also states that when analyzing the sensitivity of leak detection systems using airborne particulate or gaseous radioactivity, a realistic primary coolant radioactivity assumption should be used. Per FSAR Section 5.2.5.3.2, the gaseous and particulate airborne radiation detectors can detect a postulated step increase from 0.1 to 1.0 gpm. This assumes 85% thermal power and RCS activity based on 0.1% failed fuel.Actual RCS activity can be much less than the assumed RCS activity with 0.1% failed fuel.When Regulatory Guide 1.45 was written in 1973, actual RCS activity at most operating plants was high enough to detect a 1 gpm increase in RCS leakage in 1 hour using particulate or gaseous radioactivity detectors. Improvements in fuel performance have greatly decreased actual RCS activities. Actual RCS activities are normally less than the assumed activity in FSAR Section 5.2.5, therefore the radiation detectors may not be able to detect a 1 gpm increase in RCS leakage within 1 hour. The particulate and gaseous radioactivity detectors meet the suggested sensitivities listed in Regulatory Guide 1.45.Regulatory Guide 1.46 PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT (REV 0)The SHNPP project complies with Regulatory Guide 1.46 with exception for the postulation of break points for which the criteria of BTP MEB 3-1 has been adopted. Specific break points are contained in Appendix 3.6A.Our clarification with reference to Regulatory Position are described in the following FSAR Sections:a) Regulatory Position C.1.a through C.1.d - See Section 3.6.2.1.1.2 b) Regulatory Position C.2.a through C.2.d - See Section 3.6.2.1.1.3 c) Regulatory Position C.3.a - See Section 3.6.2.1.5(b) d) Regulatory Position C.3.b - See Section 3.6.2.1.5(a) e) Regulatory Position C.4.c - See Section 3.6.1.3(e)&(f) f) Regulatory Position C.4.d - See Section 3.6.1.2 Amendment 65 Page 33 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.47 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEM (REV. 0)The manner in which SHNPP project meets the intent of this guide is described in Sections 7.3.2.2.13 and 8.3.1.2.10.FSAR

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Sections 7.3.2, 7.6.2.2.m, and 8.3.1.Regulatory Guide 1.48 DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS (REV 0)The SHNPP project meets the intent of this guide as described below:NSSS - Westinghouse supplied components are designed using the stress limits and loading combinations presented in Sections 3.9.1 and 5.2 for ASME Code Class 1 components and in Section 3.9.3 for ASME Code Class 2 and 3 components. The conservatism in these limits and the associated ASME design requirements precludes any component structural failure.The operability of active ASME Code Class 1, 2, and 3 valves and active ASME Code Class 2 and 3 pumps (there are no active ASME Code Class 1 pumps) will be verified by methods detailed in Sections 3.9.1 and 5.2 for ASME Code Class 1 components and in Section 3.9.3 for ASME Code Class 2 and 3 components.The use of the above stated methods provides an acceptable alternate method to meeting the guidance of this Regulatory Guide.Balance of Plant-Regulatory Positions C.6a and C.6b BOP systems will not utilize ASME Code Class 2 and 3 vessels designed to ASME Section VIII, Division 1 except for the nitrogen accumulators that supply the pressurizer PORV's. These accumulators are ASME Section VIII, Division 1, but designed to Section III. They comply with regulatory positions C.6a and C.6b and are discussed in Section 9.3.1-1 and Table 3.2.1-1.Regulatory Position C.7 - This position is not applicable to SHNPP. BOP systems at the SHNPP will not utilize ASME Code Class 2 vessels assigned to Division 2 of Section VIII of the ASME Code.Regulatory Position C.8.a - The allowable stress for ASME Code Class 2 and 3 piping is not exceeded although the loading combinations listed in Table 3.9.3-7 are greater than those required by Regulatory Position C.8.a(1).The emergency loading of Regulatory Position C.8.a(2) is addressed in Table 3.9.3-11.Regulatory Position C.8.b - For the faulted loading combination, Class 2 and 3 piping designed by Ebasco will meet the stress limits provided in Table 3.9.3-11.Regulatory Position C.10.a - The allowable stress for ASME Code Class 2 and 3 pumps is not exceeded although the loading combinations listed in Table 3.9.3-7 are greater than those required by Regulatory Position C.10.a(1), except where the pump bending stresses are insignificant when compared to the membrane stresses. However, in no case will membrane stress exceed 0.75 yield stress under these conditions.Amendment 65 Page 34 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 3.9.3-8 specifies an allowable stress for the emergency plant condition as required by Regulatory Position C.10.a(2). See the response to Regulatory Position C.6.a above.Regulatory position C.10.a(3) - The SHNPP Table 3.9.3-8 meets the guidance of the Regulatory Guide Note 11 (since pump operability will be demonstrated as discussed in FSAR Section 3.9.2), except where the pump bending stresses are insignificant when compared to the membrane stresses. Therefore, for those materials where the allowable stress is limited by yield stress rather than ultimate stress, the primary membrane stress could slightly exceed the yield stress. Under these conditions, the safety function of the pump would not be impaired.Regulatory Positions C.11 and C.12 - Class 2 and 3 system pressure and temperature design conditions are determined for normal, upset, emergency and faulted plant conditions in conjunction with specified seismic events. A valve primary pressure rating is then specified to the manufacturer which is in excess of the limiting system pressure and temperature design conditions. Therefore, the requirements of Regulatory guide 1.48 are met.In addition, allowable valve stress limits for specified plant conditions and seismic loadings are specified to the manufacturer as indicated in FSAR Table 3.9.3-8. These allowable stresses are generally more restrictive than those presently proposed by the ASME Task Group on valves.FSAR

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3.9.1, 3.9.2, 5.2.Regulatory Guide 1.49 POWER LEVELS OF NUCLEAR POWER PLANTS (REV 1)The SHNPP project complies with this guide.Regulatory Guide 1.50 CONTROL OF PREHEAT TEMPERATURE FOR WELDING OF LOW ALLOY STEEL (REV 0)NSSS - Westinghouse considers that this Regulatory Guide applies to ASME Code, Section III, Class 1 components.The Westinghouse practice for ASME Code Class 1 components is in agreement with the recommendations of Regulatory Guide 1.50, except for Regulatory Positions C.1.b and C.2. For ASME Code Class 2 and 3 components, Westinghouse does not apply any of the recommendations of Regulatory Guide 1.50.In the case of Regulatory Position C.1.b, the welding procedures are qualified within the preheat temperature ranges required by Section IX of the ASME Code. Westinghouse experience has shown excellent quality of welds using the ASME qualification procedures.In the case of Regulatory Position C.2, the Westinghouse position documented in Reference 1.8-4 has been found acceptable by the NRC.Balance of Plant - Control of preheat temperatures for welding carbon and low-alloy steels in safety-related components supplied by Ebasco complies with Regulatory Guide 1.50 as follows:

1) When used, low-alloy steels will be pre-heated as required by the Regulatory Guide, unless exceptions noted within the Regulatory Guide apply (e.g., Volumetric Examinations, etc.).

Amendment 65 Page 35 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1

2) Vendor's welding procedure specifications for carbon steels specify the preheat and interpass temperatures to be in accordance with the recommendations for ASME Code, Section III, and Article D-1000.

3) All flux-bearing filler metals are specified to be low-hydrogen type.

4) Vendors are required to store all low-hydrogen electrodes in ovens at 200-300°F for eight hours following their removal from containers and prior to use.

Field Work - CP&L complies with the requirements of Regulatory Guide 1.50 as follows:

1) Welding Procedure Specifications (WPS's) utilized at HNP for welding low alloy steels in Code class 1, 2, or 3 applications have the minimum preheat and maximum interpass temperature specified (Position C.1.a).

2) Welding Procedure Specifications (WPS's) utilized at CP&L for welding low alloy steels in Code class 1, 2, or 3 applications were qualified utilizing the minimum specified temperature as allowed by ASME Section IX (Position C.1.b).

3) CP&L does not maintain the preheat on production welds until a post-weld heat treatment has been performed. However, CP&L does wrap the preheated weld joint after completion of the welding and allows the joint to slow cool. This practice has been satisfactory in the production welding accomplished to date since we have had no welds rejected because of under bead cracking (Position C.2). CP&L performs final inspections of welded joints after completion of any required post-weld heat treatment has been performed.

4) Preheat of production welding is monitored to verify that the limits on preheat and interpass temperatures are maintained (Position C.3).

5) Welding electrodes utilized for welding low alloy steel at CP&L are specified to be low hydrogen type.

Regulatory Guide 1.51 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS (REV 0)This guide was withdrawn by the NRC on July 15, 1975.Regulatory Guide 1.52 DESIGN, TESTING AND MAINTENANCE CRITERIA FOR POST ACCIDENT ENGINEERED-SAFETY-FEATURE ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (REV. 2)The SHNPP project meets the intent of this guide as described in Sections 6.5.1.1 through 6.5.1.4 and Section 9.4.1 with the following exceptions:Regulatory Position Exceptions c.3.d This section requires HEPA filters to be in accordance with MIL-F-51068. MIL-F-51068 has been canceled Amendment 65 Page 36 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 and replaced by ASME AG-1; therefore, HEPA filter requirements will be allowed to either specification.c.4.d This section required that each ESF atmosphere cleanup train should be operated at least 10 hours per month, with the heaters on (if so equipped), in order to reduce the buildup of moisture on the absorbers and HEPA filters. The duration of the monthly operation of each ESF atmosphere cleanup train was changed from requiring 10 continuous hours to 15 continuous minutes with implementation of License Amendment 156, which was based on NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month.Heaters are not required to be on during the monthly operation of each ESF atmosphere cleanup train as a result of implementation of License Amendment 170, which is also based upon TSTF-522.Regulatory Guide 1.53 APPLICATION OF THE SINGLE FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS (REV 0)The SHNPP project meets the intent of this guide as described below:NSSS Westinghouse furnished systems meet the recommendations of this Regulatory Guide as described in Section 7.1.2.7.Balance of Plant (BOP)BOP furnished systems meet the recommendation of this Regulatory Guide as described in Section 7.3.2.2.2.FSAR

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Section 7.6.2.2.b.Regulatory Guide 1.54 QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS (REV. 0)Regulatory Guide 1.54 endorses ANSI N101.4-1972. The SHNPP project complies with the requirements of ANSI N101.4-1972, as it is endorsed by this guide for protective coatings for containment surfaces (steel and concrete) and exposed surfaces of large equipment and pipe.FSAR

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Section 17.3 Regulatory Guide 1.55 CONCRETE PLACEMENT IN CATEGORY I STRUCTURES (REV 0)The SHNPP project complies with Regulatory Guide 1.55 with the following clarification:Amendment 65 Page 37 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Section 3.8.1.2 lists the codes and standards used to establish the standards of construction for the Seismic Category I Containment Building. With the exception of part UW, which pertains to pressure vessels, these same codes and standards are utilized in the construction of all Seismic Category I buildings. In addition to the applicable codes and standards listed in Section 3.8.1.2, the SHNPP specification for concrete incorporates portions of References 2, 4, 5, and 6 of Regulatory Guide 1.55.Embedded piping will be pressure tested in accordance with ANSI B 31.1, NFPA 24, ASME Section III or Branch Technical Position ETSB No. 11-1 R-1, as applicable, rather than ACI-318 part b. The maximum limits set by ACI-318 will not be exceeded, as these standards have more stringent requirements.Regulatory Guide 1.56 MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS (REV 1)This guide is not applicable to the SHNPP project.Regulatory Guide 1.57 DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS (REV 0)Regulatory Guide 1.57 is applicable to the containment penetrations only. Type I penetrations conform to Regulatory Guide 1.57, except that a fatigue analysis will not be performed since all of the piping passing through the penetration is Safety Class 2 and does not require a fatigue analysis.FSAR

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Section 3.8.2 Regulatory Guide 1.58 QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION AND TESTING PERSONNEL Conformance with Regulatory Guide 1.58 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.59, DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS (REV. 2)The SHNPP project complies with this guide.FSAR

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Section 2.4.3.Regulatory Guide 1.60 DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (REV. 1)The design response spectra for Westinghouse supplied equipment complies with this guide with the clarification that the damping values recommended and approved by the NRC in Reference 1.8-5 are used in the dynamic analysis.Amendment 65 Page 38 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The design response spectra for all other structures, systems and components complies with the guide with the exceptions described in Section 3.7.1.1.FSAR

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Section 3.7.1.Regulatory Guide 1.61 DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (REV. 0)The damping values listed in Regulatory Guide 1.61 are acceptable with the single exception of the large piping systems faulted condition value of 3 percent critical. Higher damping values when justified by documented test data have been provided for in Regulatory Position C.2. A conservative value of 4 percent critical has, therefore, been justified by testing for the Westinghouse reactor coolant loop configuration in Reference 1.8-7 and has been approved by the NRC.FSAR

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Section 3.7.1.Regulatory Guide 1.62 MANUAL INITIATION OF PROTECTIVE ACTIONS (REV 0)The SHNPP project complies with this guide as described below:NSSS - There are individual main steam isolation valve momentary control switches (one per loop) mounted on the main control board. Each switch when actuated, will isolate one of the main steam lines. In addition, there are two master switches. Each master switch actuates all three main steam line isolation and bypass valves.Manual initiation of switchover of safety injection from injection to recirculation is in compliance with Section 4.17 of IEEE Standard 279-1971 with the following comment.Manual initiation of either one of two redundant safety injection actuation main control board mounted switches provides for actuation of the components required for reactor protection and mitigation of adverse consequences of the postulated accident, including delayed actuation of sequenced started emergency electrical loads as well as components providing switchover from the safety injection mode to the cold leg recirculation mode following a loss of reactor coolant accident. Therefore, once safety injection is initiated, those components of the ECCS (see Section 6.3) which are realigned as part of the semiautomatic switchover, go to completion on low refueling storage tank water level without any manual action. Manual operation of other components or manual verification of proper position as part of emergency procedures is not precluded nor otherwise in conflict with the above described compliance to Section 4.17 of IEEE Standard 279-1971 of the semiautomatic switchover circuits.No exception to the requirements of IEEE Standard 279-1971 has been taken in the manual initiation circuit of safety injection. Although Section 4.17 of IEEE Standard 279-1971 requires that a single failure within common portions of the protective system shall not defeat the protective action by manual or automatic means, the standard does not specifically preclude the sharing of initiated circuitry logic between automatic and manual functions. It is true that the manual safety injection initiation functions associated with one actuation train (e.g., Train A) shares portions of the automatic initiation circuitry logic of the same logic train; however, a single failure in shared functions does not defeat the protective action of the redundant actuation train (e.g., Train B). A single failure in shared functions does not defeat the protective Amendment 65 Page 39 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 action of the safety function. The sharing of the logic by manual and automatic initiation is consistent with the system level action requirements of the IEEE Standard 279-1971, Section 4.17 and consistent with the minimization of complexity.Balance of Plant - The recommendations of Regulatory Guide 1.62 are complied with by the following design:a) Manual initiation of each protective action at the system level is provided.b) Manual initiation of a system level protective action initiates all required supporting systems.c) Manual initiation switches are located in the Control Room and are readily accessible by the operator.d) No single failure within the manual, automatic or common portions of the protection system can prevent initiation of the protection action by manual or automatic means.e) Manual initiation of protective action depends on the operation of a minimum of equipment.f) Manual initiations at the system level are designed to go to completion once initiated.FSAR

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Sections 7.2.2.2.3, 7.6.2.2.q, and 8.3.1.2.22.Regulatory Guide 1.63 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER - COOLED NUCLEAR POWER PLANTS (REV. 2)The SHNPP project complies with the intent of this guide as described in Section 8.3.1.2.11.Regulatory Guide 1.64 QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.64 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.65 MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS (REV. 0)Westinghouse follows the recommendations of Regulatory Guide 1.65 with the following exceptions:a) The use of modified SA-540, Grade B-24, as specified in the ASME code (Code Case 1605) is permitted by Westinghouse, but is not listed in this Regulatory Guide. Code Case 1605 has been found acceptable to the NRC for application in the construction of components for water-cooled nuclear power plants within the limitations discussed in Regulatory Guide 1.85 which are followed by the Westinghouse practice. The use of Amendment 65 Page 40 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Code Case 1605 for reactor vessel closure stud materials is not precluded by this regulatory guide.b) A maximum ultimate tensile strength of 170,000 psi is not specified by Westinghouse, as recommended by this regulatory guide. Westinghouse does not consider this exception to be a safety issue for the following reasons:The ASME Code requirement for toughness for reactor vessel bolting has precluded the regulatory guide's additional recommendation for tensile strength limitation, since to obtain the required toughness levels, the tensile levels are reduced.Westinghouse has specified both 45 ft.-lb. and 25 mils lateral expansion for control of fracture toughness determined by Charpy-V testing, required by the ASME Code, Section III, Summer 1973 Addenda and 10 CFR 50, Appendix G (Paragraph IV.A.4).These toughness requirements assure optimization of the stud bolt material tempering operation with the accompanying reduction of tensile strength level when compared with previous ASME Code requirements.Prior to 1972, the ASME Code required a 35 ft.-lb. toughness level which provided maximum tensile strength levels ranging from approximately 155 to 178 kpsi (Westinghouse review of limited data - 25 heats).After publication of the Summer 1973 Addenda to the ASME Code and 10 CFR 50, Appendix G, wherein the toughness requirements were modified to 45 ft.-lb. with 25 mils lateral expansion, all bolt material data reviewed on Westinghouse plants showed tensile strengths of less than 170 kpsi.The specification of both impact and maximum tensile strength as stated in the Regulatory Guide results in unnecessary hardship in procurement of material without any additional improvement in quality.The closure stud bolting material is procured to a minimum yield strength of 130,000 psi and a minimum tensile strength of 145,000 psi. This strength level is compatible with the fracture toughness requirements of 10 CFR 50, Appendix G (Paragraph I.C) although higher strength level bolting materials are permitted by the ASME Code.The primary concern of the regulatory position concerning a maximum tensile strength is to minimize the susceptibility of the bolting material to stress corrosion cracking. Stress corrosion has not been observed in reactor vessel closure stud bolting manufactured from material of this strength level. Accelerated stress corrosion test data do exist for materials of 170,000 psi minimum yield strength exposed to marine water environments stressed to 75 percent of the yield strength (given in Reference 2 of this Regulatory Guide). These data are not considered applicable to Westinghouse reactor vessel closure stud bolting because of the specified yield strength differences and a less severe environment; this has been demonstrated by years of satisfactory service experience.Additional protection against the possibility of incurring corrosion effects is assured by:a) Decrease in level of tensile strength compatible with the requirement of fracture toughness as described above.Amendment 65 Page 41 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 b) Design of the reactor vessel studs, nuts, and washers, allowing them to be completely removed during each refueling, permitting visual and/or nondestructive inspection in parallel with refueling operations to assess protection against corrosion, as part of the in-service inspection described in Section 5.2.4. If a stud sticks in place and cannot be easily removed from the vessel flange, the exposed surface above the stud hole is still available for inspection. The risk of thread corrosion is expected to be less on the threads in the hole than above the stud hole due to:

1) The confined area down the stud hole,

2) The presence of thread lubricant on the stud and stud hole threads, and

3) The lack of a mechanism for further concentrating the boric acid in the stud hole during refueling.

c) Design of the reactor vessel studs, nuts, and washers, providing protection against corrosion by allowing them to be completely removed during each refueling and placed in storage racks on the containment operating deck, in accordance with Westinghouse refueling procedures. The stud holes in the reactor vessel flange are sealed with special plugs before removing the reactor closure. In the event a stud sticks in place and cannot be easily removed from the vessel flange, enclosing the stud in a cover or "can" is an alternative means of keeping the stud dry while the cavity is flooded. Thus, the bolting materials and stud holes are not normally exposed to the borated refueling cavity water.FSAR

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Section 5.3.1 PCR-6575 Regulatory Guide 1.66 NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS (REV 0)This guide was withdrawn by the NRC on September 28, 1977.Regulatory Guide 1.67 INSTALLATION OF OVERPRESSURE PROTECTION DEVICES (10-73)The SHNPP project complies with this guide as discussed in Section 3.9.3.3.Regulatory Guide 1.68 INITIAL TEST PROGRAMS FOR WATER-COOLED NUCLEAR POWER PLANTS (REV 2)The SHNPP project will comply with this guide as described in Section 14.2.Regulatory Guide 1.68.1 PREOPERATIONAL AND INITIAL STARTUP TESTING OF FEEDWATER AND CONDENSATE SYSTEMS FOR BOILING WATER REACTOR POWER PLANTS (REV 1)This guide is not applicable to the SHNPP project.Amendment 65 Page 42 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.68.2 INITIAL STARTUP TEST PROGRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILITY FOR WATER-COOLED NUCLEAR POWER PLANTS (REV 1)The SHNPP project will comply with this guide as described in Section 14.2.Regulatory Guide 1.68.3 PREOPERATIONAL TESTING OF INSTRUMENT AND CONTROL AIR SYSTEMS (REV 0)The SHNPP project will comply with this guide as described in Section 14.2.12 except for Regulatory Position C.10 and C.11. Regulatory Position C.10 does not apply to SHNPP because the instrument air system does not contain any single large loads that would cause a significant perturbation on the normal instrument air pressure. Therefore, a test to verify the instrument air system's response to the conditions postulated in C.10 will not be conducted.Regulatory Position C.11 does not apply to SHNPP because of the design of the Instrument Air System. To overpressurize the system would require three failures; therefore, overpressurization is not a credible failure and applicable testing will not be done.Regulatory Guide 1.69 CONCRETE RADIATION SHIELD FOR NUCLEAR POWER PLANTS (REV. 0)The SHNPP project complies with this guide with the exceptions and clarifications described in Section 12.3.2.4.Regulatory Guide 1.70 STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS (REV. 3)The SHNPP project complies with this guide except where it conflicts with the guidance of General Letter 81-06 for conformance with 10 CFR 50.71(e). In this case, the guidance in Generic Letter 81-06 will be followed.Regulatory Guide 1.71 WELDER QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY (REV 0)NSSS - Shop Welding - Westinghouse practice does not require qualification or requalification of welders for areas of limited accessibility as described by Regulatory Guide 1.71. Experience shows that the current Westinghouse shop practice produces high quality welds. In addition, the performance of required nondestructive evaluations provides further assurance of acceptable weld quality.Westinghouse believes that limited accessibility qualification or requalification, which is in excess of ASME Code, Section III and IX requirements, is an unduly restrictive requirement for component fabrication, where the welders' physical position relative to the welds is controlled and does not present any significant problems. In addition, shop welds of limited accessibility are repetitive due to multiple productions of similar components, and such welding is closely supervised.Balance of Plant - Shop Welding - The SHNPP project complies with Regulatory Guide 1.71, with the following clarifications and exceptions:Amendment 65 Page 43 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Carolina Power & Light Company does not plan to comply with the Regulatory Positions for procured components: "C.1. The performance qualification should require testing of the welder under simulated access conditions when physical conditions restrict the welder's access to a production weld to less than 30 to 35 cm (12 to 14 in.) in any direction from the point." "C.2 Requalification is required: (a) when significantly different restricted accessibility conditions occur, . . ." Limited accessibility qualification or requalification, which exceeds ASME Section III and IX requirements, is considered an unduly restrictive requirement for shop fabrication, where the welder's physical position relative to the welds is controlled and does not present any significant problems. In addition, shop welds of limited accessibility are repetitive due to multiple production of similar components, and such welding is closely supervised. Also, critical shop welds are nondestructively tested in accordance with ASME Section III requirements which will reveal any defective welds. The adequacy of welds is determined by the nondestructive testing of welds which is in accordance with requirements of the ASME B&PV Code. If a welder is assigned to a job which he cannot perform satisfactorily, the test results of the weld joint will not be acceptable. There is no compromise in safety in such matters.NSSS & BOP - Field Welding - The SHNPP project complies with Regulatory Guide 1.71, with the following clarifications and exceptions:For field application, the type of qualification should be considered on a case-by-case basis due to the great variety of circ*mstances encountered.When a full penetration field weld joint in a reactor coolant pressure boundary system, Code Class 1, 2, and 3 has the accessibility restrictions outlined by Regulatory Guide 1.71 and does not require final volumetric NDE, only welders previously qualified for manual welding with a restricted accessibility welding test shall perform the welding. For instrument lines this is only applicable to larger than 3/4 in. nominal pipe size including the weld joint of the first isolation valve inside the Containment. Automatic welding shall be considered exempt from Regulatory Guide 1.71 since limitations are inherent to the welding process for inaccessibility of locating the welding apparatus within the confinements of the weld joint. Special qualification for limited accessibility as defined in Regulatory Guide 1.71 and this paragraph shall be required for a welder who is physically limited in his ability to perform field welding of a joint because of barriers (walls, hangers, adjacent piping equipment, etc.) as close as 12 in. or less and encroaching into the envelope of space normally used by the welder when making a similar weld without such restrictions. The necessity for use of a mirror by the welder would be considered as a requirement for special welder qualification. Requalification by a different special welder qualification will be required when significantly different access condition occurs other than those originally required by the previous special welder qualification (example, addition of a mirror, significant physical changes in welders access, etc.).Regulatory Guide 1.72 SPRAY POND PIPING MADE FROM FIBERGLASS-REINFORCED THERMOSETTING RESIN (REV 2)This guide is not applicable to the SHNPP project.Regulatory Guide 1.73 QUALIFICATION TESTS OF ELECTRIC VALVE OPERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS (REV 0)The SHNPP project complies with this guide as described below:Amendment 65 Page 44 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 NSSS - For safety-related motor operated valves located inside Containment, environmental qualification is performed in accordance with IEEE Standard 382-1972. Auxiliary safety-related equipment (e.g., stem mounted limit switches) is qualified separately. Qualification conditions (temperature, pressure, radiation and chemistry) are those specified in Part III of IEEE Standard 382-1972 for pressurized water reactor applications. Since there are no exposed organic materials, consideration of beta radiation is not required.BOP - Electric valve operators installed inside the Containment for the SHNPP project comply with this guide.FSAR

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Sections 3.10, 3.11, 8.3.1.Regulatory Guide 1.74 QUALITY ASSURANCE TERMS AND DEFINITIONS Conformance with Regulatory Guide 1.74 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.75 PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS (REV. 1)The SHNPP project meets the intent of this guide as described in Sections 8.3.1.2.14, 8.3.1.4, 7.1.2.2.1, and 7.6.2.2.f.Regulatory Guide 1.76 DESIGN BASIS TORNADO FOR NUCLEAR POWER PLANTS (REV.0)The SHNPP project complies with this guide (with the exception described below).Regulatory Guide 1.76 Revision 1 was issued for use in March 2007. This regulatory guide provides licensees and applicants with new guidance that the staff of the NRC considers acceptable for use in selecting the design-basis tornado and design-basis tornado-generated missiles that a nuclear plant should be designed to withstand. This guidance divides the United States into three regions: the Harris Nuclear Plant is located in Region 1. The NRC staff accepts the methods described in Regulatory Guide 1.76 Revision 1 to evaluate submittals from operating reactor licensees after March 2007 who voluntarily propose to initiate system modifications that have a clear nexus with the guidance provided. No backfitting is intended or approved in conjunction with its issuance. The Harris Nuclear Plant adopts the guidance provided in Regulatory Guide 1.76 Revision 1 as an optional design basis for new system modifications occurring after March 2007.FSAR

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Section 3.3.2.1.Regulatory Guide 1.77 ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD EJECTION ACCIDENT FOR PRESSURE AND WATER REACTORS (REV. 0)The SHNPP project complies with the guide with the exception described below:Westinghouse methods and criteria are documented in Reference 1.8-8 which has been reviewed and accepted by the NRC.Amendment 65 Page 45 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The results of the Westinghouse analyses show agreement with Regulatory Positions C.1 and C.3. In addition, Westinghouse utilizes the assumptions given in Appendices A and B of the Regulatory Guide. However, Westinghouse takes exception to Regulatory Position C.2 which implies that the rod ejection accident should be considered as an emergency condition.Westinghouse considers this a faulted condition as stated in ANSI N18.2. Faulted condition stress limits will be applied for this accident.FSAR

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Section 15.4.8.Regulatory Guide 1.78 EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTULATED HAZARDOUS CHEMICAL RELEASE (REV 1)The SHNPP project complies with this guide.FSAR

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Section 6.4 and Section 2.2.3.3.Regulatory Guide 1.79 PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS (REV 1)The SHNPP project will comply with this Regulatory Guide with the exception as noted in Section 14.2.7.Regulatory Guide 1.80 PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS (REV 0)Regulatory Guide 1.80 is superseded by Regulatory Guide 1.68.3.Regulatory Guide 1.81 SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTI-UNIT NUCLEAR POWER PLANTS (REV. 1)Regulatory Guide 1.81 is not applicable to the SHNPP.Regulatory Guide 1.82 WATER SOURCES FOR LONG-TERM RECIRCULATION COOLING FOLLOWING A LOSS-OF-COOLANT ACCIDENT (REV. 3)The SHNPP project satisfies the intent of this guide as described in the referenced FSAR Sections.FSAR

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Sections 6.2.2.2, and 6.5.2.Regulatory Guide 1.83 INSERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR TUBES (REV 1)The SHNPP project complies with the recommendations of this guide. Westinghouse steam generators are designed to permit access to tubes for inspection and/or plugging. The inservice inspection program is discussed in the Technical Specifications.Amendment 65 Page 46 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.84 CODE CASE ACCEPTABILITY-ASME III DESIGN AND FABRICATION The SHNPP project complies with this guide as described below:NSSS a) Westinghouse controls its suppliers to:

1) Limit the use of code cases to those listed in Regulatory Position C.1 of the Regulatory Guide 1.84 revision in effect at the time the equipment is ordered, except as allowed in item b) below.

2) Identify and request permission for use of any code cases as listed in Regulatory Position C.1 of the Regulatory Guide 1.84 revision in effect at the time the equipment is ordered, where use of such code cases is needed by the supplier.

3) Permit continued use of a code case considered acceptable at the time of equipment order, where such code case was subsequently annulled or amended.

b) Westinghouse seeks NRC permission for the use of Class 1 code cases needed by suppliers and not yet endorsed in Regulatory Position C.1 of the Regulatory Guide 1.84 revision in effect at the time the equipment is ordered and permits supplier use only if NRC permission is obtained or is otherwise assured (e.g., a later version of the Regulatory Guide includes endorsem*nt).Balance of Plant The SHNPP project complies with this guide.Regulatory Guide 1.85 CODE CASE ACCEPTABILITY-ASME III MATERIALS The SHNPP project complies with this guide as described below:NSSS a) Westinghouse controls its suppliers to:

1) Limit the use of code cases to those listed in Regulatory Position C.1 of the Regulatory Guide 1.85 revision in effect at the time the equipment is ordered, except as allowed in item b) below.

2) Identify and request permission for use of any code cases as listed in Regulatory Position C.1 of the Regulatory Guide 1.85 revision in effect at the time the equipment is ordered, where use of such code cases is needed by the supplier.

3) Permit continued use of a code case considered acceptable at the time of equipment order, where such code case was subsequently annulled or amended.

Amendment 65 Page 47 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 b) Westinghouse seeks NRC permission for the use of Class 1 code cases needed by suppliers and not yet endorsed in Regulatory Position C.1 of the Regulatory Guide 1.85 revision in effect at the time the equipment is ordered and permits supplier use only if NRC permission is obtained or is otherwise assured (e.g., a later version of the Regulatory Guide includes endorsem*nt).Balance of Plant The SHNPP project complies with this guide.Regulatory Guide 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS (REV 0)The SHNPP project will comply with this guide.Regulatory Guide 1.87 GUIDANCE FOR CONSTRUCTION OF CLASS 1 COMPONENTS IN ELEVATED-TEMPERATURE REACTORS (SUPPLEMENT TO ASME SECTION III CODE CASES 1592, 1593, 1594, 1595 AND 1596) (REV 1)This guide is not applicable to the SHNPP project.Regulatory Guide 1.88 COLLECTION, STORAGE AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS Conformance with Regulatory Guide 1.88 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.89 QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS (REV. 0)The SHNPP project meets the intent of the guide as described below:NSSS - The Westinghouse approach to satisfying the guidelines of Regulatory Guide 1.89 and IEEE Standard 323-1974 is documented in Reference 1.8-9. The Westinghouse approach to satisfying the guidelines of IEEE Standard 323-1971 is documented in WCAP-7410-L, WCAP-7709-L, and Westinghouse Supplemental Environmental Qualification Testing Program (Re:Westinghouse letter NS-CE-692, Eicheldinger to Vassallo, July 10, 1975, and NRC letter Vassallo to Eicheldinger, November 9, 1975).Balance of Plant - Class 1E electrical equipment supplied by Ebasco is qualified by the equipment manufacturer by utilizing analysis of type testing. Such qualifications incorporate the methods prescribed in IEEE-323-74, IEEE-334-71, and IEEE-317-76. Purchase specifications indicate that type testing is the preferred method of qualification. Ebasco supplied valve operators are qualified in accordance with IEEE-382-72 and Regulatory Guide 1.73.The Emergency Diesel Generating System was qualified according to the requirements of Regulatory Guide 1.89 and IEEE Standard 323-1974. Replacement parts for the Emergency Diesel Generating System may be procured to the requirements of IEEE Standard 323-1983.Amendment 65 Page 48 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FSAR

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Sections 3.10 and 3.11.Regulatory Guide 1.90 INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES WITH GROUTED TENDONS (REV 1)This guide is not applicable to the SHNPP project.Regulatory Guide 1.91 EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANTS (REV 1)The SHNPP project meets the intent of this guide as described in FSAR Section 2.2.3.Regulatory Guide 1.92 COMBINING MODAL RESPONSES AND SPATIAL COMPONENTS IN SEISMIC RESPONSE ANALYSIS (REV 1)The SHNPP project meets the intent of this guide as described below:NSSS The Westinghouse procedure for combining modal response is presented in Section 3.7.2.7B Balance of Plant The BOP design of the SHNPP meets the intent of this Regulatory Guide as described in FSAR Sections 3.7.2.1 and 3.7.3.7.FSAR

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Appendix 3.9A Regulatory Guide 1.93 AVAILABILITY OF ELECTRIC POWER SOURCES (REV 0)The SHNPP project will comply with this guide with the following clarification: Section C of the guide permits electric power sources to be removed from service for corrective maintenance activities only. As permitted by Technical Specifications, electric power sources may be removed from service for other reasons under administrative controls provided these are restored to service within the allowed out of service time intervals.FSAR

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Technical Specifications Subsection 3/4.8.Regulatory Guide 1.94 QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.94 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR in Chapter 17 (see Section 17.3)Amendment 65 Page 49 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.97 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT (REV 3)The guidance of Regulatory Guide 1.97 (Revision 3) has been implemented along with that of NUREG 0737 Supplement 1, Section 6.2, in establishing the guidelines for variables to be monitored.The compliance with these documents is detailed in a response to NRC ICSB Question 44 via CP&L letter LAP-83-405 dated September 6, 1983, CP&L letter NLS-85-109 dated June 3, 1985, and CP&L letter NLS-88-279 dated December 22, 1988.Regulatory Guide 1.98 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A RADIOACTIVE OFFGAS SYSTEM FAILURE IN A BOILING WATER REACTOR (REV 0)This guide is not applicable to the SHNPP project.Regulatory Guide 1.99 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS (REV. 2)The SHNPP project complies with the intent of Regulatory Guide 1.99 with the following clarification.The results of the analysis of the third reactor vessel surveillance capsule indicated that the scatter in the shift of reference temperature - nil ductility transition ( ) for two of the six plate surveillance specimens exceeded the criteria for "credibility" as defined by the regulatory guide. The weld surveillance data meets the credibility requirements of the regulatory guide.When surveillance data is "credible", i.e., without excessive scatter, the regulatory guide permits calculation of chemistry factors from the surveillance data, and halving of required margins applicable to reactor vessel pressure-temperature limits. Otherwise, chemistry factors are based on generic data from the regulatory guide, and full margins are to be used. SHNPP still uses the plate surveillance data to calculate chemistry factors, but has assumed the full margin terms in developing reactor vessel pressure-temperature limits.FSAR

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Sections 5.3.1, 5.3.2 Regulatory Guide 1.100 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS (REV 1)The SHNPP project complies with this guide as described below:The Westinghouse program for seismic qualification of safety-related electrical equipment to Regulatory Guide 1.100 is delineated in the latest revision of WCAP-8587 "Methodology for Qualifying Westinghouse PWR-SD Supplied NSSS Safety-Related Electrical Equipment,"together with Supplement 1 to this report. In summary, seismic qualification for all Westinghouse supplied equipment will be demonstrated by the following methods:a) For equipment not subject to high energy line break conditions which has been previously qualified, as identified in Supplement 1 to WCAP-8587, using the methods Amendment 65 Page 50 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 permitted by the 1971 version of IEEE Standard 344 (i.e., single axis sine-beat testing or analysis, after demonstration of no resonant frequency below 33Hz), no additional seismic qualification will be specified provided that:

1) It can be shown, by separate component testing and/or analysis, that there are no aging mechanisms that could prejudice the previously completed seismic qualification.

2) Any design modifications made to the equipment do not significantly affect the seismic characteristics of the equipment.

3) The adequacy of the original seismic test levels can be demonstrated as conservative by plant specific verification.

b) For new equipment, or equipment that cannot meet the provisions of a) above, seismic qualification will be performed in accordance with IEEE Standard 344-75. The method to be employed (i.e., test and/or analysis) is indicated, for the safety-related equipment in the Westinghouse PWRSD Scope of Supply, in Supplement 1 to WCAP-8587. Where multifrequency biaxial inputs are employed for testing, the methodology described in WCAP-8695, "General Method of Developing Multifrequency Biaxial Test Inputs for Bistables," will be employed. When flexible equipment size and weight precludes biaxial testing, single axis testing with justification will be utilized to meet IEEE Standard 344-1975. For rigid equipment (i.e., no resonant frequency below 33Hz), qualification may be by analysis in accordance with IEEE Standard 344-1975.All non-NSSS supplied equipment and supports are in compliance with the qualification and documentation of IEEE 344-75.Regulatory Guide 1.101 EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS (REV 1)SHNPP does not comply with this Regulatory Guide.Regulatory Guide 1.102 FLOOD PROTECTION FOR NUCLEAR POWER PLANTS (REV 1)Shearon Harris Nuclear Power Plant complies with the Regulatory Guide 1.102 Rev. 1 with the following exceptions.a) The plant grade at Elevation 260 ft. is higher than the maximum water level in the Auxiliary Reservoir or in the Main Reservoir as discussed in Sections 2.4.5 and 3.4.1.The main access road on the east side of the plant, which is constructed on an embankment through the Main Reservoir, is not classified as safety-related.b) The east and south slopes of the plant site will be subject to water waves in the Main Reservoir. These slopes have been protected by sacrificial spoil fill as discussed in Section 2.4.3.c) As discussed in Section 2.4.13, the groundwater level at the plant site will not exceed Elevation 251 ft. and the plant structures have been designed for the hydrostatic and the buoyant forces corresponding to this groundwater level.Amendment 65 Page 51 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Ponding of storm water in the plant yard, as discussed in Section 2.4.2, due to the probable maximum precipitation (PMP) is not expected to change the groundwater table and is also not expected to cause any additional hydrostatic and buoyant forces on the structures because of the selected impervious backfill placed around the structures. For conservatism the stability of the structures against floatation, as discussed in Section 3.8.4, has been checked for water at the plant grade (Elevation 260 ft.).d) The walls of safety-related structures, though not designed for the static and dynamic forces due to the ponded water in the yard, have adequate reserve strength to resist these forces. For the reason stated in c) above the walls of the plant structures have not been designed for static and dynamic forces due to the ponded water in the yard.e) The roofs of safety-related structures have not been designed for standing water on the roofs up to the top of curbs; however the water will be drained off through roof drains.Regulatory Guide 1.103 POST-TENSIONED PRESTRESSING SYSTEMS FOR CONCRETE REACTOR VESSELS AND CONTAINMENTS (REV 1)This guide is not applicable to the SHNPP project.Regulatory Guide 1.104 OVERHEAD CRANE HANDLING SYSTEMS FOR NUCLEAR POWER PLANTS (REV 0)This guide was withdrawn by the NRC on August 22, 1979.Regulatory Guide 1.105 INSTRUMENT SETPOINTS (REV 1)The SHNPP project meets the intent of this guide as described below:NSSS Technical Specifications provide the margin from the nominal setpoint to the technical specification limit to account for drift when measured at the rack during periodic testing. The allowances between the technical specification limit and the safety limit include a statistical combination of the following items: a) the inaccuracy of the instrument, 2) process measurement accuracy, 3) uncertainties in the calibration, 4) the potential transient overshoot determined in the accident analyses (this may include compensation for the dynamic effect),and 5) environmental effects on equipment accuracy caused by postulated or limiting postulated events (only for those systems required to mitigate consequences of an accident).Westinghouse designers choose setpoints such that the accuracy of the instrument is adequate to meet the assumptions of the safety analysis.The range of instruments is chosen based on the span necessary for the instrument's function.Narrow range instruments are used where necessary. Instruments are selected based on expected environmental and accident conditions. The need for qualification testing is evaluated and justified on a case-by-case basis.Administrative procedures coupled with the present cabinet alarms and/or locks provide sufficient control over the setpoint adjustment mechanism such that no integral setpoint securing device is required. Integral setpoint locking devices are supplied.Amendment 65 Page 52 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The assumptions used in selecting the setpoint values in Regulatory Position C.1 and the minimum margin with respect to the technical specification limit and calibration uncertainty is documented. Drift rates and their relationship to testing intervals is not documented.Balance of Plant Discussion of compliance is included below:a) Setpoints are established with margins between technical specification limits for the process variable and the nominal trip setpoint which include allowance for instrument inaccuracy, calibration uncertainty, instrument drift anticipated between calibration intervals.b) All instrument ranges are selected to ensure that the portion required for the setpoint is within the portion of the instrument ranges that yields the maximum accuracy and maintainability.c) The ranges selected for the instrumentation fully encompass the expected operating range of the monitored variables and the selected range is always well within the saturation limits of the instrument.d) The accuracy of all setpoints is equal to or better than the accuracy assumed in the safety analysis. Instrument intervals are chosen for the design conditions in which they are installed in order not to anneal, stress relieve, or work harden to the extent that they will not maintain the required accuracy. Design verification is included as part of the equipment qualification program as recommended in Regulatory Guide 1.89.e) Instruments important to safety have securing devices on the setpoint adjustment mechanism and/or are under administrative control. The securing device is designed such that during securing or releasing it will not alter the setpoint.f) Documentation of methodology and assumptions used in selecting setpoint values and minimum margins, drift rates and test intervals is contained in plant setpoint determination procedures/documents.g) Safety related setpoints not covered by technical specification have sufficient documentation to support the setpoint value, tolerance, and margin to system process limits.Regulatory Guide 1.106 THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR-OPERATED VALVES (REV. 1)The SHNPP project meets the intent of this guide as described in Section 8.3.1.2.18.Regulatory Guide 1.107 QUALIFICATIONS FOR CEMENT GROUTING FOR PRESTRESSING TENDONS IN CONTAINMENT STRUCTURES (REV 1)This guide is not applicable to the SHNPP project.Amendment 65 Page 53 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.108 PERIODIC TESTING OF DIESEL GENERATORS USED AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS (REV. 1)Preoperational and periodic testing of the SHNPP diesel generators will comply with this guide as described in the referenced FSAR sections except as discussed below:FSAR

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Section 8.3.1.1.2.14, 14.2.12.1.16, and 3.1.14.RG 1.108, paragraph C.2.a(5) requires that design accident loading sequence be performed immediately after the 24-hour load run. The requirements of this paragraph will not be fulfilled immediately after the 24-hour load run. Instead, this test will be performed in conjunction with the Integrated Engineered Safety Features Actuation System Test. The Diesel Engine will be operated at full load conditions to reestablish full load temperature conditions. A loss of all A.C.voltage will then be simulated to demonstrate that the diesel generator unit can start automatically and attain required voltage and frequency. Also, proper operation for the design-accident-loading-sequence to design-load requirements while maintaining voltage and frequency within limits will be demonstrated as required. This will provide for accomplishment of 24-hour full load carrying capability demonstration as soon as the Emergency Diesel Generator Systems are ready.RG 1.108, paragraph C.2.d defines the periodic test interval for the emergency diesel generators. The requirements of this section will be fulfilled, as required by FSAR Section 3.1.14, by performing the periodic tests in accordance with the schedule provided in Technical Specification Section 3/4.8.1.Regulatory Guide 1.109 CALCULATIONS OF ANNUAL DOSES TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENTS FOR THE PURPOSE OF EVALUATING COMPLIANCE WITH 10 CFR 50, APPENDIX I (REV. 1)The SHNPP project complies with this guide.FSAR

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Section 11.2.3, 11.3.3.Regulatory Guide 1.110 COST-BENEFIT ANALYSIS FOR RADWASTE SYSTEMS FOR LIGHT-WATER-COOLED NUCLEAR POWER REACTORS (REV 0)The SHNPP project is not required to address this guide.Regulatory Guide 1.111 METHODS FOR ESTIMATING ATMOSPHERIC TRANSPORT AND DISPERSION OF GASEOUS EFFLUENTS IN ROUTINE RELEASES FROM LIGHT-WATER-COOLED REACTORS (REV 1)The SHNPP project complies with this guide.FSAR Reference Section 2.3.4.Amendment 65 Page 54 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.112 CALCULATIONS OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM LIGHT WATER COOLED REACTORS (REV. 0)The SHNPP project complies with this guide.FSAR

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Section 11.2.3, 11.3.3.Regulatory Guide 1.113 ESTIMATING AQUATIC DISPERSION OF EFFLUENTS FROM ACCIDENTAL AND ROUTINE REACTOR RELEASES FOR THE PURPOSE OF IMPLEMENTING APPENDIX I (REV. 1)The SHNPP project complies with this guide and is discussed in the SHNPP Operating License Environmental Report, Section 5.2.2.1.Regulatory Guide 1.114 GUIDANCE ON BEING OPERATOR AT THE CONTROLS OF A NUCLEAR POWER PLANT (REV. 1)The SHNPP project complies with this guide.FSAR

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Chapters 13 and 14.Regulatory Guide 1.115 PROTECTION AGAINST LOW-TRAJECTORY TURBINE MISSILES (REV 1)The SHNPP project complies with the intent of this guide in that due to the limited exposure of vital equipment and the high degree of barrier protection provided, and as stated in the NRC SER Supplement No. 3, dated July 1977, low trajectory turbine missiles will not be a significant threat to the plant.FSAR

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Section 3.5.1 Regulatory Guide 1.116 QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS Conformance with Regulatory Guide 1.116 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.117 TORNADO DESIGN CLASSIFICATION (REV. 1)The SHNPP project complies with this guide except for conditions deemed acceptable using TMRE methodology that was approved by the NRC in License Amendment No. 169 [ADAMS Accession No. ML18347A385]. TMRE is an alternative methodology for determining whether protection from tornado-generated missiles is required. The methodology can only be applied to discovered conditions where tornado missile protection was not provided, and cannot be used to avoid providing tornado missile protection in the plant modification process.FSAR

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Sections 3.3 and 3.5.1.4.Amendment 65 Page 55 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.118 PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS (REV 2)Except as noted below, SHNPP complies with IEEE Standard 338-1977, Reg. Guide 1.118.With these exceptions, the project still meets IEEE 338-1971 as discussed in Section 7.1.2.18 and IEEE 338-1975 as discussed in Section 8.3.1.2.27.NSSS Westinghouse will make clear distinction between recommendations and requirements when addressing criteria. Detailed positions on the Regulatory Positions are presented below:a) Regulatory Position C.1 - Westinghouse will provide a means to facilitate response time testing from the sensor input at the protection rack to and including the input to the actuation device. Examples of actuation devices are the protection system relay or bistable.Westinghouse defines "Protective Action Systems" to mean the electric, instrumentation and controls portions of those protection systems and equipment actuated and controlled by the protection system.Equipment performing control functions, but actuated from protection system sensors is not part of the safety system and will not be tested for time response.Status, annunciating, display, and monitoring functions, except for those related to the Post Accident Monitoring System (PAMS), are considered to be control functions.Reasonability checks, i.e., comparison between or among similar such display functions, will be made. Otherwise, the clarification note in Position C.1 is observed.b) Regulatory Position C.5 - Response time testing for control functions operated from protection system sensors will not be performed. Nuclear instrumentation system detectors will not be tested for time response. The "expected environmental and mechanical configuration of the actual installation" will not be duplicated for the testing of sensors which must be removed to accomplish response time testing unless it can be shown that the duplication is practical and that the duplicated factors significantly influence the sensor time response.The Westinghouse scope protection system does not preclude the response time testing of process sensors by their removal at normal shutdown. The standard Westinghouse scope protection system does not include design provision which permit insitu testing of process or nuclear instrumentation system sensors. Nuclear instrumentation sensors are exempt from testing since "their worst case response time is not a significant fraction of the total overall system response (less than 5 percent)." This exemption is permitted by IEEE-338-1975.c) Regulatory Position C.6 - Temporary jumper wires, temporary test instrumentation, the removal of fuses and other equipment not hard-wired into the protection system will be used where applicable.Amendment 65 Page 56 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 d) IEEE Standard 338-1977, Section 5.7 states "Each test bypass condition utilized at a frequency of more than once a year shall be individually and automatically indicated to the operator in the main control room in such a manner that the bypassing of a protective function is immediately evident and continuously indicated."SHNPP takes exception to this requirement during performance of containment ventilation isolation from the containment ventilation isolation radiation monitors. During monthly testing of containment ventilation isolation, the opposite train of containment ventilation isolation is bypassed to ensure that both trains of containment vacuum relief are not made inoperable. This test is performed from the main control room and operators are made aware of the bypassed containment ventilation isolation from the radiation monitors.Balance of Plant - The SHNPP project complies with this guide.FSAR

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Sections 7.1.2.17, 7.3.2.2.10, 7.6.2.2.j, 8.3.1.2.20, and 13.5.1.3.e.Regulatory Guide 1.119 SURVEILLANCE PROGRAM FOR NEW FUEL ASSEMBLY DESIGNS (REV 0)This guide was withdrawn by the NRC on June 23, 1977.Regulatory Guide 1.120 FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS (REV 1)This guide is not applicable to the SHNPP project. For a discussion of fire protection requirements, see Section 9.5.1.Regulatory Guide 1.121 BASES FOR PLUGGING DEGRADED PWR STEAM GENERATOR TUBES (REV 0)The SHNPP project complies with this guide.Regulatory Guide 1.122 DEVELOPMENT OF FLOOR DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF FLOOR-SUPPORTED EQUIPMENT OR COMPONENTS (REV 1)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.123 QUALITY ASSURANCE REQUIREMENTS FOR CONTROL OR PROCUREMENT OF ITEMS AND SERVICES FOR NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.123 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Amendment 65 Page 57 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.124 SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 LINEAR-TYPE COMPONENT SUPPORTS (REV 1)The SHNPP project complies with the intent of this guide as described below:NSSS The Regulatory Guide states in paragraph B.1(b): "Allowable design limits for bolted connections are derived from tensile and shear stress limits and their non-linear interaction; they also change with the size of the bolt. For this reason, the increases permitted by NF-3231-1, XVII-2110(a), and F-1370(a) of Section III are not directly applicable to allowable shear stresses and allowable stresses for bolts and bolted connections.", and in paragraph C.4: "This increase of design limits does not apply to limits for bolted connections and shear stresses."As noted above, the increase in bolt allowable stress under emergency and faulted conditions is not permitted because: 1) the interaction between the allowable tension and shear stress in bolts is nonlinear, 2) the allowable tension and shear stress vary with the bolt size.Westinghouse believes that the present ASME Code rules are adequate since they do satisfy the two objectives raised in the above quoted paragraph and hence will use the present rules without further restrictions or justification. This position is based on the following:It is well recognized after extensive experimental work by several researchers that the interaction curve between the shear and tension stress in bolts is more closely represented by an ellipse and not a line. This has been clearly recognized by the ASME. The latest revision of Code Case 1644 specifies stress limits for bolts and represents this tension/shear relationship as a non-linear interaction equation (ellipse). This interaction equation has a built-in safety factor that ranges between 2 and 3 (depending on whether the bolt load is predominantly tension or shear) based on the actual strength of the bolt as determined by test (

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"Guide to Design Criteria for Bolted and Riveted Joints," Fisher and Struik, copyright 1974, John Wiley and Sons, P. 54).Balance of Plant Regulatory Guide Position C is a supplement to Subsection NF of the ASME Code, Section III, service limits. The balance of plant design is not bound by Subsection NF; but is based on MSS-SP-58, "Pipe Hangers and Supports - Materials, Design and Manufacturer," 1975 Edition.Regulatory Guide 1.125 PHYSICAL MODELS FOR DESIGN AND OPERATION OF HYDRAULIC STRUCTURES AND SYSTEMS FOR NUCLEAR POWER PLANTS (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.126 AN ACCEPTABLE MODEL AND RELATED STATISTICAL METHODS FOR THE ANALYSIS OF FUEL DENSIFICATION (REV 1)Westinghouse uses the fuel densification model presented in Reference 1.8-12 which has been approved by the NRC.Amendment 65 Page 58 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Siemens uses the fuel densification model presented in Reference 1.8-15 which has been approved by the NRC.Areva uses the fuel densification model presented in Reference 1.8-18 which has been approved by the NRC.Regulatory Guide 1.127 INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS (REV 1)The SHNPP project complies with this guide.Regulatory Guide 1.128 INSTALLATION DESIGN AND INSTALLATION OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS (REV 1)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.129 MAINTENANCE, TESTING AND REPLACEMENT OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS (REV 1)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.130 SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 PLATE-AND-SHELL TYPE COMPONENT SUPPORTS (REV 1)The SHNPP project meets the intent of the guide as described below:NSSS a) The Regulatory Guide states in Paragraph B.1: - "Allowable design limits for bolted connections are derived on a different basis that varies with the size of the bolt. For this reason, the increases permitted by NF-3224 and F-1323.1(a) of Section III are not directly applicable to bolts and bolted connections."It is the Westinghouse position that it is reasonable to allow an increase in the limits for bolted connections for emergency and faulted conditions. Further justification of this position can be found in the discussion of Regulatory Guide 1.124 on Class 1 linear type supports.b) Paragraphs C.3, C.4(a), and C.6(a) of the Regulatory Guide state that the allowable buckling strength should be calculated using a design margin of 2 for flat plates and 3 for shells for normal, upset and emergency conditions.In the design of plate-type supports, member compressive axial loads shall be limited per the requirements of Paragraph C.3 for normal, upset and emergency conditions.There are no Class l shell-type supports in the Westinghouse NSSS.c) In Paragraph C.7 of the Regulatory Guide, inclusion of the upset plant condition is inappropriate in the load combination under discussion. Westinghouse does not include the upset plant condition in this combination.Amendment 65 Page 59 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 d) In Paragraphs C.7(a) and B.1 of the Regulatory Guide, the stress limits of F-1370(c) are discussed. The criterion stated in F-1370(c), "...loads should not exceed 0.67 times the critical buckling strength of the support...."In the design of plate-type component supports, member compressive axial loads shall be limited to 0.67 times the critical buckling strength. If, as a result of a more detailed evaluation of the supports, the member compressive axial loads is shown to safety exceed 0.67 times the critical buckling for the faulted condition, verification of the support function adequacy is documented and submitted to the NRC for review. The member compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance.e) In Paragraph C.7(b) of the Regulatory Guide, the limit based on the test load given in the Regulatory Guide, T.L. x 0.7 Su /Su, is overly conservative and is inconsistent with ASME Code requirements presented in Appendix F.Westinghouse uses the provisions of F-1370(d) to determine service level D allowable loads for supports designed by the load rating method.Study of three interaction curves of allowable tension and shear stress based on the ASME Code (emergency condition allowables per XVII-2110 and faulted condition allowables per F-1370) and the ultimate tensile and shear strength of bolts (obtained from experimental work published by Chesson, Faustina, and Munse in "Proceedings of ASCE", October 1975) indicates that there is adequate safety margin between the emergency and faulted condition allowables and failure of the bolts.From this study it is observed that:

1) For the emergency condition, the safety factor (ratio of ultimate strength to allowable stress) varies between a minimum of 1.63 and a maximum of 2.73 depending upon the actual tensile stress/shear stress (T/S) ratio on the bolt.

2) For the faulted condition, the safety factor varies between a minimum of 1.36 to a maximum of 2.29, again depending upon actual T/S ratio on the bolt.

It is thus reasonable to allow an increase in these limits for the emergency and faulted conditions.

3) In Section III Subsection NA Table XVII-2461.1 the ASME Code provides the criterion for allowable stress on bolts. As per this table, the allowable stress depends upon the bolt size as well as the bolt material.

4) The structures designed to meet AISC Manual of Steel Construction have been proven to be adequately designed. It is also recognized that the ASME Code requirements for Class 1 linear type supports (Appendix XVII) have been derived from AISC Manual of Steel Construction. In Paragraph 1.5.6 of Manual of Steel Construction, AISC permits the increased allowables for "occasional loads such as wind and seismic". In view of this, restrictions by NRC on not permitting increased allowables under emergency and faulted conditions which also are "infrequent incidents and limiting faults" are not justified.

Amendment 65 Page 60 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Based on item 1 through 4 above, for the emergency and faulted conditions, Westinghouse uses allowable bolt stresses specified in Code Case 1644, latest revision, as increased according to the provisions of XVII-2110(a) and F-1370(a), respectively.Westinghouse uses the latest revision of Code Case 1644 as opposed to Revision 4 specified in the Regulatory Guide, Paragraph B.I(a).Concerning allowable shear stresses, a commonly used allowable shear stress for structural members is 0.55 Fy (Fy = yield stress). This limit is well documented in the literature and substantiated by numerous tests (see, for instance, the AISC Manual for Steel Construction, Part 2). This limit not only maintains the member load below the yield load, but keeps it well below the actual shear strength of the member. In addition, the moment carrying capacity of a structural member is not appreciably affected for shear loads corresponding to 0.55 Fy.Westinghouse uses the following criteria to make the allowable shear stress compatible with other allowable emergency and faulted condition stresses, and at the same time, keep shear stresses within proven limits: For the emergency condition, shear stresses may be increased according to the provisions of XVII-2110(a). Faulted condition allowable shear stresses should be limited to the lesser of 0.55 Fy or 0.45 Su, where Su is the material tensile strength.f) In Paragraphs B.5 and C.8 of the Regulatory Guide, Westinghouse takes exception to the requirement that systems whose safety-related function occurs during emergency or faulted plant conditions must meet upset limits. The reduction of allowable stresses to no greater than upset limits (which in reality are only design limits since design, normal and upset limits are the same for linear supports) for support structures in those systems with normal safety-related functions occurring during emergency or faulted plant conditions is overly conservative for components which are not required to mechanically function (inactive components). In addition, Westinghouse believes that emergency and faulted condition criteria are acceptable for active components. However, when these criteria are invoked for active components, any significant deformation that might occur is considered in the evaluation of equipment operability.g) Paragraph C.4 of the Regulatory Guide states: "However, all increases (i.e, those allowed by NF-3231.1(a), XVII-2110(a), and F-1370(a) should always be limited by XVII-2110(b) of Section III." Paragraph XVII-2110(b) specifies that member compressive axial loads shall be limited to 2/3 of critical buckling. Satisfaction of these criteria for the faulted condition is unnecessarily restrictive.The most significant faulted condition loads on equipment supports result from seismic disturbances and postulated loss-of-coolant accidents, both of which are dynamic events. The allowable faulted condition compressive load should not be limited to 2/3 of critical buckling because: (a) these faulted dynamic loads are of extremely short duration, and (b) support members can take impulsive loads that exceed static critical buckling load. Westinghouse will use a compressive axial load of 0.9 of critical buckling since the dynamic buckling capacity of the member is greater than the static buckling capacity.h) Paragraph C.2 of the Regulatory Guide presents two methods of estimating the ultimate tensile strength, Su, at elevated temperatures. It is believed that Method #2 is not Amendment 65 Page 61 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 conservative at elevated metal temperature (in excess of 800°F). In Westinghouse's judgment, values of Su at these elevated temperatures should be determined by test rather than via the method given in C.2(b).i) Paragraph C.6(a) of the Regulatory Guide appears to erroneously allow the use of faulted stress limits for the emergency condition. Westinghouse will interpret this paragraph as follows: "The stress limits of XVII-2000 of Section III and Regulatory Position 3, increased according to the provisions of XVII-2100(a) of Section III, should not be exceeded for component supports designed by the linear elastic analysis method."j) Westinghouse uses the provisions of F-1370(d) to determine faulted condition allowable loads for supports designed by the load rating method. The method described in Paragraph C.7(b) of the Regulatory Guide is very conservative and inconsistent with the remainder of the faulted stress limit.Balance of Plant Regulatory Guide Position C is a supplement to Section 3, Subsection NF of the ASME Code, Section III. The balance of plant design is not bound by Subsection NF, but is based on MSS-SP-58, "Pipe Hangers and Supports - Material Design and Manufacturers," 1975 Edition.Regulatory Guide 1.131 QUALIFICATION TESTS OF ELECTRIC CABLES, FIELD SPLICES, AND CONNECTIONS FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.132 SITE INVESTIGATIONS FOR FOUNDATIONS OF NUCLEAR POWER PLANTS (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.133 LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF LIGHT-WATER-COOLED REACTORS Carolina Power & Light Company concurs with the desirability of implementing a loose parts detection program in the SHNPP. Experience has shown the system's merits, functional adequacy and reliable performance in operating plants. Operating history has proven that there can be early warning benefits from experienced interpretation of system data. However, CP&L takes exception to requirements which go beyond the need for a reliable system which provides anything more than this basic reassurance. These exceptions are delineated as follows:a) The functional performance requirements for this system serve solely to provide an alert, by impact monitoring, for circ*mstances which could result from loose metallic objects in the Reactor Coolant System. The system (a) does not maintain the reactor coolant pressure boundary (b) serves no automatic reactor protection function, and (c) does not classify as an IE system as defined in IEEE-308. Consequently, the application of Class IE criteria is unjustified.Amendment 65 Page 62 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 b) Based on exception a) above, the requirements of Regulatory Guide 1.100 "Seismic Qualification of Electric Equipment for Nuclear Power Plants" are not applicable.c) Based on exception a) above, the redundancy and separation requirements for the system are not applicable, except for in-containment hardware where good engineering practice prevails.d) The use of system noise to provide functional tests is adequate. The need for calibrated simulated signals or in-containment calibration is unduly restrictive and unnecessary.Impact energy sensitivity is verifiable from system noise signatures.e) Based on exception a) above, the limiting condition for operation imposed by Regulatory Position C.5.b is not applicable. The availability of a non-Class IE system is not essential for continued safe operation.f) The implementation of the method described in Regulatory Guide 1.133 Section D is a backfit as defined by 10 CFR 50.109. Based on exception a) above, the backfit requirement cannot be justified.g) Carolina Power & Light Company also takes exception to technical requirements not enumerated here but which include, as an example, the restriction of the background noise level of the primary system in a nuclear plant to 20 percent of an arbitrary system sensitivity level simultaneously established by the guide.Regulatory Guide 1.134 MEDICAL CERTIFICATION AND MONITORING OF PERSONNEL REQUIRING OPERATOR LICENSES (REV 2)The SHNPP project complies with this guide.Regulatory Guide 1.135 NORMAL WATER LEVEL AND DISCHARGE AT NUCLEAR POWER PLANTS (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.136 MATERIAL FOR CONCRETE CONTAINMENTS (REV 0)This guide is not applicable to the SHNPP project.Regulatory Guide 1.137 FUEL-OIL SYSTEMS FOR STANDBY DIESEL GENERATORS (REV 1)Carolina Power & Light Company ensures the quality and reliability of the diesel generator fuel oil by implementing the diesel generator fuel oil surveillance requirements described in the SHNPP Technical Specifications.Carolina Power & Light Company uses the sampling and testing methods recommended in RG 1.137 with the following exceptions:Amendment 65 Page 63 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 a) Instead of the standard recommended in Part C.2.a of RG 1.137 (ASTM D975-77),CP&L will use methods described in ASTM D975-18 to test the quality and reliability of oil stored in the fuel oil supply tank and the oil used to fill or refill the supply tank.b) Instead of the two weeks recommended in Part C.2.b of RG 1.137, CP&L will complete analyses of fuel oil properties listed in applicable specifications (other than specific or API gravity, water and sediment and viscosity) within 31 days of the addition.c) Instead of the standard recommended in Part C.2.c of RG 1.137 (ASTM D270-1975),CP&L will periodically sample the fuel oil in accordance with ASTM-D6217-18 or ASTM-D7321-18 (as applicable).d) Exceptions to Appendix B to ANSI N195-1976 are described in Section 9.5.4.1 and 9.5.4.5.e) Fuel oil sampling may be performed at an offsite contracted lab that is on the Approved Suppliers List (ASL) or analyzed onsite to comply with the requirements of the Diesel Fuel Oil Testing Program as described in Technical Specification 6.8.4.q.These surveillance requirements are partly based on a Technical Report entitled, "Surveillance Requirements for Emergency Diesel Fuel Oil Systems in Nuclear Power Plants" (Ref. 1.8-14).This position endorses ASTM D975-81 and ANSI N195-1976 Appendix B with the exceptions noted above. Therefore, CP&L meets the intent of RG 1.137. ASTM D975-18 will be used instead of ASTM D975-81 to test quality and reliability of new and stored fuel oil to the current methods.Regulatory Guide 1.138 LABORATORY INVESTIGATIONS OF SOILS FOR ENGINEERING ANALYSIS AND DESIGN OF NUCLEAR POWER PLANTS (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.139 GUIDANCE FOR RESIDUAL HEAT REMOVAL (REV 0)Section D, "Implementation", of Regulatory Guide 1.139 states that the method described will be used in the evaluation of submittals for construction permit applications docketed after January 1, 1978. On this basis, CP&L feels that compliance with this regulatory guide is not mandated.However, in recognition of the NRC's prerogative to review against this guide on a case-by-case basis, the following discussion of compliance is presented. The responses are organized per the appropriate subsection of Section C, "Regulatory Position".

1. Shearon Harris is a Class 2 plant, as defined by the implementation section of BTPRSB5-1.

The safe shutdown design basis is hot standby. Thus, Shearon Harris does not fully comply with the functional requirements of Regulatory Guide 1.139.Four key functions are required to achieve and maintain cold shutdown. Means for performing these functions are described below:Amendment 65 Page 64 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1

1. Circulation of the reactor coolant can be provided first by natural circulation (that is, using the reactor core as the heat source and the steam generators as the heat sink) and then by the residual heat removal pumps.

2. Removal of residual heat can be accomplished first via the Auxiliary Feedwater System and then via the residual heat removal heat exchanger. Hot standby can be maintained by releasing steam via the safety grade steam generator safety valves. Cooldown to 350°F can be accomplished by releasing steam via operation of the steam generator power operated relief valves. Then cooldown to cold shutdown conditions can be achieved with the Residual Heat Removal System. A sufficient Seismic Category I supply of auxiliary feedwater to permit four hours' operation at hot standby plus cooldown to Residual Heat Removal System initiation conditions is provided by the condensate storage tank and the backup supply from the Emergency Service Water System.

3. Boration can be accomplished using portions of the Chemical and Volume Control System. Boric acid from the boric acid tanks can be supplied to the suction of the centrifugal charging pumps by the boric acid transfer pumps. The centrifugal charging pumps can inject the boric acid into the reactor coolant system via the safety injection flow path or the normal charging and reactor coolant pump seal injection flow paths. Makeup in excess of that needed for boration can be provided from the refueling water storage tank.

4. Depressurization can be accomplished using portions of the Chemical and Volume Control System. Either boric acid from the boric acid tanks or refueling water can be used as desired for depressurization with the flow path being via the centrifugal charging pumps and auxiliary spray valve to the pressurizer.

Plant operation conditions which allow operation of the Residual Heat Removal System (approximately 350°F, 360 psig) can be achieved in approximately 36 hours or less following plant shutdown. Section 7.4.1 of the FSAR provides more descriptive information on the equipment used for shutdown and identifies other reference sections in the FSAR.2a) With regard to isolation of the suction side of the RHRS, SHNPP meets the requirements of Regulatory Guide 1.139. Refer to FSAR Sections 5.4.7.1, 5.4.7.2.1, 5.4.7.2.4, and 5.4.7.2.6.2b) With regard to isolation of the discharge side of the RHRS, SHNPP meets the requirements of Regulatory Guide 1.139. Refer to FSAR Section 5.4.7.1.3a(1) The protection of the RHRS against overpressurization meets the requirements set forth in this part of Regulatory Guide 1.139. Refer to FSAR Sections 5.4.7.2.4, 5.4.7.2.5, and 3.2.3.a(2) When the RHRS relief valves are stuck open, the fluid discharged inside containment goes to the pressurizer relief tank and the fluid discharged outside containment goes to the boron recycle holdup tank. Refer to FSAR Section 5.4.7.2.4.Amendment 65 Page 65 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 3.b. The effect of a relief valve stuck in the open position can be mitigated by isolation of the associated RHRS loop. The minimum design capabilities of the ECCS require one RHRS loop to function. Refer to FSAR Section 6.3.2.3c. See 3a(2) above.

4. SHNPP complies with Regulatory Guide 1.139. Refer to FSAR Section 5.4.7.2.

5. SHNPP meets the requirements of Regulatory Guides 1.68 and 1.139. Refer to FSAR Sections 14.2.12.28 and 14.2.12.29. The system can be tested during operation.

6. The RHRS meets the requirements of Regulatory Guide 1.139 due to the fact that the AFWS has the ability to remote manually take suction from the SWS to provide feedwater for much longer than four hours. Refer to FSAR Sections 10.4.9.2.2 and 10.4.9.3.

7. The operational procedures will be written to comply with the intent of Regulatory Guide 1.139.

Regulatory Guide 1.140 DESIGN, TESTING AND MAINTENANCE CRITERIA FOR NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (REV 1)The SHNPP project meets the intent of this guide as described in Sections 9.4.3, 9.4.4, and 9.4.7 with the following exceptions:Regulatory Position Exceptions c.3.b This section requires HEPA filters to be in accordance with MIL-F-51068. MIL-F-51068 has been canceled and replaced by ASME AG-1; therefore, HEPA filter requirements will be allowed to either specification.3.g, 6.a, 6.b Each original or replacement batch of impregnated activated carbon used in the absorber section should meet the qualification and batch test results summarized in Table 5.1 of ANSI/ASME N509-1980 with the additional exception that the 30°C/95% relative humidity methyl iodide test is performed per ASTM D3803-1989.5.a, 5.b, 5.c, 5.d In-Place Testing to be performed per ANSI/ASME N510-1980.6.a, 6.b Laboratory tests (radioiodine removal efficiency) of representative samples of used activated carbon to be performed per ASTM D3803-1989 at 30°C and 70% relative humidity.Regulatory Guide 1.141 CONTAINMENT ISOLATION PROVISIONS FOR FLUID SYSTEMS (REV 0)Amendment 65 Page 66 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 The SHNPP project complies with this guide as described below:

1) Design criteria for closed systems used as isolation barriers is described in Section 6.2.4.

2) Containment isolation valve leakage testing will be performed in accordance with Appendix J to 10 CFR 50 as described in Section 6.2.6.

Regulatory Guide 1.142 SAFETY-RELATED CONCRETE STRUCTURES FOR NUCLEAR POWER PLANTS (OTHER THAN REACTOR VESSELS AND CONTAINMENTS) (REV 0)This guide does not have to be addressed by the SHNPP project.Regulatory Guide 1.143 DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (REV 1)The SHNPP project meets the intent of this guide. The Waste Processing Systems meet the requirements of Regulatory Guide 1.143. These guidelines are spelled out in SRP Section 11.2 and BTP ETSB 11.1 (Revision 1).The QA program described in BTP ETSB 11-1, Revision 1, is being applied with the clarification that (i) Vendor QA programs based on ASME Section VIII or ANSI B31.1 for Boiler External Piping are considered acceptable to comply with ETSB 11-1 QA requirements; and (ii) the Radwaste QA Program applies only to the pressure boundary.The hydrostatic testing of piping systems is being applied as described in BTP ETSB 11-1, Revision 1 except that, where such testing would damage equipment, the 75 psig minimum shall be waived. In this case, pneumatic testing should be performed.The Waste Processing System meets the design and construction specifications with the exception to the recommendations, of the screwed connections providing the only seal for quick disconnects, small manual and pressure relief valves, which are not available in any flanged, socket weld or butt welded type.The Waste Processing System meets the material specifications requirements of Regulatory Guide 1.143 with the exception of the flexible high pressure hoses utilized in the Modular Fluidized Transfer Demineralization System. Due to the Modular Design of this system it is necessary to use quick disconnects and high pressure hoses. It allows the operational flexibility needed for this system. These hoses were hydrostatically tested to 225 psig. They have a design operational pressure of 150 psig. The total dynamic head of the floor drain feed pumps are 117 psig. The hoses will be hydrostatically tested or replaced every four years.Regulatory Guide 1.144 AUDITING OF QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.144 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Amendment 65 Page 67 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.146 QUALIFICATION OF QA PROGRAM AUDIT PERSONNEL FOR NUCLEAR POWER PLANTS Conformance with Regulatory Guide 1.146 is addressed in the description of the Quality Assurance Program that is incorporated by reference into the FSAR Chapter 17 (see Section 17.3)Regulatory Guide 1.147 INSERVICE INSPECTION CODE CASE ACCEPTABILITY, ASME SECTION XI DIVISION I This Regulatory Guide lists those Section XI ASME Code Cases that are generally acceptable to the NRC Staff for implementation in the inservice inspection of components and supports at light-water cooled nuclear power plants. The Carolina Power & Light Company may make use of these code cases provided for in the Regulatory Guide.Regulatory Guide 1.149 NUCLEAR POWER PLANT SIMULATION FACILITIES FOR USE IN OPERATOR TRAINING, LICENSE EXAMINATIONS, AND APPLICANT EXPERIENCE REQUIREMENTS (Rev. 4)The SHNPP project complies with Regulatory Guide 1.149 which endorses ANSI/ANS-3.5-2009.Regulatory Guide 1.150 ULTRASONIC TESTING OF REACTOR VESSEL WELDS DURING PRESERVICE AND INSERVICE EXAMINATIONS REVISION 1 The SHNPP project complies with the recommendation of this guide. The Inservice Inspection Program Plan identifies the Reactor Vessel Welds requiring the use of this Regulatory Guide.Regulatory Guide 1.155 STATION BLACKOUT (August 1988)The SHNPP project complies with Regulatory Guide 1.155.FSAR

Reference:

Section 8.3.1.2.21 Regulatory Guide 1.183 ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS (July, 2000)The SHNPP project complies with the intent of Regulatory Guide 1.183 as described in the specific event analyses of FSAR Section 15.FSAR

References:

Sections 6.5.2, 15.09, 15.15, 15.26, 15.3.3, 15.4.3 (by reference to 15.3.3),15.4.7 (by reference to 15.4.3), 15.4.8, 15.6.2 (only for use of TEDE/10 CFR 50.67 dose limits; methods still remain described in Standard Review Plan), 15.6.3, 15.6.5, 15.7.1 (only for use of TEDE/10 CFR 50.67 dose evaluation method; assumptions and limits still remain described in HNP Tech Specs), 15.7.4.Amendment 65 Page 68 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Regulatory Guide 1.192 OPERATION AND MAINTENANCE CODE CASE ACCEPTABILITY, ASME OM CODE This Regulatory Guide lists those ASME OM Code cases that are generally acceptable to the NRC staff for implementation in the Inservice Testing and Examination of components at light-water cooled nuclear power plants. The Carolina power and Light Company may make use of these Code cases provided for in the Regulatory Guide.

REFERENCES:

SECTION 1.8 1.8-1 "Reactor Coolant Pump Integrity in LOCA," WCAP-8163, September 1973.1.8-2 Enrietto, J. F., "Control of Delta Ferrite in Austenitic Stainless Steel Weldments,"WCAP-8324-A, June 1975.1.8-3 Enrietto, J. F., "Delta Ferrite in Production Austenitic Stainless Steel Weldments,"WCAP-8693, January 1976.1.8-4 Caplan, J. S., "The Application of Preheat Temperatures After Welding Pressure Vessel Steels," WCAP-8577, February 1976.1.8-5 "Damping Values of Nuclear Power Plant Components," WCAP-7921-AR, May 1974.1.8-6 "Westinghouse Nuclear Energy Systems Divisions Quality Assurance Plan," WCAP-8370, Revision 7A, February 1975.1.8-7 "Westinghouse Water Reactor Divisions Quality Assurance Plan," WCAP 8370, Revision 8A, September 1977.1.8-8 Risher, D. H., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1A, January 1975.1.8-9 "Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety-Related Electrical Equipment," WCAP-8587.1.8-10 Deleted by Amendment No. 45.1.8-11 Deleted by Amendment No. 45.1.8-12 Hellman, J. M. (Ed.), Fuel Densification Experimental Results and Model for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP 8219-A (Non-Proprietary),March 1975.1.8-13 "Westinghouse Water Reactor Divisions Quality Assurance Plan," WCAP-8370 Revision 9A, October 1979.1.8-14 Strauss, K. H. "Surveillance Requirements for Emergency Diesel Fuel Oil Systems in Nuclear Power Plants" prepared for Standardized Nuclear Unit Power Plant System, September 23, 1983.Amendment 65 Page 69 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 1.8-15 Siemens Power Corporation, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," XN-NF-81-58(P)(A), Supplements 1 and 2, Revision 2 (NRC Safety Evaluation Report Issued November 16, 1983).1.8-16 Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.1.8-17 NRC letter from Conrad McCracken, NRC, to Barry McCrudden, Lehigh Testing Laboratories, Inc. "Use of ASTM-C-692-77," dated June 21, 1989.1.8-18 BAW-10231P-A, Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004.Amendment 65 Page 70 of 70

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE TITLE 1.1.1-1 ACRONYMS USED IN THE FSAR 1.1.1-2 ABBREVIATIONS USED IN THE FSAR 1.1.1-3 MAJOR BUILDINGS AND STRUCTURES 1.2.3-1 DELETED BY AMENDMENT NO. 15 1.3.1-1 DESIGN COMPARISON WITH SIMILAR FACILITIES (INITIAL FUEL CYCLE) 1.3.2-1 COMPARISON OF FINAL AND PRELIMINARY INFORMATION SIGNIFICANT DESIGN CHANGES 1.5.1-1 DELETED BY AMENDMENT NO. 48 1.5.1-2 DELETED BY AMENDMENT NO. 48 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE 1.6-2 OTHER REPORTS INCORPORATED BY REFERENCE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE 1.6-4 PROCEDURES, PROGRAMS, OR MANUALS INCORPORATED BY REFERENCE 1.7.1-1 DELETED BY AMENDMENT NO. 48 1.7.1-2 DELETED BY AMENDMENT NO. 48 1.7.1-3 DELETED BY AMENDMENT NO. 48 1.7.1-4 DELETED BY AMENDMENT NO. 48 1.7.2-1 DELETED BY AMENDMENT NO. 48 1.8-1 FUNCTIONAL LEVEL, ASSIGNMENT OF RESPONSIBILITY, AND QUALIFICATIONS CROSS REFERENCE FOR SHNPP Amendment 65 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR AAB Accident Analysis Branch AASHTO American Association of State Highway and Transportation Officials AB auxiliary boiler A/C air conditioning ABMA American Boiler Manufacturers' Association ACI American Concrete Institute ACP auxiliary control panel ACRS Advisory Committee on Reactor Safeguards A/E architect/engineer AEC Atomic Energy Commission AF alternating field AFI Air Filter Institute AFAS Auxiliary Feedwater Actuation Signal AFBMA Anti-Friction Bearing Manufacturer's Association AFS Auxiliary Feedwater System AFW Auxiliary Feedwater AGMA American Gear Manufacturers Association AGU American Geophysical Union AI authorized inspector AIA authorized inspection agency AISC American Institute of Steel Construction ALARA as low as is reasonably achievable AMCA Air Moving & Conditioning Association ANS American Nuclear Society ANSI American National Standards Institute AOV air-operated valve APCSB Auxiliary and Power Conversion Systems Branch API American Petroleum Institute ARI A/C and Refrigeration Institute ARMS Area Radiation Monitoring Subsystem AS auxiliary steam ASB Auxiliary Systems Branch ASCE American Society of Civil Engineers ASHRAE American Society of Heating, Refrigerating and Air Conditioning ASME American Society of Mechanical Engineers ASNT American Society of Nondestructive Testing ASNT-TG-IA American Society of Nondestructive Testing - Training Guide IA Amendment 64 Page 1 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR ASTM American Society for Testing and Materials ATWS anticipated transients without scram ATWT anticipated transients without trip AVT all volatile treatment AWS American Welding Society AWWA American Water Works Association BAT boric acid tank BCMS Boron Concentration Measurement System BEF Best Estimate Flow BIR Boron injection recirculation BIST Boron injection surge tank BIT boron injection tank BOL beginning of life BOP balance-of-plant B&PV boiler and pressure vessel BRS Boron Recycle System BTP Branch Technical Position BTR Boron Thermal Regeneration BTRS Boron Thermal Regeneration System BWR boiling water reactor CAPES Containment Atmosphere Purge Exhaust System CAS Compression Air System CB Control board CC Centrifugal charging CEA French Atomic Energy Commission C of E U. S. Army Corps of Engineers CF Correction factor CIAS containment isolation actuation signal CR Control Room CCS Containment Cooling System CCW component cooling water CCWS Component Cooling Water System CDC Computer design code CFR Code of Federal Regulation CHRS Containment Heat Removal System CIS Containment Isolation System CIT Corporate Investigation Team Amendment 64 Page 2 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR CMAA Crane Manufacturers Association of America, Incorporated CMTR Certified Material Test Reports CNS Corporate Nuclear Safety COR City of Raleigh COV center of vortex CP Construction Permit CPB Core Performance Branch CPDS Condensate Polishing Demineralizer System CPER Construction Permit Environmental Report CPI center pressure index CPIS containment purge isolation signal CP&L Carolina Power & Light Company CPM critical path method CPP Containment Pre-Entry Purge System CPPMU Containment Pre-Entry Purge Makeup CPRW condensate polishing regeneration waste CPRWCT condensate polishing regeneration waste collection tank CQAA Corporate Quality Assurance Audit CRACS Control Room Air Conditioning System CRD Control Rod Drive CRDM control rod drive mechanism CRDS control rod drive system CREST Committee on Reactor Safety Technology CRI Control Room Indicator CRM Chemical Remanent Magnitization CRT cathode ray tube CS carbon steel CSAS containment spray actuation signal CSB Containment Systems Branch CSIP Charging/Safety Injection Pump CSS Containment Spray System CST condensate storage tank/Central Standard Time CSTB Condensate Storage Tank Building CT Compactension CVCS Chemical and Volume Control System CVPETS Condensate Vacuum Pump Effluent Treatment System CVIS Containment Ventilation Isolation Signal Amendment 64 Page 3 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR CWS Circulating Water System DAF dynamic amplification factor DBA Design basis accident DBE Design basis earthquake DCC Daniel Construction Company DCN design change notice DDNB delayed departure from nucleate boiling DDR deficiency and disposition reports DEB double-ended break DECL double-ended cold leg DECLG double-ended cold leg guillotine DEHL double-ended hot leg DEHLG double-ended hot leg guillotine DEIT digital electro-hydraulic DEMA Diesel Engine Manufacturers Association DEP Duke Energy Progress, Inc.DEPS double-ended pump suction DEPSG double-ended pump suction guillotine DER Design electrical rating DF decontamination factor DG diesel generator DGB Diesel-Generator Building DGCAIES Diesel Generator Combustion Air Intake and Exhaust System DGFOSTS Diesel Generator Fuel Oil Storage and Transfer System DLF dynamic load factor DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DOE Department of Energy DOP dioctyl phthalate DOT Department of Transportation DPE discipline project engineer DWR Department of Water Resources DWS Demineralized Water System DWST demineralized water storage tank DWT Drop weight test Ebasco Electric Bond and Share Company ECAR East Central Area Reliability Amendment 64 Page 4 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR ECCS Emergency Core Cooling System EDT Eastern Standard Time E&CQA Engineering and Construction Quality Assurance/Quality Control EES Emergency Exhaust System EFDS Equipment and Floor Drain System EAB Exclusion area boundary EFPD effective full-power day(s)EFPH effective full-power hour(s)EFPY effective full-power year(s)EH electric hydraulic EHC Electrohydraulic Control (System)EICSB Electrical, Instrumentation, and Control Systems Branch EOC Emergency Operations Center EOL end of life EPC Engineering Planning Coordinator EPM engineering project manager EPR ethylene propylene rubber EPRI Electric Power Research Institute EPZ Emergency Planning Zone EQDP Environmental Qualifications Data Package ER environmental report ERDA Energy Research and Development Administration ESCWS Essential Services Chilled Water System ESDR engineered safeguards design rating ESF Engineered Safety Feature(s)ESFAS Engineered Safety Features Actuation System ESSA Environmental Science Services Administration EST Eastern Standard Time ERTS Earth Resources Technology Satellite ESWS Emergency Service Water System ETSB Effluent Treatment Systems Branch EVS Emergency Voice System EZB exclusion zone boundary FA forced air FAA Federal Aviation Administration FANP Framatome Advanced Nuclear Power FCC Federal Communications Commission Amendment 64 Page 5 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR FCV flow control valve FDS Floor Drain System FDT floor drain tank FERP Fire Emergency Response Plan FHB Fuel Handling Building FHS Fuel Handling System FMEA failure modes and effects analysis FMR Factory Material Research FOA forced oil air FOC Fine Offices Committee of the British Standards Institution FPC Federal Power Commission FPT feed pump turbine FR friction ratio FSAR Final Safety Analysis Report FTS Fuel Transfer System FTU Formazine turbidity unit FW feedwater FWIV feedwater isolation valve FWS Feedwater System GDC general design criteria GFFD gross failed fuel detector GM geiger mueller GWMS Gaseous Waste Management System GWPS Gaseous Waste Processing System H&V heating and ventilating HEPA high-efficiency particulate air filters HEI Heat Exchanger Institute HHSI high-head safety injection HI Hydraulic Institute HITC Hydraulic Institute Test Codes HNP Harris Nuclear Plant HSST Heavy Section Steel Technology HTP High Thermal Performance HVAC heating, ventilating, and air conditioning HX heat exchanger I&C instrumentation and control ICEA Insulated Cable Engineers Association Amendment 64 Page 6 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR ICRP International Commission on Radiation Protection ID inside diameter IEEE Institute of Electrical and Electronic Engineers IES Illumination Engineering Society IH integrated head ILRT integrated leakage rate test IPCEA Insulated Power Cable Engineers Association IRS Iodine Removal System ISI inservice inspection LBB Leak Before Break LED light-emitting diode LEFM linear elastic fracture mechanics LFDCP local fire detection control panel LFL lower flammability limit LHSI low-head safety injection LHST laundry and hot shower tank LL liquid limit LMTD log mean temperature difference LNG liquified natural gas LOCA loss-of-coolant accident LOPAR Low Parasitic LOR lower oil reservoir LP low pressure LPG liquid petroleum gas LPZ low population zone LRTS Liquid Radwaste Treatment System LSA low specific activity LSB last stage blade LTC linear translation case LTMD less than minimum detectable (concentration)LVDT linear variable differential transducers LWMS Liquid Waste Management System LWPS Liquid Waste Processing System LWR Light Water Reactor M&E Mass and Energy M&TE Measuring and Test Equipment MAAC Mid-Atlantic Area Council Amendment 64 Page 7 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR MCB main control board MCC motor control center MCES Main Condenser Evacuation System MDC moderator density coefficient MEB Mechanical Engineering Branch MFCS Main Feedwater Control System MFIS main feedwater isolation signal MFIV main feedwater isolation valve MFWLB main feedwater line break MG motor generator MIL military standards MIMS metal impact monitoring system MLW mean low water MOV motor-operated valve MPBB maximum permissible body burden MPC maximum permissible concentration MPCa maximum permissible concentration in air MPCw maximum permissible concentration in water MS main steam MSIS main steam isolation signal MSIV main steam isolation valve MSL mean sea level MSLB main steam line break MSLI main steam line isolation MSR moisture separator reheater MSS Manufacturers Standardization Society MSSS Main Steam Supply System MTC moderator temperature coefficient MTEB Materials Engineering Branch MUR-PU Measurement Uncertainty Recapture - Power Uprate MWS Makeup Water System MWST Makeup Water Storage Tank NASA National Aeronautics and Space Administration NASTRAN NASA Structural Analysis Computer Program NBS National Bureau of Standards NCDWR North Carolina Department of Water Resources NCHP North Carolina Highway Patrol Amendment 64 Page 8 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR NCRERP North Carolina Radiation Emergency Response Plan NCSB North Carolina State Building Code NCSU North Carolina State University NDRC National Defense Research Committee NDE nondestructive examination NDT nil-ductility transition or nondestructive testing NDTT nil-ductility transition temperature NEC National Electric Code NELPIA Nuclear Engineering Liability and Property Insurance Association NEMA National Electrical Manufacturer's Association NEPIA Nuclear Engineering Property Insurance Association NESCWS Non-Essential Service Chilled Water System NFPA National Fire Protection Association NGS National Geodetic Survey NIOSH National Institute of Occupational Safety and Health NIS Nuclear Instrumentation System NM nautical miles NMC Nuclear Mutual Limited NNS non-nuclear safety NOAA National Oceanic and Atmospheric Administration NPS nominal pipe size NPSH net positive suction head NPT national pipe thread NRC Nuclear Regulatory Commission NRMCA National Ready Mixed Concrete Association NSF National Sanitation Foundation/National Science Foundation NSSS Nuclear Steam Supply System NUPPSCO Nuclear Power Plant Standards Committee NWL normal water level NWS National Weather Service OA oil-to-air OBE Operating Basis Earthquake OCD Office of Civil Defense OD outside diameter OQA Operations Quality Assurance OQAP Operational Quality Assurance Plan ORE occupational radiation exposure Amendment 64 Page 9 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR ORNL Oak Ridge National Laboratory ORVHSF Old Reactor Vessel Head Storage Facility OSC Operational Support Center OSGSF Old Steam Generator Storage Facility OSHA Occupational Safety and Health Administration PA Public Address PAG Protective Action Guides PAMS Post-Accident Monitoring System PABX Private Automatic Brand Exchange PAMI Post Accident Monitoring Instrumentation PAP plant administrative procedure PCI Prestress Concrete Institute/Pellet-to-Clad Interaction PCT Peak clad temperature PG Particulate and noble gas monitor PGDS Pressurized Gas Distribution System PI plasticity index PIG Particulate, iodine, and noble gas monitor P&ID piping and instrument diagram PL plastic limit PMF probable maximum flood PMH probable maximum hurricane PMP probable maximum precipitation PMS Primary Makeup System PMW probable maximum wind PNSC Plant Nuclear Safety Committee POQAP plant operational quality assurance procedure PORV power-operated relief valve PPC pore pressure cell PPDI Plant process display instrumentation PPM Procedure Preparation Manual PRT pressurizer relief tank PS pressurizer surge PSAR Preliminary Safety Analysis Report PSB Power System Branch PSS Process Sampling System PSWS Potable and Sanitary Water System PT Peroidic Testing Amendment 64 Page 10 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR PUR Power Uprate PVC polyvinyl chloride PWR pressurized water reactor QA quality assurance QAA quality assurance audit QAB Quality Assurance Branch QA/QC quality assurance/quality control QAP quality assurance procedures QC quality control QCP quality control procedures RAB Reactor Auxiliary Building RABNVS Reactor Auxiliary Building Normal Ventilation System RABSRVS Reactor Auxiliary Building Switchgear Room Ventilation System RAM random access memory RCA reactor coolant activity RCB Reactor Containment Building RCC rod cluster control RCCA rod cluster control assembly RCDT reactor coolant drain tank RCFC reactor containment fan cooler RCL Reactor Coolant Loop RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS Reactor Coolant System RC&T Radiation Control and Test RDT Reactor Development Technology Division of the NRC RG Regulatory Guide RH relative humidity RHR Residual heat removal RHRS Residual Heat Removal System RHT recycle holdup tank RIC rotary inertia included case RIS Reservoir induced seismicity RMS Radiation Monitoring System RMWS Reactor Makeup Water System RNDT reference nil ductility temperature RO reactor operator Amendment 64 Page 11 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR RPS Reactor Protection System RPT radiation protection technician RPV reactor pressure vessel RQD rock quality designation RRS required response spectrum RSB Reactor Systems Branch RSG Replacement Steam Generator RT reference temperature RTD resistance temperature detector RTP rated thermal power RTS Reactor Trip System RTT reference transition temperature RWP radiation work permit RWST refueling water storage tank SAB Site Analysis Branch SAMA Scientific Apparatus Manufacturers Association SAR Safety Analysis Report SAT spray additive tank SC safety class SCA single channel analyzer SCFM standard cubic feet per minute SCR silicon control rectifier SCS Soil Conservation Service SDD system design description SDF spillway design flood SDR supplier deviation request SDS Steam Dump System SEB Structural Engineering Branch SERC Southeastern Electric Reliability Council SFA spent fuel assembly SFPCCS Spent Fuel Pool Cooling and Cleanup System SG steam generator SGAS Steam Generator Available Signal SGBS Steam Generator Blowdown System SGFP steam generator feedwater pump SGR Steam Generator Replacement SGTP Steam Generator Tube Plugging Amendment 64 Page 12 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR SGTR Steam Generator Tube Rupture SHEEC Shearon Harris Energy and Environmental Center SHNPP Shearon Harris Nuclear Power Plant SI safety injection SIAS safety injection actuation signal SI/EB safety injection/emergency boration signal SimVS Simulator Voice System SIS Safety Injection System SIT structural integrity test SLB status light boxes SLF seismic load factor SMA shielded metal arc SMACNS Sheet Metal and Air Conditioning Contractors National Association SPC Siemens Power Corporation SPF standard project flood SPS standard project storm SPT standard penetration test SRO senior reactor operator SRP Standard Review Plan SRSS square root of the sum of the squares SRST spent resin storage tank SRV safety relief valves SS stainless steel SSE Safe shutdown earthquake SSPC Steel Structure Painting Council SSPS Solid-State Protection System SSS Secondary Sampling System STA Shift Technical Advisor STP standard temperature and pressure SVS Site Voice System SWPS Solid Waste Processing System SWS Service Water System TB Turbine Building TC thermocouple TDC thermal diffusion coefficient TDH Total Developed Head TDS total dissolved solids Amendment 64 Page 13 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR TEFC totally enclosed fan cooled TEMA Tubular (Exchanger) Manufacturer's Association TG turbine generator TGB Turbine-Generator Building TLD thermoluminescent dosimeter TSC Technical Support Center UBC Uniform Building Code UFL upper flammability limit UHS ultimate heat sink UL Underwriters' Laboratories UNC University of North Carolina UOR upper oil reservoir UPS uninterruptible power supply USAEC U. S. Atomic Energy Commission USASI U. S. American Standards Institution USBR U. S. Bureau of Reclamation USDA U. S. Department of Agriculture USGS U. S. Geological Survey USNRC U. S. Nuclear Regulatory Commission USPHS U. S. Public Health Service UTM universal transverse mercator TGSS Turbine Gland Sealing System TGS Turbine Generator & Associated Systems VCT volume control tank VHF very high frequency VRS Volume Reduction System VSL0 valve stem leak-off VWO valve wide open WAPD Westinghouse Atomic Power Division WCAP Westinghouse Commercial Atomic Power WECT waste evaporator condensate tank WHT waste holdup tank WMT waste management tank NES Westinghouse Nuclear Energy System WPB Waste Processing Building WPBCWS Waste Processing Building Cooling Water System WPCB waste processing control board Amendment 64 Page 14 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Table 1.1.1-1 ACRONYMS USED IN THE FSAR WPS Waste Processing System Westinghouse Westinghouse Electric Corporation Amendment 64 Page 15 of 15

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-2 ABBREVIATIONS USED IN THE FSAR Words or Term Abbreviation Acre ac.acre-foot ac.-ft.actual cubic feet per minute acfm alternating current AC ampere a ampere-hour ah angular velocity l atmosphere atm atomic mass unit amu atomic percent at.%average avg average temperature Tavg before present bp billion cubic feet Mmcf brake horsepower bhp British thermal unit Btu British thermal units per lbm Btu/lbm British thermal units per hour Btu/hr.calorie cal centigram cg centimeter cm Chi/Q /Q coefficient Cr counts per minute counts/min or cpm 3cubic centimeter cm or cc 3cubic foot ft.3 cubic feet per hour ft. /hr.3 cubic feet per minute ft. /min. or cfm 3cubic feet per second ft. /sec. or cfs 3cubic feet per second-days ft /sec-days 3cubic inch in.3 cubic meter m cubic yard cy curie Ci cycles per second cps or Hz decibel Db degrees Baume' B Degrees Rankine R degrees Centigrade C degrees Fahrenheit per hour F/hr Amendment 61 Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-2 ABBREVIATIONS USED IN THE FSAR degrees Kelvin K direct current DC disintegrations per minute dpm dry bulb temperature dbT electromagnetic units emu electron volt ev exponent E feet (foot) ft.feet per minute ft./min or fpm feet per second ft./sec. or fps fluence, neutron nvt flux, neutron nv foot candle ft.-cdl foot-pound ft.-lb.gallon gal.gallons per day gal./day or gpd gallons per hour gal./hr.gallons per minute gal./min. or gpm gallons per second gal./sec. or gps gallons per year gpy giga volt ampere gva gram g Hertz Hz horsepower HP horsepower-hour HP-hr.hour hr.inch (inches) in.inch-pound in.-lb.inches per second ips or in./sec.inches of mercury absolute in. Hg abs.kiloelectron volt kev kilogram kg 3kilograms per cubic meter kg/m kilograms per second kg/sec.kilometer km kilopounds per hour kpph kilovolt Kv kilovolt-ampere kVa kilowatt Kw kilowatt-hour Kwh linear foot LF liter l Amendment 61 Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-2 ABBREVIATIONS USED IN THE FSAR logarithm, Napierian base 2.718 e mean sea level MSL mercury (absolute) Hg (abs) megawatt MW megawatt day Mwd megawatt days per metric ton of uranium MWD/MTU megawatt (electric) Mwe megawatt hour MwH megawatt (thermal) Mwt megohm Meg-ohm meter m 3microcuries per cubic centimeter µCi/cm microcuries per gram µCi/gm

 -6 micron, micro (10 ) µ micro mho µ mho mile per hour mph milli m milliampere ma millicurie mCi milliliter ml millimeter mm millimicron µm million cubic feet mcf million electron volts Mev 2

million electron volts per square centimeter mev/cm million gallons per day MGD million years my milliroentgen Mr milliroentgen per hour Mr/hr.milliroentgen equivalent man mrem mili second msec millivolt mv millivolt amperes MVA millivolt amperes reactive mvar milliwatt mw minute min.neutron multiplication factor, effective Keff neutron multiplication factor, infinity K 2neutrons per square centimeter n/cm 2neutrons per square centimeter-second n/cm -sec.ohms, million Meg ohm-1 mho Amendment 61 Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-2 ABBREVIATIONS USED IN THE FSAR one thousands pounds kip parts per million ppm parts per billion ppb pcm percent mille phase ph.poly vinyl chloride PVC pound lb.pound mass lbm pound mass per hour lbm/hr pound mass per second lbm/sec 3pounds per cubic foot lb./ft. or pcf 2pounds per square foot lb./ft. or psf pounds per hour lb./hr.pounds per second lb./sec pounds per square inch psi pounds per square inch (absolute) psia pounds per square inch (differential) psid pounds per square inch (gage) psig radius r reactive kilovolt-ampere kvar reactive volt-ampere var reactivity k/k reference temperature Tref revolutions per minute rpm revolutions per second rps roentgen R roentgen equivalent man rem roentgens per hour R/hr.root mean square rms running foot RF second sec.specific gravity sp gr 2square ( ) or sq.2 square foot ft. or sq. ft.2 square inch in. or sq. in.2 square mile mi. or sq. mi.3 standard cubic feet scf or std ft 3standard cubic feet per minute scfm or std ft. /min.standard cubic feet per second scfs Thickness T thousand cubic feet mcf thousand pounds kip Amendment 61 Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-2 ABBREVIATIONS USED IN THE FSAR thousand pounds per linear foot k/lf thousand pounds per square inch ksi ton (short ton) ton, st tonne (metric ton, 2,204.62 lb.) te, mt volt V volt alternating V AC volt ampere Va temperature of the hot leg Thot temperature of the cold leg Tcold average reactor coolant temperature Tavg vibrations per minute Vpm volt direct current V DC volts per phase per Hertz V/ph./Hz volume percent vol. percent inch water gage in. wg or wg watt W weight percent wt. percent wet bulb temperature wb yard yd.year yr.Amendment 61 Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.1.1-3 MAJOR BUILDINGS AND STRUCTURES Administration Building Auxiliary Boiler Fuel Oil Storage Tank Auxiliary Dam Auxiliary Reservoir Auxiliary Reservoir Channel Auxiliary Reservoir Separating Dike Auxiliary Transformer Concrete Containment Structure Containment Building: Containment Control Room Cooling Tower Cooling Tower Makeup Water Intake Channel Diesel Fuel Oil Storage Tank Building Diesel Generator Building Emergency Service Water and Cooling Tower Makeup Intake Structure Emergency Service Water Discharge Channel Emergency Service Water Discharge Structure Emergency Service Water Intake Channel Emergency Service Water Screening Structure Fuel Handling Building Fuel Handling Unloading Area Main Dam and Spillway Main Reservoir Main Transformer Makeup Water System Dikes Meteorological Tower Microwave Tower and Equipment House Normal Service Water Intake Structure Old Reactor Vessel Head Storage Facility Old Steam Generator Storage Facility Reactor Auxiliary Building Security Building Service Building Sewage Treatment Plant Startup Transformer Turbine Building Tank Building Warehouse Waste Processing Building Water Treatment Building 230 Kv Switchyard Hot Shop Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Hydraulic and Thermal (1 & 2) Design Parameters Total Core Heat Output, MWt 4.1 2775 2652 2775 6Total Core Heat Output, 10 Btu per hr. 4.1 9471 9051 9471.1 Heat Generated in Fuel, Percent 4.2 97.4 97.4 97.4 System Pressure, Nominal, psia 4.4 2250 2250 2250 System Pressure, Minimum Steady State, psia 4.4 2220 2220 2220 Minimum DNB Ratio at Nominal Initial 4.3 Rating Conditions Typical Cell 1.98 2.27 2.15 Thimble Cell 1.68 1.86 1.77 Minimum DNBR for Design Transients 4.3 >1.30 >1.30 >1.30 DNB Correlation 4.3 R (W-3 with modified R (W-3 with modified R (W-3 with modified Spacer Factor) Spacer Factor) Spacer Factor)Nuclear Enthalpy Rise Hot Channel Factor 4.3 1.55 1.55 1.55 Coolant Flow Total Flow Rate, 106 lb. per hr. 4.4 109.1 100.9 105.2 Effective Flow Rate for Heat Transfer, 106 lb. per hr. 4.4 102.5 96.3 100.5 Amendment 65 Page 1 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Effective Flow Area for Heat Transfer, ft.2 4.4 41.6 41.5 41.5 Average Velocity Along Fuel Rods, ft. per sec 4.4 15.6 14.5 15.1 Average Mass Velocity, 106 lb. per hr.-ft2 4.4 2.47 2.32 2.42 Coolant Temperatures F at 100 Percent Power 4.4 Nominal Inlet 556.0 542.5 546.8 Average Rise in Vessel 62.9 76.4 67.0 Average Rise in Core 66.4 70.3 67.8 Average in Core (based on enthalpy) 592.6 579.3 583.4 Average in Vessel (based on enthalpy) 588.3 576.2 580.3 Heat Transfer at 100 Percent Power 4.4 Active Heat Transfer Surface Area, ft2 48,600 48,700 48,600 Average Heat Flux, Btu per hr.-ft2 189,800 181,000 189.800 Maximum Heat Flux, Btu per hr.-ft2 440,400 434,500 440,500 Average Thermal Output, kw per ft. 5.44 (4) 5.2 5.44 (3)Maximum Thermal Output, kw per ft. 12.6 12.5 12.6 Fuel Central Temperature, F 4.4 Maximum at 100 Percent Power 3250 3150 3250 Maximum at Overpower 4700 4400 4150 Core Mechanical Design Parameters 4.2 Fuel Assemblies Number of Fuel Assemblies 157 157 157 Design RCC Canless RCC Canless RCC Canless 17 x 17 17 x 17 17 x 17 Rod Pitch, in. 0.496 0.496 0.496 Overall Dimensions, in. 8.426 x 8.426 8.426 x 8.426 8.426 x 8.426 Amendment 65 Page 2 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Fuel Weight (as UO2), lb. 181,190 181,205 181,205 Clad Weight, lb. 41,415 38,230 38,230 Number of Grids per Assembly(5) 8-Type R 8-Type R 8-Type R Fuel Rods 4.2 UO2 Rods per Assembly 264 264 264 Number 41,448 41,448 41,448 Outside Diameter, in. 0.374 0.374 0.374 Diametrical Gap, in. 0.0065 0.0065 0.0065 Clad Thickness, in. 0.0225 0.0225 0.0225 Clad Material Zircaloy-4 or M5 Zircaloy-4 Zircaloy-4 Fuel Pellets 4.2 Material UO2 Sintered UO2 Sintered UO2 Sintered Density (percent of Theoretical) 95 95 95 Diameter, in. 0.3225 0.3225 0.3225 Length, in. (3) 0.530 0.60 0.530 Rod Cluster Control Assemblies 4.3 Neutron Absorber Material 5% Cd-15% 5% Cd-15%In-80% Ag or 100% Hf In-80% Ag Cladding Material Type 304 Type 304 SS-Cold Worked SS-Cold Worked Clad Thickness, in. 0.0185 0.0185 Number of RCC Assemblies (Full/Part Length) 52/0 48/5 Number of Absorber Rods per RCC Assembly 24 24 24 Core Structure 4.4 Amendment 65 Page 3 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Core Barrel I.D./ O.D., in. 133.9/137.9 133.9/137.9 133.9/137.9 Thermal Shield Neutron Pads Neutron Pads Neutron Pads Nuclear Design Data Structural Characteristics Core Diameter, in. (Equivalent) 4.2 119.7 119.7 119.7 Core Height, in. (Active Fuel) 4.2 144 144 144 Reflector Thickness and Composition 4.5 Top - Water plus steel ~10 in. ~10 in. ~10 in.Bottom - Water plus steel ~10 in. ~10 in. ~10 in.Side - Water plus steel ~15 in. ~15 in. ~15 in.H2O/U, Molecular Ratio (lattice, cold) 2.4 2.4 2.4 Performance Characteristics Loading Technique 4.3 3 region, nonuniform 3 region, nonuniform 3 region, nonuniform Power Density, kw per liter of core 104.5 99.9 104.5 Specific Power, kw per kg UO2 38.4(6) 36.6 38.3 Feed Enrichments, w/o 2.10 2.10 2.10 Region 1 2.60 2.60 2.60 Region 2 3.10 3.10 3.10 Region 3 Amendment 65 Page 4 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Reactor Coolant System-Code Requirements Component 5.2 Reactor Vessel ASME III ASME III ASME III Class 1 & 2 Class 1 & 2 Class 1 & 2 Steam Generator ASME III Class 1 ASME III Class 1 ASME III Class 1 Tube Side Shell Side ASME III Class 2 (7) ASME Class 2 ASME Class 2 Pressurizer ASME III Class 1 ASME III Class 1 ASME III Class 1 Pressurizer Relief Tank ASME VIII ASME III Class 3 ASME Class 3 Pressurizer Safety Valves 5.2 ASME III ASME III ASME III Reactor Coolant Piping 5.2 ASME III Class 1 USAS B31.1 USAS B31.1 Principal Design Parameters of the Reactor Coolant System (100% Power)NSSS Heat Output MWt 5.1 2785 2660 2785 6NSSS Heat Output, Btu per hr. 5.1 9505 x 10 9078 x 106 9503 x 106 Operating Pressure, psi gage 5.1 2235 2235 2235 Vessel Inlet Temperature 5.1 557.4 542.5 546.8 Amendment 65 Page 5 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Vessel Outlet Temperature 5.1 620.2 610.9 613.7 Number of Loops 5.1 3 3 3 Design Pressure, psi gage 5.1 2485 2485 2485 Design Temperature, F 5.1 650 650 650 Hydrostatic Test Pressure (Cold), psi gage 5.1 3107 3107 3107 Reactor Coolant System Volume, Including Total pressurizer, ft.3 5.1 8963 9458 9438 Total Reactor Flow, gpm 5.1 292,000 265,500 278,400 Reactor Design Parameters of the Reactor Vessel Material (Vessels) 5.3 Shell & Head Plates: ASME Section II SA 32 ASME Section II SA ASME Section II SA- Grade B, low alloy 302 Grade B, low alloy 533 Grade A, B, or C steel, internally clad steel, internal clad with Class 1 or 2, Shell & with Type 304 Type 304 austenitic Nozzle Forgings: SA austenitic stainless stainless steel 508 Class 2 or 3. steel Cladding: Type 304 stainless steel or equivalent and Inconel Design Pressure, psi gage 5.3 2485 2485 2485 Design Temperature, F 5.3 650 650 650 Operating Pressure, psi gage 5.3 2235 2235 2235 Inside Diameter of Shell, in. 5.3 157 157 157 Overall Height of Vessel/Enclosure Head, ft.-in. 5.3 42-8 40-5 42-7-3/16 Amendment 65 Page 6 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Minimum Clad Thickness, in. 5.3 1/8 5/32 5/32 Principal Design Parameters of the Steam Generators Number of Units 5.4 3 3 3 Type 5.4 Vertical U-Tube with Vertical U-Tube with Vertical U-Tube with integral moisture integral moisture integral moisture separator separator separator Tube Material 5.4 Inconel Inconel Inconel Shell Material 5.4 Carbon Steel Carbon Steel Carbon Steel Tube Side Design Pressure, psi gage 5.4 2485 2485 2485 Tube Side Design Temperature, F 5.4 650 650 650 6Tube Side Design Flow, lb. per hr. 5.4 36.4 x 10 33.6 x 106 35.03 x 106 Shell Side Design Pressure, psi gage 5.4 1185 1085 1085 Shell Side Design Temperature, F 5.4 600 600 600 Operating Pressure, Tube Side, Nominal, psi gage 5.4 2235 2235 2235 Operating Pressure, Shell Side, Maximum, psi gage 5.4 1091 1005 1100 Maximum Moisture at Outlet at Full Load, percent 5.4 0.25 0.25 0.25 Hydrostatic Test Pressure, Tube Side (cold), psi gage 5.4 3107 3107 3107 Principal Design Parameters of the Reactor Coolant Pumps Number of Units 5.4 3 3 3 Amendment 65 Page 7 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Type 5.4 Vertical, single mixed Vertical, single stage Vertical, single stage flow mixed flow with bottom mixed flow with bottom suction and horizontal suction and horizontal discharge discharge Design Pressure, psi gage 5.4 2485 2485 2485 Design Temperature, F 5.4 650 650 650 Operating Pressure, Nominal, psi gage 5.4 2235 2235 2235 Suction Temperature, F 5.4 557.1 549 546.5 Design Capacity, gpm 5.4 96,600 88,500 92,800 Design Head, ft. 5.4 300 280 312 Hydrostatic Test Pressure (Cold), psi gage 5.4 3107 3107 3107 Motor Type 5.4 AC Induction single AC Induction single AC Induction single speed speed speed Motor Rating 5.4 7000 HP 6000 HP 7000 HP Principal Design Parameters of the Reactor Coolant Piping Material 5.4 Austenitic SS Austenitic SS Austenitic SS Hot Leg - I.D., in. 5.4 29 29 29 Cold Leg - I.D., in. 5.4 27.5 27.5 27.5 Between Pump and Steam Generator I.D., in. 5.4 31 31 31 Design Pressure, psi gage 5.4 2485 2485 2485 Engineered Safety Features Number of High Head Safety Injection (Charging) Pumps 6.3 3 3 3 Number of Low Head Safety Injection Pumps 6.3 2 2 2 Containment Spray-Number of Pumps 6.5.2 2 2 2 Boron Injection Tanks - Number 9.3.5 1 1 1 Amendment 65 Page 8 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Refueling Water Storage Tanks - Number 6.3 1 1 1 Hydrogen Control Systems Recombiners 6.25 2 2 2 Containment System Parameters Type 3.8 Steel lined reinforced Subatmospheric Subatmospheric concrete Inside Diameter, ft. 3.8 130 126 126 Height ft. 3.8 225 185 191 Free Volume, 106 x ft.3 3.8 2.266 1.8 1.825 Design Pressure, psig 3.8 45 45 45 Concrete Thickness Vertical Wall ft.-in. 3.8 4-6 4-6 4-6 Dome ft.-in. 3.8 2-6 2-6 2-6 Containment Leak Rate Percent per day 3.8 0.1 0.1 0.1 Electrical Systems Transmission Lines 8.2 7-230 kV 4-345 kV 3-500 kV (for 2 units) 5-138 kV Main Transformer Number 8.2 3 3 3 Startup Transformers 8.2 2 2 3 Auxiliary Transformers 8.2 2 2 3 Emergency Diesel Generators 8.2 2 2 2 Unit Batteries (125V) 8.2 3 5 4 Radioactive Waste Management System Liquid Waste Processing System 11.2 Reactor Coolant Waste Holdup Tank Amendment 65 Page 9 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Number 1 2-5,000 gal. 2-5,000 gal.2-3,000 gal. 2-1,500 gal.Capacity Each (gal.) 25,000 2-3,000 gal. 2-1,400 gal.Spent Resin Storage Tank Number 4 1 Capacity (ft.3) 500 - 1,800 gal.Secondary Waste Concentrate Tank Number 2 2 2 Capacity (gal.) 4,000 7,500 5,000 Gaseous Waste Management System 11.3 Waste Gas Decay Tank Number 10 3 2 Design Pressure (psig) 150 100 145 Volume Each (ft.3) 600 743 3,400 gal.Waste Gas Compressors Number 2 - -Design Pressure 150 - -Catalytic Hydrogen Recombiners Number 2 5 1 Solid Waste Management System 11.4 Amendment 65 Page 10 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.1-1 DESIGN COMPARISON WITH SIMLAR FACILITIES (INITIAL FUEL CYCLE Item Reference Section Shearon Harris Beaver Valley North Anna Solidification Pretreatment Tank Pump Number 2 1 1 Capacity (gpm) 35 - 50 Solidification Pretreatment Tank Number 2 1 1-1,800 gal.Capacity (gpm) 5,000 - 1-500 gal.Instrumentation Systems*Reactor Protection System 7.2 7.2 7.2 7.2 Reactor and Reactor Coolant System 7.7 7.7 7.7 7.7 Steam and Feedwater Control System 7.7 7.7 7.7 7.7 Nuclear Instrumentation 7.2 7.2 7.2 7.2 Plant Process Display Instrumentation 7.5 7.5 7.5 7.5

*This section is not suited for tabular description. SAR section numbers have been included for the location of the detailed description of each system.

NOTES:

1) There is not single design value for Axial Offset. There is however, a target value with an allowable band width, within which the plant operates. Analyses have been performed and the results presented in Chapter 4, which demonstrate that operation within the bank width results in axial power distributions no more severe than that used in the design.

2) For all three plants, the design axial power shape for DNB evaluations is a chopped cosine with peak to average value of 1.55.

3) This limit is associated with the value of Fq=2.32.

4) This value is associated with a design value of Fq=2.1. Power monitoring at this value will be performed as stated in Section 7.6.

5) Reflects current approved as-built design which may not be changed in all references SARs.

6) The specific power for Shearon Harris was based upon 94.5 percent of theoretical density.

7) The shell side of the steam generator conforms to the requirements for ASME Class 1 vessels and is so stamped as permitted under the rules of Section III.

Amendment 65 Page 11 of 11

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.3.2-1*COMPARISON OF FINAL AND PRELIMINARY INFORMATION SIGNIFICANT DESIGN CHANGES Item Reference FSAR Description of Change Erosion of Hydraulic 2.4.5.5 The riprap protection on the east and south Structures edges of the plant island has been replaced by sacrificial spoil fill. Also, the auxiliary reservoir spillway has been widened.Main Dam Low Level Release 2.4.11 Added the low level release system for System maintaining the water quality of the reservoir discharge.Auxiliary Dam Diversion 2.5.6 There is no diversion conduit through the Channel Auxiliary Dam.Tunnel Conduit Turbine 3.2.1 The tunnel under the Turbine Building was Building upgraded to Seismic Category I, since this tunnel is housing safety-related components.Containment Penetrations 3.8.2.1 Changes were made on load combinations for the steel containment penetrations. These changes are deviations from SRP 3.8.2.Electrical Penetrations 3.8.2.1 A continuous supply of pressurized nitrogen and instrument air is provided to the electrical penetrations.Seismic Qualification of 3.10 All BOP electrical equipment previously Electrical Equipment qualified in accordance with IEEE 344-1971, has been updated to the requirements of IEEE 344-1975.Reactor Core 4 Part length rods are deleted.Integrated Reactor Vessel 9.4.8 The new configuration incorporates integral Head 4.6.1 CRDM cooling capability and lifting device.Containment Design 6.2.1, The containment liner design temperature was Temperature 3.8.1.3.1.i changed to 248 F, based on the latest containment pressure and temperature analyses. Also, the dome liner will not be used to resist seismic shear stresses, in accordance with ACI 359-74.ECCS 6.3 The switchover from injection to cold leg recirculation phase is done semi-automatically.Amendment 65 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 Item Reference FSAR Description of Change Fuel Handling System 9.1.4 The system has been updated to reflect the use of two overhead cranes. Also, the fuel storage capacity has been increased.Provisions to handle and store offsite spent fuel have been made.Cooling Tower Blowdown 9.2.1 Cooling tower blowdown now discharges into Discharge System the lake at a point approximately 3.5 miles south of the plant at about one mile north of the dam. The discharge has been changed from a multipoint diffuser to a single point jet. This change reduces circulation of the cooling tower blowdown discharge into the plant makeup structure.Tornado Protection 9.4 Added tornado protection dampers to selected outside air intakes.Steam Generator Blowdown 10.4.8 The changes in the Steam Generator System Blowdown System were made to upgrade the system to Safety Class 3 requirements in order to minimize pipe rupture considerations.

• THIS TABLE IS FOR HISTORICAL INFORMATION ONLY.

Amendment 65 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Eggleston, F. T., "Safety-Related Research and Development for 4.3 B Westinghouse Pressurized Water Reactors, Program Summaries -Winter 1977, Summer 1978" WCAP-8768, Revision 1, October, 1978.TP-04124 Missile Probability Analysis for the Siemens 13 9m2 3.5 A Retrofit Design of Low-Pressure Turbine by Siemens AG, Submitted to the Nuclear Regulatory Commission as Topical Report TP-04124-NP-A, For Public Record, June 7, 2004, Siemens Westinghouse Power Corporation.

 "Pipe Breaks for the LOCA Analysis of the Westinghouse Primary 3.6 A Coolant Loop," WCAP-8082 P-A (Proprietary) and WCAP-8172-A (Non-Proprietary), January 1975. "Damping Values of Nuclear Plant Components," WCAP-7921-AR, 3.7 A May 1974. "Safety Analysis of the 8-Grid 17 x 17 Fuel Assembly for Combined 3.7, 4.2 A Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March, 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April, 1974. "Documentation of Selected Westinghouse Structural Analysis 3.9 U Computer Codes," WCAP-8252, Provision 1, May 1977. "Benchmark Problem Solution Employed for Verification of the 3.9 U WECAN Computer Program,"WCAP-8929, June 1977.

Takeuchi, K., et al., "Multiflex - A Fortran-IV Computer Program for 3.9 A Analyzing Thermal-Hydraulic- Structure System Dynamics" WCAP-8709 (February 1976).Witt, F. J., Bamford, W. H., Esselman, T. C., "Integrity of the Primary 3.9 U Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," WCAP-9288, March 1978.Bogard, W. T., Esselman, T. C., "Combination of Safe Shutdown 3.9 U Earthquake and Loss-Of-Coolant Accident Responses for Faulted Condition Evaluation of Nuclear Power Plants," WCAP-9279, March 1978.Amendment 63 Page 1 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Bloyd, C. N., Ciarametaro, W. and Singleton, N. R., "Verification of 3.9 A Neutron Pad and 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976.

 "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale 3.9 A Model Tests," WCAP-8303-P-A (Proprietary) and WCAP-8317-A (Non-Proprietary), July, 1975.

Bloyd, C. N. and Singleton, N. R. "UHI Plant Internals Vibration 3.9 A Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary) April, 1975.Cooper, F. W. Jr., "17 x 17 Drive Line Components Tests - Phase 1B 3.9 A 11, 111, D-Loop-Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December, 1974.Kraus, S., "Neutron Shielding Pads," WCAP-7870, May, 1972. 3.9 A Butterworth, G. and Miller, R. B., Methodology for Qualifying 3.10 U Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," WCAP-8587, Revision 2, February 1979.

 "Equipment Qualification Data Packages," Supplement 1 to WCAP- 3.10 U 8587, November 1978.

Morrone, A, "Seismic Vibration Testing With Sine Beats," WCAP- 3.10 U 7558, October 1971.Jarecki, S. J., "General Method of Developing Multi-frequency Biaxial 3.10 U Test Inputs for Bistables," WCAP-8624 (Proprietary), September 1975 and WCAP-8695 (Non-Proprietary), August 1975.Butterworth, G. and Miller, R. B., "Methodology for Qualifying 3.11 U Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment," WCAP-8587, Revision 2, February 1979.

 "Equipment Qualification Data Packages, Supplement 1 to WCAP- 3.11 U 8587, November 1978.

Amendment 63 Page 2 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Locante, J., "Environmental Testing of Engineered Safety Features 3.11 A Related Equipment," Vols. 1 & 2, WCAP-7410-L, December, 1970 (Westinghouse Proprietary Class II). Non-proprietary versions WCAP-7744, Vol. 1, August 1971, Vol. 2, January 1972.Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and 4.1, 4.2 A Model and Reactor Application," WCAP-8218-P-A (Proprietary) and 4.3, 4.4 WCAP-8219-A (Non-Proprietary), March, 1975.Skaritka, J. and Iorii, J. A., "Operational Experience with 4.2 B Westinghouse Cores," (Up to December 31, 1977)," WCAP-8183, Revision 8, August 1978.Beaumont, M. D., et al., "Properites of Fuel and Core Component 4.2 U Materials," WCAP-9179, Revision 1 (Proprietary) and WCAP-9224 (Non Proprietary), July 1978.Christensen, J. A., Allio, R. J. and Biancheria, A., "Melting Point of 4.2, 4.4 O Irradiated UO2," WCAP-6065, February 1965.Miller, J. V. (Ed.), "Improved Analytical Models Used in Westinghouse 4.2 A Fuel Rod Design Computations," WCAP-8720 (Proprietary) and WCAP-8785 (Non-Proprietary), October 1976.George, R. A., Lee, Y. C. and Eng, G. H., "Revised Clad Flattening 4.2 A Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary),July 1974.Risher, D., et al., "Safety Analysis for the Revised Fuel Rod Internal 4.2 A Pressure Design Basis," WCAP-8963 (Proprietary), November 1976 and WCAP-8964 (Non-Proprietary), August 1977.Eggleston, F. T., "Safety-Related Research and Development for 4.2 B Westinghouse Pressurized Water Reactors, Program Summaries -Winter 1977 - Summer 1978," WCAP-8768, Revision 2, October 1978.Demario, E. E., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," 4.2, 4.4 U WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary),February 1974.Reavis, J. R., et al., "Fuel Rod Bowing," WCAP-8691 (Proprietary) and 4.2 B WCAP-8692 (Non Proprietary), December 1975.Amendment 63 Page 3 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, 4.2 A Revision 4-A, March 1975.Westinghouse Anticipated Transients Without Reactor Trip Analysis," 4.3, 4.6, 15.1, O WCAP-8330, August, 1974. 15.2 15.4, 15.8 Langford, F. L. and Nath, R. J., "Evaluation of Nuclear Hot Channel 4.3 U Factor Uncertainties," WCAP-7308-L (Proprietary) and WCAP-7810 (Non Proprietary), December, 1971.Hellman, J. M. and Yang, J. W., "Effects of Fuel Densification Power 4.3 A Spikes on Clad Thermal Transients," WCAP-8359, July, 1974.Moore, J. S., "Power Distribution Control of Westinghouse Pressurized 4.3 O Water Reactors," WCAP-7811, December, 1971.Morita, T., et al. "Power Distribution Control and Load Follow 4.3, 4.4 A Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Non-Proprietary), September, 1974.McFarlane, A. F., "Power Peaking Factors, WCAP-7912-P-A 4.3, 4.4 A (Proprietary) and WCAP-7912-A (Non Proprietary), January, 1975.Meyer, C. E. and Stover, R. L., "Incore Power Distribution 4.3 U Determination in Westinghouse Pressurized Water Reactors," WCAP-8498, July, 1975.Barry, R. F. and Altomare, S., "The TURTLE 24.0 Diffusion Depletion 4.3, 15.4 A Code," WCAP-7213-P-A (Proprietary) and WCAP-7758-A (Non-Proprietary), January, 1975.Cermak, J. O., et al., "Pressurized Water Reactor pH - Reactivity 4.3 O Effect Final Report," WCAP-3696-8 (EURAEC-2074), October, 1968.Dominick, I. E. and Orr, W. L., "Experimental Verification of Wet Fuel 4.3 B Storage Criticality Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Non-Proprietary), December, 1975.Poncelet, C. G. and Christie, A. M., "Xenon-Induced Spatial 4.3 O Instabilities in Large PWRs," WCAP-3680-20 (EURAEC-1974), March, 1968.Amendment 63 Page 4 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Skogen, F. B. and McFarlane, A. F., "Control Procedures for Xenon- 4.3 O Induced X-Y Instabilities in Large PWRs," WCAP-3680-21 (EURAEC-2111), February, 1969.Skogen, F. B. and McFarlane, A. F., "Xenon-Induced Spatial 4.3 O Instabilities in Three-Dimensions," WCAP-3680-22 (EURAEC-2116),September 1969.Lee, J. C., et at., Axial Xenon Transient Tests at the Rochester Gas 4.3 B and Electric Reactor," WCAP-7964, June, 1971.Barry, R. F., et al., "The PANDA Code," WCAP-7048-P-A (Proprietary) 4.3 B and WCAP-7757-A (Non Proprietary), January, 1975.Barry, R. F., "LEOPARD - A Spectrum Dependent Non-Spatial 4.3, 15.4 O Depletion Code for the IBM-7094," WCAP-3269-26, September, 1963.Poncelet, C. G., "LASER - A Depletion Program for Lattice 4.3 O Calculations Based on MUFT and THERMOS," WCAP-6073, April, 1966.Olhoeft, J. E., "The Doppler Effect for a Non-Uniform Temperature 4.3 O Distribution in Reactor Fuel Elements," WCAP-2048, July, 1962.Nodvik, R. J., et al., "Supplementary Report on Evaluation of Mass 4.3 O Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium,"WCAP-6086, August, 1969.Moore, J. S., "Nuclear Design of Westinghouse Pressurized Water 4.3 B Reactors with Burnable Poison Rods," WCAP-7806, December, 1971.Nodvik, R. J., "Saxton Core II Fuel Performance Evaluation," WCAP- 4.3, 4.4 O 3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," July, 1970.Leamer, R. D., et al., "PUO2-UO2 Fueled Critical Experiments," 4.3 O WCAP-3726-1, July, 1967.Camden, T. M. et al., "PALADON - Westinghouse Nodal Computer 4.3 U Code," WCAP-9485A (Proprietary) and WCAP-9486A (Non-Proprietary), December, 1978.Amendment 63 Page 5 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status "Augmented Startup and Cycle 1 Physics Program Supplement 1," 4.3 U WCAP-8575 Supplement 1, June 1976 (Westinghouse Proprietary) and WCAP-8576, June 1976 (Non-Proprietary).Motley, F. E., Wenzel, A. H. and Cadek, F. F., "Critical Heat Flux 4.4 A Testing of 17 x 17 Fuel Assembly Geometry with 22-Inch Grid Spacing," WCAP-8536 (Proprietary), May, 1975 and WCAP-8537, May, 1975.Hill, K. W., Motley, F. E. and Cadek, F. F., "Effect of Local Heat Flux 4.4 A Spikes on DNB in Non-Uniform Heated Rod Bundles," WCAP-8174, August, 1973 (Proprietary) and WCAP-8202, August, 1973 (Non-Proprietary).F. E. Motley, F. F. Cadek, "Application of Modified Spacer Factor to L 4.4 A Grid Typical and Cold Wall Cell DNB," WCAP-7988 (Westinghouse Proprietary), and WCAP-8030-A (Non-Proprietary), October, 1972.Chelemer, H., Weisman, J. and Tong, L. S., "Sub-channel Thermal 4.4 O Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January, 1969.Motley, F. E. and Cadek, F. F., "DNB Tests Results for New Mixing 4.4 A Vane Grids (R)," WCAP-7695-P-A (Proprietary), January, 1975 and WCAP-7958-A, January, 1975.Motley, F. E. and Cadek, F. F., "DNB Test Results for R Grid Thimble 4.4 A Cold Wall Cells," WCAP-7695 Addendum 1-P-A (Proprietary),January, 1975 and WCAP-7958, Addendum 1-A, January, 1975.Hill, K. W. Motley, F. E., Cadek, F. F. , Wenzel, A. H., "Effect of 17 x 4.4 A 17 Fuel Assembly Geometry on DNB," WCAP-8296-P-A (Westinghouse Proprietary) and WCAP-8297-A (Non-Proprietary),February, 1975.Cadek, F. F., Motley, F. E. and Dominicis, D. P., "Effect of Axial 4.4 A Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid, WCAP-7941-P-A (Proprietary), January, 1975 and WCAP-7959-A January, 1975.Motley, F. E., Wenzel, A. H., Cadek, F. F., "The Effect of 17 x 17 Fuel 4.4 A Assembly Geometry on Interchannel Thermal Mixing," WCAP-8298-P-A (Proprietary), January, 1975 and WCAP-8299-A, January, 1975.Amendment 63 Page 6 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Cadek, F. F., "Interchannel Thermal Mixing with Mixing Vane Grids," 4.4 A WCAP-7667-P-A (Proprietary), January, 1975 and WCAP-7755-A, January, 1975.Hochreiter, L. E., and Chelemer, H., "Application of the THINC IV 4.4 A Program to PWR Design," WCAP-8054 (Proprietary), October, 1973, and WCAP-8195 (Non Proprietary), September, 1973.Hetsroni, G., "Hydraulic Tests of the San Onofre Reactor Model," 4.4 O WCAP-3269-8, June, 1964.Balfour, M. G., Christensen, J. A. and Ferrari, H. M., "In-Pile 4.4 O Measurement of UO2 Thermal Conductivity," WCAP-2923, March 1966.Hochreiter, L. E., Chelemer, H. and Chu, P. T., "THINC-IV An 4.4 A Improved Program for Thermal- Hydraulic Analysis of Rod Bundle Cores," WCAP-7956, June, 1973.Poncelet, C. G., "Burnup Physics of Heterogeneous Reactor Lattices," 4.4 B WCAP-6069, June, 1965.Carter, F. D., "Inlet Orificing of Open PWR Cores," WCAP-9004, 4.4 B January, 1969 (Proprietary) and WCAP-7836, January, 1972 (Non-Proprietary).Shefcheck, J. "Application of the THINC Program to PWR Design," 4.4 B WCAP-7359-L (Proprietary), August, 1969 and WCAP-7838, January, 1972.Novendstern, E. H. and Sandberg, R. O., "Single Phase Local Boiling 4.4 O and Bulk Boiling Pressure Drop Correlations," WCAP-2850-L (Proprietary) and WCAP-7916 (Non-Proprietary), April, 1966.Burke, T. M., Meyer, C. E. and Shefcheck J., "Analysis of Data from 4.4 A the Zion (Unit 1) THINC Verification Test," WCAP-8453, May 1976.Shopsky, W. E., "Failure Mode and Effects Analysis (FMEA) of the 4.6 U Solid State Full Length Rod Control System," WCAP-8976, August, 1977.Amendment 63 Page 7 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Gangloff, W. C. and Loftus, W. D., "An Evaluation of Solid State Logic 4.6, 7.1, 7.2 U Reactor Protection in Anticipated Transients," WCAP-7706-L (Proprietary) and WCAP-7706 (Non-Proprietary), July, 1971.Eggleston, F. T., Rawlins, D. H. and Petrow, J. R., "Failure Mode and 4.6 U Effects Analysis (FMEA) of the Engineering Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), April, 1976.Cooper, L., Miselis, V. and Starek, R. M., "Overpressure Protection for 5.2 U Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June, 1972 (also letter NS-CE-622, dated April 16, 1975, C.Eicheldinger (Westinghouse) to D. B. Vassallo (NRC), additional information on WCAP-7769, Revision 1).Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907- 5.2, 15.1 A P-A, April, 1984. 15.2, 15.3 15.4, 15.6 "Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME 5.2 U SA-533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals," WCAP-9292, March, 1978.Golik, M. A., "Sensitized Stainless Steel in Westinghouse PWR 5.2 B Nuclear Steam Supply Systems," WCAP-7735, August, 1971.Enrietto, J. F., "Control of Delta Ferrite in Austenitic Stainless Steel 5.2 A Weldments," WCAP-8324-A, June, 1974.Buchalet, C. and Mager, T. R., " A Summary Analysis of the April 30 5.3 B Incident at the San Onofre Nuclear Generating Station Unit 1", WCAP-8099, April 1973.

 "Reactor Coolant Pump Integrity in LOCA," WCAP-8163, September, 5.4 U 1973.

Whyte, D. D. and Picone, L. F., "Behavior of Austenitic Stainless Steel 6.1 B in Post Hypothetical Loss-of-Coolant Accident Environment," WCAP-7798 L (Proprietary), November, 1971 and WCAP-7803 (Non Proprietary), December, 1971.Amendment 63 Page 8 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Picone, L. F., "Evaluation of Protective Coatings for Use in Reactor 6.1 B Containment," WCAP-7198-L (Proprietary), April, 1968 and WCAP-7825 (Non-Proprietary), December, 1971.Geets, J. M., "MARVEL, A Digital Computer Code for Transient 6.3 B Analysis of a Multiloop PWR System," WCAP-7909, June, 1972.Marasco, F. W. and Siroky, R. M., "Westinghouse 7300 Series 7.1 A Process Control System Noise Tests," WCAP-8892-A, June, 1977.Reid, J. B., "Process Instrumentation for Westinghouse Nuclear 7.2, 7.3 B Steam Supply Systems," WCAP-7913, January, 1973.Lipchak, J. B., "Nuclear Instrumentation System," WCAP-8255, 7.2, 7.7 B January, 1974.Katz, D. N., "Solid State Logic Protection System Description," WCAP- 7.2, 7.3 A 7488-L (Proprietary), March, 1971 and WCAP-7672 (Non-Proprietary),May, 1971.Swogger, J. W., "Testing of Engineered Safety Features Actuation 7.3 B System," WCAP-7705, Revision 2, January, 1976. (Information only; i.e., not a generic topical WCAP).WCAP-9230, Report on the Consequences of a Postulated Main 15.2 U Feedline Rupture, January 1978.Hargrove, H. G., "FACTRAN - A Fortran-IV Code for Thermal 15.3, 15.4 U Transients in a UO2 Fuel Rod," WCAP-7908, June 1972.Balwin, M. S., Merrian, M. M., Schenkel, H. S. and Van DeWalle, D. J., 15.3 U "An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975.Risher, D. H., Jr. and Barry, R. F., "TWINKLE - A Multi-Dimensional 15.4 A Neutron Kinetics Computer Code," WCAP-7979-A (Proprietary) and WCAP-8028-7 (Non-Proprietary), January 1975.Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in 15.4 A Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.Amendment 63 Page 9 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Bordelon, F. M., Massie, H. W. and Zordan T. A., "Westinghouse 15.6 A ECCS Evaluation Model - Summary," WCAP-8339, July 1974.Bordelon, F. M., et al., "SATAN-VI Program: Comprehensive Space 15.6 A Time Dependent Analysis of Loss-of-Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974.Kelly, R. D., et at., "Calculational Model for Core Reflooding After a 15.6 A Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974.Bordelon, F. M. and Murphy, E. T., "Containment Pressure Analysis 15.6 A Code (COCO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974.Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Transient 15.6 A Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.Bordelon, F. M., et al., "Westinghouse ECCS Evaluation Model - 15.6 A Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non Proprietary), April 1975.

 "Westinghouse ECCS Evaluation Model - October 1975 Version," 15.6 A WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),

November 1975.Esposito, V. J., Kesavan, K. and Maul, B. A., "WFLASH, A FORTRAN- 15.6 A IV Computer Program for Simulation of Transients in Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974.Skwarek, R., Johnson, W., Meyer, P., "Westinghouse Emergency 15.6 A Core Cooling System Small Break October 1975 Model", WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary), April, 1977.Kelly, R. D., Thompson, C. M., et. al., "Westinghouse Emergency 15.6 U Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation with One Loop Out of Service for Plants Without Loop Isolation Valves," WCAP-9166, February 1978.Amendment 63 Page 10 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status Eicheldinger, C., "Westinghouse ECCS Evaluation Model, February 15.6 A 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-P-A (Non-Proprietary Version), February 1978.

 "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP- 15.6 A 8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July, 1974.

Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies," 15.6 A WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July, 1974.

 "Design Considerations for the Protection from the Effects of Pipe 3.6 U Rupture," ETR-1002, November, 1975. "Design Considerations for the Protection from the Effects of Pipe 3.6 U Rupture," ETR-1002P (Proprietary), November, 1975.

Babco*ck and Wilcox, H. W. Behnre, et al., "Methods of Compliance 5.3 A with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G", BAW-10046A, Revision 2, June 1986.J.R. Worsham III, "Fluence and Uncertainity Methodologies," BAW- 5.3 A 2241 P-A, Revision 1, Framatome Technologies, Inc., Lynchburg, Virginia, April 1999.

 "SGTR Analysis Methodology to Determine the Margin to Steam 15.6.3 A Generator Overfill," WAP-10698-P-A (Proprietary), August 1987 "Evaluation of Offsite Radiation Doses for a Steam Generator Tube 15.6.3 A Rupture Accident," Supplement 1 to WCAP-10698-P-A, March 1986 (Proprietary). "Streamline Break Mass/Energy Releases for Equipment 3.11 A Environmental Qualification Outside Containment, Report to the Westinghouse Owners Group High Energy Line Break/Superheated Blowdowns Outside Containment Subgroup," WCAP-10961, Rev. 1, (Proprietary), October 1985 "Mass and Energy Releases Following a Steam Line Rupture," 6.2 A WCAP-8822 (Proprietary) and WCAP-8860 (Nonproprietary),

September 1976.Amendment 63 Page 11 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE Review Report Reference(s) Status "Supplement 1 - Calculations of Steam Superheat in Mass/Energy 6.2 A Releases Following a Steam Line Rupture," WCAP-8822-S1-P-A (Proprietary) and WCAP-8860-S1-A (Nonproprietary), September 1986.

 "Supplement 2 - Impact of Steam Superheat in Mass/Energy Releases 6.2 A Following a Steam Line Rupture for Dry and Subatomospheric Containment Designs," WCAP-8822-S2-P-A (Proprietary and WCAP-8860-S2-A (Nonproprietary), September, 1986. "Westinghouse LOCA Mass & Energy Release Model for Containment 6.2 A Design - March 1979 Version," WCAP-10325-P-A, (Proprietary),

WCAP-10326-A (Nonproprietary), May 1983.

 "Westinghouse Mass and Energy Release Data for Containment 6.2 A Design," WCAP-8264-P-A, Rev. 1 (Proprietary), WCAP-8312-A (Nonproprietary), August 1975.

Topical Report DOM-NAF-3-NP-A, "GOTHIC Methodology for 6.2 A Analyzing the Response to Postulated Pipe Ruptures Inside Containment", Virginia Electric and Power Company (Dominion) et.al.,November 2006.Amendment 63 Page 12 of 12

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-2 OTHER REPORTS INCORPORATED BY REFERENCE Report Reference Sections Ebasco Services, Inc., 1975, "Fault Investigation," Shearon Harris Nuclear Power Plant 2.5 Units 1, 2, 3, 4, Vol.2 (Appendices)Ebasco Services, Inc., 1981, "Final Geologic Report on Foundation Conditions, Power Plant, 2.5E Dams, and Related Structures."Ebasco Services, Inc., 1981, "Final Embankment Report for the Main Dam, Auxiliary Dam, 2.5F and Auxiliary Separating Dike."Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 1.1.1-1A FLOW DIAGRAM LEGEND-PIPING AND INSTRUMENTATION SYMBOLS 5-G-0041 1.2.2-1 SITE PLAN 5-G-0003 1.2.2-2 PLOT PLAN 5-G-0002 1.2.2-3 GENERAL ARRANGEMENT-CONTAINMENT BUILDING PLAN EL. 221.00' & 236.00' UNIT 1 5-G-0011 1.2.2-7 GENERAL ARRANGEMENT-CONTAINMENT BUILDING PLAN EL. 261.00' & 286.00' UNIT 1 5-G-0012 1.2.2-11 GENERAL ARRANGEMENT-CONTAINMENT BUILDING SECTIONS-SHEET 1 5-G-0013 1.2.2-15 GENERAL ARRANGEMENT-CONTAINMENT BUILDING SECTIONS-SHEET 2 5-G-0014 1.2.2-19 GENERAL ARRANGEMENT-CONTAINMENT AUXILIARY BUILDING PLAN EL. 190.00' & 216.00' 5-G-0015 1.2.2-23 GENERAL ARRANGEMENT-CONTAINMENT GENERAL ARRANGEMENT REACTOR EL. 236.00' 5-G-0016 1.2.2-27 GENERAL ARRANGEMENT-CONTAINMENT AUXILIARY BUILDING EL. 261.00' 5-G-0017 1.2.2-31 GENERAL ARRANGEMENT-REACTOR AUXILIARY BUILDING PLAN EL. 286.00' 5-G-0018 1.2.2-35 GENERAL ARRANGEMENT-REACTOR AUXILIARY PLAN EL. 305.00' 5-G-0019 1.2.2-39 GENERAL ARRANGEMENT-REACTOR AUXILIARY BUILDING SECTIONS-SHEET 1 5-G-0020 1.2.2-43 GENERAL ARRANGEMENT-REACTOR AUXILIARY BUILDING SECTIONS-SHEET 2 5-G-0021 1.2.2-47 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING PLAN AT EL. 211.00' & 216.00' 5-G-0910 1.2.2-48 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING PLAN AT EL. 236.00' 5-G-0911 1.2.2-49 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING PLAN AT EL. 261.00' 5-G-0912 1.2.2-50 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING PLAN AT EL. 276.00' 5-G-0913 1.2.2-51 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING PLAN AT EL. 286.00' & EL. 291.00' 5-G-0914 1.2.2-52 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING SECTIONS-SHEET 1 5-G-0916 1.2.2-53 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING SECTIONS-SHEET 2 5-G-0917 1.2.2-54 GENERAL ARRANGEMENT-WASTE PROCESSING BUILDING SECTIONS-SHEET 3 5-G-0918 1.2.2-55 GENERAL ARRANGEMENT-FUEL HANDLING BUILDING PLANS-SHEET 1 5-G-0022 1.2.2-56 GENERAL ARRANGEMENT-FUEL HANDLING BUILDING PLANS-SHEET 1 5-G-0023 1.2.2-57 GENERAL ARRANGEMENT-FUEL HANDLING- BUILDING SECTIONS 5-G-0024 1.2.2-58 GENERAL ARRANGEMENT-FUEL HANDLING BUILDING SECTIONS 5-G-0025 1.2.2-59 GENERAL ARRANGEMENT-FUEL HANDLING BUILDING SECTIONS 5-G-0026 1.2.2-59A GENERAL ARRANGEMENT-FUEL HANDLING BUILDING MISCELLANEOUS PLANS 5-G-0038 1.2.2-60 GENERAL ARRANGEMENT-TURBINE BUILDING PLAN EL. 240.00' 5-G-0004 1.2.2-64 GENERAL ARRANGEMENT-TURBINE BUILDING GROUND FLOOR PLAN 5-G-0005 1.2.2-68 GENERAL ARRANGEMENT-TURBINE BUILDING MEZZANINE FLOOR PLAN 5-G-0006 1.2.2-72 GENERAL ARRANGEMENT-TURBINE BUILDING OPERATION FLOOR PLAN 5-G-0007 1.2.2-76 GENERAL ARRANGEMENT-TURBINE BUILDING SECTIONS-SHEET 1 5-G-0008 Amendment 62 Page 1 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 1.2.2-80 GENERAL ARRANGEMENT-TURBINE BUILDING SECTIONS-SHEET 2 5-G-0009 1.2.2-84 GENERAL ARRANGEMENT-TANK AREA PLANS AND SECTIONS 5-G-0033 1.2.2-86 GENERAL ARRANGEMENT-DIESEL GENERATOR BUILDING PLANS 5-G-0036 S01 1.2.2-87 GENERAL ARRANGEMENT-DIESEL GENERATOR BUILDING SECTIONS 5-G-0036 S02 2.5.6-1 MAIN DAM GENERAL PLAN 7-G-6240 2.5.6-2 MAIN DAM PROFILE & TYPICAL SECTION 7-G-6241 2.5.6-3 RESERVOIR-WEST AUXILIARY DAM GENERAL PLAN 7-G-6270 2.5.6-4 AUXILIARY DAM-PROFILE AND TYPICAL SECTIONS 7-G-6271 2.5.6-5 AUXILIARY SEPARATIONS DIKE GENERAL PLAN, PROFILE AND SECTIONS 7-G-6343 2.5.6-6 AUXILIARY RESERVOIR CHANNEL 7-G-6371 2.5.6-7 ESW-INTAKE CHANNEL PLAN, SECTIONS & DETAILS 7-G-2940 2.5.6-8 ESW-DISCHARGE CHANNEL PLAN, SECTIONS & DETAILS 7-G-2941 2.5.6-26 MAIN DAM-DIVERSION SYSTEM 7-G-6232 2.5.6-28 COOLING TOWER MAKEUP WATER INTAKE CHANNEL 7-G-2942 3.4.1-1 PLANT SITE PLAN 5-G-0003 REACTOR AND REACTOR AUXILIARY BLDG. BREAK AND RESTRAINT LOCATIONS AND JET 3.6A-1 SK-2165-MNE-R-0067 IMPINGEMENT ENVELOPES-MAIN STEAM PIPING PLANS MAIN STEAM PIPING-BREAK & RESTRAINT LOCATIONS & JET IMPINGEMENT ENVELOPES PLAN 3.6A-2 SK-2165-MNE-R-0068 SHEET 2 REACTOR AND REACTOR AUXILIARY BLDG. BREAK AND RESTRAINT LOCATIONS AND 3.6A-5 SK-2165-MNE-R-0071 IMPINGEMENT ENVELOPES-FEEDWATER PIPING PLANS 3.6A-6 TURBINE BUILDING-BREAK & RESTRAINT LOCATIONS FEEDWATER PIPING PLANS SK-2165-MNE-R-0072 REACTOR AND REACTOR AUXILIARY BLDG. BREAK AND RESTRAINT LOCATIONS AND JET 3.6A-7 SK-2165-MNE-R-0073 IMPINGEMENT ENVELOPES-FEEDWATER PIPING SECTIONS CONTAINMENT BUILDING AND TUNNEL AREA BREAK RESTRAINT LOCATIONS AND JET 3.6A-8 SK-2165-MNE-R-0074 IMPINGEMENT ENVELOPES- AUXILIARY FEEDWATER PIPING CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-9 SK-2165-MNE-R-0137 CHEMICAL AND VOLUME CONTROL PIPING PLAN CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-10 SK-2165-MNE-R-0138 CHEMICAL AND VOLUME CONTROL PIPING-PARTIAL CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-11 SK-2165-MNE-R-0139 CHEMICAL AND VOLUME CONTROL PIPING-SECTIONS AND DETAILS REACTOR AUXILIARY BLDG.-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT 3.6A-12 SK-2165-MNE-R-0140 ENVELOPES- CHEMICAL AND VOLUME CONTROL PIPING-PLANS REACTOR AUXILIARY BLDG.-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT 3.6A-13 SK-2165-MNE-R-0141 ENVELOPES- CHEMICAL AND VOLUME CONTROL PIPING SECTIONS CONTAINMENT BUILDING-BREAK & RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-14 SK-2165-MNE-R-0147 REACTOR COOLANT PIPING PLAN REACTOR AUXILIARY BLDG.-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT 3.6A-17 SK-2165-MNE-R-0151 ENVELOPES-RHR AND SAFETY INJECTION PIPING PLAN-EL. 190 REACTOR AUXILIARY BLDG.-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT 3.6A-18 SK-2165-MNE-R-0152 ENVELOPES-RHR AND SAFETY INJECTION PIPING PLAN - EL. 236 REACTOR AUXILIARY BLDG.-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT 3.6A-19 SK-2165-MNE-R-0153 ENVELOPES-RHR AND SAFETY INJECTION PIPING- PARTIAL PLANS AND SECTIONS CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-20 SK-2165-MNE-R-0154 RHR AND SAFETY INJECTION PIPING PLAN SHEET 1 Amendment 62 Page 2 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-21 SK-2165-MNE-R-0155 RHR AND SAFETY INJECTION PIPING PLAN- SHEET 2 CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-22 SK-2165-MNE-R-0156 RHR AND SAFETY INJECTION PIPING-PARTIAL PLANS AND SECTIONS BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES- REACTOR PRIMARY 3.6A-23 SK-2165-MNE-R-0163 COOLANT LOOP PIPING AND CONNECTIONS CONTAINMENT BUILDING-BREAK AND RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-3.6A-24 SK-2165-MNE-R-0176 STEAM GENERATOR BLOWDOWN AND SAMPLING PIPING REACTOR AUXILIARY BUILDING-BREAK AND RESTRAINT LOCATIONS-STEAM GENERATOR 3.6A-25 SK-2165-MNE-R-0177 BLOWDOWN PIPING COMPOSITE PIPING-SHIELDED PIPE TUNNEL-REACTOR AUXILIARY BUILDING-BREAK AND 3.6A-26 SK-2165-MNE-R-0244 RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES COMPOSITE PIPING-SHIELDED PIPE TUNNEL-REACTOR AUXILIARY BUILDING-BREAK AND 3.6A-27 SK-2165-MNE-R-0245 RESTRAINT LOCATIONS AND JET IMPINGEMENT ENVELOPES-SHEET 2 3.7.3-7 BACKFILL ADJACENT TO SEISMIC CATEGORY 1 BUILDINGS- PLANT AREA 7-G-0488 3.8.3-7 CONCRETE CONTAINMENT INTERNAL STRUCTURES-STEAM GEN. & PRESSURIZER SHIELD WALLS 7-G-0840 3.8.4-22A DIESEL FUEL OIL STORAGE TANK BUILDING SECTIONS & DETAILS 7-G-3171 3.8.4-23 PROTECTIVE MATS FOR CLASS 1 YARD DUCT RUNS AND MISCELLANEOUS 7-G-0509 3.8.4-24 CLASS 1 UNDERGROUND ELECTRIC MANHOLES MASONRY 7-G-0512 3.8.4-25 ESW SCREENING STRUCTURE MASONRY 7-G-2875 3.8.4-26 ESW-SCREENING STRUCTURE MASONRY 7-G-2876 3.8.4-27 ESW-SCREENING STRUCTURE MASONRY 7-G-2877 3.8.4-28 ESW AND CT MAKEUP INTAKE STRUCTURE MASONRY 7-G-2845 3.8.4-29 ESW AND CT MAKEUP INTAKE STRUCTURE MASONRY 7-G-2846 3.8.4-30 ESW AND CT MAKEUP INTAKE STRUCTURE MASONRY 7-G-2847 3.8.4-31 ESW AND CT MAKEUP INTAKE STRUCTURE MASONRY 7-G-2848 3.8.4-32 ESW-DISCHARGE STRUCTURE MASONRY 7-G-2895 3.8.4-34 MAIN DAM SPILLWAY PLAN AND PROFILE 7-G-6248 3.8.4-35 MAIN DAM SPILLWAY SECTIONS 7-G-6249 3.8.4-36 MAIN DAM SPILLWAY OGEE, PIER AND ABUTMENTS 7-G-6252 3.8.4-37 AUXILIARY DAM SPILLWAY PLAN AND PROFILE 7-G-6280 3.8.5-3 GENERAL LAYOUT OF THE CONTAINMENT BUILDING-FOUNDATION MAT 7-G-0610 3.8.5-4 GENERAL LAYOUT OF THE CONTAINMENT BUILDING-FOUNDATION MAT 7-G-0611 DBA TEMPERATURE PROFILE INSIDE CONTAINMENT (COMBINED LOCA/MSLB) FOR DBD-1000-V02 FIGURE CB-1 3.11.4-1 ENVIRONMENTAL QUALIFICATION AND TABLE CB-1 DBD-1000-V02 FIGURE CB-3 3.11.4-2 DBA TEMPERATURE PROFILE INSIDE CONTAINMENT (LOCA) FOR ENVIRONMENTAL QUALIFICATION AND TABLE CB-3 DBD-1000-V02 FIGURE CB-4 3.11.4-3 DBA TEMPERATURE PROFILE INSIDE CONTAINMENT (MSLB) FOR ENVIRONMENTAL QUALIFICATION AND TABLE CB-4 3.11.4-4 DBA TEMPERATURE PROFILE INSIDE MAIN STEAM TUNNEL (MSLB) (1.4 FT2 MSLB AT 102% POWER) DBD-1000-V02, FIGURE AB-1 DBD-1000-V02 FIGURE CB-2 3.11.6-1 PRESSURE PROFILE INSIDE CONTAINMENT (LOCA/MSLB) FOR ENVIRONMENTAL QUALIFICATION AND TABLE CB-2 DBD-1000-V02 FIGURE CB-5 3.11.6-2 PRESSURE PROFILE INSIDE CONTAINMENT (MSLB) FOR ENVIRONMENTAL QUALIFICATION AND TABLE CB-5 Amendment 62 Page 3 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 23.11.6-3 PRESSURE PROFILE INSIDE MAIN STEAM TUNNEL (MSLB) (1.4 FT MSLB AT 102% POWER) DBD-1000-V02 FIGURE AB-2 Table Table Title Design Document 3.11B-1 EQ PLANT LOCATION/ZONE LIST DBD-1000-V02 TABLE 3-1 DBD-1000-V02, TABLES FOR TEMPERATURE, PRESSURE, 3.11B-2 TEMPERATURE HUMIDITY, SPRAY, AND SUBMERGENCE FOR ZONES OUTSIDE CONTAINMENT DBD-1000-V02, TABLE AB01-3.11B-3 VALVE CONTAINMENT TEMPERATURES FOLLOWING ACCIDENT 02 Figure Figure Title DBD-1000-V02, FIGURES GENERAL ARRANGEMENT CONTAINMENT BLDG. PLAN EL. 221.00 & 236.00 ENVIRONMENTAL 3.11B-1 CB11-1 AND CB21-1 AND PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES CB11-1 AND CB21-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT CONTAINMENT BLDG. PLAN EL. 261.00 AND 286.00 ENVIRONMENTAL 3.11B-2 CB31-1, AND CB41-1 AND PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES CB31-1 AND CB41-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 190.00 AND 216.00 ENVIRONMENTAL 3.11B-3 AB01-1, AND AB11-1 AND PARAMETERS DURING NORMAL & AND POST-ACCIDENT ENVIRONMENTS TABLES AB01-1, AND AB11-1 GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 236.00 ENVIRONMENTAL PARAMETERS DBD-1000-V02 FIGURE AB21-3.11B-4 DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1 AND TABLE AB21-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 261.00 ENVIRONMENTAL PARAMETERS 3.11B-5 AB31-1, AND AB32-1 AND DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES AB31-1 AND AB32-1 DBD-1000-V02, FIGURE TA12-3.11B-6 ENVIRONMENTAL PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1 AND TABLE TA12-1 DBD-1000-V02 FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 286.00 ENVIRONMENTAL PARAMETERS AB32-1, AB41-1, AND AB43-1; 3.11B-11 DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES AB32-1, AB41-1 AND AB43-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 305.00 ENVIRONMENTAL PARAMETERS AB51-1, A1, AB52-1, AND 3.11B-12 DURING NORMAL & POST-ACCIDENT ENVIRONMENTS FH61-1 AND TABLES AB51-1, AB51-2, AB52-1, AND FH61-1 GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 216.00 AND 236.00 ENVIRONMENTAL DBD-1000-V02, FIGURE FH21-3.11B-13 PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1, AND TABLE FH21-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT FUEL HANDLING BLDG. - PLAN EL. 261.00 & 286.00 ENVIRONMENTAL 3.11B-14 FH31-1 AND FH41-1 AND PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES FH31-1 AND FH41-1 GENERAL ARRANGEMENT TANK AREA - PLAN EL. 236.00 & 261.00 ENVIRONMENTAL PARAMETERS DBD-1000-V02, FIGURE TA11-3.11B-15 DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1, AND TABLE TA11-1 GENERAL ARRANGEMENT WASTE PROCESSING BLDG. - PLAN EL. 236.00 ENVIRONMENTAL DBD-1000-V02, FIGURE 3.11B-16 PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS SW21-1 AND TABLE SW21-1 Amendment 62 Page 4 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document DBD-1000-V02, FIGURES GENERAL ARRANGEMENT DIESEL GENERATOR BLDG. - PLAN EL. 261.00, 280.00, & 292.00 DG31-1, DG31-2, AND DG31-3 3.11B-17 ENVIRONMENTAL PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS AND TABLES DG31-1, DG31-2 AND DG31-3 GENERAL ARRANGEMENT DIESEL OIL STORAGE TANK AREA - PLAN EL. 242.25 ENVIRONMENTAL DBD-1000-V02, FIGURE DF31-3.11B-18 PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1 AND TABLE DF31-1 GENERAL ARRANGEMENT EMERGENCY SERVICE WATER INTAKE STRUCTURE - PLAN EL. 262.00 DBD-1000-V02, FIGURE 1E31-3.11B-19 ENVIRONMENTAL PARAMETERS DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 1 AND TABLE 1E31-1 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT CONTAINMENT BLDG. PLAN EL. 221.00 & 236.00 INTEGRATED RADIATION 3.11B-20 CB11-2 AND CB21-2 AND DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES CB11-2 AND CB21-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT CONTAINMENT BLDG. PLAN EL. 261.00 & 286.00 INTEGRATED RADIATION 3.11B-21 CB31-2, AND CB41-2 AND DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES CB31-2 AND CB41-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 190.00 & 216.00 INTEGRATED 3.11B-22 AB01-2 AND AB11-2 AND RADIATION DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES AB01-3 AND AB11-2 GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 236.00 INTEGRATED RADIATION DOSES DBD-1000-V02, FIGURE AB21-3.11B-23 TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 2 AND TABLE AB21-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 261.00 UNIT 1 INTEGRATED RADIATION 3.11B-24 AB31-2 AND AB32-2 AND DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES AB31-2 AND AB32-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 286.00 INTEGRATED RADIATION DOSES AB32-2, AB41-2 AND AB43-2, 3.11B-25 TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS AND TABLES AB32-2, AB41-2 AND AB43-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT REACTOR AUX. BLDG. - PLAN EL. 305.00 INTEGRATED RADIATION DOSES AB51-2, AB52-2, AND FH61-2 3.11B-26 TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS AND TABLES AB51-3 AND AB52-2, AND FH61-2 DBD-1000-V02, FIGURES GENERAL ARRANGEMENT FUEL HANDLING BLDG. - PLANS AT EL. 216.00 & 236.00 INTEGRATED FH11-1, AND FH21-2 AND 3.11B.27 RADIATION DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES FH11-1, FH21-2 AND FH21-3 DBD-1000-V02 FIGURES GENERAL ARRANGEMENT FUEL HANDLING BLDG. - PLANS AT EL. 261.00 & 286.00 INTEGRATED FH31-2 AND FH41-2 AND 3.11B-28 RADIATION DOSES TO EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS TABLES FH31-2, FH31-3, AND FH41-2 GENERAL ARRANGEMENT TANK AREA PLAN AT EL. 236.00 INTEGRATED RADIATION DOSES TO DBD-1000-V02, FIGURE TA11-3.11B-29 EQUIP. DURING NORMAL & POST-ACCIDENT ENVIRONMENTS 2, AND TABLE TA11-2 4.2.2-9A AREVA ROD CLUSTER CONTROL ASSEMBLY OUTLINE 1364-042493 5.1.2-1 REACTOR COOLANT SYSTEM PROCESS FLOW DIAGRAM 5-G-0800 5.1.2-2 REACTOR COOLANT SYSTEM PROCESS FLOW DIAGRAM 5-G-0801 5.1.3-1 ELEVATION DRAWINGS OF REACTOR PRIMARY COOLANT LOOP PIPING AND CONNECTIONS 5-G-0163 5.4.7-1 RESIDUAL HEAT REMOVAL SYSTEM- FLOW DIAGRAM 5-G-0824 6.2.2-1 FLOW DIAGRAM CONTAINMENT SPRAY SYSTEM 5-G-0050 6.2.2-2 CONTAINMENT SPRAY PIPING CONTAINMENT BUILDING PLAN AND SECTIONS 5-G-0119 6.2.2-3 FLOW DIAGRAM CONTAINMENT COOLING SYSTEM PURGE AND VACUUM RELIEF 8-G-0517 6.2.2-10 HVAC CONTAINMENT BUILDING PLANS-SHEET 1 8-G-0519 S01 6.2.2-11 HVAC CONTAINMENT BUILDING PLANS-SHEET 2 8-G-0519 S02 Amendment 62 Page 5 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 6.2.2-12 HVAC CONTAINMENT BUILDING PLANS-SHEET 3 8-G-0519 S03 6.2.2-13 HVAC CONTAINMENT BUILDING PLANS-SHEET 4 8-G-0519 S04 6.2.2-14 HVAC CONTAINMENT BUILDING SECTIONS-SHEET 1 8-G-0520 6.2.2-15 HVAC CONTAINMENT BUILDING SECTIONS-SHEET 2 8-G-0521 6.2.2-16 HVAC CONTAINMENT BUILDING SECTIONS-SHEET 3 8-G-0521 S02 6.2.2-19 CONTAINMENT BUILDING RECIRCULATION SUMP SCREENS 1364-036778 6.3.2-1 SAFETY INJECTION SYSTEM FLOW DIAGRAM 5-G-0808 6.3.2-2 SAFETY INJECTION SYSTEM FLOW DIAGRAM 5-G-0809 6.3.2-3 SAFETY INJECTION SYSTEM FLOW DIAGRAM 5-G-0810 7.2.1-1 S01 SHEET 1-INDEX AND SYMBOLS 1364-000864 7.2.1-1 S02 SHEET 2-REACTOR TRIP SIGNALS 1364-000865 7.2.1-1 S03 SHEET 3-FUNCTIONAL DIAGRAM- NUCLEAR INSTR. AND MANUAL TRIP SIGNALS 1364-000866 7.2.1-1 S04 SHEET 4-FUNCTIONAL DIAGRAM- NUCLEAR INSTR. PERMISSIVES AND BLOCKS 1364-000867 7.2.1-1 S05 SHEET 5-FUNCTIONAL DIAGRAM- PRIMARY COOLANT SYSTEM TRIP SIGNALS 1364-000868 7.2.1-1 S06 SHEET 6-FUNCTIONAL DIAGRAM- PRESSURIZER TRIP SIGNALS 1364-000869 7.2.1-1 S07 SHEET 7-STEAM GENERATOR TRIP SIGNALS 1364-000870 7.2.1-1 S08 SHEET 8-FUNCTIONAL DIAGRAM- SAFEGUARDS ACTUATION SIGNALS 1364-000871 7.2.1-1 S09 SHEET 9-ROD CONTROLS AND ROD BLOCKS 1364-000872 7.2.1-1 S10 SHEET 10-FUNCTIONAL DIAGRAM- STEAM DUMP CONTROL 1364-000873 7.2.1-1 S11 SHEET 11-FUNCTIONAL DIAGRAM- PRESSURIZER PRESSURE LEVEL CONTROL 1364-000874 7.2.1-1 S12 SHEET 12-FUNCTIONAL DIAGRAM- PRESSURIZER HEATER CONTROL 1364-000875 7.2.1-1 S13 SHEET 13-FEEDWATER CONTROL AND ISOLATION 1364-000876 7.2.1-1 S14 SHEET 14-FUNCTIONAL DIAGRAM- AUXILIARY FEEDWATER PUMPS STARTUP 1364-000877 7.2.1-1 S15 SHEET 15-FUNCTIONAL DIAGRAM- TURBINE TRIPS RUNBACKS AND OTHER SIGNALS 1364-000878 7.3.1-1 S01 SHEET 1-SOLID STATE PROTECTION SYSTEM FUNCTIONAL DIAGRAMS 1364-000864 7.3.1-1 S02 SHEET 2-SOLID STATE PROTECTION SYSTEM FUNCTIONAL DIAGRAMS 1364-000871 7.3.1-1 S03 SHEET 3-SOLID STATE PROTECTION SYSTEM FUNCTIONAL DIAGRAMS 1364-000877 7.3.1-1 S06 SHEET 6-SOLID STATE PROTECTION SYSTEM FUNCTIONAL DIAGRAMS 1364-000876 7.3.1-1 S07 SHEET 7-SOLID STATE PROTECTION SYSTEM FUNCTIONAL DIAGRAMS 1364-000878 7.3.1-3 CONTAINMENT SPRAY SYSTEM - LOGIC & SCHEMATIC DIAGRAMS 6-G-0423 7.3.1-4 S01 SHEET 1-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.64 7.3.1-4 S02 SHEET 2-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.65 7.3.1-4 S03 SHEET 3-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.65A 7.3.1-4 S04 SHEET 4-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.66 7.3.1-4 S05 SHEET 5-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.67 7.3.1-4 S06 SHEET 6-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.67A Amendment 62 Page 6 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-4 S07 SHEET 7-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.68 7.3.1-4 S08 SHEET 8-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.69 7.3.1-4 S09 SHEET 9-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.69A 7.3.1-4 S10 SHEET 10-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.70 7.3.1-4 S11 SHEET 11-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.71 7.3.1-4 S12 SHEET 12-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.71A 7.3.1-4 S13 SHEET 13-CONTAINMENT COOLING SYSTEM, SAFETY RELATED-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.72 7.3.1-5 S01 SHEET 1-CONTAINMENT ISOLATION VALVE, MOTOR OPERATED-TYPICAL LOGIC DIAGRAMS 6-B-430 30.1 7.3.1-6 CONTAINMENT ISOLATION VALVE, AIR OPERATED-TYPICAL LOGIC 6-B-430 30.3 7.3.1-7 S01 SHEET 1-LOGIC DIAGRAMS-MAIN STEAM DRAIN LINES 6-B-430 08.7 7.3.1-7 S02 SHEET 2-LOGIC DIAGRAMS-MAIN STEAM ISOLATION VALVES FOR STEAM GENERATOR-1A 6-B-430 08.8 7.3.1-7 S03 SHEET 3-LOGIC DIAGRAMS-MAIN STEAM ISOLATION VALVES FOR STEAM GENERATOR-1B 6-B-430 08.9 7.3.1-7 S04 SHEET 4-LOGIC DIAGRAMS-MAIN STEAM ISOLATION VALVES FOR STEAM GENERATOR-1C 6-B-430 08.10 7.3.1-7 S05 SHEET 5-LOGIC DIAGRAMS-MAIN STEAM ISOLATION VALVES 6-B-430 08.11 7.3.1-7 S06 SHEET 6-LOGIC DIAGRAMS-MAIN STEAM ISOLATION VALVE BYPASS VALVES 6-B-430 08.12 SHEET 1-FEEDWATER TO STEAM GENERATOR 1A INSTRUMENT SCHEMATICS AND LOGIC 7.3.1-8 S01 6-G-0424 S01 DIAGRAMS SHEET 2-FEEDWATER TO STEAM GENERATOR 1B INSTRUMENT SCHEMATICS AND LOGIC 7.3.1-8 S02 6-G-0424 S02 DIAGRAMS SHEET 3-FEEDWATER TO STEAM GENERATOR 1C INSTRUMENT SCHEMATICS AND LOGIC 7.3.1-8 S03 6-G-0424 S03 DIAGRAMS 7.3.1-9 AUXILIARY FEEDWATER SYSTEM, MOTOR DRIVEN PUMP-LOGIC & SCHEMATIC DIAGRAM (1 SHEET) 6-G-0427 7.3.1-10 AUXILIARY FEEDWATER SYSTEM, TURBINE DRIVEN PUMP-LOGIC & SCHEMATIC DIAGRAM (1 SHEET) 6-G-0428 7.3.1-11 S01 SHEET 1-RAB EMERGENCY EXHAUST SYSTEMS LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.33 7.3.1-11 S02 SHEET 2-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.33A 7.3.1-11 S03 SHEET 3-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35BA 7.3.1-11 S04 SHEET 4-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35BB 7.3.1-11 S05 SHEET 5-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35DA 7.3.1-11 S06 SHEET 6-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35DB 7.3.1-11 S07 SHEET 7-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35E 7.3.1-11 S08 SHEET 8-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35EA 7.3.1-11 S09 SHEET 9-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35EB 7.3.1-11 S10 SHEET 10-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35CA 7.3.1-11 S11 SHEET 11-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35CB 7.3.1-11 S13 SHEET 13-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35M 7.3.1-11 S14 SHEET 14-RAB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.35N 7.3.1-12 S09 SHEET 9-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33J 7.3.1-12 S10 SHEET 10-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33K 7.3.1-12 S11 SHEET 11-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33L Amendment 62 Page 7 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-12 S12 SHEET 12-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33M 7.3.1-12 S13 SHEET 13-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33N 7.3.1-12 S14 SHEET 14-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33O 7.3.1-12 S15 SHEET 15-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33P 7.3.1-12 S16 SHEET 16-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33Q 7.3.1-12 S17 SHEET 17-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33R 7.3.1-12 S18 SHEET 18-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33S 7.3.1-12 S19 SHEET 19-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33T 7.3.1-12 S20 SHEET 20-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33U 7.3.1-12 S21 SHEET 21-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33V 7.3.1-12 S22 SHEET 22-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33W 7.3.1-12 S23 SHEET 23-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33X 7.3.1-12 S24 SHEET 24-RAB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.33Y 7.3.1-13 S04 SHEET 4-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.60D 7.3.1-13 S05 SHEET 5-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.60E 7.3.1-13 S06 SHEET 6-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.60F 7.3.1-13 S07 SHEET 7-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.60H 7.3.1-13 S11 SHEET 11-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.61D 7.3.1-13 S12 SHEET 12-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.61E 7.3.1-13 S13 SHEET 13-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.61F 7.3.1-13 S14 SHEET 14-FHB EMERGENCY EXHAUST SYSTEMS, LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.61H 7.3.1-14 S01 SHEET 1-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56D 7.3.1-14 S02 SHEET 2-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56A 7.3.1-14 S03 SHEET 3-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56E 7.3.1-14 S04 SHEET 4-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56B 7.3.1-14 S05 SHEET 5-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56F 7.3.1-14 S06 SHEET 6-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56G 7.3.1-14 S07 SHEET 7-FHB ISOLATION DAMPERS- LOGIC DIAGRAMS 6-B-430 31.56H 7.3.1-15 S01 SHEET 1-STATION SERVICE WATER SYSTEM-LOGIC AND SCHEMATIC DIAGRAMS 6-G-0425 S01 SHEET 2-STATION SERVICE WATER SYSTEM-LOGIC AND SCHEMATIC DIAGRAMS 7.3.1-15 S02 6-G-0425 S02 INSTRUMENTATION AND CONTROLS LOGIC AND SCHEMATIC SERVICE WATER TO AND FROM COMPONENT COOLING WATER HEAT 7.3.1-15A 6-B-430 21.2 EXCHANGERS 7.3.1-16 S01 SHEET 1-ESSENTIAL SERVICE CHILLED WATER SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82 7.3.1-16 S02 SHEET 2-ESSENTIAL SERVICE CHILLED WATER SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82A 7.3.1-16 S03 SHEET 3-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82B 7.3.1-16 S04 SHEET 4-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82C 7.3.1-16 S05 SHEET 5-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82D Amendment 62 Page 8 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-16 S07 SHEET 7-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82F 7.3.1-16 S10 SHEET 10-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.82I 7.3.1-16 S11 SHEET 11-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.83 7.3.1-16 S12 SHEET 12-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.83A 7.3.1-16 S13 SHEET 13-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.83C 7.3.1-16 S15 SHEET 15-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.84 7.3.1-16 S16 SHEET 16-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.84A 7.3.1-16 S17 SHEET 17-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.84B 7.3.1-16 S18 SHEET 18-ESSENTIAL SERVICE CHILLED WATER SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.84C 7.3.1-17 S01 SHEET 1-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.120 7.3.1-17 S02 SHEET 2-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.121 7.3.1-17 S03 SHEET 3-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.122 7.3.1-17 S04 SHEET 4-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.123 7.3.1-17 S05 SHEET 5-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.124 7.3.1-17 S06 SHEET 6-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.130 7.3.1-17 S07 SHEET 7-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.131 7.3.1-17 S08 SHEET 8-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.132 7.3.1-17 S09 SHEET 9-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.134 7.3.1-17 S10 SHEET 10-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.138 7.3.1-17 S11 SHEET 11-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.140 7.3.1-17 S12 SHEET 12-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.141 7.3.1-17 S13 SHEET 13-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.142 7.3.1-17 S14 SHEET 14-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.143 7.3.1-17 S15 SHEET 15-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.144 7.3.1-17 S16 SHEET 16-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.145 7.3.1-17 S17 SHEET 17-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.146 7.3.1-17 S18 SHEET 18-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.147 7.3.1-17 S19 SHEET 19-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.148 7.3.1-17 S20 SHEET 20-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.149 7.3.1-17 S21 SHEET 21-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.150 7.3.1-17 S22 SHEET 22-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.151 7.3.1-17 S23 SHEET 23-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.153 7.3.1-17 S24 SHEET 24-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.152 7.3.1-17 S25 SHEET 25-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.154 7.3.1-17 S26 SHEET 26-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.155 7.3.1-17 S27 SHEET 27-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.156 Amendment 62 Page 9 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-17 S28 SHEET 28-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.157 7.3.1-17 S29 SHEET 29-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.158 7.3.1-17 S30 SHEET 30-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.159 7.3.1-17 S31 SHEET 31-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.160A 7.3.1-17 S32 SHEET 32-CONTROL ROOM HVAC POST-ACCIDENT OAI ALARMS 6-B-430 31.160B 7.3.1-17 S33 SHEET 33-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.161 7.3.1-17 S34 SHEET 34-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.163 7.3.1-17 S35 SHEET 35-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.165 7.3.1-17 S36 SHEET 36-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.167 7.3.1-17 S37 SHEET 37-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.169 7.3.1-17 S38 SHEET 38-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.170 7.3.1-17 S39 SHEET 39-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.173 7.3.1-17 S40 SHEET 40-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.173A 7.3.1-17 S41 SHEET 41-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.174 7.3.1-17 S42 SHEET 42-CONTROL ROOM VENTILATION SYSTEM- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.174A 7.3.1-18 S01 SHEET 1-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A1 7.3.1-18 S02 SHEET 2-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A2 7.3.1-18 S03 SHEET 3-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A3 7.3.1-18 S04 SHEET 4-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A4 7.3.1-18 S05 SHEET 5-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A5 7.3.1-18 S06 SHEET 6-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A6 7.3.1-18 S07 SHEET 7-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85A7 7.3.1-18 S08 SHEET 8-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B1 7.3.1-18 S09 SHEET 9-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B2 7.3.1-18 S10 SHEET 10-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B3 7.3.1-18 S11 SHEET 11-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B4 7.3.1-18 S12 SHEET 12-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B5 7.3.1-18 S13 SHEET 1-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B6 7.3.1-18 S14 SHEET 14-ESF EQUIPMENT COOLING SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.85B7 7.3.1-19 S01 SHEET 1-DIESEL GENERATOR BUILDING VENTILATION-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.180 7.3.1-19 S02 SHEET 2-DIESEL GENERATOR BUILDING VENTILATION-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.180A 7.3.1-19 S03 SHEET 3-DIESEL GENERATOR BUILDING VENTILATION-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.180B 7.3.1-19 S04 SHEET 4-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.180C 7.3.1-19 S05 SHEET 5-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.180D 7.3.1-19 S06 SHEET 6-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.181 7.3.1-19 S07 SHEET 7-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.181A Amendment 62 Page 10 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-19 S08 SHEET 8-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.181B 7.3.1-19 S09 SHEET 9-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182 7.3.1-19 S10 SHEET 10-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182A 7.3.1-19 S11 SHEET 11-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182B 7.3.1-19 S12 SHEET 12-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182C 7.3.1-19 S13 SHEET 13-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182D 7.3.1-19 S14 SHEET 14-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182E 7.3.1-19 S15 SHEET 15-DIESEL GENERATOR BUILDING VENTILATION- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.182F SHEET 1-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S01 6-B-430 31.29 DIAGRAMS SHEET 2-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S02 6-B-430 31.30 DIAGRAMS SHEET 3-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S03 6-B-430 31.30A DIAGRAMS SHEET 4-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S04 6-B-430 31.30B DIAGRAMS SHEET 7-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S07 6-B-430 31.31 DIAGRAMS SHEET 8-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S08 6-B-430 31.31A DIAGRAMS SHEET 9-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S09 6-B-430 31.31B DIAGRAMS SHEET 10-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S10 6-B-430 31.31C DIAGRAMS SHEET 11-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S11 6-B-430 31.31D DIAGRAMS SHEET 12-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S12 6-B-430 31.31E DIAGRAMS SHEET 13-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S13 6-B-430 31.31F DIAGRAMS SHEET 14-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S14 6-B-430 31.31G DIAGRAMS SHEET 15-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S15 6-B-430 31.31H DIAGRAMS SHEET 16-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S16 6-B-430 31.31K DIAGRAMS SHEET 17-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S17 6-B-430 31.31L DIAGRAMS SHEET 18-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S18 6-B-430 31.31M DIAGRAMS SHEET 20-ELECTRIC EQUIPMENT PROTECTION ROOM HVAC SYSTEM- LOGIC & SCHEMATIC 7.3.1-20 S20 6-B-430 31.32 DIAGRAMS 7.3.1-21 S02 SHEET 2-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.73B 7.3.1-21 S03 SHEET 3-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.74 7.3.1-21 S04 SHEET 4-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.74A 7.3.1-21 S05 SHEET 5-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75 7.3.1-21 S06 SHEET 6-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75A 7.3.1-21 S07 SHEET 7-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75B 7.3.1-21 S08 SHEET 8-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75C 7.3.1-21 S09 SHEET 9-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75D 7.3.1-21 S10 SHEET 10-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.75E 7.3.1-21 S11 SHEET 11-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.76A 7.3.1-21 S12 SHEET 12-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.76B Amendment 62 Page 11 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-21 S13 SHEET 13-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.76C 7.3.1-21 S14 SHEET 14-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.77 7.3.1-21 S15 SHEET 15-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.77A 7.3.1-21 S16 SHEET 16-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78 7.3.1-21 S17 SHEET 17-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78A 7.3.1-21 S18 SHEET 18-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78B 7.3.1-21 S19 SHEET 19-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78C 7.3.1-21 S20 SHEET 20-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78D 7.3.1-21 S21 SHEET 21-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78E 7.3.1-21 S22 SHEET 22-RAB SWITCHGEAR ROOMS- LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.78F SHEET 1-FHB SPENT FUEL POOL PUMP ROOM VENTILATION SYSTEM-LOGIC & SCHEMATIC 7.3.1-22 S01 6-B-430 31.62 DIAGRAMS SHEET 3-FHB SPENT FUEL POOL PUMP ROOM VENTILATION SYSTEM-LOGIC & SCHEMATIC 7.3.1-22 S03 6-B-430 31.62B DIAGRAMS SHEET 6-FHB SPENT FUEL POOL PUMP ROOM VENTILATION SYSTEM-LOGIC & SCHEMATIC 7.3.1-22 S06 6-B-430 31.63 DIAGRAMS SHEET 8-FHB SPENT FUEL POOL PUMP ROOM VENTILATION SYSTEM-LOGIC & SCHEMATIC 7.3.1-22 S08 6-B-430 31.63B DIAGRAMS SHEET 9-FHB SPENT FUEL POOL PUMP ROOM VENTILATION SYSTEM-LOGIC & SCHEMATIC 7.3.1-22 S09 6-B-430 31.63C DIAGRAMS 7.3.1-24 S01 SHEET 1-FUEL OIL TRANSFER BUILDING VENTILATION SYSTEM 6-B-430 31.183 7.3.1-24 S02 SHEET 2-FUEL OIL TRANSFER BUILDING VENTILATION SYSTEM 6-B-430 31.183A 7.3.1-24 S03 SHEET 3-FUEL OIL TRANSFER BUILDING VENTILATION SYSTEM 6-B-430 31.183B 7.3.1-25 S01 SHEET 1-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184 7.3.1-25 S02 SHEET 2-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184A 7.3.1-25 S03 SHEET 3-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184B 7.3.1-25 S04 SHEET 4-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184C 7.3.1-25 S05 SHEET 5-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184D 7.3.1-25 S06 SHEET 6-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184E 7.3.1-25 S07 SHEET 7-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.184F 7.3.1-25 S08 SHEET 8-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.185 7.3.1-25 S09 SHEET 9-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.185A 7.3.1-25 S10 SHEET 10-EMERGENCY SERVICE WATER INTAKE STRUCTURE VENTILATION SYSTEM 6-B-430 31.185B 7.3.1-26 S01 SHEET 1-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.1 7.3.1-26 S02 SHEET 2-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.2 7.3.1-26 S03 SHEET 3-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.3 7.3.1-26 S04 SHEET 4-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.4 7.3.1-26 S05 SHEET 5-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.5 7.3.1-26 S06 SHEET 6-DIESEL FUEL OIL TRANSFER SYSTEM 6-B-430 19.6 SHEET 1-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS INSTRUMENTATION 7.3.1-27 S01 6-B-430 31.135 AND CONTROLS 7.3.1-27 S02 SHEET 2-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.136 Amendment 62 Page 12 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 7.3.1-27 S03 SHEET 3-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.137 7.3.1-27 S04 SHEET 4-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.186A 7.3.1-27 S05 SHEET 5-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.186B 7.3.1-27 S06 SHEET 6-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.186C 7.3.1-27 S07 SHEET 7-CONTAINMENT H2 PURGE SYSTEM-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 31.186D 7.4.1-4 MAIN STEAM FROM STEAM GENERATOR 1A 6-B-430 08.1 7.4.1-5 MAIN STEAM FROM STEAM GENERATOR 1B AND STEAM TO AUXILIARY FEEDWATER PUMP TURBINE 6-B-430 08.2 7.4.1-6 MAIN STEAM FROM STEAM GENERATOR 1C AND STEAM TO AUXILIARY FEEDWATER PUMP TURBINE 6-B-430 08.3 7.4.1-8 AUXILIARY CONTROL PANEL ARRANGEMENT 6-SK-E-0260 7.5.1-2 CONTROL WIRING DIAGRAM ESF SYSTEM A & B LIGHT BOX ENGRAVING 6-B-401 0602 7.5.1-4 CONTROL WIRING DIAGRAM SLB-5 FRONT VIEW ENGRAVING 6-B-401 0050E 7.5.1-5 CONTROL WIRING DIAGRAM SLB-6 FRONT VIEW ENGRAVING 6-B-401 0050F 7.5.1-6 CONTROL WIRING DIAGRAM SLB-9 FRONT VIEW ENGRAVING 6-B-401 0050J 7.5.1-7 CONTROL WIRING DIAGRAM SLB-8 FRONT VIEW ENGRAVING 6-B-401 0050H 7.5.1-14 CONTROL WIRING DIAGRAM AEP-1 STATUS LIGHT BOX SLB-10 (SA) ENGRAVING 6-B-401 0050K 7.5.1-15 CONTROL WIRING DIAGRAM AEP-1 STATUS LIGHT BOX SLB-11 (SA) ENGRAVING 6-B-401 0050L 7.6.1-10 FHB FUEL POOL B-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 04.3 7.6.1-11 FHB FUEL POOL A-LOGIC & SCHEMATIC DIAGRAMS 6-B-430 04.4 7.6.1-12 FHB FUEL POOL C&D LOGIC AND SCHEMATIC DIAGRAMS 6-B-430 05.1 7.6.1-13 FHB FUEL POOL C&D LOGIC AND SCHEMATIC DIAGRAMS 6-B-430 05.2 7.6.1-14 FHB FUEL POOL C LOGIC AND SCHEMATIC DIAGRAMS 6-B-430 05.3 7.6.1-15 FHB FUEL POOL D LOGIC AND SCHEMATIC DIAGRAMS 6-B-430 05.4 8.2.1-7 230 KV SWITCHYARD AUXILIARY ONE-LINE DIAGRAM 6-G-0025 8.2.1-8 TRANSFORMER YARD ARRANGEMENT PLAN 6-G-0172 8.3.1-1 DIESEL GENERATOR LOGIC DIAGRAM 6-G-0039 9.1.3-1 FLOW DIAGRAM FUEL POOLS COOLING SYSTEM SOUTH END 5-G-0305 9.1.3-2 FLOW DIAGRAM FUEL POOLS COOLING SYSTEM NORTH END 5-G-0307 9.1.3-3 FLOW DIAGRAM FUEL POOLS CLEAN-UP SYSTEM-SHEET 1 5-G-0061 9.1.3-4 FLOW DIAGRAM FUEL POOLS CLEAN-UP SYSTEM-SHEET 2 5-G-0062 9.1.4-5 FUEL TRANSFER SYSTEM 1364-002642 9.2.1-1 FLOW DIAGRAM CIRCULATING AND STATION SERVICE WATER SYSTEMS-SHEET 1 5-G-0047 9.2.1-2 FLOW DIAGRAM CIRCULATING AND STATION SERVICE WATER SYSTEMS-UNIT 1-SHEET 2 5-G-0048 9.2.2-1 FLOW DIAGRAM-COMPONENT COOLING WATER SYSTEM 5-G-0819 9.2.2-2 FLOW DIAGRAM-COMPONENT COOLING WATER SYSTEM 5-G-0820 9.2.2-3 FLOW DIAGRAM-COMPONENT COOLING WATER SYSTEM 5-G-0821 9.2.2-4 FLOW DIAGRAM-COMPONENT COOLING WATER SYSTEM 5-G-0822 Amendment 62 Page 13 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 9.2.2-5 FLOW DIAGRAM-COMPONENT COOLING WATER SYSTEM 5-G-0822 S01 9.2.3-1 FLOW DIAGRAM POTABLE AND DEMINERALIZED WATER SYSTEMS 5-G-0049 S02 9.2.3-2 FLOW DIAGRAM-REACTOR AUXILIARY BLDG. PRIMARY AND DEMINERALIZER WATER SYSTEMS 5-G-0299 9.2.3-3 FLOW DIAGRAM-MAKEUP WATER DEMINERALIZER 1364-020974 9.2.4-1 FLOW DIAGRAM-POTABLE AND DEMINERALIZED WATER SYSTEMS 5-G-0049 S01 9.2.8-3 FLOW DIAGRAM-HVAC ESSENTIAL SERVICES CHILLED WATER CONDENSER-SA 8-G-0498 S02 9.2.9-1 FLOW DIAGRAM-NONESSENTIAL SERVICES CHILLED WATER 8-G-0497 S02 9.2.9-2 NONESSENTIAL SERVICES CHILLED WATER SYSTEM-FLOW RATES &MISCELLANEOUS DETAILS 8-G-0497 S01 9.2.10-1 FLOW DIAGRAM-COOLING WATER SYSTEM FOR WASTE PROCESSING BUILDING-SHEET 1 5-G-0876 9.2.10-2 FLOW DIAGRAM-COOLING WATER SYSTEM FOR WASTE PROCESSING BUILDING-SHEET 2 5-G-0877 9.3.1-1 FLOW DIAGRAM-COMPRESSED AIR SYSTEM 5-G-0188 9.3.1-2 FLOW DIAGRAM-SERVICE AIR SYSTEM 5-G-0300 9.3.1-2A FLOW DIAGRAM-SERVICE AIR SYSTEM 5-G-0300 S02 9.3.1-3 FLOW DIAGRAM-INSTRUMENT AIR SYSTEM 5-G-0301 9.3.1-3A FLOW DIAGRAM-INSTRUMENT AIR SYSTEM 5-G-0301 S02 9.3.2-1 FLOW DIAGRAM-SAMPLING SYSTEM 5-G-0052 9.3.2-1A PRIMARY SAMPLE PANEL 1A FLOW DIAGRAM 1364-006781 S01 9.3.2-1B PRIMARY SAMPLE PANEL 1A FLOW DIAGRAM 1364-006781 S02 FLOW DIAGRAM-SAMPLING SYSTEM (NON-NUCLEAR) AND STEAM GENERATOR WET LAY-UP 9.3.2-2 5-G-0089 SYSTEM 9.3.2-2A FLOW DIAGRAM-SECONDARY SAMPLING SYSTEM-SAMPLING RECLAMATION SYSTEM 5-G-0089 S01 9.3.2-2B FLOW DIAGRAM-SECONDARY SAMPLING SYSTEM-SAMPLING RECLAMATION SYSTEM 5-G-0089 S02 9.3.2-3 POST ACCIDENT REACTOR COOLANT SAMPLING SYSTEM 1364-052455 9.3.3-1 FLOW DIAGRAM-REACTOR AUX. BLDG. DRAINAGE SYSTEMS 5-G-0184 9.3.3-2 FLOW DIAGRAM-CONTAINMENT TURBINE BLDG. AND TANK AREA DRAINAGE SYSTEM 5-G-0185 9.3.3-3 FLOW DIAGRAM-FUEL HANDLING BLDG. DRAINAGE SYSTEMS 5-G-0187 9.3.3-4 FLOW DIAGRAM-WASTE PROCESSING BUILDING-DRAINAGE SYSTEMS- SHEET 1 5-G-0427 9.3.3-5 FLOW DIAGRAM-WASTE PROCESSING BUILDING-DRAINAGE SYSTEMS- SHEET 2 5-G-0428 9.3.3-6 FLOW DIAGRAM-WASTE PROCESSING BUILDING-DRAINAGE SYSTEM- SHEET 3 5-G-0430 9.3.4-1 CHEMICAL AND VOLUME CONTROL SYSTEM-FLOW DIAGRAM 5-G-0803 9.3.4-2 CHEMICAL AND VOLUME CONTROL SYSTEM-FLOW DIAGRAM 5-G-0804 9.3.4-3 CHEMICAL AND VOLUME CONTROL SYSTEM-FLOW DIAGRAM 5-G-0805 9.3.4-4 CHEMICAL AND VOLUME CONTROL SYSTEM-FLOW DIAGRAM 5-G-0806 9.3.4-6 BORON RECYCLE SYSTEM- FLOW DIAGRAM 5-G-0811 9.3.4-7 BORON RECYCLE SYSTEM- FLOW DIAGRAM 5-G-0812 9.4.0-1 SYMBOLS AND ABBREVIATIONS FOR HVAC SYSTEM 8-G-0528 S02 9.4.1-1 HVAC-AIR FLOW DIAGRAM CONTROL ROOM REACTOR AUXILIARY BUILDING 8-G-0517 S04 Amendment 62 Page 14 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 9.4.2-1 HVAC-AIR FLOW DIAGRAM FUEL HANDLING BUILDING 8-G-0533 9.4.3-1 HVAC-AIR FLOW DIAGRAM REACTOR AUXILIARY BUILDING-SHEET 1 8-G-0517 S02 9.4.3-2 HVAC-AIR FLOW DIAGRAM REACTOR AUXILIARY BUILDING-SHEET 2 8-G-0517 S03 9.4.3-3 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 1 8-G-0533 S02 9.4.3-4 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 2 8-G-0533 S03 9.4.3-5 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 3 8-G-0533 S04 9.4.3-6 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 4 8-G-0533 S05 9.4.3-7 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 5 8-G-0533 S06 9.4.3-8 HVAC-AIR FLOW DIAGRAM WASTE PROCESSING BUILDING-SHEET 6 8-G-0533 S07 9.4.4-1 HVAC-AIR FLOW DIAGRAM TURBINE BUILDING 8-G-0562 HVAC-AIR FLOW DIAGRAM REACTOR AUXILIARY BUILDING SWITCHGEAR ROOMS AND EQUIPMENT 9.4.5-1 8-G-0517 S05 PROTECTION ROOMS 9.4.5-2 HVAC-AIR FLOW DIAGRAMS MISCELLANEOUS BUILDINGS 8-G-0548 9.4.9-1 HVAC-AIR FLOW DIAGRAM RAB COMPUTER, COMMUNICATION ROOM AND BATTERY ROOM 8-G-0532 S05 9.4.9-2 HVAC-COMPUTER & COMMUNICATION MECH EQUIP ROOM ROOF PLAN & REFRIG PIPING DIAGRAMS 8-G-0525 S05 9.5.4-1 FLOW DIAGRAM DIESEL FUEL OIL SYSTEM 5-G-0063 9.5.4-2 FUEL OIL PIPING SCHEMATIC 1364-007818 9.5.5-1 DIESEL GENERATOR COOLING WATER SYSTEM 1364-007812 9.5.5-2 FLOW DIAGRAM DIESEL GENERATOR SYSTEMS, UNIT 1 5-G-0133 9.5.6-1 DIESEL GENERATOR AIR STARTING SYSTEM 1364-007813 9.5.7-1 DIESEL GENERATOR LUBRICATION SYSTEM 1364-007817 10.1.0-1 FLOW DIAGRAM MAIN STEAM SYSTEM 5-G-0042 10.1.0-2 FLOW DIAGRAM EXTRACTION STEAM SYSTEM 5-G-0043 10.1.0-3 FLOW DIAGRAM FEEDWATER SYSTEM 5-G-0044 10.1.0-3A FLOW DIAGRAM FEEDWATER SYSTEM 5-G-0044 S02 10.1.0-4 FLOW DIAGRAM CONDENSATE AND AIR EVACUATION SYSTEMS 5-G-0045 10.1.0-5 FLOW DIAGRAM HEATER DRAIN AND VENT SYSTEMS 5-G-0046 10.1.0-6 FLOW DIAGRAM STEAM GENERATOR BLOWDOWN SYSTEM 5-G-0051 10.1.0-6A FLOW DIAGRAM STEAM GENERATOR BLOWDOWN HEAT RECOVERY & CLEAN UP SYSTEM 5-G-0051 S02 10.2.2-1 ASSEMBLY-LONGITUDINAL SECTION 1364-093034 10.2.2-2 TURBINE GENERATOR OUTLINE DRAWING 1364-000843 10.2.2-3 TURBINE GENERATOR OUTLINE DRAWING 1364-000844 10.2.2-5 FLOW DIAGRAM MISCELLANEOUS GAS SYSTEMS 5-G-0058 10.2.2-6 FLOW DIAGRAM MISCELLANEOUS SYSTEMS 5-G-0088 10.2.2-7 ELECTRO-HYDRAULIC FLUID SYSTEM AND LUBRICATION DIAGRAM SHEET 1 OF 3 1364-002795 S01 10.2.2-8 ELECTRO-HYDRAULIC FLUID SYSTEM AND LUBRICATION DIAGRAM SHEET 2 OF 3 1364-002795 S02 10.2.2-9 ELECTRO-HYDRAULIC FLUID SYSTEM AND LUBRICATION DIAGRAM SHEET 3 OF 3 1364-002795 S03 Amendment 62 Page 15 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-3 DESIGN DOCUMENTS INCORPORATED BY REFERENCE Figure Figure Title Design Document 10.2.2-10 TURBINE TRIP RUNBACK & OTHER SIGS W/REQUIREMENTS 1364-000878 10.4.5-1 CIRCULATING WATER SYSTEM COOLING TOWER INTAKE STRUCTURE AND CANAL 7-G-2770 10.4.5-2 CIRCULATING WATER SYSTEM COOLING TOWER INTAKE STRUCTURE AND CANAL 7-G-2771 10.4.6-1 FLOW DIAGRAM CONDENSATE DEMINERALIZER SYSTEM 1364-003628 10.4.6-2 FLOW DIAGRAM CONDENSATE DEMINERALIZER SYSTEM 1364-003629 10.4.6-3 FLOW DIAGRAM CONDENSATE DEMINERALIZER SYSTEM 1364-003630 10.4.7-1 FEEDWATER PIPING-PLANS 5-G-0071 10.4.7-2 FEEDWATER PIPING-PLANS 5-G-0072 10.4.7-3 FEEDWATER PIPING-SECTIONS 5-G-0073 10.4.7-4 AUXILIARY FEEDWATER PIPING CONTAINMENT BUILDING AND TUNNEL AREA 5-G-0074 11.2.2-1 FLOW DIAGRAM-CONTAINMENT BUILDING WASTE PROCESSING SYSTEM 5-G-0813 11.2.2-2 FLOW DIAGRAM-WASTE PROCESSING SYSTEM, WASTE HOLD-UP AND EVAPORATION 5-G-0814 11.2.2-3 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-SPENT RESIN STORAGE 5-G-0815 11.2.2-4 S01 SHEET 1-FLOW DIAGRAM-WASTE PROCESSING SYSTEM FLOOR DRAIN STORAGE AND TREATMENT 5-G-0866 11.2.2-4 S02 SHEET 2-FLOW DIAGRAM-WASTE PROCESSING SYSTEM FLOOR DRAIN STORAGE AND TREATMENT 5-G-0816 FLOW DIAGRAM-WASTE PROCESSING SYSTEM LAUNDRY AND HOT SHOWER STORAGE AND 11.2.2-5 5-G-0825 S01 TREATMENT-SHEET 1 FLOW DIAGRAM-WASTE PROCESSING SYSTEM LAUNDRY AND HOT SHOWER STORAGE AND 11.2.2-6 5-G-0825 S02 TREATMENT-SHEET 2 FLOW DIAGRAM-WASTE PROCESSING SYSTEM LAUNDRY AND HOT SHOWER STORAGE AND 11.2.2-7 5-G-0825 S03 TREATMENT-SHEET 3 11.2.2-8 S01 SHEET 1-FLOW DIAGRAM-SECONDARY WASTE TREATMENT SYSTEM 5-G-0090 11.2.2-8 S02 SHEET 2-FLOW DIAGRAM-SECONDARY WASTE TREATMENT SYSTEM 5-G-0480 11.2.2-9 FLOW DIAGRAM-MODULAR FLUIDIZED TRANSFER DEMINERALIZATION SYSTEM 1364-097532 11.3.2-1 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-GAS DECAY STORAGE 5-G-0817 11.3.2-2 FLOW DIAGRAM-WASTE PROCESSING SYSTEM, WASTE GAS COMPR, AND RECOMBINER 5-G-0818 11.4.2-1 FLOW DIAGRAM-WASTE PROCESSING SYSTEM RADWASTE SOLIDIFICATION 5-G-0826 11.4.2-3 FLOW DIAGRAM-WASTE PROCESSING SYSTEM VOLUME REDUCTION 5-G-0846 S02 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-CONCENTRATES STORAGE AND SPENT RESIN 11.4.2-4 5-G-0827 TRANSFER 11.4.2-5 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-FILTER BACKWASH SYSTEM-SHEET 1 5-G-0849 S01 11.4.2-6 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-FILTER BACKWASH SYSTEM-SHEET 2 5-G-0849 S02 11.4.2-7 FLOW DIAGRAM-WASTE PROCESSING SYSTEM-SPENT RESIN TRANSFER 5-G-0828 11.4.2-8 FLOW DIAGRAM-WASTE PROCESSING BUILDING-FILTER BACKWASH SYSTEM 5-G-0873 11.4.2-9 FLOW DIAGRAM-REACTOR AUXILIARY BUILDING-FILTER BACKWASH SYSTEM UNIT 1 5-G-0829 12.3.2-18 CONTAINMENT BUILDING LINER PENETRATIONS 5-G-0066 12.3.3-1 HVAC-REACTOR AUXILIARY BUILDING NORMAL EXHAUST EQUIPMENT ROOM EL. 286.00' 8-G-0524 S03 12.3.3-2 HVAC-REACTOR AUXILIARY BUILDING NORMAL EXHAUST EQUIPMENT ROOM SECTIONS-EL.286.00' 8-G-0539 S03 Amendment 62 Page 16 of 16

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.6-4 PROCEDURES, PROGRAMS, OR MANUALS INCORPORATED BY REFERENCE Document Document Title PLP-106 TECHNICAL SPECIFICATION EQUIPMENT LIST PROGRAM AND CORE OPERATING LIMITS REPORT PLP-114 RELOCATED TECHNICAL SPECIFICATIONS AND DESIGN BASES REQUIREMENTS EGR-NGGC-0153 ENGINEERING INSTRUMENT SETPOINTS DUKE-QAPD-001-A DUKE ENERGY CORPORATION TOPICAL REPORT QUALITY ASSURANCE PROGRAM DESCRIPTION OPERATING FLEET AD-EG-ALL-1153 ENGINEERING INSTRUMENT SETPOINT/UNCERTAINTY CALCULATIONS NOTE: AD-EG-ALL-1153 has superseded EGR-NGGC-0153, but EGR-NGGC-0153 can still be accessed. EGR-NGGC-0153 is listed in Technical Specifications and the Basis of Technical Specifications. Both procedures will be listed until EGR-NGGC-0153 is removed from Technical Specifications. The instrument uncertainty methodology has not changed in the new procedure, so this is an administrative change.Amendment 64 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.8-1 FUNCTIONAL LEVEL, ASSIGNMENT OF RESPONSIBILITY AND QUALIFICATION CROSS REFERENCE FOR SHNPP ANS 3.1 Section SHNPP Title Managers 4.2.1 Plant Manager - Harris Plant 4.2.3 Manager - Maintenance 4.2.2(a) Manager - Operations 4.3.2 Manager - Support Services 4.4.7 Manager - Training 4.4.5 Manager - Nuclear Oversight Section 4.6.1 Professional - Technical 4.6.1 Director - Design Engineering 4.6.1 Director - Engineering (Refer to FSAR Section 13.1.1) 4.6.2 (l) Shift Technical Advisor 4.6.2 (i) ALARA Analyst 4.6.2 (j) Superintendent - Work Control/Coordination 4.6.2 (j) Superintendent - Nuclear Operations Performance 4.6.2 (j) Supervisor - Design Engineering 4.6.2 (j) Supervisor - System Engineering 4.6.2 (j) Supervisor - Technical Programs 4.4.1 Senior Engineer - Reactor 4.4.4

• or 4.3.2(c) Superintendent - Radiation Protection 4.4.3 Superintendent - Environmental & Chemistry 4.4.2 Maintenance Superintendent 4.3.2(c) Supervisor - Spent Fuel Supervisors/Foreman 4.3.1 (b) Superintendent - Shift Operations 4.3.2 Administrative Supervisor 4.3.2 Superintendent - Nuclear Security 4.3.2(c) Fire Protection Coordinator 4.3.2(c) Maintenance Supervisor - Mechanical 4.3.2(c) I&C Supervisor 4.3.2(c) Electrical Supervisor 4.3.2(c) Mechanical Supervisor 4.3.2(c) Chemistry Supervisor 4.3.2(c) Health Physics Supervisor 4.3.2 Emergency Preparedness - Supervisor 4.3.2 (g) Nuclear Oversight Superintendent 4.3.2 (k) Specialist - Training Operators/Technicians/ Maintenance Personnel 4.5.2 Technician I - Engineering 4.5.2 Technician I - Health Physics 4.5.2(f) Technician II/III - Health Physics 4.5.2 Technician I - Chemistry 4.5.2(f) Technician II/III - Chemistry 4.5.2(g) Technician - QA/QC/NDE 4.5.2(h) Technical Aide - Security 4.5.2(h) Fire Protection Technician 4.5.2(h) Technical Aide - Training 4.5.2 Technician I - Maintenance 4.5.2 Technician I - I&C 4.5.2(f) Technician II - I&C 4.5.3 Electrician I 4.5.3 Planner Analyst 4.5.3 Senior Mechanic 4.5.3 Mechanic I 4.5.3(f) Mechanic II 4.5.1.2(e) Senior Control Operator 4.5.1.2(e) Control Operator 4.5.1.1(d) Auxiliary Operator Amendment 65 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 TABLE 1.8-1 FUNCTIONAL LEVEL, ASSIGNMENT OF RESPONSIBILITY AND QUALIFICATION CROSS REFERENCE FOR SHNPP( ) denotes exceptions or alternatives proposed in paragraph 3 of Section 1.8 discussion on Regulatory Guide 1.8.

• Required when individual is designated as site Radiation Protection Manager.

Amendment 65 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FIGURE TITLE 1.1.1-1 FLOW DIAGRAM LEGEND, PAGE 1 1.1.1-1a REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.1.1-2 FLOW DIAGRAM LEGEND, PAGE 2 1.2.2-1 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-2 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-3 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-4 DELETED BY AMENDMENT NO. 15 1.2.2-5 DELETED BY AMENDMENT NO. 10 1.2.2-6 DELETED BY AMENDMENT NO. 10 1.2.2-7 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-8 DELETED BY AMENDMENT NO. 15 1.2.2-9 DELETED BY AMENDMENT NO. 10 1.2.2-10 DELETED BY AMENDMENT NO. 10 1.2.2-11 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-12 DELETED BY AMENDMENT NO. 15 1.2.2-13 DELETED BY AMENDMENT NO. 10 1.2.2-14 DELETED BY AMENDMENT NO. 10 1.2.2-15 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-16 DELETED BY AMENDMENT NO. 15 1.2.2-17 DELETED BY AMENDMENT NO. 10 1.2.2-18 DELETED BY AMENDMENT NO. 10 1.2.2-19 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-20 DELETED BY AMENDMENT NO. 15 1.2.2-21 DELETED BY AMENDMENT NO. 10 1.2.2-22 DELETED BY AMENDMENT NO. 10 1.2.2-23 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-24 DELETED BY AMENDMENT NO. 15 Amendment 65 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FIGURE TITLE 1.2.2-25 DELETED BY AMENDMENT NO. 10 1.2.2-26 DELETED BY AMENDMENT NO. 10 1.2.2-27 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-28 DELETED BY AMENDMENT NO. 15 1.2.2-29 DELETED BY AMENDMENT NO. 10 1.2.2-30 DELETED BY AMENDMENT NO. 10 1.2.2-31 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-32 DELETED BY AMENDMENT NO. 15 1.2.2-33 DELETED BY AMENDMENT NO. 10 1.2.2-34 DELETED BY AMENDMENT NO. 10 1.2.2-35 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-36 DELETED BY AMENDMENT NO. 15 1.2.2-37 DELETED BY AMENDMENT NO. 10 1.2.2-38 DELETED BY AMENDMENT NO. 10 1.2.2-39 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-40 DELETED BY AMENDMENT NO. 15 1.2.2-41 DELETED BY AMENDMENT NO. 10 1.2.2-42 DELETED BY AMENDMENT NO. 10 1.2.2-43 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-44 DELETED BY AMENDMENT NO. 15 1.2.2-45 DELETED BY AMENDMENT NO. 10 1.2.2-46 DELETED BY AMENDMENT NO. 10 1.2.2-47 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-48 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-49 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-50 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-51 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE Amendment 65 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FIGURE TITLE 1.2.2-52 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-53 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-54 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-55 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-56 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-57 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-58 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-59 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-59A REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-60 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-61 DELETED BY AMENDMENT NO. 15 1.2.2-62 DELETED BY AMENDMENT NO. 10 1.2.2-63 DELETED BY AMENDMENT NO. 10 1.2.2-64 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-65 DELETED BY AMENDMENT NO. 15 1.2.2-66 DELETED BY AMENDMENT NO. 10 1.2.2-67 DELETED BY AMENDMENT NO. 10 1.2.2-68 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-69 DELETED BY AMENDMENT NO. 15 1.2.2-70 DELETED BY AMENDMENT NO. 10 1.2.2-71 DELETED BY AMENDMENT NO. 10 1.2.2-72 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-73 DELETED BY AMENDMENT NO. 15 1.2.2-74 DELETED BY AMENDMENT NO. 10 1.2.2-75 DELETED BY AMENDMENT NO. 10 1.2.2-76 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-77 DELETED BY AMENDMENT NO. 15 Amendment 65 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FIGURE TITLE 1.2.2-78 DELETED BY AMENDMENT NO. 10 1.2.2-79 DELETED BY AMENDMENT NO. 10 1.2.2-80 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-81 DELETED BY AMENDMENT NO. 15 1.2.2-82 DELETED BY AMENDMENT NO. 10 1.2.2-83 DELETED BY AMENDMENT NO. 10 1.2.2-84 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-85 DELETED BY AMENDMENT NO. 10 1.2.2-86 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.2.2-87 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 1.5.2-1 DELETED BY AMENDMENT NO. 48 Amendment 65 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 1 FIGURE 1.1.1-1 FLOW DIAGRAM LEGEND Amendment 61 Page 1 of 1

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Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.0 SITE CHARACTERISTICS ................................................................................................ 1 2.1 GEOGRAPHY AND DEMOGRAPHY ............................................................................. 1 2.1.1 SITE LOCATION AND DESCRIPTION ................................................................... 1 2.1.1.1 Specification of Location .................................................................................. 1 2.1.1.2 Site Area Map .................................................................................................. 1 2.1.1.3 Boundaries for Establishing Effluent Release Limits ........................................ 2 2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL ................................................ 2 2.1.2.1 Authority ........................................................................................................... 2 2.1.2.2 Control of Activities Unrelated to Plant Operation. ........................................... 3 2.1.2.3 Arrangements for Traffic Control ...................................................................... 4 2.1.3 POPULATION DISTRIBUTION ............................................................................... 4 2.1.3.1 Population Within Ten Miles ............................................................................. 4 2.1.3.2 Population Between Zero and Fifty Miles ......................................................... 4 2.1.3.3 Transient Population ........................................................................................ 5 2.1.3.4 Low Population Zone ....................................................................................... 5 2.1.3.5 Population Center ............................................................................................ 6 2.1.3.6 Deleted ............................................................................................................. 6 2.1.3.7 Evacuation Planning ........................................................................................ 6

REFERENCES:

SECTION 2.1 .......................................................................................... 6 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES ................ 6 2.2.1 LOCATION AND ROUTES ..................................................................................... 6 2.

2.2 DESCRIPTION

S ..................................................................................................... 7 2.2.2.1 Description of Facilities .................................................................................... 7 2.2.2.2 Description of Products and Materials ............................................................. 8 2.2.2.3 Pipelines ........................................................................................................... 8 2.2.2.4 Waterways ....................................................................................................... 8 2.2.2.5 Airports ............................................................................................................. 8 2.2.2.6 Projections for Industrial Growth ...................................................................... 9 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS ......................................................... 9 2.2.3.1 Design Basis Explosive Events ........................................................................ 9 Amendment 65 Page i of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.2.3.2 Nearby Gas Pipeline ...................................................................................... 10 2.2.3.3 Design Basis Toxic Chemicals ....................................................................... 28 2.2.3.4 Fires ............................................................................................................... 30 2.2.3.5 Collision with the Intake Structure .................................................................. 30 2.2.3.6 Liquid Spills .................................................................................................... 30 2.2.3.7 Aircraft Operations Evaluation ....................................................................... 30

REFERENCES:

SECTION 2.2 ........................................................................................ 30 2.3 METEOROLOGY ......................................................................................................... 32 2.3.1 REGIONAL CLIMATOLOGY ................................................................................. 32 2.3.1.1 General Climate ............................................................................................. 32 2.3.1.2 Regional Meteorological Conditions for Design and Operating Bases............................................................................................................. 33 2.3.2 LOCAL METEOROLOGY ..................................................................................... 37 2.3.2.1 Normal and Extreme Values of Meteorological Parameters. ......................... 37 2.3.2.2 Potential Influence of the Plant and Its Facilities on Local Meteorology ................................................................................................... 42 2.3.2.3 Local Meteorological Conditions for Design and Operating Bases............................................................................................................. 44 2.3.3 ON-SITE METEOROLOGICAL MEASUREMENTS PROGRAM .......................... 44 2.3.3.1 On-Site Operational Program ......................................................................... 44 2.3.3.2 Data Reduction .............................................................................................. 45 2.3.3.3 Maintenance and Calibration ......................................................................... 46 2.3.3.4 On Site Data ................................................................................................... 46 2.3.3.5 Regional Air Flow Trajectory Considerations. ................................................ 47 2.3.4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES ....................................... 47 2.3.4.1 Objective ........................................................................................................ 47 2.3.4.2 Calculations .................................................................................................... 47 2.3.5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES............................................ 50 2.3.5.1 Objective ........................................................................................................ 50 2.3.5.2 Calculations .................................................................................................... 51

REFERENCES:

SECTION 2.3 ......................................................................................... 52 2.4 HYDROLOGIC ENGINEERING ................................................................................... 54 Amendment 65 Page ii of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.1 HYDROLOGIC DESCRIPTION ............................................................................. 54 2.4.1.1 Site and Facilities ........................................................................................... 54 2.4.1.2 Hydrosphere ................................................................................................... 55 2.4.2 FLOOD .................................................................................................................. 61 2.4.2.1 Flood History .................................................................................................. 61 2.4.2.2 Flood Design Considerations ......................................................................... 61 2.4.2.3 Effects of Local Intense Precipitation ............................................................. 65 2.4.3 PROBABLE MAXIMUM FLOOD ON STREAMS AND RIVERS (PMF)................. 66 2.4.3.1 Probable Maximum Precipitation (PMP) ........................................................ 67 2.4.3.2 Precipitation Losses ....................................................................................... 67 2.4.3.3 Runoff Model .................................................................................................. 68 2.4.3.4 Probable Maximum Flood Flow (PMF) ........................................................... 69 2.4.3.5 Water Level Determination ............................................................................. 71 2.4.3.6 Coincident Wind Wave Activity ...................................................................... 71 2.4.4 POTENTIAL DAM FAILURES, SEISMICALLY INDUCED .................................... 73 2.4.4.1 Dam Failure Permutations ............................................................................. 73 2.4.4.2 Unsteady Flow Analysis of Potential Dam Failure.......................................... 74 2.4.4.3 Water Level at Plant Site ................................................................................ 75 2.4.5 PROBABLE MAXIMUM SURGE FLOODING ....................................................... 76 2.4.5.1 Probable Maximum Wind and Associated Meteorological Parameters .................................................................................................... 76 2.4.5.2 Surge and Seiche Water Levels ..................................................................... 77 2.4.5.3 Wave Action ................................................................................................... 77 2.4.5.4 Resonance ..................................................................................................... 78 2.4.5.5 Protection of Structures .................................................................................. 78 2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING .................................................... 79 2.4.7 ICE EFFECTS ....................................................................................................... 80 2.4.8 COOLING WATER CANALS AND RESERVOIRS ............................................... 81 2.4.9 CHANNEL DIVERSION ........................................................................................ 82 2.4.10 FLOODING PROTECTION REQUIREMENTS ..................................................... 82 2.4.11 LOW WATER CONSIDERATIONS ....................................................................... 83 Amendment 65 Page iii of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.11.1 Low Reservoir Level ....................................................................................... 83 2.4.11.2 Low Flow in Streams ...................................................................................... 85 2.4.11.3 Low Water Resulting From Surges, Seiches, and Tsunamis ......................... 85 2.4.11.4 Historical Low Water ...................................................................................... 86 2.4.11.5 Future Controls .............................................................................................. 86 2.4.11.6 Plant Requirements ........................................................................................ 86 2.4.11.7 Heat Sink Dependability Requirements ......................................................... 87 2.4.12 DISPERSION, DILUTION AND TRAVEL TIME OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS ........................... 90 2.4.12.1 Storage Tank Failure. ..................................................................................... 91 2.4.13 GROUNDWATER ................................................................................................. 92 2.4.13.1 Description and Onsite Use ........................................................................... 92 2.4.13.2 Source ............................................................................................................ 94 2.4.13.3 Accident Effects ............................................................................................. 98 2.4.13.4 Monitoring of Safeguard Requirements ......................................................... 99 2.4.13.5 Design Bases for Subsurface Hydrostatic Loading ...................................... 100 2.4.14 TECHNICAL SPECIFICATION AND EMERGENCY OPERATION REQUIREMENTS ............................................................................................... 101 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING....................... 103 2.5.0

SUMMARY

.......................................................................................................... 103 2.5.0.1 Basic Geologic and Seismic Information ...................................................... 103 2.5.0.2 Vibratory Ground Motion .............................................................................. 107 2.5.0.3 Surface Faulting ........................................................................................... 109 2.5.0.4 Stability of Subsurface Materials .................................................................. 111 2.5.0.5 Stability of Slopes ......................................................................................... 113 2.5.0.6 Embankments and Dams ............................................................................. 113 2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION ........................................... 116 2.5.1.1 Regional Geology ......................................................................................... 116 2.5.1.2 Site Geology ................................................................................................. 127 2.5.2 VIBRATORY GROUND MOTION ....................................................................... 140 2.5.2.1 Seismicity ..................................................................................................... 140 2.5.2.2 Geologic Structures and Tectonic Activity .................................................... 147 Amendment 65 Page iv of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2.3 Correlation of Earthquake Activity With Geologic Structures or Tectonic Provinces ...................................................................................... 147 2.5.2.4 Maximum Earthquake Potential ................................................................... 149 2.5.2.5 Seismic Wave Transmission Characteristics of the Site .............................. 151 2.5.2.6 Safe Shutdown Earthquake ......................................................................... 153 2.5.2.7 Operating Basis Earthquake ........................................................................ 155 2.5.3 SURFACE FAULTING ........................................................................................ 156 2.5.3.1 Geologic Conditions of the Site .................................................................... 157 2.5.3.2 Evidence of Fault Offset ............................................................................... 157 2.5.3.3 Earthquake Associated with Capable Faults ................................................ 174 2.5.3.4 Investigation of Capable Faults .................................................................... 174 2.5.3.5 Correlations of Epicenters with Capable Faults ........................................... 174 2.5.3.6 Descriptions of Capable Faults .................................................................... 174 2.5.3.7 Zones Requiring Detailed Faulting Investigation .......................................... 174 2.5.3.8 Results of Faulting Investigations ................................................................ 174 2.5.4 STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS .................. 174 2.5.4.1 Geologic Features ........................................................................................ 174 2.5.4.2 Properties of Subsurface Materials .............................................................. 175 2.5.4.3 Exploration ................................................................................................... 178 2.5.4.4 Geophysical Surveys ................................................................................... 180 2.5.4.5 Excavation and Backfill ................................................................................ 181 2.5.4.6 Groundwater Conditions .............................................................................. 183 2.5.4.7 Response of Soil and Rock to Dynamic Loading ......................................... 185 2.5.4.8 Liquefaction Potential ................................................................................... 187 2.5.4.9 Earthquake Design Basis ............................................................................. 188 2.5.4.10 Static Stability ............................................................................................... 188 2.5.4.11 Design Criteria ............................................................................................. 193 2.5.4.12 Techniques to Improve Subsurface Conditions ........................................... 194 2.5.4.13 Subsurface Instrumentation ......................................................................... 194 2.5.4.14 Construction Notes ....................................................................................... 194 2.5.5 STABILITY OF SLOPES ..................................................................................... 195 2.5.6 EMBANKMENTS AND DAMS ............................................................................. 195 2.5.6.1 General Information ..................................................................................... 195 Amendment 65 Page v of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.2 Exploration ................................................................................................... 199 2.5.6.3 Foundation and Abutment Treatment .......................................................... 211 2.5.6.4 Embankments .............................................................................................. 213 2.5.6.5 Slope Stability .............................................................................................. 220 2.5.6.6 Seepage Control .......................................................................................... 229 2.5.6.7 Diversion and Closure .................................................................................. 230 2.5.6.8 Performance Monitoring ............................................................................... 232 2.5.6.9 Construction Notes ....................................................................................... 233 2.5.6.10 Operational Notes ........................................................................................ 233

REFERENCES:

SECTION 2.5 .............................................................................................. 233 APPENDIX 2.5A ....................................................................................................................... 241

1. TABULATIONS OF BOREHOLE AND TEST PIT LOCATIONS................................. 241

2. LOGS OF BOREHOLES AND TEST PITS................................................................. 269 APPENDIX 2.5B LABORATORY ANALYSES OF FOUNDATIONMATERIALS FOR DAMS AND DIKE ............................................................................................... 269 APPENDIX 2.5C PRECONSTRUCTION BORROW MATERIAL TESTING ....................... 269 2.5C.1 INDEX PROPERTIES OF BORROW MATERIAL ............................................... 269 2.5C.2 LABORATORY TESTING PROGRAM FOR STATIC AND DYNAMIC ENGINEERING PROPERTIES........................................................................... 273 2.5C.3 RESULTS OF TESTS ON SAMPLES FROM BORINGS IN BORROW AREAS .............................................................................................. 284 APPENDIX 2.5D SEISMIC STABILITY ANALYSIS OF SEISMIC CATEGORY I DAMS AND DIKE ........................................................................................................ 288 2.5D.0 PURPOSE AND CONCLUSION .................................................................. 288 2.5D.0.1 Introduction .................................................................................................. 288 2.5D.0.2 Category I Dams: Main Dam, Auxiliary Dam, and Auxiliary Separating Dike ........................................................................................... 289 2.5D.0.3 Conventional Static and Dynamic Stability Analyses of the Category I Dams .......................................................................................... 290 2.5D.0.4 Finite Element Seismic Stability Analyses for the Category I Dams ........................................................................................................... 290 2.5D.0.5 Evaluation of the Seismic Stability of the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike ............................................................ 291 Amendment 65 Page vi of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.1 INTRODUCTION ................................................................................................. 292 2.5D.2 DESCRIPTION OF CATEGORY I DAMS ........................................................... 293 2.5D.2.1 General ........................................................................................................ 293 2.5D.2.2 Main Dam ..................................................................................................... 293 2.5D.2.3 Auxiliary Dam ............................................................................................... 293 2.5D.2.4Auxiliary Separating Dike ................................................................................. 293 2.5D.3 PROCEDURE USED IN SEISMIC STABILITY EVALUATION ........................... 294 2.5D.4 SAFE SHUTDOWN EARTHQUAKE AND LOADING CONDITIONS .................. 295 2.5D.4.1 Safe Shutdown Earthquake ......................................................................... 295 2.5D.4.2 Artificial Accelerogram ................................................................................. 295 2.5D.4.3 Comparison Between the Artificial Accelerogram and Actual Accelerograms of Earthquakes Applicable to the Site ................................. 296 2.5D.4.4 Operating Basis Earthquake ........................................................................ 297 2.5D.4.5 Loading Conditions ...................................................................................... 297 2.5D.5 MAIN DAM, EVALUATION OF SEISMIC STABILITY ......................................... 298 2.5D.5.1 Material Properties ....................................................................................... 298 2.5D.5.2 Static Stress Analysis ................................................................................... 299 2.5D.5.3 Dynamic Stress Analysis .............................................................................. 299 2.5D.5.4 Seismic Stability Evaluation ......................................................................... 300 2.5D.5.5 Conclusions for Main Dam ........................................................................... 305 2.5D.6 AUXILIARY DAM, EVALUATION OF SEISMIC STABILITY ............................... 305 2.5D.6.1 Material Properties ....................................................................................... 305 2.5D.6.2 Static Stress Analysis ................................................................................... 305 2.5D.6.3 Dynamic Stress Analysis .............................................................................. 306 2.5D.6.4 Seismic Stability Evaluation for the Maximum Cross Section ...................... 306 2.5D.6.5 Seismic Stability Evaluation for Cross Section A-44 .................................... 308 2.5D.6.6 Seismic Stability Evaluation for Cross Section A-24 .................................... 310 2.5D.6.7 Conclusions .................................................................................................. 310 2.5D.7 AUXILIARY RESERVOIR SEPARATING DIKE, EVALUATION OF SEISMIC STABILITY .......................................................................................... 310 2.5D.7.1 Material Properties ....................................................................................... 310 2.5D.7.2 Static Stress Analysis ................................................................................... 310 Amendment 65 Page vii of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.7.3 Dynamic Stress Analysis .............................................................................. 310 2.5D.7.5 Conclusion ................................................................................................... 312 2.5D.8 SEISMIC STABILITY EVALUATION FOR LOADING CONDITION II ................. 312 2.5D.8.1 General ........................................................................................................ 312 2.5D.8.2 Main Dam ..................................................................................................... 312 2.5D.8.3 Auxiliary Dam ............................................................................................... 313 2.5D.8.4 Auxiliary Reservoir Separating Dike ............................................................. 313 2.5D.8.5 Conclusion ................................................................................................... 313 2.5D.9 SEISMIC STABILITY EVALUATION FOR LOADING CONDITION III ................ 314 2.5D.9.1 Introduction .................................................................................................. 314 2.5D.9.2 Description of Dams ..................................................................................... 314 2.5d.9.3 Procedure Used In Seismic Stability Evaluation .......................................... 314 2.5D.9.4 Design Earthquake and Loading Conditions ................................................ 315 2.5D.9.5 Main Dam, Evaluation Of Seismic Stability .................................................. 316 2.5D.9.6 Auxiliary Dam, Evaluation of Seismic Stability ............................................. 319 2.5D.10 PROPERTIES OF MATERIAL M ........................................................................ 322 2.5D.10.1 Introduction .................................................................................................. 322 2.5D.10.2 Origin and Preparation of Material M ........................................................... 323 2.5D.10.3 Physical Properties ...................................................................................... 323 2.5D.10.4 Static Stress-Strain Characteristics .............................................................. 324 2.5D.10.5 Dynamic Stress-Strain Characteristics ......................................................... 326 2.5D.11 PROPERTIES OF MATERIAL Z ......................................................................... 330 2.5D.11.2 Origin and Preparation of Material Z ............................................................ 330 2.5D.11.3 Physical Properties ...................................................................................... 330 2.5D.11.4 Static Stress-Strain Characteristics .............................................................. 331 2.5D.11.5 Dynamic-Strain Characteristics .................................................................... 331 2.5D.12 PROPERTIES OF FILTERS AND ROCKFILLS .................................................. 334 2.5D.12.1 Introduction .................................................................................................. 334 2.5D.12.2 Filter Materials .............................................................................................. 334 2.5D.12.3 Rockfill Materials .......................................................................................... 337 2.5D.13 PROPERTIES OF FOUNDATION MATERIALS ................................................. 340 2.5D.13.1 Introduction .................................................................................................. 340 Amendment 65 Page viii of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.13.2 Properties of Weathered Rock ..................................................................... 340 2.5D.13.3 Properties of In-situ Residual Soil in the Foundation of Auxiliary Dam and Auxiliary Reservoir Separating Dike ............................................ 342 2.5D.14 STATIC STRESS ANALYSIS .............................................................................. 346 2.5D.14.1 Introduction .................................................................................................. 346 2.5D.14.2 General Procedure ....................................................................................... 346 2.5D.14.3 Selection of Material Properties ................................................................... 346 2.5D.14.4 Analysis of Dams ......................................................................................... 347 2.5D.15 PARAMETRIC STUDIES .................................................................................... 349 2.5D.15.1 Introduction .................................................................................................. 349 2.5D.15.2 Expected Constructed Values of Material Properties ................................... 349 2.5D.15.3 Bases for Parametric Variations ................................................................... 350 2.5D.15.4 Material Property Combinations ................................................................... 352 2.5D.16 DYNAMIC ANALYSES AND STABILITY EVALUATIONS .................................. 353 2.5D.16.1 Introduction .................................................................................................. 353 2.5D.16.2 Dynamic Analyses ........................................................................................ 353 2.5D.16.5 Determination of Equivalent Uniform Peak Shear Stresses ......................... 354 2.5D.16.4 Stress Induced During Ground Motions ....................................................... 355 2.5D.16.5 Seismic Stability Evaluation ......................................................................... 355 2.5D.16.6 Influence of Vertical Component .................................................................. 356 2.5D.16.7 Consideration of Tensile Stresses ............................................................... 356 2.5D.17 COMPARISON OF RESPONSE BY VARIOUS ANALYTICAL MODELS ............................................................................................................. 357 2.5D.17.1 Introduction .................................................................................................. 357 2.5D.17.2 Case Studied ................................................................................................ 357 2.5D.17.3 Analytical Models ......................................................................................... 357 2.5D.17.4 Response Evaluation ................................................................................... 361 2.5D.17.6 Discussion and Conclusion .......................................................................... 363 2.5D.18 BEHAVIOR DURING EVENT PRESCRIBED BY REGULATORY GUIDE 1.60 SPECTRA....................................................................................... 363 2.5D.18.1 Introduction .................................................................................................. 363 2.5D.18.2 Procedure Used ........................................................................................... 364 2.5D.18.3 Results ......................................................................................................... 364 Amendment 65 Page ix of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.18.4 Conclusions .................................................................................................. 366 Amendment 65 Page x of x

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.0 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 SITE LOCATION AND DESCRIPTION 2.1.1.1 Specification of Location The SHNPP site is located in the extreme southwest corner of Wake County, North Carolina, and the southeast corner of Chatham County, North Carolina. The City of Raleigh, North Carolina, is approximately 16 mi. northeast, and the City of Sanford is about 15 mi. southwest.Carolina Power & Light Company (now Duke Energy Progress, Inc.) has constructed a dam on Buckhorn Creek about 2.5 mi. north of its confluence with the Cape Fear River. This dam has created an approximately 4000-acre reservoir which will be used for cooling tower makeup requirements. The power block structures are located on the northwest shore of the Main Reservoir about 4.5 mi. north of the Main Dam.Coordinates of the reactor are:Latitude (North) 35° 38 00 Longitude (West) 78° 57 22 North Carolina (North) 685,444.524 Plane Coordinates (East) 2,013,001.262 Universal Transverse (North) 3,945,013.683 Mercator Coordinates (East) 685,064.389 The universal transverse Mercator zone number for the SHNPP is 17.All elevations are referred to the National Geodetic Vertical Addendum of 1929, commonly known as mean sea level (MSL).2.1.1.2 Site Area Map Maps of the site area are included as Figures 2.1.1-1 and 2.1.2-1. Indicated are the site boundary line (which is the same as the station property boundary), the principal plant structures, the exclusion area, and the principal transportation routes. The station requires approximately 10,800 ac. Duke Energy Progress, Inc. (DEP) owns all land within the site boundary lines. There are no private, residential, industrial, institutional, or commercial structures (other than those related to plant operation) within this area. However, as recreational usage increases at the Main Reservoir, some recreational structures may be constructed in accordance with DEPs land use policy.The minimum distance (+/- 25 ft.) and direction from the reactor to an exclusion area boundary is 6790 ft. ESE.Amendment 65 Page 1 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 U.S. Highway 1 passes north of the site, and several State maintained roads traverse the area, allowing access to the plant and reservoir. The CSX Corp. Railroad passes north of the plant, and the Southern Railroad crosses south of the Main Dam. Railway access to the plant is provided by a DEP rail spur which connects to the CSX Corp. Railroad.The Cape Fear River lies adjacent to the site. Use of the river near the plant is limited to small, recreational boating activities.2.1.1.3 Boundaries for Establishing Effluent Release Limits The protected area, defined for the purpose of controlling access and egress to and from the site, coincides with the plant security fence as reflected in the Security Plan. The plant security fence and the protected area and its relation to the Main and Auxiliary Reservoirs are shown on Figure 2.1.2-1. Access and egress to the protected area is controlled by a security organization on the site. Authority to enter the area is given to plant personnel, authorized contractor personnel, and authorized visitors only. Before entry is authorized to a new employee or visitor, such a person may be subject to radiological safety training and may be issued a dosimetry device for recording personal exposure to radiation. Unauthorized access is prevented by physical barriers, closed-circuit TV cameras, security force patrols, intrusion detection equipment, and access control. These measures are described in greater detail in Section 13.6 and the Security Plan, which is submitted separately. Radiation monitors are also located at the exit to the plant-protected area for radiation protection purposes at egress from the area.The effluent release limits are established in accordance with 10 CFR 20 and Appendix I to 10 CFR 50 in order to ensure that (1) the concentrations of radionuclides in gaseous effluent at the exclusion boundary do not exceed the limits set forth in Table 2, Column 1 of Appendix B to 10 CFR 20; (2) the annual average concentrations of radionuclides in liquid effluent at the point of discharge do not exceed the limits set forth in Table 2, Column 2 of Appendix B to 10 CFR 20; and (3) the cumulative liquid and gaseous radionuclide releases do not result in exposures to individuals outside the exclusion boundary in excess of the limits set forth in Appendix I to 10 CFR 50.The liquid radioactive release point (via the cooling tower blowdown discharge line) is shown on Figure 2.4.1-1. Gaseous radioactive release points are shown on Figure 9.4.0-2. This figure identifies gaseous release points (airborne effluent release points) as well as outside air intakes.Figure 9.4.0-2 references Figure 9.4.0-1 for definition of the appropriate symbols for the gaseous release points as opposed to the outside air intake. Figure 9.4.0-2 also references Figure 2.1.2 1 for definition of the distances of the gaseous release points (airborne effluent release points) to the plant exclusion boundary. The Liquid and Gaseous Waste Processing Systems are discussed in Sections 11.2 and 11.3, respectively. These radioactive releases are within the limits set forth in 10 CFR 20 and 10 CFR 50.2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL 2.1.2.1 Authority The exclusion area is shown on Figure 2.1.2-1. All lands within the exclusion area are owned by DEP. For the most part, roads which existed within the exclusion area prior to construction have been abandoned by the State and are blocked to prevent public use. The exception to this is Shearon Harris Road (SR1134) which connects with access roads to the plant (Figure 2.1.2-Amendment 65 Page 2 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1). Easem*nts have been granted to the State of North Carolina for maintenance of this road.An easem*nt has also been granted to the area Telephone Service Provider for maintenance of communications lines to the plant.Duke Energy Progress, Inc. owns and maintains the rail spur which connects the plant with commercial railroad service. Under a commercial side-track agreement, rail carriers have access to the plant over the DEP track.No mineral rights have been leased within the exclusion area and there are no rights outstanding which could allow production of either surface or sub surface minerals.Furthermore, the potential for commercial exploitation of minerals within the exclusion area is minimal, and leasing of mineral rights by DEP is not anticipated.The distance from the plant to the exclusion area boundary for each major compass direction is given in Table 2.1.2-1.2.1.2.2 Control of Activities Unrelated to Plant Operation.Activities unrelated to plant operations which will be permitted within the exclusion area (aside from transit through the area) are described below:a) Activities along the State road will generally be limited to highway/utility maintenance.This activity could take place anywhere within the easem*nt granted to the State. In the event of an emergency requiring an exclusion area evacuation, road/utility maintenance personnel will be evacuated in accordance with Security Procedures. Signs are posted along the road at the exclusion area boundary stating that the area is private property and advising persons therein that they are subject to evacuation. It is estimated that the road could be cleared of any maintenance personnel within thirty minutes.b) Activity on the rail spur will be limited to that which is directly related to delivery of rail cars to the plant. The rail spur is owned by DEP, and necessary maintenance is under the cognizance of the Company. Commercial railroad personnel in the exclusion area involved in the delivery of rail cars could be evacuated within fifteen minutes.c) Activity along the easem*nt granted to the area Telephone Service Provider consists of construction, modification, repair, and maintenance of telephone lines to the plant.Signs are posted whenever the easem*nt intersects the exclusion area boundary, stating that the area is private property and advising that anyone therein is subject to evacuation.It is estimated that no more than ten people will be involved in telephone maintenance/installation operations in the exclusion area at any one time and that they could be evacuated within thirty minutes.d) Recreational use of the land and main reservoir within the exclusion area, by the general public, is permitted. Warning signs, similar to those described above, are posted at known points of entry on the exclusion area boundary (land) and buoyed in conspicuous locations within and on the boundary of reservoir waters inside the exclusion area.Persons in these areas should be able to clear the exclusion area within one hour of notification.Amendment 65 Page 3 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 e) HNP Nuclear Security Firearms Training Range, Cary Police Department Firing Range and Wake County Fire Department Training Facility are located within the Owner Controlled Exclusion Area. It is estimated that normally not more than 75 personnel combined will occupy these areas at one time and could be evacuated within thirty (30) minutes.2.1.2.3 Arrangements for Traffic Control If it becomes necessary to control traffic into the exclusion area, the following actions will be initiated:a) Access control will be established by Plant Security/Local Law Enforcement personnel on Shearon Harris Road (SR1134) where it intersects with the exclusion area boundary in order to limit access to the area to authorized personnel.b) In a similar manner, the rail spur will be closed to rail traffic. Scheduled rail deliveries may be cancelled or postponed.c) Telephone service activities along the easem*nt in the exclusion area will be prohibited or postponed. If necessary, warning signs will be posted at the intersection of the easem*nt and the boundary.d) Control of public access to recreational land and water areas will be exercised by motorized patrols of known or likely points of entry on land and by patrol boats on the reservoirs. Plant Security/Local Law Enforcement will provide traffic control. Assistance will be requested from the county sheriff departments and the North Carolina Highway Patrol.2.1.3 POPULATION DISTRIBUTION Estimates of existing population were based on the 2012 Evacuation Time Estimate Study completed by KLD Engineering, which was derived using data from the 2010 U.S. Census.Geographical Information System (GIS) software was used to process the geographic data and associated population counts for census blocks in each of the counties surrounding the plant.The populations were then aggregated over subzones to generate a permanent resident population count.2.1.3.1 Population Within Ten Miles A map showing the 10-mile radial area of the site is presented in Figure 2.1.3-1. Concentric circles have been drawn at distances of 1, 2, 3, 4, 5, and 10 miles from the center point of the originally planned four units. The circles have been divided into 22-1/2-degree segments, with each segment centered on one of the 16 compass points. The 2010 estimates of residential population within each of these areas are presented in Figure 2.1.3-5.2.1.3.2 Population Between Zero and Fifty Miles The population within a 50-mile radius of the plant site is marked by concentrations of people in and around the cities of Raleigh (16-28 mi. NE; 2010 population of 403,892); Durham (20-30 mi.N; 2010 population of 228,330); Fayetteville (37-43 mi. S; 2010 population of 200,564); and Amendment 65 Page 4 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Cary (13-18 mi. NE; 210 population of 135,234). Several other smaller cities and towns have populations greater than 10,000. Away from these population concentrations, there is a rural-type population distribution with small towns interspersed through the area. A map showing the 50-mile radial area and identifying major cities and towns is presented in Figure 2.1.3-2.Concentric circles have been drawn at distances of 10, 20, 30, 40, and 50 miles, using the center point of the originally planned four units. The circles have been divided into 22-1/2-degree segments, with each segment centered on one of the 16 compass points.2.1.3.3 Transient Population Recreational land uses which would attract transient concentrations of people within the 50-mile radius of the site are not extensive and are limited to the Harris County Park (2 mi. SE), Jordan Lake State Recreation Area (5-12 mi. NW), Umstead State Park (20 mi. NE), Raven Rock State Park (13 mi. SSE), Eno River State Park (30 mi. N), and Falls Lake State Recreation Area (30 mi. NNE) (Figure 2.1.3-3). On occasions, there are also high concentrations of people at sporting events and at functions at the various universities in the area. The North Carolina State Fair, a 10-day event held during October of each year in Raleigh, attracted a maximum of 151,467 (single-day attendance) people in 2010 (Reference 2.1.3-1).Daily transient population concentrates in and around the major industrial areas of the region as a result of commuting patterns of workers. Figure 2.1.3-6 summarizes the transient population within the 10-mile EPZ. Reference 2.1.3-2, Appendix E, lists all public use facilities within the 10-mile Emergency Planning Zone (EPZ).2.1.3.4 Low Population Zone As stated in Reference 2.1.3-5, the analysis of the Low Population Zone (LPZ) is reviewed to assure that appropriate protective measures could be taken in this area in the event of an emergency. The SHNPP Emergency Plan is the primary document to provide this assurance.The low population zone is defined as land within a three-mile radial area as measured from the center point of the originally planned four reactors. The basis for its selection is in conformance to both the definition of "low population zone" specified in 10 CFR 100.3 and the method for determining a low population zone specified in 10 CFR 100.11.The three-mile radial area scribed from the center point of the plant is conservatively large such that it envelopes the set of low population zones scribed from the center point of each originally planned reactor. Dose calculations which show conformance to the dose limit criteria of 10 CFR 50.67 are described in Section 15.6.5.The low population zone is shown in Figure 2.1.3-1 and is superimposed on the three-mile concentric circle. Highways, railways, and waterways are identified. The Harris County Park is located 2 mi. SE of the plant. The Harris Energy & Environmental Center is located 2.1 mi. ENE of the plant. Two private nursing homes (Browns Family Care Home and James Rest Home) are located between 2 and 3 miles NE. There are no other facilities or institutions such as schools, hospitals, prisons, or beaches, within the low population zone. However, as recreational usage increases at the Harris Reservoir, additional recreational areas may be established within the low population zone in accordance with the Duke Energy Progress, Inc.land use policy.Amendment 65 Page 5 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 There are no other facilities or institutions beyond the low population zone and within 5 mi. of the plant that require special consideration when evaluating emergency plans.The daily transient population within the low population zone (three mile radius) is approximately 2,455 people. This estimate includes employees and recreational transients within the three mile radius.2.1.3.5 Population Center The nearest population centers as defined in 10 CFR 100 are Holly Springs, NC (8 mi. E), and Apex, NC (8 mi. NE). The 2010 population of Holly Springs was 24,661. This number reflects a 268% increase over the 2000 census figure of 9,192. The 2010 population of Apex was 37,746.This number reflects a 54% increase over the 2000 census figure of 20,212. Cary, NC (13-18 mi. NE), the previous nearest population center, had a 2010 population of 135,234.Holly Springs, Apex, and Cary are located at a distance greater than one and one-third times the distance from the reactor to the nearest Low Population Zone boundary.2.1.3.6 Deleted 2.1.3.7 Evacuation Planning Evacuation time studies provide Duke Energy Progress, Inc. and state/local governments site-specific information helpful to protective action decision making. The studies provide information on current population within the 10-mile Emergency Planning Zone (EPZ) and time estimates for evacuation of each area. See Table 2.1.3-6 and Figure 2.1.3-4. This information is extracted from a study, Reference 2.1.3-2, prepared by KLD Engineering in December, 2012.

REFERENCES:

SECTION 2.1 2.1.3-1 North Carolina State Fair. About Us: Attendance (1986-2013) http://www.ncstatefair.org/2013/About/Attendance.htm Retrieved 11/9/2013.2.1.3-2 KLD Engineering. Harris Nuclear Plant: Development of Evacuation Time Estimates. December 2012.2.1.3-3 Deleted 2.1.3-4 Deleted 2.1.3-5 U. S. NRC. Standard Review Plan, NUREG-0800, July 1981, Section 2.1.3.2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 2.2.1 LOCATION AND ROUTES An investigation was undertaken to locate all significant manufacturing plants, chemical plants, refineries, storage facilities, mining and quarrying operations, active military bases, transportation facilities, oil and gas pipelines, and underground gas facilities. Airplane high and Amendment 65 Page 6 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 low level flights and landing patterns (commercial, general aviation, and military) were also included in the search.There are no significant military facilities within a 25-mile radius of the plant site. The nearest active military facility is Fort Bragg (35 miles south), a support base for Army training operations.Industrial activity in the area surrounding the plant is not intensive. Durham, Guilford, Alamance, and Orange counties provide the most concentrated industrial areas within a 50-mile radius of the plant.There is some light industry at the 5600-acre Research Triangle Park, which is located approximately 20 miles NNE of the plant site. The nearest industrial facilities are near Moncure, about seven miles west of the site. There are no industrial facilities within a five-mile radius of the power block.The City of Raleigh, and to a lesser extent the City of Durham, serve as rail and highway transportation centers for the area. They are both over 15 miles from the plant site. The highways in and around these cities carry large amounts of traffic. US Highway 1, which passes approximately 6640 ft. north northwest of the plant site (Figure 2.1.1-1), has a range of 9,900 vehicles near the plant to 21,800 vehicles near Apex to 61,700 vehicles near the City of Raleigh.Rail transportation is principally for freight to and through the major cities. The CSX Railroad passes approximately 8000 feet north of the plant site. The Norfolk Southern Railroad (previously called the Southern Railroad) passed through the Main Dam area and was relocated just south of the Main Reservoir. The Durham branch of the Norfolk-Southern Railroad which ran from Bonsal, North Carolina, passed directly through the location of the plant site. A section of this branch, which ran from north of the plant site to Duncan, North Carolina, was purchased by CP&L. A small section of this purchase is utilized as an access spur to the plant. Use of the remainder of this line was discontinued.Within a ten-mile radius surrounding the plant, there is one general aviation airport, which is discussed in Section 2.2.2.5. The Cape Fear River runs southwest of the plant site, but this part of the river is not used for commercial traffic (Figure 2.1.3-1).2.

2.2 DESCRIPTION

S 2.2.2.1 Description of Facilities Tobacco manufacturing and processing is the principal industry in Durham County. Furniture manufacturing is found in Orange, Alamance, and Guilford counties. In Guilford and Alamance counties, textile manufacturing is also a very prevalent industry.The Research Triangle Park and the Raleigh/Wake County area (including Apex, Fuquay-Varina and Holly Springs) contain light industry such as electronic component manufacturing, electronic research, fiber chemistry research, pharmaceutical research, health statistics studies, and air pollution control studies. This area employs approximately 600,000 people.There is no industrial development within a five-mile radius of the plant site. However, there is a local concentration of industry which has developed in the vicinity of the Moncure community (seven miles west). Wood products, adhesive resins, and synthetic fibers are manufactured Amendment 65 Page 7 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 there by approximately 1500 employees. These facilities and their products are at such a distance from the nuclear reactor that they will pose no safety hazard to the plant site.2.2.2.2 Description of Products and Materials Sections 2.2.3 and 9.5.1 describe hazardous materials which are stored at and transported to the SHNPP site. A liquefied propane gas pipeline is located near the site, and is shown on Figure 2.2.3-1 and discussed in Sections 2.2.2.3 and 2.2.3.2. Also, there are two mining operations and five inactive quarries within a ten-mile radius surrounding the plant site. They are listed in Table 2.2.2-1 and shown on Figure 2.2.2-1.2.2.2.3 Pipelines A liquified propane pipeline (Figure 2.2.3-1) is operated by Dixie Pipeline Company and has been rerouted outside of the exclusion area north of US Highway 1. This line, 3 ft. underground, will carry approximately 1600 barrels per hour at peak flow at a maximum pressure of 1440 psi.There were three ANSI-600 through-conduit, flanged-end isolation valves installed in the relocated line, one at the midpoint and one at each end. The isolation valve locations are shown on Figure 2.2.3-1. The line will terminate at Apex, North Carolina, where the fuel will be stored and distributed for local use. The pipeline is not used for storage of gas at higher than normal pressures. There are no plans to carry any product other than liquified propane in the pipeline. There are no other petroleum operations within a ten-mile radius of the plant site.The Dixie Pipeline has easem*nts for the relocated pipeline. The conditions of the easem*nt grant the Dixie Pipeline Company a 50-foot right-of-way for their pipeline. The easem*nt grants Dixie Pipeline Company the right of ingress and egress to this right-of-way, the right to install, maintain, inspect, protect, operate, modify, replace, or remove the pipeline.2.2.2.4 Waterways River traffic on the Cape Fear River in the vicinity of the plant is limited to small boats used for pleasure. Barge traffic is not possible in the Cape Fear River upstream of Fayetteville, North Carolina (Figure 2.4.1-4). No provisions are in place to keep small boats out of the vicinity of the intake structure. However, trash screens are used in each bay.2.2.2.5 Airports Raleigh Executive Airport is located approximately six miles from the plant site. It provides no commercial services. There are from 5-10 landings per day at this airport.The nearest major airport is the Raleigh-Durham Airport, 19 miles north northeast of the plant site. There are twelve major airways branching out from this airport. Three pass within ten miles of the site (Figure 2.2.2-1). Raleigh-Durham is the only airport in the area which serves commercial traffic. Actual traffic for 1976, 1977, 1978, 1979, 1980, 1985, 1990, 1995, and 2001, and projected traffic through the year 2025 are listed in Table 2.2.2-2. The traffic is classified into air carrier, general aviation, air taxi, and military types.The closest aviation related, active military base is Pope Air Force Base, 35 miles south of the plant site and adjacent to Fort Bragg. In addition, a National Guard facility is located at Raleigh-Durham Airport.Amendment 65 Page 8 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.2.2.6 Projections for Industrial Growth As reflected by the U. S. Bureau of Economic Analysis' projections on manufacturing, between 1980 and 2020 for the six-county area of Wake, Durham, Orange, Chatham, Lee, and Harnett, industrial development will tend to expand at a rate somewhat higher than the forecast for the State of North Carolina. Using 1980 as the base year, manufacturing is expected to increase approximately 201.8 percent by 2010, and 306.2 percent by 2020.Major industrial development within a ten-mile radial distance of the plant is limited due to the lack of waste treatment and sewage facilities. Exceptions to this are in the immediate vicinity of Apex, Fuquay-Varina, and Moncure, where industrial sites are available for development.2.2.3 EVALUATION OF POTENTIAL ACCIDENTS Shearon Harris Nuclear Power Plant is situated in a non-industrialized area. However, transportation and use of some materials present a potential for explosions, fires, or release of toxic gases. The hazards associated with chemicals transported or stored in quantity in the vicinity of SHNPP were evaluated to assure appropriate design consideration. The spectrum of credible explosive events and missiles generated, as well as the delayed ignition of flammable vapor clouds, are addressed in Sections 2.2.3.1 and 2.2.3.2. The Control Room design, coupled with administrative procedures prevents the incapacitation of control room operators during postulated toxic gas episodes.The design basis events for SHNPP are discussed below.2.2.3.1 Design Basis Explosive Events Review of all combustible materials transported or stored within five miles of SHNPP revealed that the only sources which present a potential hazard are rail and/or truck transportation of high explosives, and three 10,000 gallon underground tanks containing gasoline and/or diesel fuel.In addition, combustible material is piped in the vicinity of the plant, as discussed in Section 2.2.3.2.2.2.3.1.1 Design Basis Events Arising from Transportation of Explosives The complete and instantaneous detonation, at the closest point to the plant, of one train car load of TNT (200,000 lb.) was evaluated to determine the blast loadings on critical plant structures. Loadings were calculated by methods set forth in "The Effect of Nuclear Weapons Blast," Dept. of Army Pamphlet No. 39-3, 1962, and "Design of Structures to Resist Nuclear Weapons Effects," ASCE - Manual of Engineering Practice No. 42, 1961. The maximum loads were determined to be 0.4 psi, or less, within 0.062 seconds. This loading, as well as any missiles generated by the explosion, will be satisfactorily resisted by all Seismic Category I structures and critical storage tanks. In addition, other safety related equipment which is not capable of withstanding this pressure pulse is protected against the pressure pulse by location inside an enclosure.The complete and instantaneous detonation of one truck load of TNT (approximately 50,000 lb.)would result in less severe loadings on critical plant structures.Amendment 65 Page 9 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The explosion of one of the ten thousand gallon buried gasoline tanks is considered equivalent to the detonation of approximately 100 lbs. of TNT. This assumes that the detonable mixture consists of 10 percent gasoline vapor by volume, that the remainder is air, and that the yield per pound of the mixture is the same as that of TNT.The detonation of one of these tanks is considered highly unlikely. However, if it were to occur, it would not pose a hazard to the plant safety related structures.2.2.3.2 Nearby Gas Pipeline An eight in. liquified petroleum gas (LPG) pipeline, operated by the Dixie Pipeline Company, is buried three ft. underground; it carries approximately 1600 barrels per hour at peak flow at a maximum pressure of 1440 psi. It terminates at Apex, North Carolina, where the fuel is stored and distributed for local use. The locations of this pipeline and its isolation valves are shown on Figure 2.2.3-1.The eight inch line was originally analyzed in this section as a six inch line that carried 1100 barrels per hour. The original analysis has not been changed. Rather, following the same methodology, an engineering evaluation was performed under ESR 9800222 to demonstrate that the effects of the propane gas line rupture described in this section were still acceptable with a pipeline size increase from 6" to 8". The pertinent results of this evaluation are contained in Table 2.2.3-5.The line passes in excess of 8500 ft. of the closest plant critical structures. The analyses described below form a conservative basis for evaluating plant safety-related structures. The elevation of various points along the new location of the pipeline is given on Figure 2.2.3-1.The effects on the plant safety-related structures resulting from a break in the LPG line have been evaluated on the basis of the following:a) Assumptions in Calculations:

1) Double-ended rupture or slot rupture with the slot size equal to twice the flow area of the pipeline. Rupture occurs instantaneously at the closest location to the plant.

2) The released LPG liquid-gas mixture initially escapes from the break at the critical velocity for single phase flow at the design pressure of the line (1440 psi), then drops to two phase critical flow at its saturation pressure, and finally drops to inertial flow from one end and to the flow passed by the pumps at the other end, as the pressure in the pipeline falls.

3) The temperature of the atmosphere is assumed to be 72°F. Higher temperatures would lead to somewhat higher vaporization of escaping LPG fluid, but the initial flowrate would be less due to the higher quality at the exit plane. The fluid-gas cloud size and explosive yield are largely insensitive to the assumed temperature within a temperature variation of +/- 30 degrees.

4) Propane disperses toward the plant at 1 m/sec. (Pasquill F. stability condition) as an airborne cloud, or alternatively, it drifts by gravity toward the plant.

Amendment 65 Page 10 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

5) Regardless of the potential source of ignition, a detonation of the resulting cloud is assumed to occur at selected centroids of the cloud, after the centerline concentration has reached the rich (explosive) limit.

It would take approximately 5 to 10 minutes to detect and isolate a major line leak or rupture during the day, or 30 minutes at night.b) Calculation of Flowrate Out of the Break The propane in the line will, upon the instant of the break, decompress isenthalpically to a saturation pressure of 125 psia essentially immediately because of the very large velocity of sound in the liquid. A decompression wave will travel very rapidly away from the break, leaving the fluid behind it at the saturation pressure. Since the propane would issue from the break at 72°F, approximately 1/3 of it would quickly vaporize, cooling the remainder to its boiling point of about -44°F. Hence the process of decompression is described by the throttling process in a pressure-enthalpy diagram. From such a diagram the exit plane quality, x, of the fluid can be estimated from:v = vf + x vfg where v = 2.4 ft.3/lb., vf= .0275 ft.3/lb., vg= 6.6 ft.3/lb.,vfg = vg-vf Hence x = 0.36 The value of specific volume x 2.4 ft3/lb was determined from the pressure-enthalpy diagram for isenthalpic process to ambient pressure. V is the specific volume of the 2-phase fluid.To estimate the flow rate out of the break, GCrit, Fauske's equation, (Reference 2.2.3-1) for critical two-phase mass velocity is used:1/2 GCrit =where:gc = 32.2 lbm ft./lbf sec2 k1 = (1 - x + x )x k2 = vg (1+2x-2x)+vf(2x-22x+2) k3 = [1+x(-2)-x2(-1)]and = (vg/vf)1/2 Since the derivatives are not known, they are approximated by dvg/dp vg/p; dvf/dp vf/p; dx/dp = (-1/hfg) (dhf/dp+xdhfg/dp)Amendment 65 Page 11 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2(-1/hfg) (hf/p+xhfg/p) where the differences are evaluated about the saturation pressure point( 125 psi)

= (.852/.032)1/2 = 5.16 k1 = [1 - .36 + .36 (5.16)] (.36) = 0.9 k2 = .852 [1 + 10.32 (.36) - 72] + .032 [10.32 (.36) - 10.32 - .72 (26.6) +26.6] = 3.44 ft.3/lbm vg /p = - 4.4 x 10-5 ft.5/lbm - lbf k1dvg/dp = -3.96 x 10-5 ft.5/lbm - lbf vf/p = 0 hfg = hg - hf = 291.4 - 143.4 = 148 Btu/lbm hf/p = 2.33 x 10-3 hfg/p = -1.54 x 10-3 dx/dp = 2.33 10 .36 154 10 1.2 10 .

k2dx/dp = -4.13 x 10-5 ft.5/lbf - lbm Gcrit = 1433 lb./ft.2 sec.Hence, for the given pipe area, .196 ft.2, the flowrate out of the break is at most 560 lbm/sec.,assuming that it flows out of both ends of the broken pipeline.While it may appear that such flowrates are low for upstream stagnation pressures of 1440 psi, upstream pressure decays extremely rapidly and thus the effective stagnation pressure for the line is nearly 125 psi, except for the first fraction of a second during which the flow progresses from single phase, subcooled fluid flow to the steady two-phase flow. The 125 psi represents the static pressure at the exit, which is taken to be at saturation.During the initial phase of the blowdown of the pipe (single phase flow) the maximum flowrate can be estimated from

 /

Gmax =Since the constant entropy and constant enthalpy lines essentially coincide from the initial condition of p = 1440 psi and vf = .0308 ft.3/lbm to the saturated state of p = 125 psi, vf = .0319 ft.3/lbm, the partial derivative can be evaluated as the ratio of the differences Amendment 65 Page 12 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 =1.06 lbf - lbm/ft.5 where the initial vf is taken along the isentrope at the initial pressure, vf = 0.031 ft.3/lbm.

The peak flowrate at t=0 would be 87,000 lbm/ft.2-sec. of liquid at a specific volume of 0.031 ft.3/lbm escaping with sonic velocity, i.e., 2570 ft./sec. This discharge would last a fraction of a second; i.e., the time for the wave to travel back to a location where the frictional pressure drop would equal the pressure drop to saturation. Thereafter, steady blowdown would proceed at the conservative rate of 560 lb./sec until inertial flow would be established.At a subsequent time, as the pressure in the line falls, the flow becomes inertial or two-phase critical.The portion of the piping not pump pressurized would drop rapidly to saturation pressure. The saturated liquid inertial flow, which is higher than the two phase critical flow for the same pressure, is approximately 30 lb./sec. Hence one can conservatively assume that a maximum of 30 lb./sec. would issue continuously from the unpressurized side of the break after the initial period of time required to depressurize the line to near saturation pressure. This time, for a line length of approximately seven miles, is about 40 seconds.The pump pressurized side would behave differently since the pump at peak flow can deliver approximately 70 lb./sec. This quantity is sufficient to maintain the pressure in the line at nearly 600 psi. Hence, from the pump side of the break a continuous flow of approximately 70 lb./sec.can be expected.The total long term escape flow from the break can thus be taken as approximately 100 lbm/sec.Since about 1/3 of this would vaporize, the vapor flow would correspond to a uniform source of propane vapor of 36 lbm/sec. or 324 scfs which is available for atmospheric dispersion.This vapor flowrate is of the same order of magnitude as that observed by Burgess & Zabetakis (Reference 2.2.3-2) in their investigation of a propane line break in Port Hudson, Missouri.A check on the validity of the model can be obtained by comparison of predicted and observed total quantities of propane that have escaped from actual breaks. For instance, for the Port Hudson, Missouri, break, the total release of liquid propane in barrels, estimated by Burgess &Zabetakis, during the first 24 minutes, was 750 barrels.For 100 lbm/sec. and the specific volume of saturated liquid propane (0.031 ft.3/lbm), the release during the first 24 minutes is estimated to be 795 barrels. This agreement confirms that the contribution from the subcooled or two-phase critical blowdown portion is negligible.The same methodology applied to the events at Ruff Creek (Reference 2.2.3-3) and Austin (Reference 2.2.3-4) produced the following comparison.At Ruff Creek, a 0.0174 ft.2 break is computed to have a mass flowrate of 20.5 lb./sec. during the first phase of the accident, lasting approximately 1.5 hours, and then a 5.5 lb./sec. release rate for the remaining 14.5 hours.Amendment 65 Page 13 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The predicted release quantity of propane is 1691 barrels, while approximately 1800 barrels were observed to have been released. In Austin, a 0.163 ft.2 break is computed to have released an average of 117 lb./sec. during the 1.17 hr. required to isolate the broken section.The contents of the isolated section is discharged thereafter. In this case, 6307 barrels are predicted to have been discharged versus the observed 6640.Two situations can be imagined for the ensuing clouds of propane-air mixtures that form after the break. The first one envisions a propane cloud transported by atmospheric dispersion toward the plant site. The second one assumes that the propane would form a very tenacious layer close to the ground, wherein air entrainment would occur only at the air surface of the layer. This fog-like ground-hugging cloud would advance toward the plant under its own gravity (cloud slumping), because of sloping ground.c) Calculation of Detonable Cloud Size, Assuming Atmospheric Dispersion The dimensions of the detonable plume downwind of a 324 ft.3/sec. propane source, for Category F stability and a constant invariant wind speed of 3.3 ft./sec. are given in Table 2.2.3-1.The centerline (directly downwind) concentration, Cl, of propane is determined by cl =yz u where = 324 ft.3(STP) of propane per second u = 3.3 ft./sec. (1.0 m/sec.) wind velocity and y, z are the plume dispersion standard deviations obtained from Reference 2.2.3-5.The off-centerline concentrations, are determined by

= cl exp { -1/2 [(y/y)2 + (z/z)2]}

The proper dispersion standard deviations are the continuous release standard deviations.The potential cloud configuration is plotted on Figure 2.2.3-2. The dashed portion represents the fraction of the cloud which falls within the detonable limits. These limits are reported in Reference 2.2.3-6 to be 2.8 percent and 7.0 percent, respectively, of propane in the propane-air mixture.The volume of the detonable cloud is calculated to be that of the difference between the ellipsoid enclosing all gas above the 2.8 percent mixture and the ellipsoid engulfing the gas above the 7.0 percent mixture level.Volume of ellipsoid (2.8 percent) = 1.19 x 106 ft.3 Volume of ellipsoid (7 percent) = 2.94 x 105 ft.3 Volume of detonable cloud = 9.0 x 105 ft.3 Amendment 65 Page 14 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 During the 1/2 hour period required to close the valves, the major portion of the propane vapor (90 percent) is dispersed past the lower detonable limit.d) Calculation of Effects of Detonation From an enthalpy of detonation release of 260 K cal./lb. of propane-air mixture of 4.9 percent (enthalpy of detonation is insensitive to mixture ratios between 4 and 5 percent), and a volume of mixture of 900,000 ft.3, it is possible to compute the total energy released in a hypothetical detonation of the entire detonable cloud, assuming that the whole cloud is a mixture averaging 4.9 percent of propane by volume.The total volume of propane in the detonable cloud is 4.42 x 104 ft.3 and the volume of air is 8.6 x 105 ft.3. The total weight of the mixture is approximately 68,600 .

 . . . . . / .

Total enthalpy released = 1.78 x 107 K cal.Equivalent TNT = 17.8

 ./ .

Comparison of this value with the occurrence at Port Hudson, Missouri, (Reference 2.2.3-2) shows that this estimate might really be the minimum hazard expected from the break, and that ground and atmospheric conditions may result in more propane being trapped in a detonable cloud.The absolute upper limit of the size of the detonable cloud would be a cloud where all the liquid propane which escapes from the break eventually vaporizes and mixes with air in a detonable mixture.Assuming that the heat to vaporize the liquid propane comes from the air, then it would require approximately 4.15 lb. of air at 72°F to vaporize 1 lb. of propane at -44°F.The resulting cold mixture would be 19.4 percent by weight, or 11.4 percent by volume, of propane, and it would be denser than the ambient air. Thus, it is possible to visualize a ground layer of propane-air mixture above the detonable limit, which would contribute appreciably more propane to the detonable cloud as its temperature increases than that calculated by the dispersion technique. This ground layer could eventually become dispersed in the atmosphere as its temperature increases, or it could move by gravity. For lower atmospheric temperatures, proportionately more air would be required to vaporize the propane. At near freezing temperature ( 32°F), vaporization would result in a mixture of 7 percent by volume of propane, which is the upper detonable limit.A portion of the liquid ejected from the break would be expected to be vaporized while in the form of droplets in the jet escaping from the break; the remainder of the propane would eventually evaporate as the soil and air provide the necessary heat. The rate at which the vapor is injected in the air is determined by:Amendment 65 Page 15 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

1) The amount vaporized in the jet and in the immediate vicinity. This fraction would form a heavier than air mixture which would at the same time disperse in the air and also tend to settle into a "fog-like" ground layer. Since the instantly vaporized propane (amount flashing at break) would exhibit similar behavior, the fraction of propane liquid vaporized in the jet and in the immediate vicinity can be treated, at equilibrium, as an additive to the constant vapor flow assumed for the initial dispersion calculation.

2) The amount of propane which remains liquid and experiences delayed evaporation at the ground surface.

The rate of evaporation of the liquid in contact with the ground depends on the area and depth of the liquid. Since this in turn is determined by terrain, assessment of this rate is difficult. It is reasonable to assume that an equilibrium condition can be established whereby a like amount of liquid evaporates as that which exits from the break. Less evaporation would result in an ever increasing area (or depth) of the liquid. An estimate of atmospheric clouds that could result from delayed evaporation is presented in Section 2.2.3.2e).Conservatively, however, it is possible to assume that all of the escaping propane would be atmospherically dispersed. Under this assumption, the resulting maximum detonable cloud would be that established by a constant escape of approximately 1000 ft.3/sec. of propane.In the previous calculation, the heavier than air density of the propane mixture had not been considered, but the limit in vertical dispersion was implied in the choice of z 0.5 y (characteristic of F stability conditions).Reference 2.2.3-2 recommends values of z 0.2 y. Table 2.2.3-2 and Figure 2.2.3-3 show the dimensions of the propane plume downwind of the assumed 1000 ft.3/sec. source, with z 0.2 y.This detonation would then be equivalent to the detonation of 4.06 x 105 lb. = 203 tons of TNT (assuming 100 percent yield).In actuality the yield will not be 100 percent. Reference 2.2.3-1 cites a yield of 7.5 percent.Work by Iotti, et al., (Reference 2.2.3-7) shows that the yield of a gaseous detonation is lower than that of TNT; Reference 2.2.3-7 compares overpressures calculated by assuming gaseous point sources (Reference 2.2.3-8) to overpressures obtained by Kingery (Reference 2.2.3-9) for the same yield, and those measured by Kogarko et al. (Reference 2.2.3-10), for gaseous detonations. This comparison shows that Kingery's result would have been comparable to those of References 2.2.3-8 and 2.2.3-10, if a TNT yield of 50 percent had been employed.Thus, a conservative estimate of the TNT equivalent to the detonation of the entire cloud can be obtained by using 50 percent yield and Kingery charts, see Figure 2.2.3-4, and the hypothetical detonation of the entire propane discharged by the line would result in consequences similar to the detonation of 100 tons of TNT. As explained above, this detonation cannot occur, and more realistically the detonation will be somewhere between the 8.9 tons and the 100 tons of TNT, but closer to the 8.9 tons.Table 2.2.3-3 lists the pertinent shock wave parameters for the two yields. These parameters are obtained from Reference 2.2.3-9, assuming that the center of the detonation is in excess of Amendment 65 Page 16 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 7500 ft. from the plant critical structure for the 8.9 ton detonation, and 5000 ft. from the plant critical structure for the 100 ton detonation.Table 2.2.3-3 also lists the seismic parameters at the plant site due to air blast induced ground motions. These parameters are obtained from equations in Reference 2.2.3-7.Critical plant structures are designed so that they are able to withstand the overpressures and ground motions listed in Table 2.2.3-3, hence it is concluded that a detonation of propane that has escaped from a break in the 8 in. LPG line will not result in unacceptable consequences.e) Calculation of Flammable (Detonable) Cloud Size - Ground Layer Formation and Dispersion Model The dispersion model considers the formation of a cloud by assuming that all of the propane issues from the break as a liquid, then evaporates and forms a cloud denser than air which travels by gravity until it becomes neutrally buoyant, i.e., its travel velocity is equal to or less than the prevailing wind.For a continuous release rate of liquid propane, W, the maximum pool radius, rmax is determined from the liquid pool area required to vaporize the LPG at a rate equal to the release rate, hence:

 /

rmax = (1) where B is the heat flux determinable from the following equation (Reference 2.2.3-11), which is in metric units:

 / .

30 (2)In equation (2), , k, f, g and are the surface tension, thermal conductivity, saturated liquid density, and saturated vapor density of propane at -44°F and latent heat of vaporization; Wb is an empirical constant (equal to 280 m/hr) which is the product of the diameter of a sphere that has the same volume as a bubble of average size and the frequency of bubble formation.The ground to liquid heat transfer coefficient, h, can be determined from the equation for natural convection of flat plates (Reference 2.2.3-11).

 / = 0.14 (3)

In equation (3) the subscript refers to the film properties, and k, , , Cp are the thermal

 ° conductivity, liquid density, viscosity, and heat capacity of propane at t = with tg equal to the ambient temperature.

is the volumetric expansion coefficient given by (4)Amendment 65 Page 17 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 and t = tg - t Inserting the appropriate values and converting to English units, the ambient temperature is 532 R and the heat flux B, is 33,800 BTU/ft.2/hr.This value compares well with values which can be extrapolated from the Jacob experiments (Reference 2.2.3-11) on vaporization of propane from clean surfaces, and also with values reported in LNG studies (Reference 2.2.3-12).The initial height of the vapor cloud, which is input to the gravity spreading analysis from an initial radius, rmax, is computed by determining the amount of vapor injected in the air during the time it takes the prevailing wind to traverse the pool radius, if wind is present, hence u = wind velocity hin = (5)In case of no wind, the initial height, hin, is computed by determining the equivalent instantaneous volume of a spill which would give rise to the same rmax as computed in equation (1).Vinst = (6) where the 240 is the volume ratio between liquid and vapor at -44°F, at which the propane issues from the break. For an instantaneous spill, several expressions for rmax are available.The one chosen here is from Reference 2.2.3-12.

 /

1.23 (7) with all quantities in metric units. The liquid regression rate, Q, is given by (8)The gravity spread of the cloud can be computed by

 /

2 (9) wherein c (t) is the time varying cloud density (assuming a hom*ogenized cloud at each instant of time), a is the air density at ambient temperature, g the acceleration of gravity, S the slope of the ground in ft./ft., h(t) is the time varying cloud height, and R(t) is either the radius of the cloud front at time t (for the case of flat terrain), or the distance travelled by the advancing cloud front which is confined by banks to travel along a channel of width 2W. R is the change in radius of the cloud front.As the cloud spreads, it will entrain air in an amount given by either dQ(t) = Vc 2 rdr (10a)Amendment 65 Page 18 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 when the cloud is expanding cylindrically or, dQ(t) = Vc W dr (10b) as the cloud spreads confined in a channel of width 2 W. In equation (10), is the entrainment coefficient, and Vc is the local air entrainment velocity, which is defined as:Vc = for (10a) (11a)Vc = for (10b) (11b)Integrating from r = O to r = R yields the total volume of air entrained by the cloud.for radial spreading (12a) or (12b) for initial radial spreading to W, followed by axial spreading; is a parameter which equals unity for triangularly shaped channels and two for channels with a rectangular cross section.The wind velocity, u, is additive to the radial spreading velocity upwind of the break, but subtractive downwind of it. Thus its net effect is ignored in equation (12a).The choice of the entrainment coefficient is prompted by data reported for plumes (References 2.2.3-13 and 2.2.3-14) and work by Lofquist (Reference 2.2.3-15). In Lofquist's work, the important dimensionless parameters are the Reynolds number and the densimetric Froude number, which he defines as

 / (13)

He finds that even for very low Reynolds numbers (<105), the entrainment is simply related to the F number and is independent of the Reynolds number. Figure 2.2.3-5 shows the values of derived from Lofquist's work. Examination of equations (9) and (13) shows that, in this case, wherein = c, F 2 since c > a and S is positive.From Figure 2.2.3-5, is taken as 0.1 or larger. Values reported by others are also in the range of 0.1 to 0.12. Hence a value of 0.1 is conservative (less dispersion for lower propane concentrations).Mass conservation requires that (14)Amendment 65 Page 19 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 where Mc (t) is the mass of the cloud at time t Energy conservation further requires that:(15)In equation (15), C and T are the cloud heat capacity and temperature at time t, Ca and Ta are the air heat capacity and ambient temperature in R. Qw is the heat of condensation or freezing of water vapor, which is set equal to zero in the interest of maximizing cloud travel. Qp, the rate of change of the heat content of the cloud due to the mass transfer from the pools, is computed from Qp = 98 mp; mp is the effluent rate in lb./sec., and 98 (BTU/lb. m) is the heat content of propane vapor at -44°F (the reference point is zero heat content at 0 R).Qg is the heat transfer rate from the ground to the cloud. To minimize heat transfer, it can be assumed that this heat is transferred in the natural convection regime, where:Qg = (16) with h given by equation (3).Equations (9), (14), and (15) are solved numerically given an initial cloud radius, rmax [equation (1)], height, hin [equation (5)], and temperature, -44°F, to yield the configuration of a hom*ogeneous cloud as a function of time.The progress of the cloud is either followed until the cloud concentration falls below the lower flammable limit (2.4 percent), assuming no atmospheric dispersion, or until it is stopped when the cloud velocity falls below the prevailing wind velocity, at which point atmospheric dispersion would be assumed to commence.With this approach, only average (time and space) cloud properties can be obtained.Figure 2.2.3-6 shows downwind and downslope distances achievable by a cloud of 2.4 percent concentration of propane for various wind velocities. Also shown is the final cloud height at that concentration. These are derived for the topography between the LPG line and the SHNPP, as shown on Figure 2.2.3-1.For the existing topography, low lying clouds resulting from a break in a portion of the line west of the plant could travel by gravity toward the plant until intercepted by a portion of the Auxiliary Reservoir, if the break occurred between STA 450 and 500 (shown on Figure 2.2.3-1).Otherwise, gravity would propel the cloud away from the plant, i.e., in a westerly direction.After the cloud reached the Auxiliary Reservoir, the cloud would disperse so that at the point of closest approach the fringe of the cloud would be more than 2200 ft. from safety-related structures.The cloud itself would initially have to travel a minimum of 500 ft. to reach the reservoir, after which an average channel width in excess of 300 ft. would be available for spreading.Amendment 65 Page 20 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Breaks near STA 400 could result in clouds which also could roll toward the Auxiliary Reservoir, where again channel widths of 300 ft. or more would be available.Between STA 400 and STA 340, clouds resulting from breaks could roll either northward, away from the plant, or westward into the Auxiliary Reservoir, or eastward into the depression leading to Thomas Creek. This relatively large depression is cut in half by U S Highway No. 1, so that the cloud would have to either roll over the highway or pass toward the lower elevation through three culverts underneath the highway. The average channel width of the lower depression leading to the Thomas Creek is in excess of 500 ft. at the Elevation 250 ft. contour, and 300 ft.at the Elevation 220 ft. contour.Breaks between STA 340 and approximately STA 300 could also result in clouds rolling toward Thomas Creek. East of STA 300, any break could result in clouds travelling by gravity either away from the plant or toward Little White Oak Creek. Because of the much greater distances involved, and also the somewhat larger channel widths, breaks east of STA 300 present significantly less hazard to the plant than breaks west of STA 300 and are therefore not analyzed.Figure 2.2.3-6 which involves an average ground slope of less than one percent, shows that, under extraordinary conditions in which the wind velocity closely matches the cloud velocity, the cloud can travel long distances. Although molecular diffusion, which will take place between the air and the cloud, has been ignored in the development of the previous equations, its contribution to decreasing the distances travelled by the cloud is not significant, except for long distance clouds where it becomes the predominant effect.In most instances however, the cloud will travel a distance of six to seven thousand ft. in relatively narrow channels, and less as the channel width increases.The final cloud height is relatively insensitive to the wind velocity (under the assumption of no atmospheric dispersion). The final height is determined by the break outflow; however, it is affected by channel width.An analysis of the effect of ground slope on cloud slumping concluded that ground slope is not an important factor so long as it is limited to less than one or two percent. Much higher slopes would affect the results by increasing the cloud's equilibrium speed.Except for the analyses performed for low wind velocities, it was shown that the cloud velocity soon falls below the wind speed. From that point on, clouds travelling long distances can only be sustained by assuming that atmospheric dispersion does not occur. This would of course happen only if the air flow is laminar (i.e. no disturbance), a condition which is extremely unlikely.At the point where the cloud velocity approaches the level of the wind speed, the more realistic assumption can be made that atmospheric dispersion begins. The concentration downwind can then be obtained from the equation for a continuous line source:

 = exp erf (17)

Amendment 65 Page 21 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 in which Q is the source strength in ft.3/sec., u is the wind speed in fps, and L is the width of the source (which is taken to be that of the cloud spread across the wind at the point in time at which the cloud velocity equals the wind velocity).For points along the wind direction, y = 0, equation (17) for ground levels reduces to

 =2 (18)

The virtual location of the line source is chosen upwind of the point at which the cloud and wind speed become equal, by solving the following equation u = 2 (19) where u is the computed concentration, in volume fraction, at the time when the wind velocity, u, equals the cloud velocity, . In equations (17), (18), and (19), the downwind distance does not appear explicitly, but is found by iteration for z and y, the dispersion parameters which are functions of that distance. These are the same as previously defined.Equation (18) was used to calculate the width of the source and its distance upwind that would produce concentrations of 2.4 percent of propane at a point on the ground. The results are shown in Figure 2.2.3-7 for both cases examined, i.e., y 2 y 5 z.The results shown in Figure 2.2.3-7 have been derived by assuming that the line source is pure propane. In reality, the propane concentration of the cloud at the point where atmospheric dispersion is assumed to commence, as computed from equation (19), is generally significantly lower, as a result of air entrainment accompanying gravity slumping. Thus Figure 2.2.3-7 is a very conservative upper limit of the distance downwind from the point at which gravity slumping ceases and atmospheric dispersion takes over for which the cloud is still flammable. Since it has been found that the virtual location of the line source, determined from equation 19, is upwind of the break for the channel widths and wind velocities of interest, the actual dimension downwind from the pipe break of the flammable cloud resulting from a combined gravity slumping and atmospheric dispersion is less than that predicted by Figure 2.2.3-7, which in fact considers atmospheric dispersion only. This result indicates that air entrainment during gravity slumping is more effective in reducing the propane concentration in the cloud than atmospheric dispersion. This is further evident in Figure 2.2.3-6. It is seen from this figure that very long clouds can only result from gravity slumping in those instances where the combination of channel width and mass flow produces a cloud advancing front velocity equal to the wind velocity, and when atmospheric dispersion is ignored. Otherwise, air entrainment rapidly drops the cloud bulk concentration below the flammable limit.For a channel width of 300 ft. and a wind velocity of 3.3 fps, the gravity slumping model predicts a flammable cloud length of 2500 ft. for a propane flow out of the break of 100 lbm/sec. If the channel width is increased to 800 ft., the cloud length is reduced to 2000 ft. The theory predicts that for a particular wind velocity there is a constant channel width which would result in very long clouds; two things, however, prevent this from occurring. First, the topography of the SHNPP site shows a variable channel width growing from a small dimension in the pipeline Amendment 65 Page 22 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 vicinity to a much larger dimension at the reservoirs. Second, atmospheric dispersion does take over as the cloud velocity approaches the wind velocity.From examination of Figure 2.2.3-6, cloud lengths to be expected for variable channel widths and constant wind velocity at 3.3 fps should not exceed 5000 to 6000 ft., whereas from Figure 2.2.3-7, cloud lengths in excess of 7500 ft. are not expected. Hence the credible largest cloud downwind (or down slope) dimension is between 5000 and 7500 ft.This distance is enough for the cloud to settle in either the Auxiliary Reservoir, or perhaps reach the Thomas Creek branch of the Main Reservoir depending on where the break occurs.That these are the maximum credible distances is further confirmed by the fact that the time required for the formation of such clouds is roughly comparable to the time required to isolate the break, which is approximately 1/2 hour. To generate a much larger cloud, the flow of propane from a break would have to continue uninterrupted at the assumed rate for a much longer time.For these distances, none of the flammable cloud would reach any part of the plant site. Hence, on the basis of this analysis, it is concluded that fires from such clouds would pose no hazard to the SHNPP.Similarly, the detonation of such clouds has been evaluated to present no hazard; this assessment is presented hereafter.The preceding is a simplistic analysis. The actual ground configuration is not accounted for, not even in the time varying width of the channel; and as the cloud progresses, the cloud is hom*ogenized in space and time, when in fact this will not be the case, and concentration gradients will exist within it which will affect both cloud height and velocity, hence air entrainment.Therefore another approach is used to confirm the reasonableness of the model just described.In this approach, the maximum flammable cloud extent is computed by equating the flow of propane out of the break to the quantity of propane transferred to the atmosphere across the cloud/atmosphere interface when the interface propane concentration reaches the lower flammable limit.The propane transfer rate across the cloud surface is given by N - x (N + Na) = km A(x-xb) (20) where: N = moles of propane crossing the cloud surface per unit time Na = moles of air crossing the surface per unit time A = cloud surface area km = mass transfer coefficient xb = average propane concentration in the full atmosphere (xb = 0 in this case)Amendment 65 Page 23 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 x = propane cloud concentration (assumed to be the average mole fraction of a cloud extending to either 2.4 percent (flammable limit) or 2.8 percent (detonable limit) concentration).The mass transfer coefficient, km, is given by:

 / .

0.036 (21) where: L' =distance from break to farthest reach of cloud (for continuously emitting break) ui =interface velocity at the surface (fps)

 =air viscostiy c =molar concentration of air =air density DAB =Coefficient of molecular diffusivity between two gases. . / / /

2.745 10 (22)Therein MA, MB are the molecular weights of air and of the propane cloud at the given concentration PCA, PCB are the critical pressures of air and of the propane cloud, and TCA, TCB are the critical temperatures of the same.Values of km are a function of the wind velocity (interfacial velocity) and the cloud dimension along the wind, L', and are given by km = 1.997 x 10-5 . (23)The solution to equation (20) requires knowledge of the average mole fraction of propane in the cloud. This mole fraction varies from the vicinity of the break, where it is nearly unity, to the location where it reaches the lower flammable or detonable limit. For example, for a given km and a constant mass transfer, the concentration varies inversely with the cloud surface area.An average mole fraction of propane is computed to be about 12 to 13 percent of the mixture.Flammable or detonable cloud areas for the average widths of the cloud, ranging from 300 ft.upward, and wind velocities equal to 3.3 fps, will be below 1.8 x 106 ft.2. Typical lengths of the clouds are about 6000 ft.On the basis of a cloud height of approximately 30 ft., chosen conservatively from the previous model, the total volume of detonable cloud would be 5.4 x 107 ft.3.Amendment 65 Page 24 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The average mole fraction of propane of 12 to 13 percent corresponds to an average volume percentage of approximately 15 percent. Thus, to obtain such a cloud, approximately 8.1 x 106 ft.3 propane would have to escape. At a rate of 1000 ft.3/second, it would take 2.25 hours to release this quantity of propane. Since it is estimated that only about 30 minutes would be required to detect and isolate a break in the line, clouds of this size are not expected, and the largest credible cloud will be approximately 1.2 x 107 ft.3. Only a fraction of this cloud will be within the detonable limits. If 50 percent is assumed, the volume that could detonate would be equivalent to the detonation of 118 tons of TNT. The closest approach of a cloud lying on the Auxiliary Reservoir or Thomas Creek to the critical plant structure is approximately 2000 ft. The effective center of a detonation cannot be predicted, but it could range from 2200 ft. to larger distances. Even when the center of the detonation is chosen at essentially the closest point to the Seismic Category I structures, the resulting over pressure is approximately the one psi. The plant can withstand the blast parameter values listed in Table 2.2.3-3 for the detonation. Thus, it is concluded that the detonation of flammable clouds from a LPG pipeline break present no hazards to the SHNPP.The following paragraphs discuss more fully the conclusion that no hazards are posed to the plant by either fires or detonations by a LPG pipeline break.The flammable lower limit (and detonable limit) of the cloud are reached within approximately 7000 ft. of the break when one end of the cloud is continually fed by escaping propane. This is predicted by the atmospheric dispersion model, the gravity slumping model, the combined gravity slumping and atmospheric dispersion model, and lastly by the mass transfer balance model. The mass transfer balance model, however, gives no information regarding cloud height.The atmospheric dispersion, as well as the gravity slumping, models predict cloud heights of approximately 30 ft. for the channel widths of approximately 300 ft. that exist in the vicinity of the postulated break.The average concentration of propane in a cloud that continues slumping toward the plant should fall below the lower flammable limit. However, the simplicity of the models gives no absolute assurance that this will occur.It is, therefore, conceivable that a cloud of flammable or detonable concentration could spread over the surface of either the Thomas Creek branch of the Main Reservoir or the Auxiliary Reservoir.The width available for gravity slumping in the Thomas Creek branch of the Main Reservoir is a reasonably constant 600 ft. in the immediate vicinity of the plant at the Elevation 220 ft. contour.The width of the Auxiliary Reservoir is more variable, but can be taken as an average of 1000 to 2000 ft. at the Elevation 250 ft. contour.Since the cloud height, derived on the basis of a continuously supplied, hom*ogeneous propane air cloud slumping by gravity with zero ground slope, is a conservative estimate of the actual cloud height, particularly since the propane flow should be interrupted after approximately 30 minutes, an estimate of the maximum height of the cloud in the Thomas Creek branch of the Main Reservoir and the Auxiliary Reservoir can be obtained from Figure 2.2.3-6.Amendment 65 Page 25 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 For a 400 to 600 ft. channel width, the maximum cloud height is about 15 ft. The Thomas Creek level is Elevation 220 ft. while the plant site is located at Elevation 260 ft. Hence, no portion of a flammable cloud lying in the Thomas Creek Branch of the Main Reservoir could encroach on the plant site, and therefore there is no fire hazard. The effect of detonation of such a cloud has already been addressed and found to be of no consequence.The difference in elevation between the plant site (Elevation 260 ft.) and the normal level of the Auxiliary Reservoir (Elevation 252 ft.) is eight ft. The maximum cloud height predicted for a channel width of 1200 ft. is about 9.5 ft. for a continuously fed cloud. In fact, the average width of the Auxiliary Reservoir taken at a level halfway between the Elevation 252 ft. contour (normal water level) and the plant site Elevation 260 ft., is closer to 1600 ft.; therefore, the expected maximum cloud height in the Auxiliary Reservoir would be less than eight ft.; there would be no direct fire hazard to the plant from such a cloud.f) Potential Missiles From Detonation It is difficult to assess the potential hazard to critical plant structures which could result from missiles generated by a detonation of LPG, either at the initial crater or propelled by the blast wave. Since the LPG line crosses the plant vicinity in an open area, there is little likelihood that substantial missiles would be generated other than from the place where the detonation is postulated to occur.There are three possible ways in which the detonation of the propane cloud could be initiated:

1) At the pipeline break, either by shocks from the high pressure jet of propane, static electricity, or frictional effects.

2) At structures which have operating electrical equipment that could provide initial ignition to the cloud, followed by shock propagation to a detonation.

3) Lightning.

The first and second cases are most likely to generate missiles.Work by Ahlers (Reference 2.2.3-16) on observed maximum debris distance and equivalent yield, which is reproduced on Figure 2.2.3-8, shows that the maximum range of missiles from the 8.9 tons detonation could be in excess of 7,500 ft., but that most missiles would not reach critical plant structures. Detonations of larger amounts of propane, however, could result in many missiles reaching the plant. For instance, for a 100 ton detonation, the maximum range could be in excess of 10,000 ft.Missiles which travel the longer distances, however, are expected to be smaller, since air drag will affect larger missiles proportionally more. Studies on several detonations (References 2.2.3-17 and 2.2.3-18) have shown that the mass density of missiles follows an exponential law.

 = (K) (r -3.5) where is the areal density, and r the distance from the detonation center.

Amendment 65 Page 26 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 To ascertain the probability of a missile from even the maximum credible detonation of the propane (100 ton) hitting critical plant structures, it is assumed that the total weight of the missiles is proportional to the volume of the crater which would be created by the detonation, if it had been a TNT detonation near the surface. The crater, in turn, is proportional to the detonation yield. The volume of the crater can be estimated by the scaling law:Diameter (depth) of crater = (Diameter (depth) of a crater for a 1 kt explosion) x W1/3 and the knowledge that a one kt surface detonation in dry soil results in diameters of 180 ft. and depths of 35 ft.Hence the total mass of missiles at an assumed missile density of 95 lb./ft.3 will be equal to 5.63 x 106 lb.Assuming that none of the mass falls back within the crater, then the total mass is given by M= .2 .2 .0155 K = 3.63 x 108 lb.-ft.3/2 Hence the areal density 5000 ft. away from the center of the hypothetical (considered improbable) 0.1 kt detonation is

 . . /

4.10 10 ./.

 . . /

Since the area of critical structures is approximately 105 ft.2, the total weight of missiles hitting this area for the hypothetical maximum detonation would be only 4.1 pounds.Assuming that all of this mass is concentrated in one missile, and that the missile travels at the maximum air particle velocity, up, given by up = 26.7

 /

where p is the peak overpressure at the critical structure ( .5psi), Po is the ambient pressure (14.7 psi), and Co is the ambient sonic speed (taken as 1130 ft./sec.), then the impact energy of this missile would be 45 ft.-lb., which is considerably below the energy required for penetration of the structures.Therefore, it is concluded that missiles from a propane cloud detonation would present no hazards to the SHNPP, and that no plant damage due to a LPG detonation at the plant site could occur.g) Summary and Conclusion To summarize the preceding sections, the possibility of transport of a propane cloud toward the SHNPP site has been investigated by an atmospheric dispersion model, a gravity slumping model, and a combination atmospheric dispersion-gravity slumping model.Amendment 65 Page 27 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Although the analyses demonstrate that it is possible for a flammable cloud to form in the vicinity of the plant site, fires or a detonation of the cloud will pose no hazard to the plant.2.2.3.3 Design Basis Toxic Chemicals A summary analysis of off-site and on-site toxic chemical hazards that may impact control room habitability is contained in Calculation 9-CRH. There are no significant off-site stationary sources of toxic chemicals within a five-mile radius of SHNPP.2.2.3.3.1 Releases of Toxic Materials Due to A Railroad Accident Other Than Chlorine The accidental release of toxic chemicals resulting from a railroad accident in the vicinity of the site has been calculated to be an event with a probability of occurrence of less than 10-7 per year, thus, the release of a hazardous material due to a railroad accident is not a design basis event.Three railroad segments come within five miles of the SHNPP site, which carry scheduled railroad traffic. The three segments are:a) The Bonsal-Durham segment which is 2.5 miles northwest of the plant.b) The Fuquay-Varina-Brickhaven segment which is 4.3 miles south of the plant.c) The Raleigh-Moncure segment which is 1.9 miles northwest of the plant.Only the Raleigh-Moncure segment of railroad traffic within the vicinity of the SHNPP site carries hazardous materials on a regular basis. The data used in the analysis was supplied by the Seaboard System Railroad. There was no exact break-down as to type of material transported, thus an assumption was made that of the top 25 hazardous materials shipped by rail, the distribution of shipment is even. Of these 25 materials, two are considered in the evaluation (see Table 2.2.3-4 for justifications). However, the probability of occurrence of an accidental release that could result in a release of these two materials (anhydrous ammonia and vinyl chloride), as hereafter calculated, is less than the probability for design basis events; thus, a dose analysis is not required.From the NUREG/CR 2650, SAN-82-0774 document, the statistics on number of events involving hazardous materials per mile of rail travel per year is given to be 2.2E-08. This figure is a national average, but is believed to be acceptable since no known railcar accidents have ever occurred on this segment of the railroad.The probability of such an event is given by the following equation:where:APil = Annual probability of design basis event under atmospheric class "l" involving the chemical of concern.Amendment 65 Page 28 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 P = The national average probability of a tank car containing the chemical of concern will have an accident from a mobile source per unit length of travel.N = Annual number of trips involving the chemical of concern.Mi = Annual probability of an atmospheric stability class.Dj = The length of railroad segment, in the direction of concern.Fji = Wind frequency from the sector "j" to outside air intake of the control room stability class "i".n = Number of wind direction sectors affecting the plant.The following have been determined to be the length of railroad track within 5 miles of the SHNPP site on the Raleigh-Moncure track segment:West track segment 2.0 miles West northwest segment 1.0 miles Northwest segment 0.8 miles North northwest segment 0.7 miles North segment 1.2 miles North northeast segment 2.5 miles If a train accident occurred on these segments of railroad track and the wind was blowing from that direction, an airborne cloud of hazardous materials might be transported to the SHNPP plant control room intake location.The meteorological data used in the analysis is from the SHNPP FSAR, Table 2.3.3-13, pages 2.3.3-73 and 2.3.3-74. This data is from the lowest wind measuring level on the SHNPP meteorological tower and represents a three year period of record which is representative of onsite meteorological conditions. The Pasquill stability class types "F" and "G" were the only atmospheric stability conditions used in the analysis, since under these conditions, the plume of released material would remain sufficiently concentrated so that by plume travel time over the distance from the railroad to the intake structure at the SHNPP plant, a sufficient concentration would remain so as to pose a potential concern. The frequency of calm wind conditions has been subtracted from the total wind direction frequency since under calm winds, the plume would either meander so dramatically that no concentration would be received onsite or the plume would "puddle" around the location of the release to spread in a uniform manner in all directions.The probability of an event under Pasquill stability type "F" and the probability of an event under Pasquill stability type "G" for the entire hazardous travel distance of 8.2 miles is the sum of the values calculated for each sector, resulting in a combined annual probability less than 10-7 per Amendment 65 Page 29 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 year; thus the release of a hazardous material due to a railroad accident is not a design basis event.The above railroad accident analysis was reviewed and updated in 2006 (and documented in Calculation 9-CRH). The results confirmed the overall conclusion that a railroad accident is not a credible design basis event for HNP.2.2.3.3.2 Accidental Releases of Chlorine The analyses for accidental off-site releases of chlorine are contained in Calculations CPL-XI-4 and 9-CRH. Two design basis accidents were selected: the complete loss of loading of a 20-ton tank truck on US 1 and of a 90-ton tank car on the Seaboard Railroad, both at points of nearest approach to SHNPP. The analysis found that the probability of an accident involving transient sources of chlorine was less than is required for consideration. Therefore, an accident involving the transportation of chlorine in the vicinity of Shearon Harris Plant affecting the Control Room personnel is not considered to be a credible event and need not be considered in the safety evaluation of the plant, provided that Control Room personnel have access to breathing apparatus and are trained to recognize chlorine by its odor.Positive pressure, full-face, self-contained, breathing apparatus are stored in the Control Room.2.2.3.4 Fires The only fire hazard in the vicinity of the SHNPP is the potential delayed ignition of flammable vapor clouds associated with a propane line break. This information is discussed in Section 2.2.3.2.2.2.3.5 Collision with the Intake Structure The SHNPP site is not located on a navigable waterway; therefore this section is not applicable.2.2.3.6 Liquid Spills There are no storage facilities for oil or liquids which may be corrosive, cryogenic, or coagulant, located where failure of the facility would allow these materials to be drawn into the intake structures and affect the plant's safe operation.2.2.3.7 Aircraft Operations Evaluation

REFERENCES:

SECTION 2.2 2.2.3-1 Fauske, Hans K, "Contribution to the Theory of Two-Phase One Component Flow",ANL-6633, USAEC R&D Report, 18th Ed., October, 1962.2.2.3-2 Burgess, D.S. and Zabetakis, M.G., "Detonation of a Flammable Cloud Following a Propane Pipeline Break", Bureau of Mines Report of Investigation 7752, 1973.2.2.3-3 NTSB Report PAR-78-1 "Pipeline Accident Report-Ruff Creek, PA, July, 20, 1977.2.2.3-4 NTSB Report SS-P-17 "Pipeline Accident Report-Austin, Texas, February 22, 1973.Amendment 65 Page 30 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.2.3-5 Turner, "Workbook of Atmospheric Dispersion Estimates", Public Health Service Pub. No. 999-AP-26, 1969.2.2.3-6 Benedick, W. B., Kennedy, J.D. and Morosin, B., "Detonation Limits of Unconfined Hydrocarbon-Air Mixtures" Combustion and Flame, 15. pp. 83-84, 1970.2.2.3-7 Iotti, R., Krotiuk, W.J., and DeBoisblanc, D.R., "Hazards to Nuclear Plants From On (or Near) Site Gaseous Explosions", CONF 730304, USAEC, 1973.2.2.3-8 Brode, H. L., "Numerical Solutions of Spherical Blast Waves", J. App. Phys. 26, 766, 1954.2.2.3-9 Kingey, D., "Air Blast Parameters vs Distance for Hemispherical TNT Bursts", BRL Report No. 1344, Ballistic Research Laboratories Aberdeen Proving Ground, Maryland, 1966.2.2.3-10 Kogarko, Adushkin and Lyamin, "Investigation of Spherical Detonations of Gas Mixtures" Combustion, Explosion and Shock Waves, 1, No. 2, 1965, pp. 22-34.2.2.3-11 Jacob, M. "Heat Transfer" John Wiley & Sons, Vol. 1, 1949.2.2.3-12 Germeles and Drake, "Gravity Spreading and Atmospheric Dispersion of LNG Vapor Clouds, 4th Int. Symposium on Transport of Hazardous Cargoes by Sea and Inland Waterways, U.S. Coast Guard, Jacksonville, Florida, October 1975.2.2.3-13 Morton, B.R., "The Ascent of Turbulent Forced Plumes In a Calm Atmosphere", Int.J. Air Pollution, 1, p. 184.2.2.3-14 Fisher, R.D. et al, "Prediction of the Rise of A Gaseous Plume in a Calm Atmosphere with Any Lapse Rate Profile" 2nd. Int. Clean Air Congress of Int'l Un. Air Poll., 1970.2.2.3-15 Lofquist K., "Flow and Stress Near and Interface Between Stratoped Liquids" The Physics of Liquids, 3, p. 158, 1960.2.2.3-16 Ahlers, E. B. "Debris Hazards, A Fundamental Study" DASA 1362, Illinois Institute of Technology Research Institute, Chicago, Illinois.2.2.3-17 Kaplan, K., Sears, P. and Melichar, J., "The Hazards to the Brunswick Steam Electric Power Plant Caused By An Explosion Near the Sunny Point Munitions Terminal, Section D, Live and Inert Missiles", URS Report 700-1, May, 1968.2.2.3-18 Brode, H.L., "The Hazards to the Brunswick Steam Electric Power Plant Caused By An Explosion Near the Sunny Point Munitions Terminal, Section D Missiles from Accidental Explosion-Distant Hazards."2.2.3-19 ASHRAE Handbook of Fundamentals, ASHRAE, New York, N. Y. 1967.2.2.3-20 Moody, F. J., "Maximum Flow Rate of a Single Component. Two Phase Mixture" Transcript of ASME, Journal of Heat Transfer, February, 1967, pp. 134-142.Amendment 65 Page 31 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.2.3-21 Bird, Steward, and Lightfoot, "Transport Phenomena", Ch. 21, John Wiley and Sons, 1960.2.3 METEOROLOGY 2.3.1 REGIONAL CLIMATOLOGY 2.3.1.1 General Climate The SHNPP site lies in the transition zone delineating the Coastal Plain Region and the Piedmont Region of North Carolina. The climatology of North Carolina largely depends on elevation above sea level and distance from the Atlantic Ocean. At an elevation of about 260 ft.above mean sea level and a distance of about 115 mi. from the nearest Atlantic coastline, the site area has a temperate climatic regime. Stations representing the regional climatology, their locations with respect to the site area, and their elevations above mean sea level are presented in Table 2.3.1-1. Information and data provided is based upon information available prior to the issuance of the Harris Plant's operating license.The summer months of June, July, and August are characterized by a southwesterly air flow resulting from the extension of the Azores-Bermuda high pressure system. This Gulf of Mexico and occasionally Atlantic moisture laden air produces the bulk of precipitation for these months in the form of afternoon and evening thundershowers. During this three-month period, an average of 39 days reach 90°F or hotter as reported by the Raleigh-Durham Weather Service, the nearest first-order reporting station to the site area. July is the hottest month at all stations within the site area. These months can be quite oppressive with dewpoints averaging between 66 and 67°F (Reference 2.3.1-1).The autumn months of September, October, and November show a gradual decrease of average temperature of about 10°F per month. As the sun moves south, days become shorter and correspondingly, nights become longer. The combination of residual summer moisture and increased radiational cooling due to longer nights makes this the season of highest fog frequency. Although precipitation is distributed rather uniformly on an annual basis, the autumn months tend to be the driest. Though moisture is usually available, daytime heating is not sufficiently intense to produce convective activity, and the general north-south temperature gradient does not substantially materialize to generate strong frontal precipitation. Winds tend to be from the northeast during the autumn, reflecting a change in the pressure distribution. The summer wind flow configuration of a high pressure system offshore and a low pressure system over the continent is replaced by the northerly wind flow configuration of a continental high pressure system with a low pressure system offshore. The land-sea temperature contrast favors higher pressure over the ocean in spring and summer and higher pressure over the continent in autumn and winter, thus providing the seasonal reversal of wind. The higher autumnal northeastern frequency, when compared to the winter frequency, is the result of the slower moving continental high pressure systems of autumn that tend to prolong the associated northeasterly wind flow.The winter months of December, January, and February show a shift of the wind direction frequency into the southwesterly and northwesterly quadrants responding to a westerly component added to the predominant southwest northeast bimodal distribution. January is the coldest month, averaging 18 days with a minimum temperature below 32°F at the Raleigh-Durham Weather Service (Reference 2.3.1-1). However, cold air outbreaks are either blocked Amendment 65 Page 32 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 or significantly modified by the Appalachian Mountain chain located some 150 miles to the west and northwest of the site. Most sustained winter precipitation is the result of two storm tracks.One track originates in the warm waters of the western Gulf of Mexico, crosses Florida, then skirts the Atlantic Coast as it moves northward. The second track is called the "Cape Hatteras Low," so named because the temperature contrast of the offshore Gulf Stream and the shape of the coastline just south of Cape Hatteras, North Carolina, provide excellent breeding conditions for cyclonic circulations. These two storm tracks are responsible for virtually all of the snowfall in the site area; January has the greatest average snowfall totals.The spring months of March, April, and May are characterized by consistently rising temperatures on the order of 9°F per month. Precipitation occurs in a mixed mode of frontal and convective forms. This transitional season generally has more winter than summer characteristics. The mean date of the last 32°F temperature for the area is around the first week in April (Reference 2.3.1-1). Maximum wind speeds are generally observed in this season due to the peak of the general north-south temperature gradient.2.3.1.2 Regional Meteorological Conditions for Design and Operating Bases 2.3.1.2.1 Tornadoes The SHNPP site lies within Region I for determining the Design Basis Tornado (Reference 2.3.1-2). The Region I associated Design Basis Tornado parameters are as follows:Maximum wind speed 360 mph Rotational wind speed 290 mph Translational speed 70 mph maximum; 5 mph minimum Radius of maximum rotational speed 150 ft.Pressure drop 3.0 psi Rate of pressure drop 2.0 psi/sec.Calculation of the tornado strike probability is accomplished by the following equation:

 / (1) where:

Ps = Probability that a tornado will strike a particular location during a one-year interval

 = Average number of tornadoes per year, equal to 1.46 for the SHNPP site area (Reference 2.3.1-3).

a = Average individual tornado area, equal to 2.82 sq. mi. for the SHNPP site area (Reference 2.3.1-3).Amendment 65 Page 33 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 A = Total area of concern (e.g., 1 square with 35° 30' mid/latitude) equal to 3891.15 sq.mi.Using these parameters, Ps is equal to .00106 or stated conversely, a return period of 944 years. Consequently, one would expect a tornado strike every 944 years. Since no large body of water capable of sustaining waterspouts is located near the site, these need not be considered.2.3.1.2.2 Thunderstorms Charlotte, Greensboro, and Raleigh-Durham have a mean total of 42, 47, and 46 thunderstorm days per year, respectively (References 2.3.1-4 through 2.3.1-6). The distribution by month is presented in Table 2.3.1-2. July has the most thunderstorms; these occur mainly in the afternoon and evening, and most frequently they are the scattered, air-mass type. This provides variable precipitation patterns; on the average, there is one thunderstorm every third day in July. Thunderstorms which occur in the autumn through spring period are usually the result of frontal activity, rather than the result of the convective heating process that prevails during the summer months.2.3.1.2.3 Lightning Three kinds of lightning occur in thunderstorms: cloud-to-cloud, in-cloud, and cloud-to-ground.Although cloud-to-cloud and in-cloud strokes outnumber cloud-to-ground strokes by about two to one (Reference 2.3.1-7), cloud-to-ground strokes present the only hazard to nuclear power plant safety. Table 2.3.1-3 presents the annual and seasonal frequencies of cloud-to-ground flashes for Charlotte, Greensboro, and Raleigh-Durham (latitude variations excluded). These frequencies are calculated with a technique outlined by Marshall (Reference 2.3.1-8) by using the following equation:NE = (0.1 + 0.35 sin d) (0.40 +/- 0.20) (2) where:NE = Number of flashes to earth per thunderstorm day per sq km.d = Geographical latitude of the SHNPP, equal to 35° 35' Using the conservative estimate of 0.40 + 0.20 (equaling to 0.60) in the above equation, the site area NE equals to 0.182 flashes per thunderstorm day per sq. km.2.3.1.2.4 Hail Hail is an indication of strong vertical velocities that occur in severe thunderstorms. Storms reaching hail severity stage are mostly associated with strong frontal zones during late spring to late summer; they are infrequent in the plant site area.The Raleigh-Durham Weather Service observed hail a maximum of three times in any one year during the period 1951-1978, and has never experienced hail more frequently than once in any one month. Hail occurred most frequently in May and only during the months of March through August, except for one instance in November and one instance in February (Reference 2.3.1-6).Amendment 65 Page 34 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.1.2.5 Ice Storms (freezing rain)The U. S. Weather Service defines glaze as ". . . hom*ogeneous, transparent ice layers which are built upon horizontal, as well as on vertical surfaces either from supercooled rain or drizzle, or from rain or drizzle when the surfaces are at a temperature of 32°F or lower". Although glaze may occur at air temperatures far below 32°F, the majority of ice storms occur with air temperature between 25°F and 32°F. Ice storms at the SHNPP site are caused primarily by polar front waves that have their origin in the Gulf of Mexico. Below-freezing temperatures seldom last in this area more than a few hours after ice storms. Consequently, the mean duration of ice on utility wires during the period 1928-1937 was 21 hours. The greatest radial thickness on utility wires observed during the nine winters of the period between 1928 and 1937 was .74 in. During the 10-year period of 1939-48, a total of 40 freezing precipitation days were observed at the Raleigh-Durham Weather Service; 17 of these days occurred in January.Therefore, an annual average of four freezing precipitation days per year was experienced (Reference 2.3.1-9).2.3.1.2.6 Hurricanes Sustained hurricane force winds (>74 mph) have never been recorded by the Raleigh-Durham Weather Service, although they have been observed in coastal areas of the State. Hurricanes deteriorate rapidly as they move onshore because of increased frictional drag and a loss of the energy source (water through release of latent heat of evaporation and condensation). Once onshore, the increased frictional effects have a tendency to turn the winds inward towards the hurricane's center; this yields greater vertical velocities which are capable of producing intense rainfall. Since the SHNPP site lies approximately 115 miles from the nearest coastline, the major effect on the Raleigh area due to hurricanes is heavy precipitation. A list of hurricanes that have affected the Raleigh area are listed in Table 2.3.1-4. The maximum 24-hour precipitation of 5.20 in. at the Raleigh-Durham Weather Service was the result of Hurricane Diane in August 1955. The fastest one-minute wind observed at the Raleigh-Durham Weather Service was associated with Hurricane Hazel in October 1954 (Reference 2.3.1-6). As would be expected, the intensities of wind and precipitation produced by hurricanes at the plant site are generally no greater than those produced by severe thunderstorms in the area.During the period 1901 to 1955, destruction due to tropical storms occurred about ten times in the plant site area. Consequently, one would expect a return period of a destructive tropical storm of 5.5 years (Reference 2.3.1-10).2.3.1.2.7 Extreme Winds Using a Fisher Tippett Type II extreme value distribution, Thom (Reference 2.3.1-11) has calculated and plotted the annual extreme-mile 30 ft. level, 100-year mean recurrence interval winds for the United States. From this publication, the SHNPP site extreme-mile 100-year recurrence period wind speed is 90 mph. The vertical distribution of velocity is presented in Figure 2.3.1-1, computed from the l/7 power law and the 30 ft. level, 90 mph value. Other return periods are shown by Figure 2.3.1-2.The extreme-mile wind is defined as the one-mile passage of wind with the highest speed. This includes all meteorological phenomena except tornadoes, which are dealt with separately. The extreme-mile wind does not reflect gustiness occurring during a short time interval. As an adjustment, Huss (Reference 2.3.1-12) suggests that a gust factor of 1.3 be applied to the 30 ft.Amendment 65 Page 35 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 level extreme-mile wind. Therefore, an instantaneous gust of 117 mph would be expected to occur at the 30 ft. level once within a 100-year period.The highest observed wind speed recorded at the Raleigh-Durham Weather Service is a 79 mph wind in September 1996, at Greensboro a 63 mph wind from the north in July 1932, and at Charlotte a 59 mph wind from the southwest in July 1962 (References 2.3.1-6, 2.3.1-5, and 2.3.1-4). The 90 mph extreme-mile Thom value is greater than the highest observed wind speed at Raleigh-Durham and therefore, can be used as a conservative value in additional analysis.2.3.1.2.8 Precipitation extremes Table 2.3.1-5 lists precipitation extremes along with other parameter extremes for the Charlotte, Greensboro, Pinehurst, Asheboro, Moncure and Raleigh-Durham stations (References 2.3.1-4, 2.3.1-5, 2.3.1-6, and 2.3.1-13). The maximum monthly precipitation total observed in the plant site area was 13.88 in. at Pinehurst in July, 1959. The maximum 24-hour precipitation recorded in the site area was 8.96 in. at Asheboro in August, 1966. Conversely, the minimum monthly precipitation recorded in the site area was a trace at Charlotte in October, 1953. The maximum monthly snowfall for the region was a 22.9 in. total which fell in Greensboro during January, 1966. The maximum 24-hour snowfall total for the site area was 14.3 in. which fell in Greensboro in December, 1930.The site area on ground snow load, 100-year mean recurrence interval is 15 lbs. per sq. ft.(Reference 2.3.1-14). The minimum design roof snow load is determined by multiplying the on-ground snow load of 15 lbs. per sq. ft. by the basic snow load coefficient of 0.8. This gives a value of 12 lbs. per sq. ft. as the 100-year recurrence interval roof snow load. Additionally, the 48-hour probable maximum winter precipitation occurs in December and is approximately 19.2 in. (water equivalent) using a 200 sq. mi. reference area (Reference 2.3.1-15). The month of December averages only 1 in. of snowfall, therefore, it is very improbable that the probable maximum winter precipitation would fall in the form of snow. The floors of the unroofed areas of the Tank Building and the roofs of other safety-related structures where water due to probable maximum winter precipitation (PMWP) can accumulate, are structurally capable of withstanding the dead-load combination due to the PMWP and 12 lb./ft.2 (snowpack load), in combination with the dead-loads defined in Sections 3.8.1.3.1 and 3.8.4.3.1.2.3.1.2.9 Atmospheric conditions The extent of vertical mixing is a major factor in determining atmospheric diffusion characteristics. As a rule, mixing depths are characterized by a diurnal cycle of a nighttime minimum and a daytime maximum. The nighttime minimum is the result of surface radiational cooling which produces stable conditions, frequently coupled with low level temperature inversions or isothermal layers.The mid-afternoon maximum is attributable to surface heating which produces instability and convective overturning through a larger portion of the atmosphere. Mean mixing depths also show a seasonal cycle of a winter season minimum and a summer season maximum.Holzworth has shown this (Reference 2.3.1-16) by listing monthly mean maximum mixing depths. Table 2.3.1-6 lists these results for Greensboro (nearest data point to the plant site).The lowest mean maximum mixing depth occurs in January (390m), and the greatest mean maximum depth occurs in June (1790m).Amendment 65 Page 36 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Low level temperature inversions also inhibit vertical mixing. Hosler (Reference 2.3.1-17) has compiled frequencies based on the percent of total hours of occurrence of an inversion or isothermal layer based below 500 ft.; the frequency of low level temperature inversions for Greensboro are presented in Table 2.3.1-7. The summer season averages inversions about 33 percent of all hours. Comparatively, the winter season averages inversions approximately 43 percent of all hours.Cases of high air pollution potential occur during periods of stagnating anticyclones which exhibit low surface winds, no precipitation, and shallow mixing depths that result from a subsidence inversion. These conditions occur most frequently at the plant site during the fall months, particularly October. According to Korshover (Reference 2.3.1-18), about 32 cases of autumnal atmospheric stagnation that lasted four days or more occurred during the 35-year period from 1936 to 1970. A total of four cases that lasted seven days or more were recorded during the same 35-year period.2.3.1.2.10 Ultimate heat sink The meteorological data that was used for evaluating the performance of the ultimate heat sink (Main or Auxiliary Reservoir) are discussed in Section 2.4.11.7.2.3.2 LOCAL METEOROLOGY 2.3.2.1 Normal and Extreme Values of Meteorological Parameters.The local meteorology is based upon SHNPP on site data collected from January 14, 1976 through December 31, 1978, and offsite data from Charlotte, Greensboro, Raleigh-Durham, Moncure, Asheboro, and Pinehurst. Normal and mean data for the National Weather Service stations of Charlotte, Greensboro, and Raleigh-Durham are based on the 1941-1970 recording period; Moncure, Asheboro, and Pinehurst normal temperature data are based on the 1951-1973 recording period, normal precipitation data on the 1941-1970 period. Data provided is based on information available prior to issuance of the Harris Plant's operating license.2.3.2.1.1 Wind Wind direction and speed distributions are essential parameters for determining site characteristic diffusion climatology. Onsite joint frequency distributions of direction and speed by stability class and a summary of all winds, as outlined by Regulatory Guide 1.23 (Rev. 0)(Reference 2.3.2-1), for the period January 1976 through December 1978 are given by Tables 2.3.3-12 and 2.3.3-13. Tables 2.3.3-16 and 2.3.3-17 show joint frequencies for the lower and upper level wind direction and speed, respectively, by atmospheric stability class. Annual and seasonal wind roses for Raleigh (Reference 2.3.2 2), Greensboro (Reference 2.3.2-3), and Charlotte (Reference 2.3.2-4) are illustrated by Figures 2.3.2-1 through 2.3.2-6, respectively.The Raleigh-Durham Weather Service (1955-1964) joint frequency distribution of wind direction and speed by Pasquill stability classes is given in Table 2.3.2-1. Pasquill stability classes were determined by the STAR method. Stability classes F and G were combined into F stability.Observations of wind speeds less than 3 knots were directionally distributed according to the frequency of occurrence of speeds from 3 to 6 knots in each direction category.Amendment 65 Page 37 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Despite differing techniques used to determine atmospheric stability (delta temperature method for on-site data and the STAR method for Raleigh data), the on-site joint wind frequencies for the SHNPP site (Tables 2.3.3-12 and 2.3.3-13) compare favorably to those compiled for Raleigh. Neutral (D) and slightly stable (E) stability classes occur more frequently at both stations. However, stable (F) and extremely stable (G) stability classes are more frequent at the on-site meteorological station. This is due in part to some nighttime cold air drainage into the broad, shallow basin in which the site is located (see Section 2.3.2.2.2).The characteristic northeast-southwest bimodal frequency distribution is evident at all locations and is depicted by the on-site wind rose given in Figure 2.3.2-7. Average wind speeds from the area offsite stations are rather uniform, ranging from 6.9 mph at Charlotte to 7.9 mph at Raleigh-Durham.The on-site lower level (12.5m) mean wind speed based on 1976-1978 data is 4.6 mph. This on site value is about 35 percent lower than the 7.9 mph value observed at the Raleigh-Durham Weather Service. Differing time periods, averaging methods, and instrumentation account, in part, for the lower on site wind speed value.However, topography probably is the single most influential factor determining the lower average on-site wind speed. The SHNPP site lies in a broad, shallow 200 ft. deep basin that extends about 10 miles in directions west through south of the site (See Section 2.3.2.2.2).Hypothetically, the basin has a discoupling effect on the wind flow in the site area, particularly during nighttime hours when cold air drainage into the basin can be expected. This colder air is denser than the surrounding environment and hence difficult to displace. Therefore, turbulence and mean wind speed tend to be suppressed.From the seasonal wind roses, the southwesterly component is more evident in the spring, summer, and winter seasons. The northeasterly fall season distribution is the result of the combination of a trend toward continental high pressure systems introducing a northerly wind flow and the relatively slow movement of these synoptic systems due to weak upper steering currents prevailing at this time of year. Winds from the southeastern quadrant are rare and for the most part precede warm frontal passages.Wind persistence is defined here as the number of consecutive hours during which the wind direction was from the same 22.5° direction sector. Table 2.3.2-2 shows the number of occurrences of persisting wind directions by stability class as recorded at both upper and lower on-site levels of operation at the SHNPP site for January 1976 through December 1978. The maximum persistent wind for the upper level was from the south-southwest and lasted 37 hours ending at 1:00 p.m. on December 9, 1978. The same synoptic pattern produced the maximum persistent wind direction at the lower level, which lasted 30 hours. Maximum persisting winds at both levels were of the same direction and end time. Figure 2.3.2-8 is a graph of the number of persisting wind direction hours for all directions versus cumulative probability of wind persistence occurrence for the same period of record as Table 2.3.2-2. An estimate of the percentage of the total time a known number of wind persistence hours occurs can be taken directly from Figure 2.3.2-8. For example, 10-hour wind direction persistence from any one of the 16 compass directions occurred about 2 percent of the total hours.Sustained winds greater than 50 knots have occurred only twice in the past 24 years as recorded by the Raleigh-Durham Weather Service. A one-minute average 69 mph wind from the southwest was recorded during a thunderstorm on July 21, 1962. The maximum site area Amendment 65 Page 38 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 one-minute average wind of 73 mph from the west northwest was recorded during Hurricane Hazel on October 15, 1954 (Reference 2.3.2-6). A complete list of hurricanes affecting the site area, the amount of precipitation, and fastest-mile wind associated with each is given by Table 2.3.2-3. As would be expected, the intensities of wind and precipitation produced by hurricanes at the plant site are generally no greater than those produced by severe thunderstorms in the area.2.3.2.1.2 Temperature Monthly and annual summaries of climatological normal maximum, minimum, and average temperatures for Raleigh-Durham (Reference 2.3.2-6), Greensboro (Reference 2.3.2-7),Charlotte (Reference 2.3.2-5), Moncure (Reference 2.3.2-8), Pinehurst (Reference 2.3.2-8), and Asheboro (Reference 2.3.2-8) are given in Tables 2.3.2-4 through 2.3.2-9. Monthly and annual on-site average temperature data for January 1976 through December 1978 is presented in Table 2.3.2-10. The mean maximum and minimum temperature data from the onsite meteorological station is shown in Table 2.3.2-11. The site area diurnal temperature range spans from about 20°F in the winter and summer seasons to around 25°F in the transitional autumn and spring months (Reference 2.3.2-9). Measured maximum and minimum temperature extremes for the offsite stations are summarized in Table 2.3.2-12. The lowest temperature recorded was -9°F in January 1985 at Raleigh Durham and the highest recorded temperature was 107°F at Moncure in July 1952 (Reference 2.3.2-5 through 2.3.2-8).2.3.2.1.3 Water Vapor Mean monthly and annual dew point temperatures and corresponding absolute humidity values for Raleigh-Durham, Charlotte, and Greensboro are given in Table 2.3.2-13 (Reference 2.3.2-9).Monthly and annual on-site dew point temperatures for the period January 1976 through December 1978 are given in Table 2.3.2-14. The on-site average dew point of 47.4°F compares very well to the 48°F average dew point observed at Raleigh-Durham. On site winter dew point temperatures tend to be lower and summer values a little higher. The maximum persisting 12-hour surface dew point temperature for the period of record for the site area is approximately 77°F; it occurred during a period of extended air flow trajectories from the Gulf of Mexico (Reference 2.3.2-9).Diurnal variations of relative humidity for Charlotte, Greensboro, and Raleigh-Durham are given in Tables 2.3.2-15 through 2.3.2-17 respectively for local standard times of 1:00 a.m., 7:00 a.m.,1:00 p.m., and 7:00 p.m. (Reference 2.3.2-5 through 2.3.2-7). The 7:00 a.m. and 1:00 p.m.times correspond to the general maximum and minimum respective values of the diurnal relative humidity cycle, with 1:00 a.m. and 7:00 p.m. providing approximate midrange values. The late summer to early fall maximum of early morning (7:00 a.m.) relative humidity values also results in the same seasonal maximum of radiational fog frequency.2.3.2.1.4 Precipitation Precipitation is rather uniformly distributed on an annual basis in the site region. Tables 2.3.2-4 through 2.3.2-9 give climatological normal monthly and annual precipitation amounts for area recording stations (Reference 2.3.2-5 through 2.3.2-7). On-site precipitation totals are summarized in Table 2.3.2 18. Climatologically, July has a tendency to be the wettest month, October the driest, but the variance is small such that the region does not possess a "wet" and Amendment 65 Page 39 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 "dry" season. Extreme precipitation amounts for area recording stations are listed in Table 2.3.2-12 (Reference 2.3.2-5 through 2.3.2-8). The extreme rainfall rates summary for the on-site facility for the January 1976 through December 1978 period is shown in Table 2.3.2-19.The on-site extreme rainfall rates for all time periods included in the table occurred on the same date, March 21, 1976, with a maximum 24-hour precipitation total of 4.41 in.Yearly, the site area receives precipitation one day in three. Table 2.3.2-20 displays precipitation statistics for the site area stations of Raleigh-Durham, Greensboro, and Charlotte (Reference 2.3.2-10). These statistics are presented for the months of January, April, June, and October which are considered representative of the four seasons. Table 2.3.2-20 indicates that, on the average, precipitation intensities of July are about double those of January for the site region. Table 2.3.2-20 further characterizes the more convective nature of higher intensity, shorter duration July precipitation versus lower intensity, longer duration January precipitation.Generally, winter precipitation duration is about twice as long as that of July. However, daily rain totals are generally smaller in winter due to the low precipitation rates that offset the longer winter durations. The transitional April and October months seem to fit the winter regime better, partly due to precipitation dependence on slower moving rain systems in the transitional seasons; however, in mid-winter the systems are larger in area and of more uniform intensities than in the transitional period. On-site data showing the number of hours with measurable precipitation by month and year, including the overall average for the January 1976 through December 1978 period, is depicted in Table 2.3.2-21.Seasonal and annual precipitation wind roses from the Raleigh-Durham Weather Service (Reference 2.3.2-2) are illustrated by Figures 2.3.2-9 and 2.3.2-10. On-site precipitation wind roses for the period January 1976 through December 1978 are presented on Figure 2.3.2-11. A northeast-southwest wind frequency distribution is the dominant flow regime during precipitation periods for both stations. Extreme precipitation totals for nearby representative stations are shown by Table 2.3.2-12 along with measured extreme snowfall totals (References 2.3.2-5 through 2.3.2-8).Table 2.3.2-25 gives joint frequencies for the lower wind direction (10m) and lower wind speed by precipitation rate classes. The precipitation rate classes are divided into 0.2 inches per hour increments.Table 2.3.2-26 gives joint frequencies for the upper level (60m) wind direction and upper wind speed by precipitation rate classes. These precipitation rate classes are also delineated into 0.2 inches per hour increments.Wind directions from the northeast quadrant predominate with lower precipitation rates associated with the more common synoptic scale storms that produce general rainfall of lighter intensities.Higher precipitation rates are coupled with a southwesterly wind flow associated with the less frequent mesoscale thunderstorms that produce high intensity rainfalls for short durations.2.3.2.1.5 Fog For the period 1950-1977, heavy fog (visibility 1/4 miles) occurred at Raleigh-Durham on an average of 36 days per year, with the fall and winter months showing the greater number of days of nearly 4 per month (Reference 2.3.2-6). The most common type of fog occurring in the Amendment 65 Page 40 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 SHNPP area is ground fog as a result of nighttime radiational cooling. For this purpose, fog may be defined as a stratus cloud occurring at the surface with its top at the base of the radiationally induced temperature inversion. Ordinarily ground fog occurs more frequently in the early morning hours near sunrise when the daily minimum surface temperature is reached. It is usually shallow and disappears shortly after sunrise (Reference 2.3.2-11). Seasonally, the greater frequencies of occurrence are the fall and winter as a result of the combination of relatively moist air at the surface due to summer vegetation and the increasingly longer nights in the early fall. Also, the continental anticyclone which develops in late summer results in low surface winds, especially near sunrise, to give a consistently high frequency of fog occurrences during August and September; however, the fog persists for only about 5 hours. Strong nightly radiational cooling coupled with a predominance of anticyclonic circulation, which results in marked stability in the lower atmosphere, produces the greater fog frequencies.2.3.2.1.6 Atmospheric Stability Table 2.3.2-22 gives onsite frequencies of Pasquill Stability categories for the 1976-1978 period.Temporal variations of frequencies within the individual stability classes are small. Almost 50 percent of all hours fall into either neutral (D) or slightly stable (E) stability categories. Nearly 20 percent of all hours fall into the extremely stable (G) stability category. Extremely unstable (A),moderately unstable (B), and slightly unstable (C) stability categories combined occur only approximately 16 percent of the total hours.Stability persistence data for the on-site meteorological station is given in Table 2.3.2-23. The longest stability persistence period recorded during the 1976-1978 period occurred in the neutral (D) stability category; it lasted 62 hours ending at 12 a.m. on January 2, 1978. Four periods of neutral (D) stability persisted between 49 and 72 hours during the period 1976-1978.Neutral stability primarily occurs during day-night transitions; extended periods occur under cloudy skies with moderate winds in both day and night. Over 50 percent of neutral stability occurrences persisted two hours or less; the fewer, longer persistence periods occurred during cloudy sky, moderate wind conditions. The many occurrences of short duration neutral stability persistence offset the fewer number of longer persistence periods resulting in an average duration of 4.2 hours.Extremely stable (G) stability is prevalent under nighttime conditions of light surface winds and clear skies which allow maximum surface radiational heat loss. These restrictions tend to limit the maximum persisting hours of the extremely stable stability to shortly before sunset to shortly after sunrise, a maximum of approximately 16 or 17 hours. Consequently, five periods of very stable (G) stability lasting 16-19 hours were recorded at the on-site meteorological station from 1976-1978. Once G stability is set up, it is likely that the initial conditions will persist throughout the night resulting in the on-site average G stability persistence duration of 7.6 hours.2.3.2.1.7 Monthly Mixing Heights Mixing height data is presented in Section 2.3.1.2.9.Amendment 65 Page 41 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.2.2 Potential Influence of the Plant and Its Facilities on Local Meteorology 2.3.2.2.1 Cooling tower impact on local meteorology Data used to evaluate the potential impact of the cooling tower on local meteorology is based upon information available prior to the issuance of the Harris Plant's operating license. The natural draft cooling tower that is used to dissipate waste heat to the atmosphere is not expected to have a significant influence on local meteorology. This is due primarily to the height of discharge (approximately 520 ft. above plant grade). After leaving the tower, the plume may rise another 1000 to 3000 ft., depending on wind speed and atmospheric temperature conditions. At these elevations, the additional water and heat added to the atmosphere will not significantly affect conditions at ground level.At full power, between 8,000 and 12,000 gpm of water (depending on weather conditions) will be evaporated and discharged to the atmosphere by the tower. Under most meteorological conditions, the discharge will condense upon leaving the tower and will be visible (as condensed water vapor) until it is evaporated to invisibility after mixing with the drier (unsaturated) air in the atmosphere. The length of the visible plume depends on the temperature and humidity of the atmosphere. Colder and more humid weather is conducive to longer plumes. Most of the time, the visible plume will extend only a short distance from the tower and then disappear by evaporation. This was shown in a study at Keystone, where 97.3 percent of the time the plume length was less than 5000 ft. (Reference 2.3.2-12). On very humid days, when longer plumes are expected, there may be a naturally occurring overcast.On such occasions it is difficult to distinguish the cooling tower plume from the overcast.After a quick increase of the invisible plume radius after it leaves the tower, entrainment or mixing with ambient air keeps the visible plume radius constant or decreases it. Long persistent visible plumes occur with stable air, hence, vertical mixing is very limited and plumes tend to flatten after injection into a layer due to initial plume rise and then maintain a constant horizontal dimension normal to the wind due to lateral mixing. The Keystone photographs show this to be the case (Reference 2.3.2-12).The plume lengths and orientation with respect to the plant were determined for all hours with visibilities greater than one-half mile during the three-year period. Plume frequencies were calculated in 820-foot (250 meter) plume length intervals. Plume characteristics were categorized by season and annual average.Figure 2.3.2-12 shows the annual cumulative frequency of plume lengths from the tower. A large proportion of the plumes (about 85 percent) will be confined to plant property; 97 percent of the plumes will have lengths less than 1.5 km, and 99.6 percent less than 2.5 km. The maximum plume length expected is 3.5 km, occurring on average once every three years; plumes 3.0 km in length can be expected about one hour per year, and 2.0 km plumes about 10 hours per year.The nearest airport to the plant is the Raleigh-Durham Airport, located 29 km northeast of the plant. There will be no safety hazard created to air traffic.Table 2.3.2-24 shows seasonal frequencies of 1, 2, and 3 km plumes associated with wind direction. The greatest frequency of the plumes occurs during winter and fall months. The largest plumes are expected during the winter; this is due to the fact that condensation is Amendment 65 Page 42 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 enhanced and plume lengths increase with increasing ambient moisture content and decreasing temperature. The greatest frequency of plumes is associated with north to northeast and south to southwest winds, which indicates the importance of colder temperatures (winds with northerly component) and greater moisture (winds with southerly component) in producing plumes.Figure 2.3.2-13 shows the annual frequency of plumes associated with the operation of the natural draft tower. Results are similar to those given for Table 2.3.2-24. The greatest frequency of plumes is expected north of the plant, and the longest plumes are expected southwest of the plant.Ground fogging could occur if ground elevations in the plant vicinity were comparable to plume heights; however, the release elevation at the plant is approximately 780 ft. msl and the highest ground elevations in the vicinity are 430 ft. msl (8 km southeast of the site), and 400 ft. msl (10 km west of the site). Plumes will easily clear these areas without considering the rise of the plume above the release elevation.Extended visible plumes will probably occur during periods of high humidity when restricted visibility occurs naturally. An average frequency of 845 hours per year of naturally occurring fog was reported during the three year period 1955-1957; visibilities were less than one-half mile for 124 hours per year during this period and less than three-eighths of a mile for 90 hours per year. The tower will therefore only slightly increase the severity of the condition.Ice formation on structures is not expected to occur if the structure is lower than half the cooling tower height. From a 520 ft. tower and a plume rise that extends 1,000 ft. above the tower in the most stable case, the plume will not ordinarily pass across any structure having a height less than the cooling tower. The only exception to this will be at high-wind speeds. While plumes will be very short when winds are strong, occasionally the wake effect of the tower will cause the plume to curl below the lip. Flow around the cylindrical natural draft tower quickly removes the downwash, and it either ascends or evaporates. The downwash at Keystone was not observed to go more than approximately 75 ft. below the lip of the tower.There are no large safety-related plant structures or other nearby structures which could be affected by icing from this cause. During times of naturally occurring snowfall, it is conceivable that snow conditions could be more intense under the plume and cause greater accumulation on the surrounding area and roadways. This should not create any greater hazard, since normal precautions taken by travelers in such circ*mstances would be adequate. Such an effect is expected to be very local if it occurs.A theoretical study showed that collection of water by rain falling through the plume contributes less than one percent at rainfall rates greater than 1 mm/hr. Therefore, significant increases in rainfall are not expected in the area surrounding the tower.No synergistic effects of cooling tower operation at the site location have been identified.Gaseous effluents will be released from the plant from rooftop vents approximately 120 ft. above grade. None of these vents release significant amounts of heat; therefore, there will be only a small plume rise. As stated previously, the cooling tower plume will be at a much higher elevation. If the two plumes eventually mix, it would be well downwind where any water droplets in the cooling tower plume would have evaporated, and the gas concentrations in the plant effluent plume would be well diluted.Amendment 65 Page 43 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 A very small fraction of the water circulating through the cooling towers will be carried as small droplets in the rising air which leaves the tower top. This drift rate fraction (defined as kg of water per second leaving the tower top divided by the kg of water per second circulating through the tower heat exchange section) will average about 2 x 10-5 (or 0.002 percent).2.3.2.2.2 Topographic Features The SHNPP site lies within a shallow basin, as depicted by Figures 2.3.2-14 through 2.3.2-17 which give plots of elevation versus distance from the plant center by directional sectors.Generally, within 10 miles of the site, elevations above mean sea level gradually increase from the plant grade of 260 ft. to around 400 ft. in all but the west-southwest, southwest, north-northwest, north, and north-northeast sectors.Topographic features within a 5-mile radius (as modified by the plant) are shown on Figure 2.3.2-18. Filling of the main reservoir south and southeast of the plant adds an additional heat and moisture source to the area. As a result, a slight increase of speed and frequency of southerly winds are expected. Additionally, the heat source will tend to destabilize the nighttime surface inversion, thereby reducing the frequency of occurrence of Pasquill Class G stability.Topographic features within a 50-mile radius are shown on Figure 2.3.2-19. In general, the terrain slopes upward northwest of the site area, averaging about 10 ft. per mile, to an elevation of about 800 ft. at 50 miles from the plant site. The terrain through the north through west sectors is gently rolling, ranging from about 100 ft. to 500 ft. above mean sea level.2.3.2.3 Local Meteorological Conditions for Design and Operating Bases Local meteorological data have not been used for design and operating basis considerations other than those conditions referred to in Sections 2.3.4 and 2.3.5. However, local meteorological data are conservative with respect to the design and operating parameters, and are discussed in the sections outlined below:Ultimate Heat Sink 9.2.5 Tornadoes 3.3.2 Winds 3.3.1 Flooding 3.4 2.3.3 ON-SITE METEOROLOGICAL PROGRAM 2.3.3.1 On-Site Operational Program Collection of SHNPP onsite meteorological data began in March 1973. A 200-ft. guyed, open-latticed tower supports the lower and upper levels of instrumentation. Wind direction, wind speed, wind variance (sigma theta), and two ambient temperatures are monitored at both the lower and upper levels. Two channels of differential temperature between the upper and lower levels are monitored simultaneously. Solar radiation and precipitation are collected near ground level. The wind sensors are mounted on 12-foot fiberglass booms oriented perpendicular to the general NE-SW prevailing wind flow to minimize tower shadow effects. The temperature probes Amendment 65 Page 44 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 and relative humidity sensor are housed in aspirated shields mounted on 8-foot fiberglass booms. Current operational sensor elevations are displayed in Table 2.3.3-3.The meteorological tower is located approximately 1.1 miles northeast of the reactor complex.The base of the tower is at approximate plant grade level of 260 ft. above mean sea level. A topographical map showing the meteorological tower with respect to the reactor complex is given on Figure 2.1.2-1.An environmentally controlled shelter, which houses the meteorological system datalogger and remote data access equipment, is located about 40 ft. northwest of the tower, perpendicular to the prevailing wind flow to minimize air trajectory deviations. A complete illustration of the meteorological facility layout is presented in Figure 2.3.3-1.The datalogger acquires signals from the meteorological sensors and converts those signals into engineering units. 15-minute averaged values for each parameter are stored internally within the datalogger memory. The memory is battery-backed to prevent loss of stored data during power outages. A date/time stamp on each 15-minute average recorded data set corresponds to the end time of the data interval. The data is transmitted to the plant computer system for display and archiving. The Company's offsite meteorological consultant will be permitted access to the stored data for download.2.3.3.2 Data Reduction A host computer located in the office of the Company's meteorological consultant retrieves the meteorological data from the HNP datalogger system daily (except weekends and Holidays).This data is reviewed for potential immediate data problems by meteorological personnel. The datalogger data is then rigorously checked for consistency with the Raleigh-Durham National Weather Service data. Erroneous data is then discarded prior to insertion into the historical data base. The edited 15-minute averaged data is then stored on magnetic history tapes.Routine computer outputs include:

1. Monthly Data Summaries listing maximum temperature, minimum temperature, average temperature, barometric pressure, precipitation, solar radiation, and dew point temperature as a daily average and monthly average.

2. Hourly averages of precipitation, barometric pressure, ambient temperature, differential temperature, dew point temperature, upper and lower level wind direction and wind speed, upper and lower level wind direction variance (sigma theta), Pasquill stability classes (as outlined in Regulatory Guide 1.23) computed from the average of the two delta temperature systems, and accumulated solar radiation (langlies/minute).

3. The 15-minute averages of all parameters, except precipitation which is displayed as a 15-minute total value.

4. Joint wind frequency distributions (as outlined in Regulatory Guide 1.23) for both upper and lower levels showing average wind speeds and number of unrecovered data hours.

Amendment 65 Page 45 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.3.3 Maintenance and Calibration System reliability is supported by performance of the following maintenance:

1. The datalogger input channels are calibrated semi-annually.

2. The wind sensors are calibrated or replaced with NIST-traceable calibrated sensors semi-annually.

3. The precipitation collector is calibrated semi-annually.

4. The barometric pressure, relative humidity and solar radiation channel sensors are calibrated or replaced with NIST-traceable calibrated sensors annually.

5. The temperature sensors are thermistors purchased with NIST-traceable calibration documentation. Since thermistors are inherently stable (100 month drift <0.01°C),

routine sensor calibration or replacement is not performed. Deviation between the two ambient/differential temperature channel indications provides early warning of a problem with one of these channels.Routine analysis of the accumulated system data, including comparison to appropriate alternate weather data sources, provides an opportunity to screen for inconsistent of erratic data.2.3.3.4 On Site Data Westinghouse System on-site joint wind frequency distributions (compiled per Regulatory Guide 1.23) for both upper and lower sensor elevations for the period January 14, 1976, at 4:00 p.m.EST through December 31, 1978, at 11:00 p.m. EST is presented in Tables 2.3.3-6 through 2.3.3-13. Upper and lower level annual joint wind frequency distributions for January 14, 1976, at 4:00 p.m. EST to December 31, 1976, at 11:00 p.m. are displayed in Tables 2.3.3-6 and 2.3.3-7, respectively. Data recovery percentages for this period are 96.3 percent for the lower level and 95.8 percent for the upper level.Tables 2.3.3-8 and 2.3.3-9 depict the annual joint wind frequency distributions for the period January 1, 1977, at 12:00 a.m. EST through December 31, 1977, at 11:00 p.m. EST for the upper and lower levels, respectively. Data recovery percentages for this period are 93.6 percent for the lower level and 91.8 percent for the upper level.Annual joint wind frequency distributions for both upper and lower levels for January 1, 1978, at 12:00 a.m. EST through December 31, 1978, at 11:00 p.m. are given by Tables 2.3.3 10 and 2.3.3 11, respectively. Data recovery percentages for this period are 97.9 percent and 98.6 percent for the upper and lower levels, respectively.Both upper and lower level distributions for the entire three-year period are given by Tables 2.3.3-12 and 2.3.3-13, respectively. Data recovery percentages for this period are 95.2 percent for the upper level and 96.2 percent for the lower level.All on-site joint wind frequency distributions were compiled by using the delta temperature stability classifications as outlined by Regulatory Guide 1.23.Amendment 65 Page 46 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Average on site wind speeds for the total three year period at the lower and upper levels are 4.6 mph and 8.9 mph, respectively. Representation of the data to long-term, climatological averages is discussed in Section 2.3.1 and 2.3.2.Joint wind frequency distributions by stability class for data recorded via the data logger sensor system for the period February 1, 1979, through January 31, 1980, are given in Tables 2.3.3-14 and 2.3.3-15.Table 2.3.3-16 provides joint frequencies for the lower level (10mm) wind direction and lower wind speed by atmospheric stability class (per Regulatory Guide 1.23). These frequencies represent three year monthly totals for the period 1976-1978.Table 2.3.3-17 provides joint frequencies for the upper level (60m) wind direction and upper wind speed by atmospheric stability class (per Regulatory Guide 1.23). The frequencies represent the three year monthly totals for the period 1976-1978.2.3.3.5 Regional Air Flow Trajectory Considerations.Meteorological data and analysis of the preceding sections have included onsite and representative offsite stations both within and outside an 80 km radius of the plant. Because of the hom*ogeneous nature of the topography and climatology of the parameters that govern atmospheric transport processes, the analysis presented in the preceding sections is also sufficient to characterize transport processes to within an 80 km radius.All wind data used in analysis were from the first order Weather Service Stations, Raleigh-Durham, Greensboro, and Charlotte. All have exposures typical of airport locations. Table 2.3.3-18 shows wind instrument height above ground level for each of these stations.2.3.4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES 2.3.4.1 Objective On site data from SHNPP for the period of January 1976 through December 1978 (Table 2.3.3-

13) has been used to evaluate the design basis accident meteorology for the SHNPP site. The design basis accidents are postulated (and evaluated in Chapter 15) to characterize highly unlikely physical events, upper limit radioactivity concentrations, and doses at on-site and off-site locations. Among the basic inputs to the accident analyses are the meteorological parameters which determine the dilution capacity of the atmosphere.

2.3.4.2 Calculations Diffusion calculations for accidental or short-term releases of radionuclides were performed in accordance with the criteria provided in Draft NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants,"September 30, 1977. It was assumed that the releases emanate from a point source at ground level; no advantage is claimed from effluent emissions at elevated release points. The effluent plume is assumed to spread according to a Gaussian dispersion model, except during atmospheric conditions exhibiting neutral (D type) or stable thermal structure (E, F and G type) conditions as defined by Regulatory Guide 1.23 accompanied by light wind speeds less than 6 Amendment 65 Page 47 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 m/sec. During the expected case, atmospheric dispersion includes the consideration of the "plume meander" using a method described in Draft Regulatory Guide 1.145.2.3.4.2.1 The diffusion model for eight hours or less. Design basis accident /Qs are calculated using one of the following three formulae:

 /Q (1) / (2) /

or

 / (3) where:

/ = is the relative concentration (sec/m3) at ground level

 = is 3.14159 = is the wind speed (m/sec) at ten meters above ground level y = is the lateral plume spread (m), as a function of atmospheric stability, wind speed

( ) and downwind distance from release. For distances to 800 meters, y =My: M being a function of atmospheric stability and wind speed (see Figure 2.3.4-1). For distances greater than 800 meters, y = (M-1)y 800 m + y.y = is the lateral plume spread (m), as a function of atmospheric stability and distance, (Figure 2.3.4-2).z = is the vertical plume spread (m), as a function of atmospheric stability and distance (Figure 2.3.4-3).A = is the smallest vertical plane, cross-sectional area (m2) of the building from which the effluent is released.During conditions of neutral (D) and stable (E, F, and G) stability when the wind speed at the 10-meter level is less than 6 m/sec, credit for horizontal plume meander was considered by determining /Q using equation (1). This value was used if it was less than the higher of the /Q values obtained using equations (2) or (3). Otherwise the /Q value used was the greater value calculated from either Equation (2) or (3).Wind velocities were grouped as indicated in Table 2.3.4-1. /Q values were computed for each stability class defined by Regulatory Guide 1.23 using the corresponding wind group for each of the 16 cardinal directions. The critical values are summarized in Table 2.3.4-5.Amendment 65 Page 48 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.4.2.2 Selection of Downwind Distances In order to allow for changes in the airflow trajectory, plume segmentation (particularly in the light wind, stable condition), wind speed and direction frequency variations from year to year, the following procedure was used to determine the distance at which the calculations of atmospheric dilution (/Q) were made.For each of the 16 cardinal wind direction sectors, the distance used for /Q computations at the minimum exclusion area boundary was the minimum distance from the original SHNPP plant center to the nearest point of the exclusion area boundary within a 45-degree sector centered on the compass direction of interest.2.3.4.2.3 Choice of Dilution Factor (/Q)To choose the appropriate /Q value for the dose assessment analysis, cumulative probability distributions of the /Q values, as determined from the method described in Section 2.3.4.2.1 at a specified distance from the plant center were constructed for each of the 16 cardinal compass point directions (22-1/2 degree direction sectors). Each directional probability distribution was normalized to 100 percent. Since the joint frequency table data was used to calculate the /Q values, the cumulative probability distribution function was computed such as to envelop the data points.The effective probability level (Pe) for the selection of the /Q value in each direction sector was defined by the following equation:

 /

Pe = (4)Where:P = probability level.N = total number of hours having wind and stability data in the meteorological data record.n = total number of hours having wind flow in the direction of interest.S = total number of Sectors (16).For the realistic accident assessment /Q determination, P should be selected as 50 percent.Note that Pe can exceed 100 percent if n is sufficiently small. In those directions, the selection of a /Q value may be ignored unless the /Q values for that sector are very high when compared with /Q values of Pe in other direction sectors.For each effective probability level (Pe = 5% or Pe = 50%), the /Q values for each of the cardinal directions are compared to each other and the highest value is utilized as the atmospheric dilution factor to be assumed in the accident analysis.Amendment 65 Page 49 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.4.2.4 Results of dilution computations Using the described procedure and the joint frequency data compiled from on-site meteorological data, /Q values were calculated for the SHNPP site. Table 2.3.4-2 presents the distances to the nearest SHNPP site boundary for the 16 cardinal directions. Table 2.3.4-3 summarizes the minimum distances from the original SHNPP plant center to the exclusion boundary and the distances used in the accident dilution assessment as prescribed by NRC Draft Regulatory Guide 1.145, September 1977. Table 2.3.4-4 presents a summary of the zero-to two-hour, 5 and 50 percentile /Q values at the exclusion boundary for each direction. The dilution factor which is exceeded 5 percent of the time is 6.17 x 10-4 seconds/meter3 in the south sector at the minimum exclusion boundary. The dilution factor of 5.1 x 10-4 seconds/meter3 was calculated for west sector at the actual site boundary.Values of /Q for other points of interest are presented in Table 2.3.4-5.The dilution factors for the outer boundary of the low population zone (LPZ) have been calculated for 0-8 hours, 8-24 hours, 1-4 days, and 4-30 days for each wind-direction sector.(Minimum distance to the outer boundary of the LPZ is three miles). The /Q value was determined for the appropriate time period at the distance of interest in each direction sector by using a logarithmic interpolation between the calculated value that was selected in that sector of interest for the 0-2 hour period and the annual average (8760 hour) value at the distance of interest in that direction sector. The computation of the annual average value is described in Section 2.3.5. The appropriate time period was selected from the interpolation and the highest sector /Q value was selected.Table 2.3.4-5 is a summary of the worst, 5 and 50 percentile values of /Q for all time periods and distances of interest.As discussed in Section 2.3.2, the on-site data sample was considered to be conservatively representative of meteorological conditions at the site. Observed differences between the three year on-site data period (Section 2.3.3.4) and ten year joint frequency distribution from the Raleigh Durham Airport (Section 2.3.2.1.1) reflect the bimodal wind direction distribution characteristic of the area. Although the percentage of calm winds annually at the site are about one-half of those recorded at the airport (probably due to differences in instrumentation) the percentage of stable atmosphere conditions are about 30 percent above those of the Raleigh-Durham Airport (probably due to differences in Raleigh-Durham "STAR" Stability Classification and the Regulatory Guide 1.23 Stability Classification System).Topographic influences at the site are discussed in Section 2.3.2.2. The area of gently rolling hills will have only slight effect on short-term diffusion estimates. It is anticipated that after creation of the SHNPP reservoir system, a slight improvement in diffusion conditions will be observed by alteration of the area's micrometeorological regime.2.3.5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES 2.3.5.1 Objective The on-site meteorological record (January 1976 through December 1978) has been used to provide estimates of annual average atmospheric dilution factors (/Q) for locations at selected distances out to 50 miles from the plant. These /Qs are used to evaluate the dispersion of Amendment 65 Page 50 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 radionuclides released during routine plant operations through air pathways. The on-site meteorological data recorded on magnetic tape, has been provided to the NRC separately.2.3.5.2 Calculations On-site annual joint frequencies of wind direction, wind velocity, and stability class were determined from hourly averages. The temperature differences were measured between the 11.03 meter and 59.85 meter temperature sensing levels. These parameters were used as input to a computerized Gaussian model, which calculates annual average /Q values for distances out to 50 miles from the Shearon Harris Nuclear Power Plant. The basic equation used in the diffusion model is:

 /

(5) 3 (6)Where:

 = average effluent concentration normalized by source strength at distance x and direction k; = mid-point values of the ith wind speed class; = vertical (z) spread of the effluent at distance x for jth stability class; = joint probability of the ith wind speed class, jth stability class, and kth wind direction.

x = downwind distance from release point or building;

 = reduction factor due to radioactive decay at distance x for the ith wind speed class; = reduction factor due to plume depletion at distance x for the ith wind speed class, jth stability class, and kth wind direction; = correction factor for air recirculation and stagnation at distance x and kth wind direction; and Dz = the building height which is used to describe the dilution due to the building wake.

Equation (5) yields the /Q using the maximum building wake dilution allowed by the NRC; the computer code uses the higher value of (/Q) calculated from Equation (5) and (6).Amendment 65 Page 51 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The computer code used to generate the annual long-term values is described in NUREG-0324, "XOQDOQ Program For the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Plants", September 1977. The recirculation factors for an inland location are specified as input, along with the exclusion boundary distances and the special points of interest.The annual average /Q values for the exclusion boundary, site boundary and low population zone boundary are presented in Tables 2.3.5-1, 2.3.5-2, and 2.3.5-3 respectively. Values for annual average /Q no decay, undepleted; annual average /Q 2.260 day decay, undepleted; annual average /Q 8.0 day decay, depleted; and annual average D/Q relative deposition for incremental distances to 50 miles are presented in Tables 2.3.5-4, 2.3.5-5, 2.3.5-6, and 2.3.5-7, respectively.The annual average dilution factors given in this section are likely to be quite conservative for the reasons given in Section 2.3.4.2.5.

REFERENCES:

SECTION 2.3 2.3.1-1 U. S. Department of Commerce, Environmental Data Service, "Climatic Atlas of the United States," June, 1968.2.3.1-2 Regulatory Guide 1.76, "Design Basis Tornado for Nuclear Power Plants," April, 1974.2.3.1-3 "Technical Basis for Interim Regional Tornado Criteria," WASH-1300, U.S. Atomic Energy Commission, Office of Regulation, May, 1974.2.3.1-4 Charlotte, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.1-5 Greensboro, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.1-6 Raleigh-Durham, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.1-7 Byers, H. R., "General Meteorology," McGraw Hill, Inc., New York, New York, 1974.2.3.1-8 Marshall, J. Lawrence, "Lightning Protection," John Wiley & Sons, 1973.2.3.1-9 Bennett, I., "Glaze-Its Meteorology and Climatology, Geographical Distribution and Economic Effect," Headquarters Quartermaster Research and Engineering Command, U. S. Army, Natick, Mass. Tech. Rep. EP-105, 1959.2.3.1-10 "Climatography of the United States No. 60-31, Climates of the States, Climate of North Carolina," U. S. Department of Commerce, Environmental Science Services Administration, Environmental Data Service, Silver Spring, Maryland, First Printed February 1960, Revised June 1970, Page 19.Amendment 65 Page 52 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.1-11 Thom, H. C. S., "New Distribution of Extreme Mile Winds in the United States," ASCE Environmental Engineering Conference, Dallas, Texas, February 1967.2.3.1-12 Huss, P. O., "Relation Between Gusts and Average Wind Speeds," Report No. 140, David Gugenheim Airship Institute, Akron, Ohio, 1946.2.3.1-13 "Climatography of the United States No. 60, Climate of North Carolina," U. S.Department of Commerce, National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina, Reprinted July 1978.2.3.1-14 ANSI A58.1 "Building Code Requirements for Minimum Design Loads in Buildings and Other Structures," American National Standards Institute (1972).2.3.1-15 "Seasonal Variation of the Probable Maximum Precipitation East of the 105th Meridan for Areas from 10 to 1000 Square Miles and Durations of 6, 12, 24, and 48 Hours,"Hydrometeorological Report No. 33, U.S. Department of Commerce, U. S.Department of Army Corps of Engineers, Washington, D. C., April 1956.2.3.1-16 Holzworth, G. L., "Estimates of Mean Maximum Mixing Depths in the Contiguous United States," U. S. Weather Bureau Research Station, Cincinnati, Ohio, May 1964.2.3.1-17 Hosler, C. R., "Low-Level Inversion Frequency in the Contiguous United States,"Monthly Weather Review, Vol. 89, September 1961, Pages 319 to 332.2.3.1-18 Korshover, Julius, "Climatology of Stagnating Anticyclones East of the Rocky Mountains, 1936-1970, NOAA Technical Memorandum ERL ARL-34, U. S.Department of Commerce, National Oceanic and Atmospheric Administration, October, 1971.2.3.2-1 Regulatory Guide 1.23 (Rev. 0), "On site Meteorological Programs," February 17, 1972.2.3.2-2 "Wind Distributions by Pasquill Stability Classes (Star Program), Raleigh, North Carolina, 1955-1964," U. S. Department of Commerce, Environmental Science Services Administration, National Weather Records Center, Asheville, North Carolina.2.3.2-3 "Wind Distributions by Pasquill Stability Classes (Star Program), Greensboro, North Carolina, 1966-1970," U. S. Department of Commerce, National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.2-4 "Wind Distributions by Pasquill Stability Classes (Star Program), Charlotte, North Carolina, 1966-1970," U. S. Department of Commerce, National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.2-5 Charlotte, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.Amendment 65 Page 53 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.3.2-6 Raleigh-Durham, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.2-7 Greensboro, North Carolina, 1978, "Local Climatological Data, Annual Summary with Comparative Data," National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina.2.3.2-8 "Climatography of the United States No. 60, Climate of North Carolina," U. S.Department of Commerce, National Oceanic and Atmospheric Administration, National Climatic Center, Asheville, North Carolina, Reprinted July 1978.2.3.2-9 U. S. Department of Commerce, Environmental Data Service, "Climatic Atlas of the United States," June, 1968.2.3.2-10 Saucier, W. J., "Some Features in the Diurnal Variations of Rain and Wind Over North Carolina," Water Resources Research Institute, University of North Carolina, June 1972.2.3.2-11 Byers, H. R. "General Meteorology," McGraw Hill, Inc., New York, New York, 1974, pp. 277-279.2.3.2-12 Bierman, G.F. et. al., "Characteristics, Classification, and Incidence of Plumes from Large Natural Draft Cooling Towers," presented at the American Power Conference, 33rd Annual Meeting, April, 1971.2.4 HYDROLOGIC ENGINEERING 2.4.1 HYDROLOGIC DESCRIPTION 2.4.1.1 Site and Facilities The Shearon Harris Nuclear Power Plant is located between Tom Jack Creek and Thomas Creek. These creeks are two of the tributaries of Whiteoak Creek; Whiteoak Creek is a tributary of Buckhorn Creek. A complete description of the site area is presented in Section 2.1. The location of the site relative to the various tributaries of Buckhorn Creek and the Cape Fear River is shown on Figure 2.4.1-1.The Seismic Category I structures which should be considered from the hydrologic standpoint include the Main Dam, the Auxiliary Dam, their respective spillways, and those safety related structures located on the plant island (see Section 3.2). The top of the Main Dam, the top of the Auxiliary Dam, and nominal plant grade are all at Elevation 260 ft., while the maximum water level at the corresponding locations resulting from the probable maximum flood coincident with the corresponding designed wind velocity are 243.1 ft., 258.0 ft., and 257.7 ft., MSL respectively (see Tables 2.4.5-1, -2).Emergency service water and cooling tower makeup water for the power plant are supplied through the Emergency Service Water and Cooling Tower Makeup Water Intake Structure. The preferred source of emergency service water is the Auxiliary Reservoir. Water from the Auxiliary Reservoir passes through the Emergency Service Water Intake Channel and the Amendment 65 Page 54 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 associated intake screening structure to the Emergency Service Water and Cooling Tower Makeup Water Intake Structure. The cooling tower makeup water source and the secondary source of emergency service water is the Main Reservoir.Water from the Main Reservoir passes through the Cooling Tower Makeup Water Intake Channel to the Emergency Service Water and Cooling Tower Makeup Water Intake Structure.The Emergency Service Water is discharged from the Emergency Service Water Discharge Structure into the Auxiliary Reservoir. The decks of the Emergency Service Water and Cooling Tower Makeup Water Intake Structure, the Emergency Service Water Intake Screening Structure, and the Emergency Service Water Discharge Structure are all at Elevation 262 ft.MSL. This elevation is 2.0 ft. above the plant grade, 4.0 ft. above the probable maximum water in the Auxiliary Reservoir, and 18.9 ft. above the probable maximum water in the Main Reservoir. All three of these structures are Seismic Category I reinforced concrete structures and are capable of withstanding wave action during the probable maximum flood and the probable maximum hurricane.As shown on Figure 2.4.1-2, the southern half of plant grade is above the existing ground, while the northern half extends to the cutting region of the pre-existing ground contours. Filling and cutting of the pre-existing ground has altered the existing local drainage pattern. A description of on-site water bodies and drainage patterns is presented in Section 2.4.1.2.1.5. Most of the runoff from the plant grade is drained by means of a graded ground surface with inlet structures and associated underground reinforced concrete pipe. Along the peripheral areas of the plant island, the drainage system consists of open ditches and underground reinforced concrete pipe.The reinforced concrete pipe drain to the Main Reservoir or to the sides of the Emergency Service Water Intake and Discharge Channels.The plant drainage system is designed for a storm of five in. per hour rainfall intensity. The maximum accumulation of rainwater on the plant island during the local probable maximum precipitation storm is 14.8 in. (see Section 2.4.2.3). For protection of Seismic Category I structures and safety-related systems, see Section 3.4.1.1. Storm runoff will flow into the Main Reservoir through the underground reinforced concrete pipe and into the Auxiliary Reservoir via the Emergency Service Water Intake and Discharge Channels. Should the flow through the drainage system for the plant island area become blocked during a period of such rainfall intensity, the plant island is capable of being drained by overland flow on the open roads and ground surface directly to the Main Reservoir or the Emergency Service Water Intake and Discharge Channels. Sediment buildup in the Emergency Service Water Channels and Auxiliary Reservoir is monitored in accordance with the requirements of Reg. Guide 1.127, Rev.1.2.4.1.2 Hydrosphere 2.4.1.2.1 Surface and Groundwater Hydrology The principle source of water for the SHNPP is a storage reservoir system, which consists of two reservoirs. The Main Reservoir, situated on Buckhorn Creek, is impounded by an earthen dam located just below the confluence of Whiteoak Creek and Buckhorn Creek, while the Auxiliary Reservoir, located on Tom Jack Creek, is formed by an earthen dam situated to the west of the plant island. There are two creeks adjacent to the plant site; Tom Jack Creek to the west and Thomas Creek to the east. No pre-existing ponds or impoundments were located within the boundary of the plant island.Amendment 65 Page 55 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The location of the Cape Fear River basin is shown on Figure 2.4.1-3. Details of the Cape Fear River drainage basin and its relation to Buckhorn Creek are shown on Figure 2.4.1-4.2.4.1.2.1.1 Buckhorn Creek Buckhorn Creek has its headwaters in the vicinity of Holly Springs and Apex, North Carolina, and flows along a southwesterly course to its confluence with the Cape Fear River about 12 miles northwest of the town of Lillington. As shown on Figure 2.4.1-1, Buckhorn Creek has five tributaries above the Main Dam: Tom Jack Creek, Thomas Creek, Little Whiteoak Creek, Whiteoak Creek, and Cary Creek. These five creeks, together with the remainder of Buckhorn Creek's basin, drain a watershed area of approximately 79.5 sq. mi. The entire drainage basin lies near the eastern edge of the Piedmont Plateau, with elevations from 450 ft. to 150 ft. MSL.There is no flow record of the Buckhorn Creek drainage basin before 1972, except for a total of 13 days of low flow records taken at a station near Holly Springs, North Carolina, during isolated periods of drought in the general area. However, the Middle Creek basin lies adjacent to the eastern border of the Buckhorn Creek basin and also has its headwaters in the vicinity of Apex.A permanent United States Geological Survey stream gaging station on Middle Creek near Clayton, North Carolina, gages the runoff from 80.7 sq. mi. Records have been maintained at this gaging station since October 1939, and the overall average annual runoff for the period 1939 through 1978 is 15.53 in. Due to the immediate proximity and similar size of these basins, the overall average flow of Buckhorn Creek correlates reasonably well with that of Middle Creek. Therefore, flow data of Buckhorn Creek from October 1939 through May 1972 were derived by utilizing the records of flow of Middle Creek modified by multiplying the ratio of their drainage areas. In order to synthesize flow data for Buckhorn Creek prior to this period, six other streams with long term flow records and comparable drainage areas which are in the same general area were analyzed and correlated with those of Middle Creek for the overlapping period of record. These include the following:USGS Gaging Station Records Drainage Areas (sq. mi.)a) Little River near Princeton 1930-1969 229 b) Deep River at Ramseur 1922-1969 346 c) Deep River at Randleman 1928-1969 120 d) Reedy Fork near Gibbonsville 1928-1969 133 e) Flat River at Bahama 1925-1969 150 f) Eno River at Hillsboro 1927-1969 67 Coincident flow data of Middle Creek and the six streams above were programmed and correlation equations were developed for each stream. Flow data of Little River near Princeton, North Carolina, has the best correlation with that of Middle Creek; the correlation coefficient is 0.93. Flow data from 1930 through 1939 was therefore synthesized for Middle Creek by utilizing its correlation with Little River.The correlation equation is:Amendment 65 Page 56 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 y = 1.92 + 0.415 x - 0.000118 x2 where y is the monthly flow in cfs of Middle Creek, and x is the corresponding monthly flow in cfs of Little River.The flow of the Deep River at Ramseur, North Carolina, and Randleman, North Carolina shows good correlation with the data of Middle Creek. Their correlation coefficients are 0.65 and 0.63, respectively. Therefore, flow data were synthesized for Middle Creek for the period from 1924 to January 1930 by utilizing the correlation equations developed for the Deep River at both stations:Deep River at Ramseur:y = 0.49 + 0.329 x - 0.000282 x2 and Deep River at Randleman:y = 4.36 + 0.744 x - 0.001383 x2 where y is the monthly flow in cfs of Middle Creek, and x is the corresponding monthly flow in cfs of Deep River. The average value of data from those two stations was used when coincident records were available.With the synthesized data (1924-1939) and the observed data (1939-1981) of Middle Creek, the monthly flow data of Buckhorn Creek for the period 1924-1981 were obtained by drainage area direct ratio relationship, and they are shown in Table 2.4.1-1.The average values of Middle Creek at the Clayton, North Carolina, gage station (D.A. = 80.7 sq. mi.) for the 42-year period (1939-1981) and the 58-year period (1924-1981) are 90.8 cfs and 88.5 cfs, respectively, while the corresponding synthesized values of Buckhorn Creek (D.A. =79.5 sq. mi.) are 89.4 cfs and 87.2 cfs, respectively.Subsequent to the above analysis, a review was made of the flows in the New Hope River near Pittsboro, North Carolina, which has flow records dating back to January 1949. This river has a basin area of 285 sq. mi. and is located about seven miles from Buckhorn Creek basin at the closest proximity. The overall average runoff of this river is 12.85 in. With an intention to synthesize Buckhorn Creek flow data from the data of the New Hope River, it was found that since 1955 there has been some flow diverted into the river basin above the station by the city of Durham; therefore, the flow data at the station has been distorted. If the flow data of the New Hope River were used to synthesize flows of Buckhorn Creek, lower values would be obtained during drought periods than those obtained by the correlation with Middle Creek. However, the effect on storage use would be minor.Since June 1972, a stream-gaging station has been installed on Buckhorn Creek in the vicinity of the Main Dam site in order to accumulate actual flow data prior to the impoundment of water in the Main Reservoir. The US Geological Survey installed and operates the Buckhorn Creek gaging station. The drainage area at this station is 74.2 sq. mi. Table 2.4.1-2 shows the actual observed monthly flow data and the calculated values derived from the drainage area ratio in Amendment 65 Page 57 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 relation to that of Middle Creek (D.A. equal to 80.7 sq. mi.). A correlation analysis has been made between the data from the observed and derived flow of Buckhorn Creek during the period of the stream-gaging station operation. It was found that the correlation coefficient is 0.95 during this period of time. The result of the analysis is shown on Figure 2.4.1-5.From the above correlation analysis, it can be concluded that the synthesized flow data of Buckhorn Creek from 1924 to 1981 can be utilized in hydrological studies.2.4.1.2.1.2 Cape Fear River The Cape Fear River basin is oblong in shape; its greatest width is about 60 miles and its length is about 200 miles. The Cape Fear River is formed by the confluence of the Deep and Haw Rivers. It flows generally southeast 198 miles and empties into the Atlantic Ocean at Cape Fear, 28 miles below Wilmington, North Carolina. The basin has a total area of 9136 sq. mi. of which 3127 sq. mi. are located above the confluence of the Deep and Haw Rivers.The Cape Fear River is an estuary with the tidal reach extending to Lock and Dam No. 1, about 39 miles above Wilmington. The river is navigable to Fayetteville, with a channel width of generally 400 ft. and a depth ranging from 30 to 35 ft. from the Atlantic Ocean to Wilmington; thence a 200-foot width and 25 ft. depth from Wilmington to Navassa; and a depth of eight feet with varying widths for the remaining distance to Fayetteville.The bankfull flood flow at Fayetteville is about 35,000 cfs, and at Lock No. 2 (River mile point 99, see Figure 2.4.1-4), it is about 20,000 cfs. The average width of the flood plain is approximately 2.2 miles. The difference between high and low stages is 69 ft. at Fayetteville and 44 ft. at Lock No. 2. The maximum flood flow of 150,000 cfs occurred on September 19, 1945, at Lillington and the minimum flow of 11 cfs occurred on October 14 and 15, 1954, at the same location. The monthly average flow data at Buckhorn Dam, shown in Table 2.4.1-3, were obtained from the records at Lillington by a drainage area ratio of 3196 sq. mi. at Buckhorn Dam and 3440 sq. mi. at Lillington.2.4.1.2.1.3 Tributaries The Cape Fear River has two major tributaries above the Buckhorn Dam (which is located nearby the confluence of the Buckhorn Creek and the Cape Fear River), the Haw and Deep Rivers, both of which originate in Forsyth County, North Carolina. The Deep River has a total length of 116 miles and a drainage area of 1422 sq. mi. The Haw River is about 90 miles in length and drains approximately 1705 sq. mi. Both rivers originate at elevations of about 1000 ft. MSL and have numerous falls and rapids, with the Haw River having the steepest gradient.The water surface elevation of the junction of the two rivers is about 158 ft.Other major tributaries downstream of the withdrawal point include the Black River and the Northeast Cape Fear River. The former has a drainage area of 1563 sq. mi. and joins the Cape Fear River at river mile point 44. The latter drains a basin of 1738 sq. mi. and enters the Cape Fear River at Wilmington.There are numerous minor tributaries including Upper Little River, Little River, Rockfish Creek, and Buckhorn Creek.Amendment 65 Page 58 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.1.2.1.4 Dams, Reservoirs, and Locks There are a number of regulating structures and reservoirs on the Cape Fear River. The locations of these structures and reservoirs are shown on Figure 2.4.1-4. Lock and Dam Numbers 1, 2 and 3 are located at river mile points 67, 99, and 123 respectively. Buckhorn Dam is at river mile point 192, and its spillway crest is at Elevation 158.18 ft.In addition to the existing Lockville Dam and Carbonton Dam on the lower reach of the Deep River, the Corps of Engineers has studied additional development of water resources for the Cape Fear River Basin. A summary of this additional development is shown in Table 2.4.1-4.The completion of the proposed plan will furnish a minimum continuous flow of 600 cfs at Lillington.2.4.1.2.1.5 On-site Waterbodies The Buckhorn Creek drainage system involves five tributaries (Tom Jack Creek, Thomas Creek, Little Whiteoak Creek, Whiteoak Creek, and Cary Creek). These five creeks together with Buckhorn Creek are affected by the impoundment of the Main Dam. Furthermore, the Auxiliary Dam creates the Auxiliary Reservoir on Tom Jack Creek near the plant site. The plant site is bounded by Thomas Creek on the east side and extends westward to Tom Jack Creek. A nameless tributary of Little Whiteoak Creek had been located on the western half of the plant island where grading has now raised the elevation to 260 ft. MSL.A drawing of the topography near the plant site is presented on Figure 2.4.1-6. This figure shows the small drainage areas and their divides before construction of the project. The drainage pattern can easily be visualized from this figure, which in comparison with Figure 2.4.1-2 of the plant site and drainage shows that construction of the project did not materially change the drainage pattern.No existing ponds or impoundments are located within the boundary of the plant island area; however, there are numerous small farm ponds located in the reservoir area, some of which have been inundated by filling the reservoirs. They are shown on Figure 2.4.1-7.2.4.1.2.1.6 Other Waterbodies There are other reservoirs, lakes, and ponds in addition to those described in Sections 2.4.1.2.1.4 and 2.4.1.2.1.5 in the vicinity of the plant site.The rivers, creeks, lakes, reservoirs, and ponds existing, prior to filling of the reservoirs, within 5 miles and 25 miles radius of the plant site are shown on Figures 2.4.1-7 and 2.4.1-8, respectively.2.4.1.2.1.7 Groundwater The sources of groundwater in the vicinity of the site are the bedrock units of the Sanford Formation of the Newark Group (Triassic). They consist of claystone, shale, siltstone, sandstone, conglomerate, and fanglomerate. Exceptions to this lithology are the thin diabase dikes in the bedrock. The Triassic rocks are overlain by a thin layer of dense clayey soils and saprolite.Amendment 65 Page 59 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The primary permeability of the Triassic aquifer is very low. However, the aquifer has secondary permeability due to the fractures which are filled with water below the water table.The fractures are common to depths of 100 ft, but become less prevalent and tight below that depth. Below about 400 ft, the fractures are closed and sealed to water flow. For a discussion of groundwater characteristics and usage in the plant vicinity, see Section 2.4.13.2.4.1.2.2 Effects of Normal or Accidental Release of Contaminants on Water Users 2.4.1.2.2.1 Contaminants Normal releases of contaminants into the hydrosphere will have negligible effects on surface and groundwater users (see Section 2.4.12.1). Should an accidental release of contaminants occur, adverse effects, if any, will be restricted to the area within the plant island. The only water user within the plant island is the plant itself. The use of water includes makeup to the cooling tower basin, the intake for the Emergency Service Water System, intermittent HVAC cooling water makeup, and fire protection.Dilution of contaminants (refer to Section 2.4.12), should they enter the reservoirs, will be great enough to reduce concentrations below the limits of 10 CFR 20. The effect of accidental spills into the groundwater is discussed in Section 2.4.13.3.2.4.1.2.2.2 Water Users There are no known domestic surface water users of Buckhorn Creek within the proposed reservoir area. There are no surface water users of Buckhorn Creek downstream of the SHNPP project. There are no known domestic potable water supply intakes on the Cape Fear River between Buckhorn Creek and Lillington, North Carolina. The nearest source of potable water supply downstream of the site is at Lillington, North Carolina, approximately 12 miles downstream on the Cape Fear River.Industrial and municipal surface water uses of the Cape Fear River downstream of Buckhorn Dam are shown in the Tables 2.4.1-5 and 2.4.1-6, respectively. River drainage basin areas at the points of withdrawal are included to indicate the additional flow available, as compared with the drainage area of 3196 sq. mi. at Buckhorn Dam. Most of the water withdrawn is returned to the Cape Fear River.Carolina Power & Light Company's Brunswick Plant, located 19 miles south of Wilmington at Southport, N.C., nominally withdraws cooling water from the Cape Fear River. However, this user is not included on Table 2.4.1-5 since the withdrawal is within the tidal reaches of the river and does not constitute a consumptive use of river flow. The outfall of the Brunswick Plant is located on the Atlantic Ocean. The drainage area at the plant is 9090 square miles, and the withdrawal and discharge are both 1900 mgd.Discussions with North Carolina State University's agricultural staff, U.S. Department of Agriculture, and county extension chairmen indicate that there are no known withdrawals for irrigation from the Cape Fear River. The principal economic crop in the Cape Fear basin is tobacco; however, the land along the Cape Fear River is not generally suited to production of tobacco. Tobacco is grown in the uplands and irrigation water, if used, is taken from farm ponds or wells. The lands along the Cape Fear River are either wooded, pasture or used for crops that are not generally irrigated in North Carolina.Amendment 65 Page 60 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.2 FLOOD 2.4.2.1 Flood History Before 1972, there were no flood records available for Buckhorn Creek. Records of flood flows since November of 1939 are available for the Middle Creek basin. Based on the ratio of drainage areas of Middle Creek (D.A. = 80.7 sq. mi.) and Buckhorn Creek (D.A. = 79.5 sq. mi.),the corresponding maximum historical peak flows for Buckhorn Creek near its confluence with the Cape Fear River are listed in Table 2.4.2-1 and the frequency analysis (Reference 2.4.2-1) of the data is shown in Figure 2.4.2-1.Table 2.4.2-2 shows the recorded data after 1972 at the USGS gage station on Buckhorn Creek near Corinth (D.A. = 74.2 sq. mi.) together with the corresponding estimated values based on the Middle Creek data from the drainage area relationship. The maximum measured flood flow of Buckhorn Creek of 6920 cfs correlates very well with the estimated value of 7820 cfs, as the recorded data of Middle Creek indicates that a dam failure occurred in this flood.The Cape Fear River has flow records at Lillington dating back to December 1923. The maximum flood flows at Buckhorn Dam are derived from the data at the Lillington gage by an adjustment by drainage area ratio (D.A. = 3440 sq. mi. at Lillington, D.A. = 3196 sq. mi. at Buckhorn Dam). These flows are shown in Table 2.4.2-3. The frequency analysis (Reference 2.4.2-1) of these data is presented on Figure 2.4.2-2. The maximum flood flow of 139,370 cfs occurred on September 19, 1945.2.4.2.2 Flood Design Considerations The SHNPP safety related structures and facilities are protected against all floods and flood waves caused by probable maximum events, such as the probable maximum flood (PMF), and the probable maximum hurricane (PMH). The tops of the Main Dam and the Auxiliary Dam are above the probable maximum flood water level in the reservoirs. The top of the Auxiliary Reservoir Separating Dike is allowed to be below the PMF stillwater level in the Auxiliary Reservoir, and the dike is designed to withstand the wave action under the PMF condition. The Emergency Service Water Intake channel is protected by sacrificial spoil fill along the shore bank of the Main Reservoir. The sacrificial spoil fill is allowed to be eroded during probable maximum events. After the event, the eroded portion will be inspected, restored, and stabilized where required. The design of this fill is discussed in Section 2.4-10. The plant island, where many safety related facilities are located, has a grade elevation above the PMF water level in both reservoirs.The PMF in the Cape Fear River is not considered because of the large difference in elevation between the river bank (approximate elevation 160 ft. MSL) and the top of the Main Dam (elevation 260 ft. MSL) the nearest Seismic Category I structure to the river. The backwater effect of the PMF in the river through Buckhorn Creek is comparatively small.The downstream face of the Main Dam is protected by a layer of oversized rock as indicated on Figure 2.5.6-2 for possible wind wave action whenever the backwater reaches the Main Dam.No specific design basis exists for downstream slope protection of the Main Dam. The rockfill shell does not require special slope protection because the Cape Fear River 500-year-flood backwater effect on Buckhorn Creek near the downstream face of the Main Dam is not expected to result in wave action on the dam. This is due to protection afforded by a small Amendment 65 Page 61 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 downstream fetch which severely limits the size of wind-generated waves. The oversize rock zone on the downstream face is primarily a construction-related feature. During construction of the Main Dam, oversize rocks were plucked from each of the rockfill lifts in order to meet specifications. Where the oversize rocks were within practical limits (20 to 30 inches) they were placed near the downstream face in order to reduce handling of oversize material and provide additional protection to the downstream face.Since the drainage area of Buckhorn Creek is small in comparison with that of the Cape Fear River at Buckhorn Dam, the construction of the Main Dam and the Auxiliary Dam of the project will have no significant effect on the 100-yr. and 500-yr. flood levels in the Cape Fear River.Consequently, the floor level shown on Figure 2.4.2-3 represents both the pre-construction and post-construction conditions.The 100-yr. and 500-yr. floodplains adjoining the Cape Fear River in the vicinity of Buckhorn Creek are shown in Figure 2.4.2-3. The corresponding plains for Buckhorn Creek and the SHNPP reservoirs adjacent to the plant island are shown in Figure 2.4.2-4.The flood profiles in the Cape Fear River are based on the following data provided by the U.S.Army Corps of Engineers (Reference 2.4.2-2):Amendment 65 Page 62 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 100-yr. Flood Water Standard Project (approx. 500-yr.)Location Level (Ft. above MSL) Flood Water Level (Ft. above MSL) 10,000 ft. Upstream of 168.5 186.5 Buckhorn Dam Upstream Side of 165.5 182.0 Buckhorn Dam Downstream Side of 159.5 182.0 Buckhorn Dam 4 miles Downstream of 147.0 172.0 Buckhorn Dam The flood water level profile slopes uniformly between the two locations upstream of the Buckhorn Dam as well as between the two locations downstream of the Buckhorn Dam.The pre-construction flood profiles of Buckhorn Creek for the 100-yr. and 500 yr. floods were calculated using the HEC-2 computer program (Ref. 2.4.2-3). The 100-yr. and 500-yr. flood flows in Buckhorn Creek before plant construction were obtained from Figure 2.4.2-1 as 9,900 cfs and 16,000 cfs, respectively, at its confluence with the Cape Fear River. Based on these flows, the corresponding flows in the tributaries of Buckhorn Creek were estimated according to their drainage area ratios. Since the normal creek channel is rather shallow, the creek cross-sections for the flood flows were principally scaled from a 1/12000 scale map at 1000 to 2000 feet intervals. In addition, available project construction maps for the area below the Main Dam and the USGS 1/24000 map of the area adjacent to the Cape Fear River were also used.Manning's n-values of 0.04 and 0.045 were selected for the main and flood channels, respectively, in the flood profiles computation.The floodplains adjoining Buckhorn Creek and its tributaries were delineated from the 1/12000 contour map as shown in Figure 2.4.2-4.The construction of the Main Dam and Auxiliary Dam of the plant will reduce the magnitude of the flood flows downstream of the plant because of the storage capacity of the two reservoirs created by the dams. Again, based on the drainage area ratio between that at each dam location and that of the entire Buckhorn Creek, the 100-yr. and 500-yr. floods adopted for the floodplain delineation are:Flood At Main Dam At Auxiliary Dam 100-yr 8850 cfs 215 cfs 500-yr 14300 cfs 350 cfs Amendment 65 Page 63 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Both the Main Dam and Auxiliary Dam have uncontrolled spillways to release floods. The spillway rating curves for these dams are shown in Figures 2.4.3-3 and 2.4.3-4. The corresponding flood level in each reservoir was determined by applying the flood flows to the appropriate rating curve. Since the reservoirs are rather small, no backwater effect in the reservoirs was taken into consideration when the floodplains adjoining the reservoirs were delineated.The floodplains adjoining the reach of Buckhorn Creek between the Main Dam and Cape Fear River after the construction of the Main Dam were not studied since the flood levels will be less than before construction.The construction of the plant will increase the extent of the floodplains above the Main and Auxiliary Dams in Buckhorn Creek and reduce the flood magnitude below the Main Dam. The water level (WL) and storage capacity (SC) of both reservoirs at 100-yr. and 500-yr. flood are:Main Reservoir Auxiliary Reservoir Flood WL (ft) MSL SC (Ac ft) WL (ft) MSL SC (Ac ft) 100-yr. 234.0 142 x 103 252.5 5.25 x 103 500-yr. 239.0 174 x 103 252.8 5.35 x 103 The storage capacities are obtained from Figure 2.4.3-5 and 2.4.3-6, the reservoir area and capacity curves, using the calculated water levels.The pre-construction and post-construction floodplains for the portion of Buckhorn Creek that is influenced by the plant construction are entirely within the site boundary. There are no existing structures within these floodplains other than those constructed for plant use. These structures were designed to preclude adverse effects due to the probable maximum flood. Additional structures may be constructed to support the recreational use of the Main Reservoir. It is expected that the effect of floods will be considered in the design of these structures based on a cost/risk assessment.Since the Cape Fear River floodplains are not increased due to plant construction, any pre-existing structures in these areas are not subject to increased risk of flood damage due to plant construction.In determining the probable maximum water level in the reservoirs, the precipitation produced floods in Buckhorn Creek were determined and the water levels in the reservoirs at the beginning of the PMF or during the PMH were considered under the normal operation condition.For compliance with NRC Regulatory Guide 1.59 (Design Basis Floods for Nuclear Power Plants), see Section 1.8. The events considered are summarized below:a) PMF in Buckhorn Creek and a designed wind wave activity in the reservoirs when the resulting PMF water levels in the reservoir are taken into consideration.b) PMH wind wave activity when the water levels in the reservoirs are at the normal operation level.Amendment 65 Page 64 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The probable maximum water levels are 243.1 ft., 258.0 ft., and 257.7 ft. at the Main Dam, Auxiliary Dam, and around the plant island, respectively (see Section 2.4.3.6.2). Flood protection design of safety related structures and facilities for flood waters for those heights include water-proofing, riprap, and sacrificial fill. The detailed flood design basis is addressed in Section 2.4.10. Tsunamis and seiches are events not applicable to the SHNPP.See Section 2.4.4 for additional discussion of dam failures.2.4.2.3 Effects of Local Intense Precipitation The general area of the site is subject to local intense precipitation. The plant island faces the Main Reservoir, and its grade is 40 ft. above the normal water level of the Main Reservoir and 8 ft. above the normal water level of the Auxiliary Reservoir. The site is not expected to experience any long term accumulation of ice and snow; thus ice or snow melt is not considered for flooding effects. As described in Section 2.4.1.1, most of the site is drained to the Main Reservoir, while a small portion of the run-off is drained to the Auxiliary Reservoir via the Emergency Service Water Intake and Discharge Channels. Therefore, the site drainage does not pose any potential problems. The local probable maximum precipitation (PMP) is determined by the method described in Section 2.4.3.1 and is assumed to be the same as the PMP for one sq. mi. which is the smallest area considered in the method of determining the PMP. Table 2.4.2-4 presents the time distribution of this PMP. Since losses in unpaved areas are not considered during the PMP, the capacity for the plant site drainage for run-off is four in.per hour. As a result, the accumulated water depth during the PMP considered is approximately 14.8 in. as shown in Table 2.4.2-4. The protections of Seismic Category I structures and safety related systems on the plant island against local flood are discussed in Section 3.4.1.1.All safety related buildings other than the Emergency Service Water Intake Structure, and Discharge Structure have structural features surrounding their roofs that would impound rainwater on the roofs assuming that the roof drains are plugged. In general, the ponding is caused by curbing whose height varies depending on the roof but is a maximum of one (1) foot above the high point of the surrounded roof. In addition to curbing around roof edges, the portions of the Reactor Auxiliary Building roofs which wrap around the west side of the containment building is partially surrounded by taller structures. Also, the tank building has two areas without roofs where walls enclose the tanks. The roof plans of all safety related buildings where ponding can occur are shown in Figure 2.4.2-5. Top elevation of the curbs and high points of each roof are also indicated in the figure. The figure does not include the ESW Screening Structure which, although it has a curbed area with drains on the roof, would not impound any significant volume of water with its drain plugged. Water would instead spill into the structure through louvered ventilation openings before draining into the Auxiliary Reservoir.The effects of spillage into the structure have been evaluated to have no impact on the ESW safety function.No scuppers or openings have been provided in the curbs. If the regular roof drains are assumed to be plugged during a local intense PMP event, the storm water will pond on the roof and overflow the curbs. For the local intense PMP event as given in Table 2.4.2-4, the water level on all roofs will exceed the top of the surrounding curb by less than three (3) inches except for some areas of the Reactor Auxiliary Building roof which are surrounded by higher walls. In these areas the accumulated water depth will exceed the top elevation of the curb by a maximum of 1 1/2 feet. The maximum water levels, including the cascade flow from higher roof levels, are indicated on Figure 2.4.2-5.Amendment 65 Page 65 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The open areas of the Tank Building, which are surrounded by 25 foot high walls (see Figure 1.2.2-84), do not overflow, however, rainwater will accumulate to a depth of 23.36 feet.The floor of the unroofed areas of the Tank Building and the roofs of all safety related buildings where water accumulates are strong enough to withstand the ponding loads in addition to other dead and live loads that can reasonably be expected to occur coincident with the PMP. The varying depths of water on a given roof due to the slope of the roof were accounted for in determining the structural adequacy.2.4.3 PROBABLE MAXIMUM FLOOD ON STREAMS AND RIVERS (PMF)The probable maximum flood (PMF) has been defined as an estimate of the hypothetical flood characteristics that are considered to be the most severe "reasonably possible" at a particular location based on comprehensive hydrometeorological analysis of critical runoff-producing precipitation and hydrologic factors favorable for maximum flood runoff (Reference 2.4.3-1). The objective of this study is to obtain a flood estimate that has a probability of occurrence near zero or a return period of near infinity.Using the above definition as a guide, the PMF's for the SHNPP were developed as follows:a) The Buckhorn Creek drainage basin was first analyzed under its natural, pre-construction condition. A unit hydrograph was developed for the entire drainage basin.b) After construction of the Main Dam, the drainage basin above the dam is 71.0 sq. mi.,wherein the area inundated is about 8.7 sq. mi., or about 12 percent of the entire basin.As a result, the inflow hydrograph to the Main Reservoir above the Main Dam is considerably altered from its natural condition; first, because of the reduced distance from the ridge line to the Main Reservoir surface; and secondly, because of the effect of direct rainfall on the reservoir area caused by the impoundment. In order to have a detailed estimate of the PMF, the drainage basin was divided into ten sub-basins, two of which (Sub-basins I and II) are located below the Main Dam site. Unit hydrographs were then developed for each sub-basin.c) The probable maximum precipitation (PMP) was applied to the unit hydrograph with the appropriate infiltration losses to develop the estimated flood hydrograph for each sub-basin, as well as for the entire drainage basin.d) An antecedent precipitation, which has an intensity of 1/2 PMP, and the PMP were also applied to the unit hydrograph with appropriate infiltration losses to develop the estimated flood hydrograph for each sub-basin in order to have a more conservative estimate of the PMF still water level in the Main and Auxiliary Reservoirs.e) The total inflow into the Main Reservoir behind the Main Dam is the summation of the outflow from all the sub-basins located above the Main Dam.f) After obtaining the inflow hydrograph, the PMF was then routed through the reservoirs to estimate the PMF still water level in the reservoirs.The following discussions are based on the guidance presented in Regulatory Guide 1.59, Revision 2, Appendix A.Amendment 65 Page 66 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.3.1 Probable Maximum Precipitation (PMP)The PMP is the theoretical greatest precipitation over the applicable drainage area that would produce flood flows that have virtually no risk of being exceeded. The PMP depths used in this study were developed from the U.S. Weather Bureau's "Hydrometeorological Report No. 33" (Reference 2.4.3-2).Their values for durations of 6, 12, 24, and 48 hours for the drainage basin of Buckhorn Creek (D.A. = 79.5 sq. mi.) at its confluence with the Cape Fear River at the Main Dam (D.A. = 71.0 sq. mi.), and at the Auxiliary Dam (D.A. = 2.43 sq. mi.) are shown in Table 2.4.3-1. Since the smallest drainage area considered in Reference 2.4.3-2 is 10 sq. mi., the intensity of the PMP of 10 sq. mi. was adopted for the PMP of the area of 2.43 sq. mi. For the PMP of the local drainage study, Hydrometeorological Report Nos. 51 and 52 of NOAA and Corps of Engineers were used (see Table 2.4.3-4). To allow for basin shape and the improbability of exact center of storms occurring over a particular drainage area, the PMP depths were reduced by a factor of 10 percent for the drainage areas of 79.5 sq. mi. and 71.0 sq. mi. as recommended by the Army Corps of Engineers (Reference 2.4.3-3), while those values for the drainage area of 2.43 sq. mi.and for the local drainage study were not reduced.To facilitate the application of the unit hydrograph to compute the probable maximum flood (PMF), unit rainfall periods of l-hour increments were required. These increments were derived from a 6-hour rainfall distribution curve applicable to the North Carolina region, as shown on Figure 18 of "Design of Small Dams" (Reference 2.4.3-4). Furthermore, these increments were rearranged in accordance with the criteria recommended in Hydrometeorological Reports Nos.33 (Reference 2.4.3-2) and 40 (Reference 2.4.3-5) for generating the PMF. The time distributions of PMP used are shown in Table 2.4.3-2. In view of the small areas of the drainage basins, the spatial distribution of the PMP was not considered.2.4.3.2 Precipitation Losses The HEC-1 computer program (Reference 2.4.3-6) was used to determine the precipitation losses. This program yields the optimized parameters of the unit hydrograph and loss rates for a stream basin by best reconstitution of an observed hydrograph. The data used in this optimization procedure are rainfall, drainage area, antecedent flow, and recession flow characteristics. The best reconstitution is considered to be that for which the weighted squared deviations between the observed and reconstituted hydrographs are a minimum. The univariate gradient search method was used in the optimization procedure.Since the establishment of a gaging station on Buckhorn Creek near Corinth in 1972, several floods have occurred in Buckhorn Creek (Table 2.4.2-2). Among these floods, that which occurred on February 2, 1973 is the most severe, even when using the flood records of the adjacent Middle Creek basin (which date back to 1940), to estimate Buckhorn Creek flows.Therefore, the reconstitution of the flood hydrograph for the storm of February 1973 was performed using recording (showing hourly values during the storm) and non-recording (showing the total value of the entire storm) precipitation data and recorded flow rates.Parameters such as precipitation loss and initial loss rate were determined as part of the optimization processes using the HEC-1 computer program. The output from the optimization, showing the observed and reconstituted hydrographs of Buckhorn Creek near Corinth, N.C.,Amendment 65 Page 67 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 together with the recording and non-recording precipitation data at various stations, is shown in Figure 2.4.3-1.Other output from the optimization are the HEC-1 loss parameters (STRKR, DLTKR, RTIOL, and ERAIN, see Reference 2.4.3-6 for their definitions) which are presented in Table 2.4.3-3.They were used in computing the PMF's for the entire Buckhorn Creek basin and its sub-basins.2.4.3.3 Runoff Model The unit hydrograph parameters for Buckhorn Creek near Corinth, along with the loss rate parameters, were obtained by using the HEC-1 computer program as mentioned in Section 2.4.3.2. The lag time (tp) and the basin constant (CT) thus obtained were utilized to develop unit hydrographs for the entire Buckhorn Creek basin and its sub-basins.By application of Snyder's synthetic unit hydrograph relationships (References 2.4.3-4 and 2.4.3-7) and the basin geometric parameters, L (the length of the longest water-course from the point of interest) and Lca (the length of water course from the point of interest to the intersection of the perpendicular from the center of gravity of the basin to the stream alignment) of the Buckhorn Creek drainage basin above Corinth, the basin constant CT can be derived from the known lag time, tp. Consequently, lag times for the entire Buckhorn Creek basin and its sub-basins can be obtained from known basin geometric parameters. The basin constant (Cp) and lag time (tp) for each basin are the input required for the HEC-1 computer program to generate each sub-basin's unit hydrograph.The unit hydrograph for the whole basin was used for predicting the PMF for the natural preconstruction condition, while those for the sub-basins were used for studying the PMF after completion of the project. Figure 2.4.3-2 shows a map of the entire Buckhorn Creek drainage basin area and the sub-basin areas for the Auxiliary Reservoir and Main Reservoir, and the area between the Main Reservoir and just above its confluence with the Cape Fear River.Table 2.4.3-3 shows the Snyder and loss parameters of the whole Buckhorn Creek basin and each sub-basin area.The sub-basin area identified as IX in Table 2.4.3-3 is that area north of the dashed line shown above the Auxiliary Dam on Figure 2.4.3-2. All parameters of the unit hydrograph were determined based only on this area. The drainage area between the dam and the dashed line was divided into lake area (0.63 sq. mi.) and land area (0.07 sq. mi.) and treated separately for inflow into the Auxiliary Reservoir. The total inflow into the Auxiliary Reservoir is comprised of three separate inflows: overland inflow from the 1.73 sq. mi. drainage area, direct rainfall on the 0.63 sq. mi. lake area, and direct runoff from the 0.07 sq. mi. residual land area. This accounts for the entire 2.43 sq. mi. drainage area behind the Auxiliary Dam.As a result of the construction of the Main Dam and the Auxiliary Dam, two reservoirs were formed. There is a spillway associated with each dam. The spillway crest at the Main Dam has a net length of 50 ft. with a pier at its mid-length, while the spillway crest at the Auxiliary Dam has a length of 170 ft. Both spillways are hydraulically designed for an ogee shape with a design head (HO) and an upstream dam height (P) of 30 ft. and 10 ft., respectively, for the Main Dam Spillway, while the corresponding values for the Auxiliary Dam Spillway are 5 ft. and 7 ft.,respectively. The equation of discharge is:Amendment 65 Page 68 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Q = CsL He3/2 where Q is the discharge in cfs, L is the effective length of the spillway crest in feet and He is the total head on the crest of the spillway in feet including the velocity of approach head. The coefficient of discharge (Cs) is obtained from the curves shown on Figures 249 and 250 of "Design of Small Dams" (Reference 2.4.3-4) with known values of HE/Ho and P/Ho. When the values of He/Ho reduce from 1 to 0.5, Cs varies from 3.95 to 3.64 for the Main Dam Spillway, and from 3.85 to 3.54 for the Auxiliary Dam Spillway.The effective length of the Main Dam Spillway crest is determined from the formula (Reference 2.4.3-4).L = L' -2 (NKp + Ka) He where L' is the net length of the spillway crest in feet, N is the number of piers, Kp is the pier contraction coefficient, and Ka is the abutment contraction coefficient. For the present case, Kp and Ka are equal to 0.01, N is equal to 1, and L' is 50 ft.; consequently L is 49.4 ft.Since the head on the crest of the Auxiliary Dam Spillway ranges from 1 to 6 ft., the effect of the end contractions is insignificant; therefore the effective length of the Auxiliary Dam Spillway crest is taken as equal to the actual length, 170 ft. The rating curves for both spillways are shown on Figures 2.4.3-3 and 2.4.3-4.For flood routing through the reservoirs, the Modified Pulse Method was used. This method is contained in the HEC-1 computer program. The capacity curves for both reservoirs are shown on Figures 2.4.3-5 and 2.4.3-6 and the corresponding rating curves for the spillways provide data for input to the computer program for the routings.The crest of the Main Dam Spillway is at El. 220 ft. MSL and that of the Auxiliary Dam Spillway is at El. 252 ft. MSL.2.4.3.4 Probable Maximum Flood Flow (PMF)Application of the PMP shown in Table 2.4.3-2 to the unit hydrographs derived from the HEC-1 computer program resulted in the PMF for the entire Buckhorn Creek basin as well as for its sub-basins. With the input shown in Table 2.4.3-3, the resulting PMF includes the initial loss and infiltration loss during the PMP. No base flow was considered in the estimate of the PMF since the mean flow in Buckhorn Creek is 87.2 cfs as shown in Table 2.4.1-1. This amount is equivalent to 1.1 cfs per sq. mi. of drainage area, which is insignificant when compared to the PMF flow.The nearest Seismic Category I structure to the Cape Fear River is the Main Dam, which has its top at Elevation 260 ft. MSL, and the elevation of the Cape Fear River bank near Buckhorn Dam is in the vicinity of 160 ft. MSL. This large difference in elevation precludes any over-topping of the Main Dam due to backwater effects of the PMF on the Cape Fear River. Therefore, the PMF and the flood induced by the failure of dams upstream of the Buckhorn Dam in the Cape Fear River are not considered.Amendment 65 Page 69 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.3.4.1 Buckhorn Creek Drainage Basin Under Its Natural Condition The PMF hydrograph for Buckhorn Creek in its natural condition prior to the construction of the reservoirs is shown on Figure 2.4.3-7. The peak flow is 52,000 cfs, which would occur about 29 hours after the beginning of the PMP storm.2.4.3.4.2 Drainage Basin Above the Auxiliary Dam By considering the drainage area above the Auxiliary Dam as a sub-basin of the drainage area above the Main Dam, the intensity of the PMP of the drainage area of 71.0 sq. mi. is used to estimate the PMF. The PMF flow includes the overall runoff from the drainage area of 1.73 sq.mi., the direct rainfall on the reservoir surface area of 0.63 sq. mi., and the direct runoff from rainfall excess on the residual land area of 0.07 sq. mi. The flow was then routed through the Auxiliary Reservoir, and the peak flow is reduced from the combined overland runoff and direct rainfall of 5950 cfs to 3670 cfs, as shown on Figure 2.4.3-8.In addition, a study with a separate and more severe local PMF was also performed. The PMP with its intensity related to a drainage area of 2.43 sq. mi. was used, as shown in Table 2.4.3-2.Since the drainage area involved is rather small, the maximum possible antecedent moisture condition was assumed, and infiltrated and retention were neglected to derive the PMF.Consequently, the PMP excess equals the PMP values shown in Table 2.4.3-2. This is the most conservative assumption that can be made with respect to rainfall. The storm period considered for this analysis is 36 hours since time to peak inflow is well within this time period, and the incremental hourly rainfall beyond this period is negligible. A unit rainfall period of 1-hour increments was adopted, and the 1 hour unit hydrograph characteristics shown in Table 2.4.3-3 were used for determining the overland flow from the drainage area of 1.73 sq. mi. In addition to the overland flow, the direct rainfall on the sum of the lake area and residual land area totaling 0.70 sq. mi. was accounted for separately, with an assumed time of concentration of zero. After obtaining the PMF hydrograph with a peak flow of 8270 cfs, the PMF was routed through the Auxiliary Reservoir, reducing the peak flow to 5030 cfs.Figures 2.4.3-8 and 2.4.3-9 show the inflow flood hydrographs from overland flow and direct rainfall on the lake and residual land area, and the outflow hydrograph for both cases studied above.2.4.3.4.3 Drainage Basin Above the Main Dam The drainage area controlled by the Main Dam is 71.0 sq. mi. This area is composed of 2.43 sq. mi. above the Auxiliary Dam mentioned in Section 2.4.3.4.2, 46.37 sq. mi. of seven sub-basin areas shown on Table 2.4.3-3, and 8.68 sq. mi. and 13.52 sq. mi. of the reservoir water surface area and residual land surface area, respectively, for direct rainfall and runoff. The PMF hydrographs for each of the sub-basins were determined separately by the HEC-1 computer program, then they were summed to determine the overland PMF inflow to the Main Reservoir.Due to the proximity of all sub-basins, no adjustment was made to reflect lag time due to their relative positions in producing the peak instantaneous overland inflow.The average hourly inflow from direct rainfall on the lake surface and the rainfall excess on the residual land areas was also determined separately for the drainage area controlled by the Main Reservoir. In addition to this inflow, the discharge from the Auxiliary Reservoir was added to obtain the total inflow to the Main Reservoir during the flood.Amendment 65 Page 70 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 As shown on Figure 2.4.3-10, the overland inflow hydrograph, including the release from the Auxiliary Reservoir, has a peak of 58,790 cfs 19 hours after the start of the storm. Also shown on Figure 2.4.3-10 is the runoff from direct rainfall on the reservoir surface and residual land areas. The instantaneous combined peak inflow at the Main Dam site from all sources indicated by Figure 2.4.3-10 is about 161,710 cfs, which includes release from the Auxiliary Reservoir.This combined peak occurs about 11 hours after the start of the storm. The combined inflow hydrograph was routed through the reservoir, and an outflow hydrograph was developed with a peak flow of 11,030 cfs.As indicated by a comparison of Figures 2.4.3-7 and 2.4.3-10, the project will afford some flood protection to the area downstream of the Main Dam during a major storm. For the probable maximum flood, the peak outflow is reduced from 52,000 cfs for the natural condition to 11,030 cfs after construction of the project.An analysis was also made of the PMF approaching the Main Reservoir by assuming conservatively that the PMF begins five days after the start of a less severe storm such as the standard project flood resulting from 1/2 PMP. For this assumed antecedent condition, the reservoir level at the end of the fifth day resulting from the standard project flood would be about Elevation 225.2 ft. MSL, and the corresponding discharge would be about 2000 cfs. Starting with this flood surcharge at the beginning of the probable maximum flood, the peak outflow is 14,190 cfs, peaking about 33 hours after the start of the PMP storm. Figure 2.4.3-11 shows the inflow and outflow hydrographs for the probable maximum flood following the standard project flood.2.4.3.5 Water Level Determination The water levels in the Main Reservoir and the Auxiliary Reservoir are related to the safety of the safety related structures such as the Main Dam, the Auxiliary Dam, and those structures situated at the plant site. With the known PMF flow over the spillways of both reservoirs, the PMF stillwater levels in the reservoirs were determined by the corresponding spillway rating curves shown on Figures 2.4.3-3 and 2.4.3-4.The PMF flow over the Auxiliary Reservoir Spillway is 3670 cfs, and the corresponding water level elevation in the reservoir is 255.2 ft. MSL. When the more severe local PMP is considered, the corresponding PMF flow is 5030 cfs, and the water level elevation is 256.0 ft.MSL. The reservoir water level elevation hydrographs for both cases are shown on Figures 2.4.3-8 and 2.4.3-9.The PMF flow over the Main Reservoir Spillway is 11,030 cfs without the antecedent standard project flood, and the resulting water level elevation is 236.2 ft. MSL. When the antecedent standard project flood is taken into consideration, the PMF flow becomes 14,190 cfs, and the water level elevation in the reservoir is 238.9 ft. MSL. Both reservoir water level elevation hydrographs are shown on Figures 2.4.3-10 and 2.4.3-11.2.4.3.6 Coincident Wind Wave Activity The coincident wind wave activities were determined in accordance with the procedures and methods presented in the U.S. Army Corps of Engineers' ETL 1110-2-221 (Reference 2.4.3-10) and in the Shore Protection Manual (Reference 2.4.3-8). For this study, the first reference was used to determine the wave characteristics, while the second reference was employed in Amendment 65 Page 71 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 computing the wave runup. Since no long term wind records are available for the plant site, the maximum wind velocity charts in Reference 2.4.3-10 were utilized to determine the design wind velocity. The PMH wind speed was taken from Section 2.4.5.3.1.2.4.3.6.1 Wind Setup, Wave Height, and Wave Period The wind setup, wave height, and wave period are a function of the effective fetch length, wind speed, wind duration, and water depth. These values are shown in Tables 2.4.5-1 and 2.4.5-2 for various critical locations where the maximum wind runup occurs. Figure 2.4.5-1 provides the locations of various fetchs used in computing the wind setup, wave height, and wave period.2.4.3.6.2 Wave Runup Wave runup is a function of wave height, wave period, slope of the approach, bottom structure slope, and the water depth at the toe of the structure. The upstream faces of the Main Dam and the Auxiliary Dam are protected by riprap; the former has a slope of 1 (vertical) to 2 (horizontal) and the latter has a slope of 1 (vertical) to 2 1/2 (horizontal). By considering the slope and characteristics of the protected surface and the wave steepness (ratio of wave height to wave length), the relative wave runup (ratio of wave runup to wave height) was obtained (Reference 2.4.3-8). The wave runup for wave steepness greater than 0.08 was estimated by extrapolating.The wave runup thus obtained was primarily based on the results of small scale hydraulic model tests. The results are presented in Tables 2.4.5-1 and 2.4.5-2.The maximum wave runup at the Main Dam is 4.1 ft. This value in combination with the wind setup, (0.1 ft.) and the PMF stillwater elevation, (238.9 ft.MSL as determined in Section 2.4.3.5) produces a probable maximum water level at the Main Dam of approximately 243.1 ft. MSL due to the PMF in the Main Reservoir coincident with wave activity. This maximum water level is 16.9 ft. below the top of the Main Dam, 260 ft. MSL.The maximum wave runup on the upstream face of the Auxiliary Dam is 1.9 ft. This value in conjunction of the wind setup, (0.1 ft.), and the PMF stillwater level (256.0 ft. MSL in the Auxiliary Reservoir determined in Section 2.4.3.5) results in a probable maximum water level at the Auxiliary Dam of approximately 258.0 ft. MSL. This maximum water level is 2.0 ft. below the top of the Auxiliary Dam.On the plant island, the southerly fill portion of the Emergency Service Water Intake Channel and the embankment faces of the plant island which face the Main Reservoir are protected by sacrificial spoil fill. The fill at the southeast corner has a slope of 1 (vertical) to 5 (horizontal) from plant grade 260 ft. MSL to Elevation 245 ft. MSL (a variable-width berm at that elevation),and a slope of 1 (V) to 10 (H) from the berm to the existing ground. The locations of this fill are shown on Figure 2.4.1-2. Wave runup is estimated for a 1 (V) to 10 (H) smooth slope for the sacrificial fill area, and a 1 (V) to 5 (H) slope for the natural ground surface adjoining the Auxiliary Reservoir. The results are shown in Tables 2.4.5-1 and 2.4.5-2.The plant is generally protected from wind-generated waves by high ground from all quadrants.The most critical wind fetch at the site is for a wind coming from the west over the Auxiliary Reservoir. The maximum wave runup and wind setup level on the side of the plant island is 1.7 ft. For a maximum PMF stillwater level of 256.0 ft MSL in the Auxiliary Reservoir, the maximum water level is estimated to be 257.7 ft. MSL, 2.3 ft. below the plant grade.Amendment 65 Page 72 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 For design of safety related structures against static and dynamic effects of water waves, see Sections 2.5.6.4.5 and 3.8.4.3.1.2.4.4 POTENTIAL DAM FAILURES, SEISMICALLY INDUCED The following discussions in this section are based on the guidance presented in Regulatory Guide 1.59, Revision 2, Appendix A.2.4.4.1 Dam Failure Permutations A map of the entire Buckhorn Creek drainage basin area at the plant site, and the area up to its confluence with the Cape Fear River, is shown on Figure 2.4.3-2. There are no existing water control structures in this drainage basin other than the Main Dam, the Auxiliary Separating Dike, and the Auxiliary Dam.Figure 2.4.1-1 shows the location of the dikes and the dams with respect to the plant site. The Auxiliary Separating Dike and Auxiliary Dam are located on Tom Jack Creek to the west of the plant site, whereas the Main Dam is located on Buckhorn Creek downstream and south of the plant site.The Auxiliary Reservoir was formed by the construction of the Seismic Category I Auxiliary Dam and Spillway. The Auxiliary Separating Dike and Auxiliary Reservoir Channel form a cooling loop in the reservoir. The maximum water level in the Auxiliary Reservoir, as indicated in Tables 2.4.5-1 and 2.4.5-2, is expected to be Elevation 258.0 ft. MSL for a probable maximum flood (PMF) with 52.9 mph wind. The top of the Auxiliary Dam and the plant island grade are at Elevation 260 ft. MSL. The PMF is assumed to occur with the Auxiliary Reservoir at a normal water level of 252.0 ft. MSL (Auxiliary Dam Spillway crest level).The Auxiliary Reservoir Channel, a Seismic Category I structure, is designed to remain stable when subjected to the Safe Shutdown Earthquake or the most severe cases of other postulated natural phenomena. The Auxiliary Reservoir Channel is sized such that, if any earth slippage did occur, the flow due to the PMF would still pass through the channel at a velocity not in excess of 2 ft/sec, and no substantial differential head (less than one foot) would be created across the Auxiliary Separating Dike. A typical cross section of the Auxiliary Reservoir Channel is shown on Figure 2.5.6-6.The top of the Auxiliary Separating Dike is at Elevation 255 ft. MSL. The Auxiliary Separating Dike is a Seismic Category I structure. Due to the low differential head across the Auxiliary Separating Dike, its failure would not cause loss of water or storage capacity or result in failure of any other structure. Due to short-circuiting of the cooling pathway, the ability of the Auxiliary Reservoir to perform its emergency cooling water function would be adversely affected by the loss of the Auxiliary Separating Dike. Under this postulated condition, the Main Reservoir would function as a backup source of cooling water with discharge of water into the Auxiliary Reservoir and then into the Main Reservoir over the Auxiliary Dam Spillway. The pathway thus established exceeds the heat dissipation requirements normally provided by the Auxiliary Reservoir alone, and it is more than adequate for plant requirements.The Auxiliary Dam is a Seismic Category I structure with the top of the dam at Elevation 260 ft.This will prevent overtopping of the dam due to maximum wave runup associated with a 52.9 mph wind coincident with the PMF (see Table 2.4.5-1). The preferred source of emergency Amendment 65 Page 73 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 service water for the plant is the Auxiliary Reservoir. In the unlikely event of a failure of the Auxiliary Dam, emergency service water will be supplied from the Main Reservoir.Failure of the Auxiliary Dam will not induce failure of the Main Dam since the two dams are widely separated (over four miles). As discussed in Section 2.4.4.2, the rise in water level in the Main Reservoir would be approximately 1.5 ft. to Elevation 221.5 ft. and there would be no wave action to impose unusual forces on the Main Dam.The Main Reservoir was formed by the construction of a Seismic Category I dam and spillway on Buckhorn Creek. The maximum water level, including wave runup and wind setup in the Main Reservoir as indicated in Table 2.4.5-1, is expected to be at Elevation 243.1 ft. MSL. The flood is assumed to begin five days after the start of a less severe storm such as the standard project flood. For this assumed antecedent condition, the Main Reservoir level at the end of the fifth day as a result of the standard project flood would be Elevation 225.2 ft. MSL. Starting with this flood surcharge at the beginning of the PMF, the maximum still water flood level would be Elevation 238.9 ft. MSL and wave runup for a 50.4 mph wind would be to Elevation 243.1 ft.MSL. The top of Main Dam is at Elevation 260 ft. MSL.Failure of the Main Dam would not induce failure of the Auxiliary Dam. As the lowest ground level at the Auxiliary Dam is only 10 to 15 ft. below the normal water level in the Main Reservoir, rapid drawdown of the Main Reservoir would have a negligible effect on the downstream face of the Auxiliary Dam. Furthermore, the dam is conservatively designed using a loading condition of no water downstream. If the Main Dam failed, emergency service water for the plant would be available from the Auxiliary Reservoir.U.S. Highway No. 1 and the relocated Norfolk Southern Railroad, Durham Line, cross portions of the Auxiliary Reservoir, as shown on Figure 2.1.1-1. The highway and the railroad embankment have reservoir water on both sides which are connected by culverts.Embankment failure would not affect the safety related function of the Auxiliary Reservoir.The Plant access road on the east side of the plant site crosses the Main Reservoir at Thomas Creek with the top of the road at approximate Elevation 243 ft. MSL. The embankment has Main Reservoir water on both sides and its failure would not affect the safety related function of the Main Reservoir, Main Dam, or Auxiliary Dam. An additional temporary construction road crosses Thomas Creek upstream of the plant access road. The failure of the construction road cannot adversely affect the safety related functions of the Main Reservoir, Main Dam, or Auxiliary Dam.Should any of these embankments fail in such a way that the culverts are blocked, the volume of water that would be impounded is insignificant to the cooling function of the reservoirs. There are additional roads which cross fingers of the Main Reservoir. The failure of these roads would not reduce the amount of water available to the plant, nor would they cause failure of any safety related structures.2.4.4.2 Unsteady Flow Analysis of Potential Dam Failure An analysis was made of the effect of failure of the Auxiliary Dam on the Main Reservoir and Main Dam. The following conservative assumptions were made:Amendment 65 Page 74 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 a) The Auxiliary Dam fails instantaneously (disappears) with the water level in the Auxiliary Reservoir conservatively chosen to be at Elevation 253 ft. MSL and at Elevation 220 ft.MSL in the Main Reservoir.b) The wave height and velocity developed at the Auxiliary Dam site travels downstream without any attenuation or dampening through Tom Jack Creek just downstream of the Auxiliary Reservoir.Based on the assumption of an instantaneous dam failure and the method of analysis given in "Water Waves" by J.J. Stoker (Reference 2.4.4-1), the height and velocity of the hydraulic bore were calculated to be 10 ft. and 44 ft. per second, respectively. The corresponding instantaneous discharge rate at the time of failure is 1,744,000 cfs. The velocity of the negative wave moving upstream was determined on the basis of the average depth (d) at each section using the formula v = , where g is 32.2 ft/sec2. The time required for this wave to move upstream and reflect back to the Auxiliary Dam site is the total time required to empty the Auxiliary Reservoir down to Elevation 220 ft. MSL.On the basis of the instantaneous velocities and cross sectional areas at the stations upstream of the reservoir, the time required to completely empty the Auxiliary Reservoir was determined to be approximately 400 seconds. However, the Auxiliary Reservoir would not drain completely, since the normal water level in the Main Reservoir (Elevation 220 ft. MSL) is higher than the bottom of the Auxiliary Reservoir.The time required for the positive wave to move through Tom Jack Creek into the Main Reservoir is approximately 200 seconds, during which time about 144 x 106 cu ft of water would be emptied from the Auxiliary Reservoir. Since the volume of storage between Elevation 220 ft.MSL and 230 ft. MSL (10 ft. higher) in Tom Jack Creek is 154 x 106 ft.3 which is greater than the volume emptied from the Auxiliary Reservoir, the bore cannot be maintained due to the lack of continuing supply of energy, and it becomes a non-linear wave with a highly dissipative impulse.For this condition, an exponential type of decay can be assumed; therefore the bore will quickly dissipate and enter the Main Reservoir with little disturbance. The average rise in the water level of the Main Reservoir would be approximately 1.5 ft., which results from a uniform spread of the total amount of water stored behind the Auxiliary Dam over the surface of both reservoirs to Elevation 221.5 ft. MSL.Since there is no wave action in the Main Reservoir, the failure of the Auxiliary Dam would impose no unusual forces on the Main Dam.2.4.4.3 Water Level at Plant Site The plant site, as indicated on Figure 2.4.1-1, is bounded by the Main Reservoir on the east, south, and south-west sides, and by the Auxiliary Reservoir on the west and north-west sides.The maximum water level in the Auxiliary Reservoir would be at Elevation 258.0 ft. MSL; and the maximum water level in the Main Reservoir would be at Elevation 243.1 ft. MSL (see Tables 2.4.5-1 and 2.4.5-2). The Auxiliary Reservoir maximum water level of Elevation 258.0 ft. MSL includes wave runup and wind setup superimposed on the PMF. Failure of the Auxiliary Dam, Auxiliary Reservoir Separating Dike, or Main Dam would not result in any rise of water level above Elevation 258.0 ft. MSL and therefore the plant site with its grade at Elevation 260 ft.MSL will not be flooded by dam failure.Amendment 65 Page 75 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.5 PROBABLE MAXIMUM SURGE FLOODING The probable maximum hurricane (PMH) could cause a water level change of the Main Reservoir and the Auxiliary Reservoir. The resulting high water levels, if not considered in the project design, could affect the safety of the Main Dam, Auxiliary Dam, or other safety related structures located at the plant island. The design considerations afforded these facilities due to flooding are discussed in the following sections. Preceding the PMH, the normal water level elevations in the Main Reservoir and in the Auxiliary Reservoir are 220 ft. MSL and 252 ft. MSL, respectively.2.4.5.1 Probable Maximum Wind and Associated Meteorological Parameters The meteorological characteristics used to calculate the probable maximum hurricane (PMH) are those reported by the U. S. National Oceanic and Atmospheric Administration (NOAA) in their unpublished report HUR 7-97 (Reference 2.4.5 1). HUR 7-97 describes the PMH as "a hypothetical hurricane having that combination of characteristics which will make it the most severe that can probably occur in the particular region involved". The hurricane should approach the point under study along the critical path at a critical speed. The hurricane characteristics used in establishing the PMH include:a) Central pressure index (CPI) - the minimum surface pressure in the eye of the hurricane.b) Radius of maximum wind (R) - the distance from the eye of the hurricane to the locus of maximum wind.c) Forward speed (T) - the rate of forward movement of the hurricane.d) Maximum gradient wind (Vgx) - the absolute highest wind speed in the belt of maximum wind.e) Peripheral pressure (Pn) - the surface pressure at the outer limits of the hurricane where hurricane circulation ends.HUR 7-97 presents values for each of those characteristics for each degree of north latitude along the east coast. Single values are presented for CPI and Pn, and three values are given for both R and T. Since Vgx is dependent upon Pn, CPI, and R, three values are also given for this parameter.At the latitude (35.6N) where the SHNPP is located, the following PMH characteristics are recommended in HUR 7-97:a) CPI: 26.91 in. Hg.b) R for: small radius storms (RS) - 7 nautical miles (NM); medium radius storms (RM) -17NM; large radius storms (RL) - 34 NM.c) T for: slow forward speed (ST) - 5 knots; medium forward speed (MT) - 18 knots; high forward speed (HT) - 40 knots.d) Vgx for RS: 148.9 mph Amendment 65 Page 76 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 RM: 147.2 mph RL: 144.1 mph e) Pn: 30.90 in. Hg.The maximum gradient wind (Vgx) for various radii (R) of maximum wind shown above is a 30-foot, 10 minute average over-water wind speed. When a hurricane moves inland, its intensity is reduced. The overwater wind speed is then reduced by an adjustment ratio which is a function of its overland travel time. SHNPP is located approximately 140 miles inland from the coast line. With a forward speed of 40 knots, the overland travel time is approximately 3-1/2 hours, and the adjustment ratio is 0.84, as recommended by HUR 7-97. If 147 mph is adopted as the maximum gradient over-water wind speed, the maximum gradient overland wind speed at the site would be 123 mph.2.4.5.2 Surge and Seiche Water Levels For the SHNPP site the only dynamic mechanism considered to be credible for the production of high water levels is the probable maximum wind discussed in Section 2.4.5.1. Section 2.4.5.3 discusses the effects of the probable maximum wind on the plant reservoirs and the resulting wave activity.2.4.5.3 Wave Action 2.4.5.3.1 Wave Generating Wind Activity Since the 123 mph wind speed at the site, as discussed in Section 2.4.5.1, is a 30-ft., 10 minute average over-land wind speed, and if it is assumed that a wind duration time of 30 minutes is required to generate the maximum wave, then a further correction to account for the variation of wind speed with duration should be made. The ratio of the 30-minute wind speed (V30) to the 10-minute wind speed (V10) is given by the equation (Reference 2.4.5-2):V30/V10= 1.45 - (0.07) (ln 1800) = 0.925, and V30 = 0.925 V10 = 0.925 x 123 = 114 mph.To account for the increase in wind speed at the site as the air trajectory moves over the smoother and more uniform water surface of the reservoir, the ratio of over-water to overland wind speed for a conservatively long fetch of 0.5 miles was applied. According to Reference 2.4.5-3, this ratio is 1.08 which would result in a 123-mph wind speed. This wind speed, in conjunction with the normal water levels in the Main Reservoir and the Auxiliary Reservoir, was used to estimate maximum wave runup. All wind wave activities were determined in accordance with the procedures and methods described in the Shore Protection Manual (Reference 2.4.5-4).Amendment 65 Page 77 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.5.3.2 Wind Setup, Wave Height, and Wave Period In view of the 40 ft. difference in elevation between the top of the Main Dam and the normal water level in the Main Reservoir, wave action on the Main Dam during the PMH was not studied. The wind setup, wave height, and wave period for the critical locations at the Auxiliary Dam and around the plant island were calculated based on known values of fetch, water depth, and wind speed. These values are listed in Tables 2.4.5-1 and 2.4.5-2. Each critical fetch for which wave runup has been calculated is shown on Figure 2.4.5-1.2.4.5.3.3 Wave Runup Following the same procedure as described in Section 2.4.3, wave runup at the Auxiliary Dam and plant island has been calculated.Assuming a PMH windspeed of 123 mph, the maximum runup is 3.8 ft. at the Auxiliary Dam.This value, combined with a wind setup of 0.4 ft. and the normal water level in the Auxiliary Reservoir, results in a maximum water level elevation of 256.2 ft. MSL, which is 3.8 ft. below the top of the Auxiliary Dam. Similarly, the maximum runup at the plant island is 2.7 ft. This value, combined with the wind setup of 0.2 ft. and the normal operation water level in the Auxiliary Reservoir, results in a maximum water elevation of 254.9 ft. MSL, which is 5.1 ft. below the grade elevation of the plant island.U. S. Highway 1, which crosses a finger of the Auxiliary Reservoir, is elevated above the maximum reservoir water levels; the pre-existing alignment of the road and the pre-existing drainage structures under the road have not been changed except for lengthening of the culverts to allow for possible future widening of the road.The plant access roadway to the site crosses the Thomas Creek area of the Main Reservoir and is protected from flooding since the lowest elevation of the roadway at the plant site is 243 ft.MSL. The maximum wave runup and wind setup level is 240.2 ft. MSL as shown in Table 2.4.5-2.The relocated Norfolk Southern Railroad, Durham Line, crosses portions of the Auxiliary Reservoir, as shown on Figure 2.1.1-1. The railroad embankment has reservoir water on both sides. The top of the rails is set at a minimum elevation of 262 ft. MSL, and the top of the embankment is at approximate Elevation 260 ft. MSL; these elevations are approximately eight feet above normal water level in the Auxiliary Reservoir and are above all wave runup heights.2.4.5.4 Resonance Wave amplification due to "harbor resonance" will not occur on either reservoir at the plant site because the wind fetch is approximately 100 times longer than the significant wave length. The resonance due to such a high mode, if it does occur, would not have an appreciable effect.Normally, only the first few modes of resonance are of concern, that is, the wave length would have to be at least 500 ft.2.4.5.5 Protection of Structures The only Seismic Category I safety-related facilities that require design consideration due to wave action are those associated with the reservoirs; i.e., the Main Dam, Auxiliary Dam, Amendment 65 Page 78 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Auxiliary Separating Dike, Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, the Emergency Service Water Screening Structure, Emergency Service Water Intake and Discharge Structures, the Cooling Tower Makeup Water Intake Channel, and the Emergency Service Water and Cooling Tower Makeup Water Intake Structure.The upstream face of the Main Dam, both upstream and downstream faces of the Auxiliary Dam, and both sides of the Auxiliary Separating Dike are protected by riprap designed for the worst postulated wave action. Additional description of the riprap design is in Section 2.5.6.The downstream face of the Main Dam is protected by a layer of oversized rock, as indicated on Figure 2.5.6-2. As discussed in Section 2.4.2.2, the backwater effects of the Cape Fear River on the downstream face of the Main Dam are not expected to be significant. However, protection of the downstream face, as described in Section 2.5.6.4, serves as an additional safety precaution.The Emergency Service Water and Cooling Tower Makeup Water Intake Structure, the Emergency Service Water Screening Structure, and the Emergency Service Water Discharge Structure are designed to withstand forces that could result from the worst postulated flood and wind conditions.The plant intake and discharge channels and the Auxiliary Reservoir Channel are designed to withstand the worst postulated natural phenomena, as described in Section 2.4.8.The embankment of the plant island along the Main Reservoir is protected by sacrificial spoil fill, as shown on Figure 2.4.1-2. The berm of the sacrificial spoil fill at Elevation 245 ft. MSL is above the maximum Main Reservoir water level of 240.2 ft. MSL (see Table 2.4.5-2) and it has a width of 300 feet on the south and southeast exposures. The nearest Seismic Category I structure (Diesel Generator Building) on the plant island is about 2000 ft. from the edge of the Main Reservoir at its normal water level of 220 ft. MSL and approximately 900 ft. from the edge of the plant island grade at Elevation 260 ft. MSL. The southerly edge of the Emergency Service Water Intake Channel is also 1000 ft. from the edge of the Main Reservoir at its normal water level of 220 ft. MSL. The extent of erosion due to the two worst fetches (S and SSE, see Table 2.4.5-2) is estimated to be 150 ft. resulting from a PMH having a duration of 48 hours.This estimation is based on methods described in References 2.4.5-5 and 2.4.5-6. The 300 foot wide sacrificial spoil fill, therefore, provides a very conservative design. Additional description of the sacrificial spoil fill is in Sections 2.4.3.6.2 and 2.5.6.All safety related structures on the plant island are protected from high water levels up to Elevation 261 ft. which is higher than any anticipated flood levels due to wave runup in the reservoirs or direct rainfall on the plant island. For further discussion see Sections 2.4.1.1 and 3.4.1.2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING The areas of the U. S. that are susceptible to tsunamis are those bordering on the Pacific Ocean or the Gulf of Mexico. As it is located approximately 140 miles inland on the Atlantic coast, the Shearon Harris Nuclear Power Plant is not subjected to tsunamis.Amendment 65 Page 79 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.7 ICE EFFECTS Ice formation in this locality is not expected to be severe enough under any circ*mstances to jeopardize the operation of the Cooling Tower Makeup Water System or the Emergency Service Water System. Minimum average temperatures, as recorded at Raleigh Airport, for the months of December, January and February are 30.5, 30.0 and 31.1°F, respectively.The Emergency Service Water Screening Structure draws cooling water by gravity from the Auxiliary Reservoir, the preferred source of emergency service water, through the Emergency Service Water Intake Channel. Pump suction for all pumps are located more than 10 ft. below the low water level. Therefore, considerable icing would be required in order to affect the operation of these pumps. This amount of icing is unlikely, based on data available from the National Oceanic and Atmospheric Administration Office at the Raleigh-Durham airport which show ice formation on large bodies of water in Central North Carolina is limited to minor freezing along the shoreline.However, in the unlikely event that ice is present in the Auxiliary Reservoir the Emergency Service Water Screening Structure is protected from ice entering the intake bays by means of a concrete baffle wall which extends from the deck (Elevation 262 ft MSL) to Elevation 247.5 ft MSL. This is below the normal Auxiliary Reservoir level. In addition each bay has a course screen consisting of 3-in. by 3/8 in. bars with a clear space of 3-in. between bars. Any ice fragments smaller than 3-in. that pass through the course screens will be picked up and disposed of by the traveling screens. Ice buildup on the traveling screens will be prevented by the use of heated hoods and by continuously running the screens if potential icing conditions warrant. The inlet of the gravity discharge pipe is located more than eight ft. below low water.The Emergency Service Water and Cooling Tower Makeup Intake Structure on the Main Reservoir serves as the backup source of emergency service water and the sole source of Cooling Tower makeup water. This structure draws water from the Main Reservoir through the Cooling Tower Makeup Water Intake Channel. The structure is protected from ice entering the pump bays by means of a concrete baffle wall which extends from the deck to one foot below the normal water level. In addition each pump bay has a coarse screen and traveling screen.The coarse screens and traveling screens are of the same description and have the same function as those for the Emergency Service Water Screening Structure described above. In addition, the traveling screens in the bays housing the emergency service water pumps are equipped with heated hoods. The pump suction inlet of all pumps in the Emergency Service Water and Cooling Tower Makeup Intake Structure is located more than 10 ft. below low water level. Therefore considerable icing would be required in order to affect the operation of these pumps.Ice formation in the Emergency Service Water Discharge Channel cannot jeopardize emergency service water system performance since the elevation of the Emergency Service Water System discharge point is above the high water Level resulting from the PMF.The plant will be shutdown and cooled down if ice formation in either intake channel would jeopardize the emergency service water supply.Amendment 65 Page 80 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.8 COOLING WATER CANALS AND RESERVOIRS The safety related cooling water channels (canals), reservoirs, and water control structures within the reservoir system of the Shearon Harris Nuclear Power Plant consist of the Main Reservoir, the Auxiliary Reservoir, the Auxiliary Reservoir Separating Dike, the Auxiliary Reservoir Channel, the Emergency Service Water Intake and Discharge Channels, and the Emergency Service Water and Cooling Tower Makeup Intake Channel.The design bases and operating modes of the reservoir system are described in relation to the safety-related Emergency Service Water System, Ultimate Heat Sink, and the Cooling Tower Makeup Water System; these discussions appear in Sections 2.4.11, 9.2.1, 9.2.5, and 10.4.5.Shearon Harris Nuclear Power Plant complies with NRC Regulatory Guide 1.127 (refer to Section 1.8) and Ebasco Specification CAR-SH-CH-24, "Reservoir, Dams and Dike Instrumentation Program (Non-Nuclear Safety)." In addition, the North Carolina Utilities Commission requires a dam inspection program involving independent consultants. As a minimum, the inspection program will include the water-control structures discussed in Section C.2 of Regulatory Guide 1.127. Periodic monitoring of embankment instrumentation will be performed. The Emergency Service Water Channels and Auxiliary Reservoir are monitored for sediment buildup.The Shearon Harris Nuclear Power Plant reservoir system constitutes the only water bodies that are of concern regarding protection of plant facilities from flood and wave runup; discussion of the protection of channels and reservoirs is contained in Sections 2.4.2, 2.4.3, 2.4.4, and 2.4.5.The only locations where potential blockage is of concern to safe plant operation are the Emergency Service Water Intake and Discharge Channels, and the Auxiliary Reservoir Channel. These channels are Category I structures and are designed to remain stable when subjected to the Safe Shutdown Earthquake or the most severe cases of other postulated natural phenomena (see Section 2.5.6). In the unlikely event of a slide of the earth slopes, the size of the channels is sufficient to pass the minimum required service water flow at a maximum velocity of 2 ft. per second under the conditions of maximum drawdown of the Main Reservoir and the Auxiliary Reservoir, as indicated in Section 2.4.11. Channel plans and sections are shown on Figures 2.5.6-6, 2.5.6-7, 2.5.6-8, and 2.5.6-28.The use of screens for the Emergency Service Water Screening Structure and the Emergency Service Water and Cooling Tower Makeup Intake Structure, the location of the intake structures, and the maximum velocity of 2 ft. per second in the channels provide assurance that no blockage of the intake structures, damage to the intake structures or damage to the emergency service water pumps can occur.The effects of failure of the Auxiliary Separating Dike are discussed in Section 2.4.4.The design bases for reservoir operation during periods of low water level are discussed in Section 2.4.11.Uncontrolled spillways at both dams are designed to provide release of flood waters so that the reservoir water levels do not exceed the design bases of the dams (see Section 2.4.3). There is no operational need for emergency storage evacuation of reservoir inventories, however, a low level release system is provided for the Main Reservoir to adjust reservoir discharge water Amendment 65 Page 81 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 quality to assure that downstream release criteria are met. Incorporated into the Main Dam spillway, it consists of three (3) Howell Bunger valves located in the central pier and side abutments of the spillway. The valves have intakes in the reservoir at different elevations and locations. The arrangement is shown in Figures 2.5.6-1, 3.8.4-34, and 3.8.4-36 and the discharge capacity curves are shown in Figure 2.4.8-1.The Howell Bunger valve in the central pier is a 24-inch valve with center line at El. 206.7 ft MSL. A 36-inch diameter steel pipe with intake at El. 195.0 ft MSL in the reservoir conveys water to the valve. The valves in the two abutments of the spillway are 36-inch valves with center lines at El. 213.0 ft. MSL. The intake for the West abutment valve is in the abutment at El. 213.0 ft MSL, whereas the East abutment has its intake inside the reservoir at El. 180.0 ft MSL, connected to the valve by a 48-inch diameter steel pipe.2.4.9 CHANNEL DIVERSION In view of the lack of historical evidence of any realignment of Buckhorn Creek, future realignment is considered to be extremely remote. Moreover, realignment in such a way that the runoff of the drainage basin would be diverted away from the reservoir system is impossible due to the contours of the basin.Likewise, due to the topography of its valley, realignment of the Cape Fear River is also considered extremely remote. As explained in Section 2.4.1, the comprehensive development plan for the Cape Fear River Basin proposed by the Corps of Engineers will function to control flood flow on the river and rather than diverting flow, will ensure the availability of a 600 cfs minimum flow at Lillington.Regardless of the continued availability of runoff to the reservoirs, the safety of the plant cannot be jeopardized by diverted flow. Operational commitments require shut down of the plant when reservoir water levels reach designated low points. At these low points there will still be sufficient water in the reservoirs to achieve safe shut down of the plant.2.4.10 FLOODING PROTECTION REQUIREMENTS The safety related facilities will not be affected by flooding resulting from reasonably possible combinations of the probable maximum flood (PMF), probable maximum wind (PMW), and the probable maximum precipitation (PMP).Sections 2.4.3, 2.4.4, and 2.4.5 discuss the maximum water levels in the Main and Auxiliary Reservoirs and around the plant island, where most of the safety related facilities are located.Section 2.4.2.3 discusses the water level on the plant island due to the PMP.The facilities located on the plant island will not be subjected to any flooding, as the plant grade, at Elevation 260 ft. MSL, is 2.3 ft higher than the maximum water levels around the plant island.The protections of structures on the plant island against the PMP are discussed in detail in Section 3.4.1.1. For the area between west face of Fuel Handling Building and the retaining wall see Section 3.4.1. Flood protection for safety related systems is discussed in Sections 3.4.1.The Emergency Service Water Screening Structure, the Emergency Service Water Discharge Structure, and the Emergency Service Water and Cooling Tower Makeup Intake Structure are Amendment 65 Page 82 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 designed so that their decks are above all calculated water levels. Since they extend down below normal water levels, they are also designed to withstand forces that could result from the worst postulated flood, wave, and wind conditions.Safety related facilities other than those located on the plant island are the Main Dam, the Auxiliary Dam, the Auxiliary Reservoir Separating Dike, and the Auxiliary Reservoir Channel.The dams, dike, and channel are discussed in Section 2.4.4, 2.4.8, and 2.5.6. The top of the Main Dam is at Elevation 260 ft MSL, which is 16.9 ft above the maximum water level at the Main Dam. Therefore, the dam will not be overtopped. The top of the Auxiliary Dam is at Elevation 260 ft MSL, which is 2.0 ft higher than the maximum water level in the Auxiliary Reservoir at the Auxiliary Dam.The Auxiliary Separating Dike, with its crest at Elevation 255 ft. MSL, will be subjected to overtopping due to waves generated by the probable maximum hurricane (PMH) wind (123 mph) acting on the normal reservoir level, Elevation 252 ft MSL or winds up to 54.4 mph combined with the PMF level, Elevation 256 ft MSL. The upstream and downstream slopes of the dike are protected by riprap, as discussed in Section 2.5.6. The failure of the Auxiliary Reservoir Separating Dike, as discussed in Section 2.4.4, will not result in loss of water or storage capacity and the Main Reservoir will function as a back-up source of cooling water to provide an adequate cooling circuit through the reservoirs.As discussed in Section 2.4.4, the Main and Auxiliary Dams will not be subjected to any dynamic forces due to flooding, other than local wave action which is dissipated by the use of riprap. The dams and associated spillways have been designed for hydrostatic forces corresponding to the PMF levels in the two reservoirs.2.4.11 LOW WATER CONSIDERATIONS 2.4.11.1 Low Reservoir Level The synthesized flows of Buckhorn Creek for 1924 to 1981 were analyzed to determine the most critical low flow period. This period was determined to be a 19 month historical drought of May 1980 through November 1981. The actual analysis was carried out for May 1980 through May 1982 to demonstrate reservoir recovery.Although the earliest synthesized Buckhorn Creek flow data dates back only to 1924, there are good precipitation records at the Raleigh National Weather Station which go back to 1867. A review of these records indicates that the lowest annual precipitation occurred in 1933, with a total of 29.93 in. Near record lows were also experienced in 1930, 1941, 1951, 1968, and 1977.It is conceivable, therefore, that the minimum flow experienced during the period of synthesized runoff may represent the lowest values of runoff dating back to 1867, the beginning of precipitation data. However, drought frequency analyses were made based on the period when flow could be synthesized from regional streamflow data, and therefore the results should be conservative.2.4.11.1.1 Worst Critical Period From a seven-year reservoir operation study for the period 1973 through 1980, the mean water level in the Main Reservoir is 219.4 ft. MSL as discussed in the SHNPP Environmental Report, Section 2.4.2.3.2.2.3. This elevation varies depending on actual inflow conditions, consumptive Amendment 65 Page 83 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 use, and downstream releases. Calculations were made to determine the reservoir level fluctuation during the worst drought period of record for Buckhorn Creek under the worst monthly evaporation conditions. Table 2.4.11-18 provides results of analysis for the period May 1980 through May 1982 with the assumption that the reservoir was at the minimum predicted normal operation level at 216.3 ft. MSL.The total annual evaporation of 63.67 inches for the worst meteorological conditions was used in Table 2.4.11-18. This estimate of natural evaporation was obtained from a method of maximization using the meteorological data shown in Table 2.4.11-3. The basis for selecting these monthly meteorological data was the maximum monthly average temperature of record at the first-order National Weather Station at Raleigh, North Carolina during the period 1931-1970.Plate 2 of the Environmental Science Service Administration (ESSA) Climatic Atlas (Reference 2.4.11-2) shows the mean annual lake evaporation in the Raleigh area to be approximately 41.0 in. Plate 3 of the Atlas shows a mean annual Class A pan coefficient of 75.5 percent for the site area, and Plate 5 indicates a standard deviation of annual Class A pan evaporation of about 3 in. The estimates of evaporation under normal meteorological conditions of 53.35 in. and under worst meteorological conditions of 63.67 in. are considerably more conservative than the 41 in.shown on Plate 2 of the Atlas. Another direct indication of the degree of conservatism in these estimates is provided by a calculation of the number of standard deviations from the average evaporation value of 41 in. The estimated normal evaporation of 53.35 in. under normal meteorological conditions is 4 standard deviations above the Atlas average. The evaporation under worst conditions (63.67 in.) represents 7.5 standard deviations above the Atlas average.The period had a 4-month average flow of 10.5 cfs, a 7-month average of 11.5 cfs and a 12-month average of 25.1 cfs and an average flow over the drought period of 26.6 cfs (DA = 71 sq.mi.).The maximum storage use was 34833 ac. ft. and the minimum water level was 209.4 ft. MSL.2.4.11.1.2 100-Year drought The average 4-month, 7-month, and 12-month flow of Buckhorn Creek corresponding to the 100-year drought discussed in Section 2.4.11.2 was utilized in the analysis. Evaporation rates are for worst monthly evaporation conditions. As shown in Table 2.4.11-19, the maximum Main Reservoir storage utilized for this extreme drought condition would be 30342 ac. ft. drawing the reservoir down to Elevation 211.0 ft. MSL.As discussed in Section 2.4.11.7, the Tech Spec minimum Main Reservoir Level is 206 ft. to ensure flow requirements to safety related heat exchangers, cooled by Emergency Service Water, are met. The unit will be shutdown if the water level in the Main Reservoir falls below 206 ft.A study was performed to address the statistical probability of Main Reservoir level falling below a level of 215.0 ft. MSL. A Main Reservoir impoundment analysis was performed using historical synthesized monthly average flows of Buckhorn Creek taken from Table 2.4.1-1 and Table 2.4.11-18. From this study, average monthly reservoir levels were calculated for the period from 1923 to 1982 (713 months). These monthly levels generally trended down each summer and fall and then recharged to normal pool levels during the winter and spring. This coincides with the normal climatological precipitation patterns for this region. However many months experienced low precipitation and occasionally this occurred for many consecutive Amendment 65 Page 84 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 months. During a few of these abnormally lengthy dry periods the Main Reservoir level actually dropped below elevation 215.0 ft. and is summarized below:

1. October 1951 through January 1952 (4 months)

2. July 1981 through January 1982 (7 months)

Both of these dry periods were preceded by an unusually dry winter and spring whereby the reservoir was not recharged to normal pool level. Then during the climatologically drier summer and fall, the reservoir levels dropped quickly to less than 215.0 ft. The lowest level calculated was in November 1981 when the reservoir fell to 212.9 ft.The monthly average reservoir levels for each of the 713 months in the study were evaluated.The calculations showed there is less than a 1.0% chance for any given month that the Main Reservoir level will fall below a level of 215.0 ft. MSL.2.4.11.2 Low Flow in Streams Short duration minimum flow in Buckhorn Creek has little effect on the project due to the large storage in the Main Reservoir. Therefore, isolated periods of less than four months were not considered. Drought periods of four months and seven months were determined for the period of synthesized record of Buckhorn Creek from 1924 to 1981. The two worst drought periods were the periods February 1951 through January 1952 and August 1980 through July 1981. The 1950-1951 period had a 4-month average of 5.8 cfs, a 7-month average of 8.5 cfs, and a 12-month average of 26.5 cfs at the Main Dam. The 1980-1981 period had a 4-month average low flow of 10.5 cfs, a 7-month average low flow of 11.5 cfs and a 12-month average low flow of 25.1 cfs.Based on the available data and the principle of the Markov chain process, synthesized flow of Buckhorn Creek was generated according to a statistical method described in Reference 2.4.11-

2. The frequency analysis of the synthesized flow was then utilized to estimate a severe drought having a return period of 100 years. The following gives the average minimum flow for the various 100-year return period drought durations:

Buckhorn Creek Flow at Main Dam (DA=71.0 sq.mi.) (cfs)Average minimum 4-month flow 3.7 Average minimum 7-month flow 6.9 Average minimum 12-month flow 23.2 2.4.11.3 Low Water Resulting From Surges, Seiches, and Tsunamis The water level in the Auxiliary Reservoir is maintained at a minimum elevation of 250 ft. MSL by pumping water from the main reservoir. Since the bottom of the Emergency Service Water Intake Channel is at Elevation 238 ft. MSL, a minimum water depth of 12 ft. will be maintained in the channel. During the probable maximum hurricane (PMH), the maximum wave height would be 5.4 ft. and the maximum wind set-up would be 0.4 ft. (see Table 2.4.5-1). Therefore the lowering of the water level resulting from a PMH induced surge in comparison with the 12-foot Amendment 65 Page 85 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 depth of intake water would not significantly affect the intake capacity of the Emergency Service Water System.The normal operation water level in the Main Reservoir is at Elevation 220 ft. MSL. Since the bottom of the Cooling Tower Make-Up Water Intake Channel is at Elevation 194 ft. MSL, this 26-foot difference provides sufficient margin to withstand the lowering of the water level in the Main Reservoir by surges due to the PMH because the resulting wave height would be 6.7 ft.and the wind setup would be 1.0 ft. (see Table 2.4.5-2).Due to its inland location the site is not subject to tsunamis (see Section 2.4.6).As the site is located in a region of minor seismic activity and less severe meteorological events, the occurrence of a seiche is a remote possibility. In the unlikely event of a seiche, the wave amplitude would be such that surges due to the PMH described above still present the worst case.The absence of expected icing concerns and the design consideration given to unlikely ice formation in the reservoir system is discussed in Section 2.4.7.2.4.11.4 Historical Low Water Historical low flow data for Buckhorn Creek were derived from the recorded data of Middle Creek because only limited low flow data during isolated periods is available for Buckhorn Creek prior to June 1972, as described in Section 2.4.1.2.1.1. Based on the drainage areas ratio, the calculated minimum flow data for Buckhorn Creek from 1940 to 1978 are presented in Table 2.4.11-12. Calculated data, based upon actual measured data as adjusted by drainage area relationships, are also shown in this table for periods starting with 1973, the first whole water year following the establishment of a USGS gage station. As can be seen from this table, Buckhorn Creek has experienced some periods with no flow. The frequency analyses for 1, 7, 30, and 60 consecutive days low flow are shown on Figures 2.4.11-2, 2.4.11-3, 2.4.11-4, and 2.4.11-5, respectively. The water level in the Main Reservoir associated with the drought periods on record has been studied in detail; the minimum water level elevation (conservatively based on four unit operation) following a subsequent 30-day emergency condition, is 204.4 ft.MSL as discussed in Table 2.4.11-14.2.4.11.5 Future Controls No future uses of Buckhorn Creek which could significantly affect the natural creek flow are foreseen at this time. The minimum flow conditions for the creek are presented in Section 2.4.11.2.Since the ability of safety related facilities to function adequately under drought conditions has been confirmed in Section 2.4.11.1, provisions for stream flow augmentation for plant use are not required.2.4.11.6 Plant Requirements Safety related water requirements for the plant are supplied by the Emergency Service Water System, described in Section 9.2.1. The Emergency Service Water System is a once through, open cycle design with respect to the reservoir system of SHNPP.Amendment 65 Page 86 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Service water during normal plant operation is supplied from and returned to the Cooling Tower.The make-up requirement to the Cooling Tower from the Main Reservoir constitutes a major plant use during normal operation; Cooling Tower make-up is a maximum of 16,000 gpm operating at peak evaporative rates (12,000 gpm evaporation, 4,000 gpm blowdown, 10 gpm drift).Additional quantities of reservoir water pumped by the Cooling Tower make-up pumps are diverted for use as make-up to the Water Treatment System (1,028 gpm), as Auxiliary Reservoir makeup (2,000 gpm), by the screen wash pumps (540 gpm), and by the self-cleaning strainers of the trash collection system (360 gpm). The total withdrawal from the Main Reservoir is therefore a maximum of 19,938 gpm.However, Cooling Tower blowdown (4,000 gpm), screen wash water, and strainer backwash are returned to the Main Reservoir. The net consumptive use of reservoir water is therefore 15,038 gpm for the plant under maximum evaporative conditions, assuming all secondary services of the Cooling Tower make-up pumps are required simultaneously.During emergency service water operation, the Auxiliary Reservoir is the preferred source of water and the discharge is returned to the Auxiliary Reservoir. When the Auxiliary Reservoir is not available for supplying emergency service water, the Main Reservoir will serve as the backup source with discharge to the Auxiliary Reservoir.The safety related cooling water flow requirement for various modes are provided in Table 9.2.1-1. The availability of water during the 30-day period following an accident is discussed in Section 2.4.11.7 and Table 2.4.11-14. Figures 3.8.4-25 through 3.8.4-31 show the intake structures including sump invert elevation and configuration, minimum design operating water level, and pump submergence elevation.The thermal effluent released during emergency shutdown will have a temperature not exceeding 120°F, while the radioactive release is maintained as low as reasonably achievable.The effluent will be discharged into the Auxiliary Reservoir through the Emergency Service Water Discharge Channel. The design bases for effluent mixing and dispersion from the Service Water System under emergency conditions are the same as for the Circulating Water System, described in Section 2.4.12.The 100-year drought water level elevation in the Main Reservoir has been incorporated in the study of the previously mentioned 30-day period following an accident (see Table 2.4.11-14).2.4.11.7 Heat Sink Dependability Requirements The ultimate heat sink is a complex of water sources, including associated retaining structures and any canals or conduits connecting the sources with the cooling water intake structures of the plant. The sources of water for the normal and emergency cooling modes are the Cooling Tower and the Main or Auxiliary Reservoirs, respectively. The retaining and conveyance systems consist of the Main Dam, the Auxiliary Dam, the Auxiliary Reservoir Separating Dike, the Auxiliary Reservoir Channel, the Emergency Service Water Intake and Discharge Channels, and the Cooling Tower Makeup Water Intake Channel. These facilities are described in Sections 2.4.4 and 2.4.8.Amendment 65 Page 87 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The normal source of plant service water is the circulating water Cooling Tower (see Section 9.2.1). The preferred source of emergency cooling water, if the non-safety related Cooling Tower or its associated components are not available, is the Auxiliary Reservoir. The backup source of emergency service water is the Main Reservoir if the Auxiliary Reservoir is not available. Emergency Service Water from either source is discharged to the Auxiliary Reservoir.The Auxiliary Reservoir will perform its function as the ultimate heat sink in the event of a loss of service water from the Cooling Tower. Figure 2.4.3-6 shows the area-capacity curve for the Auxiliary Reservoir. A minimum level of 250 ft. Elevation in the Auxiliary Reservoir is maintained at all times during normal operation by creek inflow above the Auxiliary Dam and by pumping water from the Main Reservoir. The Main Reservoir is normally maintained at a level of 220 ft.MSL, but may decrease to 206 ft. MSL during normal operation. The UHS analysis uses the Main Reservoir level of 205.7 ft. as the starting point to determine final UHS temperature and level and if an adequate volume of water exists to remove the heat generated by the plant.Level was determined to decrease to 203.6 ft. MSL during emergency conditions. While at their minimum normal operation levels, both the Auxiliary and Main Reservoirs, taken separately, are more than adequate to permit emergency shutdown and cooldown of the plant.The Main Reservoir will function as a cooling reservoir in the case where the Auxiliary Reservoir is not available. The Auxiliary or Main Reservoir will function as cooling reservoirs during an emergency.Emergency and normal shutdown heat loads are discussed in Section 9.2.1. The service water system heat exchanger design basis temperature is 95°F, and the estimated maximum service water system inlet temperature for the postulated conditions of emergency shutdown and cooldown is 94.6°F. 94.6°F is acceptable as a maximum pre-accident temperature since the Ultimate Heat Sink analysis does not account for thermal stratification which would result in a maximum, post-accident (30-day) pump suction temperature below 95°F (Ref. SW 0085). The meteorological conditions postulated in this case are considered to be the most critical conditions for maximum inlet temperature. The basis for these conditions is a statistical analysis of the meteorological parameters directly related to water temperature; that is, solar radiation, ambient air temperature, dew point temperature, and wind speed. The meteorological conditions that maximize the service water temperature are high solar heating, high ambient air temperature, high relative humidity, and low wind speed. The worst meteorology for one day occurred on June 27, 1952, and the worst month occurred between July 18 and August 15, 1949. The average values of these meteorological parameters for 1-day to 30-day periods were calculated based on these days to estimate the maximum service water system inlet temperature. The limiting analysis calculated a maximum post-accident reservoir temperature of 94.6°F using a composite 9-day period including the worst 8-day consecutive meteorology (7/22/49 - 7/29/49) plus the worst 1-day (6/27/52). The auxiliary Reservoir initial water temperature of 82.2°F is assumed in the analysis based on the July reservoir equilibrium water temperature for the normal meteorological conditions. The analysis also found that a pre-accident reservoir temperature of 94°F would result in a final, 1-day, post-LOCA temperature of 95.17°F, which is just slightly above the ESW design-basis temperature of 95°F. This is acceptable because the reservoir analysis does not account for thermal stratification which would result in a pre-accident temperature below 94°F and a post-accident temperature below 95°F. The analysis accounts for auxiliary reservoir inventory loss via valve seat leakage into a completely empty main reservoir. (Reference Calc SW-0085.)Amendment 65 Page 88 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The normal meteorological conditions of each month for the site area are presented in Table 2.4.11-2 and the corresponding reservoir equilibrium water temperatures (in °F) are:JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC 39.6 43.2 52.6 62.9 73.1 79.5 82.2 80.7 74.3 62.2 49.5 39.1 The consumptive use of water during the emergency period is a function of plant heat rejection and evaporation. Using the most severe monthly evaporation rates based on the worst single month meteorological data (July, 1932, listed in Table 2.4.11-3), the total drawdown of the Auxiliary Reservoir for the 30-day emergency period was computed for accident conditions for one unit. In estimating the evaporation rate, the Auxiliary Reservoir initial water temperature was conservatively assumed to be 95°F.For the purpose of the calculation, rainfall and inflow are assumed negligible during the emergency period. The Auxiliary Reservoir water level decreases only 1.5 ft. Therefore, the reservoir provides an adequate supply of water during the emergency period.The analysis is performed to demonstrate the adequacy of the Main Reservoir to provide at least a 30-day supply of water for emergency shutdown and cooldown of one unit in the event that the level in the Main Reservoir drops to Elevation 205.7 ft. MSL. This analysis is based upon the assumptions that:

1. Rainfall during the 30-day emergency period is zero,

2. Buckhorn Creek inflow is zero,

3. The auxiliary reservoir water level is maintained at a minimum elevation of 250 ft. MSL during normal operation, and the spillover elevation to provide the spill rate of 200 cfs into the Main Reservoir is 252.5 ft. MSL.

4. Worst single month meteorological data (July 1932) is used for the maximum evaporation.

5. The Main Reservoir and the Auxiliary Reservoir initial water temperatures are 95°F.

The Main Reservoir water level at the end of the 30-day emergency period would be Elevation 203.6 ft. MSL which is above the minimum operating level of the Service Water and Cooling Tower Make-up Water Pumps.The UHS analysis uses the Main Reservoir level of 205.7 ft. as the starting point to determine final UHS temperature and level, and if an adequate volume of water exists, to remove the heat generated by the plant. However, to ensure the flow requirements for safety related heat exchangers cooled by Emergency Service Water (see Section 9.2.1) are met, the UHS minimum Main Reservoir level is 206 ft.Ample warning of low water level in the Main Reservoir will be available since the low water conditions produced by the 100-year drought would be of such unusual severity that the conditions will be obvious weeks prior to reaching MSL critical level. Plant operators, especially, will be cognizant of the condition in the Main Reservoir due to the unusual make-up requirements of the Auxiliary Reservoir during these weeks. In addition, a level transmitter is Amendment 65 Page 89 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 located at the Emergency Service Water and Cooling Tower Make up Water Intake Structure which alarms in the Control Room when the water level in the Main Reservoir decreases to Elevation 207 ft. MSL.Section 2.4.11.1 discusses the reservoir capacity for heat dissipation and the assumed concurrent water losses under normal operation and emergency conditions, respectively. For the description of compliance with Regulatory Guide 1.27, see Section 1.8.Other than service water, the Auxiliary Reservoir is the source of firefighting water and certain non-safety related HVAC cooling water makeup. It is not postulated that a fire and an accident that requires plant shutdown would occur simultaneously. The firefighting water system has a design capacity of 2500 gpm, and it is activated only if a fire occurs. HVAC cooling water makeup occurs on an intermittent basis and requires a maximum of 100 gpm. These requirements are considered insignificant.Section 2.4.4.1 discusses the capability of each (Main & Auxiliary) dam to withstand the failure of the other dam.Volumes of potential sediment deposit from Buckhorn Creek were determined to be on the order of 460 and 20 ac-ft., respectively, in the Main and the Auxiliary Reservoirs for the length of the plant life of 40 years. These amount only to 0.7 percent and 0.4 percent of the respective reservoir capacities at their normal water levels.The USGS sediment sampling data for Buckhorn Creek at Corinth, North Carolina were analyzed to obtain a sediment rating curve, as shown on Figure 2.4.11-9. This rating curve was combined with 40 years of synthetic daily streamflow to estimate the quantity of direct sediment inflow from Buckhorn Creek. To generate the synthetic daily streamflow data, the streamflow records of Buckhorn Creek were supplemented with the records of Middle Creek at Clayton, North Carolina and two computer programs (References 2.4.11-4 and 2.4.11-5) were used.In view of the small quantities predicted, the effects of sediment deposit on reservoir operations and cooling capacities will be negligible. The critical levels of reservoir drawdown during drought or emergency periods discussed above will be essentially unaffected because the sediment aggregation found for the reservoirs is not significant enough to cause any appreciable changes to the reservoir area-capacity curves. The same argument applies to the effects of sedimentation on reservoir cooling capacities.During the License Renewal review, the evaluation of sedimentation in the Main and Auxiliary Reservoirs was considered. Sedimentation effects were determined to be negligible and would be managed by monitoring activities during the period of extended operation. Refer to Chapter 18 for additional information.2.4.12 DISPERSION, DILUTION AND TRAVEL TIME OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS Accidental discharges of radionuclides to the surface waters could conceivably occur due to failure of storage tanks which contain potentially radioactive liquids or due to inadvertent discharge of liquid waste to cooling tower blowdown. Inadvertent discharges are precluded by automatic termination of liquid waste discharge on high radiation signal (see Section 11.2).Amendment 65 Page 90 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.12.1 Storage Tank Failure.The contents in the tanks located in either the Waste Processing Building (Figures 1.2.2-47 through 1.2.2-54) or the Tank Building (Figure 1.2.2-84) are potentially radioactive.The Waste Processing Building is a Seismic Category I concrete structure which contains such tanks as the floor drain tank, laundry and hot shower tank, and waste hold-up tanks. The building is designed to retain the liquid contents of all tanks containing potentially radioactive liquids. The tanks are located below plant grade, and therefore their failure would not result in runoff to the surface waters.The Seismic Category I Tank Building contains large capacity tanks, such as the refueling water storage tank (RWST), the condensate storage tank, and the reactor makeup water storage tank (RMWST). Each of these tanks is housed in an individual compartment capable of retaining the tank contents.As discussed in Section 11.2, an inadvertent release of radionuclides via the normal liquid effluent release pathways will be automatically terminated. For the purpose of investigating the effects of an accidental release of liquid effluents in surface waters, the unlikely failure of a storage tank and its concrete enclosures is postulated to occur.The RWST has both the highest specific and total activity of the three tanks located above grade in the Tank Building. Table 2.4.12-1 lists the concentrations of nuclides contained in the refueling water storage tank.Any accidental releases into the Main Reservoir are afforded dilution by the Main Reservoir, Buckhorn Creek, and the Cape Fear River before reaching Lillington, the location of the closest downstream surface water user (see Tables 2.4.1-5 and 2.4.1-6). Any spillage from the plant island would enter the reservoir in its northern reaches and would discharge into the Cape Fear River of the Main Dam Spillway at the reservoir's extreme southern end. Any spillage would essentially achieve complete mixing before being discharged since the flow through the reservoir is relatively slow.Even though it is not considered likely, if the entire contents of the RWST (470,000 gals, see Table 6.2.2-9) were released to the lake instantaneously, the dilution provided by the volume of water in the Main Reservoir, conservatively chosen to be 62,000 acre feet (volume of the Main Reservoir when two ft. below normal operating level), would be on the order of 105.Also, note that the following assumptions add to the conservatism of the analysis. The radionuclide content in the RWST was maximized by assuming the reactor coolant activity to be based on 1 percent failed fuel. The design basis activity in Table 11.1.7-1 is also hom*ogeneously mixed in the RWST. Further, it was assumed that the Spent Fuel Pool Filtration System was not operating during refueling. The instantaneous mixing of the tank contents in the reservoir is also a conservative assumption. The RWST is located approximately 1500 ft.from the reservoir inside the Tank Building. Even spilled liquid would have to travel 1500 ft.overland where some absorption and retention of radionuclides in the soil would take place.Instantaneous mixing removes from consideration the travel time (and consequently the radioactive decay) for the water to migrate to the reservoir and neglects the possibility that portions of the waste would be prevented from reaching the reservoir by retention in the soil. If Amendment 65 Page 91 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the real case were analyzed, absorption and retention would lessen the amount of activity present in the reservoir at any given moment.Specific concentrations at Lillington, provided in Table 2.4.12-1, show the results of plant operation (reservoir volume - 2.92x109ft3 and reservoir discharge = 43 cfs). The reservoir flow was diluted by the minimum annual average flow of 1350 cfs (occurred in 1981, see Table 2.4.1-3) in the Cape Fear River. Any natural runoff into Buckhorn Creek between the Main Dam and its confluence with the Cape Fear River was conservatively assumed to be negligible. It can be seen from Table 2.4.12-1 that the C/MPC at Lillington is well below the allowable concentrations of 10 CFR 20.2.4.13 GROUNDWATER The project is located essentially within the Buckhorn Creek watershed of the Piedmont providence near the Fall line, the physiographic limit between the Coastal Plain and the Piedmont Plateau, in east-central North Carolina. Buckhorn Creek is a tributary of the Cape Fear River. The towns of Bonsal, New Hill, and Corinth are located within the watershed.Based on Raleigh Durham Airport (RDU) data, the average annual temperature in the area is 59.1 F and the mean annual precipitation is 42.5 in. This section presents groundwater conditions, sources, and usage of the aquifer in the area and at the site (Figure 2.4.1-1).2.4.13.1 Description and Onsite Use 2.4.13.1.1 Regional Groundwater Conditions The entire drainage area of Buckhorn Creek northwest of the Jonesboro Fault is underlain by Triassic rocks of the Newark Group. The drainage area of Buckhorn Creek that is located southeast of the Jonesboro Fault is relatively small and is underlain by Paleozoic crystalline rocks and igneous intrusives, as well as metamorphic rocks of the Carolina Slate Belt. Both the Triassic and Pre-Triassic rocks are overlain by clayey soils and saprolite.a) Overburden - The plant area is covered with residual soils derived from the underlying rocks. Numerous soil borings drilled at the plant island, as well as in the Auxiliary Reservoir area, confirm the existence of up to about 15 ft. of clayey soil and saprolite overlying the Triassic rocks.Excavation and mapping of trenches in the plant site area, as well as excavation and borings for the Site Fault Investigation (Reference 2.4.13-1), also indicate the preponderance of clayey and silty loam soils.b) Triassic Rocks - The source of groundwater in the area is the rock units of the Sanford Formation of the Newark Group (Triassic). They consist of claystone, shale, siltstone, sandstone, conglomerate, and fanglomerate. An exception to this lithology is the intrusion of thin diabase dikes in the rock (Reference 2.4.13-1 and 2.4.13-2). The dikes were mapped in connection with the site fault investigation in the plant and Auxiliary Dam areas (Reference 2.4.13-1). The diabase rock is weathered near the surface and is unweathered below depths of about 20 ft.The primary permeability of the Triassic rocks is very low and the rocks appear to be essentially dry. Some lenses of relatively higher permeability exist within the Triassic rocks.Amendment 65 Page 92 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 However, they are not extensive and are surrounded by materials with relatively lower permeability. The Triassic rocks have fractures resulting from stress releases; the fractures provide secondary permeability in the rocks and are filled with water below the water table.The fractures are common to depths of about 100 ft., but become less prevalent and are tight below that depth. Below about 400 ft., the fractures are closed and sealed to water flow, as shown by tests and by experience gained through well drilling in the area.Recharge in the area occurs by percolation of precipitation through the overburden; however, most of the precipitation is either returned to the atmosphere through evapo-transpiration or becomes surface runoff. The predominance of surface and near-surface deposits with extremely low permeability results in rapid runoff of precipitation. Therefore, natural recharge to the aquifer occurs at a very low rate.The precipitation which percolates downward is confined laterally by the diabase dikes and vertically by the absence of open fractures or joints at depth in the Triassic rocks.Numerous attempts to develop groundwater supplies from deep Triassic rocks have not been successful since these rocks are tight and relatively dry. However, groundwater is developed in the Triassic basin from hornfels zones adjacent to diabase dikes. The relationship of dikes and fractures to groundwater flow is illustrated diagrammatically on Figure 2.4.13-3.The use of groundwater in the site region is limited because of the low yield of the aquifer; most of the wells are for domestic use. A few small towns in the area (see Table 2.4.13-1 and Figure 2.4.13-4) use the Traissic rocks as a source of water. However, total groundwater usage is still small, as discussed in Section 2.4.13.2.2.4.13.1.2 Site Groundwater Conditions Investigations conducted at the site reveal that geologic and hydrologic conditions in the site area are essentially the same as the regional conditions described in Section 2.4.13.1.1.The plant site is located on a ridge bounded by Thomas Creek to the east, Tom Jack Creek to the west, and White Oak Creek to the southeast. These creeks are tributaries of Buckhorn Creek, which enters the Cape Fear River about seven miles south-southwest of the site. The plant site has been graded to Elevation 260 ft. msl. The pre-graded site elevations ranged from about 210 ft. to 280 ft. msl; the land surface generally sloped towards the east and southeast.a) Top Layer - The Soil Conservation Service soil survey of Wake County, 1970, classified the site soils as the Creedmoor-White Store Association (Reference 2.4.13-3). Some typical engineering properties of the Creedmoor White Store soil series, as mapped in the site area and taken from the Soil Conservation Service soil survey of the Wake County, 1970, are listed below. They indicate that the Creedmoor-White Store soil conditions are relatively impervious. The surficial clay and saprolite zones prevent ready recharge to the rocks below them, as indicated by the dry state of the rocks (Reference 2.4.13-1).CREEDMOOR-WHITE STORE ASSOCIATION CREEDMOOR SOIL (Typical Profile)Percentage Passing Permeability Shrink-Swell Depth (in.) Texture Sieve No. 200 (0.074 (in./hr.) Potential

 )

Amendment 65 Page 93 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 0-12 sandy loam 30-45 2.0 - 6.3 Low 12-29 clay loam 35-85 0.63 - 2.0 Moderate 29-58 clay 70-95 0.2 High 58-96 clay 35-90 0.2 Moderate b) Triassic Rocks - The plant site and peripheral lands are underlain by Newark Group rocks (Triassic). The tightness of the rock formations, as a result of compaction and cementation, is evident from most of the cores extracted from the borings. Surface percolation of precipitation is controlled by the location of joints and fractures in the rocks. The low permeability of the Triassic rocks suggests that any movement of groundwater within the rocks also will be controlled by the interconnecting patterns of joints and fractures. Groundwater at and around the site occurs principally within jointed rock, generally at depths of 30 to 90 ft. below the original ground surface. Within the Newark Group, larger reserves of groundwater occur in the proximity of diabase dikes.Several small dikes were found in the plant area; groundwater supplies for use during plant construction have been developed in the proximity of the dikes.Seven wells with a total capacity of about 200 gal./min. were completed during 1973 and are being used during the construction phase. Additionally, eight new wells, which increased the total capacity of all wells to about 450 gal./min., were developed in the proximity of diabase dikes during 1977-1979; three more in 1980 and two more in 1981 (Table 2.4.13-2).2.4.13.1.3 Onsite Use of Groundwater Site wells are listed in Table 2.4.13-2 and are shown on Figure 2.4.13-1. Groundwater is being used at the site during the construction phase for (1) concrete batch plant and concrete placement, (2) office and plant use, and (3) grouting. Groundwater is not expected to be used for plant operation after the plant potable water system becomes operational. Estimated monthly groundwater consumption at the site for March, 1978, through February, 1980, is shown in Table 2.4.13-3. The estimated plant water requirements through the year 1982 are shown in Table 2.4.13-4. Carolina Power & Light Company is the principal user of groundwater within two miles of the plant; there are only two domestic users within two miles of the plant, and both are up gradient near the 7,000 ft. radius boundary.2.4.13.2 Source 2.4.13.2.1 Regional Use of Groundwater Public wells within ten miles of the site are listed in Table 2.4.13-1 along with their maximum pumpage rates, water levels, and use; Figure 2.4.13-4 shows the locations of the wells.The nearest communities that are using groundwater for public water supply are Holly Springs and Fuquay-Varina; both are in Wake County. Holly Springs, about seven miles east of the plant site, has two wells which supply a total of about 40,000 gallons per day. Fuquay-Varina, about ten miles southeast of the plant site, has eight wells which supply a total of about 400,000 gallons per day. The wells produce water from crystalline rocks of the Carolina Slate Belt; in the plant area, the same crystalline rocks are buried a few thousand feet beneath the Triassic sediments. The Holly Springs and Fuquay-Varina wells are not located in the Triassic Basin.Amendment 65 Page 94 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The closest community downstream of the plant site is Corinth, approximately five miles to the southwest, where a group of houses have individual wells of minimal production from the Triassic, Newark group aquifer. The well depths range from 62 ft. to 140 ft., while their production varies from 0.5 to 13 gpm. The relative yields of the wells at Corinth are all less than 0.10 gpm per foot of uncased hole.2.4.13.2.2 Groundwater Levels and Movement A piezometric-level map (Figure 2.4.13-1), based on water-level measurements taken before commencement of full-scale plant construction, shows that the general groundwater movement in the plant area at that time was to the southeast toward White Oak Creek. Most of the original site-area piezometers have been lost due to construction activities; therefore, sixteen new piezometers were constructed in December, 1979. A piezometric-level map, based on water-level readings in production wells, in the sixteen new piezometers, and in two old piezometers, taken during the winter of 1979-1980, is shown on Figure 2.4.13-2. The map is based on the highest water levels observed during this period (Tables 2.4.13-5 and 2.4.13-6) and they do not necessarily represent static water levels. Figure 2.4.13-2 shows that the general direction of groundwater movement at the site is still to the southeast toward White Oak Creek. However, the water levels have been significantly altered due to the ongoing pumpage from the site wells.Figure 2.4.13-2 shows that cones of depression have developed on the northeastern and southwestern sides of the plant. Three of the piezometers have since been abandoned and one piezometer was destroyed as indicated on Table 2.4.13-2.2.4.13.2.3 Aquifer Characteristics The Triassic rocks underlying the plant site constitute the principal aquifer in the area of the plant and reservoirs. The thin layer of overburden overlying the Triassic rock consists of clayey soils and saprolite which yield little or no usable groundwater.The Triassic rocks of the aquifer are quite thick and widespread in extent. However, because of compaction and cementation of individual rock layers, it can be regarded only as a minor aquifer. Yields from known wells in the area generally range up to 20 gpm, but average only about 5 gpm or about 0.03 gpm/ft. of well (References 2.4.13-2 and 2.4.13-4). Generally, the principal areas of groundwater storage in the Triassic Basin are found near diabase dikes which have intruded the Triassic sediments. Twelve wells which were developed in the proximity of the dikes in the site area are providing water for use during construction of the plant.Even though Triassic rocks constitute the major groundwater source within the site environs, they exhibit very low permeability for groundwater storage and movement. Of the 57 wells with an average depth of 158 ft. that have been constructed in the Triassic formation in western Wake County, 16 percent yield less than 1 gpm, while the average production rate is 5 gpm.Such relatively low permeability also explains why the Triassic formation is the lowest producing groundwater source in the region (Reference 2.4.13-2). Numerous borings carried out for soils and geologic information in the plant site and reservoir areas confirm the very low permeability of the Triassic formation. The permeabilities of materials in the Auxiliary Reservoir and SHNPP site areas, based on pump-in packer tests, are summarized in Table 2.4.13-7.Six site wells located in the proximity of diabase dikes yielded specific capacity values from 24-hour driller's tests that range from 0.16 gpm/ft. to 0.59 gpm/ft. The specific capacity values correspond to transmissivity values of about 40 ft.2/day to 130 ft.2/day (Reference 2.4.13-5).Amendment 65 Page 95 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The beds below the surface clay and saprolite zones appear to have two distinct components of permeability. There is a very low permeability in the materials themselves. The second component of permeability comes from fractures that have resulted from stress release. This is the principal component which is measured as permeability during pump-in tests at the site.The fractures in the Triassic rocks are filled with water below the water table. The fractures are common to depths of 100 ft. but become less prevalent and tight below that point. Below 400 ft., the fractures are closed and sealed to water flow, as shown both by tests and by experience gained through private well drilling in the area. This fracture relationship is illustrated on Figure 2.4.13-3.Small amounts of groundwater were encountered in some trenches where fractured rock was evident. After their excavation, the trenches continued to hold surface runoff due to the low permeability of the fine-grained soil and rock materials.Down-hole pressure testing of the soils and of the Sanford formation was carried out in borings located in the plant site area and in the main dam area (see Sections 2.5.1 and 2.5.6). In the plant site area, 10 ft. intervals were tested under pressures up to 110 psi in borings BP62, BP68, and BP70 (Figure 2.5.1-14). Intervals tested ranged from depths of 10 ft. to 145 ft.Several isolated zones registered small water losses under high pressure.The results of the pressure tests, coupled with the soil conditions observed and mapped in the trenches, confirm that the soils and foundation materials and the permeabilities listed in Table 2.4.13-7 are representative of those found at the site. At the plant site area, the few zones that exhibited small water losses during pressure testing were isolated intervals that are located between dense, impervious rock layers which registered no water losses during pressure testing. The impermeable zones ranged in thickness from 10 to 50 ft. above and below each interval that had a water loss.Hydrogeologic information from borings and published data indicate that the small water losses in the above mentioned borings were due primarily to fracture confluence instead of formation texture or permeability changes in the Sanford formation of the Newark Group.2.4.13.2.4 Effects of Groundwater Usage The population in the vicinity of the plant is small and groundwater usage is minimal due to low yields of wells. Most of the land within a two-mile radius, and some beyond this distance, has been acquired by Carolina Power & Light Company. Therefore, the population in the plant vicinity is not likely to increase much and groundwater usage will remain essentially the same.The yield of the Triassic aquifer is low and only a limited supply of groundwater is obtainable from the proximity of diabase dikes. Therefore, any increase in groundwater usage will be limited because of the poor permeability and storage characteristics of the aquifer.Groundwater is being utilized at the site during construction. Table 2.4.13-4 shows the total site groundwater use for the years 1980 through 1982.Site groundwater usage is expected to gradually decrease due to the decline in construction activities.Amendment 65 Page 96 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Figures 2.4.13-1 and 2.4.13-2 compare the pre-construction piezometric level to that existing during the winter of 1979-1980. The groundwater levels, which have been affected considerably, are declining due to pumpage from the site wells. Cones of depression have developed on the northeastern and the southwestern sides of the plant around wells which are being pumped. Directions of groundwater movement have been reversed in the proximity of some wells, as depicted by the cones of depression. The levels are expected to return to near normal as the construction use of water declines.The reservoirs at the Shearon Harris Nuclear Power Plant site comprise a total of approximately 4,417 acres in surface area and contain approximately 77,500 acre-feet of water at the normal pool elevations. The main reservoir operating level is Elevation 220 ft. MSL, and the elevation of the Auxiliary Reservoir is Elevation 252 ft. MSL. Three main cones of depression have groundwater levels lower than Elevation 190 ft. MSL. When the reservoirs are at operating levels, the subsurface flow of water is toward the cones of depression from the two reservoirs and the water levels in the cones of depression are expected to gradually achieve partial recovery. After construction is completed and the groundwater level has recovered, groundwater will move toward the Main Reservoir.Water is supplied to the reservoirs by stream flow, direct precipitation and runoff, and an insignificant quantity of groundwater influent from springs of intercepted permeable zones associated with intrusive rocks where they are in hydraulic contact with the reservoirs.Because of the impervious nature of the soils and country rock, there is only insignificant interchange of water between the reservoirs and the aquifer. This condition is verified as shown in Figure 2.4.13-5. Note that the water levels in piezometers 8A and LP13 are at elevations 102.5 ft. (affected by pumping) and 189.3 ft., respectively, while the water level in the emergency intake canal, approximately 50 feet from both wells, is at elevation 245 ft.In Table 2.4.13-7, the results of permeability determinations from downhole pressure tests show that permeability values for the country rock range from 0.0096 to 0.265 gallons per day per square foot (gpd/ft2) within the plant site. According to the USDA Soil Conservation Service soil survey of Wake County, 1970, the permeability values of the upper 96 inches of soil range from 29.9 gpd/ft.2 to 94.2 gpd/ft.2 in the uppermost 12 inches of sandy loam, from 9.4 to 29.9 gpd/ft.2 in the next 17 inches of clay loam, and 3 gpd/ft.2 in the next 79 inches of clay. The saprolite zones below the surficial clay have much lower permeability values, as mentioned above, and prevent ready movement of water from the surface to the deeper soils.The lack of data points outside the immediate vicinity of the plant island makes it impossible to prepare an accurate map of the piezometric surface in the offsite areas. However, in Figures 2.4.13-6 and 2.4.13-7 pre-construction, current, and post-construction water-level conditions in the plant island area are illustrated. The post-construction water levels are anticipated to closely duplicate the preconstruction conditions except where altered by the plant structure and, to some extent, in the immediate proximity of the reservoir and canals.The permeability values of the soils and saprolite that underlie the reservoir are so low as to require near vertical gradients to drive even a small amount of water from the reservoir bottom to the water table. In areas where there may be a flow of water from the reservoir to the water table, the steep hydraulic gradient will confine the flow path to within approximately 100 feet of the shoreline. Where fracture systems of intrusive dikes may be in hydraulic contact with the reservoir and the head relationships are such as to allow flow from the reservoir into the aquifer, Amendment 65 Page 97 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the gradients will be less than in the country rock, but the flow path will be narrow and confined very closely to the fractured zones in the dikes. According to the observed behavior of water in the fracture system during the pumping test on wells 13 and 15, it is possible that measurable changes in the water level may occur a few hundreds of feet from the reservoir in such fracture systems. The reservoirs will produce no observable effects on the groundwater levels outside the Shearon Harris Nuclear Power Plant site.2.4.13.3 Accident Effects The contents in the tanks located inside the Waste Processing Building or in the Tank Building, including the reactor make-up water storage tank, the refueling water storage tank, and the condensate storage tank are potentially radioactive.2.4.13.3.1 Groundwater Pathway The only possible groundwater path to water users following a radioactive spill would be seepage through the soil. The plant and peripheral lands are underlain by the Triassic, Newark Group aquifer. An accidental release of radionuclides at the site can be assumed conservatively to percolate downward to the aquifer instantaneously. The general direction of groundwater movement in the aquifer at the site is toward the southeast. However, ongoing pumpage at the site for construction water has altered the flow direction locally toward the pumping wells (Figure 2.4.13-2).The value for porosity in the groundwater movement analysis was based on a measured value for permeability for the fracture system of the intrusive-rock dike between wells 13 and 15 (Figure 2.4.13-8). Inasmuch as hard-rock fracture systems are heterogeneous and anisotropic, hydraulic characteristics for these systems can be grouped only in a broad category. In the system between wells 13 and 15, the measured permeability value of 2841 gpd/ft.2 compares with the lower part of the scale of values for gravel as given in Walton, pp. 33-36 (Reference 2.4.13-6). Values were estimated for porosity and "effective porosity" (specific yield) by using the same relative position as "permeability" on scales of these values given in that publication.The range of values for permeability of gravel is given as 1,000 to 15,000 gpd/ft.2.Proportionally, the value of total porosity is estimated at 31 percent and the value of effective porosity (same as specific yield in Walton, 1970) is estimated at 17 percent.Assuming the maximum parameters, it is established that the minimum time required for the groundwater to reach the closest community downstream from the plant would be about 144 years. This time estimate is based upon the following parameters: Corinth is the nearest town, approximately five miles to the southwest, where residents have wells of minimal production from the Triassic, Newark Group (Figure 2.3.2-18). The maximum measured site coefficient of permeability is 520 ft./yr. (Table 2.4.13-7). The maximum measured site hydrologic gradient is 0.06 ft./ft. towards the SE from the Waste Processing Building (Figure 2.4.13-2). The effective porosity is 0.17.The effective travel time of radionuclides which may contaminate the aquifer following a tank rupture would be considerably greater due to absorption and ion exchange on the underlying rock. The distribution coefficients (Kd) for cesium and strontium, the critical radionuclides, are assumed to be 20 and 2, respectively. These values were taken from Table VII 3-7 of Appendix VII of WASH 1400 and are conservative when compared to values reported in the literature Amendment 65 Page 98 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 (Reference 2.4.13-7). The calculated retention factors using these values for Kd, an effective porosity of 0.17 and a bulk dry weight density of 2.6 (Table 2.5.4-1; 162.8 lbs/ft.3) are 307 for cesium and 32 for strontium. Using these retention factors, the travel time for Cs-137 and Sr-90 for transport to the nearest community would be:Cs-137 = (144 yrs) (307) = 4.4 x 104 yrs Sr-90 = (144 yrs) (321) = 4.6 x 103 yrs Assuming tritium to be in the form of water, the effective travel time for tritium would be 144 years. Based upon these effective travel times, radioactive decay would reduce the amount of tritium, Cs-137, and Sr-90 which could potentially reach Corinth to negligible levels.Although the hydraulic gradients at the site vary considerably, the maximum gradient is about 0.06 ft./ft. toward the southeast from the Waste Processing Building (Figure 2.4.13-2). The distance between the Waste Processing Building and nearest site well that has created a cone of depression is about 2000 ft. Any spills could travel toward the well.The permeability of the aquifer material adjacent to the site dikes is significantly higher than values for the country rock; a value of 500 ft./yr. could represent the dike zone or some extensive fracture zone in the country rock.Assuming that the aquifer is hom*ogeneous, with a conservative value for the coefficient of permeability of 500 ft./yr. and a porosity of 30 percent, the water would travel at a rate of about 100 ft./yr. and would cover a distance of 2000 ft. in about 20 years.Table 2.4.13-7 shows that all but two permeability values for the materials at the plant site are less than 10 ft./yr. These two values could represent fractures which may not be extensive.The aquifer material, which has a permeability of 10 ft./yr., a porosity of 30 percent, and a hydraulic gradient of 0.06 ft./ft., would have a water movement rate of 2 ft./yr. and it would take about 1000 years for water to move through 2000 ft.The closest community downstream from the plant site is Corinth (Figure 2.3.2-18),approximately five miles to the south-southwest, where residents have wells of minimal production from the Triassic, Newark Group. Taking the above two groundwater movement rates of 100 ft./yr. and 2 ft./yr. it would take about 54 years and 13,200 yrs., respective]y, for the water to reach Corinth.2.4.13.4 Monitoring of Safeguard Requirements Fourteen piezometers that were installed in 1979, as well as two pre-construction piezometers and one new well, are available at the plant site. The piezometers and site wells provide data on water levels, hydraulic gradient, and direction of flow. Water levels in piezometers and site wells are measured periodically and analyzed to assess the effect of construction on the site groundwater regime. Water samples from three wells were analyzed to determine baseline water quality parameters (Table 2.4.13-8).Table 2.4.12-1 shows the C/MPC at Corinth are approximately one-half of the allowable 10 CFR 20 concentrations. The closest privately-owned well to the reservoir is on N. C. State Road 1128 at approximately 600 feet from the shoreline. The ground surface elevation at this Amendment 65 Page 99 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 well is greater than 30 feet above the normal pool level. The direction of groundwater flow would be from the well to the reservoir, in this case. Inspection of the topographic maps of the area indicates the expected direction of groundwater flow all around the reservoir to be towards the reservoir. Possible exceptions may be in the stream valley immediately and around the dam, and within a few feet of the general shoreline as the gradients adjust to the water levels in the reservoir.The chemical and biological requirements for the plant make-up water are quite stringent and dictate that the high quality of the reservoir water must be maintained. Should any reservoir water seep into the surrounding streams, it would be filtered within the aquifer and would be of better quality than the water in the receiving streams.2.4.13.5 Design Bases for Subsurface Hydrostatic Loading The subsurface portions of Seismic Category I structures on the plant island are designed for hydrostatic loadings with groundwater at Elevation 251 ft. MSL. A permanent dewatering system is not utilized for the Shearon Harris Nuclear Power Plant. Groundwater occurring in widely separated joints in the rock did not significantly affect construction. Any rain or surface water that accumulated during construction was pumped out by sump pumps.The pre-construction piezometric-level map, shown on Figure 2.4.13-1, indicates that the piezometric levels were higher than Elevations 251 ft. MSL under some sections of the plant.However, the lack of significant inflow of groundwater in the completed plant block excavation indicates that groundwater in the rock occurs only in widely separated joints and bedding planes. Except for the west side of the fuel handling building and the adjacent portion of the waste processing building the perimeter of the plant structure up to the top of the foundation mats is in direct contact with rock that is essentially impermeable, and the portion between the plant structures and rock above the top of the mats has been backfilled with residual soil which is of very low permeability (estimated to be less than 10 ft./yr). Additionally, the winter 1979-1980 piezometric-level map (Figure 2.4.13-2) shows that water levels beneath the plant area are well below Elevation 251 ft. MSL.The source of surface water higher than the design basis groundwater level is the Emergency Service Water Intake Channel of the Auxiliary Reservoir, which has an operating level at Elevation 252 ft. MSL, the closest point of which comes to within about 300 ft. of the plant island. The Auxiliary Reservoir will not raise the groundwater elevation beneath the plant island above Elevation 251 ft. MSL for the following reasons:a) The residual soil underlying the reservoir is of very low permeability, as indicated by testing.b) After the Auxiliary and Main Reservoirs are filled, groundwater will move from the reservoirs to the cones of depression created by the pumpage from wells. However, the major portion of the groundwater flow will be from the Auxiliary Reservoir to the Main Reservoir.c) Groundwater from the Auxiliary Reservoir will start moving toward the plant island at an Elevation of about 252 ft. MSL. However, the water level will be at a much lower level than Elevations 251 ft. MSL by the time it reaches the plant island due to the hydraulic Amendment 65 Page 100 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 head loss as it flows through the low permeability materials for a distance of about 300 ft.d) Drainage for the retaining wall and seepage through open joints in the retaining wall further reduce the hydraulic head beneath the plant area.2.4.14 TECHNICAL SPECIFICATION AND EMERGENCY OPERATION REQUIREMENTS As described previously in this section, the Main and Auxiliary Reservoirs have been designed and constructed to withstand without operational restrictions the most adverse hydrological conditions, including flooding, that can occur. Actions that will be taken to assure that reservoir conditions are within the analyzed parameters are as follows:a) The plant will be shut down when the water level in the Main Reservoir falls to the minimum level specified in Technical Specifications. This action will assure adequate pump submergence and cooling capability in the Main Reservoir.b) The plant will be shut down when the water level in the Auxiliary Reservoir falls to the minimum level specified in Technical Specifications. This action will assure adequate pump submergence and cooling capability in the Auxiliary Reservoir.

REFERENCES:

SECTION 2.4 2.4.2-1 Guidelines for Determining Flood Flow Frequency, Bulletin No. 17A of the Hydrology Committee, U.S. Water Resources Council, Washington, D.C., Revised June 1977.2.4.2-2 Letter, K.B. Old, Jr., Wilmington District, Corps of Engineers to C. H. Zee of Ebasco, dated June 2, 1982 and telephone call between K. B. Old, Jr. and D. Hunter of Ebasco, dated June 4, 1982.2.4.2-3 HEC-2, Water Surface Profiles, Hydrologic Engineering Center, U. S. Army Corp of Engineers.2.4.3-1 USNRC Design Basis Floods for Nuclear Power Plants, NRC Regulatory Guide 1.59, Rev 2, August 1977.2.4.3-2 U.S. Weather Bureau, Seasonal Variation of the Probable Maximum Precipitation East of the 105th Meridan for Area from 10 to 1000 Square Miles and Durations of 6, 12, 24, and 48 Hours. Hydrometeorological Report No. 33, April 1966.2.4.3-3 U.S. Army Corps of Engineers, Policies and Procedures Pertaining to Determination of Spillway Capacities and Freeboard Allowances for Dams, Engineer Circular No.1110-2-27, August 1, 1966.2.4.3-4 U.S. Bureau of Reclamation, Design of Small Dams, Second Edition, 1973.2.4.3-5 U. S. Weather Bureau, Probable Maximum Precipitation Susquehanna River Drainage above Harrisburg, Pa., Hydrometeorological Report No. 40, May, 1965.Amendment 65 Page 101 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.3-6 U. S. Army Corps of Engineers, HEC-1, Flood Hydrograph Package, Hydrologic Engineering Center, January, 1973.2.4.3-7 Creager, Justin, and Hinds, "Engineering for Dams" Vol. I, John Wiley & Sons, 1947.2.4.3-8 U. S. Army Coastal Engineering Research Center, Corps of Engineers, Shore Protection Manual, 1977.2.4.3-9 Deleted by Amendment No. 10.2.4.3-10 U.S. Army Corps of Engineers, Wave Runup and Wind Setup on Reservoir Embankments, Engineer Technical Letter No. 1110-2-221, November 19, 1975.2.4.4-1 Stoker, J. J., "Water Waves: The Mathematical Theory with Applications," Intersigns Publishers, Inc., NY, NY, 1957.2.4.5-1 Meteorological Characteristics of the Probably Maximum Hurricane, Atlantic and Gulf Coasts of the United States, Interim Report No. 7-97 U. S. Weather Bureau, May 1968.2.4.5-2 Myers, Holm and McAllister, "Handbook of Ocean and Underwater Engineering",Chapter 12, McGraw-Hill, 1969.2.4.5-3 Wave Runup and Wind Setup on Reservoir Embankments, Engineer Technical Letter No. 1110-2-221, Army Corps of Engineers, November 19, 1976.2.4.5-4 Shore Protection Manual, U. S. Army Coastal Engineering Research Center, Corps of Engineers, 1973.2.4.5-5 Shore Erosion by Storm Waves, Misc. Paper No. 1-59, U. S. Army Corps of Engineers, 1959.2.4.5-6 River Training and Bank Protection. Flood Control Series No. 4, United Nations.2.4.11-1 Patterson, W.D., J.L. Leporati, and M.J. Scarpa, "The Capacity of Cooling Ponds to Dissipate Heat", Proc. of the American Power Conference, Vol. 33, 1977 pp 446-456.2.4.11-2 Environmental Science Service Administration Climatic Atlas, Environmental Data Services, Environmental Service Administration, U.S. Department of Commerce, June 1968.2.4.11-3 Fiering, M.B. and Jackson, B.B. "Synthetic Streamflows," American Geophysical Union, Washington, D.C. 1971.2.4.11-4 HEC-4: Monthly Streamflows Simulation. Hydrologic Engineering Center. U.S.Army Corps of Engineers, Feb. 1971.2.4.11-5 Daily Streamflow Simulation. Hydrologic Engineering Center. U.S. Army Corps of Engineers. 1968.Amendment 65 Page 102 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.4.13-1 Fault Investigation, Shearon Harris Nuclear Power Plant, Units 1, 2,3,4. For Carolina Power & Light Company, Ebasco Services, Inc. New York, February, 1975.2.4.13-2 Geology and Groundwater Resources in Raleigh Area, North Carolina, Groundwater Bulletin No. 15, North Carolina Department of Water Resources, 1968.2.4.13-3 Soil Survey, Wake County, North Carolina, U. S. Soil Conservation Services, U.S.Dept. of Agriculture, in cooperation with North Carolina Agricultural Experiment Station, November, 1970.2.4.13-4 Geology and Groundwater in the Durham Area, North Carolina, Groundwater Bulletin No. 7, North Carolina Department of Water Resources, 1966.2.4.13-5 Estimating the Transmissibility of Aquifers From the Specific Capacity of Wells, U. S.Geological Survey Water Supply Paper No. 1536-I, 1963.2.4.13-6 Walton, W. C. 1970. Groundwater Resource Evaluation New York, McGraw-Hill Book Co., Inc., 664 pp.2.4.13-7 NUREG/CR-0912 Vol., 1981. Geoscience Data Base Handbook for Modeling a Nuclear Waste Repository.2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 2.5.0

SUMMARY

2.5.0.1 Basic Geologic and Seismic Information 2.5.0.1.1 Regional Geology The SHNPP site is located in the Deep River Triassic Basin, a trough-like topographic lowland located mostly within the Piedmont Plateau physiographic province. The upland surface of the Piedmont Plateau is an ancient, deeply weathered erosion surface. It is characterized by gently rolling topography with shallow valleys and rounded divides. Upland elevations range from 300 to 600 ft. above sea level along the eastern border of the plateau to about 1500 ft. above sea level at the foot of the Blue Ridge scarp. Immediately west of the Deep River basin, the upland surface rises to the west and north from between 350 and 375 ft. in northern Lee County to about 600 ft. above sea level in the northwest corner of Chatham County, with the slope of the surface averaging about 10 ft. per mile. Elevations in the Deep River Triassic lowland are generally 50 to 200 ft. lower than those of the Piedmont upland which borders its western and northeastern margins; as a result these borders are marked by abrupt erosional escarpments in many places. Elevations range from less than 160 ft. above sea level along the Cape Fear River to more than 500 ft. in some places in the northern part of the basin. Local relief is less than 100 ft. except near the main streams. The lowland is generally more intricately dissected than the Piedmont upland to the west, and streams are better adjusted to the structure of the bedrock. Most streams have steep valley sides and narrow bottoms, but some streams such as the Deep, Haw, and Cape Fear Rivers, are bordered by terraces which form extensive flat areas.Amendment 65 Page 103 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The igneous and metamorphic rocks which underlie the Piedmont Plateau can be divided into several broad northeast-southwest trending belts on the basis of differences in metamorphic grade. Metamorphic grade is highest in the westernmost belt, the Inner Piedmont, which contains upper amphibolite grade gneisses and schists. Rocks of lower amphibolite grade are present in the Charlotte Belt, which borders the Inner Piedmont on the east, and the Raleigh belt, which lies near the eastern margin of the Piedmont. Metavolcanics and metasediments of mostly greenschist grade characterize the Carolina Slate Belt, which lies between the Charlotte Belt and the Raleigh Belt, and the Eastern Slate Belt, which flanks the Raleigh Belt on the east and forms the easternmost portion of the Piedmont. The Deep River Triassic Basin is a sediment-filled trough located between the Carolina Slate Belt on the west and the Raleigh Belt on the east.The Carolina Slate Belt rocks form a section that is believed to be at least 30,000 ft. thick in North Carolina. This section consists largely of metavolcanic rocks and subordinate metasediaments, of Late Precambrian and Cambrian age, intruded in places by granitic plutons.Volcaniclastic rocks, including siltstone, claystone, and graywackes are interbedded with rhyolitic, dacitic, andesitic, and basaltic volcanic rocks, including tuff, lapillistone, pyroclastic breccia, and lava flows. These rocks are deformed into a series of northeast-trending folds, the major one being the Troy anticlinorium near the center of the southern portion of the belt. Minor folds to the east and northeast of the anticlinorium are either asymmetrical or overturned to the southeast and have axial plane dips that range from vertical to less than 60° northwest. Folds to the west and southwest of the anticlinorium are broad and open and only slightly asymmetrical with axial planes dipping steeply northwest. Some rocks have closely spaced slaty cleavage, hence the name "Slate Belt"; however, many others are rather massive with little or no cleavage. Slaty cleavage is particularly common in the Gold Hill - Silver Hill fault zone, a major fault zone along the southwest border of the Carolina Slate Belt. Radiometric age dating indicates that the volcaniclastic sequence is mainly 650-500 million years old, although some members may be older, and that many of the plutons, ranging from gabbro to granite in composition, are contemporaneous with or slightly younger than the volcanic rocks.Greenschist grade metamorphic granitic plutons on the eastern margin of the belt have been dated at 326 million years (Lilesville pluton) and 285 million years. (Wilton pluton).The Raleigh Belt is essentially a southward-plunging anticlinorium which disappears beneath a cover of Coastal Plain sediments south of Smithfield. The axial surface of the anticlinorium strikes N20° E and dips 70° E. The axial portion of the anticlinorium is occupied by a granitic core, generally referred to as the Rolesville granite, which is about 50 miles long and 10 to 15 miles wide. The granite is flanked by amphibolite grade metamorphic rocks consisting mostly of biotite-muscovite gneiss and schist with some interlayered hornblende gneiss. Near the Rolesville granite the gneisses and schists are injected in places by pegmatites up to 50 ft. or more in width. Small bodies of altered ultramafic rock are common in the northwest part of the belt. Scattered small plutons of granitic rock are present in the northern and western part of the belt; those that have been dated radiometrically have ages in the 325-265 million years range.The metamorphic rocks on both flanks have steep dips in most places and in some places appear to have been isoclinally folded. A northeast trending mylonite zone thought to represent a major fault, the Nutbush Creek Fault, is reported to cut the west flank of the anticlinorium.Apparent offset on the fault is right-lateral.The Deep River Triassic Basin is a structurally complex, northeast trending trough containing a wedge-shaped block of mostly clastic sediments belonging to the Upper Triassic Newark group.Maximum thickness of the sedimentary wedge is believed to be 10,000 ft. or more. Strata within Amendment 65 Page 104 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the basin generally dip as well as thicken toward the southeast. The sediments are mostly of fluvial or lacustrine origin and are characterized by abrupt lateral changes in texture and composition. The oldest rock unit, the Pekin formation consists largely of red or brown, fine-grained clastic rocks with a few beds of conglomerate in the lower part. The middle unit, the Cummock formation, contains two coal beds in its thickest part in the southeastern part of the basin; one bed is generally less than two ft. thick and the other is generally less than four ft.thick. The remainder of the formation consists of gray or black shale, claystone, siltstone, and sandstone. The youngest unit, the Sandford formation, consists mostly of red or brown fine-grained clastics except along the southeast edge of the basin, where it consists of conglomerate and fanglomerate. The rocks of the basin are broken by two systems of normal faults, northeast-trending longitudinal faults and northwest-trending minor cross faults. The faults divide the basin into triangular or diamond-shaped blocks with dimensions as small as one mile by two miles. The Jonesboro Fault zone, which forms the eastern border of the Basin is a diagonal dip slip fault with a total vertical displacement of 5,000 to 10,000 ft. and unknown right lateral displacement. Major longitudinal faults within the basin have vertical displacements of several hundred ft. and minor cross faults have displacements from a few feet to a few hundred feet In many places the Triassic sediments are intruded by diabase dikes of late Triassic or early Jurassic age. The dikes range up to 300 ft. wide and up to seven miles long. Most dikes trend N15° - 40°W, but more northerly trends are common in the northern part of the basin.The geologic history of the central and eastern Piedmont region is poorly known because fossil-bearing strata are extremely rare and geochronology is based largely on radiometric dating of igneous events. The geologic record suggests that island arc volcanism was the dominant activity from Late Precambrian through Cambrian time. A period of major deformation of early volcanogenic deposits around 600 million years ago formed the major folds of the Carolina Slate Belt. This deformation was accompanied or closely followed by emplacement of granitic plutons. The deformational event was followed by renewed volcanic activity possibly indicating development of a new island arc system during Cambrian time. This volcanism continued through late Cambrian time and was followed in early Ordovician time by a metamorphic event which produced greenschist metamorphism in Carolina Slate Belt rocks. Another major deformational event occurred during Devonian time which involved major movement in the Gold Hill fault zone, greenschist metamorphism, and emplacement of granitic plutons in the Charlotte belt. The last clearly indicated major Palezoic event was emplacement of granitic plutons in the Charlotte and Carolina Slate Belts during Pennsylvanian time. The earliest clearly recorded Mesozoic event is the deposition of late Triassic sediments in subsiding northeast-trending troughs in the eastern (Deep River - Wadesboro Basin) and western (Dan River Basin) parts of the Piedmont. In the Deep River Basin normal fault movement along segments of the Jonesboro fault system and the resulting differential subsidence caused eastward tilting of sedimentary strata. Accumulation of the sedimentary wedge was followed by continued movements in the Jonesboro fault zone and development of cross-basin faults. Emplacement of diabase sills and dikes followed formation of the cross faults and continued into Jurassic time.Final movement of the Jonesboro fault during late Triassic-early Jurassic time was followed by widespread zeolite mineralization related either to lowgrade burial metamorphism or to high heat flow and hydrothermal activity. Little is known of late Mesozoic and Tertiary history. The region apparently has been relatively stable tectonically since late Mesozoic time. Crustal movement has largely been limited to vertical isostatic adjustments possibly related to periodic uplift of the Appalachians to the west and subsidence of the Coastal Plain to the east.Amendment 65 Page 105 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.0.1.2 Site Geology The SHNPP site is located near the eastern edge of the Cape Fear River drainage basin. The plant site is on an upland area of gently sloping hills and ridges located between Tom Jack Creek on the west and Thomas Creek on the east. Elevations of hill tops and ridge crests are mostly between 250 and 275 ft. and local relief is generally less than 60 ft. The area is mostly in woodland with scattered small farms. Drainage from the site is southeast through Tom Jack and Thomas Creek to Whiteoak Creek, which flows southwestward into Buckhorn Creek, which in turn flows southward and empties into the Cape Fear River about a quarter mile below Buckhorn Dam.The soils around the site are mostly residual soils derived from sedimentary rocks and diabase dikes underlying the area. Soil depth ranges from 0 to 15 ft., but is commonly between five and 10 ft. The soil is generally thinnest over sandstone and thickest over diabase dikes. Most residual soil is silty clay in texture, but silty sand may be found along streams and in limited areas overlying sandstone. Residual soils observed in trench excavations were medium stiff to hard. Permeability values of most soils are extremely low, resulting in rapid precipitation runoff.The site is underlain by gently dipping rocks of the Upper Triassic Sanford formation. The bedrock is mostly siltstone and fine-grained sandstone interbedded with subordinate shale, claystone, and conglomerate. These rocks consist mostly of alluvial fan, stream channel and floodplain deposits and are characterized by abrupt changes in composition and texture, both horizontally and vertically. Beds range in thickness from less than an inch to a maximum of 20 ft. They interfinger and overlap into compact masses with no structural weakness. Several north to northwest trending diabase dikes of Triassic-Jurassic age have intruded the Triassic bedrock in the site vicinity. These dikes are near vertical and one to 15 ft. thick. Bedrock adjacent to the dikes is commonly baked to a dark gray or black color. Most dikes are deeply weathered to a mixture of clay and rounded cobbles of residual diabase. Layering in the Triassic sedimentary rocks strikes N5°-15°E and dips 9° to 17° to the southeast. Three joint sets are present; the two dominant sets are vertical, one striking N40°-50°E and the other N20°-30°W. A third set trends north-northwest and dips 55° to 70° to the southwest. A minor fault uncovered in the plant excavation trends nearly east-west across the site. The fault is a normal fault with downthrow on the south. The fault surface is somewhat undulatory with dips ranging from vertical to 55° southward. The vertical component of movement is between 80 and 100 ft; the maximum horizontal offset could not be determined, although dikes intruded during the period of movement are offset by an amount ranging from a minimum of one-half foot to a maximum of 13 ft. Drag folding of Triassic beds is present on the hanging wall of the fault. A detailed investigation of this fault, described in Section 2.5.3, determined that the fault is a minor tensional normal fault whose last movement was prior to 150 million year ago.Several small, non-capable faults were found in the foundations of Main Dam structures. These are described in Section 2.5.0.3. No other significant structural features were found.The Triassic strata underlying the site have very low primary permeability and yield little or no groundwater except where secondary permeability is provided by fractures, which are common to depths of 100 ft. Below 100 ft. fractures are less common and tighter and below about 400 ft.are closed and sealed to water flow. Highest groundwater yield is found adjacent to the diabase dikes, which act as impermeable barriers to groundwater flow. Piezometric data from boreholes in the plant area indicate that groundwater movement in the plant area is, in general, to the southeast toward WhiteOak Creek.Amendment 65 Page 106 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Historical records of earthquake activity indicate that the site is virtually aseismic. There is little history of felt earthquakes in the site area and no historical accounts of the behavior of the site during the few earthquakes which have been felt. A seismic monitoring network covering the plant site began operation in 1977; it has recorded no local earthquakes through four years of continuous monitoring. During historical time no earthquake has occurred within 60 miles of the site. No geological evidence of Holocene earthquake activity in the vicinity around the site was found.The geologic history of the site through Paleozoic time is poorly known, since the only Paleozoic rocks exposed in the plant area are Raleigh Belt gneisses and schists exposed in the main dam foundation south of the plant. It is uncertain whether these rocks represent Precambrian continental crust of possible Grenville age upon which Slate Belt volcanogenic deposits were formed or whether they represent Slate Belt rocks which have undergone a higher degree of metamorphism than other Slate Belt rocks. The sedimentary rocks underlying the site were deposited in Late Triassic time contemporaneously with normal faulting along the Jonesboro fault zone. Filling of the Deep River Basin, mainly from eastern sources, continued until rocks now exposed at the site were buried under several thousand feet or more of sediments. A change in stress orientations resulted in the formation of the Shearon Harris site fault and other cross-basin faults. Three separate periods of dike emplacement occurred contemporaneously with the movement of the site fault. A likely sequence of events was (1) sequential intrusion of the two oldest dikes and continuation of fault movement (2) crystallization of zeolites (low-barium laumontite and saponite) in the fault-dike intersections (3) intrusion of the third dike as fault movement continued (4) cessation of movement on the fault, and (5) further crystallization of zeolites harmstome, heulandite, barite, and high-barium laumontite in dike fault intersections.These events probably culminated during Jurassic time. The site was probably buried under Coastal Plain deposits similar to those that overlap the basin to the southwest during Cretaceous time and may have remained buried under Coastal Plain cover for much of Tertiary time. Exhumation of the present day surface may have occurred as late as Pliocene time. The site has been relatively stable tectonically since late Jurassic time, with tectonic activity limited largely to minor vertical movements.2.5.0.2 Vibratory Ground Motion The region in the immediate vicinity of the site is characterized by low-level seismicity.Moderate levels of earthquake activity occur in the surrounding region at distances greater than 130 miles from the site. During the period 1754-1977, eight earthquakes of epicentral intensity VII or greater occurred within about 200 miles of the site. Six were of intensity VII, the closest of which occurred at a distance of about 133 miles from the site. The Charleston, South Carolina earthquake of August 31, 1886, which occurred about 200 miles south of the site, has an intensity of X and was probably felt with an intensity of VI in the site area. The earthquake in Giles County, Virginia on May 31, 1897, which occurred about 160 miles northwest of the site, had an intensity of VIII and was probably felt in the site area with an intensity of about V.The relationship between earthquake activity and known geologic structures or tectonic provinces in the southeastern region of the United States is poorly known because earthquakes are not associated with visible surface faulting and because seismograph station coverage of the region has been inadequate to determine focal depths or focal mechanisms of most historical earthquakes. Three seismic zones have been recognized which border the site region. These are the Central Virginia seismic zone, a relatively narrow zone of activity located in the Piedmont province and oriented obliquely to the NE-SW structural trend; the South Amendment 65 Page 107 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Carolina-Georgia seismic zone, a broad zone spanning both the Piedmont and Coastal Plain Provinces and oriented transversely to the regional structure of the crystalline basem*nt rock; and the Southern Appalachian seismic zone, which includes the Blue Ridge and the Valley and Ridge Provinces from southwestern Virginia to central Alabama and parallels regional structural trends. However, the many faults mapped in each of these zones have no record of surface rupture during historical times.Historical seismicity related to the filling of large reservoirs in the southeastern Piedmont is limited to a few reservoirs in the South Carolina-Georgia seismic zone. These reservoirs range in volume from 0.5 km3 to 2.5 km3 and water depths near seismically active areas range from 25 m to 110 m. All earthquakes reported to date have had local magnitudes less than five.Reservoir-induced seismicity has not been reported from the Piedmont of North Carolina despite the presence of several large reservoirs, some of which cross major faults.Based on historical seismicity, the maximum potential earthquake which might affect the site would be a recurrence of the Charleston, South Carolina earthquake of 1886 which was probably felt as an intensity VI at the site. The largest earthquakes in the site region which are not attributable to any particular geologic structure or seismic zone have been of intensity V.However, it is considered possible that some intensity VII earthquakes in the eastern Piedmont and the Coastal Plain may have been related to exposed or buried Triassic Basins. Therefore, a shock of intensity VII occurring in the Deep River Basin is considered to be the maximum potential earthquake.Seismic wave transmission characteristics of the site were determined from seismic refraction measurements and ambient vibration measurements. Compressional wave velocities range from 1250 to 2000 ft./sec. for residual soils and/or highly weathered bedrock, from 5000 to 7150 ft./sec. for weathered and/or fractured bedrock, and from 10,900 to 13,650 for sedimentary bedrock. Measured shear wave velocities were 2500 ft./sec. for weathered and fractured rock and 5600 ft./sec. for sound bedrock; a velocity of 500 ft./sec was assumed for residual soil based on previous experience under similar conditions. Computed values of Poisson's ratio were 0.44 for residual soil, 0.37 for weathered and fractured bedrock, and 0.35 for sound bedrock. Observed characteristic frequencies of the site from ambient ground motion measurements were 100, 55 1/2, and 25 Hz. The maximum observed level of ground motion, which occurs in the vertical component, was 0.48 x 10-3 in./sec. at about 25 Hz.The safe shutdown earthquake is designated as an intensity VII earthquake occurring close to the site. The resulting maximum horizontal ground acceleration at foundation level within the competent bedrock at the site is estimated to be less than 12 percent of gravity. In order to provide an additional margin of conservatism, a value of 15 percent of gravity is assigned as the maximum horizontal ground acceleration. All safety related structures and systems are designed to assure safe plant shutdown for two horizontal excitations and one vertical excitation simultaneously. Seismic Category I systems and components are designed for a minimum of 10 loading cycles under safe shutdown earthquake conditions. The horizontal and vertical response spectra for the SSE, prepared in accordance with NRC Regulatory Guide 1.60, are presented on Figures 2.5.2-12 and 2.5.2-13.The operating basis earthquake is designated as one with half the accelerations of the safe shutdown earthquake and equivalent to an Intensity VI earthquake near the site. The corresponding horizontal acceleration at foundation level in the bedrock would be less than 7.5 percent of gravity. The horizontal and vertical response spectra for the OBE, prepared in Amendment 65 Page 108 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 accordance with Regulatory Guide 1.60 and scaled to .075g horizontal ground acceleration, are presented on Figures 2.5.2-14 and 2.5.2-15.2.5.0.3 Surface Faulting At the time of preparation of the PSAR for the SHNPP, the only fault known to exist within five miles of the plant site was the Jonesboro Fault, whose trace approximates the course of Buckhorn Creek three miles southeast of the site. Site investigations in the plant and Auxiliary Reservoir area, which included 12,125 ft. of trenching at depths from two to twelve ft. ,numerous geologic borings 100 to 250 ft. deep, and approximately 5000 linear ft. of seismic refraction survey lines, failed to uncover any evidence of surface faulting. However, during excavation activities a minor normal fault, herein referred to as the Site Fault, was exposed in the foundations of the plant Waste Processing Building and approximately 20 small faults were exposed in the foundations of Main Dam structures about 4.5 miles south of the plant site.2.5.0.3.1 Jonesboro Fault The Jonesboro Fault is a northeast-trending diagonal slip fault whose total length exceeds 100 miles. It forms the eastern edge of the Deep River Basin and marks the contact between Triassic sedimentary rocks to the west and Paleozoic volcaniclastic and crystalline rocks to the east. It is nearly vertical with 8,000 to 10,000 ft. of vertical displacement and an unknown amount of right-lateral displacement. In places south of the site it offsets diabase dikes and is overlain by undisturbed, flat-lying Cretaceous sedimentary deposits. The age of last movement on the Jonesboro is bracketed between intrusion of Late Triassic-Jurassic diabase dikes and the deposition of Cretaceous sediments overlying it. Therefore, the Jonesboro Fault is not considered to be a capable fault.2.5.0.3.2 Site Fault A comprehensive investigation of the Site Fault was performed for Carolina Power & Light Company by Ebasco Services Incorporated in order to determine (1) age of last movement on the fault, (2) fault length, (3) vertical and horizontal components of movement on the fault, and (4) alignment and attitude of the fault from the plant excavation to and through the Auxiliary Reservoir area.A program of trenching perpendicular to the trend of the fault, supplemented by further exploratory borings, was carried out to trace the fault beyond the limits of the excavation.Magnetometer surveys were used to locate and trace diabase dikes which served as markers to aid in determining fault displacement. This program traced the fault for a total of 8000 ft. across the plant and Auxiliary Reservoir area and located five diabase dikes trending north northwest across the fault. It was found that the fault, when exposed in sedimentary beds, exhibited a southerly dip between 55 and vertical. Drag folding was present on the hanging wall of the fault in all exposures, but bedding planes of strata on the northern or foot wall were rarely disturbed.Nine core borings were completed in sedimentary rocks on both sides of the fault to determine the vertical component of offset along the fault. The vertical component of movement was determined to be greater than 80 ft. and less than 100 ft. as measured at three locations. The horizontal component of movement on the fault was determined from offsets of diabase dikes, which are essentially vertical. The horizontal offsets range from 0.5 ft., up to a maximum of 13 ft. A large horizontal component of movement is precluded by the undulatory nature of the fault Amendment 65 Page 109 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 surface, which changes strike about every 300 ft. Movement on the fault was primarily tensional with a minor left-lateral component.Remote sensing imagery, including ERTS, SLAR, and Skylab as well as conventional aerial photos, were analyzed in an attempt to determine the total length of the fault. The attempt was unsuccessful because the site fault was not detected by any imagery technique. Several hundred regional and local linear features were identified, but none were identified as capable faults on the basis of imagery evaluation. Nineteen lineaments were considered significant to the site and required field investigation; none that were field checked were identified as capable faults.Field observations of undisturbed soil, saprolite, and, in one place, a post-Triassic sedimentary deposit overlying the fault indicated only that the age of last movement was probably greater than one million years. Therefore, detailed laboratory studies of samples from diabase dikes, associated sedimentary rocks, and undeformed crystals of secondary zeolites found in the fault gouge were undertaken to determine the geochronology of dike emplacement and fault movement. These studies included paleomagnetic age dating of diabase dike samples, potassium-argon age dating of diabase dike and zeolite samples, and strontium isotope studies of diabase, zeolites, and associated sedimentary rock.Potassium-argon age dates on the diabase dikes range from a minimum of 168 million years to 260 million years, while dike ages based on remanent magnetization determinations ranged from an absolute minimum of 150 million years to a maximum of about 225 million years. The intrusion of the dikes at the site was largely contemporaneous in time with movement on the fault, but the youngest of the dikes showed slight displacement by the fault, indicating that some movement took place after dike emplacement. Secondary zeolite minerals were present in fault gouge at dike-fault intersections which showed no evidence of deformation by fault movement.Minimum potassium-argon age dates on the zeolites indicated ages up to 35 million years.These minimum ages were believed to be spuriously low because of argon loss from the zeolites since their formation. A comparison of strontium 87/86 ratios of the zeolites, dike rocks, and associated sedimentary rocks established that the diabase dikes were the source of the chemical ingredients which formed the zeolites. These minerals are formed at temperatures between 100 and 225 degrees centigrade. The two events which could be associated with such temperatures in the Triassic basin are the emplacement of the diabase dikes and a period of regional low-grade burial metamorphism, both of which occurred prior to 150 million years BP.Since the secondary minerals were emplaced prior to 150 million years ago and have not been disturbed by subsequent faulting, last movement on the fault was prior to that time.As a result of these investigations, CP&L concluded in its Fault Investigation Report (Reference 2.5.1-29) that the Site fault was not a capable fault and would not move under reservoir loading or other proposed construction.On January 6, 1976, Carolina Power & Light Company was formally notified by the NRC that the fault discovered at the Shearon Harris Nuclear Power Plant is not a capable fault as defined in Appendix A to 10 CFR Part 100 (see Reference 2.5.3-5 and Supplement 3 to SER dated July 1977). This fault is further discussed in Section 2.5.1.2.3, where it is referred to as a minor high-angle fault. The NRC also requested that seismic monitoring be performed at the site to confirm the NRC staff conclusion that the proposed reservoirs at the site will not cause fault movement during and after filling.Amendment 65 Page 110 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 In response to the NRC's request, CP&L submitted a proposal on February 13, 1976, to establish a seismic monitoring network which would encompass the SHNPP plant site area.Although this proposal called for monitoring to begin in January of 1979, the network was installed and became operational on September 30, 1977, in order to obtain more definitive baseline data prior to the filling of the reservoirs.2.5.0.3.3 Main Dam Faults The minor faults exposed in the foundations of the main dam structures are in Paleozoic crystalline rocks. The faults are all minor normal faults with lengths measured in 10's of feet and displacements measured in inches.Because the small amount of movement along these faults took place prior to deformation-mineralization which occurred more than 225 million years ago, the faults are not considered to be capable faults as defined in Appendix A to 10 CFR Part 100. Reports on these small faults are catalogued in the Foundation Report, Appendix 2.5E. The NRC concurred with this conclusion based on detailed reports which were submitted and on field inspections by their geological staff. The NRC's concurrence was primarily verbal. However, copies of internal NRC memoranda from Mr. Sydney Miner to Mr. Olan D. Parr state concurrence with CP&L's findings that certain faults are considered non-capable. Also, in IE Inspection Report Nos. 50-400/79-07, 50-401/79-07, 50-402/79-06, and 50-403/79-06 concerning inspections conducted by Mr. John R. Harris of the NRC, Region II, certain faults are identified as being considered as non-capable.2.5.0.4 Stability of Subsurface Materials The plant is founded on gently dipping, well-consolidated Triassic siltstones and sandstones.Rock beds range in thickness from a few inches to a maximum of around 20 ft., and are commonly lenticular. Depth of weathering is commonly between five and 10 ft. but tends to be shallower over sandstones and thicker over diabase dikes. Joints are irregularly spaced at intervals of a few feet and are mostly vertical. The dominant joint set is oriented N40° - 50° E, a secondary set is oriented N20° - 30° W, and a tertiary trends north-northwest and dips 55° to 70° to the southwest. An east-west trending, non-capable normal fault with drag folds on its downthrown south side crosses the plant and Auxiliary Reservoir areas. No other folds, faults, shears, or zones of structural weakness were noted in the plant foundations.Programs of subsurface exploration based primarily on trenching and borehole sampling and drilling were conducted for both the preliminary site investigation and the site fault investigation.In addition, the floors and walls of excavations for plant foundations were mapped geologically.No areas of actual or potential surface or subsurface subsidence, uplift, or collapse were found.The static and dynamic engineering properties of subsurface materials were determined by laboratory testing of rock samples obtained from core drilling and by field geophysical measurements. The index properties determined were dry density and rock quality designation (RQD) for selected rock samples and grain size distribution for selected residual soil samples.The static modulus of deformation was computed from unconfined compression tests.Compressional wave and shear wave velocities were determined from seismic refraction measurements in the field. These velocities were used in the computation of Poisson's ratio and the dynamic modulus of deformation.Amendment 65 Page 111 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The plant excavation includes the foundations for the Waste Processing Building, four Turbine Buildings, a Fuel Handling Building, four Reactor Auxiliary Buildings, four Containment Buildings, and four Tank Buildings. It encompasses a total area of approximately 837,500 sq. ft.and has a total volume of approximately 1,200,00 yd.3. Excavation began with the leveling of the ground surface in the plant area to 260 ft. elevation. After leveling, unclassified soil materials were excavated to ripper refusal and excavation was completed to final grade by controlled blasting. Slopes were excavated at 1:1 in overburden and at 1:4 in bedrock. After inspection and approval of final excavation surfaces by a geologist, the foundation surface was cleaned of loose materials and mapped geologically. Dewatering in the excavation was accomplished by intermittent use of sump pumps. Treatment methods used for foundation protection after excavation included slush grouting of joints to control groundwater seepage, placement of drain pipes at locations of seeps to prevent buildup of excessive pressures, shotcreting of some slopes and placement of a seal coat on foundation surfaces. Selected backfill material was compacted between structures and rock surfaces to meet requirements of 95 percent Standard Proctor Density with moisture control at +/- 4 percent of Standard Proctor optimum moisture content and a maximum permeability value of 10 ft./yr.A plot of piezometric levels across the site area show that the piezometric surface, in general, slopes southeastward, indicating that groundwater movement is toward the Whiteoak Creek arm of the Main Reservoir. Groundwater is confined mostly to joints and fractures in bedrock and to the bedrock zones adjacent to diabase dikes, which tend to act as barriers to groundwater movement; residual soils and unfractured bedrock have extremely low permeability and yield little or no groundwater. Downhole water pressure tests at borehole depths ranging from 10 ft.to 145 ft. indicate that permeabilities range from approximately 5 ft./yr. to less than 300 ft./yr. in sedimentary rock units underlying the site area. Twenty-four hour pump tests in several wells drilled near diabase dikes during 1977-78 indicated specific capacities ranging from 0.16 gpm/ft.to 0.59 gpm/ft. Minor seepage of groundwater from joints and fractures occurred in the plant excavation, particularly after rains. This water, as well as surface-water runoff, drained into sumps and was removed by occasional pumping.All Seismic Category I structures within the plant area, except for Seismic Category I Electrical Manholes, Seismic Category I Underground Electrical Conduits, and some Seismic Category I pipes, are founded on sound rock which will not amplify ground motion from earthquakes. The ground acceleration at the foundation levels of structures supported on sound rock is equal to the baserock acceleration. The maximum horizontal accelerations of the bedrock were chosen to be 15 percent of gravity for the safe shutdown earthquake and 7.5 percent of gravity for the operating basis earthquake. The foundation of the plant has no potential for liquefaction because it consists of hard sound rock. The ground acceleration at the level of individual manholes was determined by an amplification analysis of ground motion through a vertical soil column between the bedrock and the manholes. The acceleration values obtained from the analysis were further increased by 50 percent for the equivalent static analysis of each manhole structure. Horizontal acceleration values at the level of individual manholes, as obtained from the analyses, are 0.25 g for the SSE and 0.14 g for the OBE; vertical acceleration values are 0.19 for the SSE and 0.10 for the OBE.Static stability analysis of the plant island structures included settlement, bearing capacity, and lateral earth pressures. The average settlements under static loading computed for the various structures are very small, ranging from 0.008 ft. to 0.035 ft. The differential settlements should be even smaller and, therefore, structurally tolerable. Since the settlements consist of the pseudo-elastic compression of the bedrock, they will occur essentially upon load application.Amendment 65 Page 112 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The computed ultimate bearing capacity is 714 tons per sq. ft., but the design bearing capacity chosen is 25 tons per sq. ft., which provides a factor of safety of 28 compared with the ultimate bearing capacity.2.5.0.5 Stability of Slopes The plant site has no natural slopes whose failure could adversely affect the safety of the nuclear power plant.2.5.0.6 Embankments and Dams Two dams were constructed in the Buckhorn Creek watershed to impound cooling water for the Shearon Harris Nuclear Power Plant. The Main Dam impounds a reservoir used primarily for cooling tower makeup water which has a normal water level elevation of 220 ft. and a water surface area of approximately 4,000 acres. The Main Reservoir also serves as a backup source of emergency service water. The Auxiliary Dam impounds a reservoir for emergency service water which has a minimum pond elevation of 250 ft. and a surface area of 317 acres. An Auxiliary Separating Dike and Auxiliary Reservoir Channel control the flow of discharged emergency service water through the east and west arms of the Auxiliary Reservoir. The dike, constructed across the east arm of the reservoir, prevents discharged emergency service water from flowing directly back to the emergency service water intake area. The Auxiliary Reservoir Channel connects the east and west arms of the Auxiliary Reservoir, allowing emergency service water discharge to enter the west arm of the reservoir for maximum cooling before circulating back to the intake area.The Main Dam, Auxiliary Dam, Auxiliary Separating Dike, Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, and Emergency Service Water and Cooling Tower Makeup Intake Channel are designed and constructed to Seismic Category I criteria and to withstand the effects of natural phenomena. The slope of the dams, dike, and channels are designed to a factor of safety of 1.5 under static conditions, 1.2 for simultaneous OBE and 100-year return period flood level, and 1.1 for simultaneous SSE and 25-year return period flood level. The simultaneous OBE and 100-year return period flood level was not analyzed for the ESW and Cooling Tower Makeup Intake Channel since the simultaneous SSE and 25-year return period flood level analysis was more conservative.2.5.0.6.1 Main dam The Main Dam is located on Buckhorn Creek about 4.5 miles south of the plant site and about 2.5 miles north of the Cape Fear River. It is a rockfill dam with a maximum height of 108 ft. and a length of approximately 1550 ft. at the berm elevation of 260 ft. Its outside slopes are 2.0 horizontal to one vertical. The main dam's spillway, with a crest elevation of 220 ft., is uncontrolled. The spillway crest has a net length of 50 ft. with a pier at its midlength.The Main Dam and Spillway are located approximately 3000 ft. southeast of the Jonesboro Fault and are underlain by pre-Triassic igneous and metamorphic rocks consisting of granite, hornblende-mica gneiss, quartz-feldspar gneiss, and mica schist. The predominant foliation strikes approximately N55°E and dips 30° to 60° northwest. Most joints are spaced two to three ft. apart. The two dominant joint sets are steeply dipping, one striking northeast and the other northwest. In most places the bedrock is covered by residual soils or by alluvium in the valley bottoms. Exploration in the Main Dam area consisted of surface geological reconnaissance and Amendment 65 Page 113 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 several subsurface exploration programs for the purpose of evaluating foundation conditions for the Main Dam and Spillway and for exploration and sampling of possible sources of borrow materials for use as earth and rockfill for the dam. Subsurface investigations included Seismic Refraction Surveys, detailed foundation mapping in excavations for main dam structures, excavation of trenches and test pits for borrow material sampling, and borehole drilling, testing, and sampling.The Main Dam has a core of compacted silty clay and clayey silt material protected on each side by two 8-ft.-thick transitional filter zones and a rockfill shell. The core is founded on suitable rock (defined as rock material that cannot be removed on a production basis with a single-tooth ripper of a D-8 tractor or equivalent) and the rockfill shell is founded on weathered rock (defined as material which cannot be removed on a production basis with the blade of a D-8 tractor-dozer. The materials comprising the core, fine and coarse filters, and rockfill shell met specified design properties as discussed in detail in Section 2.5.6. The downstream face of the dam is protected by a layer of oversized rock and the upstream face by a four ft. thickness of riprap.The stability of the Main Dam was evaluated by the slip circle method and the finite element method. The slip circle method was used to determine the static, pseudo-static, and pseudo-dynamic stability of the dam. The results of these analyses demonstrate that the slopes of the Main Dam will have an adequate factor of safety under all postulated design conditions. The finite element analysis was made to evaluate the seismic stability of the dam. Results indicate that the dam has ample safety margins during the SSE and the OBE.Seepage control at the Main Dam is affected by the use of an impervious clay core founded in a cut-off trench and by installation of a grout curtain in the foundation rock. The cut-off trench was excavated to suitable rock, and the grout curtain, utilizing neat cement, was emplaced to a depth of 50 ft. below the floor of the cut-off trench.2.5.0.6.2 Auxiliary Dam The Auxiliary Dam is located across the Tom Jack Creek Basin arm of the Main Reservoir, adjacent to the southwest boundary of the plant site. It is an earth dam approximately 3903 ft.long with a maximum height of approximately 72 ft. and a berm elevation of 260 ft. Its outside slopes are 2.5 horizontal to one vertical. The dam's spillway is an uncontrolled concrete ogee section with a crest length of 170 ft. and crest elevation of 252 ft. The basis for its hydraulic design is the probable maximum flood (PMF).The auxiliary dam area is in the Deep River Basin and is underlain by clastic sedimentary rocks which, like those underlying the plant site, belong to the lower part of the Triassic Sanford Formation. The bedrock consists of four major lithologic units: medium-to coarse-grained sandstone, fine-to medium-grained sandstone, siltstone, and shaly siltstone. These units grade into one another both laterally and vertically, and all intermediate combinations are present.The strata strike N5°-15°E with dips ranging from 9° to 17° to the southeast. The two dominant joint sets are vertical, one striking N40°-50°E and the other N20°-30°W. A third set trends north northwest and dips 55° to 70° to the southwest. The plant site fault, as detailed in the Shearon Harris Fault Investigation Report (Reference 2.5.1-29), crosses the Auxiliary Dam at Station 4 +23, striking N87°E and dipping 65° to 75° southeast. This fault has been demonstrated to be non-capable, as discussed in Section 2.5.3.Amendment 65 Page 114 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The foundation exploration program for the Auxiliary Dam included test borings, test trenching, and seismic wave velocity measurements. The Shearon Harris fault investigation included borehole drilling and excavation of exploratory trenches in the reservoir area a few hundred feet north of the Auxiliary Dam axis. Further drilling along the dam axis and detailed geologic mapping of the foundation were done during construction. Exploration and sampling for the borrow area testing program included auger borings in the area located between the Auxiliary Dam and the Auxiliary Reservoir Separating Dike and auger boring and test pits in the area which was selected as the borrow area adjacent to the Auxiliary Dam Spillway.The Auxiliary Dam has a compacted core of silty clay and clayey silt material protected by a transition filter zone and a random rockfill shell on each side. The downstream shell is provided with two horizontal drainage blankets, each three ft. thick, which are connected with the transition filter zone adjacent to the core of the dam. In addition, a 200 ft. wide drainage layer is provided under the shell in each of two areas where pre-existing creeks had been located. The foundation of the dam was excavated into weathered rock and the cutoff trench was excavated to suitable rock. The materials comprising the core, filters, and rockfill shell met specified design properties as discussed in detail in Section 2.5.6.The random rockfill surfaces are protected by a layer of riprap in the areas of wave action.Surface areas outside the wave action zones are protected by oversized rock.The static, pseudo-static, and pseudo-dynamic stability of the Auxiliary Dam was determined by the slip circle method. In addition, sliding wedge analyses were made to verify the sliding stability in the abutment areas. The results of these analyses demonstrate that the slopes of the Auxiliary Dam will have an adequate factor of safety under all postulated design conditions. A finite element dynamic analysis made to evaluate the seismic stability of the dam indicates that the dam has ample safety margins during the SSE and the OBE.Seepage control was affected by the use of an impervious core founded in a cut-off trench and by istallation of a grout curtain in the foundation rock. The cut-off trench was excavated to suitable rock and the grout curtain, utilizing neat cement, was emplaced to a depth of 50 ft.below the floor of the cut-off trench.2.5.0.6.3 Auxiliary Reservoir Separating Dike The Auxiliary Reservoir Separating Dike is located across the east arm of the Auxiliary Reservoir about 1700 ft. north of the Auxiliary Dam. It is approximately 1200 ft. long with a maximum height of approximately 55 ft. Its outside slopes are 2.5 horizontal to one vertical.The bedrock underlying the Auxiliary Reservoir Separating Dike consists of clastic sedimentary rocks of the Triassic Sanford Formation and is similar in most respects to the strata underlying the Auxiliary Dam.Foundation conditions were explored by means of borehole drilling and measurement of seismic wave velocities along the centerline of the dike and by geologic mapping of the foundation excavation.The Auxiliary Reservoir Separating Dike has a core of compacted silty clay and clayey silt material protected by a random-rockfill shell which is size graded near the core, with the finer material placed adjacent to the core. The core and rockfill shell are founded either on Amendment 65 Page 115 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 weathered rock or on a thin layer of stiff residual soil overlying weathered rock. The material comprising the core and rockfill shell met specified design properties as discussed in detail in Section 2.5.6. Slope protection is provided by riprap placed on random rockfill. Since water level is the same on both sides of the dike, provisions for seepage control through the dam foundation were unnecessary.The stability of the Auxiliary Reservoir Separating Dike was evaluated by the same methods used for the Main Dam and the results of the stability analyses indicated that the slopes of the Auxiliary Reservoir Separating Dike will have an adequate factor of safety under all postulated design conditions.2.5.0.6.4 Channels The Emergency Service Water Intake and Discharge Channels are conservatively designed to carry the service flow required for normal and emergency shutdown of SHNPP. The intake channel is approximately 3580 ft. long and 50 ft. wide at its invert elevation of 238 ft. The discharge channel is approximately 2170 ft. long. The width at its invert (Elevation 240 ft.) is 50 ft. to 80 ft. The walls of both channels have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock. Portions of the slopes were shaped to grade by backfilling with an impervious material. Diabase dikes were capped with concrete where they crossed channels.The Auxiliary Reservoir Channel is approximately 1570 ft. long and 140 ft. wide at its invert elevation of 235 ft. Its walls have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock. The Auxiliary Reservoir Channel is sized to carry the maximum ultimate discharge of the Service Water System coincident with a PMF flow for the upstream drainage basin.The Emergency Service Water and Cooling Tower Makeup Intake Channel is approximately 2500 ft. long and 45 ft. wide at its invert elevation of 194.0 ft. The walls of the channel have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock on the north side of the channel and two horizontal to one vertical in rock on the south side. The channel was cut entirely through hard residual soil and rock. Therefore, there were no fill sections except local backfill around intake structure and local small structures such as manholes, etc.All four channels are underlain by gently dipping strata of the Triassic Sanford Formation.Bedrock is lithologically and structurally similar to that at the Auxiliary Dam. Foundation conditions were explored by means of borehole drilling, sampling, and testing.The stability of the channels was analyzed by means of the slip circle method. Results demonstrate that the slopes of the channels have an adequate factor of safety under all postulated design conditions.2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION 2.5.1.1 Regional Geology The Shearon Harris Nuclear Power Plant is located in the Deep River Triassic Basin near the eastern edge of the Piedmont Plateau in central North Carolina. Physiographically, North Amendment 65 Page 116 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Carolina is divided into three natural divisions: the Coastal Plain on the east, the Piedmont Plateau in the center, and the Appalachian Mountains on the west, as shown on the Regional Physiographic Map, Figure 2.5.1-1. The Deep River Basin is a northeast-southwest trending topographic trough that lies near the eastern edge of the Piedmont Plateau and is occupied by a complex wedge-shaped downfaulted block of Triassic rocks.As defined in the FSAR, the Deep River Triassic Basin extends southwest from near Oxford in Granville County to west of Carthage in Moore County. It is a structural and topographic trough almost 100 miles long and five to 20 miles wide. It is bounded on the west, north, and east by pre-Triassic metamorphic and igneous rocks of the Piedmont Plateau. On the south and southeast, along the inner edge of the Coastal Plain, there is a cover of post-Triassic (Coastal Plain) deposits that overlap and cover parts of the Triassic and older rocks.The Deep River Basin is divided into three principal parts (Figure 2.5.1-2). The Durham Basin, in which the plant is located, is the northernmost part; the Sanford Basin is the southernmost; the Colon cross structure separates them.a) The Sanford Basin - The Sanford Basin extends southwest from Colon in Lee County to the end of the Deep River Basin in central western Moore County, a total distance of 32 miles. It has a maximum width of 14 miles and includes parts of Lee, Moore, and Chatham Counties. Carthage is near the south-western end and Sanford is near the northeastern end of the basin.b) The Colon Cross Structure - The Colon Cross structure, eight miles long and five miles wide, is entirely in northeastern Lee County. It extends westerly from the northeast edge of the county to the village of Colon, which is four miles northeast of Sanford.c) The Durham Basin - The Durham Basin, about 52 miles long with a maximum width of 20 miles, extends southwesterly from near Oxford in Granville County to the northeastern border of Lee County. It includes parts of Chatham, Wake, Orange, Durham, and Granville Counties. The City of Durham, for which it is named, lies in the north-central part of the basin. The plant, located in the south-central part, is 19 miles south of Durham and 16 miles southwest of Raleigh.2.5.1.1.1 Regional Physiography The Shearon Harris Nuclear Power Plant site is located in the Piedmont Plateau physiographic province. The upland surface of the Piedmont Plateau is an ancient, deeply weathered erosion surface developed on crystalline and volcaniclastic rocks of Precambrian and Paleozoic age.The Piedmont lowlands are developed on Triassic sedimentary rocks occupying structural troughs in the Piedmont Plateau.The Piedmont upland surface is characterized by broadly undulating topography with narrow, shallow valleys and rounded divides. Upland elevations range from 300 to 600 ft. above sea level along the eastern border of the Piedmont to around 1500 ft. at the foot of the Blue Ridge scarp to the west. Immediately west of the Deep River Basin the upland surface rises to the west and north from between 350 and 375 ft. in northern Lee County to about 600 ft. in the northwest corner of Chatham County, with the slope of the surface averaging about 10 ft. per mile. In a few places monadnocks are present which have elevations several hundred feet higher than the plateau surface. The most prominent monadnocks in the eastern part of the Amendment 65 Page 117 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Piedmont are the Uwharrie Mountains in Randolph and Montgomery Counties. The Occoneechee Mountains are a less prominent series of knobs and ridges in Chatham and Orange Counties.The Deep River Triassic lowland surface is generally 50 to 200 ft. lower than that of the Piedmont Plateau upland, which borders the lowland on the west, north, and east, and that of the coastal Plain upland, which borders it on the southeast and south. Consequently, the lowland is bordered by abrupt erosional escarpments in many places. Elevations range from less than 160 ft. near the Cape Fear River to more than 500 ft. in some places in the northern part of the lowland. Local relief is generally less than 100 ft. except near the main streams and along some border escarpments. The lowland surface is more intricately dissected than the Piedmont upland surface, and stream courses in the lowland are better adjusted to the structure of the bedrock. Stream valleys are wider in the lowland than in the upland and narrow flood plains are present along the rivers and their larger tributaries. Most smaller tributaries have steep valley sides and narrow valley bottoms. The most extensive flat areas in the lowland are formed by terraces which border some of the main streams such as the Haw, Deep, and Cape Fear Rivers.The Triassic lowland is bordered on the south and southeast by the Coastal Plain physiographic province, whose inner edge forms a north-facing escarpment with a maximum height of about 250 ft. above the lowland at Carthage. Elevations of the inner margin of the Coastal Plain range from 400 ft. to a maximum of 580 ft. The upland surface on the inner part of the Coastal Plain is characterized by broad interstream divides and flat, wide valleys; surface relief is generally less than that of the Piedmont upland and Triassic lowland surfaces.2.5.1.1.2 Regional Stratigraphy and Lithology The Piedmont Plateau in North Carolina is underlain mostly by crystalline and volcaniclastic rocks of Precambrian and Paleozic age (Figure 2.5.1-3). These rocks can be divided into several broad northeast-southwest trending belts on the basis of differences in metamorphic grade. Metamorphic grade is highest in the westernmost belt, the Inner Piedmont, which contains upper amphibolite grade gneisses and schists. Rocks of lower amphibolite grade are present in the Charlotte Belt, which borders the Inner Piedmont on the east, and in the Raleigh Belt, which lies near the eastern margin of the Piedmont. Metavolcanic rocks and metasediments of mostly greenschist grade characterize the Carolina Slate Belt, which lies between the Charlotte Belt and the Raleigh Belt, and the Eastern Slate Belt, which flanks the Raleigh Belt on the east and forms the easternmost portion of the Piedmont. The Deep River Triassic Basin is a sediment-filled trough located along the eastern margin of the Carolina Slate Belt; it forms the western margin of the Raleigh Belt in places.The southeastern part of the site region is underlain by Atlantic Coastal Plain sediments of Cretaceous and younger age. Cretaceous sediments overlap the southern margins of the Eastern Slate Belt and the Raleigh Belt and the southeastern margin of the Deep River Basin.Scattered small deposits of Eocene age overlie Cretaceous sediments and crystalline rocks of the Raleigh Belt in places.A generalized regional stratigraphic column compiled from several sources is shown in Figure 2.5.1-3a. There are no published regional isopach maps pertinent to the Shearon Harris site.The deposits underlying the plant (Triassic sedimentary rocks) are characterized by many lateral and vertical variations. Thus, data obtained from design and exploratory borings cannot Amendment 65 Page 118 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 be adapted to generate an isopach map because there is no continuous, easily identifiable marker horizon in the Triassic rocks.2.5.1.1.2.1 Precambrian and Paleozoic Rocks Precambrian and Paleozoic rocks in the site region have been grouped into three broad categories. These are (1) metavolcanic rocks of the Carolina and Eastern Slate Belts; (2) gneisses and schists of the Raleigh Belt; and (3) intrusive rocks of mostly granitic composition which are present in the Raleigh Belt and both Slate Belts.2.5.1.1.2.1.1 Metavolcanic Rocks The metavolcanic rocks of the Carolina and Eastern Slate Belts include lithic and crystal tuffs, welded flow tuffs, flows, volcanic breccias, volcanic conglomerates, graywacke conglomerates, argillites, slates, phyllites, and thin limestone beds. (Reference 2.5.1-1) which have intertonguing relationships. Their aggregate thickness is believed to be at least 30,000 ft. in the Carolina Slate Belt. These rocks have been subjected to low-grade, greenschist facies metamorphism and in places have a well-developed slaty cleavage. The simple classification used on the Geologic Map of North Carolina (Reference 2.5.1-2) places the metavolcanics in three major groups: bedded argillites (volcanic slate), felsic volcanics, and mafic volcanics. The areal distribution of these groups in the site region is shown on Figure 2.5.1-3. The following descriptions of these three groups is based largely on References 2.5.1-1 and 2.5.1-2.a) Bedded Argillites - The rocks mapped as bedded argillites include finely bedded meta-shale exhibiting graded bedding, slate, phyllite, and phyllonite as well as argillite.Graywacke and tuff are present in places. This group is represented by slates in the western part of the site region and by phyllites and phyllonites in the eastern part of southern Wake County. The rocks are deeply weathered in most places and outcrops of fresh rock are rare.b) Felsic Volcanics - This group consists largely of volcanic fragmental and flow materials.The fragmental rocks range from rhyolitic to dacitic in composition and consist mostly of coarse and fine tuffs and subordinate breccias. The coarse tuff predominates and contains the breccia and fine tuff as interbedded bands and lenses. In places the finer material grades into bedded argillite. The flows are essentially rhyolite and occur as narrow bands or lenses interbedded with the tuff and breccia. Lenses of bedded slate and mafic volcanics too small to show on Figure 2.5.1-3 are also present within this group. Felsic volcanics weather to a light gray soil commonly underlain by yellow to light-red clay.c) Mafic Volcanics - Mafic volcanics consist largely of flows ranging from andesite to basalt in composition and of fragmental rocks of mostly andesitic composition. The fragmental rocks are mostly tuffs, breccias, conglomerates, and graywackes. Andesite and basalt flows occur as narrow bands and lenses interbedded with the fragmental rocks. The andesite is dark green and coarsely porphyritic in places, but most commonly it is fine grained and massive. The basalt is dark gray to nearly black and commonly amygdaloidal. Most rocks in this group weather to a dark red to maroon clayey residuum.Amendment 65 Page 119 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Radiometric age dating of the metavolcanics indicates that they were formed during Late Precambrian and Cambrian time. R-Sr whole rock and U-Pb zircon ages of felsic metavolcanics from the Carolina Slate Belt range from 620 to 520 m.y. (References 2.5.1-3 through 2.5.1-7).However a few plutons which intrude the metavolcanics have ages older than 620 m.y.(References 2.5.1-6 and 2.5.1-8). Near Chapel Hill some metavolcanics are apparently older than the Chapel Hill pluton, which yields a R-Sr whole rock age around 700 m.y. (Reference 2.5.1-6).2.5.1.1.2.1.2 Gneisses and Schists Metamorphic rocks of amphibolite grade present in the Raleigh Belt include hornblende gneisses, mica gneisses and schists, and felsic gneisses. The hornblende gneisses are mostly medium-to-coarse-grained, massive amphibolites or well-foliated hornblende gneisses. They commonly contain plagioclase (oligoclase) and quartz in various amounts and lesser amounts of biotite and epidote.Mica gneisses and schists are medium to coarse grained and consist predominantly of feldspar, quartz, muscovite, and biotite; biotite is more common in the gneisses and muscovite is more common in the schists. Feldspar is predominantly plagioclase but microcline and orthoclase are common.Felsic gneisses are light-colored, medium-grained rocks characterized by high quartz and microline content and a predominance of muscovite over biotite. Interlayered with these gneisses are garnet-mica schist and two persistent belts of graphite schist.The age of the Raleigh Belt gneisses and schists is poorly known. As discussed further on, they have been intruded in several places by granitic plutons which have Rb-Sr whole rock ages around 300 m.y. (Reference 2.5.1-9), and thus cannot be younger than Pennsylvanian. The fact that the gneisses and schists form part of a large anticlinorium flanked on both sides by Slate Belt rocks suggests that they are as old as, or older than, the metavolcanics. If they represent the "basem*nt" on which the Carolina Slate Belt rocks were deposited, their age may be one billion years old or older (Reference 2.5.1-10).2.5.1.1.2.1.3 Intrusive Rocks The gneisses, schists, and metavolcanics in the site region have been intruded by a wide variety of igneous rocks, of which granitic rocks are by far the most common. These intrusive bodies range in size from small, lens-like bodies to large single plutons and plutonic complexes.Granitic intrusives in the site region include granite, adamellite, granodiorite, quartz diorite, and tonalite. Most of these can be classified into two broad groups, one pre-dating major regional metamorphism and the other post-dating it. Pre-metamorphic granites characterize the western part of the site region and post-metamorphic granites the eastern part. Pre-metamorphic granites which have been dated radiometrically have ages which range from around 705 m.y. to 519 m.y. Black and Fullagar (Reference 2.5.1-6) report Rb-Sr whole rock ages of 705 m.y. for the Chapel Hill pluton and 638 m.y. for the Meadow Flats granodiorite which is a few miles north of Chapel Hill, and Fullagar (Reference 2.5.1-4) obtained a Rb-Sr age of 519 m.y. for the Farrington complex south of Chapel Hill. U-Pb zircon ages of 575 m.y. were obtained for the Roxboro metagranite (References 2.5.1-5 and 2.5.1-11) and 650 m.y. for the Flat River intrusive complex northeast of Durham (Reference 2.5.1-8). The pre-metamorphic nature of intrusives in Amendment 65 Page 120 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 this group is indicated by foliation that is parallel to regional trends and by microtextures that show cataclasis and recrystallization (Reference 2.5.1-12).Post-metamorphic granitic plutons contain massive, even grained rocks that show little evidence of metamorphism. Most rocks in this group are medium to coarse grained and light to medium gray or light to medium pink. Only a few of these plutons in the site region have been dated radiometrically. These include the Lilesville granite in the southwest corner of the site region, the Wilton pluton on the west border of the Raleigh Belt in Granville County, and the Castalia pluton which straddles the Raleigh Belt-Eastern Slate Belt boundary in the northeastern part of the site region. Ages obtained by the Rb-Sr whole rock method were 313 m.y. for the Castalia pluton, 285 m.y. for the Wilton pluton, and 326 m.y. for the Lilesville pluton (Reference 2.5.1-9).The Rolesville batholith, which covers an area of 1710 sq km east of Raleigh, was classified as possible syn-metamorphic by Butler and Ragland (Reference 2.5.1-12). It consists of mostly medium-to-coarse-grained, foliated granitic rocks in its main portion. According to Parker (Reference 2.5.1-13) it was formed in two magmatic stages during the mid-Paleozic; however, its exact age is unknown. Structural mapping by Becker and Farrar (Reference 2.5.1-14) indicates that the batholith post-dates the formation of isoclinal folds in the surrounding country rock and pre-dates a second deformational event which refolded the isoclinal folds and produced muscovite-biotite foliation in the granite. The main part of the batholith is apparently younger than the Precambrian and Cambrian plutons found in the western part of the site region and older than the 313 m.y. old Castalia pluton, which is poorly foliated.Rocks mapped as diorite-gabbro on Figure 2.5.1-3 as a whole are intermediate in composition between true diorite and gabbro. However, bodies of almost normal diorite are present in southeastern Guilford, southern Caswell, and southern Person Counties, and bodies of almost normal gabbro are found in northern Person and northeastern Granville counties (Reference 2.5.1-2). Included in the diorite-gabbro group are some granite-diorite bodies consisting predominantly of diorite. No radiometric ages have been reported for diorite-gabbro intrusions in the site region. Fullagar (Reference 2.5.1-4) suggests that some diorite-gabbro bodies located elsewhere in the southeastern Piedmont are 385 m.y. old or older.The rocks shown as dunite on Figure 2.5.1-3 are actually lenticular bodies of serpentinite, soapstone, and other altered ultramafic rock. Individual lenses range in outcrop length from a few hundred feet to about three miles (Reference 2.5.1-15). All are poorly exposed and are mapped chiefly by the distribution of float of ultramafic composition on the ground surface.2.5.1.1.2.2 Mesozoic Rocks Mesozoic rocks in the site region comprise Triassic sediments deposited in the Deep River Basin, Coastal Plain sediments of Cretaceous age, and diabase dikes and sills of Triassic-Jurassic age.2.5.1.1.2.2.1 Triassic Sedimentary Rocks The Triassic sedimentary rocks of the Deep River Basin have been discussed by Campbell and Kimball (Reference 2.5.1.16) and Reinemund (Reference 2.5.1-17), who divided them into three formations: (1) a lower unit, the Pekin Formation, consisting predominantly of fine-grained clastic sediments, (2) a middle unit, the Cumnock Formation, which is coal-bearing in the Sanford Basin, and (3) an upper unit, the Sanford Formation, which constitutes the bedrock in Amendment 65 Page 121 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the vicinity of the plant site. The following descriptions of these formations are from Reinemund (Reference 2.5.1-17).The Pekin Formation lies unconformably on the metavolcanics and intrusive rocks of the Carolina Slate Belt and crops out in a belt along the northwest side of the Deep River basin.The total thickness of exposed rocks in this formation is 1800 to 4000 ft. in the Sanford Basin, about 3500 ft. in the Colon cross structure, and 3000 ft. at the south end of the Durham Basin.Red, brown, or purple claystone, siltstone, shale, and fine-grained sandstone are the predominant rock types in the Durham and Sanford Basins, and a few beds of conglomerate are present in the basal 50 to 500 ft. of the formation. Conglomerate is abundant in all parts of the formation in the Colon cross structure.The Cumnock Formation, which lies conformably on the Pekin, is present mainly in the Sanford Basin, where it crops out in a narrow belt. This formation is absent in the Colon cross structure; in the Durham Basin it is exposed for a distance of only 3 1/2 miles along strike in the southernmost part of the basin. The maximum thickness of the Cumnock Formation is 750 to 800 ft. in the eastern part of the Sanford Basin, where two beds of coal are present in the lower part of the formation. The lower bed, known as the Gulf coal bed, is generally less than two ft.thick, and the upper bed, the Cumnock coal bed, is generally less than four ft. thick. Below the coal beds, the Cumnock Formation contains about 250 ft. of gray shale, claystone, siltstone and sandstone and overlying the coal beds is about 500 ft. of gray and black shale and claystone.Coal is absent in the western part of the Sanford Basin, where the Cumnock Formation is about 500 ft. thick and consists mostly of siltstone and sandstone. At the southern end of the Durham Basin the Cumnock consists mostly of silty shale, siltstone, and sandstone, and contains only a few inches of coal.The Sanford Formation crops out along the southeastern side of the Deep River Basin. It is more than 3000 ft. thick in the Sanford Basin, 2000 to 3000 ft. thick at the southern end of the Durham Basin, and from 500 to 600 ft. thick at the southwestern end of the Colon cross structure. Throughout most of the Deep River Basin, outside of the Colon cross structure, the formation consists of fine-grained red, brown, or purple clastic sediments. The rocks of this formation coarsen toward the southeast and form a nearly continuous belt of conglomerate and fanglomerate along the southeastern edge of the basin. The formation consists largely of conglomerate and fanglomerate in the Colon cross structure.2.5.1.1.2.2.2 Triassic Jurassic Diabase Diabase intrusions, generally regarded as Late Triassic or Jurassic in age, are common in the Raleigh Belt, Carolina Slate Belt, and the Deep River Basin, although only the larger dikes are shown on Figure 2.5.1-3. These intrusions most commonly have the form of dikes which range from a few inches to several hundred feet in width and from a few feet to several miles in length.Sills and sill-like masses of diabase are present in the Deep River Basin and range up to 400 ft.in thickness. According to Reinemund (Reference 2.5.1.17) diabase intrusives occupy about four percent of the Deep River Basin.The diabase dikes include both olivine-normative and quartz-normative types (Reference 2.5.1-18). Typical diabase is dark gray to greenish black, fine to medium grained, and consists of about 50 percent plagioclase, 25 to 30 percent augite, 10 to 20 percent olivine, and accessory magnetite, ilmenite, apatite, pyrite, titanite, biotite, and amphibole. Many of the dikes in the Deep River Basin and the Carolina Slate Belt show the effects of zeolitization.Amendment 65 Page 122 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Reconnaissance mapping of the dikes in the eastern Piedmont by Burt and others (Reference 2.5.1-19) indicates that two major dike trends are present, one N10°-30°W and the other north-south. A third, less prominent group of dikes, trends east-northeastward.2.5.1.1.2.2.3 Cretaceous Sediments Coastal Plain deposits, generally regarded as Cretaceous in age, overlap the entire southern portion of the site region (Figure 2.5.1-3). The basal unit of the Coastal Plain deposits is shown as the Tuscaloosa Formation on the Geologic Map of North Carolina (Reference 2.5.1-2). The Tuscaloosa Formation can be divided into lower and upper members. The lower member rests unconformably on Triassic rocks of the Deep River Basin and crystalline rocks of the Piedmont.According to Conley (Reference 2.5.1-20) the basal part of this member consists of gray carbonaceous clays containing lignitized plant remains and amber, with interbedded thin, gray and olive sand beds. Above the base the clays are less carbonaceous and lighter gray in color.The upper part consists of light olive clayey sand beds containing thin clay beds. The lower member of the Tuscaloosa was probably deposited in a marine environment (References 2.5.1-20 and 2.5.1-21).The upper member of the Tuscaloosa Formation unconformably overlaps the lower member as well as Piedmont crystalline rocks and Deep River Basin sediments. It consists of a heterogeneous sequence of lenticular clays, muddy sands, clean sands, and pebbly sands (Reference 2.5.1-21). Its base is unconsolidated gravel composed of rounded quartz, ranging from one to six in. in diameter, in a matrix of kaolinitic clay and clayey sand (Reference 2.5.1-20). Most authorities consider this member to be of fluvial origin.The Tuscaloosa Formation is overlain by the Black Creek Formation which, according to Swift and Heron (Reference 2.5.1-21), has an interfingering contact with the Tuscaloosa's upper member. The Black Creek Formation consists of laminated, medium dark gray to dark gray clay interbedded with medium gray to yellow orange sands. Primary sedimentary structures within the formation indicate that it was deposited in a near-shore marine environment.2.5.1.1.2.3 Cenozoic Rocks Cenozoic rocks in the site region include Tertiary marine deposits, high level surficial deposits of possible Tertiary age, and Quaternary terrace gravels and alluvium. The Tertiary marine deposits (Castle Hayne Limestone and the Yorktown Formation) are shown on Figure 2.5.1-3; the other Cenozoic deposits are mostly too small to be shown at the scale of Figure 2.5.1-3.The main outcrop belt of the Eocene Castle Hayne Limestone lies to the southeast of the site region. It lies unconformably on older rocks and at one time evidently covered much of the Coastal Plain and even extended onto the Piedmont Plateau (Reference 2.5.1-2). It is present in the site region as scattered deposits lying on crystalline rocks of the Raleigh Belt in Wake and Johnston Counties and overlying Cretaceous rocks in the southern part of the site region. The formation consists in part of light gray fossiliferous limestone and in part of light-gray marl (Reference 2.5.1-2).The Yorktown Formation of Miocene age overlaps Eastern Slate Belt rocks in places on the eastern margin of the site region. It consists largely of clay, sand, and shell marl in surface exposures (Reference 2.5.1-2).Amendment 65 Page 123 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 High-level, unconsolidated gravels, sands, and clays of unknown age occupy a belt 10 to 40 miles wide along the western border of the Coastal Plain in North Carolina. These deposits lie unconformably in part on older Coastal Plain sediments and in part on crystalline rocks of the easternmost Piedmont (Reference 2.5.1-22). They form discontinuous sheets and patches at elevations of from 270 to about 500 ft. above sea level. These high-level deposits have not been mapped on a regional scale, but Reinemund (Reference 2.5.1-17) mapped "high-level surficial deposits" in the Deep River Basin. However, Reinemund considered all of the Coastal Plain deposits high-level gravel, not recognizing the upper member of the Tuscaloosa Formation which directly underlies the gravel deposits throughout Moore County (Reference 2.5.1-20).Conley (Reference 2.5.1-20) proposed the name Pinehurst Formation for the nonfossiliferous gravel and sand deposits which unconformably overlie the upper member of the Tuscaloosa Formation, and cap all the higher Coastal Plain Hills in central and western Moore County. The formation is normally brown or grayish brown, commonly iron stained, and in places it is cemented with hematite or, less commonly, limonite. Hematite concretions and kaolinitic clay concretions are commonly interspersed throughout the formation (Reference 2.5.1-20).Reinemund (References 2.5.1-17) thought it possible that the high-level surficial deposits he mapped included materials as old as Cretaceous and as young as Pleistocene, while Conley (Reference 2.5.1-20) thought it conceivable that the surficial gravels could be Late Miocene, Pliocene, or early Pleistocene age.Terrace deposits are present along the Deep, Haw, and Cape Fear Rivers and along some of their larger tributaries (Reference 2.5.1-17). These deposits consist predominantly of friable silty or sandy clay enclosing irregular patches or thin lenses of sand and gravel. They are probably of Pleistocene age (Reference 2.5.1-17).Alluvium is present along most streams flowing across the Deep River Basin and the Coastal Plain. It consists predominantly of chocolate-brown and grayish-brown silt with some gray organic clays.2.5.1.1.3 Regional Structure and Tectonic Features The site region is characterized structurally by northeast-trending belts of isoclinally folded and metamorphosed Precambrian and Paleozoic rocks which are partly overlain by two generally wedge-shaped, northeast-trending masses of southeast-dipping, mostly Mesozoic sediments.Triassic rocks of the Deep River Basin constitute the smaller sediment mass and the Cretaceous rocks of the Inner Coastal Plain the larger.2.5.1.1.3.1 Precambrian and Paleozoic Features As discussed in Section 2.5.1.1.2, the Precambrian and Paleozoic rocks of the site region can be divided into three northeast-trending belts on the basis of metamorphic grade. Metamorphic grade is highest in the Raleigh Belt (amphibolite facies) and is lowest in the Slate Belts which flank it (greenschist facies).2.5.1.1.3.1.1 Carolina Slate Belt The metavolcanic rocks of the Carolina Slate Belt have been deformed into a series of northeast-trending folds that range from broad and open to tightly compressed (Reference 2.5.1-23). Two major fold structures have been recognized in Carolina Slate Belt rocks of the Amendment 65 Page 124 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 site region. These are the Troy Anticlinorium in the southern portion and the Virgilina Synclinorium in the northern portion.The Troy Anticlinorium is the dominant feature of the Carolina Slate Belt in Randolph, Moore, and Montgomery Counties. It trends northeast and plunges toward the southwest (Reference 2.5.1-20). It is over 30 miles wide between the Pee Dee River on the west and the northwestern corner of Moore County on the east. A series of normally doubly plunging folds ranging in wavelengths from one to three miles are developed on the southeastern flank of the anticlinorium (Reference 2.5.1-20). These minor folds are either asymmetrical or overturned toward the southeast and have axial plane dips that range from vertical to less than 60 degrees northwest (Reference 2.5.1-1). Minor folds on the west flank of the anticlinorium are broad and open and only slightly asymmetrical with axial planes dipping steeply northwest.The Virgilina Synclinorium, divided by the North Carolina - Virginia border, is the major Carolina Slate Belt structure, and is located in eastern Person County and northwestern Granville County. Its width in North Carolina is about 32 km (Reference 2.5.1-5). The synclinorium trends northeast but its axial trace cannot be located exactly because there is some faulting in the axial region of the structure and the axial plane cleavage is younger than the fold buckling which formed the synclinal structure. Numerous small folds of about one mile wavelength are present on the flanks of the synclinorium; these folds plunge gently northeast or southwest (Reference 2.5.1-5). According to Tobisch and Glover (Reference 2.5.1-24), folding began before the onset of metamorphism and the buckle folds produced by the initial deformation were later modified by metamorphism and a cleavage-forming deformational event which tightened the early folds where it was coaxial with them or refolded them where it was not. Glover and Sinha (Reference 2.5.1-5) have presented evidence that the major syncline was produced by a major compressional event about 575-620 m.y. ago during the Late Precambrian or early Cambrian, and that the Slate Belt regional metamorphism probably occurred between 520 and 300 m.y. ago.No faulting of regional extent has been mapped in Carolina Slate Belt rocks of the site region (excluding the Jonesboro Fault discussed below). Conley (Reference 2.5.1-20) mapped two groups of faults of local extent in Moore County. One group trends northeast parallel to the axes of folds, and the other trends northwest across fold axes. In Person County wrench faulting is responsible for at least a mile of left lateral offset of the fold axis of the Virgilina Syncline (Reference 2.5.1-5).Faulting of regional extent is present outside the site region along the western edge of the Carolina Slate Belt, where the Gold Hill-Silver Hill fault zone marks a sharp break between low-grade Slate Belt rocks on the east and plutonic rocks of the Charlotte Belt on the west (Reference 2.5.1-23). Butler and Fullagar (Reference 2.5.1-25) have established on the basis of radiometric age dating that major movement along the Gold Hill Fault probably occurred during the Devonian period.2.5.1.1.3.1.2 Raleigh Belt The Wake-Warren Anticlinorium is the dominant structural feature of the Raleigh Belt. As defined by Parker (References 2.5.1-15 and 2.5.1-26) it is at least 30 miles wide and about 90 miles long in North Carolina; it continues beyond the state boundary into Virginia. The axial plane of the anticlinorium strikes N20°E and dips 70 degrees southeast. The axis plunges southward at a very steep angle of perhaps 70 degrees (Reference 2.5.1-15). The axial portion Amendment 65 Page 125 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 of the anticlinorium is occupied by a granitic batholith, generally referred to as the Rolesville granite, which is about 50 miles long and 10 to 15 miles wide. The granite core is flanked by metamorphosed rocks which dip steeply in most places and appear to have been closely and, in places, isoclinally folded. A Barrovian-type metamorphic facies series, ranging from lower greenschist to middle almandine-amphibolite facies, is concentric with the batholith (Reference 2.5.1-13).2.5.1.1.3.1.3 Mesozoic Features The major Mesozoic structural feature in the site region is the Deep River Basin, a complex, wedge-shaped block of rocks occupying a northeast-trending, trough-like depression in the older rocks in the Piedmont Plateau. The northwestern edge of this basin is formed in part by the unconformable and geographically irregular contact of Triassic sediments with underlying rocks and in part by northeast-trending longitudinal faults. The southeastern edge is formed by a fault zone, the easternmost fault of which is the Jonesboro Fault (Reference 2.5.1-27), a northeast-trending diagonal slip fault with a total vertical displacement of 5,000 to 10,000 ft. and unknown right-lateral displacement. Within this block, the Triassic sedimentary rocks generally dip about 15° southeast. Local reversals of dip eastward and northward also occur as a result of faulting and intrusion of diabase dikes (Reference 2.5.1-27). Two systems of normal faults, northeast-trending major longitudinal faults and northwest-trending minor cross faults, have broken the rocks of the southern half of the Deep River Triassic Basin into irregular blocks as small as 1 km x 3 km (Reference 2.5.1-27).The Jonesboro Fault and related faults just to the west are among the most significant in North Carolina. They are part of a northwest-dipping fault zone with a vertical displacement of 5,000 to 10,000 ft. For more than 100 miles, the Jonesboro Fault forms the contact between Triassic and older rocks along the southeastern side of the Deep River Basin. The other related faults are known only from geophysical studies. Five other major longitudinal faults, all paralleling the Jonesboro Fault, but not part of the fault zone, are the Deep River, Gulf, Indian Creek, Governors Creek, and Crawley Creek faults. There are also several faults that branch from the Crawley Creek and Governors Creek faults, and a number of minor longitudinal faults. These five, their branches, and other minor longitudinal faults all occur in the northwestern part of the Sanford Basin west of the Cape Fear River. No faults of this type were mapped by Reinemund in the Durham Basin and none were recognized during the siting and design investigations of the SHNPP area. Harrington (Reference 2.5.1-28) mapped minor longitudinal faults along the northwestern edge of the Durham Basin, but these are farther from the SHNPP area than those in the northwestern part of the Sanford Basin.2.5.1.1.4 Regional Seismicity The region in the immediate vicinity of the site is characterized by low-level seismicity. During historical time no earthquake epicenters have been recorded within 40 miles of the plant site.Moderate levels of earthquake activity occur in the surrounding region at distances greater than 130 miles from the site. During the period 1754-1977 eight earthquakes of epicentral intensity VII or greater occurred within about 200 miles of the site. Six were of intensity VII, the closest of which occurred at a distance of about 133 miles from the site. The Charleston, South Carolina earthquake of August 31, 1886, which occurred about 200 miles south of the site, had an intensity of X and was probably felt with an intensity of VI in the site area. The earthquake in Giles County, Virginia on May 31, 1897, which occurred about 160 miles northwest of the site had an MM intensity of VIII and was probably felt in the site area with an intensity of about V.Amendment 65 Page 126 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Further details on regional seismicity are given in Section 2.5.2.2.5.1.1.5 Regional Geologic History The geologic history of the central and eastern Piedmont region is poorly known because fossil-bearing strata are extremely rare and geochronology is based largely on radiometric dating of igneous events. The geologic record suggests that island arc volcanism was the dominant activity from Late Precambrian through Cambrian time. A period of major deformation of early volcanogenic deposits around 600 m.y. ago formed the major folds of the Carolina Slate Belt.This deformation was accompanied or closely followed by emplacement of granitic plutons. The deformational event was followed by renewed volcanic activity possibly indicating development of a new island arc system during Cambrian time. This volcanism continued through late Cambrian time and was followed in early Ordovician time by a metamorphic event which produced greenschist metamorphism in Carolina Slate Belt rocks. Another major deformational event occurred during Devonian time which involved major movement in the Gold Hill fault zone, greenschist metamorphism, and emplacement of granitic plutons in the Charlotte belt. The last clearly indicated major Paleozoic event was emplacement of granitic plutons in the Charlotte and Carolina Slate Belts during Pennsylvanian time. The earliest clearly recorded Mesozoic event is the deposition of late Triassic sediments in subsiding northeast-trending troughs in the eastern (Deep River - Wadesboro Basin) and western (Dan River Basin) parts of the Piedmont.In the Deep River Basin normal fault movement along segments of the Jonesboro fault system and the resulting differential subsidence caused eastward tilting of sedimentary strata.Accumulation of the sedimentary wedge was followed by continued movements in the Jonesboro fault zone and development of cross-basin faults. Emplacement of diabase sills and dikes followed formation of the cross faults and continued into Jurassic time. Final movement of the Jonesboro Fault during Jurassic time was followed by widespread zeolite mineralization related either to low-grade burial metamorphism or to high heat flow and hydrothermal activity.Little is known of late Mesozoic and Tertiary history. The region apparently has been relatively stable tectonically since late Mesozoic time.Crustal movement has largely been limited to vertical isostatic adjustments possibly related to periodic uplift of the Appalachians to the west and subsidence of the Coastal Plain to the east.2.5.1.2 Site Geology The Shearon Harris project site includes the Shearon Harris Nuclear Power Plant and two cooling water reservoirs. The power plant, Auxiliary Reservoir, and most of the Main Reservoir are located in the Deep River Basin and, as shown in the site area geologic map in Figure 2.5.1-4, are underlain by Triassic sedimentary rocks. The Main Reservoir Dam is located approximately 3000 ft. southeast of the Jonesboro Fault and is underlain by pre-Triassic crystalline rocks.2.5.1.2.1 Site Physiography The Triassic sediments of the Deep River Basin are more easily eroded than the igneous and metamorphic rocks of the Piedmont Plateau or the porous sand and gravel deposits of the Coastal Plain. As a result, the Triassic basin is a trough-like topographic lowland for most of its length. It is bounded on the northwest, north, and east by the upland surface of the Piedmont Plateau and on the southeast and south by the upland surface of the Coastal Plain.Amendment 65 Page 127 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The Triassic lowland is an intricately dissected surface with a local relief generally less than 100 ft. Interstream divides are sharp and narrow near the main streams, but away from the main drainage lines they become higher, broader, and flatter. In the southern part of the Durham Basin (the area of interest), the general slope is toward the Cape Fear River, which crosses the Triassic lowland at the south end of the Durham Basin. Most of the streams in the Traissic lowland have a dendritic pattern and are actively deepening their valleys. In general, valley sides are comparatively steep and valley bottoms are narrow, but narrow flood plains are present along some of the larger streams.Elevations in the Triassic lowlands range from less than 160 ft. above sea level along the Cape Fear River to approximately 500 ft. above sea level in the northern part of the Durham Basin. In the general Power Plant and Auxiliary Reservoir area, the lowland is some 200 to 300 ft. above sea level and 50 to 200 ft. below the adjacent surfaces of the Piedmont Plateau and the Coastal Plain.These gently rolling hills and low-gradient streams are a result of a long period of erosion that has stripped away large volumes of relatively soft Triassic rocks, leaving the Deep River Triassic Basin as a muted topographic low. Because of the muted topographic features and the long period of topographic inactivity that produced them, there are no landslides, areas of uplift or subsidence, or other natural features which could be potentially hazardous to the plant.There are also no activities or man-made features in the area which have the potential for affecting site safety.2.5.1.2.2 Site Stratigraphy and Lithology The rocks of the site area are predominantly well-consolidated Triassic sedimentary rocks. A few diabase intrusive rocks are also found in the vicinity of the site (Figure 2.5.1-4). Remnants of the flat-lying poorly consolidated sediments of the Coastal Plain, terrace gravels, and Holocene alluvial deposits occur in the vicinity of the site. The pre-Triassic crystalline rocks which underlie the Main Dam are described in Section 2.5.6.2.5.1.2.2.1 Triassic Sedimentary Rock The Triassic sedimentary rocks of the Deep River Basin are clastic fluvial deposits: claystone, shale, siltstone, sandstone, conglomerate, and fanglomerate. They are characterized by abrupt changes in composition. In some places coarse sediments predominate; in others fine-grained materials are present. Lithologic units are seldom more than a few feet thick. In general, three-fourths of the rocks in the Deep River Basin are red, brown, or purple; the rest are gray or black.The terms "arkosic" or "feldspathic," as used in this report and applied to the Triassic rocks of the southern part of the Durham Basin, refer to any rock that has an appreciable quantity of feldspar (at least enough to be readily recognized in a hand specimen). Many of the sandstones are arkosic but few contain sufficient feldspar (25 percent) to be true arkoses.The sediments of the Deep River Basin are composed largely of debris from nearby pre-Triassic metamorphic and igneous rocks; in places they contain much debris from nearby granite intrusive bodies. These sediments were deposited as alluvial fans, stream-channel and flood plain deposits, and lake and swamp deposits.Amendment 65 Page 128 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The Triassic bedrock underlying the plant site and adjacent areas is covered by overburden that is partly residual, partly transported. On the upland areas, the overburden consists of residual yellow, sandy clay and sandy loamy soil derived from the weathering of underlying rock. This soil is usually only two to six ft. thick, but in areas of erosion, it has been mostly removed by running water. In stream valleys and bottoms, yellow, sandy, clayey alluvium has often accumulated to thicknesses of from two to four ft. to as much as 10 to 12 ft. Beneath this overburden, the Triassic sedimentary rocks are usually dense, compact, and only slightly weathered, with a variably developed, thin saprolite zone.In some areas, notably that near the west abutment and spillway of the Auxiliary Dam, as shown on Figure 2.5.1-4, the Triassic rocks are overlain by younger sedimentary rocks, which were studied intensively only at the place where they overlie and are undisturbed by the site fault (Section 2.5.3). There they were correlated with the undifferentiated high-level deposits described by Reinemund (Reference 2.5.1-17).The sedimentary rocks of the Deep River Triassic Basin have been divided into three units which from oldest to youngest are the Pekin, Cumnock, and Sanford Formations. The Pekin and Sanford Formations are mostly red, brown, or purple siltstone, claystone, shale, sandstone, conglomerate, and fanglomerate. The Cumnock Formation consists of gray and black claystone, shale, siltstone, fine-grained sandstone, and two beds of coal.a) Pekin Formation - The Pekin Formation, oldest of the three Triassic formations in the Deep River Basin, lies unconformably on pre-Triassic metamorphic and igneous rocks and crops out in a narrow belt one to three miles wide along the northwestern side of the basin. In general, the Pekin Formation is not particularly well exposed in the Deep River Basin and knowledge of its lithologic character is based on the examination of many small, scattered outcrops. The best exposures of this formation occur in the Wadesboro Triassic Basin near the town of Pekin, Montgomery County, for which it is named.The total thickness of exposed rocks in the Pekin Formation is from 1,800 to 4,000 ft.At the southern end of the Durham Basin, the thickness is approximately 3,000 ft.Practically all the rocks in this formation are red, brown, or purple. In general they are medium to fine-grained clastic rocks consisting of claystone, shale, siltstone, and sandstone, with a few beds of conglomerate and fanglomerate near the base of the formation.The eastern limits of the Pekin Formation are well to the northwest of the plant site; therefore, rocks of this formation are of no importance at the site.b) Cumnock Formation - The Cumnock Formation lies conformably on the Pekin Formation and crops out in a narrow belt in the southern end of the Durham Basin. It consists of gray and black claystone, shale, siltstone, fine-grained sandstone, and two beds of coal.The formation was named for the Cumnock coal mine where 460 ft. of strata were exposed in the shaft. In the southern part of the Durham Basin the Cumnock Formation is only a few hundred feet in thickness and width of outcrop. The northernmost outcrop mapped by Reinemund (Reference 2.5.1-17) is about one mile southwest of Merry Oaks in Chatham County. The formation lies well to the southwest of the plant site.c) Sanford Formation - The Sanford Formation was named for the town of Sanford, which lies in a belt where this formation is more than 10 miles wide. It lies conformably on the Amendment 65 Page 129 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Cumnock Formation in the Durham and Sanford Basins, but apparently lies unconformably on the Pekin Formation in the Colon cross structure, where the Cumnock Formation is missing. The Sanford Formation is more than 3,000 ft. thick in the Sanford Basin and 2,000 to 3,000 ft. thick in the southern edge of the Durham Basin. It borders the southeastern edge of the Deep River Basin throughout most of its length and reaches a width of several miles in the Sanford Basin and the southern part of the Durham Basin. Rocks of the Sanford Formation underlie almost two-thirds of the southern half of the Deep River Triassic Basin, including the plant site.The Sanford Formation is composed of clastic sediments consisting of claystone, shale, siltstone, sandstone, conglomerate, and fanglomerate, more than three-fourths of which are red, brown, or purple. The remaining are various shades of gray. In the Durham Basin, rocks of the upper part of the Sanford Formation, consisting of conglomerate and fanglomerate, are exposed in a zone one to two miles wide along the eastern and southeastern side of the basin adjacent to the Jonesboro Fault. Between the zone of conglomerates and fanglomerates and the zone of rocks of Pekin age along the western and northwestern side of the basin, fine grained sediments of the lower part of the Sanford Formation, consisting of shale, siltstone, and sandstone, are present. The Shearon Harris Nuclear Power Plant Site is located in this area of fine-grained sediments (Figure 2.5.1-4).The sediments of the Sanford Formation underlying the plant site and much of the southeastern part of the Durham Basin were deposited as alluvial fans and stream channel and flood plain deposits. These materials are characterized by abrupt changes in composition and texture, both horizontally and vertically. They contain few distinctive beds and subdivisions that are consistently mapable. The beds vary in thickness from less than an inch to a maximum of 15 to 20 ft. As a result, exposures only a few feet apart may vary considerably in texture and composition.Notwithstanding these variations in composition and texture, the beds and lenses interfinger and overlap into compact masses that show no structural weakness.Because the rocks dip gently to the southeast, the plant is not founded on any single layer. As shown in the foundation report (Appendix 2.5E), layers of fine to medium sandstone are the most common foundation rock.2.5.1.2.2.2 Triassic-Jurassic Intrusive Rocks Triassic sedimentary rocks in the Deep River Basin have been intruded by Triassic-Jurassic dikes, sills, and sill-like masses. The dikes are from a fraction of an inch to more than 300 ft.wide and from a few ft. to more than seven miles long. Sills and sill-like intrusives vary from a few inches to more than 200 ft. thick. These intrusives, commonly classed as diabase, occupy about four percent of the total area of the Deep River Trissic Basin.Sills and sill-like intrusives are almost completely confined to the Cumnock Formation in the Deep River Coal Field. They are most abundant between the towns of Gulf and Haw Branch, where one-third to one-half of the Cumnock Formation is occupied by sills and sill-like intrusives as much as 400 ft. thick. The only exposed sill of any appreciable size known to occur in the Sanford Formation crops out near Euphonia Church in western Lee County. No sills or sill-like masses are known to occur in the southern part of the Durham Basin.Amendment 65 Page 130 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Diabase dikes are abundant in the Sanford Basin and the Colon cross structure. They generally follow northwest-trending joints and cross faults along which displacements are less than 50 ft.Dikes in the Sanford Basin trend N25 to 40W, whereas those in the Colon cross structure and the southern end of the Durham Basin trend N15 to 20W or N-S.2.5.1.2.2.3 Post-Triassic-Jurassic Sedimentary Deposits South of the plant site, the Triassic sedimentary rocks, the younger intrusive igneous rocks, and the trans-basin faults are overlain by undisturbed, flat-lying weakly consolidated Cretaceous marine sediments (References 2.5.1-20 and 2.5.1-21). These could, however, be remains of the Jurassic burial event described in Section 2.5.3 and therefore could be pre-Cretaceous.Probable up-dip, non-marine remnants of these deposits are found nearer the plant site, as are probable later Tertiary deposits (which record sea-level fluctuations) and Quaternary terrace gravels. None of these deposits were formed by tectonic movements.2.5.1.2.3 Site Structural Geology The only major structural feature present in the site vicinity is the Jonesboro Fault, whose trace crosses the lower end of the Main Reservoir less than a mile north of the Main Dam, as shown on Figure 2.5.1-4. This fault is covered in places by unbroken Cretaceous sediments southwest of the Main Reservoir Dam and is considered to be inactive. A minor high-angle fault was discovered in the foundation of the plant during excavation. This fault was subjected to an intensive investigation (Reference 2.5.1-29) which led to the conclusion, with which the NRC concurred, that the fault is not capable (see Reference 2.5.3-5). The fault and the investigation of its capability are described in Section 2.5.3, where it is referred to as the site fault. Other minor faults, all judged to be non-capable, were mapped in the foundation of the Main Dam in the pre-Triassic crystalline rocks. Most of these faults are only a few tens of feet long with only several inches of displacement. Written reports on these features as presented to the NRC are included in Appendix 2.5.E.Folding in the site area is most common in gneisses and schists exposed in the foundation of the Main Reservoir Dam. These rocks appear to have undergone several periods of folding with isoclinal folding predominating. The magnitude of this folding could not be determined because of the limited area of exposure of the crystalline rocks. The only folding observed in Triassic sedimentary rocks were drag folds resulting from movement along the site fault.Joints are common in both the pre-Triassic crystalline rocks and the Triassic sedimentary rocks.The dominant joint set in the crystalline rocks at the Main Dam strikes approximately N60°-70°E and dips 50° to 70° to the southeast. Another set strikes N20°-35°W and dips 70° to 90° southwest. Joints in the Triassic rocks of the power plant and Auxiliary Dam areas are most prominent in the sandstones and siltstones. Three joint sets are present. The two dominant sets are approximately vertical, one striking N40°-50°E and the other N20°-30°W. A third set strikes north-northwest and dips 55° to 70° southwest. Most of the joints are tight and do not extend vertically for more than a few feet.2.5.1.2.3.1 Geophysical Studies of Structural Features a) Gravity Studies - To define geological structures in the basin, Mann and Zablocki (Reference 2.5.1-30) made gravity measurements over the Deep River Wadesboro Basin to establish (1) location of concealed structures within the eastern Triassic Basin, (2) location Amendment 65 Page 131 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 of the northwestern and southeastern borders of the basin where they are covered by a post-Triassic overlap, and (3) determination of the southeastern extent of the Sanford Basin.They took measurements with a Worden Gravimeter at 1,200 stations spaced about one mile apart. These measurements were relative to the gravity base station at Chapel Hill established by Woollard and Mann in 1956.Assuming a specific gravity of 2.67, Mann and Zablocki (Reference 2.5.1-30) computed Bouguer anomalies, shown on Figure 2.5.1-5, with an isomilligal interval of 5 mgals. They also made a residual anomaly map with a contour interval of 5 mgals in an attempt to separate the deep-seated features from surface and near-surface geologic effects. The Triassic area is shown in dashed lines on Figure 2.5.1-5. The surface and near-surface geologic effects and the probable basem*nt and intrabasem*nt variations in lithology mask to a degree the limits of the Triassic basin.Two gravity highs border the basin, one on the northwestern side of the Durham Basin, the other on the southeastern side of the Wadesboro Basin. Relative gravity lows, indicated by the 5 and 0 mgal isogal lines, exist in the Wadesboro, Sanford, and Durham Basins. The only negative gravity areas are located in the southeastern part of the Durham Basin and east of the Sanford Basin; these measure approximately -5 mgals. The gravity low and negative anomaly areas were chiefly attributed by Mann and Zablocki (Reference 2.5.1-30) to especially great thicknesses of Triassic rocks. The isogals cross the Jonesboro Fault at right angles in the southern part of the Durham Basin, but north and southwest of this area they cross obliquely. The isogals on the residual map also cross the Jonesboro Fault obliquely for most of its trace. Mann and Zablocki (Reference 2.5.1-30) believed that the primary reasons this great fault is masked are variations in the basem*nt and near-surface lithology surrounding the basin.The Colon cross structure, the constriction between the Durham and Sanford Basins, is a northwest-trending anticlinal warp. As a result of this feature, the isogals surrounding this area have a saddle-like appearance. Mann and Zablocki (Reference 2.5.1-30) correlated the longitudinal faults, cross faults, and diabase dikes mapped by Reinemund (Reference 2.5.1-17) with finger-like deviations of the isogals in the Sanford Basin. On that basis, they postulated that similar projections in the Wadesboro and Durham Basins stem from the same type of structural features.The negative-anomaly area along the eastern border of the Durham Basin south of Raleigh is probably associated with the large granite body in that area, the gravitational effects of which extend into that part of the basin indicated by the -5 mgal isogal. With the exception of this area, the deeper parts of the eastern Triassic basin are enclosed by 5, 10, and 15-mgal isogals. Mann and Zablocki (Reference 2.5.1-30) correlated the negative anomaly in the center of the Sanford Basin with the network of longitudinal faults and cross faults indicated on the geologic map (Figure 2.5.1-3). These faults may have caused the basem*nt in this area to subside more than in other parts of the Deep River Wadesboro Basin.Figure 2.5.1-6 is a map of the Deep River-Wadesboro Basin showing the location of traverse lines across the basin. The profile from Pittsboro to Raleigh through Cary, and the profile from Siler City to Sanford were selected for further examination, because the power plant site lies in the area bounded by these profiles.Amendment 65 Page 132 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The interpretation of gravity anomalies is complicated by the unknown configuration and physical properties of underlying geological structures. A useful approach to interpretation can be made by application of "direct methods" described by Tanner (Reference 2.5.1-31),which obtain the configuration of a body directly from the observed anomaly, provided that the general shape and physical properties of the body are specified . Direct methods have been successfully used to interpret gravity anomalies caused by sedimentary basins and granite bodies (References 2.5.1-31 and 2.5.1-32).Anthony Watts, of Lamont-Doherty Geological Observatory, Columbia University, used a computer program he developed to compute the basem*nt configuration, assuming a certain density contrast. Figure 2.5.1-7 shows the residual anomaly values for the traverse from Pittsboro through Cary to Raleigh. It also shows the computed values for the basem*nt configuration illustrated in that figure, for a density contrast of -.015 gm/cm3. The computed values agree well with observed data. The basem*nt rocks in the western half of the basin appear to lie closer to the surface than the rocks in the (graben-like) eastern half, which are probably 5.2 km thick. The maximum measured anomaly difference, about 12 milligals, is between the basin sediments and the adjacent bedrock. The great displacement of the Jonesboro Fault and related faults is dramatically emphasized in this profile.The residual anomaly values along the profile through the northern part of the Sanford Basin from Siler City through Sanford are shown on Figure 2.5.1-8. This basin appears wider near the crystalline basem*nt than does the Durham Basin. The basem*nt floor is closest to the surface near the northwestern side. The maximum value of the residual gravity anomaly is about 8 mgals and model computations indicate the maximum thickness to be about 2.2 km for a density contrast of -0.15 gm/cm3. The geologic model that can explain the residual anomaly is indicated at the bottom of Figure 2.5.1-8.Conclusions The isogals of the Bouguer and residual anomaly gravity maps cannot be correlated satisfactorily with either the outline of the basin or the entire trace of the Jonesboro Fault.The fault discovered at the power plant is a minor feature, not reflected in these records. It is clear from an examination of the Bouguer map that the deepest parts of the eastern Triassic basin are located in the northwestern and southeastern part of the Sanford Basin and the southern part of the Wadesboro Basin.Although the geologic features are not exactly defined on the gravity maps, the gravity profiles clearly show the basem*nt configuration of the Deep River Wadesboro Basin.Model calculations suggest a deep accumulation of Triassic sediments and also show the Jonesboro Fault as a steeply dipping feature on the southeastern side of the basin. To locate dikes and/or faults in the area, it is necessary to have station spacing of about one-tenth of a mile or less; therefore, the gravity measurements of Mann and Zablocki cannot be used to locate these features. Recontouring and computer modeling of available gravity data have not revealed any information unfavorable to the plant site. Figures 2.5.1-7 and 2.5.1-8 show schematic cross-section models of a Triassic basin which fit the gravity data.b) Aeromagnetic Studies - To further define geologic structures in the area, previously compiled aeromagnetic studies were examined. The examination revealed northwest-trending lineaments near the site (Figure 2.5.1-9), representing diabase dike swarms in this otherwise magnetically quiet area. The apparent low-order lineament aligned east-west Amendment 65 Page 133 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 near the site is probably not a true magnetic lineament, but a normal error involving the east-west flight alignment of the aeromagnetic survey. The en echelon magnetic signature of the diabase dike swarm passing immediately through the site is discussed further in Appendix D of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).2.5.1.2.3.2 Remote Sensing Imagery Positive imagery acquired for the fault investigation by Side-Looking Airborne Radar has been examined exhaustively. Conventional high- and low-altitude aerial photography, Skylab imagery and ERTS composites have been examined in great detail. Location, length and alignment of several hundred linear features were identified. None were identified as capable faults on the basis of imagery evaluation, nor were any of these linear features that were field checked identified as faults.Ground-truth assessments in the field indicated no evidence of recent earth movements in the site area. The site fault was not detected by any imagery techniques. Appendix C of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29) contains a full description of the remote sensing imagery evaluation.2.5.1.2.3.3 Stresses Prouty (References 2.5.1-33) noted that the absence of folding of sediments in the Durham Basin seemed to confirm the absence of strong compressional forces since the time the sediments were laid down, an observation consistent with data and interpretations of the Deep River Basin over the intervening years.The relationship between residual stress and contemporary tectonic stress must be considered.The levels of stresses locked into some rocks by past, no longer active stress patterns are probably greater than stress levels under present tectonic conditions. For example, in some underground works it is commonly found that the crystalline rocks are under sufficiently high stress to induce explosive fragmentation of rocks, known commonly as pop-outs and rock bursts, around underground openings.In contrast, throughout the Deep River Coal Field, students of mining activities, including Dr. J. L. Stuckey, report lack of apparent stress in underground openings, some as deep as 800 ft. below the surface. Dr. Stuckey recalls that during visits to the mines in the early 1920s, Howard N. Butler, the manager of Carolina Mine, felt that supporting sets in workings were often unnecessary, although installed by custom. The histories of these mines do not indicate incidence of wall or crown failure. The present basic stress condition in the rocks can be described as an "at rest" condition.Sbar and Sykes (Reference 2.5.1-34) presented evidence which suggests that the locations of earthquakes in eastern North America are controlled by unhealed faults or fault zones in the presence of high stress. The orientation of the faults with respect to the stress field may be one factor that determines on which faults strain release occurs. If this hypothesis is assumed valid, then by mapping both stress and faults unhealed by metamorphism within plates, it may be possible to assess earthquake risk at specific locations.Metamorphism, which includes crystallization in fault zones, is associated with deep burial and high temperature and is often accompanied by hydrothermal activity. If a region like the site Amendment 65 Page 134 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 area has been quiet and stable and not undergone great burial or great uplift since the Mesozoic, the possibility of seismicity is extremely low. Consequently, an unhealed fault in such a region of a subdued, stable surface indicates the region has been quiet and free of great earth movements. In an area of unhealed faults, it is necessary to determine individually the time of last movement of any particular fault, since the lack of healing by metamorphism may point directly to the crustal stability of that region. Other signs of deformation, such as oversteepened slopes, landslides, high and sharp terrain, and offset stream courses are more indicative of cause for concern in areas of unhealed faulting. Harrington (Reference 2.5.1-35), in discussing his analysis of the west border of the Durham Triassic Basin, is quoted as follows:

 "Field evidence indicates that there have been no ascending mineralizing waters in post-sedimentation time. This means that in the granites and slates, except for the manganese oxide precipitates and colloidal clays which have been carried in by meteoric waters, the fault cracks are still open. The slickensides developed during faulting are often preserved as casts of manganese or clay."

The fault investigated at the plant site is an unhealed fault along which movement has not occurred since the Late Jurassic, more than 150 million years B.P. The only metamorphic event that has affected the rocks at the site was a low-grade burial event that occurred prior to that time, as indicated by chemical remanent magnetization superimposed on the original magnetization of the diabase dikes.A sensitive indicator of the relative stress state, though not the orientation, may be the condition of the ground water in the rock mass. The Triassic basin sediments are themselves highly impermeable rocks, known, however, to contain relatively small quantities of water trapped between diabase dikes that act as vertical aquicludes and compartmentalize the basins. In the Dunbarton Basin of South Carolina-Georgia, however, the occurrence of overpressurized ground water has been reported (Reference 2.5.1-36). This is in a region of some seismic activity.No overpressurized conditions have been found in the Durham Basin. Bain (Reference 2.5.1-

37) did not note overpressured groundwater conditions in the Deep River Basin. Prouty (Reference 2.5.1-33) described a deep well (1,640 ft.) drilled in Triassic sediments in Durham and failed to note any pressure condition. In some instances, artesian groundwater has been an indicator of excessive stresses in a rock mass. Bain and Thomas listed 349 wells in Triassic rocks in 1966 (Reference 2.5.1-38) and reported that a few flowing wells are present in Durham and Chatham counties and that these are normal artesian wells. Shallow wells, down to 300 ft.,

in the Triassic rocks average only 0.08 gpm per foot of uncased hole and show a marked reduction in yield below 100 ft. May and Thomas (Reference 2.5.1-39) reported on 84 wells in Triassic rocks in the Raleigh area, the average depth of which is 153 ft., which produce an average of 0.04 gpm per foot of well. Reinemund (Reference 2.5.1-17) described groundwater in the Deep River Basin and found no artesian wells. He described the logs of 26 deep borings of which he personally logged 5 (deepest 2,354 ft.), but did not report water conditions. Artesian flow from these borings did not occur (Reinemund, verbal communication, 1975). The absence of overpressure groundwater conditions in the Durham Basin is further evidence of a lack of contemporary high levels of stress.A search for geologic evidences of Holocene earthquake activity in the region surrounding the plant site was made but none were found. Geologic evidences suggesting the occurrence of earthquakes as found in areas of known seismic activity include over-steepened slopes; large Amendment 65 Page 135 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 landslides; areas of trapped surface water, with poorly developed drainage; stream offsets; and other topographic irregularities. None of these features have been found in the site area. A number of investigators have suggested that clastic dikes found in the Coastal Plain might be associated with earthquake activity. Heron, Judd, and Johnson (Reference 2.5.1-40) have concluded that clastic dikes in the Carolina Coastal Plain were formed by filling of fractures that probably developed in weathered rock as a result of slump or hillside creep. They found no evidence to indicate that any of the clastic-filled fractures in the Carolina Coastal Plain were formed by tectonic activity.Some investigators feel that the Cape Fear Arch has been alternately a positive and negative feature since the Cretaceous, long after the sediments of the Triassic basin were deposited.Ferenczi (Reference 2.4.1-41) reviewed the literature up to that time on the Cape Fear Arch and other features of the Coastal Plain. This review indicates a general agreement that the Cape Fear Arch has been alternately a structural high and structural low through geologic time since the earliest Jurassic deposits in the Coastal Plain. Swift and Heron (Reference 2.5.1-21) also report some changes with geologic time.Certainly the Cretaceous and post-Cretaceous rocks of the Atlantic Coastal Plain are incapable of storing any significant level of stress beyond those normal body stresses associated with lithostatic pressure without soft ground rupture, similar to that which occurs in the Gulf of Mexico Coastal Plain, where the ruptures are not associated with seismicity.Brown, Miller and Swain (Reference 2.5.1-42) concluded that the region of the plant site is in a phase of crustal deformation in which the east-west-trending fault at the site would be under secondary compressive horizontal force. This alignment of present stresses is favorable to the stability of the site fault.An assessment of the geologic and structural history of the development of the Triassic Basin and subsequent events is outlined in the following sequence:a) The Jonesboro Fault and other major faults of the Deep River Triassic Basin are probably reactivated older structural trends in the basem*nt rocks. Reactivation or initiation of tensional, normal-type movement, together with a possible lateral component on the Jonesboro Fault, was followed by the deposition of Triassic sediments, which continued to be deposited during progressive movement on the Jonesboro and other longitudinal and cross faults. The episode of Triassic deposition accumulated sediments which were probably several thousand feet thicker than those remaining today and which spread over a much broader area. This load overconsolidated the deeper sediments and resulted in the lithification of the remaining rocks, with resulting high specific gravity, above normal seismic velocities, and low permeability.b) During this time span, shallow rooted movement on the site fault took place as a complement to deeper rooted continuing movement on the Jonesboro and probably other trans-basin faults. As the fault movements continued in Triassic-Jurrassic time, diabase dikes were intruded. Some of the dikes contain abundant amygdules at present surface levels, indicating that the dikes were intruded when the ancestral surface was less than about 1,000 ft. above the present surface. This indicates that a major period of erosion followed deposition of the original thicknesses of Triassic sediments but occurred before intrusion of the diabase dikes. At this time, movement on the Jonesboro Fault remained tensional, with a right-lateral slip component. Movement on the site fault Amendment 65 Page 136 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 was primarily tensional, with a minor left-lateral component. These movements produced a clockwise sense of rotation between the two faults.This episode of movement on the site fault and other faults, the folding of the Triassic rocks to form the Colon cross structure and other more minor folds, and intrusion took place during late Triassic-Jurassic time. The zeolite laumontite was crystallized in the fault gouge from the diabase.Emplacement of a pluton of granitic rock a few miles east of the site may have occurred late in the history of movement on the Jonesboro Fault. The pluton is expressed on the gravity map of the Triassic basin as a negative anomaly. On the aeromagnetic map, it is expressed as an irregular positive anomaly and appears to encroach into the basin. In side looking radar, the pluton is reflected as a gray tonal change and appears circular in east-west imagery. Field reconnaissance confirms the presence of the large granite body, but its relationship to the Jonesboro Fault is not defined.c) After intrusion of the dikes, major movement continued on the Jonesboro Fault, south of the site. This movement may have ended more quickly north of the site, since a dike, probably of this period, appears to cross the Jonesboro Fault with very little off-set.Alternately, the dike was emplaced during last stages of movement on the fault.Negligible to small last movement took place on the site fault after intrusion of the youngest dikes, which are Jurassic in age.d) During later Jurassic time, the surface of the Triassic sediments, which was only slightly higher than the present surface, was buried under a load of sediments which may have been as thick as 9,000 ft. or as thin as 2,000 ft. This burial is suggested by the regional low-grade metamorphic event determined from chemical remanent magnetization of the dikes prior to 150 m.y. BP, and the crystallization of higher temperature secondary minerals, including zeolites, in the gouge of the site fault at that time. The differences in possible depth of burial noted above reflect lack of precise knowledge of the heat flow regime in the rocks at that time. If a normal continental heat versus depth relationship is postulated to provide the temperatures needed for crystallization of the secondary zeolite minerals, then the depth of burial was great. However, if the continental crust was thin or absent at the time of burial, higher temperatures would have been achieved under much shallower burial.The depth of Jurassic burial was relatively great or spread over much of the Piedmont, since subsequent erosion of this material and quantities of underlying Piedmont and rejuvenated Appalachian rocks have furnished about 600 million cubic miles of land-derived rock material to the Coastal Plain, more than half of it Cretaceous in age. Whatever the depth of burial at the site during Jurassic time, the site fault was not moved subsequently, since the secondary minerals that crystallized in the fault gouge are not cataclastically strained or broken.e) Following the Jurassic metamorphic event, major transportation of sediments across the Triassic basin took place and resulted in deposition of great quantities of late Jurassic-Cretaceous marine sediments in the Coastal Plain, extending inshore to points west of the basin. Since the Jonesboro and other faults do not off-set these materials, and to Amendment 65 Page 137 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the north appear to offset only slightly a dike of Jurassic age, last movement on the Jonesboro Fault and the site fault took place no later than the late Jurassic.f) Since the late Jurassic period, the site area has been remarkably stable. Less than 1,000 ft. of Triassic rocks have been eroded, and they have not been further folded or deformed. No faults younger than Miocene are found in the site region. It is evident that groundwater levels at the dikes have never been lower than at present because secondary minerals in the diabase dikes and fault gouge could only have been preserved below groundwater level. Streams are currently downcutting in Triassic rocks, indicating that they have not been relatively lower than at present. Levels of aggradation of sediments associated with higher stands of sea level such as during the Cenozoic are preserved in the site region. Examples include the exposure of the site fault underlying sediments at Trench FET-19W (see Section 2.5.3) (though this deposit could be a late Jurassic remnant) and Quaternary terrace deposits found along the Cape Fear River.g) Diabase dike intrusion at the site was contemporaneous with movement on the site fault, which continued over a relatively long span of time before and during dike intrusion. The diabase dikes range in age from an absolute minimum of 150 million years to a maximum of about 225 million years on the basis of remanent magnetization studies and from a minimum of 168 million years to a maximum of 260 million years based on potassium-argon dating.2.5.1.2.5 Site Engineering Geology A comprehensive series of subsurface investigations was conducted to evaluate the engineering, geologic, and seismologic characteristics of the plant site and the sites of various reservoir-related structures and to sample soils for possible use as borrow materials. Figure 2.5.1-10 is an index map for FSAR figures which show the locations of boreholes, exploration trenches, test pits, and geophysical surveys in the power plant, Auxiliary Dam, and Main Dam areas. Subsurface investigations in the power plant vicinity (Figures 2.5.1-11, 2.5.1-12, 2.5.1-13, and 2.5.1-14) are discussed below. Those in the Auxiliary Dam and Main Dam areas are discussed in Section 2.5.6. Numerous boreholes drilled for these subsurface investigations are located outside the map areas outlined in Figure 2.5.1-10. These boreholes were not critical to Seismic Category I structures and are not discussed in the FSAR; however, their locations are included in the tabulations of borehole locations presented in Appendices 2.5.A and 2.5.C.Figure 2.5.1-11 shows the locations of boreholes (Series D and Series P boreholes) and exploration trenches excavated for a preliminary subsurface investigation of bedrock composition, orientation, and quality across the sites of the power plant and the Auxiliary Dam.Soils encountered in the boreholes were described in accordance with the Unified Soils Classification System, and RQD (Rock Quality Designation) values were obtained by calculating the ratio of core four in. or more in length to the length of the full core run. Twelve thousand ft.of trenches were excavated during these investigations to supplement the information obtained from borings. Layout of two of the trenches was planned in relation to regional geology. Trench No. 1 was oriented perpendicular to the regional strike of lithologic units and Trench No. 2 was oriented perpendicular to the trend of regional cross faults. Logs of these bore holes and trenches are included in Appendix 2.5A.The locations of subsurface investigations conducted for design of the power plant and adjacent structures' foundations are shown in Figures 2.5.1-12, 2.5.1-13, and 2.5.1-14. Boreholes in the Amendment 65 Page 138 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BP series (Figures 2.5.1-12 and 2.5.1-14) are design borings for the power plant; logs of these borings and results which are of downhole water pressure tests that were conducted on selected borings in this series are included in Appendix 2.5A. Laboratory test results on representative samples from the boreholes are included in Section 2.5.4 and Appendix 2.5B.The BC series of boreholes (Figures 2.5.1-12 and 2.5.1-13) was drilled to investigate foundations for the emergency service water channels and the cooling tower make-up water intake channel; borehole logs are included in Appendix 2.5A and laboratory test results are in Section 2.5.4 and Appendix 2.5B. Boreholes in the BD series (Figures 2.5.1-12 and 2.5.1-13) and the BX series (Figure 2.5.1-12) were drilled to investigate foundations for proposed dikes (BD boreholes) and a proposed auxiliary dam (BX boreholes) which were included in preliminary project plans. Logs of these borings are on file with CP&L but are not included in the FSAR because the structures for which they were drilled were eliminated from the final project plans. The BCT series boreholes (Figure 2.5.1-12) were drilled in the foundations of the plant cooling tower; the logs of these boreholes are not included in the FSAR because the cooling tower is not classified as Seismic Category I structures. The borehole logs are, however, on file with CP&L. The BB series of boreholes (Figure 2.5.1-12) consists of auger borings drilled to sample soils in a proposed borrow area. The logs of these auger borings and the logs of the TPY series of borrow area test pits are included in Appendix 2.5C. The seismic refraction and shear wave velocity surveys whose locations are shown in Figure 2.5.1-12 are discussed in Section 2.5.2.Additional subsurface exploration consisting of 21 borings and 4000 ft. of trenches was completed in the power plant and Auxiliary Reservoir Dam area in 1974 during the Shearon Harris fault investigation, which is described in Section 2.5.3. Locations of these boreholes and trenches are shown in Figures 2.5.1-15 and 2.5.1-16. Borehole logs and trench wall sections are presented in Section 2.5.3.In order to document geologic conditions during construction the excavations for the foundations of the power plant, the Main Dam, and the Auxiliary Dam were mapped at a scale of 1 in. equals 10 ft. The Auxiliary Separating Dike was mapped at a scale of 1 in. equals 50 ft. These maps are included in Appendix 2.5E.None of the subsurface investigations revealed any zones of alteration, irregular weathering profiles, structural weakness, unrelieved residual stresses in bedrock (Section 2.5.1.2.3.3), or activities of man in the area which might adversely affect the site. The siltstones and some of the fine sandstones at the plant and Auxiliary Dam slake over a period of days on repeated wetting and drying. To prevent this, water was not used in final excavation clean-up. Rocks were blown clean using air lances, and kept covered during periods of inclement weather until concrete could be placed. Concrete was placed only on clean, fresh, unaltered rock.As described in Section 2.5.2, there is little history of felt earthquakes in the site area. There are no historical accounts of the behavior of the site during the few earthquakes which have been felt. There is, however, no evidence of adverse behavior and no reason to expect any.2.5.1.2.6 Site Groundwater Conditions Site groundwater conditions are described in Section 2.4.13.Amendment 65 Page 139 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2 VIBRATORY GROUND MOTION A review of seismic history of the region and an examination of evidences for Quaternary faulting or other forms of ground deformation provide information necessary for establishing the seismic design basis for vibratory ground motion. Earthquake activity is correlated with geologic structures or tectonic provinces (when earthquakes may not be reasonably correlated with specific structures), and earthquakes significant for determining SSE (safe shutdown earthquake) are identified. The design acceleration at the site is estimated by using an attenuation relationship appropriate to the site region. Response spectra for SSE and OBE (operating basis earthquake), for horizontal and vertical ground motion, are developed by using Regulatory Guide 1.60.2.5.2.1 Seismicity Table 2.5.2-1 is a catalog of earthquakes that occurred within about 200 miles of the site through 1981. Figure 2.5.2-1 (Reference 2.5.2-23) is an epicenter map of the southern Appalachian region for the period 1754-1971. Figure 2.5.2-1a shows an epicenter map of earthquakes, from the earliest time through 1981, occurring within 50 miles from the site. A large scale map showing earthquakes within 200 miles from the site, for the period 1698-1981, along with seismic zones, after Bollinger (Reference 2.5.2-23), is presented in Figure 2.5.2-17.The region in the immediate vicinity of the site is characterized by a low level of seismicity.However, a moderate level of earthquake activity has occurred in the surrounding region at distances greater than about 130 miles from the site.Beginning in 1977, bulletins of the Southeastern U.S. Seismic Network describing the "Seismicity of the Southeastern United States" were issued. In these bulletins the hypocenters of a large number of small magnitude earthquakes were published, thus indicating a significant improvement in detection and location capability in the region. However, the level of activity remained rather low. During the period January 1978 through December 1981, only 9 earthquakes of magnitude 3.0 or above occurred within a radius of 200 miles from the site.Only one of these (October 8, 1979 magnitude 3.6) was of magnitude larger than 3.5. All of these earthquakes occurred at distances greater than 140 miles from the site. Also during this period only three earthquakes were reported to have occurred within a radius of 50 miles from the site. The earthquakes of February 25, 1978 (magnitude 2.2), March 4, 1981 (magnitude 2.8), and October 3, 1981 (magnitude 1.1) occurred at distances of 39, 42, and 31 miles, respectively, from the site.During the period reported (1698-1981), eight earthquakes of epicentral intensity VII and above occurred within about 200 miles of the site. Six were of intensity VII, the closest occurring about 133 miles from the site. The 1897 Giles County, Virginia earthquake, MM intensity VIII, occurred approximately 160 miles northwest of the site; the 1886 Charleston earthquake, MM intensity X, about 200 miles south of the site.An earthquake series of considerable significance occurred near New Madrid, Missouri on December 16, 1811, January 23, 1812 and February 7, 1812, each with MM intensity XI - XII, at an approximate distance of 570 miles from the site.Effects of some of these major earthquakes are discussed below:Amendment 65 Page 140 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 a) Charleston, South Carolina Earthquakes of August 31, 1886 These were the strongest shocks in the southeastern United States in historical times.There were two main shocks, at 9:51 p.m. and 9:59 p.m. EST on August 31, 1886. Two epicentral tracks were identified, one near Summerville, 16 miles northwest of Charleston.The approximate epicentral coordinates are 32.9° North, 80.0° West, about 200 miles south of the site. The shocks were preceded by explosion-like sounds near Summerville on August 27th and 28th. The main shocks were followed by aftershocks, some of rather high intensity, into the next day. The duration of the first shock was about 35 to 40 seconds.The shocks were felt in an area of about 2,000,000 square miles. The area within a distance of about 100 miles of Charleston was strongly shaken. The most serious reports came from the major population center of Charleston, where people were terrified and damage was extensive. About 60 people were killed. The shocks were accompanied by roaring sounds.The degree of damage to structures could generally be correlated with the type of design and construction as well as with local geologic conditions. Much of Charleston is constructed on "made land", including filled-in creeks. Structures in these areas were severely damaged. Heavy masonry and brick buildings, especially those constructed using poor-quality mortar, were often severely damaged, while well-built wooden houses (which would tend to respond elastically to earthquake motion) were generally much less damaged.Ground waves were reported. Near the epicentral points, cracks, craterlets and sand boils were noted and railroad rails were bent. People were thrown to the ground; many chimneys fell.The shock was felt as far away as Boston, Milwaukee, New York, Cuba, and Bermuda. At Savannah, about 90 miles from the epicenter, about 300 chimneys were damaged. At Augusta, about 100 miles away, about 100 chimneys fell and a dam fissured and broke. At Raleigh, about 215 miles from the epicenter, the shock was reported strongly felt, with instances of cracked walls and fallen chimneys. Raleigh lies within the Rossi-Forel intensity VI Zone. It is probable that the shocks were felt in the site area with intensity of about VI.Occasional earthquake activity in the Charleston area has continued to the present time.b) Union County, South Carolina, Earthquake of January 1, 1913 The shock occurred at 1:28 p.m. EST on January 1, 1913. The epicenter was about 34.7° North, and 81.7° West, about 175 miles southwest of the site. The maximum intensity of the shock was VII-VIII (VIII on Rossi-Forel scale, Reference 2.5.2-2). It was felt in an elliptical area, 45 miles by 25 miles, trending north-northeast south-southwest. Wave-like motions of the ground were reported in several places. The shock was accompanied by noises like thunder. It was probably felt in the site area with an intensity on the order of I-III. The total area affected was about 43,000 square miles, with the shock felt in North Carolina, eastern Tennessee, and southeastern Georgia.On June 26, 1945, an intensity VI shock occurred in this same area near Murray Lake, South Carolina. Chimneys were reported cracked near the epicenter. Rumbling noises also preceded this shock.Amendment 65 Page 141 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 c) Giles County, Virginia, Earthquake of May 31, 1897 The shock of maximum intensity VIII occurred at 1:58 p.m. EST on May 31, 1897. The epicenter was about 37.3° North, 80.7° West, about 160 miles northwest of the site; the total affected area was about 280,000 square miles. The shock was felt from Georgia on the south to Pennsylvania on the north and from the Atlantic Coast to Indiana and Kentucky, but most strongly at Pearisburg, Giles County, Virginia.Old brick houses and chimneys were cracked, and bricks were shaken from chimney tops.Fissures appeared in the ground, and small landslides were noted.At Narrows, Virginia, on the New River, near the West Virginia border, it was claimed that a motion like the ground swell of the ocean was observed. Large rocks rolled down the mountains. The shock was accompanied by loud sounds. At Raleigh, North Carolina, two shocks were felt, and a few chimneys were damaged. The shock was preceded by four shocks between May 3rd and May 31st, and was followed by aftershocks until June 6. This shock was probably felt in the site area with an intensity of about V.There have been numerous additional shocks in the Giles County area, the most recent in 1968 (intensity IV, near Narrows, Virginia).d) Shocks of December 22, 1875 The main shock of maximum intensity VII occurred at 11:45 p.m. EST on December 22, 1875. The epicenter was probably about 37.6° North, 78.5° West (near Richmond, Virginia),about 133 miles northeast of the site. The total affected area was about 50,000 square miles. This shock, preceded by a minor shock on March 10, was felt over a relatively large elliptical area extending from Baltimore, Maryland, southwest to Greensboro, North Carolina, and from the Atlantic Coast westward to Greenbrier County, West Virginia. Near the epicenter, five shocks occurred in quick succession. Bricks were shaken from chimneys in Goochland and Powhatan Counties, and shingles were shaken from a roof at Manakin, Virginia. A chimney collapsed in Wilmington, North Carolina. At Richmond, Virginia, the shock lasted 20 to 30 seconds, and deep rumbling was noted. There were no reports of the shock having been felt in the site vicinity. Numerous other small shocks have occurred in the Richmond-Charlottesville-Arvonia region; the largest possibly had intensities approaching that of the 1875 shock. Such shocks occurred in 1774, 1833, 1852, 1885, and 1907. Minor activity has occurred as recently as 1966 (intensity V near Richmond).2.5.2.1.1 Initiation of Seismicity Associated with Reservoir Impoundments Through 1978, 64 instances of reservoir induced seismicity (RIS) have been reported worldwide. Although most reservoirs located in aseismic terrain fail to induce seismicity, some earthquakes have occurred near reservoirs in regions that were previously considered aseismic.Because a fault at the SHNPP underlies one of the reservoir sites, the causes and potential for reservoir-related seismicity have been reviewed.Earthquakes were first related to reservoirs when local earthquakes were felt shortly after Lake Mead began to fill in the mid-1930's. They culminated in a magnitude five shock about a year after the reservoir had filled to 80 percent of its capacity. For a number of years thereafter, small local earthquakes showed close correlation in numbers and energy release with seasonal Amendment 65 Page 142 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 peak loads. After some time, correlation was as close with unloading as with loading. However, sometimes there was no direct correlation (Reference 2.5.2-3).A. G. Galanopolos (Reference 2.5.2-4) reported a correlation of small to moderate earthquakes felt in Attica, Greece, over the period 1931 to 1965 with peak loadings of Marathon Lake. The lake is relatively small, with a capacity of 4 x 107m3. Earthquake epicenters were as far as 20 km from the lake, yet most occurred when the lake was overflowing. Galanopolos (Reference 2.5.2-4) has also shown definite correlation of numerical peaks of microearthquakes with seasonal load peaks in the case of Vaiont Dam and Reservoir in Italy.In Australia, a moderate amount of seismic activity has been reported in association with Sydney Water Board Projects. Most of the earthquakes occurred at some distance from the reservoir sites; however, the filling of Lake Eucumbene was associated with moderate seismicity.The Kremasta Dam and Reservoir in Greece are in a seismically active area. Filling of the lake began on July 21, 1965, and the first tremors were felt in December of that year. An exceptionally long sequence of small local shocks started after damming of the Acheloos River and impoundment of the artificial lake. The number of foreshocks could be correlated with the increase of reservoir loading. The main shock occurred soon after water level had reached the maximum height of 120m. Subsequently, many aftershocks have been recorded and felt (References 2.5.2-5 and 2.5.2-6).The most noteworthy examples of reservoir associated seismicity in aseismic areas are the earthquakes associated with Kariba Dam Project in Africa and the Koyna Dam Project in India.Koyna Reservoir is on the Deccan Lava near the crest of a regional monoclinal structure. Hot springs and other evidences of activity are known along the axis of the structure.Many have considered the area aseismic and placed it in zone zero of the seismic zoning map of India (Reference 2.5.2-7). The filling of the reservoir was accompanied by numerous microearthquakes and by mid-September, 1967, some one hundred epicenters had been located within the reservoir area.On September 13, 1967, two magnitude 5.0 to 5.5 earthquakes caused minor damage; on December 11, 1967, a magnitude 6.5 earthquake caused extensive damage. The epicenter of the December 11 shock was within a few kilometers of the dam. Within the following 14 days, six aftershocks, magnitudes 5.5 to 6.2, occurred in the general area. Focal depths varied downward to 20km (References 2.5.2-8 and 2.5.2-9).On the Zambezi River, the Kariba Dam, a concrete arch 120m high, which created the largest reservoir in the world, developed a classical case of reservoir induced seismicity. The storage reaches 16,000 x 107m3. The dam is founded on closely folded and faulted Precambrian rocks.The reservoir is over 250km long, covering mostly Paleozoic sediments and lavas of a tectonic rift valley. Numerous faults border the Paleozoic formations, but the region was seismically quiescent before the dam was constructed and was considered aseismic.Impounding of Kariba Reservoir started in 1955, and the dam was completed in 1959.Significant shocks were first felt in 1961, and seismic activity steadfastly progressed with the rise of water level which crested in 1963 (Reference 2.5.2-10). The peak seismic activity followed almost immediately, with nine felt shocks, magnitudes 5.1 to 6.1, from August through Amendment 65 Page 143 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 November, 1963. Epicenters were localized near the deepest parts of the storage. Correlation between increased seismic activity and rising water level in the reservoir can be graphically demonstrated.The sequence of seismic events near the Kariba Reservoir is undoubtedly such that significant activity began following impounding of the reservoir, and then slowly decayed over a number of years. Such behavior and a foreshocks-peak-aftershocks pattern is reported to be characteristic of reservoir induced seismicity (Reference 2.5.2-9).Another documented case of reservoir-related earthquakes has recently been reported by Shen et. al. (Reference 2.5.2-11). The Hsin Feng Kiang Dam in Kwang Tung province, China was begun in July, 1958. The reservoir covers 390 sq. km. and impounds 1,150 x 107m3 of water.The dam is 440 m long and has a maximum height of 105 m. Before construction began, the immediate vicinity of the site exhibited the typical scattered seismicity of China.Within a month after first impounding water in October, 1959, seismic activity started; as water level rose, so did frequency of shocks. The first earthquakes were in the region of the dam itself, but progressively the activity spread to other regions. There were relatively few events under the reservoir; the great majority were within a kilometer of the water's edge. Each new rise in water level seemed to stimulate fresh activity; the depths of earthquakes were typically 4 or 5 km. Finally, in March, 1962, a magnitude 6 event occurred within 1 km. of the dam, at a depth of 5 km. The epicentral intensity was VIII - a violent shock.During the last 20 days before the main shock, there was marked reduction in seismic activity everywhere in the reservoir region; and the steadily occurring small events moved toward the epicenter of the main shock. Since March, 1962, aftershocks have continued, with characteristics conventionally expected of an aftershock sequence. Up to 1972, 258,247 shocks with magnitude equal to or greater than 0.2 were recorded.Another occurrence of reservoir-related seismicity is that at Keban Dam and Reservoir in eastern Turkey. The reservoir filled after closure of the diversion tunnels in the fall of 1973.Maximum reservoir volume is 3,000 x 107m3, at a height of 212 m. The dam and reservoir are in a tectonically active zone between the Anatolian Fault to the north and the eastward-trending northern extension of the African Rift Zone to the south. In the 10 years preceding impoundment, eleven events (the largest of magnitude of 5.6) with epicenters in the area, were recorded at more than five international stations. A microseismic array has been in operation around the reservoir since September, 1973. The background level of microseismic events, prior to impounding, was 10-15 events per month. As the reservoir approached 100 m depth in the spring of 1974 and leveled off, the incidence of microseismic events rose to 80-90 per month, with a peak number of events lagging behind maximum pool level by about two months.The number of events declined under a stable 100 m pool depth in 1974 to the background level of 10-15 events per month. This is perceived as a normal foreshock sequence, with more events expected during flood rises in pool level. To date, microseismic epicenters generally coincide with known active or recently active faults in the reservoir area.Talwani (Reference 2.5.2-12) has documented four cases of temporal and spatial seismicity modification at reservoirs in the Piedmont of South Carolina. Three of these cases (Lakes Jocassee, Keowee, and Monticello) are instances of RIS, while the fourth case (Clark Hill) is questionable.Amendment 65 Page 144 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The relationship of reservoir construction to local earthquake situations is observational and statistical. The existence of a seismic effect from reservoir impounding is best documented by cases such as Koyna, Kremasta, Kariba, Leak Mead, Camarillas, Talbingo, Nure, Hsin Feng Kiang, Keban, and a number of others. Earthquake swarms in the areas of these reservoirs were previously unknown. Since reservoir impoundment, observed seismic activity has been centered in, and only in, the immediate vicinity of the reservoirs. It seems evident that water impoundment triggered local earthquakes in these cases.On the other hand, many reservoirs, some approximating Lake Kariba in volume, have been impounded with little or no association with local earthquakes.Packer et. al. (Reference 2.5.2-13) constructed a multivariate probabilistic model for the conditional probability of RIS at a reservoir characterized by its depth, volume, stress regime, and geologic setting. Of these four variables, depth and volume showed the strongest correlation with RIS.Various mechanisms to explain RIS have been put forward. They include increased stress due to water load (Reference 2.5.2-10), decreased effective stress due to increased pore pressure (Reference 2.5.2-14), stress corrosion in silicate rocks (Reference 2.5.2-15), and argillization of the materials of the weak structural plane and the reduction of its shearing strength (Reference 2.5.2-11). The loading and pore pressure mechanisms seem to be the most reasonable. In any case, the amount of shear stress generated by even the largest reservoirs is two orders of magnitude too small to cause the fracturing of intact rock and an order of magnitude too small to cause movement along pre-existing fracture planes. Therefore, the stress change brought about by reservoir impoundment can only act as a triggering mechanism for release of pre-existing stress.The impoundment of every reservoir modifies the stress regime in the region immediately surrounding the reservoir. Whether or not these stress changes cause RIS depends upon many factors. The primary factors are the pre-existing state of stress surrounding the reservoir, the magnitude of the induced stress, and the geologic and hydrologic conditions of the site.2.5.2.1.2 Seismicity and the Plant Reservoirs The reservoirs at the SHNPP site are small compared to reservoirs associated with seismic activity. Table 2.5.2-2 lists a few reservoirs associated with seismic activity, compared with the depth and volume of the Main and Auxiliary Reservoirs for the SHNPP. While there is no definite critical water depth for causing RIS, Packer's work (Reference 2.5.2-13) suggests that the likelihood of RIS being associated with a water depth of 19 m is extremely remote.Talwani (Reference 2.5.2-12) has observed cases of RIS in the Piedmont of South Carolina.Because of this observation and due to the fact that every reservoir produces some modifications in the local stress field, the pre-existing state of stress, the magnitude of the induced stress, and the geologic and hydrologic setting of the SHNPP have been examined to seek out any possible correlations that might exist with the RIS sites in the South Carolina Piedmont.Amendment 65 Page 145 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The SHNPP site does not fall within any of Bollinger's zones of southeastern seismicity and no over pressured groundwaters have been noted in the Durham Basin (Section 2.5.1.2.3.3). This is not the case for the four RIS sites in South Carolina. All four lie within Bollinger's South Carolina-Georgia Seismic Zone. This zone has exhibited historical seismic activity and over pressured groundwaters have been reported (Section 2.5.1.2.3.3). Based upon these data it appears that the SHNPP site is located in a region of lower ambient stress than the RIS sites in South Carolina.The maximum magnitude of the induced stress caused by reservoir impoundment is related to its depth and volume. Table 2.5.2-2 lists the reservoirs with RIS in South Carolina and the SHNPP site reservoirs. The maximum depth of the SHNPP site reservoirs is 19 meters. This is less than one half of the depth of the Monticello Reservoir, which is the shallowest South Carolina site of RIS. In addition the total volume of the SHNPP Reservoirs is less than 20 percent of the volume of the smallest South Carolina RIS site. The maximum stress change due to loading at the SHNPP site is 1.85 bars. This is also the maximum possible increase in pore pressure (assuming that the depth to the water table is negligible). This value is less than half the calculated stress change induced by the smallest South Carolina RIS site.All four South Carolina RIS sites are located upon gneissic rocks of the Piedmont physiographic province. The SHNPP site reservoirs are located predominately upon clastic sediments of the Durham Basin. The site geology is discussed in detail in Section 2.5.1.2. The permeability values obtained at the site are generally low (Section 2.4.13). This suggests that the migration of pore pressures to depth will be inhibited.A comparison of the SHNPP site reservoirs to four reservoirs with RIS in South Carolina shows no correlation in pre-existing stress, water depth, water volume, geologic or hydrologic conditions. Therefore, there is no reason to expect the SHNPP site reservoirs to induce any significant seismic activity.2.5.2.1.3 Plant Seismic Monitoring Network On January 6, 1976, the NRC formally notified CP&L that the NRC had concurred with CP&L's conclusion that the fault discovered at the SHNPP is not a capable fault as defined in Appendix A to 10 CFR Part 100. The NRC requested seismic monitoring at the site to confirm their conclusion that the proposed reservoirs at the site will not cause fault movement during and after filling.In response, CP&L submitted a proposal dated February 13, 1976, to establish a seismic monitoring network encompassing the SHNPP site area. Although this proposal called for monitoring to begin in January 1979, the network was installed and became operational on September 30, 1977, in order to obtain more definitive baseline data prior to filling the reservoirs.The seismic monitoring network consists of an array of stations, Numbers 1 through 4, covering the SHNPP site area. Each contains a Teledyne Geotech Model 18300 vertical-component, short-period seismometer. One station, No. 4, also has two horizontal-component Model 18300 seismometers aligned N-S and E-W. Data from Stations 1 through 3 are transmitted via UHF radio to the SHNPP meteorological tower. Station 4 data are transmitted to the meteorological tower via commercial telephone lines. At the meteorological tower, data from all stations are multiplexed and transmitted by telephone data lines to the central recording facility in Raleigh Amendment 65 Page 146 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 and are recorded on 16 mm microfilm by a Teledyne Geotech RF 400 Develocorder. Carolina Power & Light Company committed to operate the SHNPP seismic monitoring network continuously until two years after the filling of the Auxiliary Reservoir, which reached full-pond in March 1983.During this period, seismic events detected in the site area were logged and discussed in quarterly reports to the NRC.The station locations are:Surface Rock Station Coordinates Elevation Magnification (Lat., Long.) (m) (1 Hz)Triassic Sedimentary 1 35°37'55" N 86.0 134,000 78°58'49" W Piedmont Crystalline 2 35°33'17" N 74.7 213,000 78°59'23" W Piedmont Crystalline 3 35°35'36" N 94.5 67,000 78°54'16" W Triassic Sedimentary 4 35°39'57" N 93.0 169,000 78°54'16" W 2.5.2.2 Geologic Structures and Tectonic Activity A detailed description of the geologic structures of the region and tectonic activity is presented in Section 2.5.1.2.5.2.3 Correlation of Earthquake Activity With Geologic Structures or Tectonic Provinces In recent years a number of authors have discussed the seismicity of various parts of eastern North America and its relation to various tectonic features. A major difficulty in such analysis is the near absence of instrumental data.Woollard (Reference 2.5.2-16) inferred a well-defined seismic belt associated with the Appalachian Mountain Belt. He also mentioned a seismic trend that connects Charleston to the Appalachian Belt. McClain and Meyers (Reference 2.5.2-17) observed a zone of relatively high seismicity in the southern Appalachian Mountain areas of Virginia, eastern Tennessee, western Carolinas, northern Georgia, and Alabama. This zone coincides with the southern portion of the Appalachian Mountain Geologic province, its seismicity probably representing minor adjustments of the highly disturbed rocks. Nuttli (Reference 2.5.2-18) observed that the distribution of epicenters throughout the eastern United States is diffuse and that epicenters do not appear to be associated with long, active faults. Also, eastern earthquakes are not associated with visible surface faulting. Ferguson and Stewart (Reference 2.5.2-19) observed that certain parts of North Carolina appear to be more seismic than others although no faults have been defined in these areas.A marked alignment of epicenters across western North Carolina and on into Virginia was first observed by MacCarthy (Reference 2.5.2-20), who postulated that such clear-cut linear alignment probably represents a real tectonic alignment, a belt of structural weakness in the uplifted Blue Ridge. He hypothesized that there is a zone of weakness, with numerous small movements that from time to time generate rather localized tremors and disturbances. Oliver Amendment 65 Page 147 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 and Isacks (Reference 2.5.2-21) observed that in the Appalachian region, earthquake epicenters follow areas of highest topographic elevations.Fox (Reference 2.5.2-22) observed that the Blue Ridge and Piedmont provinces are associated with moderate seismicity, i.e., shocks of intensity less than VII. Fox also observed that seismic activity is minor in the extensive area between Richmond, Virginia and Charlotte, North Carolina.From analysis of earthquake distribution in the southeastern United States, Bollinger (Reference 2.5.2-23) concluded that epicenters occur in four zones: two that are parallel and two that are perpendicular or oblique to the dominant northeast trend of Appalachian structure (Figure 2.5.2-2). He identified the following seismic zones:a) Southern Appalachian Seismic Zone - extends from western Virginia to central Alabama in the Valley and Ridge, and Blue Ridge provinces.b) Northern Virginia - Maryland Seismic Zone - A diffuse northward extension of the above.c) Central Virginia Seismic Zone - a relatively narrow, isolated zone of activity in the Piedmont province, offset from the above two zones, and oblique to the NE-SW structural trend.d) South Carolina - Georgia Seismic Zone - a broad zone spanning the Piedmont and the Coastal Plain provinces and transverse to regional structure.The area around the SHNPP site is located between the Central Virginia, Southern Appalachian, and South Carolina-Georgia seismic zones defined by Bollinger (Reference 2.5.2-

23) (see Figure 2.5.2-2). The SHNPP site area has exhibited no seismicity. The Piedmont province, in which the site is located, is active only in central Virginia, in South Carolina, and in northwest Georgia. Appreciable earthquake activity in the Coastal Plain province occurs only in South Carolina.

In general, seismic activity occurs primarily in the Valley and Ridge and Blue Ridge provinces, although, even considering epicentral errors, earthquakes occur in all four of the geologic provinces in the southeastern United States (Figure 2.5.2-3). Furthermore, activity is oriented both parallel and transverse to the dominant northeast-trending Appalachian structures. Many thrust faults and normal faults that have been mapped in each province are very ancient and have no records of surface rupture during recent geologic time. Bollinger (Reference 2.5.2-24) identified three factors which make it difficult to correlate earthquakes with near-surface geological structures:a) Inadequate seismograph-station density to determine focal depths or focal mechanisms.b) Minor versus major seismic-zone characteristics, especially with regard to surface expressions of earthquake faulting.c) Great difference in time scales between seismicity data and regional geologic data.Amendment 65 Page 148 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Focal depths and mechanisms are the critical seismic parameters that may be directly correlated with geology. Focal depths, for example, can indicate whether the casual fault was in the sedimentary section, basem*nt rocks or deep within the crust. Focal mechanism data can yield strike and dip of the fault plane and the direction of slip on it.The second factor, relating the levels of seismicity and tectonism, is intrinsic to the region. The lack of surface faulting during earthquakes prevents immediate correlation with many major and minor faults mapped in the area by geologic techniques.The third relates to the relevant time frames. When earthquake data (spanning more than 200 years) are compared with the geologic data (spanning more than one billion years), the time interval difference is seven orders of magnitude. Thus, the task is deciding which, if any, of the ancestral geologic features are pertinent to observed seismicity.2.5.2.4 Maximum Earthquake Potential Figure 2.5.2-3 shows that earthquakes occur in all four geological provinces in the area.Seismic activity is oriented parallel and transverse to the dominant northeast-trending Appalachian structures. The many thrust and normal faults mapped in each province do not have a record of surface rupture during historical times; therefore, detailed correlation of seismicity and geology in this region is impossible without detailed geological and geophysical investigations. Moreover, an appropriate tectonic (or seismotectonic) map for the eastern United States is unavailable at present. Under the circ*mstance, the seismic zones defined by Bollinger (Reference 2.5.2-23) on the basis of spatial distribution of earthquakes and geodetic and tide gage data appear to provide a reasonable basis for the investigation of maximum earthquake potential and evaluation of seismic risk.The largest historical earthquakes associated with each seismic zone in the region are identified in the following:a) Southern Appalachian Seismic Zone The Giles County, Virginia, May 31, 1897, MM intensity VIII, and Gadsden, Alabama, January 27 and 28, 1905, MM intensity VII-VIII events, were the two largest earthquakes in this zone. Minimum distance between zone and site is about 165 km. An earthquake with epicentral intensity VIII would be barely perceptible at this distance and would have no damage potential at the SHNPP site.b) Northern Virginia - Maryland Seismic Zone The Luray, Virginia, April 9, 1918, MM intensity VI and northern Virginia, September 5, 1919, MM intensity VI events, were the two largest earthquakes in this zone. The minimum distance between this zone and the SHNPP site is about 315 km. An intensity VI earthquake would not be felt at this distance.c) Central Virginia Seismic Zone The Petersburg, February 21, 1774, MM intensity VII and Richmond, December 22, 1875, MM intensity VII events, were the two largest earthquakes in this zone. The Amendment 65 Page 149 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 minimum distance between this zone and the SHNPP site is about 155 km. An intensity VII earthquake would be barely perceptible at this distance.d) South Carolina - Georgia Seismic Zone The largest earthquake in this zone occurred near Charleston, South Carolina on August 31, 1886, and was felt with MM intensity X in the epicentral region. Figure 2.5.2-4 is an isoseismal map of this earthquake, prepared by Bollinger (Reference 2.5.2-25). It shows that the earthquake was probably felt with a MM intensity V (but not more than VI) in the SHNPP site area.From studies related to the Charleston, South Carolina earthquake of 1886, Rankin (Reference 2.5.2-26) concluded the following.

 "The extent and boundaries of the Charleston block are not well known, but it does appear to underlie a sizeable area of the emerged and submerged Coastal Plain. The Orangeburg scarp appears to coincide with the northwestern boundary of the Charleston block and may be structurally controlled. "Why did the 1886 earthquake occur in the Charleston-Summerville area rather than elsewhere in the Charleston block? Is it reasonable, in fact, to restrict the probability of a recurrence of an 1886 earthquake to the Charleston block at all? Clearly, we need to know more about the Charleston block and about the nature and location of the boundaries of this block."

Without prejudice to the geographical boundaries of the Charleston block and without considering whether or not a recurrence of an "1886 earthquake" may be restricted to the Charleston block, a conservative determination of the vibratory ground motion may be made by considering the recurrence of such an earthquake in the South Carolina-Georgia seismic zone at its minimum distance, about 130 km, from the site. Using the attenuation relation, I = Io + 2.87 - 0.000 52R - 2.88 log R, (where R is the specified distance in km.)developed by Bollinger (Reference 2.5.2-25) for a 50 percent fractile of the data set, a site MM intensity of 6.71 is obtained.e) New Madrid Seismic Zone Three large earthquakes, each with epicentral MM intensities XI-XII, occurred on December 16, 1811; January 23, 1812; and February 7, 1812, near New Madrid, Missouri. Figure 2.5.2-5 is an isoseismal map for the earthquake of December 16, 1811.A microearthquake network study, currently under progress at St. Louis University, reveals that the activity in this region, which may be described as the New Madrid seismic zone, is strongly localized (Reference 2.5.2-27). The minimum distance of the New Madrid seismic zone from the site is about 900 km. From the attenuation relation, Amendment 65 Page 150 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 I = Io + 2.35 - 0.00316 R - 1.79 log R, presented by Gupta and Nuttli (Reference 2.5.2-28), it follows that a recurrence of the December 16, 1811 earthquake in the New Madrid seismic zone at the point nearest the site would be felt with a MM intensity of 5.2 in the vicinity of the site.f) Background Seismicity in the Region In the site region the largest historical earthquakes have been of MM intensity V. These cannot easily be attributed to any particular geologic structure or seismic zone. Included are two North Carolina coast earthquakes on January 18, 1884, and March 5, 1958, and one western North Carolina earthquake on August 26, 1916.2.5.2.5 Seismic Wave Transmission Characteristics of the Site To determine seismic wave transmission characteristics of the site, the following site geophysical surveys were made:a) Seismic refraction lines to define bedrock topography.b) A seismic refraction line to detect both compression wave and shear wave arrivals for evaluation of dynamic soil and rock properties.c) A downhole velocity survey to further define dynamic rock properties.d) Ambient noise studies to determine predominant periods of ground motion due to background noise levels.Figure 2.5.2-6 shows the location of these studies. The following sections describe each phase of the geophysical exploration.2.5.2.5.1 Seismic Refraction Surveys Refraction surveys were made along six seismic lines totaling approximately 5,000 linear ft.Seismic lines 1 through 4 were completed at selected locations adjacent to their respective exploration trench numbers. Seismic lines 5 and 6 extend southeast from Trench 2 to the vicinity of Borings P6 and P7, respectively. These lines are approximately perpendicular to Trench 2.The recording equipment used for this refraction investigation were a portable Electro-Tech ER-75-12 refraction seismograph and Electro-Tech EV-5-4 geophones with a natural frequency of 14 cycles per second. The geophones were fitted with spike attachments for coupling with the underlying soil. Geophone spacings of 25 to 50 ft. were used. Explosives (normally 5 lb.charges of Nitramon-S placed in 10 ft. deep drill holes) were detonated in the center, ends and beyond the ends of each line.The geophysical crew consisted of two geophysicists who supervised the field investigation, operated the recording instruments, and made preliminary interpretations of geophysical data in the field. A licensed powderman handled and placed the explosive charges.Amendment 65 Page 151 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The results of the geophysical refraction survey, showing profiles of various strata and compressional-wave velocities in soils and underlying bedrock, are given in Figures 2.5.2-7 through 2.5.2-10. The velocities of compressional wave propagation in the upper soils and underlying rock were computed from the plotted data. In addition to the geophysical profiles, the plots of the time-distance data resulting from the survey are shown immediately above the corresponding profile. The accuracy of the calculated depth to bedrock is considered to be within 10 percent for the major portion of the survey; however, in areas where sound bedrock is indicated at shallow depths, the precision is probably less.Velocities of the different strata, as evaluated from the refraction surveys, are as follows:a) Residual soils and/or highly weathered rock; the velocity range is 1,250 to 2,000 ft./sec.b) Weathered and/or fractured bedrock; the velocity range is 5,000 to 7,150 ft./sec.c) Sedimentary bedrock (probably unaltered); the velocity range is 10,900 to 13,650 ft./sec.2.5.2.5.2 Shear-Wave Velocity Survey A shear-wave velocity survey was conducted along a 3,400 ft. section of Trench 1. Shear-wave velocities were computed from the recordings of two Sprengnether Engineering Seismographs placed at 350 ft. intervals along portions of the trench. Shot holes were located at varying distances, up to 3,400 ft. from the farthest geophone. Traces from the Sprengnether seismographs, and eight 1-component geophones placed at 100 ft. intervals were recorded on an Electro-Technical SDW-100 oscillograph.Survey results, with compressional wave velocities and computations of Poisson's Ratio, are summarized in Table 2.5.2-3. These data refer to subsurface conditions in the vicinity of Boring P5, which are considered typical.2.5.2.5.3 Downhole Velocity Survey A downhole velocity survey in Boring P6 provided a check on the compressional wave velocities measured during the seismic refraction surveys. The boring was cased with steel casing to 20 ft. below the ground surface. Small explosive charges were buried approximately 5 ft. deep at distances of 10 to 25 ft. from the boring. In the boring, seismic response to the detonations was detected with a 12-trace geophone cable and was recorded with an Electro Technical Labs M E amplifier and a SDW-100 oscillograph.Figure 2.5.2-11 shows results of the downhole velocity survey. The compressional velocity of the bedrock measured in this survey is greater than the corresponding compressional velocity measured by seismic refraction, because seismic refraction surveys record average dynamic bedrock properties over a lateral distance, while downhole velocity surveys record the properties at an isolated point. Thus, compressional velocities measured during the seismic refraction survey are more representative of the actual dynamic bedrock properties. A zone of slower velocity material (9,250 ft./sec.) is apparent between depths of 93 and 130 ft.Amendment 65 Page 152 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2.5.4 Ambient Vibration Measurements The last phase of geophysical field studies was measuring the level of ambient ground motion (microtremors) caused by various natural and man-made sources. On September 18, 1970, measurements were taken at the five locations shown on Figure 2.5.2-6; all were taken when no equipment or drills were in operation at the site.A direct-writing Model VS-1100, Sprengnether Engineering Seismograph recorded ambient ground motion. A VS-1100D amplifier was used with the seismograph, resulting in a maximum gain level of 2,000. The three components of ground motion measured were radial, vertical and transverse.Table 2.5.2-4 shows the results of ambient ground-motion measurements. The observed characteristic frequencies of the site are 100, 55 1/2, and 25 Hz. The maximum observed level of groundmotion, which is in the vertical component, is 0.48 x 10-3 inches/second at about 25 Hz. Location No. 5 at the northwestern end of Trench 1, appears to be the quietest location, whereas locations 3 and 4, near Borings P6 and P7, respectively, appear to be least quiet. See Section 2.4.13 for groundwater levels, as they may affect seismic velocities in unconsolidated/loosely consolidated sediments.2.5.2.6 Safe Shutdown Earthquake To establish criteria for the SSE, the degree of remotely possible ground motion has been considered in light of the seismic history and the geologic structure of the region and the site.This included the effects upon the site of the recurrence of the earthquakes discussed in Section 2.5.2.1, when located so that their epicenters, or regions of maximum intensity, were at minimum distance from the site. The maximum effect at the site would be caused by a recurrence of the 1886 Charleston earthquake. The result would be a MM intensity 6.7 earthquake if it were assumed to occur in the South Carolina-Georgia seismic zone at its minimum distance from the site, or VI, if it were considered localized to a structure near Charleston.The general pattern of seismicity within the Piedmont was also considered. Shocks in the Piedmont, such as the 1875 intensity VII shock near Richmond, have been known to occur in the vicinity of Triassic basins. An intensity VII shock near Wilmington, Delaware in 1871 has not been related to known geologic structure. Because of limited documentation, its exact epicenter cannot be fixed.Coastal Plain sediments mask the bedrock in the area south and east of Wilmington, and much of the bedrock geology is still unknown; consequently, it is possible this shock is related to some unidentified Triassic structure. Hence, the remote possibility of a shock of intensity VII occurring in the Deep River Basin, close to the site, has been considered.Based on the foregoing, the SSE is designated an intensity VII earthquake with its epicenter near the site. At foundation level within the competent bedrock at the site, the maximum horizontal ground acceleration due to such a shock is estimated to be less than 12 percent of gravity.In order to provide an additional margin of conservatism, a value of 15 percent of gravity is assigned as the maximum horizontal ground acceleration. All safety related structures and Amendment 65 Page 153 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 systems have been designed to assure safe plant shutdown for two horizontal excitations and one vertical excitation simultaneously.The SSE is postulated as a shock of low magnitude close to the site, of less than 10 seconds expected duration, and producing no more than 5 to 10 cycles of motion at the maximum predicted level. This pattern of motion should be similar to the Golden Gate (San Francisco) earthquake of 1957 or the Helena, Montana, earthquake of 1935. Records from these earthquakes indicate only 2 to 3 cycles of strong motion; therefore, the selection of 10 cycles of strong motion for the SSE is conservative.Seismic Category I systems and components, therefore, have been designed for a minimum of 10 loading cycles under SSE conditions. Figures 2.5.2-12 and 2.5.2-13 show the horizontal and vertical response spectra for the SSE, prepared in accordance with Regulatory Guide 1.60.Recent seismic network monitoring supports the observation that the SHNPP site is located in an area of low historic seismicity that is lower than that of neighboring areas of the Piedmont Province to the north and south in the eastern United States (EUS). Inasmuch as MMI = VII is the largest intensity earthquake known to have occurred in the Piedmont Province in the southeastern United States and only one event of that intensity has been associated with Triassic basin areas in the Piedmont (the Richmond, VA earthquake of December 23, 1875), it was concluded above that a random MMI = VII event near the site provided an adequately conservative basis for the SSE design response spectrum for the SHNPP.In consideration of the above discussion, the maximum random earthquake near the Harris site is estimated to be associated with MMI = VII. There do not exist sufficient instrumental data to define an intensity - magnitude relationship for the Piedmont Region, nor are there numerous near-field strong motion records from the region to estimate a site-specific spectrum. As a result, we adopt general intensity - magnitude relationships and data obtained in other regions.Studies of earthquake intensities and magnitudes indicate that MMI = VII would be associated with an earthquake of local magnitude ML approximately equal to 5.3. For purposes of generating a site-specific spectrum, it is appropriate to consider strong motion records within one-half magnitude unit of this value, i.e., 4.8 ML 5.8.As an alternative to developing new site-specific spectra to compare to the Harris SSE, one option is to use available compilations of spectra for the appropriate magnitude range. One compilation is presented in Appendix A of NUREG/CR-1582, Volume 4, prepared by Lawrence Livermore National Laboratory (LLNL). Figure 4-7 of that document presents mean and mean-plus-one-standard deviation spectra for records in the magnitude range 4.8 ML 5.8 obtained at distances less than 25 km. (The authors of that document consider these spectra representative of MMI = VII motion ). These spectra, which are for five percent damping, are reproduced in Figure 2.5.2-18; also shown in that figure is the SSE spectrum for the Harris site, for five percent damping. The LLNL spectra were originally published in 1979 and do not include more recent data, nor do they include data from prior earthquakes which indicate low amplitude motions. Generation of new site-specific spectra which include these additional data in the appropriate magnitude and distance range may ultimately be appropriate, but all these additional data are from NON-EUS sites. For the present, it is sufficient to compare the Harris SSE spectrum with the LLNL spectra, to obtain an indication of the SSE's validity.The LLNL spectra shown in Figure 2.5.2-18 were generated for two classes of site conditions (rock and soil). The rock spectra are most appropriate for comparison to the SSE spectrum Amendment 65 Page 154 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 because the Shearon Harris facility is founded on competent lithologic units (Triassic) as discussed in the FSAR. For completeness, both the rock and soil spectra are compared to the Shearon Harris SSE spectra in Figure 2.5.2-18.These comparisons show that the Harris SSE design spectrum (five percent damping) envelopes or matches the rock mean +1 (one standard deviation) site specific spectrum over the entire frequency range. From this, we conclude that the Harris SSE spectrum is appropriately conservative for representing the maximum random earthquake in the Piedmont province, for which the facility is designed.As an aside, we note that the soil mean +1 spectrum is somewhat higher than the SSE spectrum above 12 hz. The high frequency end of the soil spectrum is of questionable reliability for several reasons:a) It has essentially the same values at high frequencies as the ML = 5.8 soil curve based on 5.3 ML 6.3 events (NUREG/CR-1582, Volume 4, Figure 4-8 compared to Figure 4-7) whereas the rock site-specific spectra differ systematically at all frequencies as expected for larger vs. smaller magnitude events.b) The soil spectra for ML 5.3 were derived by including in the analysis three Japanese records with peak accelerations as high as 0.6 g. Reanalysis of these accelerograms to account for instrument correction generally reduced the values at high frequencies: the 0.6 g. peak acceleration was reduced to 0.5 g., for example (Matushka, Personal Communication, 1982). Moreover, there may have been soil resonance effects affecting these records which would not be relevant for a rock site.Inasmuch as the rock site-specific spectrum is the most appropriate for the Shearon Harris site conditions, we conclude from the comparisons presented that the SSE design is adequately conservative for the maximum random earthquake that would occur near the site.2.5.2.7 Operating Basis Earthquake The study discussed in Section 2.5.2.6 examined the degree of ground motion which is considered possible during the economic life of the facility based on the seismic history of the region and the site area. It is concluded that it is unlikely that the site would be subjected to ground motion above intensity VI during the life of the facility; therefore, the accelerations of the operating basis earthquake were taken as half those of the SSE. The corresponding horizontal acceleration at foundation level in the bedrock would be less than 7.5 percent of gravity.Accordingly, the OBE is designated an intensity VI earthquake, with its epicenter near the site.Figures 2.5.2-14 and 2.5.2-15 show the horizontal and vertical design response spectra for the OBE, prepared in accordance with NRC Regulatory Guide 1.60 and scaled to .075g horizontal ground acceleration.Figure 2.5.2-16, Earthquake Occurrence Probability, plots the seismic history within 250 miles of the site. Assuming the seismicity of the area to be relatively uniform within the 500-mile diameter circle, a parallel relationship was drawn representing the earthquake occurrence probability within a 10-mile radius. This is a very conservative approximation, since earthquake occurrence in the region is not random and uniform but is generally confined to well-defined Amendment 65 Page 155 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 local areas far distant from the site. On this statistical basis, it is conservatively concluded that the site will undergo intensity VI ground motion (the OBE) on the order of once every 3,500 years.2.5.3 SURFACE FAULTING Studies of regional and site geology, detailed in Section 2.5.1, provided information on overall geologic conditions in the area. In addition, to provide assurance against the presence of capable faults in the area as well as to furnish basic engineering data, a number of trenches and boreholes were completed at locations shown on Figures 2.5.1-10 through 2.5.1-14.The trenches varied in depth from 2 to 12 ft., depending on the nature of the overburden and the depth to rock too hard to cut with a backhoe. In the upland areas the overburden, consisting of residual yellow sandy clay and sandy loam, is only 2 to 4 ft. thick. It is quite uniform in texture and composition and stands well in the sides of the trenches. Where this overburden is only 2 to 4 ft. thick, rocks too hard to cut with a backhoe were usually reached at depths of 5 to 6 ft.Along stream courses and other low areas, yellow sandy, clayey alluvium is commonly 3 ft. to 10 ft. thick; it generally caved badly, and satisfactory information could not be obtained on the bedrock in portions of trenches. In these areas trenching was supplemented by drill holes 40 to 100 ft. deep, in order to determine the character and soundness of the bedrock beneath the alluvium. A number of drill holes, 100 to 250 ft. deep, were put down to determine the soundness and loadbearing capacity of the rocks underlying the plant site.The plant site area was explored by four major trenches having a total length of slightly over 12,000 ft. They were supplemented by five minor trenches, 50 to 100 ft. long, and numerous drill holes, 50 to 152 ft. deep. Figure 2.5.1-11 shows trench locations; detailed logs of the four major trenches are presented in Appendix 2.5A.Minor cross faults are characteristic of the Triassic bedrock surrounding the site. However, the exploratory trenches at the site revealed no evidence of cross faults. Several small diabase dikes were exposed in Trenches 1 and 2, west and south of the proposed reactor site.During excavation of the Waste Processing Building, a small cross fault, hereafter referred to as the "site fault," was discovered. Thus, two faults, the Jonesboro fault and the site fault are relatively close to the site. Further studies were made to determine their possible relationship to each other, and the capability of the site fault (Reference 2.5.1-29).As a result of these studies, the Jonesboro Fault and the site fault are considered to be of the same general age; the site fault, however, having a history of movement later than that of the Jonesboro fault, along which movement occurred for a very long time that included the beginning of deposition of Triassic sediments and that ended after the intrusion of Jurassic diabase dikes. Movement along the site fault occurred after deposition and lithification of several thousand feet of Triassic Basin sediments and ended shortly after intrusion of the latest of the Jurassic dikes.Evaluation of the small vertical and horizontal components of movement of the site fault suggests it is short, with shallow-roots, compared to the Jonesboro Fault, along which movement began much earlier. Both faults are considered to be rooted in the crust. These shallower upper crustal fractures are in contrast to the deeper seated north-northwest oriented lines of tension revealed by the contemporaneous intrusion of diabase dikes, probably from the Amendment 65 Page 156 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 upper mantle. Such upper mantle material, intruded as diabase dikes into the Triassic Basin and the Piedmont, did not follow either of the faults.The site fault and the Jonesboro Fault belong to a conjugate set of normal faults. Stresses that produced them caused rotational deformation, emphasized by major motion along the Jonesboro Fault, the primary fault of the set. The site fault is an antithetic normal fault of this conjugate set. Normally the primary and antithetic fault planes tend to have parallel strikes, but in regions of differential vertical motion, the strike of the antithetic set may not be parallel; however, the geometry of the set precludes intersection of the faults. East of the site, stresses were most probably released by faults parallel to the site fault in an "En Echelon" arrangement.Since the Late Jurrasic, the site area has been remarkably stable. The Triassic rocks have not been further faulted, and no faults offsetting strata younger than Miocene have been found in the site region.The site fault is overlain by sedimentary rock which has not been offset or otherwise disturbed by movement on the fault. Evidence from comparative lithology, depth of oxidation, and soil-profile thickness indicates the sedimentary rock is older than one million years and probably much older, perhaps as old as Jurassic.Intrusion of diabase dikes at the site occurred before and during fault movement, which continued for a relatively long time. The dikes range in age from an absolute minimum of 150 million years to a maximum of about 225 million years, based on remanent magnetization studies, and from a minimum of 168 million years to a maximum of 260 million years, based on Potassium/Argon (K/Ar) dating.The Jonesboro Fault and the site fault are relict structural features in a tectonically very stable region. Neither they nor other faults in the site area are capable faults. Further details of these investigations are presented below and in the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).Several small non-capable faults were found in the foundation of the Main Dam as described in Section 2.5.6.2.2.5.3.1 Geologic Conditions of the Site Details of the geologic conditions of the site are presented in Section 2.5.1.2.5.3.2 Evidence of Fault Offset 2.5.3.2.1 The Jonesboro Fault The southeastern side of the Deep River Basin is formed by the Jonesboro Fault, a northeast-southwest trending diagonal slip fault, dipping northwest, with vertical displacement of 8,000 to 10,000 ft. and unknown right-lateral displacement. For more than 100 miles along the southeast side of the Deep River Basin, the fault is the contact between Triassic and Paleozoic rocks.There are five other major northeast trending longitudinal faults within the northern half of the Deep River Basin. Of the six, the nearest to the site is located four miles to the southeast. It is the only one identified in the site area investigation (see Figure 2.5.1-4). The Jonesboro Fault Amendment 65 Page 157 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 and the other major faults of the Deep River Basin are probably reactivated older structural trends in the basem*nt rocks.In addition, a system of northwest-trending minor cross faults fractured the rock of the southern half of the Deep River Basin. During the site investigation, no evidence was discovered to suggest the Jonesboro Fault is offset by cross faults.Campbell and Kimball (Reference 2.5.1-16), who believe that final movement on the Jonesboro Fault was the last episode in the structural development of this region, reported that the fault cuts all other structures.Reactivation or initiation of tensional, normal-type movement along some possible lateral components on the Jonesboro Fault was followed by deposition of Triassic sediments. Prouty (Reference 2.5.1-33) concluded that movement on the Jonesboro Fault continued sporadically from the beginning of sedimentation until after it ended.Magnetic and reconnaissance surveys were conducted on diabase dikes and "cross faults" occurring along the Jonesboro Fault in an effort to date the site fault. Five of these locations, shown on Figure 2.5.3-1, were studied and mapped. Magnetic data was collected at each location using a portable proton magnetometer along traverses which were surveyed by Brunton compass. This magnetometer data was used along with information obtained from standard field investigation methods to correlate (based on geometric relationships) similarities and differences between dikes encountered on either side of the Jonesboro Fault. In addition, seismic and plane table surveying methods were used at location 5. The information gathered at locations 1-5 indicates that none of the dikes mapped at these locations are continuous across the Jonesboro Fault. Evidence, although inconclusive, suggests that at least one pair of dikes at each of the first four locations were at one time continuous and have subsequently been offset by the Jonesboro Fault. The amount of offset between dikes at southern locations 1-4 varies from 1300 ft. at location 4 to 420 ft. at location 3. However, possible offset at the northernmost area, location 5, is only 60 ft. An alternative interpretation at location 5 is that diabase was intruded into the fault zone concurrent with last movement of the fault.Consequently, dating of the last movement of the Jonesboro Fault by the absolute dating of diabase does not appear possible at locations 1-4 and is questionable at location 5.The age of last movement on the Jonesboro Fault therefore bracketed between the intrusion of Late Triassic-Jurassic dikes and the deposition of the Cretaceous marine sediments overlying it.Appendix M of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29) discusses the Jonesboro Fault diabase dike relationship fully.2.5.3.2.2 The Site Fault The site fault was discovered in the excavation of the plant Waste Processing Building. Figures 2.5.1-15 and 2.5.3-2 show the fault in relation to the plant excavation; Figures 2.5.1-4 and 2.5.3-3 show the fault in relation to the geology of the site.Amendment 65 Page 158 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.3.2.2.1 Investigation Procedures and Results 2.5.3.2.2.1.1 Locating the Fault After discovery of the site fault, it was traced some 8,000 ft. east and west by digging short trenches normal to the fault. When exposed in sedimentary beds, the fault exhibits a southerly dip between 60 and 90 degrees, always exhibits drag folding on the hanging wall, and seldom exhibits any disturbance of bedding planes on the northern or foot wall. The fault tends to become oversteepened in coarser grained sandstones and is nearly vertical adjacent to diabase dikes offset by it.In all cases, identification of the site fault was positive. Furthermore, 4,000 linear ft. of trenching revealed no additional faults. Details of procedures and observations along the fault are described in Appendix P of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).The alignment of the fault and detailed trench maps are shown on Figures 2.5.1-15, 2.5.1-16, 2.5.3-4, and 2.5.3-5.2.5.3.2.2.1.2 Identification of Linear Features by Remote Sensing Investigation of the fault discovered in the excavation of the plant included extensive use of remote sensing techniques to seek other linear features in the site region and area. These techniques included conventional high and low altitude aerial photography Side Looking Airborne Radar, including false color enhancement, and Skylab and Landsat imagery.Location, length, and alignment of several hundred linear features were identified. Throughout this study, interpretations of lineaments were conservative. Only those features identified with some certainty were recorded; other marginal alignments were not included.In the field, direct visual observation was made (geologist and photogeologist) of at least one checkpoint along each significant lineament, with emphasis on possible manifestations of ground movement. Lineaments close to the site, of unusual length, or parallel to known regional fault trends were given special attention.No linear features were identified as capable faults on the basis of imagery evaluation, nor were any that were field checked identified as faults. The site fault was undetected by any imagery technique.No evidence of surface or near-surface faulting was found near the site or regionally during remote sensing investigations and fuel checks, nor was any evidence found of other features potentially inimical to the safety of the plant. Remote sensing investigations are discussed in detail in Appendix C of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).2.5.3.2.2.1.3 Locating Diabase Dikes A magnetic survey, using a proton magnetometer, was made across the SHNPP plant area in conjunction with the fault investigation. The purpose was to locate and trace all diabase dikes in the plant area. Readings were taken at 10 to 50 ft. intervals along several east-west traverses approximately parallel to the fault zone. Each dike was traced within the site by taking numerous readings along 25 to 200 ft. long traverses perpendicular to the estimated strike of Amendment 65 Page 159 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 each dike. Only the highest reading, usually representing the down-dip edge of the dike, along each traverse was recorded.Readings recorded along the trace of each dike were plotted on profile sheets. Complete results of site magnetic surveys are given in Appendix A of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).2.5.3.2.2.1.4 Vertical Component of Fault Offset Nine core borings were completed in sedimentary rocks on either side of the site fault to determine the vertical component of offset. The cores were carefully studied for marker beds that could be correlated across the fault and from which offset could be determined. Correlation of marker beds in the site area proved to be extremely difficult because sediment lithology changes radically over short lateral distances.On the west wall of the plant excavation, near the fault exposure on the north, or upthrown, block, there is a distinctive, thick lens of medium to coarse-grained, white to gray, arkosic sandstone. As shown on Figure 2.5.3-6, boring FB-1-74 in the excavation on the south or down-thrown side of the fault, penetrated this same sandstone lens, thus determining the vertical offset. The vertical component of offset at this location was determined to be 83 ft.Petrographic analyses were made of samples taken from the sandstone outcrop and from the sandstone in the boring. These analyses indicate that the petrography of the two sandstone samples is sufficiently similar to have come from different thin sections cut from the same hand specimen. However, this is not conclusive, since many arkosic lenses near the site are petrographically similar.Borings in sedimentary rocks were made outside the plant excavation in two areas that straddled the fault trace to determine vertical offset in those areas. One area was about 800 ft.east of the plant excavation; a second was about 1,200 ft. west of it (Figure 2.5.1-15). In each area, four borings were made, consisting of a pair on each side of the fault to help identify marker beds on one side, which could be correlated across the fault to the second pair. All borings were drilled at least 30 ft. from the fault to minimize effects of drag folding along the fault.East of the excavation, two sandstone marker beds were correlated across the fault, as shown on Figure 2.5.3-7. The vertical component of fault offset in the east area was thus determined to be 98 ft.West of the excavation, the four borings passes through a sandstone marker bed which had been offset about 92 ft., as shown on Figure 2.5.3-8. The results of this program indicate that vertical offset on the fault is greater than 80 ft. but less than 100 ft.2.5.3.2.2.1.5 Horizontal Component of Fault Offset Marker beds in the sedimentary rocks could not be used for determining horizontal offset at the fault because of the gentle dip of beds, lack of distinctive physical characteristics, and lateral depositional lithology changes over short distances. The near-vertical diabase dikes proved to be the best references for determining horizontal offset. However, offset of dikes only shows postintrusive movement, not necessarily the total offset of the sediments.Amendment 65 Page 160 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The best exposure of dike offset is illustrated on Figure 2.5.3-9, which shows the offset of composite East Dike 2 at several elevations. The apparent horizontal offset of the dike segments is quite small for a dike of irregular thickness to have been offset vertically above 90 ft., as were the sedimentary rocks. This suggests that the dike segments have not been offset vertically as much as the sediments. Figures 2.5.3-9 and 2.5.3-10 show that the easternmost segment of the dike displays more apparent horizontal offset than the smaller, westernmost segment at each elevation. This could be explained by separate emplacement of the dike segments during the period of fault movement. The easternmost segment should show more displacement because it is older and has been subjected to a longer period of activity on the fault. Paleomagnetic data (see Appendix G, Shearon Harris Fault Investigation Report, Reference 2.5.1-29) show that the smaller westernmost dike segment is younger than the other segments. This suggests that the sequence of events was (1) movement on the fault, (2) intrusion of the easternmost dike segment, (3) continued movement along the fault, (4) intrusion of the central dike segment, (5) continued movement along the fault, (6) intrusion of the westernmost dike segment, (7) minor continuing movement along the fault, (8) crystallization of laumontite, (9) final movement on the fault, and (10) low-grade burial metamorphism, with crystallization of zeolites harmotome and heulandite.The minimum horizontal offset is 0.5 feet, ranging up to a maximum of 13 feet. A large horizontal component of movement is precluded in that the fault changes strike about every 300 feet.2.5.3.2.2.1.6 Width of Fault Gouge Zone The fault-gouge zone varies from a few inches to about 3 ft. in width. Figure 2.5.3-10 illustrates the varying width of the zone; Figure 2.5.3-11 shows the zone in relation to offset at one of the dikes.2.5.3.2.2.1.7 Age of Movement on the Site Fault The last movement on the site fault was geologically ancient. Independent lines of evidence strongly indicate there has been little or no movement on it in the past 150 million years or more.a) Evidence from Secondary Mineralization A number of secondary minerals are found in the fault gouge at the intersection of the fault by diabase dikes. These include zeolites harmotome, heulandite, and laumontite occurring in the fault gouge near both West Dike 3S and East Dike 2. Minimum age of the zeolites minerals has been determined from their K/Ar content. That these are minimum ages is supported by other special studies. Other secondary minerals identified include calcite, f-saponite, pyrite, and barite. Coal is also present in minor amounts. All secondary materials and coal are found below the present groundwater table. Coal and pyrite occur with carbonaceous shale in the fault plane at the diabase dike-fault intersection at East Dike 2. F-saponite occurs as a secondary mineral in the fault gouge at East Dike 2. While these minerals (other than the zeolites) and coal probably cannot be dated, they formed in the fault plane subsequent to latest movement.K/Ar data firmly establish that heulandite sample 4 KA 74-210 is greater than 10 million years old, as shown with other zeolite ages on Figure 2.5.3-12, but, as the Sr87/Sr86 ratios Amendment 65 Page 161 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 shown on Figure 2.5.3-13 indicate, heulandite is probably cogenetic with emplacement of the diabase dikes and is therefore more realistically 150 million years old.That the age from K/Ar data is much less than that of the dike is easily accounted for by the recognized tendency for zeolites to lose argon. The intact condition of some of the very brittle, delicate zeolites indicates that they were formed after faulting, so that the last movement on the fault was certainly more than 10 million years ago, most probably before the final cooling of the dike about 200 million years ago or at least during a burial metamorphic event before 150 million years ago.The remaining K/Ar minimum ages for zeolite samples on Figure 2.5.3-12 range up to 35.3 +/-14 million years before present (B.P.) for sample 2-ETB. The K/Ar studies further demonstrate that significant amounts of argon are readily lost during preparation of samples for mass spectrometric analyses. It is believed that even at room temperature, argon will be lost from zeolites under vacuum, because of diffusion due to reduced partial pressure outside the crystal structure. Because of this, the calculated dates given for this last group of analyses must be considered absolute minimum ages. From the nature of zeolites, it is apparent that most of the argon is lost naturally, prior to mass spectrometric analysis.Indeed, since the potassium in zeolites resides in large cation exchange locations, there is little tendency for the argon to accumulate.Studies of the Sr87/Sr86 ratios of zeolite samples indicated the likely source of the zeolites which occurred in the fault zone at the intersections with diabase dikes. The Sr87/Sr86 ratios of the zeolites in diabase and fault gouge are consistent with zeolite formation from late-phase hydrothermal solutions derived from the cooling diabase dike and contaminated by no more than 20 percent of the strontium dissolved from the surrounding fault-gouge material.Since the volume of the enclosing sedimentary rocks is much greater than the volume of the diabase, it is extremely unlikely that the zeolites could have precipitated from groundwater solutions with a strontium content derived from the sedimentary rocks reduced in Sr87/Sr86 ratio by later contact with the diabase dike.Zeolite sample 2-ETB has a Sr87/Sr86 ratio so close to that of the adjacent dike that it appears to have undergone very little strontium-cation exchange. This sample, and perhaps all the zeolites, formed with a Sr87/Sr86 ratio of less than 0.7060. The zeolites could not have formed with a Sr7/Sr86 ratio of much less than 0.704 to 0.705, unless they formed from strontium in magma or rocks derived directly from the upper mantle. The initial Sr87/Sr86 ratio of the diabase was 0.704. Therefore, it appears that the only suitable source of the strontium in the zeolites is the same as that of the diabase. Therefore, the diabase and zeolites are genetically related.Secondary mineral samples obtained from the fault zone at East Dike No. 2 and West Dike 3S were studied. Zeolite minerals were not observed in the fault zone at West Dike 3; in the fault zone, away from the dikes; nor in representative rock samples from surrounding Triassic sediments.The zeolite mineral assemblages are hydrothermal and/or "burial" metamorphic and are always associated with the diabase dikes. Evidence for this association can be found in their occurrence only in the dike-fault intersections or as small veins or amygdule fillings in the dikes, and in their absence from the fault zone away from the dikes.Amendment 65 Page 162 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Although it is difficult to determine the exact temperature of formation of these minerals, a conservative assessment would be about 150 +/- 50 C for heulandite and harmotome, and 200 +/- 50 C for laumontite and f-saponite. Ambient temperatures in the sedimentary rocks at these shallow depths of intrusion (as suggested by undeformed amygdules and crystal-lined open cavities in the dikes) were certainly lower than 100 C (around 40 C at 0.3 km with a 50C/km geothermal gradient). There is no evidence of activity in this area since the middle Mesozoic. Hence, it is reasonable to assume that the heat necessary for these hydrothermal solutions could only have come from the dikes. Magmas of this composition begin crystallizing at about 1250 C and are completely solid at about 1140 C. Thus, crystallization of the secondary minerals occurred after the dikes had completely solidified and mostly cooled. The wall rocks immediately adjacent to the dike would have reached 200 C in a geologically short time after intrusion of the dike, given the following reasonable conditions:

1) liquidus temperature of 1250 C

2) depth of intrusion of 0.3 km

3) geothermal gradient of 50 degrees C/Km

4) width of dike of 1 m Obviously, the larger the dike, the higher the geothermal gradient, and the deeper the depth of intrusion, the longer the time that is required for the dike and adjacent rocks to cool.

If some of these secondary minerals owe their existence to the later burial metamorphic event, rather than deuteric hydrothermal activity, then continued downfaulting of the graben and sediment accumulation must have buried the dikes much deeper than the level of crystallization. Assuming a high geothermal gradient (50C/km), a temperature of 150 C requires a depth of burial of almost three km; it would have been necessary for this much overburden to have been removed since middle Mesozoic time. The Jurassic burial-metamorphic event, more than 150 million years B.P. (before present) is the only such event recorded as paleomagnetic chemical remanent magnetization (CRM) in the dike rocks.All the secondary minerals can be found as free-growing crystals in small druses, vugs, and cavities, or as mosaics in veins in the fault zone, so there is no doubt that these minerals are younger than at least the initial movement on the fault. It is critically important to know whether there are any zeolites younger than the last movement on the fault, a question that can be answered by examining each zeolite mineral megascopically and microscopically, looking for evidence of cataclastic deformation or fault-induced strain in the crystals. Thus, zeolites can be readily studied for evidence of age and movement; for this reason, they were the only minerals considered in these studies.Some of the blade-like laumontite crystals examined exhibit undulatory extinction; others did not. Such extinction does not necessarily indicate that crystals were deformed by movement on the fault. Zeolite crystals are very fragile; even grinding the thin-section may cause some strain. Dehydration of laumontite to leonhardite by exposure to air perhaps can also cause strain. Optical properties of this "laumontite" indicate that it is now leonhardite, although it was true laumontite when the x-ray diffraction analysis was made.Amendment 65 Page 163 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 However, there is some stronger evidence to suggest that at least some of the laumontite in the fault zone was cataclastically deformed. A photomicrograph of a laumontite vein in sample ED2-5 (see Appendix H of Shearon Harris Fault Investigation Report, Reference 2.5.1-29) shows clearly that this material has been cataclastically deformed, as evidenced by shearing, and also shows mechanical disaggregation and rotation of laumontite grains.All this deformation could not have occurred in preparing the thin-section. Hence, there was some movement on the fault after crystallization of at least some laumontite.Photomicrographs of harmotome and heulandite show anhedral to subhedral polycrystalline mosaics in veins cutting the fault zones. Undulatory extinction is slight to nonexistent, and the cataclastic deformation seen in some laumontite is not observed. The perthite-appearing structure in the harmotome is complex twinning, a characteristic of almost all harmotome crystals, which can appear superficially as undulatory extinction, but the sharp interface between twin domains and the simultaneous extinction of many domains in a simple grain rule out simple undulatory extinction.Moreover, the small, delicate, pink heulandite crystals of sample D3S-SHNPP (see Appendix H of Shearon Harris Fault Investigation Report, Reference 2.5.1-29) that occur in apparently undisturbed veins at inclined angles to the fault plane could not have withstood movement on the fault without being deformed. Both megascopically and in the microdetails of thin sections, the zeolite minerals do not show effects of mechanical deformation.Therefore, the final movement on the fault can be bracketed between the crystallization of laumontite and harmotome at East Dike 2 and before crystallization of heulandite at West Dike 3S.If the dikes and surrounding rocks cooled rapidly, it would seem coincidental that the crystallization of laumontite and harmotome or (laumontite and heulandite) would bracket the time of last movement on the fault, assuming they were derived from the same hydrothermal solutions. The time interval between laumontite and harmotome crystallization would have been relatively short. It has been reported that Mesozoic diabase dikes of eastern North America underwent burial metamorphism several million years after the dikes were intruded. Paleomagnetic dating showed that the diabases at the Shearon Harris site underwent burial metamorphism about 20 million years after dike intrusion. Accordingly, it is reasonable to conclude that laumontite was associated with the original deuteric hydrothermal alteration of the dike shortly after its crystallization, whereas harmotome and heulandite were formed during the burial metamorphism some 20 million years later. It is concluded that the undisturbed secondary zeolites in the fault gouge have minimum radiometric ages ranging up to about 35 million years but are more realistically older than 150 million years B.P.(before-present).The material presented above has largely been extracted from reports of individual consultants. Their full reports and findings are presented in Appendices B, E, F, and H of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).b) Evidence from Soil, Saprolite, and Sediment Exposures

1) Soil and Saprolite Overlying Triassic Sedimentary Rocks Amendment 65 Page 164 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The fault has not moved during formation of existing soil and saprolite on Triassic sedimentary rocks, as seen in trenches and cuts. Clay mineralogy studies revealed this to be an inplace weathering profile.The soil is described as White Store, a brownish yellow, fine sandy clay to clay. A complete description is included in Appendix I of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29). As seen in outcrop, hand specimen, or microscopically, the residual soil horizon has not been disturbed by the fault. Below this soil horizon, weathering decreases with depth; from 2 to 15 ft. below it, the material is classified as saprolite. Saprolite is thin over the sedimentary rocks because of the impermeable soil and because the sedimentary rocks approach chemical equilibrium under nearsurface conditions much more closely than do the diabase intrusives. Thin secondary clay infillings, primarily montmorillonites and clay-sized micas, are undisturbed along the fault plane and in adjacent joints and fractures. Plaster and Sherwood (Reference 2.5.3-1) state that residual soils developed in a temperate non-glaciated area in the mid-Atlantic region of the United States are quite old, and typically show classical A, B, C, and D profile development. They indicate that a strong profile development would be expected, because residual soils in this area have been variously estimated to be as old as Miocene. More rapid erosion has robbed the plant-site soils of the opportunity to develop this strong profile (see Appendix I of the Shearon Harris Fault Investigation Report, Reference 2.5.1-29). Thorp, quoted in Buol and others (Reference 2.5.3-2),states that a red-yellow podsolic soil, which belongs to a group that includes the soil developed over the Triassic sedimentary rocks at the plant site, is preserved in Georgia under early Pleistocene deposits; it may be Pliocene, and therefore older than one million years.

2) Soil and Saprolite Overlying Diabase Igneous Intrusive Rocks No movement occurred on the fault during weathering and formation of the soil-saprolite profile, 35 ft. thick, overlying the fault trace at East Dike 2. This weathering profile is undisturbed in the saprolite above the fault plane, and there is no differential depth of weathering across the fault. If movement along the fault had occurred during the formation of the thick saprolite, the soft clay materials would have been easily deformed. Any deformation would have been preserved as shearing of the clays, resulting in rearrangement of the normal weathering profile, both in the saprolite and as differential depth of weathering across the fault. This is not the case; the profile is uniform.

The only available evidence tracing the fault across East Dike 2 in the saprolite zone is the apparent offset of individual apophyses of the diabase and at some elevations the deposition in the fault trace by groundwater of what appears to be secondary manganese. This finding is consistent with that of Harrington (Reference 2.5.1-28).Evaluation of the minimum time required to form a saprolite cover over 35 ft. thick at East Dike 2 is based on the following considerations:a) There is no evidence of shearing in the saprolite or residual soil overlying hard rock in East Dike 2 at the fault.Amendment 65 Page 165 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 b) Groundwater levels at the site have been higher than at present, or stable since the Late Jurassic, because the ages of secondary minerals in the fault gouge are of that order of age. These secondary minerals are found only below the mixing zone of oxidation that underlies the groundwater table; they were leached out of the diabase at elevations above this zone.Therefore, it can be interpreted that the secondary minerals would have been leached out of the gouge had the groundwater table been lower during any long period since their formation. These minerals do not form below 100 C, and there is no evidence of a high temperature regime in these rocks since the period of regional low-grade metamorphism following intrusion of the diabase dikes. The minerals occur only in association with the dikes.A dynamic equilibrium exists between the development of the weathering profile of saprolite and surface erosion. Such equilibrium is necessary, but the degree to which it can be refined is questionable. At some places near the site, diabase is weathered to depths as great as 90 ft.; therefore, the equilibrium postulated is largely controlled by such factors as degree of slope, and others relating to the degree of protection of a location from surface erosion over long periods; i.e., mineralogy, climate, and degree of fracturing.The diabase dikes tend to act as barriers to lateral groundwater movement. All springs known in association with them flow from hornfels adjacent to the dikes. The degree of oxidation and saprolitization of the diabase is greatest near the ground surface and above the groundwater table. From the groundwater table down to approximately 30 ft., changes in the degree and intensity of oxidation are very gradual, decreasing with depth. Below 30 ft., however, there is a relatively rapid change with increasing depth to unweathered hard rock. For example, at the East Dike 2 fault, this transition occurs and is complete at a depth of 42 ft.Carson (Reference 2.5.3-3) studied the maximum depth of oxidation in glacial deposits of the Olympic Peninsula in Washington. The greatest depth occurs in the oldest of these deposits, the relatively permeable Wedekind Creek Formation, which is oxidized to depths below 30 ft.,largely above groundwater level. The age of this deposit is 530,000 to 700,000 years B.P. Near the site fault the depth of oxidation below the groundwater table in the diabase of East Dike 2, a profoundly less permeable though coarser grained material, is approximately 40 ft. On the basis of comparative depth of oxidation, the diabase has not been disturbed by movement on the fault for more than 500,000 years.The development of weathering rinds on basalt clasts in the glacial deposits studied by Carson (Reference 2.5.3-3) is analagous to the development of spheroidal weathering rinds commonly seen in diabase at the plant site. This weathering process, carried to its extreme phase, results in the complete decomposition of the diabase, which is chemically unstable under oxidizing near-surface conditions. In the material Carson studied, the greatest mean thickness of such rinds is 6 mm, on clasts from Amendment 65 Page 166 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the Wedekind Creek Formation. According to Carson, the youngest age for this formation is 530,000 years B.P.; therefore, extrapolation of a weathering-rind formation rate of 6 mm/530,000 years would indicate that the complete weathering to clay of a 6-ft. diameter boulder of the basalt would take some millions of years. Inasmuch as the diabase is commonly weathered to clay to depths greater than 30 ft. at East Dike 2 and 90 ft.elsewhere, it may be concluded that the time of last movement of the fault, as indicated by lack of shearing in the saprolite, is at least a few millions of years B.P.Certain variables, however, introduce uncertainty into such a comparison.One of these is the variation Carson found in the basalt clasts he studied and the possibility they underwent some degree of weathering prior to their deposition. Another variable is the relationship of the rate of weathering and oxidation to the location of individual basalt fragments, with respect to groundwater level and the rate of groundwater movement. Still another is the chemical differences between the basalts studied by Carson and the diabase found at the site. Taken together, however, the evidence of a state of equilibrium between erosion and weathering, the impermeability of the diabase, the depth of oxidation, and the rate of speroidal weathering clearly suggest extremely great age for the undisturbed saprolite profile in diabase at the fault.Undisturbed secondary clay infilling crosses the fault in weathered diabase at East Dike 2. Undisturbed spheroidal weathering in diabase was observed in places.

3) Undisturbed Sedimentary Rocks Overlying the Fault In trench FET-19W, the fault trace underlies an uncemented sedimentary deposit, a relationship illustrated on Figure 2.5.3-14. In this exposure the trace of the fault does not offset the alluvial rock unit overlying the fault. A strongly developed soil profile was observed in the alluvium. Fault movement has not occurred since deposition of the alluvial sediments. It has not been possible to date the sediments by radiometric methods because radioactive isotopes are absent. The alluvium is oxidized throughout its entire depth; therefore any pollen or microfossils would have been leached out. Because of the lithologic character, distance from the Cape Fear River, and the compactness of these deposits resting unconformably on Triassic bedrock traversed by the fault, it is concluded that these sediments should be included in the group mapped and described by Reinemund (Reference 2.5.1-17) as "high level surficial deposits." The material is a dense, compact clayey sand containing small lenses of clean, medium to coarse angular quartz sand. At the base, a thin layer of rounded quartz gravel incorporates material from the underlying Triassic sedimentary rocks. Reinemund (Reference 2.5.1-17) summarized the age relations of these deposits as follows:

 "Within the mapped area, the evidence indicates only that the deposits are post-Triassic. Observations made a few miles east of the Deep River field, by Stephenson, indicate that some of the materials are post-Cretaceous, and the deposits as a whole have generally been regarded as Tertiary, possibly Pliocene. It Amendment 65 Page 167 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 is possible, however, that these deposits included materials as old as Cretaceous and as young as Pleistocene."At minimum distance, these deposits are over five miles from Cape Fear River; Quaternary deposits are markedly closer. Reinemund describes Quarternary terrace deposits as increasing in age with elevation and distance from the river, which is reasonable, and remarks that the higher of these terraces may be prePleistocene.The material overlying the fault at trench FET-19W is uniformly graded sand, markedly different from Reinemund's Quarternary terrace deposits which were found in field investigations to be grossly gap graded, coarse gravels, cobbles and small boulders in a silt-clay matrix. The Quaternary terrace deposits appear to have been formed in low-energy environments in which letdown of coarser clasts from older, higher terraces tool place into the silt-clay regime of deposition.The soil profile developed in the undisturbed sedimentary rock unit in trench FET-19W is approximately 6 ft. thick, a thickness consistent with that of the soil profile of similar quartz sands, known to be Pliocene, in middle Coastal Plain sediments.This correlation is further evidence that the deposit at trench FET-19W is older than one million years.Since the deposit seen in trench FET-19W is at approximate Elevation 240 ft., and there are other outcrops of these deposits at elevations above 600 ft., it may be presumed that the deposits, as described by Reinemund, were some 400 ft. thick in this area, and have subsequently been eroded away, an interpretation confirmed by John Reinemund during this examination of the outcrop in December, 1974.Reinemund interprets these materials as probably Tertiary but agrees they could be as old as Jurassic. These deposits may represent updip outcrops of strata identified as Cretaceous by Conley (Reference 2.5.1-20) and Swift and Heron (Reference 2.5.1-21). They could represent a remnant of the Jurassic burial episode described in Section 2.5.1.1.5.Since the deposit is perched some 25 ft. above the Holocene stream channel to which its surface drains, and is distinctly different than the gray sand silts of the present stream channel, it cannot be of Holocene age.c) Evidence from General Geology The fault is in a geologic setting of about 150 to 250 million years B.P. Faults displacing strata younger than Paleocene were neither found during this study nor previously in the Carolina Piedmont-Coastal Plain. Therefore, by association, the fault is considered a very ancient feature in a historically and geologically aseismic area. Ancient, currently stable faults in the site region do not offset present topographic features or drainage patterns. The topographic relief is subdued.Conley (Reference 2.5.1-20) states that cross faults can be traced into this Triassic basin and that these faults have displaced Paleozoic Slate Belt rocks as much as a mile along strike but have displaced the Triassic sedimentary rocks only a few hundred feet.Amendment 65 Page 168 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 This indicates that major movement on the faults occurred before the Triassic, with only minor subsequent movement.There has been no movement on the Jonesboro Fault, one of the great faults crossing North Carolina, since deposition of Cretaceous marine sediments over it. Major faults of this type, as well as minor faults, sometimes undergo recurrent movement, with a different sense of movement than before, when stresses in the earth's crust trigger macroseismic activity. Recurrent movement has not occurred since the pre-Cretaceous on the Jonesboro Fault or on the cross faults described by Conley. After minor deformation following deposition, about 150 million to 250 million years B.P., the Triassic basin must have been quite stable, because sedimentary beds are not closely folded.From the geologic setting and additional points developed in other sections of this report, it is concluded that the site fault moved late in the history of the other areal/regional faults, is millions of years old, and is a relict feature, inactive and not capable.d) Evidence from Geometry and Field Observations The general spatial relationship of the dikes, site fault and plant as discussed below is shown on Figures 2.5.1-15 and 2.5.1-16.

1) East Dike 2 From evidence throughout the hard rock exposure developed by excavation on composite East Dike 2, faulting occurred after crystallization of the youngest of the three dike segments. At the dike-fault intersection, however, there are marked differences in the apparent offset among the three dike bodies constituting what is called East Dike 2, as shown on Figures 2.5.3-9 and 2.5.3-10.

The East Dike 2 exposure supports the following important interpretations:a) Sedimentary beds are not offset across the three individual dike bodies, as shown on Figure 2.5.3-9. Therefore, the dike does not follow a discernible fault.The three bodies were separately intruded and crystallized separately.b) The vertical component of movement along the fault after intrusion of the youngest dike segment must have been small, at most a few tens of feet.c) The horizontal component of movement after intrusion of the youngest dike segment also must have been small, possibly a foot or less.d) The differences in horizontal offset of the individual dike bodies reflect their sequential intrusion during the period of movement on the fault, as seen in the table below, from Figure 2.5.3 9:HORIZONTAL OFFSETS IN FEET DIKE SEGMENT Elevation Youngest/Western Central Oldest/Eastern 247 1 - 3 1 - 3 6 244.8 2 - 3.5 4 - 6 5-6 243.1 3 3 - 4 6 Amendment 65 Page 169 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 238.5 ? 5 - 6 8 234.6 2 - 6 6 - 7 4-8 218.5 6 - 7 6 - 7 10 210.3 5 8 - 9 10 - 13 Therefore, in spite of the overall vertically sinuous nature of the dike segments, the paleomagnetically oldest eastern dike segment is clearly displaced horizontally, and probably vertically through a depth range of 37 ft., more than the younger western segment. The horizontal relationship is demonstrated on Figure 2.5.3-10, where the two easternmost dike segments are shown offset and deformed, with left lateral sense, by a splay in the fault, while the western segment is not affected, strongly suggesting that the splintering movement took place prior to intrusion of the western segment.Sequential intrusion of dike segments during faulting proves the contemporaneous nature of movement on the fault with intrusion of the dikes, which are known radiometrically and paleomagnetically to be more than 150 million years old. These conditions also offer supporting evidence that the site fault is a minor, late contemporary feature to the Jonesboro Fault. Last movement on the Jonesboro Fault was before deposition of Cretaceous sediments that are present south of the plant site. The diabase dike identified as Location 5 in Appendix M of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29) may have been intruded during the last phase of movement on the Jonesboro Fault.

2) West Dike 3S West Dike 3S intersects West Dike 3 at a point several hundred feet north of their intersection with the fault (See Figure 2.5.3-15). West Dike 3S is younger than West Dike 3. Deeper excavation revealed a second intersection of the dikes, about 8 ft. lower than the first. At the lower exposure, the smaller West Dike 3S intrudes but does not cross the larger West Dike 3, proving that the smaller dike is younger. The outcrop pattern of the dike intersection is unequivocal. Textural differences in the two dikes are distinct. Paleomagnetic sample PM 7 from West Dike 3, at the intersection between the two dikes, shows a reset of remanent magnetization in the larger dike, also providing further confirmation that West Dike 3S is the younger (Figure 2.5.3-11).

Sedimentary beds on either side of West Dike 3S are not offset. At Elevation 249 ft., the outcrop of the intersection of West Dike 3S and the fault is such that West Dike 3S, if any apparent offset can be detected, is only a few in. out of line. This is seen on Figure 2.5.3-11. Subsequent excavation to Elevation 233 ft. revealed relative change in location of all key elements of the intersection. Analysis of these changes indicates that they could not have been produced by fault offset if the dike were as planar at the fault as elsewhere.The exposure of the intersection of West Dike 3S and the fault at Elevation 239 ft. offers strong evidence that West Dike 3S was intruded after movement on the fault (Figure 2.5.3-15). As observed and described, the dike is straight everywhere but at the fault, where, however, it would be expected to have been straight originally if intruded before faulting. The irregular aspect of the dike at this exposure is apparently related to intrusion, not faulting, because the adjacent sedimentary bedding appears to be less disturbed than would be expected if faulting had caused the existing dike configuration Amendment 65 Page 170 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 (Figure 2.5.3-15). Furthermore, as can be seen on Figures 2.5.3-11 and 2.5.3-16, diabase of the West Dike 3S overlaps the sheared clay of the fault zone; thus, two apophyses at the end of the northern limb of this exposure appear to offer compelling evidence that the diabase was intruded after faulting. There is a smaller, football-shaped apophysis immediately west of the stub end of the southern limb of West Dike 3S. Zeolite samples were obtained from this apophysis and adjacent fault gouge; a detailed portion of this area is shown diagramatically on Figure 2.5.3-17.Where apophyses enclose country rock against the main body of the dike, the intensity of alteration in the surrounded rock is greater than in other immediately adjacent areas.This is consistent with evidence that alteration at this point was proceeding from two sides, from both the diabase dike and the apophyses.Clay mineralogy studies of gouge material at the stub ends of the dike at Elevation 239 ft. did not reveal characteristic changes expected from thermal alteration. The ends of the dike so overlap the clay gouge that a zone of sheared clay only one fourth to one inch thick is exposed.At lower excavation elevations, burned and baked typical fish-scale clay gouge is found in place at the stub ends of the dike segments and extends as a band between the segments. During the final stages of crystallization and initial cooling of the dike, some very slight movement must have occurred on the fault, sweeping thermally altered material off the most protrudent incursions of the dike into the gouge. In contrast, if the main movement along the fault had occurred after dike emplacement, the thermally altered material would have been wiped from the ends of the dike at all points and chaotically rearranged within the fault gouge. In East Dike 2,, a similar intrusion, not associated with faulting, is shown on Photo V of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29). This illustrates the precise structural control of the dike emplacement.Figure 2.5.3-18 represents possible solutions to fault movements, focusing on the West Dike 3S-site fault intersection. West Dike 3S is shown at true dip (about 80°) to illustrate what happens during various fault movements. From general aspect, either Panel 3 or 4 could be interpreted to represent the movement of the fault on the basis of the change in dip of the diabase, as shown in Panel 3, or on the basis of the alignment of the two ends, as shown in Panel 4. However, these solutions are ruled out, because the dip of the diabase remains the same both north and south of the fault intersection.

3) West Dike 3 West Dike 3, which was not excavated to depths sufficient to expose hard rock at the fault intersection, exhibited about 10 ft. of apparent left-lateral offset, as shown on Figure 2.5.3-15.

4) West Dike 4 This dike was not excavated, because the fault crossing underlies a trunk rail line.

5) West Dike 5 Amendment 65 Page 171 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 West Dike 5 is unique in having pronounced, probably superficial, westerly dip north of the fault and right-lateral apparent offset. It was not excavated to depths sufficient to expose hard rock at the fault intersection. The anomalous, near-surface westerly dip of this dike may be the result of the intrusion into broken ground adjacent to the fault.e) Evidence on Dike Intrusion Sequence from Petrography and Chemistry Petrographic and chemical work on the dikes provided a critical test for field observations concerning relative ages of dikes and site fault. Field evidence established that composite East Dike 2 was intruded during fault movement; West Dike 3S was intruded after most movement had occurred, but before a final, minor element of movement; and West Dike 3 (Figure 2.5.3-13) was probably intruded early during movement and offset about 10 ft. left laterally.At first inspection, the chemical variation diagrams (Figures 1 and 2 of Appendix H of the Shearon Harris Fault Investigation Report, Reference 2.5.1-29) contradict field observations concerning the relative ages of West Dikes 3 and 3S. West Dike 3S comprises the more mafic rocks and would be expected to be older. However, the generalization that the more mafic rocks are intruded first only holds in the case of non-porphyritic rocks, i.e., when 100-percent magma rather than a "crystal mush" is intruded and the chemical trend on the variation diagram is a true "liquid line of descent." West Dike 3 was intruded as a nearly 100-percent magma, while West Dike 3S was intruded as a "crystal mush." Both were derived from the same crystallizing magma chamber (comagmatic), which was tapped early, before many crystals had formed; West Dike 3 was the result. A subsequent tap of the chamber, after many crystals (phenocrysts now) had formed, produced West Dike 3S.A test of this hypothesis would be whether the groundmass of West Dike 3S is less mafic than that of West Dike 3. Unfortunately, a chemical analysis of the West Dike 3S groundmass is not available and would be difficult to determine accurately. Modal analysis would not suffice, as the critical factor is the composition of individual minerals in the groundmass. The only mineral composition readily determined without a microprobe is that of plagioclase; fortunately, plagioclase alone can provide an adequate test of the hypothesis.Comparative results for plagioclases in West Dike 3 with those in West Dike 3S are given in Table 2 of Appendix H of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29). West Dike 3 plagioclases are anorthite-rich, thus agreeing with field observations that West Dike 3 was intruded first.Other reasoning can argue that West Dike 3 was intruded before West Dike 3S, which is more altered deuterically than West Dike 3, contains more volatiles, and contains hydrous silicate-filled amygdules. Isotope data on Sr87/Sr86 for the dikes and associated zeolites indicate that the water necessary to form the zeolites came from the diabasic magmas, rather than from groundwater or sediment pore water. This water presumably also altered the dikes deuterically. The magma must have been supersaturated with respect to the fluid phase when the dike crystallized, as indicated by presence of amygdules. During crystal fractionation, volatiles are enriched in the magma so that late-stage differentiates are more volatile-rich than early differentiates. This agrees with Amendment 65 Page 172 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 a younger age than West Dike 3 for the more volatile-rich and more highly altered West Dike 3S.In conclusion, chemical and petrographic data cannot conclusively determine the relative age of East Dike 2 with respect to the other two dikes. However, they substantiate the field observations that West Dike 3S is younger than West Dike 3. This subject is more fully discussed in Appendix H of the Shearon Harris Fault Investigation Report (Reference 2.5.1-29).2.5.3.2.2.2 The Fault as an Oriented Plane of Weakness Brown, Miller, and Swain (Reference 2.5.1-42), after their exhaustive work on the structural architecture of the region, concluded that the area including the plant site is under east-west horizontal compressive force. They do not associate the secondary level and orientation of such stress with movement along faults aligned east-west in the contemporary stress field.Their conclusions confirm present stability of the site fault.Observations in mines of the Deep River Coal Field, where mine openings in both sedimentary rocks and diabase dikes extend to depths of 800 ft. below the surface (and 600 ft. below sea level) have not revealed over-stressed rock. Timbered supports in one of these mines have been considered unnecessary for long-range stability, by the mine manager. Observation of representative room and pillar openings after supports were removed at the conclusion of mining indicated that large spans were self-supporting (Reference 2.5.3-4).These reported conditions, in contrast to other areas where pop-outs and rock bursts are common occurrences at depths as little as 60 ft. below the surface because of excessive stored stresses in the rocks, indicate that the Triassic rocks of this area, including that of the site fault, are in an at-rest, stable condition.Kiersch (Reference 2.5.1-36) reports overpressurized deep groundwater in Triassic sedimentary rocks of the Dunbarton Basin, South Carolina-Georgia. This should be a very sensitive indicator of stress levels in these impermeable rock types, compartmentalized as they are by impervious diabase dikes. Occurrence of such overpressurized groundwater in the Durham basin, where the plant site is located, has not been reported in data covering some 450 wells in the Durham Basin. This also leads to the conclusion that these rocks are in an at-rest, stable condition.Talwani (Reference 2.5.2-12) has observed cases of reservoir induced seismicity (RIS) in the piedmont of South Carolina. Because of this observation, the pre-existing state of stress, the magnitude of the induced stress, and the geologic and hydrologic setting of the Shearon Harris Nuclear Power Plant have been examined to seek out any possible correlations that might exist with the RIS sites in the South Carolina Piedmont. This study is discussed in detail in Section 2.5.2.1.2. The comparison of the Shearon Harris site reservoirs to the four cases of reservoir induced seismicity in South Carolina shows no correlation in pre-existing stress, water depth, water volume, geologic or hydrologic conditions. Therefore, there is no reason to expect the SHNPP site reservoirs to induce any significant seismic activity.Amendment 65 Page 173 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.3.3 Earthquake Associated with Capable Faults A letter to Carolina Power & Light Company from the NRC (Reference 2.5.3-5), concurs with the Shearon Harris Fault Investigation Report (Reference 2.5.1-29) in concluding that the site fault is not a capable fault as defined in Appendix A to 10 CFR Part 100.In summary, there are no capable faults that could affect the site and therefore this section is not applicable to SHNPP.2.5.3.4 Investigation of Capable Faults For the reasons given in Section 2.5.3.3, this section is not applicable to SHNPP.2.5.3.5 Correlations of Epicenters with Capable Faults For the reasons given in Section 2.5.3.3, this section is not applicable to SHNPP.2.5.3.6 Descriptions of Capable Faults For the reasons given in Section 2.5.3.3, this section is not applicable to SHNPP.2.5.3.7 Zones Requiring Detailed Faulting Investigation For the reasons given in Section 2.5.3.3, this section is not applicable to SHNPP.2.5.3.8 Results of Faulting Investigations For the reasons given in Section 2.5.3.3, this section is not applicable to SHNPP.2.5.4 STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS 2.5.4.1 Geologic Features The plant is founded on well-consolidated Triassic sandstone, siltstone, shale, and claystone.The individual beds are generally lenticular; hence, rock types change abruptly horizontally and vertically. No areas of active or potential surface or subsurface subsidence, uplift, or collapse have been found; none are thought likely to be present.Geologic evidence suggests that a considerable thickness of Triassic rocks has been eroded from the site; it is probable that some unconsolidated sediments of Cretaceous and younger age have also been deposited and eroded from the site. Most erosion probably occurred during the Mesozoic Era. Evidence suggests that erosion over the last few million years has been relatively slight.Generally, groundwater in the site area is sparse, as the rocks have very low primary permeability. Water does occur in joints, fractures and bedding planes, however, and reaches useable quantities near some diabase dikes.The depth of weathering in the bedrock varies essentially from 5 to 10 ft., depending upon the type of underlying rock. Local exceptions to this were found along the eastern portions of Amendment 65 Page 174 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Trenches 2 and 3 (for locations see Figure 2.5.1-11), where residual soils are from 0 to 4 ft.thick, and over diabase dike materials, where weathering was observed to a depth of 15 ft. or more. Other zones of irregular weathering and alteration have not been found.Joints are somewhat irregularly spaced in the site area, most of them at intervals of a few feet and steeply dipping. The predominant trend direction is approximately northeast. A few joints trend approximately north-northwest; north-south trending joints are uncommon. Some joints parallel bedding. High angle joints in other directions and joints with moderate dips occur sporadically. Joints observed adjacent to diabase dikes are generally perpendicular to the strike of the dike, and extend 2 to 10 ft. laterally from the dike. When rock is weathered, light gray clay forms in the joints and fractures. Joints and other geologic features at the plant, the Main Dam and the Auxiliary Dam were mapped at a scale of 1 inch = 10 ft. (1 inch = 50 ft. outside of the core trench of the Main Dam). The maps are included in the Foundation Report which is Appendix 2.5E.Low-angle (0 degrees to 45 degrees to the horizontal) and high-angle (46 degrees to 90 degrees to the horizontal) fractures were found at various elevations in the P and D series test borings shown in Appendix 2.5A. Fractures were most common in test borings P16-B and P17.Several borings penetrated thin layers of gray or green clay that occupied fractures in claystone strata.Many fracture surfaces in the weathered rock were slickensided, but slickensided fracture surfaces were generally absent in the joints and fractures in the fresh rock forming the plant foundation; they are not thought to be related to tectonic activity. During excavation, a fault was discovered in the foundation of the Waste Processing Building; studies of this noncapable feature are reported i Section 2.5.3 and in the Shearon Harris Fault Investigation Report (Reference 2.5.1-29). No other folds, faults, shears, or other zones of structural weakness were noted at the plant excavation.As noted in Section 2.5.1.2.3.3, there is no evidence of residual stresses in the bedrock.The siltstones and a few sandstones in the plant area are subject to spalling if repeatedly wet and dried over a period of weeks. To minimize this problem, once the last phase of excavation cleanup was begun, the rocks were protected against inclement weather. Final cleanup used only air lances; water was not applied. After rock was cleared, it was moistened and the 4 in.thick concrete protective mat was applied, followed by the design concrete seal mat. In some areas, depending on the construction schedule, the design seal mat was placed after final cleanup without the 4 in. protective mat. No other potentially hazardous rock or soil conditions were noted.Geologic conditions of the site are discussed fully in Section 2.5.1.2.5.4.2 Properties of Subsurface Materials 2.5.4.2.1 General The site is underlain by clastic sediment, principally shale, claystone, siltstone, and sandstone.These materials are characterized by changes in composition and texture, both horizontally and vertically. The beds vary in thickness from a few inches to a maximum of 15 to 20 ft.Amendment 65 Page 175 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Notwithstanding these variations in composition and texture, the beds and lenses interfinger and overlap into compact masses that show no structural weakness.All safety related plant structures are founded on this compacted mass of clastic sediments.The plant is built on sound rock and plant grade is established at Elevation 260 ft. Plant structure foundations utilize reinforced concrete mats and spread footings to spread loads to an acceptable bearing value determined by laboratory tests. A discussion of dams, dikes, spillways, and channels, and the related properties of the subsurface materials at these locations, is presented in Section 2.5.6.Test results for subsurface materials, in addition to those presented in this section, are included in Appendix 2.5B.2.5.4.2.2 Static Properties Static properties that were determined included index properties, compressive strength, and deformation properties.2.5.4.2.2.1 Index Properties 2.5.4.2.2.1.1 Dry Density Dry density determinations were performed on selected rock samples and are summarized in Table 2.5.4-1. The average dry density for the rock material is 162.8 pounds per cubic foot.2.5.4.2.2.1.2 Grain Size Analysis Grain size distribution (GSD) curves for the residual soil samples are presented on Figures 2.5.4-1 through 2.5.4-53. Tests were performed in accordance with ASTM D-422.2.5.4.2.2.1.3 Core Recovery Core recovery is defined as the length of core recovered divided by the length of the core run, expressed as a percentage. Core recovery values were computed for all rock cores, and results are presented in logs of borings in Appendix 2.5A.2.5.4.2.2.1.4 Rock Quality Designation Rock quality designation (RQD) is defined as the total length of sound rock core pieces, 4 in.long and over, divided by the length of the core run and expressed as a percentage (Reference 2.5.4-1). In cases of cores broken by mechanical fractures during drilling or handling (i. e. the fracture surfaces were fresh, irregular breaks rather than natural joint surfaces), the fresh broken pieces were fitted together and were considered as one piece, provided they formed the requisite length of 4 in. RQD values were computed for all cores and results are presented in the logs of borings in Appendix 2.5A. The average RQD value below Elevation 235 ft. for each BP boring in the plant foundation area is shown in Table 2.5.4-2. The average RQD value below Elevation 235 ft. in the foundation area is 92.7 percent.Amendment 65 Page 176 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.2.2.2 Strength Determinations 2.5.4.2.2.2.1 Unconfined Compression Test The compressive strength of selected rock core samples was determined by unconfined compression tests. Tests were performed in general accordance with ASTM D-3148. The average unconfined compressive strength for the rock material is 8148 psi. Individual test results are presented in Table 2.5.4-3. Stress-strain curves, developed by utilizing the test results, are presented on Figures 2.5.4-54 through 2.5.4-65.2.5.4.2.2.2.2 Triaxial Compression Test Triaxial compression tests were performed on selected rock core samples. The test results are summarized in Table 2.5.4-4. Individual stress-strain curves based on the triaxial compression tests are presented on Figure 2.5.4-66. A Mohr's circle plot, developed by utilizing both the unconfined and triaxial compression test results, is presented on Figure 2.5.4-67. The rock samples utilized in the triaxial compression tests exhibited an average friction angle () of 45 degrees and cohesion (C) of 2400 psi based on the Mohr envelope shown on Figure 2.5.4-67.2.5.4.2.2.3 Deformation Properties 2.5.4.2.2.3.1 Poisson's Ratio Poisson's ratio () was computed, as discussed in Section 2.5.4.7, using shear wave (Vs) and compressional wave (Vp) velocities determined from field geophysical data. The shear wave and compressional wave velocities of the subsurface materials were seismically determined, as discussed in Sections 2.5.2.5 and 2.5.4.4. Values computed for Poisson's ratio based on wave velocity measurements results are summarized in Table 2.5.2-3.In addition, values of Poisson's ratio were determined from laboratory unconfined compression testing in which both axial and radial strains were measured. The results of the axial and radial strain measurements, and the computed Poisson's ratio values from laboratory testing, are presented on Figures 2.5.4-68 through 2.5.4-91 and summarized in Table 2.5.4-5.The results computed for Poisson's ratio, especially those based on field geophysical data, are in good agreement with typical values for sandstones and siltstones available in the literature (Reference 2.5.4-2).2.5.4.2.2.3.2 Static Modulus The static modulus of deformation was computed from the unconfined compression tests.Values determined for static modulus are summarized in Table 2.5.4-3. These values are then reduced by a factor dependent on rock quality designation, RQD (Reference 2.5.4-1), as discussed in Section 2.5.4.7.2.5.4.2.3 Dynamic Properties The dynamic properties of the rock material were investigated through field and laboratory measurements of compressional and shear wave velocities. The dynamic modulus was then computed from the wave velocities, as discussed in Section 2.5.4.7.Amendment 65 Page 177 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.2.3.1 Field Geophysical Measurements 2.5.4.2.3.1.1 Compressional Wave Velocity (Vp)The compressional wave velocity (Vp) was determined through geophysical measurements made during seismic refraction surveys. A downhole velocity survey was performed to provide a check on the compressional wave velocity measurements obtained during the seismic refraction surveys. Discussion of the seismic survey program for geophysical exploration are presented in detail in sections 2.5.2.5 and 2.5.4.4. The test results are summarized in Table 2.5.2-3.2.5.4.2.3.1.2 ShearWave Velocity (Vs)Shear wave velocity (Vs) measurements were determined by using Sprengnether Engineering Seismographs. Details of the tests are discussed in Section 2.5.4.4. The results of these tests for each stratum are summarized in Table 2.5.2-3.2.5.4.2.3.2 Laboratory Geophysical Measurements The shockscope, an instrument developed by Dames & Moore (Reference 2.5.4-3) to measure the velocity of cmpressional wave propagation in laboratory samples, was utilized in measuring the compressional wave velocity of selected rock core samples. In the shockscope tests, samples are subjected to a physical shock and the time necessary for the shock wave to travel the length of the sample was measured using an oscilloscope. The velocity of compressional wave propagation was then computed. The compressional wave velocity (Vp) measurements based on the shockscope test are presented in Table 2.5.4-6. The velocity measured in the shockscope is used for correlation with field velocity measurements obtained in geophysical refraction and uphole surveys. Additional testing to determine the compressional wave velocities (Vp) of rock samples was conducted by Law Engineering in general accordance with ASTM D-2845. The results are also presented in Table 2.5.4-1.2.5.4.3 Exploration 2.5.4.3.1 General Field investigations, at locations shown on Figures 2.5.1-11 through 2.5.1-16, were performed to evaluate the engineering geologic and seismologic characteristics of the site. Field exploration consisted of: 1) an engineering geologic survey of the site and surrounding areas, 2) a preliminary test boring program, borings for design of plant structures and additional borings for the site fault investigation, 3) a trench excavation program, also expanded for the fault investigation, and 4) geologic mapping of plant foundations. The exploration also included installation and monitoring of wells and piezometers. Discussions of the installation and monitoring of wells and piezometers are presented in Sections 2.5.4.6 and 2.5.4.13.The initial exploration program was performed by or conducted under the technical direction of Dames & Moore geologists and engineers. Moore, Gardner and Associates, Incorporated of Asheboro, North Carolina made surveys to establish horizontal and vertical controls at the site.Law Engineering Testing Company completed borings for design of structures and did geotechnical testing of materials as required. Ebasco Services, Incorporated geologists carried out the site fault investigation and mapped the plant foundation for Carolina Power & Light Amendment 65 Page 178 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Company (CP&L) during construction, and CP&L personnel performed the magnetometer traverses.2.5.4.3.2 Engineering Geologic Survey A comprehensive survey was conducted to identify the engineering geologic characteristics of the site and surrounding area. It included detailed inspections of 1) rock cores from test borings, 2) surface features, 3) exposed road cuts, 4) excavated trenches, 5) bedrock outcrops, and 6) a Brunton Compass survey. Geologic maps, literature, gravity-survey data, aerial photographs, and topographic maps were examined. Representatives of local and state agencies, universities, and private organizations were interviewed to obtain engineering geologic data.2.5.4.3.3 Geologic Borings Borings were drilled to investigate the surface soils, underlying sedimentary bedrock and the several diabase dikes in the investigation area. Preliminary borings of the P and D series (Figures 2.5.1-11) explored the plant and Auxiliary Dam vicinity, which were followed by closely spaced borings for foundation design of specific plant structures. Boring locations are shown on Figures 2.5.1-12 through 2.5.1-14. During construction additional borings were completed for the site fault investigation at locations shown on Figures 2.5.1-15 and 2.5.1-16.Project borings, including those for dams, bridges, and other structures totaled about 1250 and are tabulated by purpose with coordinates and depths in Appendix 2.5A.As indicated in the tabulation, most logs of the borings in the plant vicinity are also included in the Appendix. Logs for some proposed but abandoned structures such as the East Auxiliary Dam and for structures which are not Seismic Category I, such as the Cooling Tower, are not included in the FSAR, but are maintained in CP&L files. Logs for the fault investigation are shown on Section 2.5.3 figures.Truck-mounted and skid-mounted rotary wash, wire-line drilling rigs were employed in the field programs. Drillers used a 3 7/8 in. diameter rotary core barrel to penetrate and sample the soils at each test boring location. When rock was encountered, a 2 1/8 in. double-tube core barrel and diamond drilling bit were used to advance the borings and collect continuous samples of bedrock.A standard two in. outside diameter split-barrel sampler with an inside diameter of one and three-eighths in. was used to obtain soil samples for soil classification and for laboratory tests.In soils, the standard penetration test was made at every change of strata; and within strata, at intervals not exceeding five ft. For the fault investigation, some of the borings were sampled continuously using the standard penetration test.Cased borings of sufficient size were made to accommodate either a three in. diameter thin-wall tube (Shelby Tube) sampler, a three in. diameter piston sampler, or a Denison-type, double-tube core barrel in order to obtain relatively undisturbed soil samples. Twenty-five pound bag samples of soil were obtained from uncased auger borings. The rock drilling program produced a rock core sample approximately two in. in diameter. Rock cores were examined in the field, logged, and stored in standard core boxes for future use and testing.Amendment 65 Page 179 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 All borings were logged in detail by geologists. The soils encountered were described in accordance with the Unified Soil Classification System. RQD values (Reference 2.5.4-1),calculated for each core length and indicated on the boring logs, indicate the ratio of core, four or more inches in length, to the full core run.2.5.4.3.4 Trench Excavation Program More than 12,000 ft. of trenching was performed at the site during the original site studies to supplement the information obtained from the borings. Additional trenches were excavated during the fault investigation.Locations of the four major trenches are shown on Figures 2.5.1-11, 2.5.1-15, and 2.5.1-16.Portions of Trenches 1 and 2 are adjacent to the plant site; trenches 3 and 4 are on the Auxiliary Dam alignment. Trenches were generally excavated to refusal of the backhoe equipment. Most trenches were 2 to 10 ft. deep; however, those in residual soil derived from diabase dikes were advanced to the full limit of the excavating equipment, 15 ft., without meeting refusal. Trenches for the fault investigation were excavated by backhoes and bulldozers to depths of 5-40 ft. One trench for the fault investigation required blasting.All excavations were inspected in regular increments to evaluate lithology, quality, and continuity of the bedrock as well as overburden composition and consistency. Trench logs prepared for the original site studies are provided in Appendix 2.5A. Hand-penetrometer tests were made at intermittent intervals.Layouts of trenches 1 and 2 were planned with regard to the regional geology. Trench 2 was normal to possible cross-faults, whereas Trench 1 was normal to the regional structural trend.Thus, trenches with these orientations would traverse any longitudinal and cross-faults that exist in the investigation area. Geological cross-sections of the trench excavations are presented in Figures 2.5.4-92 through 2.5.4-98.Borings P9 through P18 and D11 through D18 (Figure 2.5.1-11) were drilled in the stream valleys which crossed Trenches 1 and 3 as a substitute for trenching. The loose and deeper overburden soils in these areas made excavation and data acquisition impossible by trenching operations alone.2.5.4.3.5 Geologic Mapping The plant excavation has been mapped at a scale of 1 in. = 10 ft. The foundation rock was found to be sound and competent. No features inimical to the safety of the plant were noted.Full discussion and geologic maps of each surface of the foundation are presented in the Foundation Report (Appendix 2.5E). Logs of all borings and trenches are presented in Appendix 2.5A.2.5.4.4 Geophysical Surveys The following geophysical surveys were conducted at the site:a) Surveys to determine seismic wave transmission characteristics of the site (discussed in Section 2.5.2.5).Amendment 65 Page 180 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 b) Magnetometer traverses (discussed in Section 2.5.3.2.1).c) Microearthquake monitoring in the site area (discussed in Section 2.5.2.1.3).2.5.4.5 Excavation and Backfill Rough excavation for the Shearon Harris plant structures was completed in 1974 as shown on Figure 2.5.3-2. Excavation activities were resumed in 1977 and continued to completion in 1979.A topographic map of the ground surface prior to excavation is shown on Figure 2.5.1-15. A plot plan and also a section of the excavation, as well as a plan showing the location of backfilled areas, are shown on figures presented in Appendix 2.5E. Additional geologic maps and sections of the excavation are included in the Foundation Report for the plant presented in Appendix 2.5E. The excavations for the Main Dam, Auxiliary Dam, Auxiliary Reservoir Channel, Emergency Service Water Intake Channel, and Emergency Service Water Discharge Channel are described in Section 2.5.6.The plant excavation included the foundations for the Waste Processing Building, four Turbine Buildings, a Fuel-Handling Building, four Reactor Auxiliary Buildings, four Containment Buildings, and four Tank Buildings. It had a maximum lateral extent of approximately 925 ft. in the north-south direction and 990 ft. in the east-west direction, and encompassed a total area of approximately 837,500 sq. ft. Floor elevations of the various levels within the excavation ranged from Elevation 234 ft. for the shallowest part to Elevation 179 ft. for the deepest level. The total volume of the excavation was approximately 1,200,000 cubic yards. After cancellation of Units 3 and 4 the area west of the Fuel Handling Building has been backfilled to plant grade. The backfill is supported by a retaining wall to isolate the Fuel Handling Building from the backfill (see Section 2.5.4.5.3). After cancellation of Unit 2, the Turbine, Reactor Auxiliary Building and Containment 2 areas have been backfilled to Elevation 261 ft., Elevation 242 ft. and Elevation 236 ft., respectively. This backfill is supported by the retaining wall to isolate the Fuel Handling Building.All excavation and backfill were performed in accordance with Ebasco Specification No. CAR-SH-CH-8, which is included in Appendix 2.5I.2.5.4.5.1 Excavation Plant excavation began with the leveling of the ground surface in the plant area to Elevation 260 ft. Depressions in the surface were brought up to grade with random fill consisting of local Triassic sedimentary rock material and residual soils. After leveling, unclassified soil materials were excavated to ripper refusal by common excavation using bulldozers and scrapers. The excavation was completed to final grade by controlled blasting in successive lifts and by using power shovels and backhoes to load the broken rock onto trucks for removal up an access ramp cut into the south end of the excavation.Slopes were excavated through the overburden at 1:1. Slopes in bedrock were excavated at 1:4. Rock slopes were shaped by presplitting in accordance with specifications which did not permit heavy blasting closer than 3 ft. to the rock which forms the final foundation of concrete structures. Presplitting did not completely preserve the rock surface of the slopes because adverse jointing across bedding planes which dipped toward the excavation, as in the west wall Amendment 65 Page 181 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 of the excavation, allowed small blocks to slide down bedding planes toward the excavation.This situation was somewhat aggravated during the period from late 1974 to early 1977 when excavation was halted. All loose blocks of rock were removed when excavation work was resumed.Weathering of the sandstones exposed in the excavation during the period of project delay was negligible. Weathering of the finer-grained rocks, caused by wetting of clays contained in these rocks, did not exceed several inches below excavation levels established in 1974. The coating provided by the uppermost disintegrated rock tended to protect the underlying material from further rapid advance of the weathering process.However, weathering along natural joints, fractures, and bedding planes as in the fine-grained rocks necessitated some additional excavation of foundation areas to assure adequate intact condition of the rock prior to placement of the concrete seal coat. The excavated rock was replaced by equivalent concrete so that the stability of structures founded on the rock was not affected.Treatment methods used for foundation protection after excavation included slush grouting, shotcreting, and placement of drain pipes as well as placement of seal coating. Following final excavation, joints were filled with slush grout after which the invert rock was protected with Class F mortar prior to placement of the concrete seal mat over completed foundation areas. In certain instances the foundation mat structural concrete was poured directly against the rock surface after treatment with slush grout and Class F mortar. Also, in Units 2, 3 and 4 (subsequently cancelled) a vertical mat was formed and poured with concrete. Drain pipes were placed through the shotcrete at the locations of seeps to prevent the buildup of excessive pressure.A geologist was resident at the site throughout the cleanup of the excavation to assure that foundations were prepared as designed and to document foundation conditions by geologic mapping. Following inspection and approval by the geologist, loose materials were removed from excavated surfaces by backhoe and hand tools, and the surface was blown clean with compressed air. A geologic map was then made of the rock surfaces at a scale of 1 in. = 10 ft.,as documented in the Final Foundation Report for the plant site. During construction the geologist inspected all foundation surfaces and assured that any defects were treated prior to recommending approval.2.5.4.5.2 Dewatering A permanent dewatering system is not utilized for the plant. Groundwater seepage into the excavation was minimal because of the low permeability of the rock. Most of the inflow was due to rainfall, although minor seepage of groundwater did occur along some joints and bedding planes. Drainage was accomplished by the intermittent use of sump pumps. This drainage procedure had no adverse effect on the quality and condition of the foundation.2.5.4.5.3 Backfill Materials and placement requirements for backfill around Seismic Category I plant structures are discussed in Ebasco Specification CAR-SH-CH-8 "Excavation, Backfill, Filling, Grading",presented in Appendix 2.5I. The bulk of the backfill consisted of a mixture of silt and clay derived from the local Triassic sedimentary rocks. Selected backfill material was compacted Amendment 65 Page 182 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 between structures and rock surfaces to meet requirements of 95 percent standard proctor density with moisture control at +/- 4 percent of optimum moisture content and a maximum permeability value of 10 ft./yr.The fill and placement requirements used to support Seismic Category I pipes and conduits are described in Section 3.7.3.12.The sources of backfill materials were various in-progress excavations in the plant area. At completion of backfilling (later) cubic yards of fine-grained random fill, (later) cubic yards selected backfill, (later) cubic yards of crushed rock, (later) cubic yards of riprap, and (later) cubic yards of class D concrete had been placed adjacent to Category I plant structures.These volumes do not include backfill placed during construction of the intake and discharge channels, as discussed in Section 2.5.6, or random backfill placed in large exploration trenches beyond the limits of the plant excavation.2.5.4.5.3.1 Properties of Backfill Material 2.5.4.5.3.1.1 Grain Size Grain size distribution (GSD) curves for the backfill material used for the plant and channel areas are presented on Figures 2.5.4-1 through 2.5.4-53.2.5.4.5.3.1.2 Compaction Standard proctor compaction tests were performed on the backfill material to determine the optimum moisture content and maximum density required in the placement of the material. The results are presented on Figures 2.5.4-106 through 2.5.4-110.2.5.4.5.3.1.3 Triaxial Shear Consolidated undrained triaxial shear tests with pore pressure measurements were performed on selected backfill material. A Mohr-Coulomb failure envelope summarizing the undrained triaxial shear test results is presented on Figure 2.5.4-111. Individual stress-strain curves as well as Mohr's circle plots are presented on Figures 2.5.4-112, 2.5.4-113, and 2.5.4-120 through 2.5.4-129.2.5.4.6 Groundwater Conditions 2.5.4.6.1 Groundwater Conditions Relative to Stability of Safety-Related Facilities Pre-excavation boring and piezometric records in the plant island area indicate that piezometric levels range from about 240 to 272 ft. above mean sea level and that they follow the original topography. However, the completed plant block excavation reveals that groundwater occurs in the widely separated joints and fractures in the Triassic rocks and that unfractured rock materials remain dry. The higher piezometric levels represent perched conditions, as indicated by down-hole pressure tests showing several impermeable zones at various depths.Except for the west face of Fuel Handling Building the perimeter of the plant structure up to the top of the foundation mat is in direct contact with rock which is essentially impermeable. The Amendment 65 Page 183 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 portion between the plant structure and rock has been backfilled with residual soil which is of very low permeability (estimated to be less than 10 ft./yr.). Additionally, the winter 1979-1980 piezometric-level map (Figure 2.4.13-2) shows that water levels beneath the plant area are well below 251 ft.The source of surface water higher than the design basis groundwater level is the Emergency Service Water Intake Channel of the Auxiliary Reservoir, which has an operating pool elevation of 252 ft., the closest point of which comes to within about 300 ft. of the plant island. The Auxiliary Reservoir will not raise the groundwater elevation beneath the plant island above an elevation of 251 ft. for the following reasons.a) The residual soil underlying the reservoir is of very low permeability, as indicated by testing.b) After the Auxiliary Reservoir and the Main Reservoir are filled, groundwater will move from these reservoirs to the cones of depressions created by the pumpage from wells.However, the groundwater movement ultimately will be from the Auxiliary Reservoir to the Main Reservoir after construction is completed.c) Groundwater from the Auxiliary Reservoir will start moving toward the plant island at an elevation of about 252 ft. However, the water level will be at a much lower elevation than 251 ft. by the time it reaches the plant island due to the hydraulic head loss as it flows through the low permeability materials for a distance of about 300 ft.2.5.4.6.2 Design Criteria for the Control of Groundwater Levels or Collection and Control of Seepage The subsurface portions of Seismic Category I structures in the plant island are designed for hydrostatic loadings with groundwater at Elevation 251 ft. A permanent dewatering system is not utilized for the Shearon Harris Nuclear Power Plant. Groundwater occurring in widely separated joints in the rock did not significantly affect construction. Any rain or surface water accumulated during construction was pumped out by sump pumps.The design plant grade is Elevation 260 ft. for the minimum safety factor for load combinations, including the flood buoyant force. The groundwater drainage will be from the Auxiliary Reservoir (Elevation 252 ft.) and the plant site to the Main Reservoir (Elevation 220 ft.);therefore, groundwater levels will tend to remain below the design level.2.5.4.6.3 Dewatering Requirements After excavation at the site, seepage from some of the joints and fractures occurred a few days after rains, but the volume was small compared to the volume of surface-water runoff into the excavated areas. Where necessary, water was drained into sumps and pumped out of the excavated areas; continuous pumping was not required. Below the Containment Building mat, a porous concrete network drains into a permanent sump. For the Containment Building, a waterproof membrane was also placed under the mat. Consequently, a permanent dewatering system for lowering groundwater levels is not needed. Dewatering of Unit 2 Containment mat will be included as a portion of the outside area drainage.Amendment 65 Page 184 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.6.4 Groundwater Conditions During Construction Groundwater was encountered in the joints and fractures, but the rest of the rock materials were dry. Minor seepage occurred from some of the joints and fractures for a few days after rains; however, joints and fractures were slush grouted to minimize their ability to serve as water conduits. Isolated areas of continuous water seepage through weep holes in the seal mat concrete occurred in lower elevations. Otherwise, drainage into the excavated areas was limited to surface water which was collected in sumps and pumped out as required.2.5.4.6.5 Field and Laboratory Permeability Tests Down-hole water pressure tests were conducted in selected borings and the results are shown in Appendix 2.5A. In the plant site area, these tests were performed at 10 ft. intervals under pressures up to 110 psi at depths ranging from 10 to 145 ft. Several isolated zones registered very small water losses under high-pressure tests, which are recorded in Appendix 2.5A. Table 2.4.13-7 shows permeability values based on test results in these borings.2.5.4.6.6 Monitoring and Fluctuations of Groundwater Levels Tables 2.4.13-5 and 2.4.13-6 show water levels in the site wells and site piezometers, respectively, observed at weekly intervals during the winter of 1979-1980. Comparison of Figures 2.4.13-1 and 2.4.13-2 shows the difference between the preconstruction piezometer levels and the winter of 1979-1980 water levels.2.5.4.6.7 Groundwater Movement The general direction of groundwater movement in the plant area (Figure 2.4.13-2) is southeast, toward White Oak Creek. Groundwater flow, gradients, and velocities are discussed in Section 2.4.13.2.5.4.6.8 Potential for Subsidence Except for the west face of the Fuel Handling Building, the perimeter of the plant structures up to the top of the foundation mat will be in direct contact with the essentially impermeable Triassic rock. The rock surface west of the Fuel Handling Building is protected by structural concrete. The top clay/saprolite layer has been removed at the site; therefore, the potential for subsidence is negligible.2.5.4.7 Response of Soil and Rock to Dynamic Loading All Seismic Category I structures within the plant area, except the Class IE electrical manholes and Class IE underground electrical conduits and some Seismic Category I pipes, are founded on sound rock. There will be no amplification in the sound rock. The acceleration at the foundation levels of those structures supported on sound rock is equal to the baserock acceleration.The maximum horizontal accelerations which will be experienced during a safe shutdown earthquake are 15 percent of gravity. For an operating basis earthquake, the maximum accelerations will be one-half those of the SSE; i.e., 7 1/2 percent of gravity. A further Amendment 65 Page 185 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 discussion of the earthquake design basis is contained in Section 2.5.4.9, and a complete discussion is presented in Section 2.5.2.The soil-structure interaction analyses for the structures founded on bedrock were performed by using the fixed-base approach based on the lumped mass-spring method. A general discussion of the analyses is presented in Section 3.7.2.4.The engineering properties of the bedrock at the site used for dynamic soil-structure interaction analyses are discussed below:Poisson's ratio was estimated from the following equation derived from solid mechanics:

 =

where Vp and Vs are respectively the compressional and shear wave velocities determined from field geophysical data (Table 2.5.2-3). The average value of Poisson's ratio used in the analyses is 0.35, which is consistent with values reported in the literature (Reference 2.5.4-2).As a basis with which to compare the dynamic modulus values, the static modulus of deformation was calculated from unconfined compression tests, and is presented in Table 2.5.4-3 and then reduced by a factor dependent on rock quality designation, RQD, (Reference 2.5.4-1). This reduction procedure is necessary when calculating moduli values from laboratory tests on intact cores. Intact cores tested in the laboratory do not take field discontinuities into consideration.The dynamic modulus was determined from the following equation from solid mechanics:E =where is the mass-density of the rock.Discussions of the determinations of the mass density and compressional and shear wave velocities are presented in Sections 2.5.4.2 and 2.5.4.4.To be conservative, the effect of confining pressure on the modulus was neglected. Since moduli can normally be assumed to vary as a function of applied stress, this effect was examined by comparing results of laboratory velocity measurements. Therefore, the variation of moduli as a function of strain rate was observed. Since static tests for rock exhibit strains of about 10-2 to 10-3 in/in and field geophysical measurements were conducted at strains of about 10-6 in/in, it is possible to determine if any significant modulus variation occurs. The moduli values of the competent rock were shown to be insensitive to changes in strain rate above and below that expected of the SSE.The formulae used in computing the spring constants for the foundations consisting of rigid rectangular footings or mats resting on elastic half space are presented in Table 2.5.4-8.Amendment 65 Page 186 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The horizontal spring constant, kx, considers the sliding of the foundation on its supporting material. To consider the effect of resistance of rock material surrounding the foundation, a spring constant was used which is analogous to the vertical spring constant, oriented to the horizontal.The Class IE Electrical Manhole Structures consist essentially of very rigid reinforced concrete boxes fully buried in the surrounding soil. Because of their relatively small sizes, the individual structures are assumed to be single mass points excited by the same accelerations as for the surrounding soil mass. The ground acceleration at the level of individual manholes was determined by an amplification analysis of ground motion through a vertical soil column between the bedrock and the manholes by using the computer program SHAKE developed by the University of California, Berkeley.The program SHAKE computes the responses associated with vertical propagation of shear waves through the linear viscoelastic system. The program is based on the continuous solution of the wave equation adapted for use with transient motion through the Fast Fourier Transform algorithm. The nonlinearity of the shear modulus and damping is accounted for by the use of equivalent linear soil properties using an iterative procedure to obtain values for modulus and damping compatible with the effective strain in each layer.The input soil and rock properties were obtained from the field geophysical measurements and the dynamic laboratory testing, and are summarized in Appendix 2.5C. The model is divided into a layer system and each layer is completely defined by its characteristics of shear modulus, critical damping ratio, unit weight and thickness. Both of the horizontal and vertical safe shutdown earthquake motions were input separately at the base of the model (bedrock).Through an iterative process, subsequently the strain compatible solutions were obtained, and the new motions at the top of each layer were computed. Vertical soil column models are shown on Figure 3.7.2-8.The ground accelerations at the level of manholes obtained through the above analysis were further increased by 50 percent for the equivalent static analysis of each structure.The accelerations used for design of the manholes, as obtained from the above procedure, are as follows:a) Horizontal SSE Acceleration: 0.25g b) Vertical SSE Acceleration: 0.19g c) Horizontal OBE Acceleration: 0.14g d) Vertical OBE Acceleration: 0.10g The response of buried pipelines to dynamic loading is discussed in Section 3.7.3.12, while those of the dams, dike, and channels are discussed in Section 2.5.6.2.5.4.8 Liquefaction Potential The foundation of the plant is hard, sound rock, and has no potential for liquefaction.Amendment 65 Page 187 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.9 Earthquake Design Basis The plant site lies in an aseismic area; no earthquakes have been reported within 40 miles of the site. Eight earthquakes of epicentral intensity VII and above have occurred within about 200 miles of the site. The largest of these is the Charleston earthquake of August 31, 1886, of Modified Mercalli Intensity X.Most earthquakes in the region occur in four major seismic zones (Figure 2.5.2-2); the Southern Appalachian Seismic Zone, the Northern Virginia-Maryland Seismic Zone, the Central Virginia Seismic Zone, and the South Carolina - Georgia Seismic Zone. If the largest earthquake in any one of these zones is assumed to occur at the point in that zone nearest the plant, the largest site intensity that can be calculated, using Bollinger's (Reference 2.5.2-25) attenuation relationship for the Southeast, is 6.71. This results from the Charleston earthquake of 1886, and is larger than the motion calculated for the New Madrid earthquakes of 1811-1812.To be conservative, we have assumed that the SSE will be an earthquake of intensity VII, and that this earthquake will occur close to the site, although no earthquakes have been recorded within 40 miles of the site. Such an earthquake would result in an acceleration of 0.12 g at the site. To be conservative, a value of 0.15 g has been adopted.The shock is expected to be of less than 10 seconds duration, producing no more than 5-10 cycles of motion at the maximum predicted level. This level of motion should be similar to the Golden Gate (San Francisco) earthquake of 1957, or the Helena, Montana earthquake of 1935.Since these earthquakes resulted in only 2 or 3 cycles of strong motion, selection of 10 cycles is conservative. Horizontal and vertical response spectra for the SSE, prepared in accordance with Regulatory Guide 1.60 (see Section 1.8) are presented on Figures 2.5.2-12 and 2.5.2-13.The operating basis earthquake accelerations are considered to be one-half those of the SSE (namely, 0.075g), equivalent to an Intensity VI earthquake near the site. The horizontal and vertical response spectra of the OBE are presented on Figures 2.5.2-14 and 2.5.2-15.Complete discussion of the earthquake design basis is presented in Section 2.5.2.2.5.4.10 Static Stability The geologic conditions underlying the site are discussed in detail in Section 2.5.4.1. The in-situ rock properties are discussed in Sections 2.5.4.2 and 2.5.4.5. Static stability analyses for the dams, dike and channels are discussed in Section 2.5.6. This section discusses static analyses of the plant island structures, including settlement, bearing capacity, and lateral earth pressures.Major facilities are founded on fresh sound siltstone, sandstone, shale, and claystone of the Sanford Formation. The rock provides adequate support for the units under static conditions.2.5.4.10.1 Settlement 2.5.4.10.1.1 Methods of Analysis The average settlement of the foundation under combined static and dynamic loading is found by using the equation derived from Boussinesq for the normal displacement of a semi-infinite Amendment 65 Page 188 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 elastic solid under the action of a normal load (Reference 2.5.4-4). The equation can be expressed as:S =where: S = average settlement

 = displacement coefficient dependent on the shape, loaded surface, and distribution of load (Reference 2.5.4-4) = Poisson's ratio EE = design static modulus P = total applied load A = total area on which load is applied 2.5.4.10.1.2 Selection of Deformation Properties Two parameters are required to define the in-situ static deformation properties, the elastic modulus and Poisson's ratio. The frequency and nature of geologic discontinuities are significant factors in determining the properties of the rock and their effect can be incorporated in the design equations by use of the reduction factor applied to Young's modulus.

2.5.4.10.1.2.1 Design Static Modulus EE The representative tangent modulus ET is selected based on the statistical method by using calculated values from unconfined compression and triaxial tests on selected core specimens (see Section 2.5.4.2 and Table 2.5.4-3).Using the reduction factor described above, the representative tangent modulus, ET, is reduced to incorporate the effects of in-situ geologic discontinuities. The design elastic modulus is:EE = ET (RF) where: EE = design static modulus ET = average value of laboratory tangent modulus RF = reduction factor based on RQD.The reduction factor is based on the RQD of the rock, as discussed in Section 2.5.4.2. The average value of the laboratory modulus selected is 1.7 x 106 psi. The average RQD is 92.7 percent (Table 2.5.4-2). Based on a conservative RQD of 75 percent, the reduction factor RF is 40 percent (Reference 2.5.4-5). The value of the design elastic modulus obtained is 6.8 x 105 psi.Amendment 65 Page 189 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.10.1.2.2 Static Poisson's Ratio Values of Poisson's ratio, , were obtained from both laboratory compression tests on selected core specimens (discussed in Section 2.5.4.2), and from field geophysical measurements (discussed in Section 2.5.4.4). The average design value of the selected is 0.35 based on field geophysical measurements. This value is consistent with the values reported in the literature (Reference 2.5.4-2).2.5.4.10.1.3 Computed Settlement The average settlements under static loading computed for the various structures are presented in Table 2.5.4-9.It can be seen that these settlements are very small. The differential settlements should be even smaller and, therefore, structurally tolerable. Since the settlements consist of the pseudo-elastic compression of the underlying rock, they will occur essentially upon load application.2.5.4.10.2 Bearing Capacity 2.5.4.10.2.1 Method of Analysis The ultimate bearing capacities for strip foundations on rock can be obtained by using the following equation from Stagg (Reference 2.5.4-4).qult = 0.5 BN + CNc + qNq where: = effective unit weight B = width of the foundation mat C = cohesion q = surcharge pressure and N, NC, and Nq are bearing capacity values depending on the friction angle of the foundation rock.This equation is identical with Terzaghi's bearing capacity equation (Reference 2.5.4-6).For square foundations, the corresponding bearing capacity equation is:qult = 0.4 BN + 1.2 CNc + qNq 2.5.4.10.2.2 Selection of Strength Properties The appropriate properties required for the bearing capacity computation are the density and shear strength of the rock.Amendment 65 Page 190 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4.10.2.2.1 Density The results of laboratory density tests are described in Section 2.5.4.2. The average density of the rock, as shown in Table 2.5.4-1, is 162.8 pcf. A value of 160 pcf was used for design calculations.2.5.4.10.2.2.2 Shear Strength The shear strength of the foundation rock was determined from laboratory unconfined and triaxial tests on selected rock cores. The details of the tests are discussed in Section 2.5.4.2.The results are summarized in Tables 2.5.4-3, 2.5.4-4, and Figure 2.5.4-67. Based on the unconfined compression test data, the average shear strength of the rock is about 4000 psi. To be conservative, this shear strength is reduced by the factor RF described in Section 2.5.4.10.1.2.1 earlier. The value of the shear strength used for bearing capacity analysis is 1600 psi.From the Mohr's circle plot shown on Figure 2.5.4-67, the average friction angle of the rock is about 45 degrees. For design, however, the friction angle of the foundation rock is conservatively chosen to be equal to zero.2.5.4.10.2.3 Computed Bearing Capacity Based on a friction angle of zero, the bearing capacity equation becomes:qult = 1.2CNc + q By disregarding the surcharge term q and putting NC = 5.14 (Reference 2.5.4 6),qult = 6.2C The computed ultimate bearing capacity is 714 tsf.The design bearing capacity chosen is 25 tsf, which provides a factor of safety of 28 compared with the ultimate bearing capacity (Reference 2.5.4-7).2.5.4.10.3 Lateral Earth Pressures The Plant Island Structures were designed for static and dynamic earth pressures.The following assumptions were implemented for purposes of this analysis:a) Water level was assumed to be at Elevation 251 ft. (Grade level Elevation 260 ft.).b) The depth of backfill considered for at-rest pressure calculations extends from grade level (Elevation 260 ft.) to the top of the foundation mat (elevation varies with structures).c) Pressure due to groundwater was considered only for the depth between Elevation 251 ft. and the top of the foundation mat.Amendment 65 Page 191 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 d) The total lateral pressure is equal to the effective earth pressure plus the hydrostatic pressure.2.5.4.10.3.1 Method of Analysis 2.5.4.10.3.1.1 Static Earth Pressure The foundation walls of the structures were designed for at-rest earth pressure and hydrostatic loading. The at-rest earth pressure coefficient, Ko, was computed by using the following formula:K = 1 - Sin where is the internal friction angle of the backfill soil. The value of the friction angle is discussed in Section 2.5.4.10.3.2.It is assumed that full at-rest pressure is developed by backfill, disregarding the possible compensating effect of the rock layer in reducing at-rest pressure against the walls.2.5.4.10.3.1.2 Dynamic Earth Pressure A dynamic lateral earth pressure analysis was performed for all seismic Category I structures.The procedures and parameters utilized in this analysis are presented below.The backfill against the exterior walls of the structures is divided into two levels or regions, as shown on Figure 2.5.4-130.a) The first level - depth d1 - is confined by the exterior wall surface on one side and by the soil on the other.b) The second level - depth d2 - is confined by the exterior wall surface on one side and by the rock on the other.For the first level it is assumed that full passive pressure may be realized.To compute the pressure on the face of the wall (between points A and B) due to an earthquake, it is necessary to compute the movement of the wall as it varies with depth below grade. The design earth pressure is then obtained from the strain vs. pressure coefficient curve shown on Figure 2.5.4-131 using the computed strain.The strain is computed as the wall movement at a particular depth divided by the length of the horizontal component of the Rankine failure surface at that depth.The movement of the wall relative to the soil for the first level, d1, is the arithmetic sum of the movement obtained from the dynamic analysis and the maximum soil movements as determined by using the deflection of the 15 percent damping curve of the ground response spectra for the corresponding soil column frequency, as discussed in Section 3.7.2.4A.For the second level - depth d2 - it is assumed that confinement of the backfill by the very steep rock face will prevent the formation of a passive pressure wedge.Amendment 65 Page 192 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 To compute the state of pressure on the face of the wall (between points B and C) due to an earthquake, the strain at various elevations is computed by dividing the total relative movement of the wall with respect to the rock at that level by the width of the backfill layer - between the concrete wall and the rock - at that level.The movement of the wall relative to the soil over region d2 is obtained by the same procedure as that employed for region d1 (see above).Once the strain is computed for various levels, it is related to the triaxial test stress-strain curves for the correct overburden pressure used as v for the backfill material to obtain the corresponding pressure for each level. This pressure strain curve is shown on Figure 2.5.4-131 and is modified from data presented in the literature (Reference 2.5.4-2).2.5.4.10.3.2 Selection of Design Properties The properties of the backfill used in the computation of the static and dynamic earth pressures are summarized below:a) Friction angle = 20 degrees b) Cohension C = 400 psf c) Dry Density = 115 pcf (silty clay) d) Saturated Density = 130 pcf e) Submerged Density = 67.6 pcf f) Coefficient of at-rest pressure Ko = 1 - Sin = 0.7 These backfill properties were based on results of laboratory testing which are presented in Section 2.5.4.5. The values for friction angle and cohesion were based on results of laboratory triaxial tests with pore pressure measurements on samples compacted to a dry density equal to 95 percent of the maximum dry density obtained from the Standard Proctor test. The actual test results are shown on Figures 2.5.4-106 through 2.5.4-113 and 2.5.4-120 through 2.5.4-129.2.5.4.11 Design Criteria The design criteria and methods of analysis used in the stability studies of settlement, bearing capacity, and earth pressure are discussed in detail in Section 2.5.4.10.The settlement analyses were based on the assumption that settlement is due essentially to the pseudo-elastic compression of the rock. The formula used in the computation of settlement is based on the theory of elasticity. In view of the very high safety factor against bearing capacity failure, the assumption of elasticity is valid. The modulus obtained from laboratory compression testing of intact core has been properly adjusted for discontinuities in the rock mass. The results indicated that the total, and hence the differential settlement, should be small.The bearing capacity of the foundation is computed by using Terzaghi's bearing capacity equation (Reference 2.5.4-6) which is generally conceded to be conservative. For added Amendment 65 Page 193 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 safety, the shear strength of the rock, obtained from the laboratory compression test of the rock core, was reduced to account for presence of discontinuities in the rock mass. The factor of safety based on a design bearing capacity of 25 tsf is given in Section 2.5.4.10.2.3; however, the required factor of safety is three. The design bearing capacity of 25 tsf is consistent with what is allowed by most building codes for sound rock. The maximum computed vertical bearing pressure on the foundation rock from any plant structure is 39 ksf which results in a computed factor of safety of 36.The coefficient of lateral earth pressure used for design under static conditions is the "at-rest" value, which combined with the water table at Elevation 251 ft., results in a safe total lateral pressure condition.The dynamic lateral pressure is established from structural displacements obtained from the dynamic analysis. The strain is computed by dividing the total relative movement of the wall with respect to the rock at that level by the width of the backfill (i.e., between the wall and the rock at that level). The stress is obtained by utilizing the stress-strain relationship for passive loading conditions reported in the literature (Reference 2.5.4-2). This results in a dynamic lateral earth pressure that is conservative.2.5.4.12 Techniques to Improve Subsurface Conditions In preparing the final excavation, all weathered and broken rock was removed, and dust and fine materials were cleaned off the rock surface by using air lances. While this final clean-up was going on, the resident geologist inspected the surface to ensure that no loose blocks were present; any found were removed with picks and bars.After horizontal surfaces had been cleaned and mapped, joints and fractures were slush grouted until refusal. If leakage of grout occurred at lower elevations, the leaking fracture was normally caulked with rags or oakum and wedges. Where the leakage was from a feature which could not be caulked, several pours of slush grout were made, allowing each to set before continuing with another placement. This process continued until leakage ceased.Vertical faces were covered with four in. wire mesh suspended from pieces of rebar cemented into the face. In certain instances this procedure proved unnecessary and for economic reasons, foundation mat or structural concrete was placed directly against the rock walls, with no loss of design intent. Rock bolts and rock anchors were generally not required.2.5.4.13 Subsurface Instrumentation During construction, close survey control was maintained on a series of monuments on horizontal and vertical faces to ensure there was no movement. Piezometers were installed to monitor groundwater levels during excavation as discussed in Sections 2.5.4.6 and 2.4.13.2.5.4.14 Construction Notes LATER Amendment 65 Page 194 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.5 STABILITY OF SLOPES There are no natural slopes at the project site the failure of which could adversely affect the safety of the nuclear power plant. Stability of slopes of man-made structures, such as the Main Dam, the Auxiliary Dam, the Auxiliary Reservoir Separating Dike, and emergency service water channels is discussed in Section 2.5.6.2.5.6 EMBANKMENTS AND DAMS 2.5.6.1 General Information 2.5.6.1.1 Purpose The primary purpose of the dams discussed in this section is to impound water for the circulating water and service water systems for the Shearon Harris Nuclear Power Plant. The Main Dam impounds a reservoir with a normal water level at Elevation 220 ft. and a water surface area of approximately 4,000 acres. During normal operation, the Main Reservoir functions as a storage reservoir and is used as the source of cooling tower makeup water. The Main Reservoir also serves as an alternative source of emergency service water supply or ultimate heat sink.The Auxiliary Dam impounds a reservoir with a minimum normal water level at Elevation 250 ft.and a surface area of approximately 317 acres. The Auxiliary Reservoir serves the Emergency Service Water System. An Auxiliary Separating Dike across the east arm of this reservoir acts as a barrier to prevent discharged emergency service water from flowing directly back to the emergency service water intake area. The Auxiliary Reservoir Channel conveys discharged emergency service water into the west arm of the reservoir so that maximum cooling can be attained before the discharged water circulates back to the intake area.The Auxiliary Reservoir must remain operative under the safe shutdown earthquake (SSE) condition. Consequently, the Auxiliary Dam, Auxiliary Separating Dike, Auxiliary Reservoir Channel, and Emergency Service Water Intake and Discharge Channels are Seismic Category I facilities. The Main Dam and Emergency Service Water and Cooling Tower Makeup Intake Channel, because of their importance, have also been classified as a Seismic Category I structure.2.5.6.1.2 Location A location map, which shows the plant area, Main Reservoir, Main Dam, Auxiliary Reservoir, Auxiliary Dam, Auxiliary Separating Dike, Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, and Cooling Tower Makeup Water Intake Channel, is presented on Figure 2.5.1-10. The location of these facilities is based on a siting study performed by Ebasco Services Incorporated in October, 1970 (Reference 2.5.6-1).The Main Dam is located on Buckhorn Creek about 4.5 miles south of the plant site and about 2.5 miles north of the Cape Fear River.The Auxiliary Dam is located across the Tom Jack Creek arm of the Main Reservoir, adjacent to the southwest boundary of the plant site.Amendment 65 Page 195 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The Auxiliary Separating Dike is located about 1700 ft. north of the Auxiliary Dam, between the emergency service water intake area and the emergency service water discharge area. The Auxiliary Reservoir Channel is located northwest of the Auxiliary Separating Dike. The Emergency Service Water Intake and Discharge Channels are located on the plant island southwest and northeast of the plant, respectively. The Cooling Tower Makeup Water Intake Channel is located southeast of the plant.2.5.6.1.3 General design features The Main Dam, Auxiliary Dam, Auxiliary Reservoir Separating Dike, Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, and Emergency Service Water and Cooling Tower Makeup Intake Channel are designed and constructed to Seismic Category I criteria and also to withstand the effects of credible combinations of natural phenomena. The slopes of the dams, dike, and channels are designed for a factor of safety of 1.5 under static conditions, 1.2 for simultaneous OBE and 100 year return period flood level, and 1.1 for simultaneous SSE and 25 year return period flood level.The simultaneous OBE and 100-year return period flood level was not analyzed for the ESW and Cooling Tower Makeup Intake Channel since the simultaneous SSE and 25-year return period flood level analysis was more conservative.2.5.6.1.3.1 Main dam and spillway The Main Dam is a rockfill dam approximately 1550 ft. long at the crest, Elevation 260 ft. It is founded on rock and has a maximum height of approximately 108 ft. It has a core of compacted silty clay and clayey silt material protected on each side by two 8-ft.-thick fine and coarse filter zones and a rockfill shell. The outside slopes are two horizontal to one vertical. The Main Dam plan is shown on Figure 2.5.6-1, and a profile and section are shown on Figure 2.5.6-2.The general plan and details of the Main Dam Spillway are shown on Figures 3.8.4-34, 3.8.4-35, and 3.8.4-36.The basis for the hydraulic design of the spillway is to accommodate the probable maximum flood (PMF). The spillway is uncontrolled and consists of two ogee sections, each 25 ft. wide, separated by a concrete pier, with a crest at Elevation 220 ft.The average velocity of flow in the spillway's approach channel is approximately 1 ft./sec. for a 100-year return period flood and approximately 8 ft./sec. for the PMF. The channel is cut into rock for a distance of approximately 200 ft. upstream of the ogee crest structure and in soil approximately 105 ft. further upstream. The invert and sides of the channel cut in rock are lined with concrete. The concrete lining has dowels across longitudinal and transverse contraction joints. The side lining and part of the invert lining adjacent to the ogee structure are secured to rock by rock anchors. The transverse contraction joints in the concrete lining are spaced at approximately 40 ft. centers.The portion of the approach channel which is excavated entirely in soil is designed for an average flow velocity of 4 ft./sec. for the PMF. The invert and sides of the channel are protected by riprap placed on bedding. The riprap is sized to withstand a velocity of 8 ft./sec. to permit the use of the same size riprap for the soil side slopes above the concrete lined portion of the channel.Amendment 65 Page 196 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The discharge channel downstream of the ogee crest structure has a super-critical slope of 0.02 ft. per ft., except for a portion of the channel upstream of the stilling basin which is sloped at two horizontal to one vertical to join the discharge channel to the stilling basin. The average velocity of flow in the channel for the PMF is approximately 45 ft. per second; however, the velocity increases to approximately 72 ft./sec. at the stilling basin. The Froude number of incoming flow at the stilling basin is six. The stilling basin has concrete chute blocks and a dentated end sill.The spillway's discharge channel and stilling basin are excavated in rock with side slopes of one horizontal to four vertical. The rock in the invert and on the sides is lined with concrete that is secured to bedrock by rock anchors. The top of the lining has a minimum freeboard of 3 ft.above the PMF water surface profile. Transverse contraction joints in the concrete lining are spaced at approximately 40 ft. centers. Transverse drains, consisting of perforated concrete pipe placed in a trench filled with crushed rock, are located at each transverse joint in the invert lining in order to minimize uplift. The transverse drains lead to a longitudinal collector drain which has its outlet in the stilling basin. The side lining has drainage holes drilled into rock in order to relieve water pressure. Drainage holes, provided in the bed lining, are drilled 20 ft.deep into the rock in order to drain the rock and reduce uplift.The average velocity in the discharge channel downstream of the stilling basin is approximately 6 ft./sec. for the PMF. The discharge channel is cut in rock for a distance of approximately 125 ft. downstream of the concrete end sill of the stilling basin, which precludes undermining of the structure. The sides and invert of the channel, downstream of the rock cut section, are excavated in soil and protected by sacrificial rockfill laid on bedding.2.5.6.1.3.2 Auxiliary dam and spillway The Auxiliary Dam is a random rockfill dam approximately 3903 ft. long, a maximum structural height of approximately 72 ft., and a crest at Elevation 260 ft. Its outside slopes are 2.5 horizontal to one vertical. It has a core of compacted silty clay and clayey silt material protected on each side by a transition filter zone and a random rockfill shell. The downstream shell is provided with two horizontal drainage blankets, each 3 ft. thick, which are connected to the transition filter zone adjacent to the core of the dam. In addition, a 200 ft. wide, 3 ft. thick drainage layer is provided under the shell in each of two areas where pre-existing creeks had been located. The core of the dam is founded on weathered rock and the core (cutoff) trench is excavated to suitable rock. The filters and random rockfill shells are founded on weathered rock at the center of the dam and on firm residual soil near the abutments. The Auxiliary Dam plan is shown on Figure 2.5.6-3 and a profile and sections are shown on Figure 2.5.6-4.The general plan and details of the Auxiliary Dam Spillway are shown on Figures 3.8.4-37 and 3.8.4-38. The spillway is an uncontrolled concrete ogee section with a crest length of 170 ft.and crest at Elevation 252 ft. The basis for its hydraulic design is the probable maximum flood (PMF).The ogee crest of the spillway is joined to the stilling basin by a sloping apron. The stilling basin is conservatively designed for a tailwater level corresponding to a 100 year drought water level in the Main Reservoir. Since the main reservoir water level varies from a low water level of Elevation 204.4 ft. during drought conditions (coincident with emergency shutdown) to the maximum still water level of Elevation 238.9 ft. during the PMF, the sloping apron allows for proper formation of the hydraulic jump at all tailwater levels. The incoming flow at the stilling Amendment 65 Page 197 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 basin has a maximum velocity of approximately 45 ft./sec. and a Froude number of 10. Control of the velocity of water in the stilling basin is provided by concrete chute blocks and baffle piers.All Auxiliary Dam Spillway structures are founded on rock. Over-excavated areas were filled with concrete. Rock slopes in the excavation for the sloping apron and stilling basin have a slope of approximately one horizontal to four vertical. Rock in the invert and sides of the structures are lined with concrete secured to rock by rock anchors. The lining slabs are joined by dowels across longitudinal and transverse contraction joints. The transverse joints are spaced at approximately 25 ft. centers. Transverse drains, consisting of perforated concrete pipe laid in a trench filled with crushed rock, are located at each transverse joint in the invert to minimize uplift. The drains have outlets in the invert. The side lining has drainage holes drilled into rock to relieve water pressure. Drainage holes drilled into rock to drain the rock and to reduce uplift have been provided through the invert.The side walls of the stilling basin and sloping apron, which joins the ogee crest to the stilling basin have sufficient height to retain the embankment fill and to confine the hydraulic jump within the side walls for all possible locations of the formation of the jump. The walls have a minimum freeboard of 2 ft. at the end of the jump. The additional height required for the side walls is provided by construction of retaining walls on top of the rock.A discharge channel is located downstream of the stilling basin. The invert of the channel immediately downstream of the stilling basin is cut into rock for a distance of approximately 300 ft., which precludes undermining of the stilling basin. On the east side of the channel, up to Station 40+00 and on the west side up to Station 20+00, where the sides of the discharge channel are excavated in soil, the sides are protected by riprap placed on bedding below Elevation 225 ft. to preclude side erosion. The riprap is sized to withstand a velocity of 6 ft./sec.The top of the discharge channel excavation is more than 400 ft. from the toe of the Auxiliary Dam at its nearest location.2.5.6.1.3.3 Auxiliary Reservoir Separating Dike The Auxiliary Reservoir Separating Dike is approximately 1200 ft. long and has a maximum height of approximately 55 ft. Its outside slopes are 2.5 horizontal to one vertical. The dike has a core of compacted silty clay and clayey silt material protected by a random rockfill shell which is graded near the core with the finer materials adjacent to the core. The core and rockfill shell are founded either on weathered rock or on a thin layer of stiff residual soil overlying weathered rock (see Appendix 2.5E, Figure 9). The upstream and downstream slopes of the Auxiliary Reservoir Separating Dike are protected by riprap as shown on Figure 2.5.6-5. The plan, section, and profile of the Auxiliary Reservoir Separating Dike are shown on Figure 2.5.6-5.2.5.6.1.3.4 Auxiliary Reservoir Channel The Auxiliary Reservoir Channel is approximately 1570 ft. long and 140 ft. wide at its invert Elevation of 235 ft. Its sides have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock. The plan, profile, and sections are shown on Figure 2.5.6-6.The Auxiliary Reservoir Channel is sized to carry the maximum ultimate discharge of the Service Water System coincident with the PMF flood flow for the upstream drainage basin.To facilitate construction of the Auxiliary Separating Dike, a diversion channel, 25 ft. wide with an invert at Elevation 225 ft., was excavated in the Auxiliary Reservoir Channel.Amendment 65 Page 198 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.1.3.5 Emergency service water channels The Emergency Service Water Intake and Discharge Channels (Figure 2.5.1-10) are designed to carry the service water flow required for normal and emergency shutdown of SHNPP.2.5.6.1.3.5.1 Emergency service water intake channel The Emergency Service Water Intake Channel is approximately 3580 ft. long and 50 ft. wide at its invert elevation of 238 ft. The channel walls have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock. Where cut in soil, an impervious lining 50 ft. thick was provided to form the side-slopes of the channel. Areas along the channel invert which contained sandy alluvium material were removed and replaced with an impervious material to a depth where insitu impervious material was encountered. The plan, profile, and sections of the channel are shown on Figure 2.5.6-7.2.5.6.1.3.5.2 Emergency service water discharge channel.The Emergency Service Water Discharge Channel is approximately 2170 ft. long. The width of the channel varies from 50 ft. to 80 ft. at the channel's invert Elevation of 240 ft. The channel walls have a slope of two horizontal to one vertical in soil and one horizontal to four vertical in rock. The plan, profile, and sections of the channel are shown on Figure 2.5.6-8.2.5.6.1.3.5.3 Emergency Service Water and Cooling Tower Makeup Intake Channel.The Emergency Service Water and Cooling Tower Makeup Intake Channel is approximately 2500 ft. long and 45 ft. wide at its invert elevation of 194.0 ft. The channel walls have a slope of two horizontal to one vertical in soil, one horizontal to four vertical in rock on the north side of the channel, and two horizontal to one vertical in rock on the south side. The plan, profile and sections of the channel are shown on Figure 2.5.6-28.2.5.6.2 Exploration A comprehensive series of exploration programs was conducted in the main dam and auxiliary reservoir areas for evaluation of foundation conditions and to locate and sample possible sources of borrow and quarry materials for construction of the dams. An index map of FSAR figures, which show the locations of various exploration activities, is presented on Figure 2.5.1-10.2.5.6.2.1 Main dam area The Main Dam area, as discussed in this section, includes the Main Dam, the Main Dam Spillway, and the main dam impervious borrow and quarry areas.2.5.6.2.1.1 Local geologic features The site of the Main Dam and Spillway (Figure 2.5.1-4) is approximately 3000 ft. southeast of the Jonesboro Fault; it is underlain by pre-Triassic igneous and metamorphic rocks which have no direct relationship to the Triassic sedimentary rocks which lie on the northwestern side of the fault and underlie the plant site. The main dam quarry area (Appendix 2.5F, Figure 12) is also located in pre-Triassic crystalline rocks. Main dam impervious borrow areas M and W (see Amendment 65 Page 199 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Appendix 2.5F, Figures 9 and 10), are located on the northwestern side of the fault in Triassic sedimentary rocks of the Sanford Formation, the same rock formation that underlies the plant and Auxiliary Dam site.The lower part of this formation underlies the plant site, whereas the upper conglomeratic part underlies the borrow areas.All Seismic Category I structures in the main dam area are on the southeastern side of the Jonesboro Fault. Therefore, the following discussion of lithology and structure is limited to the igneous and metamorphic rocks underlying the main dam and spillway area.2.5.6.2.1.1.1 Lithology Significant rock exposures in the main dam area are limited to stream beds and railroad cuts. In most places, the rock is covered by residual soil (saprolites) or by alluvium in the valley bottoms.As exposed in the excavations for the main dam core trench, diversion conduit, and spillway, the bedrock is generally hard, strong, and fresh to moderately weathered. Four main types of rock were distinguished for the purpose of foundation mapping. These are (1) granite, (2) hornblende-mica gneiss, (3) mica schist, and (4) quartz feldspar gneiss. The distribution of the rocks types is shown on the geologic maps of the Main Dam (Appendix 2.5E, Figures 3 through 6).

1. Granite - Rocks of granite composition are the most common; they underlie most of the left abutment of the Main Dam, most of the upstream part of the diversion conduit, and a major part of the spillway. These rock are typically hard to very hard, strong, and medium grained. They are composed predominately of feldspar (orthoclase and plagioclase) and quartz, with minor amounts of biotite, chlorite, muscovite, epidote, and pyrite. They are light gray when fresh, creamy white when slightly weathered, and tan or buff when moderately to highly weathered. The granites are commonly highly foliated, although foliation is weak or absent in places. In man places they are interlayered with intricately folded mica schist and/or hornblende-mica gneiss.

Xenoliths of hornblende gneiss are common in the granite where foliation is weak or absent. Much of the highly foliated granite appears to contain indistinct compositional layering. The granitic rocks normally weather to light gray, fine silty to coarse sandy saprolite; the upper portion of such saprolite is commonly weathered to tan, slightly clayey sand, silt, or silty sand.

2. Hornblende-Mica Gneiss - Hornblende-mica gneiss is the second most common rock type in the foundations; it underlies much of the right abutment of the dam, the downstream part of the diversion conduit, and most of the downstream part of the spillway. This gneiss is a hard, strong, medium to fine-grained rock composed of hornblende and subordinate amounts of plagioclase and biotite. Chlorite and pyrite are common accessory minerals. The gneiss is gray to black when fresh, bluish or greenish gray when slightly weathered, and rusty brown when moderately weathered.

Foliation in the gneiss is poorly to very poorly developed; in places the foliation is so weak that the rock resembles diorite. This rock type normally weathers to a red or tan saprolite composed of micaceous sandy silt; the upper portion of the saprolite is commonly weathered to a deep-red clayey sandy silt or sandy clayey silt.Amendment 65 Page 200 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

3. Mica Schist - Mica schist occurs as complexly folded layers 0.01 to 10 ft. thick within the granite and as thin layers along joint surfaces within the hornblende-mica gneiss.

The schist is soft to moderately hard, weak to moderately strong, moderately weathered, fine grained, highly fissile, and composed predominately of chlorite and biotite with subordinate muscovite and amphibole. Some schist layers contain abundant quartz. This rock type normally weathers to a very micaceous silty saprolite; the upper portion of the saprolite is commonly weathered to a red micaceous clayey silt.

4. Quartz-Feldspar Gneiss - Quartz-feldspar gneiss occurs in the central section of the main dam cutoff trench, in a small area at the intersection of the cutoff trench and the diversion conduit, and in part of the spillway. This rock is light gray to dark brown, fine to medium grained, moderately hard, moderately strong, and slightly to moderately weathered. It is characterized by close interlayering of gneissic rock composed mostly of quartz and feldspar with schistose rock composed of biotite and muscovite or, less commonly, hornblende and mica. Individual layers are from 0.01 to 10 ft. thick. A micaceous, quartz-rich variety of gneiss containing few schistose layers is present in the spillway area. The higher quartz content of the gneiss distinguishes it from rocks mapped as granite.

Other rock types in the foundations occur mostly as veins and small pods. Quartz veins are common in all rock types, and residual boulders of vein quartz are abundant on the ground surface. Most quartz veins cross-cut the foliation of the host rock, but some are parallel to the foliation. Veins and pods of quartz-feldspar pegmatite are also common in most areas of the foundations. Small veins and veinlets of epidote occur mostly in the hornblende-mica gneiss.2.5.6.2.1.1.2 Structure The orientation of the foliation and compositional layering in rocks in the main dam area is highly variable due to the highly complex nature of deformation apparent in these rocks, particularly in the layered quartz-feldspar gneiss, and in the mica schist and hornblende-mica gneiss layers in the granite. The rocks appear to have been affected by several periods of folding, with isoclinal folding predominating. However, no consistent pattern in the orientation of the folds is discernable in the foundation rocks. In most places the foliation is parallel or sub-parallel to the compositional layering. Northeast strike directions and northwest dip directions predominate. Foliation striking north-northwestern and dipping southwest is also present in places, but is much less common.Joints are most common in the granite and the hornblende-mica gneiss. The dominant joint set strikes approximately N 60°-70° E and dips 50° to 70° to the south. Another set strikes N 20°-35° W and dips 70° to 90° southwest. Joint spacing ranges from a few inches to a few feet.Several minor, non-capable faults with lengths measured in tens of feet were mapped in the foundation of the Main Dam. Most of the faults have strikes between N30°E and N75°E and dip steeply to the southeast. A few strike northwest and have steep northeast or nearly vertical dips. A majority of the faults exhibit right lateral strike separation, but those showing left lateral separation are also common. Apparent displacement is rarely more than two feet and is commonly less than 6 in. Detailed investigations of the faults, described in Section 2.5.6.2.1.2.2, indicate that any movement along the faults must have occurred prior to or during Amendment 65 Page 201 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 deformation-mineralization processes which terminated more than 225 million years ago.Written reports on the faults as presented to the NRC are cataloged in Appendix 2.5E.2.5.6.2.1.2 Exploration programs An extensive exploration program was conducted in the vicinity of the Main Dam in order to evaluate foundation conditions for various reservoir-related structures and for highway and railroad relocations, and to explore and sample potential sources of borrow and quarry materials for use as impervious fill and rockfill for the Main Dam. Exploration consisted of a surface geological reconnaissance survey and subsurface investigations including detailed excavation mapping, seismic refraction surveys, borehole drilling and sampling, and excavation of test pits and trenches for soil sampling.The locations of various preconstruction exploration activities in the vicinity of the Main Dam are shown on Figures 2.5.6-9 and 2.5.6-10. The purpose of the various borehole series shown on the figures is indicated in Appendix 2.5A. Appendix 2.5A includes a tabulation of preconstruction boreholes drilled in the plant and reservoir areas, including some boreholes located in areas not shown on Section 2.5 figures. The boreholes not shown on the figures were drilled for the make-up water system, for highway and railroad relocations, and for proposed structures which were deleted from the final project design (e.g. east auxiliary reservoir structures, skimmer wall, afterbay dam, various channels, dikes, and saddle dams).The following discussion is limited to the exploration program which was related to construction of the Main Dam and Spillway.2.5.6.2.1.2.1. Geological reconnaissance A geologic reconnaissance survey of the main dam and spillway area was made in 1972 by Law Engineering Testing Company. The purpose of this reconnaissance was to determine the trends of the geologic structures in the area. Special attention was given to any structural planes of weakness that could affect the stability of the cut slopes for the main dam spillway channel.The surface reconnaissance consisted of traversing those areas likely to contain rock or saprolite exposures. In addition to the traverses, a series of inspection trenches were excavated with a backhoe near and approximately parallel to the spillway centerline. Detailed observations of the structural characteristics of the bedrock and saprolite were made in the trenches, but detailed trench logs were not prepared for the trenches because they were excavated only for observational purposes and not for sampling and testing.At each rock or saprolite exposure, the rock type (or saprolite parent rock type) was noted, and the strike and dip of foliation and joint planes were measured with a Brunton compass. The data were recorded for later map plotting and analysis.The results of this reconnaissance survey is contained in Appendix 2.5G.2.5.6.2.1.2.2 Excavation Mapping After initial excavation was completed in the diversion conduit, core trench, and spillway invert and walls, the foundations of the Seismic Category I structures (i.e., areas to be mapped at a Amendment 65 Page 202 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 scale of 1 in. to 10 ft.) were cleaned by a backhoe, hand tools, and by compressed air or compressed-air-water jets.In general, a 50 ft. reach of core trench was cleaned at a time, then a survey crew marked a grid pattern of 10 ft. centers in preparation for mapping by a geologist. Compositional layers, joints, fractures, faults, veins, pegmatites, and other mappable features in the foundation were located with respect to the grid by measuring with a tape. Attitudes of planar features, such as joints and foliation, were measured with a Brunton compass and plotted. Areas that contained faults or other anomalous features, or that required special detail were mapped at a scale of 1 in. to 5 ft.A summary of foundation features and faults encountered at the site is presented in Table 2.5E-2.When mapping was completed, a second geologist checked the field map and noted areas of discrepancies, if any, for correction. Final drafts were checked against the original field maps by the person who performed the mapping. The geologic maps are presented in the Geologic Report on Foundation Conditions Power Plant, Dams, and Related Structures (Appendix 2.5E, Figures 3 through 6).Foundation areas not mapped in plan at a scale of 1 in. to 10 ft. were mapped at a scale of 1 in.to 50 ft. These areas were cleaned with a bulldozer blade in smooth areas and a Gradall in irregular areas. A grid was surveyed on 25 or 50 ft. centers.In accordance with established CP&L procedures, unusual features and suspected faults encountered in the foundations were promptly reported to the NRC staff along with submission of a tentative evaluation of the capability of the feature. This was followed by an investigation of each feature which was documented by detailed geologic mapping, photography, and, if appropriate, other investigative methods, such as petrographic analysis. None of the features were covered by construction until they had been inspected by the NRC staff and judged by them to be non-capable based on their inspection and their evaluation of the investigation results.During mapping at the Main Dam, 28 features were reported to the NRC; of these, 22 were determined to be faults. Most of the faults were only a few tens of feet long, with only several inches of displacement. Several schistose zones which pass through the foundation were also reported and investigated. All but one of the reported features were encountered in the right abutment and spillway area, as shown on the geologic maps (Appendix 2.5.E, Figures 3 through 6). Table 2.5E-2 lists locations of the features and dates of the reports. The written reports on the features, as presented to the NRC, are listed in Table 2.5E-3.Regardless of the type of feature, it was documented that any fault movements occurred prior to or during deformation-mineralization processes which terminated more than 225 million years ago. Therefore, none of the features affect the safety of the Seismic Category I structures, and are not capable faults, as that term is defined in Appendix A to 10 CFR 100.2.5.6.2.1.2.3 Seismic Refraction Survey A seismic refraction survey that was conducted in the main dam area included one survey line (Seismic Line No. MD5) along the centerline of the Main Dam and another (Seismic Line MD6)Amendment 65 Page 203 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 along the centerline of the spillway. The locations of the lines are shown on Figures 2.5.6-9 and 2.5.6-10; seismic lines MD1 through MD4, also seen on these figures, were established during siting studies and were not used in the design of the Main Dam or Spillway.The purpose of the survey was to determine the general excavation conditions and depth of rock along the survey lines. A Dresser RS-4A seismograph unit was used in the investigation.An alignment of twelve geophones was used to register the arrival of direct and refracted waves; the wave arrival times were recorded on linagraph paper to provide a semi-permanent record. The twelve geophones used to pick up the direct and refracted waves were placed at approximately 20 ft. intervals over a controlled distance of up to 220 ft. Energy sources consisted of small "Kinepak" explosives detonated 5 ft. from the first geophone and slightly larger explosives detonated 100 ft. from the first geophone. Explosives were detonated at both ends of the alignment to record data in a forward and reverse direction. The long shots were made to detect arrivals of waves refracted from the higher velocity layers at depths that would not normally be detected by the short shots. The forward and reverse shots were required to permit an evaluation of the sloping and uneven boundaries of the subsurface layers of different velocities.The records obtained in the field were analyzed, and graphs were developed for initial compressive wave arrival time versus the distance from the explosions. Pertinent corrections were applied to adjust the data for sloping or undulating terrain. The corrected data were used to compute the velocity of propagation of the compressive wave in the various layers and the depth to the layer boundaries beneath each geophone. The depths were calculated by a computer and the results plotted on seismic profiles. These profiles are contained in CP&L files, but are not included in the FSAR because the seismic refraction results were superseded by detailed geologic mapping of rock formations in excavations of the main dam cutoff trench and spillway.2.5.6.2.1.2.4 Borehole Drilling and Sampling A series of boreholes were drilled in the main dam and spillway area for exploration and sampling of foundation materials for the Seismic Category I structures. The locations of the boreholes, designated as BM1 through BM51, are shown on Figures 2.5.6-9 and 2.5.6-10.Boring and sampling methods were in accordance with ASTM D-1586. A standard 2 in. outside diameter split barrel sampler with an inside diameter of 1 3/8 in. was used to obtain soil samples for soil classification and for laboratory tests. In soils, the standard penetration test was made at every change of strata and within strata at intervals not exceeding 5 ft. Undisturbed soil samples were obtained from cased borings of sufficient size to accommodate either a 3 in.diameter thin wall Shelby tube sampler, a 3 in. diameter piston sampler, or a Denison-type double-tube core barrel. Rock core borings were made through the casings used for soil borings. Core drilling was in accordance with ASTM D-2113. NQ wireline or NX coring was used to recover rock core samples approximately 2 in. in diameter. Rock cores were examined in the field, logged, and stored in standard core boxes for future use and testing. Logs of the foundation borings are presented in Appendix 2.5A.A series of boreholes, consisting mostly of uncased auger borings, were drilled to define a potential source of borrow materials (Borrow Area M) for use as impervious fill in the Main Dam.Twenty-five-pound bag samples of soil were collected from the borings for laboratory analysis.The locations of the boreholes (BB806 through 809, 811 through 815, 858 through 860, and 862 Amendment 65 Page 204 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 through 869) are shown on Figure 2.5.6-9. Borehole logs are included in Appendix 2.5A.Results of laboratory analyses are given in Appendix 2.5C.After construction of the Main Dam began, eighteen additional auger borings were made in Borrow Area M. The locations of the borings and the results of laboratory tests on soils collected from the boreholes are given in Appendix 2.5F, Figure 9.Six NX wireline borings and 38 air track percussion borings were made immediately north of the Main Dam to outline a quarry site for use as a source of rockfill for the Main Dam. The locations of the boreholes are shown on Appendix 2.5F, Figure 11 (Area "A"). The NX wireline borings were logged and RQD values were determined for the core. The logs of the boreholes are included in Appendix 2.5F.During construction of the Main Dam, NX core borings were drilled to a depth of 50 ft. on 40-ft.centers located 5 or 10 ft. upstream and/or downstream of the main dam centerline. The holes, which served as primary grout holes, were logged and RQD values were determined for the core in order to verify that foundation conditions were as expected. The locations of the boreholes are shown in Appendix 2.5E, Figure 12. Details of the borings are in Appendix 2.5E.2.5.6.2.1.2.5 Test pits and trenches Six test pits were excavated in Main Dam Borrow Aera M during PSAR investigations and an approximately 300=pound representative soil sample was obtained from each test pit for laboratory testing. Each sample contained the proper proportion of the different types of soil observed in the pit. The locations of these pits, designated as TPM1 through TPM6, are shown on Figure 2.5.6-9. Test pit logs are contained in Appendix 2.5A. Laboratory test results are included in Appendix 2.5C.During construction, twenty-eight test trenches were excavated by backhoe in Borrow Area M for further soil sampling. Twenty-five-pound bag samples of soil, representing a composite mix of the vertical profile of each trench, were collected by hand excavation. The locations of these trenches and the results of laboratory testing of the soil samples are shown on Appendix 2.5F, Figure 9.Further exploration for borrow materials was required because Borrow Area M was found to be undesirable as a source of impervious fill. Additional test trenches were excavated by backhoe or bulldozer in order to locate an alternative borrow area (Borrow Area W) northeast of Borrow Area M. The locations of the trenches and the results of laboratory tests on trench samples are included in Figure 10 of Appendix 2.5F.2.5.6.2.2 Auxiliary dam, dike, and channel areas 2.5.6.2.2.1 Local geologic features The sites of the Auxiliary Dam, Auxiliary Separating Dike, Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, Emergency Service Water and Cooling Tower Makeup Channel, and Auxiliary Dam Borrow Area Z are located in the Deep River Triassic Basin; they are underlain by sedimentary rocks of Triassic age. The rocks, like those underlying the plant site, belong to the lower part of the Sanford Formation. However, the Amendment 65 Page 205 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 strata underlying the dam and dike sites occupy a lower position in the stratigraphic succession, and consequently are older.2.5.6.2.2.1.1. Lithology The sedimentary rocks underlying the auxiliary dam area are characterized by four major lithologic units: medium to coarse-grained sandstone, fine-to medium-grained sandstone, siltstone, and shaley siltstone. These units grade into one another both laterally and vertically, and all intermediate combinations are present.The medium to coarse-grained sandstone is arkosic and ranges in color from light gray to chocolate brown as the content of silt and clay in the matrix increases. It is conglomeratic in part.The fine to medium-grained sandstone is typically chocolate brown, with silt and clay present in the matrix. In some places it is light-gray, clean, well sorted, and composed predominantly of quartz.The siltstone ranges in color from chocolate brown to dark gray or greenish gray. It is commonly mottled and contains pebbly, sandy, shaly, and/or carbonaceous beds.The shaley siltstone is chocolate to dark brown, green, purple, or gray. It is composed predominantly of silt and clay, although one recurring variety is pebbly.The sandstones are the most resistant to excavation and slaking and tend to form resistant ridges in the foundation area for the random rockfill zone of the Auxiliary Dam.The shaley siltstones are the least resistant to slaking and form a characteristic hackly weathered surface after about a week of surface exposure.2.5.6.2.2.1.2 Structure The Auxiliary Dam area is located near the eastern margin of the Durham Triassic Basin, which is bounded on the east by the northeast-trending Jonesboro Fault. Smaller normal faults, transverse to the Jonesboro Fault, are common in the basin. The plant site fault, as detailed in Section 2.5.3, crosses the Auxiliary Dam at Station 4+23, striking N 87° E and dipping 65° to 75° southeast. This fault has been demonstrated to be non-capable.The sedimentary strata exposed in the auxiliary dam area strike N5°-15°E with dips ranging from 9 to 17 degrees southeast. The two dominant joint sets are vertical, one strikes N 40°-50° E and the other N 20°-30° W. A third set trends north-northwest and dips 55° to 70° to the southwest.2.5.6.2.2.2 Exploration Programs Most of the exploration programs conducted for the plant site also included some or all of the auxiliary reservoir area. The initial engineering geology survey of the plant site and surrounding area, conducted by Dames and Moore in 1970, included test borings, trench excavations, and a seismic refraction survey along the axis of the Auxiliary Dam. The foundation exploration program for Seismic Category I structures included test borings in the foundations of the Amendment 65 Page 206 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Auxiliary Dam, Auxiliary Dam Spillway, Auxiliary Separating Dike, Auxiliary Reservoir Channel, and Emergency Service Water Intake and Discharge Channels. Test trenches and seismic wave velocity measurements were made in the auxiliary dam foundation. Exploration and sampling for the borrow area testing program included auger borings in Borrow Area X, located between the Auxiliary Dam and the Auxiliary Separating Dike (Figure 2.5.6-12) and auger borings and test pits in Borrow Area Z, located adjacent to the Auxiliary Dam Spillway (Figure 2.5.6-11). The Shearon Harris Site Fault investigation in 1974 (Reference 2.5.1-29) included exploratory trenches and borehole drilling in the Auxiliary Reservoir area a few hundred feet north of the auxiliary dam axis. Further drilling along the auxiliary dam axis and detailed foundation mapping in the Auxiliary Dam and Auxiliary Separating Dike excavations were performed during construction.2.5.6.2.2.2.1 Excavation mapping The foundation of the Auxiliary Dam was geologically mapped by using the general procedure described in Section 2.5.6.2.1.2.2. The foundation was excavated to suitable rock. Those areas where suitable rock was exposed in the core trench were hand cleaned and air blown prior to mapping. Mapping in the core trench was performed on a 10 ft. grid tied to the dam centerline at a scale of 1 in. to 10 ft. Mapping in the impervious core area outside the core trench was performed on a scale of 1 in. to 50 ft. on a 25-foot grid. The maps are presented on Appendix 2.5E Figures 7 and 8.Areas of the auxiliary dam foundation outside the core trench and impervious core area were excavated to weathered rock or firm residual soil with the blade of a D-8 tractor-dozer and sketch mapped on a scale of 1 in. to 50 ft. on a 25-ft. grid tied to the dam centerline (Appendix 2.5E, Figure 8).The foundation of the Auxiliary Reservoir Separating Dike was excavated to firm residual soil, as required by the specifications, in the shallower parts of the excavation, and to weathered rock in the deeper parts. After excavation, the foundation surface was cleaned by a bulldozer, mapped geologically at a scale of 1 in. to 50 ft. and then photographed. No faults or other unusual features were observed and no foundation treatment was required. A geologic map of the Auxiliary Reservoir Separating Dike foundation is shown on Appendix 2.5E Figure 9.The Auxiliary Reservoir Channel, Emergency Service Water and Cooling Tower Makeup Intake Channel, and Emergency Service Water Intake and Discharge Channels were excavated to design grade and slope by scrapers and dozers with rippers. Shallow blasting was required in portions of these channels. The channel excavations were inspected by a geologist; geologic mapping of the excavations was not required.2.5.6.2.2.2.2 Borehole drilling Borings were drilled to investigate the rock composition, orientation, and quality across the auxiliary dam site for the PSAR. Boring areas included the Auxiliary Dam Spillway, the auxiliary dam axis, and a line that extends several hundred feet east of the east abutment of the dam.The location of the borings, designated as D2 through D19, are shown on Figure 2.5.1-11.Truck and skid-mounted rotary wash, wire-line drilling rigs were employed in the boring program. Drillers used a 3 7/8 in. diameter rotary core barrel to penetrate and sample the thin layer of soils at each test boring location. When rock was encountered, a 1 3/4 in. double-tube core barrel and diamond drilling bit were used to advance the boring and collect continuous Amendment 65 Page 207 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 samples of rock. The borings were logged in detail by an engineering geologist; they are shown in Appendix 2.5A. The soils encountered were described in accordance with the Unified Soil Classification System. RQD values, which indicate the ratio of core which was 4 in. or more in length to the full core run, were calculated for each core length and indicated on the boring logs.Additional borings were drilled for exploration and sampling of the foundations for all Seismic Category I structures in the Auxiliary Reservoir. Drilling and sampling methods were the same as those described in Section 2.5.6.2.1.2.4. Boreholes BX1 through BX16 were drilled in the auxiliary dam foundation; BX40 through BX61 in the auxiliary dam spillway foundation, BD16 through BD20 in the Auxiliary Reservoir Separating Dike foundation, and BC45 through BC48, BC165, and BC166 in the Auxiliary Reservoir Channel. Locations of the borings are shown on Figures 2.5.6-11 and 2.5.6-12. Boreholes BC71 through BC74, BC116, BC17, BC120 through BC122, BC161, BC162, and BC179 through BC190 were drilled for the Emergency Service Water Intake Channel, and BC151 and BC170 through BC178 for the Emergency Service Water Discharge Channel. The locations of these boreholes are shown on Figures 2.5.6-7 and 2.5.6-

8. Logs of all of the foundation borings are contained in Appendix 2.5A. Additional boreholes, BC-191 through BC 199, were drilled for the Emergency Service Water and Cooling Tower Makeup Intake Channel, but the cores were not recovered. These boreholes provided rock depth information only since the cores were not recovered.

Thirty-two uncased auger borings (BB155 through BB186) were drilled in auxiliary reservoir Borrow Area Z in order to obtain 25-lb. bag samples of soil for laboratory testing. Four cased borings, BB5 through BB8, were also drilled in Borrow Area Z. The locations of the boreholes are shown on Figure 2.5.6-11; logs of the borings are included in Appendix 2.5A.Six additional boreholes were drilled during 1974 for test purposes in the Auxiliary Reservoir area as part of the Shearon Harris Site Fault Investigation. The locations of the borings, designated as TB-1-74 through TB-6-74, are shown on Figures 2.5.1-15 and 2.5.1-16. Water pressure tests were conducted in the three borings located near the Auxiliary Dam Spillway (TB-1-74, TB-2-74, and TB-3-74). The other three borings, located on the east side the reservoir between the Auxiliary Dam and the Auxiliary Reservoir Separating Dike, were used for overburden permeability tests. The borings are discussed further in Section 2.5.4.3.2; borehole logs and test results are contained in Reference 2.5.1-29.Additional NX core borings were drilled to a depth of 50 ft. on 40 ft. centers along the axis of the Auxiliary Dam during construction. These borings, which served as the primary holes for the grout curtain, were logged and RQD values were determined for the cores to verify that foundation conditions were as expected. Locations of the boreholes are shown on Appendix 2.5E Figure 14; borehole logs are maintained in CP&L files.2.5.6.2.2.2.3 Trenching and test pits During the PSAR investigation in 1970, two trenches with a combined total length of 4500 ft.were excavated in the auxiliary dam and spillway area in order to evaluate the lithology, quality, and continuity of the rock, and the composition and consistency of the overburden. One trench, approximately 3500 ft. long (identified as Trench No. 3), was excavated along the axis of the Auxiliary Dam. The other trench (identified as Trench No. 4) was excavated across the spillway area. The locations of the trenches are shown on Figure 2.5.1-11. Trenches were generally excavated to refusal to the backhoe equipment that was used over most of the area. Most trenches were 2 to 10 ft. in depth. All excavations were inspected in regular increments to Amendment 65 Page 208 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 evaluate overburden and rock conditions, and hand penetrometer tests were performed at intermittent intervals. Trench logs were prepared from the observations and are included in Appendix 2.5A.Two test trenches were excavated in the foundation for the Auxiliary Dam with a Case 580B backhoe for the purpose of obtaining undisturbed representative block samples of the dam's foundation soils. The location of the trenches (identified as TPA1 and TPA2) are shown on Figure 2.5.6-11. Four (identified as TPA1 and TPA2) are shown on Figure 2.5.6-11. Four undisturbed block samples were recovered from the test trenches; three from TPA1 and one from TPA2. The block samples were cut in-situ to approximately 1-ft. cubes, sealed with wax, and placed in wooden boxes. The trench logs are included in Appendix 2.5A.During construction of the channels, one excavation in the Emergency Service Water Intake Channel and two excavations in the Auxiliary Reservoir Channel were made to obtain undisturbed block samples of soil in order to further define in-situ residual soil properties. The locations of these excavations and laboratory test results on the samples collected from them are included in Appendix 2.5K.Twenty-two trenches were excavated in the auxiliary reservoir area in order to trace the westward extension of the Shearon Harris Site Fault, which is located parallel to, and about 700 ft. north of, the auxiliary dam axis. The trenches were oriented approximately north-south and ranged in length from less than 100 ft. up to 700 ft. Trenches were excavated with a Link-Belt LS-4000 crawler hydraulic backhoe with 1 1/4 cubic yard bucket. In general, trench depths ranging from seven to 13 ft. were required to provide a satisfactory definition of fault features.All but three of the trenches provided good to excellent exposures of the fault trace. The fault was not observed in trenches FET-9W, 17W, and 18W, where caving conditions, resulting from a combination of deep alluvium and groundwater, precluded obtaining desired exposures.Trench locations and trench wall sections are shown on Figure 2.5.3-4.Four test pits were excavated in auxiliary dam Borrow Area Z. An approximately 300-lb.representative soil sample was obtained from each test pit. Each sample contained the proper proportion of the different types of soil observed in the pit. The locations of the pits, designated as TPZ1 through TPZ4 are shown on Figure 2.5.6-11; the test pit logs are shown in Appendix 2.5A.2.5.6.2.2.2.4 Seismic refraction survey A seismic refraction survey was conducted to define the topography of the rock surface along the auxiliary dam axis and across the spillway. The location of the survey lines, identified as Seismic Lines Nos. 3A, 3B, and 4, are shown on Figures 2.5.6-11 and 2.5.6-12.The recording equipment used for the refraction investigation was an Electro-Tech, ER-75-12 transistorized, portable, 12-channel, refraction seismograph. Electro-Tech EV-5-4 geophones having a natural frequency of 14 cycles per second were used. Geophone spacings of 25 and 50 ft. were used. Explosives were detonated in drill holes approximately 10 ft. deep at the ends, center, and beyond the ends of each line.The results of the seismic refraction survey showing profiles of various strata, are presented on Figure 2.5.2-9. These results show compressional wave velocities in soils and underlying rock.The velocities of compressional wave propagation in the upper soils and underlying rock were Amendment 65 Page 209 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 computed from the plotted data. In addition to the interpreted cross-section, the plots of the time-distance data resulting from the survey are presented immediately above the corresponding profile.The accuracy of the calculated depth to rock is considered to be within 10 percent for the major portion of the survey; however, in the areas where rock is indicated at shallow depths, the precision is probably less.2.5.6.2.2.2.5 Seismic Wave Velocity Measurements Seismic wave velocity measurements were made at locations along the axes of the Auxiliary Dam and Auxiliary Reservoir Separating Dike. The locations where measurements were made are shown in Appendix 2.5C. The main purpose of the measurements was to determine compression-wave (P-wave) velocities (Vp), shear-wave (S-wave) velocities (VS), and Rayleigh-wave (R-wave) velocities (VR) of in-situ residual soil; in addition, the seismic wave velocity of the transitional material and the upper portion of the weathered rock were determined. Two methods were used to measure seismic wave velocities; (1) pulse arrival measurements of compression-wave (P-wave) velocity (Vp) and shear-wave (S-wave) velocity (VS) and (2) steady-state vibration measurements of Rayleigh-wave (R-wave) velocities (VR).2.5.6.2.2.2.5.1 Pulse Arrival Measurements of PWave andSWave Velocities Pulse arrival measurements were made by using a Sprengnether VS 1200 seismograph and a three-component geophone; a sledge hammer impact was used as the energy source. An Electro-Tech vertical geophone, located adjacent to the impact station, was used to provide zero time.Pulse arrivals were recorded for both vertical and horizontal impacts; several records were made at each location to examine the repeatability of each measurement. The three-component geophone recorded the propagated seismic waves in three planes at right angles.P-wave pulse arrivals were measured by the horizontal component of the geophone that was oriented along the line between the geophone and the impact station. S-wave pulse arrivals were measured by the two components of the geophone oriented perpendicular to the line between the geophone and impact station.To create maximum S-wave energy, horizontal impacts were oriented perpendicular to the direction of the measurement line; in addition, this minimized the P-wave energy. By reversing the impact direction, the S-wave was reversed; by comparing the two records, which were symmetrical with respect to the time axis, accuracy of the interpretation was increased.The impact-to-receiver distance was measured to an accuracy of +/-0.1 ft; the time of the first pulse arrival was scaled from the records to an accuracy of +/-1 millisecond (msec). Velocities were calculated by dividing the impact-to-receiver distance by the time.At the auxiliary dam locations, P-wave and S-wave velocity measurements were made for impact receiver spacings of (1) 15 ft. and 20 ft. in the residual soil layer; (2) 10 ft., 12.7 ft., 20 ft.,and 22 ft. in the transitional material; and (3) 12 ft. and 26 ft. in the weathered sandstone. At the auxiliary reservoir separating dike locations, P-wave and S-wave velocity measurements were made for impact-receiver spacings of 25 ft. for residual soil and 20 ft. for weathered rock.Amendment 65 Page 210 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The results of the compressional wave and shear wave velocity measurements are presented in Appendix 2.5C.2.5.6.2.2.2.5.2 Steadystate vibration measurements of Rwave velocities Steady-state vibration measurements were made by using a Heathkit audio generator (1 G-72),a Dyna Kit Mark III preamplifier, and a Goodman vibrator to generate R-waves. The velocities of the R-waves were measured by using two Electrotech EV-17 vertical geophones; their response was observed on a Tectronix R-5030 dual beam oscilloscope.R-wave velocities were measured by determining the frequency required to create an in-phase response of two geophones spaced at a selected distance. The frequency was varied in increments of one hertz by the audio generator. The geophone response was displayed on the oscilloscope screen, and the in-phase response was determined to an accuracy of +/-0.5 hz for frequency ranges of approximately 30 hz to 110 hz, and +/-5 hz above frequencies of 110 hz.The distance between geophones was measured to an accuracy of +/-0.1 ft.Results of the Rayleigh wave velocity measurements are presented in Appendix 2.5C.2.5.6.3 Foundation and Abutment Treatment Foundations for the Main Dam, Auxiliary Dam, Auxiliary Reservoir Separating Dike, and the four channels (Emergency Service Water Intake and Discharge Channels, Emergency Service Water and Cooling Tower Makeup Channel, and Auxiliary Reservoir Channel) were excavated and treated in accordance with project specifications (Appendix 2.5I). For control of seepage through the foundations of the Main and Auxiliary Dams, a cutoff trench was excavated in the foundation to suitable rock (defined as rock material that cannot be moved on a production basis with a single-tooth ripper of a D-8 tractor or equivalent) and a grout curtain was emplaced for each dam.After approval of foundation reaches by the NRC, open fractures were filled with slush grout and dental concrete was placed to modify steep slopes.Grouting in the cutoff trenches at both dams was completed with neat cement in accordance with project specifications. Consolidation holes were drilled to 20 ft. depths on 10 ft. centers.Primary curtain holes were cored and logged to 50 ft. depths on 40 ft. centers. Secondary curtain holes without coring were then splitspaced between the primary holes. All curtain holes were water tested in five stages and then grouted in a minimum of three stages. Check holes were grouted for both consolidation and curtain grouting as required to assure closure. Grout takes were very low at both dams, as anticipated.Details of the foundation excavation and treatment, including foundation maps, borehole plans and sections, grout takes, mixes, and pressures are given in Appendix 2.5E. Summaries for the Main Dam, Auxiliary Dam, Auxiliary Separating Dike, and the three Seismic Category I channels are given below.2.5.6.3.1 Main Dam The Main Dam is founded on granite and gneisses, with interlayered schists, as described in Section 2.5.6.2.1.1. The foundation of the dam was excavated to weathered rock (defined as Amendment 65 Page 211 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 material which cannot be removed on a production basis with the blade of a D-8 tractor-dozer) and the cutoff trench (core trench) was excavated to suitable rock, which consists mostly of fresh to slightly weathered crystalline rock. Approximately 300,000 cubic yards of material were excavated to expose the foundation.After initial excavation, the foundation was cleaned and mapped geologically, as described in Section 2.5.6.2.1.2.2. Foundation geology of the Main Dam is shown on Appendix 2.5E Figures 4, 5, and 6. A number of minor faults and anomalous features, with offsets measured in inches, were noted during mapping. The faults, which are discussed in more detail in Section 2.5.6.2.1.2.2, were determined to be non-capable after suitable study by Ebasco geologists and inspection by NRC staff geologists.In general, grout takes for both consolidation and curtain holes were very low, averaging 0.02 bags per ft. of hole for consolidation holes and 0.03 bags per ft. of hole for curtain holes. A total of 24,403 linear ft. of holes was drilled, which required a total of 1,424 bags of cement for grouting and backfill. Grout hole locations for the Main Dam are on Appendix 2.5E Figures 11 and 12.2.5.6.3.2 Auxiliary Dam The Auxiliary Dam is founded on gently-dipping sedimentary rocks. The foundation of the dam was generally excavated into weathered rock. The impervious core is founded on weathered rock and the cutoff trench (core trench) was excavated to suitable rock. The filters and random rockfill shells are founded on weathered rock at the center of the dam and on firm residual soils near the abutments. Approximately 402,000 cubic yards of material were excavated to expose the foundation of sedimentary rock beneath the dam's shell and to provide for the cutoff trench.The foundation was cleaned and mapped geologically, as described in Section 2.5.6.2.2.2.1.Foundation geology and foundation elevations for the Auxiliary Dam are provided on Appendix 2.5E Figures 7 and 8. The plant site fault, exposed as anticipated at Station 4+23, was the only unusual feature encountered in the excavation. This fault, which was demonstrated to be non-capable, is discussed in detail in Section 2.5.3. It was treated by a special grout pattern, as detailed in Appendix 2.5E.Total drilling for consolidation and curtain grouting was 32,630 linear ft., which required a total of 2,652 bags of cement for grouting and backfill. Grout takes were low, averaging only 0.06 bags per ft. of hole for the curtain. Grout hole locations for the Auxiliary Dam are shown on Appendix 2.5E Figures 13 and 14.2.5.6.3.3. Auxiliary separating dike The Auxiliary Separating Dike is founded on gently-dipping sedimentary rocks. The geologic map of the foundation for the Auxiliary Separating Dike indicates that most of the excavation extended through firm soil into weathered rock. No faults or other unusual features were noted.Neither cutoff trench nor grouting was required because there is no differential head, as the water level is the same on both sides of the dike. The exposed foundation materials were mapped to verify that they are equal to, or better than, what was assumed for design. Prior to embankment placement, the entire foundation was proof-rolled to assure stability.Amendment 65 Page 212 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.3.4 Channels Each of the four Seismic Category I channels were excavated to design grade and slope.Subsequent foundation inspections by geologists indicated that neither geologic mapping nor special treatment were required. Fill sections in the Emergency Service Water Intake and Discharge Channels were constructed by using impervious fill. Moisture, density, and permeability controls were utilized to ensure the integrity of the fill sections (see Specification CAR-SH-CH-4, Appendix 2.5I).2.5.6.4 Embankments 2.5.6.4.1 General The locations of the two Seismic Category I dams, the Seismic Category I Auxiliary Separating Dike, and the four Seismic Category I channels (the Emergency Service Water Intake and Discharge Channels, Emergency Service Water and Cooling Tower Makeup Channel, and the Auxiliary Reservoir Channel) are shown on Figure 2.5.1-10.2.5.6.4.1.1 Main dam The Main Dam is approximately 1550 ft. long and has a maximum height of 108 ft. The general plan, typical cross section, and longitudinal profile of the Main Dam are shown on Figures 2.5.6-1 and 2.5.6-2. The Main Dam has a core of compacted silty clay and clayey silt material. The core is protected on each side by 8-ft. thick fine and 8-ft. thick coarse filter zones, and a rockfill shell. The core is founded on suitable rock and the rockfill shell is founded on weathered rock.Riprap slope protection placed on crushed rock bedding is provided on the upstream face, as shown on Figure 2.5.6-2. The downstream face is protected by an oversized rock zone, as shown on Figure 2.5.6-2.The rock at the foundation of the Main Dam consists of granite, gneisses, and schists, as discussed in Section 2.5.6.2.2.5.6.4.1.2 Auxiliary dam The Auxiliary Dam is approximately 3903 ft. long and has a maximum structural height of approximately 72 ft. The general plan, typical cross sections, and longitudinal profile of the Auxiliary Dam are shown on Figures 2.5.6-3 and 2.5.6-4. The Auxiliary Dam has a core of compacted silty clay and clayey silt material protected on each side by a single transition filter zone and random rockfill shell; there are two downstream blanket drains in the shell and two 200 ft. wide, 3 ft. thick drainage layers in areas where preexisting creeks had been located.Riprap slope protection placed on crushed rock bedding is provided, as shown on Figure 2.5.6-

4. The core of the dam is founded on weathered rock and the core (cutoff) trench is excavated to suitable rock. The filters and random rockfill shells are founded on weathered rock at the center of the dam and on firm residual soil near the abutments.

The rock at the foundation of the Auxiliary Dam consists of Triassic sandstones and siltstones.Amendment 65 Page 213 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.4.1.3 Auxiliary reservoir separating dike The Auxiliary Reservoir Separating Dike is approximately 1200 ft. long and has a maximum height of approximately 55 ft. A typical cross section, longitudinal profile, excavation plan, and fill plan of the Auxiliary Reservoir Separating Dike are shown on Figure 2.5.6-5. The dike has a core of compacted silty clay and clayey silt which is protected by a random rockfill shell. The rockfill shell is graded with the finer material placed adjacent to the core and the coarser particles placed to the outside. The core and rockfill shell are founded on weathered rock or locally on stiff residual soil overlying weathered rock. Riprap slope protection placed on crushed rock bedding is provided, as shown on Figure 2.5.6-5.The engineering properties of the rock and residual soil at the foundation of the Auxiliary Reservoir Separating Dike are essentially the same as those at the foundation of the Auxiliary Dam (see Section 2.5.6.2).2.5.6.4.1.4 Channels The profiles and typical sections of the Auxiliary Reservoir Channel and Emergency Service Water Channels are shown on Figures 2.5.6-6, 2.5.6-7, 2.5.6-8, and 2.5.6-28.The Seismic Category I channels were constructed mainly by excavation either into residual soil or into rock. Portions of the slopes of the Emergency Service Water Intake Channel were shaped to grade by backfilling with random fill and/or modified random fill materials. Portions of the slopes of the Emergency Service Water Discharge Channel were shaped to grade by backfilling with random fill material. The side slopes of the channels are two horizontal to one vertical in soil and one horizontal to four vertical in rock.2.5.6.4.2 Material properties and placement 2.5.6.4.2.1 Fill materials The fill materials used for construction of the Main Dam, Auxiliary Dam, Auxiliary Reservoir Separating Dike, and Seismic Category I channels are in accordance with established project specifications (Appendix 2.5I).The criteria for the material used in the core of the dams are that 40 percent of the material passes the No. 200 sieve and the plasticity index of the placed material is greater than ten. In order to provide the specified materials, an extensive borrow investigation was performed, as described in Section 2.5.6-2. The logs for the auger borings in Borrow Area Z are shown in Appendix 2.5A. Grain size analysis, Proctor compaction test results, and triaxial shear test results are shown in Appendix 2.5C. The logs for the test pits and test trenches in the borrow areas are shown in Appendix 2.5A. The grain size analyses and Proctor compaction test results for representative test pits are shown in Appendix 2.5C.The results of testing at the Main Dam core material from Borrow Area W are shown on Appendix 2.5F Figure 10.The results of laboratory investigations of the static and dynamic properties of the core material for the Auxiliary Dam and Auxiliary Separating Dike are presented in Appendices 2.5C and 2.5D.Amendment 65 Page 214 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 An additional laboratory testing program was conducted to evaluate the use of Material Z in portions of the auxiliary dam core, compacted at a moisture content as high as three percent above the optimum moisture content (see Section 2.5.6.9). This program included static drained and undrained shear strength tests, cyclic shear strength tests, and cyclic properties tests. The results are shown in Appendix 2.5C. These tests were performed on materials compacted to approximately 97 percent Standard Proctor Density at a moisture content of four percent above optimum moisture.The test results indicated that for the particular locations in the auxiliary dam core, this material could be compacted at a moisture content as high as four percent above optimum moisture and still exhibit the required static and dynamic properties necessary for dam stability.The test results, therefore, provided justification for increasing the required core moisture content during compaction to three percent above optimum moisture.2.5.6.4.2.2. Foundation Materials From samples of representative borings made at the auxiliary dam and spillway sites, grain size analyses have been performed on surface soils, and unconfined compression tests and laboratory determination of compressional wave velocities have been made on rock samples.The tests are listed in Table 2.5B-2 and the results are shown in Table 2.5B-4 and pages 2.5B-25 through 2.5B-41, and 2.5B-66 through 2.5B-80 of Appendix 2.5B.Selection of the average to most conservative soil samples was based upon determining their properties for use as foundation materials under the dam shell or for use as fill materials. The tested rock samples were selected to obtain average to conservative rock properties of in-situ rock.From borings made at the main dam site, grain size analyses were performed on the surface soils, and unconfined compression tests and laboratory determination of compressional wave velocities were made on rock samples. The tests are listed in Table 2.5B-1 and the results are shown in Table 2.5B-3 and pages 2.5B-5 through 2.5B-24 and 2.5B-42 through 2.5B-65 of Appendix 2.5B.Water pressure tests were performed and the results for boring BM-8 through BM-10 and BM-13 through BM-16 are shown in Appendix 2.5A.The results of the laboratory investigation of the static and dynamic properties of the foundation soil for the Auxiliary Dam and Auxiliary Separating Dike are shown in Appendices 2.5C and 2.5D.2.5.6.4.3 Placement Requirements All embankment materials, including those for the dams, dike, and channel slopes, were placed in accordance with project specifications shown in Appendix 2.5I.Field testing and construction control procedures used to assure that the required soil properties were obtained are included in the specifications for the dams and embankments. Quality control and quality assurance compliance requirements are documented in the plant's excavation and backfill procedure.Amendment 65 Page 215 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Grain size distribution tests before and after compaction were performed to determine breakdown of the rockfill and random rockfill materials. In-place density and permeability tests were conducted after compaction. These tests verified that, in the test fills, the design strength and permeability that were assumed for the rockfill in the analysis provide a sufficient margin of safety. Strength tests were performed by the U. S. Army Corps of Engineers on test fill materials. Compaction procedures, equipment, and techniques established by the in-place test fill program were used during actual dam construction. The results of the programs were submitted to the NRC for review and approval prior to the start of actual dam construction; the results are included in Appendix 2.5H.The maximum compacted lift thickness for the impervious core was 8 in. and for the transition filter zone materials, it was 16 in. During construction of the filters, compaction tests were taken to assure compliance with the specifications.The specification allowed a maximum compacted lift thickness of 2 ft. for the rockfill and random rockfill shell materials; the actual maximum lift thickness was 2 ft. as determined by an in-place test fill program.In-place test fill programs were performed to establish a method of construction that yielded effective and efficient compaction of rockfill and random rockfill materials. To accomplish the objectives, studies of roller passes and lift thicknesses of the random rockfill and rockfill material were performed. The test fills were a minimum of three lifts in height and were constructed by using vibratory rollers having a dynamic force of not less than 40,000 pounds.2.5.6.4.4 Gradation Requirements and Compaction Criteria Representative laboratory tests were performed to determine compaction criteria for all engineered backfill.The material in the impervious core of the Main Dam was compacted to 97 percent Standard Proctor density at plus or minus 2 percent of optimum moisture content. The lifts of the impervious core of the Main Dam shown in Appendix 2.5F Figure 14 were compacted to 100 percent Standard Proctor at a moisture content between optimum and plus 4 percent. The impervious core of the Auxiliary Dam below Elevation 225 ft. was compacted to 97 percent Standard Proctor density at plus or minus two percent of optimum moisture content. The impervious core of the Auxiliary Dam above Elevation 225 ft. was compacted to 97 percent Standard Proctor density at plus three to minus one percent of optimum moisture content. The Auxiliary Separating Dike central core was compacted to 97 percent Standard Proctor density, below Elevation 220 ft. and 100 percent Standard Proctor density above Elevation 220 ft. at plus or minus two percent of optimum moisture content. The results of the tests for the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike are discussed in Section 2.5.6.4.2.1.Compaction criteria for filter materials were determined on the basis of static and dynamic analyses and their ability to perform satisfactorily as a filter. Static and dynamic analyses indicated that the relative density of filter materials should be equal to or greater than a certain value, as discussed in the following paragraphs, to assure adequate stability under static and dynamic loading conditions. Over-compaction of filter materials was avoided.In dams with wide sloping cores (Main and Auxiliary Dams), the transition filters do not form a structural member to the same extent as in dams with narrow cores, since the transition filters Amendment 65 Page 216 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 are held in place by the weight of the rock or random rock shells. The criteria for compacting filter layers have been reported by Newmark (Reference 2.5.6-2) and Terzaghi and Lacroix (Reference 2.5.6-3).On the basis of the dynamic analysis presented in Appendix 2.5D, average relative density criteria were determined for the filters. The fine and coarse filters for the Main Dam were specified to be compacted to an average relative density of 75 percent, except for the upstream coarse filter above Elevation 220 ft. which was specified to be compacted to an average relative density of 80 percent. The filters for the Auxiliary Dam were specified to be compacted to an average relative density of 75 percent below Elevation 220 ft. and to an average relative density of 80 percent above Elevation 220 ft. For both dams, a five percent increase in the specified average densities was permitted during construction. The actual test results were analyzed statistically.The criteria was determined on the basis that the filters would be constructed by using satisfactory and uniform materials and methods, that construction would be continuously inspected, and that an adequate number of relative density tests, in accordance with the provision of ASTM D2049, "Relative Density of Cohesionless Soils", would be made. For relative density determination, the minimum density was determined in accordance with ASTM D2049. Maximum density was determined either by test fills that utilized actual construction equipment and placement procedures, or in accordance with ASTM D2049, whichever gave the higher density.In order to assure the design intent of the dynamic analysis and to establish a practical and workable construction criteria, the test results requirement was that only ten percent of the test densities were allowed to fall below the specified requirement, with no single value below 90 percent of the specified requirement (i.e., 67 percent for 75 percent specified; 72 percent for 80 percent specified) and that no filter (excluding blanket drains) had more than five percent of the test densities between 90 and 95 percent relative density. Therefore, the actual relative density values achieved in the construction of the dams exceeded the specified average relative density values. This procedure enabled the attainment of the densities necessary to ensure adequate embankment strength and filter flexibility while minimizing settlement. Thus, the filter performs its function of stability during dynamic conditions, and also as a transition filter should any healing of cracking in the core be necessary.The filters used for the dams consist of processed, very well graded, coarse, cohesionless mixtures of hard, dense, durable rock materials. In the Main Dam, the fine filter materials are less than 1/2 in. in size, with a maximum of 13 percent passing the No. 200 sieve; the coarse filter materials are less than 6 in. in size, with a maximum of 21 percent passing the No. 4 sieve.In the Auxiliary Dam, the transition filter materials are less than 3 in. in size, with a maximum of 15 percent passing the No. 200 sieve. Gradation criteria over the full range of particle sizes are provided in the Project specifications (Appendix 2.5I) for all of the filter zones.The gradation criteria for rockfill used in the Main Dam were that a minimum of 75 percent of the material ranged in size from 1/4 in. to 22 in. The maximum rock size did not exceed a 22 in.intermediate dimension and was not greater than 90 percent of the lift thickness. The gradation criteria for random rockfill used in the Auxiliary Dam and Auxiliary Reservoir Separating Dike were that a minimum of 75 percent of the material ranged in size from No. 10 Sieve to 22 in.The maximum rock size was 90 percent of the lift thickness that was determined by the test fill programs.Amendment 65 Page 217 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The stability analyses of the Seismic Category I dams and dike presented in Appendix 2.5D have been made by using the expected constructed values of the static and dynamic properties of the locally available materials with reasonable variations in the properties. Breakdown of the rockfill material caused by compaction, especially of the random rockfill material for the Auxiliary Dam and Auxiliary Reservoir Separating Dike, was taken into account when both the static and dynamic properties were selected. The selection of the strength parameters were based on laboratory test results and experience with rockfill and random rockfill materials, as well as on literature values of similar gradations of materials. Literature (References 2.5.6-4, 2.5.6-5, 2.5.6-6, and 2.5.6-7) on similar materials within the specified gradations show that the selected strength parameters are conservative and justifiable without deleting the minus 2 in. materials.The random rockfill material used in the Auxiliary Dam and Auxiliary Reservoir Separating Dike consists of sedimentary rocks which were expected to break down during handling and compaction. The low strength properties utilized for these materials considered the breakdown characteristics. In addition, three horizontal drainage blankets, each 3 ft. thick, are provided in the downstream shell of the Auxiliary Dam. The blankets are connected to the transition filter zone adjacent to the core of the dam. The lowest blanket drains into the existing creek. The upper two blanket drains connect to the downstream riprap bedding material (see Figures 2.5.6-3 and 2.5.6-4). The transition filter blankets provide collection and positive drainage of internal embankment seepage.As previously stated in Section 2.5.6.4.3, the maximum compacted lift thickness for the rockfill and random rockfill shell materials was determined on the basis of in-place test fill programs.The test fill sections were constructed by using vibratory rollers having a dynamic force of not less than 40,000 pounds. The maximum compacted lift thickness of the shell materials was 2 ft.Tests were performed on the test fill sections to demonstrate that the specified gradation for rockfill was achieved. The test results were submitted to the NRC for their review and approval prior to the start of construction of the dams. The triaxial shear strengths of the rockfill and random rockfill materials were measured by the U. S. Army Corps of Engineers using the results of tests on test fill compaction materials. Large diameter test specimens (15 in.) were used.The rockfill and random rockfill both exhibited shear strength parameters of: friction angle () =40 degrees and cohesion, (C) = 0 psf.The rockfill and random rockfill surfaces of the dam and dike structures are protected by a layer of riprap in the areas of wave action. The riprap is sized to withstand the wave forces on each structure. The riprap provided to protect the rockfill and random rockfill is bedded on crushed rock. Bedding materials are placed beneath the riprap and are graded to prevent movement of the bedding materials. Riprap and bedding requirements are defined by project specifications (Appendix 2.5I).2.5.6.4.5 Slope Protection In the surface areas other than where riprap is used (the wave action zones), rocks ranging up to the dimensions of Class A riprap were placed.The crest of the Main Dam is at Elevation 260 ft. The maximum wave runup in the Main Reservoir is discussed in Section 2.4.5. The upstream face of the dam is protected from the dynamic wave forces resulting from the maximum wave runup and wind setup level by a 4 ft.thickness of riprap. The downstream face of the Main Dam is protected by a layer of oversized rock, as indicated on Figure 2.5.6-2.Amendment 65 Page 218 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The crest of the Auxiliary Dam is at Elevation 260 ft. Both faces of the Auxiliary Dam are protected against the most severe wave action by a four ft. thickness of riprap.The crest of the Auxiliary Separating Dike is at Elevation 255 ft.; both faces of the dike from Elevation 235 to the crest are protected by a 4 ft. thickness of riprap.The design basis of wave protection is that recommended by Karl F. Taylor in his paper "Slope Protection on Earth and Rockfill Dams," presented in the 11th International Congress on Large Dams (Reference 2.5.6-14). The average size of riprap is based on the following formula:0.388 (b cot )3/5 =where: W50 = Average stone weight (lbs.)H = Wave height (ft.) (Table 2.4.5-1)L = Wave length (ft.)d = Depth of water at the toe of slope (ft.)

 = Angle of slope from the horizontal a = 0.20 empirical factors related to cot b = 0.75 The gradation criteria of the riprap material limits the weight of the maximum size rock to about 4 W50 (W50 is the average stone weight) and the weight of the minimum size rock to 1/4 W50.

The material consists of hard and dense sandstone, conglomerate, or granitic rock fragments.The service life was evaluated by conducting sodium sulfate tests to assist in the determination of rock weathering or rock deterioration potential.The riprap is founded on a bedding of crushed rock graded to prevent movement of the bedding material into or through the riprap. The bedding material gradation is based upon the filter criteria given in Reference 2.5.6-8.Specifications were used to assure that the riprap was carefully placed so that the rocks form an interlocked rough surface. Inspection during construction ensured compliance with the above specifications. The length to width ratio of the riprap is controlled by the specifications. The specified maximum length to width ratio is 3.3 to 1. For each type of riprap used, the maximum size, minimum size, and average size were specified, as well as a required percentage being within a size range.The Auxiliary Dam was designed to withstand the static and dynamic water pressure forces resulting from the maximum wave runup and wind setup on one face of the dam coincident with a wave trough on the opposite face.The Main Dam was designed to withstand the static and dynamic water pressures resulting from the maximum wave runup and wind setup on the upstream face of the dam.Amendment 65 Page 219 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The Auxiliary Separating Dike was designed to withstand the static and dynamic water pressure forces resulting from the maximum wave runup and wind setup.The channels were designed to withstand the static and dynamic water pressures resulting from the maximum wave runup and wind setup. Any local erosion of the side slopes of the channel due to wave action will not reduce the capability of the channel to sufficiently perform its function.The embankment of the plant island along the Main Reservoir is protected by sacrificial spoil fill, as shown on Figure 2.4.1-2. Further discussion is presented in Section 2.4.5.2.5.6.5 Slope Stability 2.5.6.5.1 General The slope stability of the SHNPP earth and rockfill Main Dam, Auxiliary Dam, Auxiliary Reservoir Separating Dike, and Channels was evaluated by using the results of field exploration, laboratory testing, and analytical study.The Main and Auxiliary Dams are zoned embankment dams constructed of three materials.The impervious core consists of compacted silty clay and clayey silt, the filters are composed of compacted granular materials, and the dam shells consist of compacted rockfill or random rockfill. Typical main dam and auxiliary dam cross sections are illustrated on Figures 2.5.6-2 and 2.5.6-4, respectively. The Auxiliary Reservoir Separating Dike consists of an impervious core with a random rockfill shell. Figure 2.5.6.5 shows a typical cross section of the Auxiliary Reservoir Separating Dike.The Main and Auxiliary Dams, the Auxiliary Separating Dike, the Auxiliary Reservoir Channel, the Emergency Service Water Intake and Discharge Channels, and Emergency Service Water and Cooling Tower Makeup Channel are designed as Seismic Category I structures. The side slopes are designed to provide adequate factors of safety under static and dynamic loadings.The minimum factor of safety against slope stability failure of the dams under static conditions is 1.5. The dams are also designed to a factor of safety of 1.2 for simultaneous OBE and 100 year return period flood, and to a factor of safety of 1.1 for simultaneous SSE and 25 year return period flood. In addition to the slip circle analysis, a two dimensional finite element model was used to evaluate the stability of the dams under dynamic loading conditions.The seismic design of the dams accounts for hydrodynamic pressures due to horizontal earthquake loads by increasing the hydrostatic loads as described by Zangar and Haefeli (Reference 2.5.6-9).Profiles and typical sections of the channels are shown on Figures 2.5.6-6, 2.5.6-7, 2.5.6-8, and 2.5.6-28. The Auxiliary Reservoir Channel, Emergency Service Water Intake and Discharge Channels, and Emergency Service Water and Cooling Tower Makeup Channel were constructed by excavating into hard residual soil or bedrock. However, portions of the slopes of the Emergency Service Water Intake Channel were shaped to grade by backfilling with random fill and modified random fill materials. The Emergency Service Water Discharge Channel was predominantly cut through rock. The earthen slopes were designed to a minimum factor of safety of 1.5 for static loads and 1.1 for dynamic SSE loading, at any water level.Amendment 65 Page 220 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.5.2 Field exploration Field exploration of the foundation conditions at the locations of the Main Dam, Auxiliary Dam, Auxiliary Separating Dike, and Category I channels is described in Section 2.5.6.2.2.5.6.5.3 Laboratory Testing The laboratory testing program included index property and compaction tests of individual and composite bulk samples of the impervious core materials for the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike. Based on the results of these tests, representative composite bulk samples from the borrow area for the Main Dam and representative composite bulk samples from the borrow area for the Auxiliary Dam and Auxiliary Separating Dike were prepared.Reconstituted specimens from each composite bulk sample were prepared at the density and moisture content equal to the field placement density and moisture content. The following tests were made on the reconstituted samples: static triaxial, cyclic torsion, and cyclic triaxial. In addition, index property and cyclic triaxial tests were performed on undisturbed samples of the foundation soils from the auxiliary dam and auxiliary separating dike areas.The results of laboratory investigation of the static and dynamic properties of the core materials for the Main and Auxiliary Dams and Auxiliary Separating Dike and the foundation soils for the Auxiliary Dam and Auxiliary Separating Dike are presented in Appendix 2.5D. Supplementary test results for the material from Borrow Area Z, used as core material for the Auxiliary Dam, are discussed in Appendix 2.5C. Results of tests on material from Borrow Area W, used as core material for the Main Dam, are shown in Appendix 2.5F.Static consolidated drained, consolidated undrained, and unconsolidated undrained triaxial tests provide values of the static strength characteristics. The cyclic torsion tests provide values of modulus and damping at very low levels of strain. Cyclic triaxial tests, made at low and intermediate levels of strain (by using controlled strain testing procedures), yield additional data on modulus and damping. Cyclic triaxial tests were made at various levels of strain (by using controlled stress testing procedures) to obtain cyclic strength characteristics, and modulus and damping values at the strain levels.Based on the results of cyclic torsion and triaxial tests, curves of modulus versus strain and damping versus strain were constructed. Curves of the most probable upper and lower bound values were also established. This is schematically illustrated on Figure 2.5.6-13.Cyclic triaxial tests were performed on compacted silty clay and clayey silt specimens of dams and dike core materials. These controlled stress cyclic triaxial tests were conducted for an appropriate range of initial effective confining pressure, 3c, and for three initial effective principal stress ratios Kc. The results of the tests were utilized to establish the shear stress required to cause five percent strain (in five and ten cycles) as a function of the normal effective stress. The duration of the safe shutdown earthquake (SSE) determined the appropriate number of cycles, (i.e., five and ten). Typical cyclic strength characteristics are illustrated schematically on Figure 2.5.6-14.Similar tests were also conducted on selected undisturbed samples of the foundation soil at the location of the Auxiliary Dam and Auxiliary Reservoir Separating Dike. The modulus, damping values, and cyclic strength characteristics of the filter material and the rockfill were based on Amendment 65 Page 221 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 available data for similar material, as discussed in Appendix 2.5D. Appropriate variations of the values were incorporated in the analyses.Static rockfill properties that were used in the analyses for the dams and dike described in Appendix 2.5D were based on data published in the literature. The values of these static properties were proved to be reasonable, based on the results of subsequent laboratory testing, as described in Appendix 2.5H.The laboratory testing program for the foundation materials and linings of the channels consists of classification, grain size analysis, and Proctor compaction tests on the disturbed residual soil samples from auger borings, and triaxial shear test on the recompacted samples. A similar compaction effort was exerted on samples of modified random fill.To further define in-situ residual soil properties, undisturbed block samples, approximately one cubic foot in volume, were obtained from the Emergency Service Water Intake and Auxiliary Reservoir Channels. A series of unconsolidated undrained, consolidated drained, and consolidated undrained triaxial shear tests were performed to develop drained and undrained shear strength parameters.Additional bag samples of channel lining material were tested for triaxial shear strength at moisture contents in excess of those specified for field placement. In this manner, very conservative shear strength parameters were obtained for stability analysis.The test results are presented in Appendix 2.5K.Based on the laboratory tests, the residual soil in the channel areas was classified to be fine to medium sandy clayey silt.A detailed study of soil shear strength and standard penetration blow counts of the standard penetration tests from the boring logs was performed by using correlation proposed by Peck, Hansen, and Thornburn (Reference 2.5.6-10). Figure 2.5.6-15 shows the blow counts of the standard penetration tests versus depth in the in-situ residual soil; blow counts and soil shear strength increase sharply with depth. The properties of in-situ residual soil in the foundation of channels are similar to those in the foundation of the Auxiliary Dam and Auxiliary Reservoir Separating Dike, as described in Appendix 2.5D.2.5.6.5.4 Analysis Procedures The stability of the Seismic Category I dams and dike was determined by the slip circle method and the finite element method. In establishing the stability of the Auxiliary Dam, the sliding wedge method was also used to evaluate the stability of the dam. The stability of Seismic Category I channels was determined by using the slip circle method.2.5.6.5.4.1 Static and Pseudo-Static Stability Analysis of Dams, Dike and Channels The static and pseudo-static stability of the dams, dike, and channel slopes was determined by a computer program that utilized the simplified Bishop slip circle method. The computer program is outlined in the ICES-LEASE-1 program user's manual (Reference 2.5.6-11).Amendment 65 Page 222 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 LEASE-1 is a sub-system of ICES designed to perform stability analysis of slopes by the method of slices. The failure surfaces are assumed to be circular arcs (Reference 2.5.6-12).The computer program locates the radius that has a minimum factor of safety at each of a specified set of trial centers. Depending on the conditions to be modeled, either drained or undrained strength parameters were used.LEASE-1 has the ability to include the effects of seismic forces through seismic commands.Horizontal and vertical seismic coefficients are applied and the factors of safety are evaluated under seismic loads. The factor of safety is defined as the ratio of the moment of the available shearing forces on the trial surface to the net moment of the driving forces.The basic assumptions of this method of analysis are:

1. The soil behaves as a Mohr-Coulomb material.

2. The factors of safety of the cohesive component of strength and the frictional component of strength are equal.

3. The factor of safety is the same for all slices.

The factors considered in the analyses include:

1. Properties of soil on the failure surface at the base of the slice, including unit weight, cohesion, and angle of internal friction.

2. Reservoir water levels and piezometric data.

3. The inclination of the failure surface at the bottom of the slice.

4. The dynamic acceleration due to an earthquake as input as an additional static load in the pseudo-static analysis.

Sliding wedge analyses were made for the Auxiliary Dam in order to verify the sliding stability in the abutment areas where a thin horizontal layer of material with low strength exists within the weathered rock. The layer of low strength material was assumed to be in the most vulnerable location with respect to sliding.The sliding wedge stability analysis involves an active soil wedge being mobilized against a neutral horizontal block and a passive resisting wedge. The factor of safety is calculated as the ratio of the sum of the resisting forces in the horizontal direction to the sum of the driving forces in the horizontal direction. This procedure is outlined in Navy Design Manual DM-7 (Reference 2.5.6-13).During the initial plant design, the static and pseudo-static stability of the dams, dikes and channel slopes was determined by a computer program, ICES-LEASE 1, which utilized the simplified Bishop Slip Circle Method. STABL 5M was selected as the computer program to reanalyze the Emergency Service Water and Cooling Tower Makeup Intake Channel (Reference 2.5.6-15). STABL 5M determines the factor of safety against slope instability by the method of slices (simplified Bishop-Circular Shaped Failure Surface) which is the identical method for ICES-LEASE 1, and STABL 5M utilizes the simplified JANBU method (applicable to Amendment 65 Page 223 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 General Shape Failure Surfaces) and the Spencer Method (applicable to any type failure surface). This program has been verified to comply with QA standards per the qualification evaluation (Reference 2.5.6-15).2.5.6.5.4.2 Finite element dynamic analysis An analytical study was made of representative sections of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike for conditions representative of the safe shutdown earthquake (SSE) selected for the site. Each section was modeled by an appropriate finite element system (see Figures 2.5.6-16).A full discussion of the dynamic finite element analyses, including the methodology, material properties, and the results are included in Appendix 2.5D. A brief synopsis of the dynamic finite element analysis that was performed prior to the start of construction is provided in the following paragraphs:As a starting point for the dynamic analysis, the static stresses within the dams were calculated by using currently available static finite element analysis procedures. Non-linear material properties were used in the calculations.The dynamic response of the dam-foundation system was computed by the finite element method in which the continuum is idealized by elements in plane strain and all components of the stress tensor are incorporated. The strength of the dam and foundation materials was determined on the basis of triaxial tests by imposing on specimens cyclic stresses which simulate those induced by an earthquake. The tests provided cyclic strength parameters representative of field conditions. Therefore, any possibility of the development of tensile stresses was accounted for by the method of analysis and the method of determination of cyclic strength parameters, i.e., the tensile stresses were not neglected.The dynamic stresses induced within the dam based on the SSE were calculated by using the dynamic finite element computer program Quad IV. The modulus and damping values for each element were selected on the basis of the strain that would be induced in the element during the applied earthquake motion. The time history of earthquake motion that was applied to the base of the dam was appropriate to provide a satisfactory estimate of the expected frequency characteristics of the SSE.The static finite element analyses that were performed provided value of initial normal effective stresses at any location throughout the dam. The values were then used in conjunction with data similar to those shown on Figure 2.5.6-14 in order to determine the cyclic shear stresses that are required to cause five percent strain at any location within the dam. The dynamic finite element analyses provided values of shear stresses that could be induced during earthquake motions at any location throughout the dam. The effect of vertical ground motions was also evaluated. The safety of the dam was then evaluated by comparing the shear stresses that could be induced during earthquake motions with the shear stresses that are required to cause five percent strain. The latter stresses exceeded the induced stresses by an appropriate margin; therefore the dam will have an adequate safety factor against failure during the SSE.A typical evaluation of the failure potential along a selected plane is illustrated on Figure 2.5.6-

17. As shown in the upper part of the figure, the plane represents average conditions within a finite zone of the dam. The middle part of the figure shows the distribution of the shear stresses Amendment 65 Page 224 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 induced during the earthquake, (ti)N, and the shear stresses required to cause five percent strain (tf)N. The ratio (tf)N (ti)N is presented in the lower part of Figure 2.5.6-17. This ratio, which may be considered to represent a factor of safety, was required to have a minimum value of 1.1. A value of 1.1 is judged appropriate because an adequate degree of conservatism was used in interpreting the test data and in selecting the material properties for the analyses. In addition, appropriate variations of the material properties (eg. modulus and damping values, as illustrated on Figure 2.5.6-13) were incorporated in the analyses.Some changes in design and material properties were found to be necessary during construction. The effects of these changes on the results of the dynamic finite element analyses are discussed in Appendix 2.5F.2.5.6.5.5 Material properties 2.5.6.5.5.1 Main dam The material properties that were used in the slip circle analyses for the main dam impervious core, transition filters, and rockfill shell are:MAIN DAM MATERIALS PROPERTIES MATERIAL WEIGHT (pcf) STRENGTH MOIST SAT. TAN C Impervious Core 137 142 30 0.5773 200 Fine Filter 130 135 35 0.7002 0 Coarse Filter 135 140 40 0.8391 0 Rockfill 130 145 40 0.8391 0 The values listed above were determined as described in Section 2.5D.10, 2.5D.12, and 2.5D.13 of Appendix 2.5D, and Appendix 2.5F.2.5.6.5.5.2 Auxiliary dam and auxiliary reservoir separating dike.The material properties used in the slip circle analyses for the auxiliary dam impervious core, transition filter, and random rockfill shell and for the auxiliary reservoir separating dike impervious core and random rockfill shell are shown in the table below. The shear strengths of the random rockfill were conservatively taken to be the same as those of the impervious core.AUXILIARY DAM AND DIKE MATERIAL PROPERTIES MATERIAL WEIGHT (PCF) STRENGTH MOIST SAT. TAN C C' Impervious Core 128 135 30 0.5773 100 300 Filter 135 140 37 0.7535 0 0 Random Rockfill 130 138 30 0.5773 0 300 Residual Soil 128 134 30 0.5773 150 150 The static (C) and dynamic (C') strength parameters were derived from tests described in Sections 2.5D.11, 2.5D.12, and 2.5D.13 of Appendix 2.5D, and Section 2.5.6.4.2.1.Amendment 65 Page 225 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The cohesion component of shear strength, C', was established by dynamic testing and is utilized with the friction angle, , to develop the dynamic strength of the materials during the pseudo-static analyses.For the sliding wedge analyses of the Auxiliary Dam, the shear strength of the horizontal seams of low strength material within the rock in the abutment areas is shown on Figure 2.5.6-18.Amendment 65 Page 226 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6.5.5.3 Channels The material properties used in the slip circle analysis for the Auxiliary Reservoir Channel and Emergency Service Water Intake and Discharge Channels are:CHANNELS MATERIAL PROPERTIES END OF CONSTRUCTION STATIC LONG PSEUDO-STATIC RAPID MATERIAL ANALYSIS TERM ANALYSIS ANALYSIS DRAWDOWN ANALYSIS pcf psf deg pcf psf deg pcf psf deg pcf psf deg Modified Random fill (Aux, ESW 135 1470 11.5 140 120 25 140 660 17 140 660 17 Channels)Random fill (Aux, 135 1470 11.5 140 120 25 140 660 17 140 660 17 ESW Channels)Residual+ Soil 137 420 34 138 410 29 138 385 27 138 385 27 (ESW Channels)Residual+ Soil 149 420 34 149 410 29 149 385 27 149 385 27 (Aux Channel)(Cooling* Tower and ESW Water Channel)Residual Soil 138 385 20 138 0 36 138 385 20 138 385 26 Rock Bedding Not Applicable Not Applicable 165 250 19 Not Applicable Plane Steep Rock Slope Not Applicable Not Applicable 165 5000 0 Not Applicable

• Note: These values are for the most part conservatively extrapolated from the values of the other two channels, etc.

+The values above are based on the results of laboratory tests presented in Appendix 2.5K.2.5.6.5.6 Results of slope stability analyses.2.5.6.5.6.1 Static and pseudo-static evaluation of the main dam, auxiliary dam, and auxiliary separating dike.The computed safety factors are shown on Figures 2.5.6-19, 2.5.6-20, 2.5.6-21, and 2.5.6-22.The tabulations show the range of values that were obtained for the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike under static and pseudo-static loading conditions. In addition, the static factors of safety for the Auxiliary Dam, based on the sliding wedge method of analysis, are shown on Figure 2.5.6-18. The static safety factors were calculated as described above with no Amendment 65 Page 227 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 consideration of seismic forces. The pseudo-static safety factors were calculated by utilizing the SSE and OBE base accelerations applied directly to the individual slip circles. The pseudo-static analysis for the OBE utilizes the static material properties in order to be consistent with conventional approaches which use earthquakes zone seismic coefficients and static properties.The pseudo-static analysis for the SSE utilizes the dynamic material properties (described in Section 2.5.6.5.5) that are realistic and compatible with actual seismic considerations.The minimum factors of safety with and without earthquake loading are shown in Tables 2.5.6-1, 2.5.6-2, and 2.5.6-3.The results of the static and pseudo-static stability analyses demonstrate that the slopes of the Seismic Category I reservoir structures have an adequate factor of safety under all postulated design conditions.The analysis was performed for the main reservoir water level at Elevation 250 ft. Because dynamic effects are maximized on submerged slopes, analysis of embankment slopes for lower water levels yields higher safety factors. The higher safety factors result from the fact that imposed dynamic forces remain relatively constant, irrespective of water levels, while lower water levels maximize effective stresses and thus the shear resistance. Therefore, with the Main Reservoir at Elevation 220 ft, the safety factors will be greater than those listed on Figures 2.5.6-19, 2.5.6-20, and 2.5.6-21. The results of the analyses presented for the Auxiliary Reservoir Separating Dike on Figure 2.5.6-22 are not changed.2.5.6.5.6.2 Dynamic finite element analysis of dams Results of the dynamic finite element analyses are presented and discussed in Appendix 2.5D.The results indicate that adequate safety margins exist in the Main Dam, the Auxiliary Dam and the Auxiliary Reservoir Separating Dike during the SSE and OBE.The analyses presented in Appendix 2.5D for the Main Dam and the Auxiliary Dam are for a water level at Elevation 250 ft. in the Main Reservoir. The operating water level in the Main Reservoir is Elevation 220 ft. Results of analyses presented in Appendix 2.5D indicate that the Main and Auxiliary Dams are stable and would maintain their integrity if the SSE occurs when the water level in the Main Reservoir is at Elevation 220 ft.2.5.6.5.6.3 Stability of channel slopes 2.5.6.5.6.3.1 Emergency Service Water and Cooling Tower Makeup Channel, Emergency Service Water Intake and Discharge Channel, and Auxiliary Reservoir Channel.Typical cross sections of the channels are shown on Figures 2.5.6-6, 2.5.6-7, 2.5.6-8, and 2.5.6-

28. Four loading conditions were investigated; long term static, pseudo-static (earthquake), end of construction, and rapid drawdown. The pseudo-static analysis incorporated horizontal and vertical seismic coefficients of .15g and .10g, respectively. Water levels were varied to provide the most conservative factors of safety under seismic loading conditions.

The rapid drawdown analysis modeled an extreme condition whereby channel water levels dropped from extreme high water level to the channel floor. Except for the Emergency Service Water and Cooling Tower Makeup Channel, whereby channel water levels dropped from Amendment 65 Page 228 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 extreme high level to water elevation of 204.4 feet. Utilizing a water drop from extreme high water to channel floor was considered to be conservative and unrealistic (Reference 2.5.6-15).The strength parameters that were utilized in the long term static analysis of the emergency service water intake and discharge channel fill sections were obtained from triaxial tests on samples compacted at moisture contents four percent above optimum. Placement requirements limited field moistures to within plus or minus two percent of optimum. Hence, the strength parameters and the resulting factors of safety are very conservative.The Auxiliary Reservoir Channel was cut entirely through hard residual soil or rock. Therefore, there were no fill sections. The Emergency Intake and Discharge Channels consisted of cut and fill sections. Both cut sections and fill sections were analyzed. Three representative cross sections conservatively modeled all channel slopes.All channel slopes are stable under static and seismic loading. Figures 2.5.6-23, 2.5.6-24, and 2.5.6-25 present the results of the stability analyses. Stability analyses results for the Emergency Service Water and Cooling Tower Makeup Channel Slopes are contained in the qualification evaluation (Reference 2.5.6-15).2.5.6.6 Seepage Control 2.5.6.6.1 General Seepage control for the Main and Auxiliary Dams is provided by impervious cores and by grout curtains in the foundation rock. Graded filters are placed as protection on each side of the impervious cores.2.5.6.6.2 Control of seepage through the dams The criteria for the materials used in the core of the dams are that 40 percent of the material passes the No. 200 sieve and the plasticity of the placed core material is greater than 10. The very low permeability (10-8 cm/sec) of the compacted core material (see Appendix 2.5D) effectively reduces the seepage through the dams to a negligible amount.As indicated on Figure 2.5.6-2, the transition filter zones for the Main Dam, which are comprised of fine and coarse filter zones, are founded upon weathered rock on the upstream side of the core of the dam. On the downstream side of the core, only the coarse filter is founded on weathered rock. An additional single transition filter zone, graded from the maximum size of crushed rock in the coarse filter to the minimum size of crushed rock in the fine filter, is extended to suitable rock under the fine filter zone. Wherever the single transition filter comes into contact with the downstream main dam cutoff trench face, the face is treated with concrete, as described in project specifications (Appendix 2.5I). The fill concrete treatment acts as an additional seal on the downstream cutoff trench face, as well as aiding in compaction of the filter against the rock face.The transition filter zone and concrete on the downstream side of the main dam cutoff trench was added to preclude the possibility of piping of core materials into the weathered rock, since jointing in the hard granitic rock at the Main Dam may be developed. The piping potential in the foundation is even more remote when the grouting program described in Section 2.5.6.3 is considered.Amendment 65 Page 229 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The transition filter zone for the Auxiliary Dam, both on the upstream and downstream sides of the core, is founded on weathered rock, as indicated by the geologic mapping shown in Appendix 2.5F and by Figure 2.5.6-4. Thus the transition filter zones are founded on weathered rock on the basis of evaluation and analysis of the rock in combination with the extensive grouting program. In addition, the width of the impervious core in each of the dams effectively precludes the potential for piping in the core and foundation contact by reducing the gradient.There are no rapid drawdown design conditions for the Main Dam or Auxiliary Separating Dike.The Auxiliary Dam may have a rapid drawdown condition on the downstream side induced by a drop in the Main Reservoir PMF water level at Elevation 238.9 ft. to Elevation 209 ft. if a hypothetical mechanistic failure of the Seismic Category I Main Dam should occur. As indicated in Appendix 2.5D, the stability analysis of the Auxiliary Dam was performed for a rapid drawdown from Elevation 250 ft. to Elevation 209 ft. The gradations of specified materials provide adequate drainage in the shell sections of the embankments. The shell materials were placed during construction such that the finer rock and random rock materials were placed near the filters and graded to the larger materials toward the outside; this assures the shortest possible drainage paths from the finer shell materials to the filter transition materials or to the more coarse rockfill materials. The rockfill material used in the Main Dam consists of granitic rocks which did not break down excessively during handling and compaction, and therefore, provides a positive drainage in the shell sections for internal embankment seepage.2.5.6.6.3 Control of seepage through foundations and abutments Seepage through the foundations and abutments at the Main and Auxiliary Dams is controlled by extending the impervious silty clay and clayey silt core in the cutoff trench excavated through the weathered rock to the top of sound rock, and by installation of grout curtains. At the Auxiliary Separating Dike, no cutoff trench or grouting is required since water level is the same on both sides of the dike.Details of excavation of the cutoff trenches and backfilling with compacted impervious core material are discussed in Appendix 2.5F.Details of installation, as well as the depth and lateral extent of the grout curtains, are presented in Section 2.5.6.3. Details, including water-test results and grout takes, are included in the final foundation report (Appendix 2.5E).Piezometers in the Main and Auxiliary Dam and seepage measurement devices in the Main Dam were installed at locations shown on Figures 2.5.6-1 and 2.5.6-3 to monitor seepage through the dam foundations. Seepage is expected to be minimal on the basis of the very low water-test values and the extremely low grout takes, which averaged only 0.02 to 0.06 gabs per foot of curtain hole drilled. Additional information on piezometer installation and pore pressure monitoring is provided in Section 2.5.6.8.2.5.6.7 Diversion and Closure 2.5.6.7.1 Main dam diversion and closure A Diversion structure was provided in the Main Dam to pass the 10 year return period flood.The details of the main dam diversion system are indicated on Figure 2.5.6-26.Amendment 65 Page 230 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The diversion system consisted of corrugated metal pipe with cast-in-place reinforced concrete encasem*nt under the entire main dam section and reinforced concrete intake and discharge structures. Details of the intake and discharge structures are shown on Figure 2.5.6-27.The reinforced concrete diversion structure for the Main Dam was formed by using twin 10-ft.diameter galvanized corrugated metal pipes as an inner form for the reinforced concrete encasem*nt structure.The material for the portion of the metal corrugated pipes encased in concrete under the dam for the diversion structure is in accordance with Specification AASHO M167, "Structural Plate for Pipe, Pipe Arches, and Arches." The corrugated metal pipe extending upstream and downstream of the Main Dam complies with Specification AASHO M36 "Zinc-Coated (Galvanized) Corrugated Iron or Steel Culverts and Underdrains."The galvanized corrugated metal pipe remained in place after construction. Since the central portion of the diversion structure under the core of the dam was completely filled with concrete to plug the pipe, there will be no flow of water through the conduit. Due to the interlocking corrugation of the plug and the concrete encasem*nt structure, the possibility of the plug movement is minimized. The other sections of pipe are also galvanized and corrosion will be minimal. However, corrosion will not affect the integrity of the dam nor the diversion structure.The reinforced concrete diversion structure for the Main Dam was founded upon suitable rock, was designed to Seismic Category I standards, and is capable of withstanding the effects of the most severe phenomena associated with the site. The diversion structure was constructed with construction joints as indicated on Figure 2.5.6-26. At the vertical construction joints in the reinforced concrete diversion structure, under the dam core, a 3/8 in. thick PVC waterstop was provided. At each horizontal construction joint in the reinforced concrete diversion structure under the dam core, a minimum of 3/16 in. thick mild steel waterstop was installed.The conduits were plugged by a single concrete plug in each conduit, approximately 100 ft.long, in the central area of the dam; the plug extends the full width of the impervious core as indicated on Figure 2.5.6-26. Each plug was designed to withstand the total hydraulic head across the dam.The concrete plug was sealed by a pressure grouting system that was installed along the length of the plug to assure that the pipe was completely filled.The corrugated pipe acts as a key to transfer the plug loading to the reinforced concrete diversion structure.The reinforced concrete diversion structure has two vertical to one horizontal sloping sides to minimize post construction differential settlement. The material used for backfill against the diversion structure is the same as that used in the core and filter zones. An envelope of pervious fill comprised of a 3 ft. thick layer of compacted, well graded crushed rock was provided around the diversion structure in the rockfill zone downstream of the inclined embankment filters to control seepage and to avoid concentrated loading on the structure. An envelope of a 3 ft. thick layer of compacted well graded crushed rock around the diversion structure was provided in the rockfill zone upstream of the inclined embankment filters to avoid concentrated loading on the structure. The compaction criteria for the fill around the diversion structure in the filter zones and for the crushed rock in the rockfill zone was the same as that Amendment 65 Page 231 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 used for the filter zone in the embankment. The backfill material that was placed against the diversion structure was placed by using hand compaction methods when necessary. Heavy hauling equipment was not permitted to cross the conduit until the backfill reached a height of 3 ft. above the top of the structure. Above this point, machine compaction methods were used.During preliminary preparation of the main dam site, a diversion channel was constructed immediately northeast of the existing stream bed, and the stream flow was diverted through it.Cofferdams were then constructed adjacent to the diverted creek and tied into the west abutment to protect the lower elevation work areas from periodic flooding. Next, the area underlying the conduit was mapped and the foundation was prepared. Construction of the concrete-encased conduit followed. After the conduits outside the limits of the dam were placed, upstream and downstream cofferdams were constructed across the valley. The original cofferdam was then removed and the water began flowing through the conduits. Mapping and foundation preparation along the remainder of the core trench was completed, followed by the construction of the dam. Upon completion of the dam, concrete plugs were placed in the conduit and the reservoir began to fill.2.5.6.7.2 Auxiliary Dam Diversion No diversion conduit was installed for the Auxiliary Dam. Diversion of the stream during the construction of the dam was performed by construction methods which ensured that the dam was constructed safely, efficiently, and met the design specification.An upstream cofferdam was constructed at Elevation 225 ft. Water level was maintained at an acceptable level by pumping from behind the upstream cofferdam to Tom Jack Creek downstream of the core trench. When all mapping, grouting and cleaning were complete in the area between approximately Station 28+0 and 31+0, impervious backfill was placed in this portion of the core trench from the upstream cofferdam to a point downstream of the dam. The diversion channel was constructed in this backfill to approximately Elevation 220 ft. Portions of the dam on both sides of this diversion were completed to approximately Elevation 235 ft. At this time, water level behind the upstream cofferdam was controlled by pumping while the required portions of the diversion embankment are removed (impervious material in areas required to be filter material or random rockfill). The dam embankment in the area previously occupied by the diversion was then constructed to Elevation 235 ft. Construction of the dam continued with water level maintained by pumping and dam embankment.2.5.6.8 Performance Monitoring Instruments for monitoring the performance of the dams were installed in accordance with project specifications.The program for periodic monitoring of instrumentation and periodic inspection of the embankments is also contained in the project specifications.The Main Dam has settlement monuments, piezometers, and seepage monitors as shown on Figures 2.5.6-1 and 2.5.6-2. The Auxiliary Dam has settlement monitors and piezometers, as shown on Figure 2.5.6-3. The Auxiliary Separating Dike has settlement monuments, as shown on Figure 2.5.6-5.Amendment 65 Page 232 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The retaining wall including the deadmen west of the Fuel Handling Building have settlement and/or lateral movement markers as shown on Figure 3.8.4-43. The representative specimens of the retaining wall tie rods are buried in the soil backfill as shown on Figure 3.8.1-44. The frequency of monitoring settlement and lateral movement of the retaining wall is shown on Figure 3.8.4-42. The retrieval and inspection frequency of the tie rods is also shown on Figure 3.8.4-42.2.5.6.9 Construction Notes Several changes in construction details became necessary during construction. These changes include:a) Borrow Area M originally proposed as a borrow source for the impervious core of the Main Dam, proved to be unsuitable since it did not meet gradation and plasticity requirements. An alternative source near the Main Dam, designated Borrow Area W, was used for the impervious core material.b) Field changes were made in the construction of the diversion and closure system for the Auxiliary Dam (see Section 2.5.6.7).c) Due to the deviations in moisture content requirements, portions of the original impervious core backfill of the Auxiliary Dam and the Auxiliary Separating Dike were removed and replaced with new compacted backfill.d) The impervious liner of the emergency service water intake channel fill sections, originally placed in accordance with performance specification, was removed and replaced with new compacted backfill in accordance with specified moisture and density control requirements.2.5.6.10 Operational Notes This section will be provided after the reservoirs have become operational.

REFERENCES:

SECTION 2.5 2.5.1-1 Conley, J. F., and Bain, G. L., 1965, Geology of the Carolina Slate Belt West of the Deep River-Wadesboro Triassic Basin, North Carolina: Southeastern Geology, v. 6,

p. 117-138.

2.5.1-2 Stuckey, J. L., and Conrad, S. G., 1958, Explanatory text for geologic map of North Carolina, Department of Conservation and Development Division of Mineral Resources, Bulletin 71.2.5.1-3 Hills, F. A., and Butler, J. R., 1969, Rubidium-strontium dates for some rhyolites from the Carolina slate belt of the North Carolina Piedmont (abs): Geological Society of America Special Paper 121, p. 445.2.5.1-4 Fullagar, P. D., 1971, Age and origin of plutonic intrusions in the Piedmont of the southeastern Appalachians: Geological Society of America Bulletin, v. 83, p. 2845-2862.Amendment 65 Page 233 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.1-5 Glover, Lynn, III, and Sinha, A. K., 1973, The Virginia deformation, a late Precambrian to Early Cambrian orogenic event in the central Piedmont of Virginia and North Carolina: American Journal of Science, v. 273-A (Cooper Volume), p.234-251.2.5.1-6 Black, W. W., and Fullagar, P. D., 1976, Avalonian ages of metavolcanics and plutons of the Carolina slate belt near Chapel Hill, North Carolina: Geological Society of America Abstracts with Programs, v. 8, p. 136.2.5.1-7 Wright, J. E., and Seiders, V. M., 1977, U-Pb dating of zircons from the Carolina volcanic slate belt, central North Carolina: Geological Society of America Abstracts with Programs, v. 9, p. 197-198.2.5.1-8 McConnell, K. I., Glover, Lynn, III, and Sinha, A. K., 1976, Geology of the Late Precambrian intrusive complex and associated volcanic rocks along the Flat River near Durham, North Carolina: Geological Society of America Abstracts with Programs, v. 8, p. 226-227.2.5.1-9 Fullagar, P. D., and Butler, J. R., 1979, 325 to 265 m.y. - old granitic plutons in the Piedmont of the southeastern Appalachians: American Journal of Science, v. 279, p.161-185.2.5.1-10 Glover, Lynn, III, Mose, D. G., Poland, F. B., and Bobyarchick, A. R., 1978, Grenville basem*nt in the eastern Piedmont of Virginia: implications for orogenic models:Geological Society of American Abstracts with Programs, v. 10, p. 169.2.5.1-11 Briggs, D. F., Gilbert, M. C., and Glover, Lynn, III, 1978, Petrology and regional significance of the Roxboro Metagranite, North Carolina: Geological Society of America, v. 89, p. 511-521.2.5.1-12 Butler, J. R., and Ragland, P. C., 1969, A petrochemical survey of plutonic intrusions in the Piedmont, southeastern Appalachians, U.S.A.: Contributions to Mineralogy and Petrology, V. 24, p. 164-190.2.5.1-13 Parker, J. M., III, 1977, Structure of Raleigh Belt of eastern Piedmont in Wake County, North Carolina: Geological Society of America, Abstracts with Programs, v.9, p. 173.2.5.1-14 Becker, S. W., and Farrar, S. S., 1977, The Rolesville batholith, in Costrain, J. K.,and others, Evaluation and targeting of geothermal energy resources in the southeastern United States: Virginia Polytechnic Institute and State University, Blacksburg (for U. S. Energy Research and Development Administration), p. A p. A-77.2.5.1-15 Parker, J. M., III, 1968, Structure of easternmost North Carolina Piedmont:Southeastern Geology, v. 9, p. 117-131.2.5.1-16 Campbell, M. R., and Kimball, D. D., 1923, The Deep River Coal Field of North Carolina: North Carolina Department of Conservation and Development, Bulletin 33,

p. 95.

Amendment 65 Page 234 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.1-17 Reinemund, J. A., 1955, Geology of the Deep River Coal Field, North Carolina: U. S.Geological Survey Professional Paper 246, p. 159.2.5.1-18 Weigand, P. W., and Ragland, P. C., 1970, Geochemistry of Mesozoic dolerite dikes from eastern North America: Contributions to Mineralogy and Petrology, v. 29, p.195-214.2.5.1-19 Burt, E. R., Carpenter, P. A., III, McDaniel, R. D. and Wilson, W. F., 1978, Diabase dikes of the eastern Piedmont of North Carolina: North Carolina Division of Land Resources, Geological Survey Section Information Circular 23, p. 12.2.5.1-20 Conley, J. F., 1962, Geology and mineral resources of Moore County, North Carolina: North Carolina Department of Conservation and Development, Bulletin 40,

p. 40.

2.5.1-21 Swift, D.J.P., and Heron, S. D., Jr., 1969, Stratigraphy of the Carolina Cretaceous:Southeastern Geology, v. 10, p. 201-245.2.5.1-22 Stuckey, J. L., 1965, North Carolina: Its geology and mineral resources: North Carolina Department of Conservation and Development, Raleigh, p. 550.2.5.1-23 Sundelius, H. W., 1970, The Carolina slate belt in Fisher, G. W., and others (eds.),Studies of Appalachian Geology: Central and Southern: New York, Interscience Publishers, p. 351-369.2.5.1-24 Tobisch, O. T., and Glover, Lynn, III, 1971, Nappe formation in part of the southern Appalachian Piedmont: Geological Society of America, v. 82, p. 2209-2230.2.5.1-25 Butler, J. R., and Fullagar, P. D., 1977, The Gold Hill fault zone in the Carolinas: age of movement and southwestern extension: Geological Society of America, Abstracts with Programs, v. 9, p. 125.2.5.1-26 Parker, J. M., III, 1979, Geology and mineral resources of Wake County: North Carolina Department of Natural Resources and Community Development, Division of Land Resources, Geological Survey Section, Bulletin 86, p. 122.2.5.1-27 Bain, G. L., and Harvey, B. W., 1977, Field guide to the geology of the Durham Triassic Basin: Carolina Geological Society, 40th Anniversary Meeting, October 7-9, 1977, p. 83.2.5.1-28 Harrington, J. W., 1951, Structural analysis of the west border of the Durham Triassic Basin: Geological Society of America Bulletin, v. 62, p. 149-158.2.5.1-29 Ebasco Services, Inc., 1975, Fault Investigation, Shearon Harris Nuclear Power Plant, Units 1, 2, 3, 4, v.1, v.2 (Appendices). Docketed.2.5.1-30 Mann, V. I., and Zablocki, F. S., 1961, Gravity features of the Deep River-Wadesboro Triassic Basin of North Carolina: Southeastern Geology, v. 2, p. 191-215.Amendment 65 Page 235 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.1-31 Tanner, J. G., 1967, An automated method of gravity interpretation: Geophysical Journal of the Royal Astonomical Society, v. 13, p. 339-347.2.5.1-32 Watts, A. B., 1972, Geophysical investigations east of the Magdalen Islands, southern Gulf of St. Lawrence: Canadian Journal of Earth Sciences, v. 9, p. 1502-1528.2.5.1-33 Prouty, W. F., 1931, Triassic deposits of the Durham Basin and their relation to other Triassic areas of eastern United States: American Journal Science, 5th series, v. 21,

p. 437-490.

2.5.1-34 Sbar, M., and Sykes, L. R., 1973, Contemporary Compressive Stress and Seismicity in Eastern North Carolina: An Example of Intraplate Tectonics: Geological Society of America Bulletin, v. 84, p. 1861-1922.2.5.1-35 Harrington, J. W., 1948, The west border of the Durham Triassic basin (Ph.D.dissertation, University of North Carolina, Chapel Hill).2.5.1-36 Kiersch, G. A., 1972, Triassic Basin geological considerations: Chapter 10 in Savannah River Bedrock Storage Project, Interim Report: Parsons, Brinkerhoff, Quade, and Douglas report to E. I. DuPont/Atomic Energy Commission, p. 36.2.5.1-37 Bain, G. L., 1972, Interim report data availability for feasibility study of east coast Triassic basins for waste storage: U. S. Geological Survey Open File Report, Morgantown, West Virginia, p. 150.2.5.1-38 Bain, G. L., and Thomas, J. D., 1966, Geology and Groundwater in the Durham Area, North Carolina, North Carolina Department of Water Resources, Groundwater Bulletin No. 6, p. 147.2.5.1-39 May, V. J., and Thomas, J. D., 1968, Geology and groundwater resources in Raleigh area North Carolina; North Carolina Department of Water and Resources, Groundwater Bulletin 15, p. 135.2.5.1-40 Heron, S. D., Jr., Judd, J. B., and Johnson, H. S., Jr., 1971, Clastic dikes associated with soil horizons in the North and South Carolina Coastal Plain: Geological Society of America Bulletin, v. 82, p. 1801-1810.2.5.1-41 Ferenczi, I., 1960, Structural control of the North Carolina Coastal Plain:Southeastern Geology, v. 1, p. 105-115.2.5.1-42 Brown, P. M., Miller, J. A., and Swain, F. A., 1972, Structural and stratigraphic framework and spatial distribution of permeability of the Atlantic Coastal Plain, North Carolina to New York: U. S. Geological Survey Professional Paper 796, p. 79.2.5.2-1 Bollinger, G. A., 1975, A catalog of southeastern United States earthquakes 1754 through 1974, Research Division Bulletin 101, Department of Geological Sciences, Virginia Polytechnic Institute and State University, Blacksburg, Virginia 24061.Amendment 65 Page 236 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2-2 Taber, S., 1913, The South Carolina earthquake of January 1, 1913, Seismological Society of America Bulletin, v. 3, p. 6-13.2.5.2-3 Carder, D. S., 1968, Reservoir loading and local earthquakes, Ann. Engr. Geol. and Soils Engr. Symp., Boise, Idaho.2.5.2-4 Galanopolos, G. A., 1967, The influence of the fluctuation of Marathon Lake elevation on local earthquakes in the Attica Basin area, Ann. Geol. Pasy Helliniques,

v. 18, p. 281-306.

2.5.2-5 Rothe, J. P., 1969, Earthquakes and reservoir loadings, 4th World Conference on Earthquake Engineering, Santiago, Chile.2.5.2-6 Comminakis, P., Drakopoulos, J., Moumoulidis, G., and Papazachos, B., 1968, Foreshock sequences of the Kremasta earthquake and their relation to the water loading of the Kremasta artificial lake, Ann. Geofis, (Rome), v. 21, p. 39-71.2.5.2-7 Housner, G. W., 1969, The seismic events at Koyna Dam, presented at the 11th annual symposium on rock mechanics, Univ. of California, Berkeley, California.2.5.2-8 Gupta, H., Narain, H., Rastogi, B. K., and Mohan, I., 1969, A study of the Koyna earthquake of December 10, 1967, Seismological Society of America Bulletin, v. 59,

p. 1149-1162.

2.5.2-9 Gupta, H. K., Rasogi, B. K., and Narain, H., 1972, Common features of the reservoir associated seismic activities, Seismological Society of America Bulletin, v. 62, p.481-492.2.5.2-10 Gough, D. I. and Gough, W. I., 1970, Load induced earthquakes at Lake Kariba, Paper II, Geophysical Journal of the Royal Astronomical Society, v. 21, p. 79-101.2.5.2-11 Shen, C. K., Chen, H. C., Huang, L. S., Yong, C. J., Chang, C. H., Li, T. C., Wang, T.C. and Lo, H. H., 1974. Earthquakes Induced by Reservoir Impounding and their Effect on the Hsin Feng Kiang Dam, Peking, China. Scientia Scinica, v. 17, p. 239-272.2.5.2-12 Talwani, P., 1979, Induced Seismicity and Earthquake Prediction Studies in South Carolina: Eighth Technical Report to U. S. Geological Survey, Contract No. USGS 14-08-001-14553.2.5.2-13 Packer, D. R., Cluff, L. S., Knuepfer, P. L., and Withers, R. J., 1979, Study of Reservoir Induced Seismicity: Final Technical Report to U. S. Geological Survey, Contract No. USGS 14-08-001-16809.2.5.2-14 Snow, D. T., 1973, The geologic, hydrologic, and geomorphic setting of earthquakes at Lake Kariba Final Technical Report to U. S. Geological Survey, Contract No.USGS 14-08-0001-13079.Amendment 65 Page 237 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2-15 Kisslinger, C., 1975, Review of theories of mechanisms of induced seismicity, Firth International Symposium on Induced Seismicity, September 15-19, 1975, Banff, Canada, Rep. 24 p.2.5.2-16 Woollard, G. P., 1969, Tectonic activity in North America in the earth's crust and upper mantle, American Geophysical Union, Monograph 13, p. 125-137.2.5.2-17 McClain, W. C. and Meyers, O. H., 1970, Seismic history and seismicity of the southeastern region of the United States ORNL 4582; UC-51, Oak Ridge National Lab., Oak Ridge, Tenn. 46 p.2.5.2-18 Nuttli, O. W., 1974, Seismic hazard east of the Rocky Mountains, presented at the ASCE National Structural Engineering Meeting.2.5.2-19 Ferguson, J. F. and Stewart, D. M., 1974, A brief survey of the seismicity of North Carolina in the 18th and 19th centuries, Earthquake Notes, p. 33.2.5.2-20 MacCarthy, G. R., 1956, A marked alignment of earthquake epicenters in western North Carolina and its tectonic implications, Jour. of the Elisha Mitchell Sci. Soc., v.72, p. 274-276.2.5.2-21 Oliver, J. and Isacks, B., 1971, Seismicity of the eastern United States, Earthquake Notes, v. 43, p. 30.2.5.2-22 Fox, F. L., 1970, Seismic geology of the eastern United States, Association of Engineering Geologists Bulletin, v. 7, p. 21-43.2.5.2-23 Bollinger, G. A., 1973, Seismicity of the southeastern United States, Seismological Society of America Bulletin, v. 63, p. 1785-1808.2.5.2-24 Bollinger, G. A., 1973, Seismicity and crustal uplift in the southeastern United States, American Journal of Science, v. 273A, p. 396 408.2.5.2-25 Bollinger, G. A., 1977, Reinterpretation of the intensity data for the 1886 Charleston, South Carolina, earthquake, in Studies Related to the Charleston, South Carolina, Earthquake of 1886 - A Preliminary Report, Geological Survey Professional Paper 1028-B, p. 17-32.2.5.2-26 Rankin, D. W., 1977, Studies related to the Charleston, South Carolina, earthquake of 1886 - introduction and discussion, in Studies Related to the Charleston, South Carolina, Earthquake of 1886 - A Preliminary Report, Geological Survey Professional Paper 1028 A, p. 1-15.2.5.2-27 Stauder, W., Kramer, M., Fisher, G., Schaefer, S. and Morrissey, S. T., 1976, Seismic characteristics of southeast Missouri as indicated by a regional telemetered microearthquake array, Seismological Society of America Bulletin, v. 66, p. 1953-1964.2.5.2-28 Gupta, I. N. and Nuttli, O. W., 1976, Spatial attenuation of intensities for central U. S.earthquakes, Seismological Society of America Bulletin, v. 66, p. 743-751.Amendment 65 Page 238 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.2-29 Bollinger, G. A., 1969, Seismicity of the central Appalachian states of Virginia, West Virginia, and Maryland - 1758 through 1968, Seismological Society of America Bulletin, v. 59, p. 2103-2111.2.5.2-30 Bollinger, G. A., and Visvanathan, T. R., 1977, The seismicity of South Carolina prior to 1886, in Studies Related to the Charleston, South Carolina, Earthquake of 1886 -A Preliminary Report, Geological Survey Professional Paper 1028 - C, p. 33-42.2.5.2-31 Coffman, J. L., and Von Hake, C. A., Editors, 1973, Earthquake History of the United States (revised edition through 1970), Publication 41-1, Environmental Data Service, NOAA, U. S. Dept. of Commerce, Boulder, Colorado.2.5.2-32 Gutenberg, B., and Richter, C. F., 1954, Seismicity of the earth and associated phenomena, 2nd edition, Princeton University Press, Princeton.2.5.2-33 Visvanathan, T. R., 1977, Earthquakes in South Carolina, 1698-1974, Department of Geology, University of South Carolina, Union Regional Campus, Union, S. C.2.5.2-34 Nuttli, O. W., 1973, The Mississippi Valley earthquakes of 1811 and 1812; intensities, ground motion and magnitudes, Seismological Society of America Bulletin, v. 63, p. 227-248.2.5.3-1 Plaster, R.W. and Sherwood, W.G., 1971, Bedrock Weathering and Residual Soil Formation in Central Virginia; Geol. Soc. Amer. Bull., V.82, pp. 2313-2326.2.5.3-2 Buol, S.W., Hole, F.D., and McCracken, R.J., 1973, Soil Genesis and Classification:Iowa State University Press, Ames, Iowa, p. 360.2.5.3-3 Carson, R.J. III, 1970, Quaternary Geology of the South-Central Olympic Peninsula:Unpublished Ph.D. Thesis, Univ. of Washington, Seattle, Washington.2.5.3-4 Stuckey, Dr. J.L., Personal communication.2.5.3-5 Butler, W.R., 1976, Letter to Carolina Power & Light Company from W. R. Butler, representing the U. S. Nuclear Regulatory Commission, January 7, 1976.2.5.4-1 Deere, D. J., Hendron, A. J., Jr. Patton, F. D., and Cording, E. J., "Design of Surface and Near-Surface Construction in Rock," Symp Rock Mech, 8th, Minnesota, 1966 (AIME, 1967).2.5.4-2 Lambe, T. W., and Whitman, R. V., Soil Mechanics, J. Wiley & Sons, 1969.2.5.4-3 Strom, J. A. and Fischer, J. A. "The Shockscope" Dames & Moore Bulletin No. 22, April, 1962.2.5.4-4 Stagg, K. G., "In-situ Tests on the Rock Mass," Rock Mechanics in Engineering Practice, edited by Stagg and Zienkiewicz, J. Wiley & Sons, London, 1968.2.5.4-5 Hendron, A. J. Jr., "Mechanical Properties of Rock, Rock Mechanics in Engineering Practice, edited by Stagg and Zienkiewicz, J. Wiley & Sons, London, 1968.Amendment 65 Page 239 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.4-6 Terzaghi, K., and Peck, R. B., Soil Mechanics in Engineering Practice, J. Wiley &Sons, 2nd Ed., 1967.2.5.4-7 Peck, R. B., Hanson, W. E., and Thornburn, T. H. Foundation Engineering, J. Wiley

 & Sons, 2nd Ed., 1974.

2.5.4-8 Richart, F. E., Jr., Hall, J. R., and Woods, R. D., Vibrations of Soils and Foundations, Prentice-Hall, 1970.2.5.6-1 "Phase II Site Study, WhiteOak Creek", Ebasco Services Inc., New York, October, 1970.2.5.6-2 Newmark N.W. "Effect of Earthquakes on Dams and Embankments," Geotechnique, Vol. 15, June 1965.2.5.6-3 Terzaghi, K., and Lacroix, Y., "Mission Dam", Geotechnique, Vol. 14, March, 1964.2.5.6-4 Embankment-Dam Engineering Casagrande Volume, John Wiley & Sons, New York, 1972.2.5.6-5 Department of the Army Corps of Engineers, Study #526, "Shear Strength of Rock Fill", Unpublished Paper dated August 17, 1970.2.5.6-6 Testing Institute for Soil Mechanics and Foundation Construction, Technical University in Darmstadt, W. Germany, "Triaxial Shear Tests", A Report on Keban Dam, Ebasco drawing number 4550-2783, May, 1970.2.5.6-7 Wilson, S.D., and Marano, D. "Performance of Muddy Run Embankment", Journal of Soil Mechanics and Foundation Division, ASCE, Vol. 94, No. MS4, July 1968.2.5.6-8 Design of Small Dams, U.S. Department of Interior, Bureau of Reclamation, 1961.2.5.6-9 Zangar, C.N., and Haefeli, R.J., "Electric Analog Indicates Effect of Horizontal Earthquake Shock on Dams", Civil Engineering, Vol. 278, April, 1952.2.5.6-10 Peck, R.B., Hanson, W.E., and Thornburn, T.H., Foundation Engineering, J. Wiley &Sons, 2nd, Ed., 1974.2.5.6-11 ICES Lease - 1 Computer Program, "Slope Stability Analysis", Soil Mechanics Publication No. 235, Department of Civil Engineering, MIT, April, 1969.2.5.6-12 Bishop A. W., "The Use of the Slip Circle in the Stability Analysis of Slopes",Geotechnique, Vol. 5, 1955.2.5.6-13 Department of the Navy, Design Manual: Soil Mechanics: Foundations and Earth Structures, Washington D.C. 2nd, Ed., 1971.2.5.6-14 Taylor, Karl V., "Slope Protection on Earth and Rockfill Dams," Transactions of the Eleventh International Congress on Large Dams, June 1973.Amendment 65 Page 240 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5.6-15 Ebasco Report CAR 1440.122 Parts I and II, "Cooling Tower and Emergency Service Water Intake Channel - Upgrade to Seismic Category I", 1991.APPENDIX 2.5A Preconstruction Exploration

1. TABULATIONS OF BOREHOLE AND TEST PIT LOCATIONS Note: The locations of all known preconstruction project borings are tabulated in this section.

Boreholes drilled for Seismic Category I structures (including cancelled Units 2, 3, & 4 structures) or to sample borrow for such structures are indicated by an asterisk (*). Borehole logs for these borings are included in Section 2 of this appendix. Logs of other borings are maintained in CP&L files.1a. P Series - Preliminary Subsurface Investigation - Plant Vicinity Coordinates Borehole Depth Number North East (feet)(feet)

 *P-1 686,303 2,011,540 150 *P-2 685,641 2,011,847 150 *P-3 685,943 2,012,509 150 *P-4 684,274 2,011,345 150 *P-5 684,706 2,012,280 151 *P-6 685,013 2,012,942 250 *P-7 685,433 2,013,849 150 *P-8 For locations see Figure 2.5.1-11 47 *P-9 For locations see Figure 2.5.1-11 64 *P-10 For locations see Figure 2.5.1-11 66 *P-11 For locations see Figure 2.5.1-11 65 *P-12 For locations see Figure 2.5.1-11 63 *P-13 For locations see Figure 2.5.1-11 65 *P-14 For locations see Figure 2.5.1-11 63 *P-15 For locations see Figure 2.5.1-11 63 *P-16B For locations see Figure 2.5.1-11 60 *P-17 For locations see Figure 2.5.1-11 152 *P-18 For locations see Figure 2.5.1-11 62 *P-19 For locations see Figure 2.5.1-11 53 *P-20 For locations see Figure 2.5.1-11 60 *P-21 For locations see Figure 2.5.1-11 48 *P-43 684,929 2,012,761 154 *P-46 685,110 2,012,676 150 *P-101 685,000 2,013,389 136 *P-102 685,084 2,013,571 150 *P-103 685,168 2,013,752 103 *P-104 685,252 2,013,933 150 *P-105 685,336 2,014,115 150 *P-106 685,181 2,013,305 150 *P-107 685,265 2,013,486 250 *P-108 685,349 2,013,668 150 *P-109 685,517 2,014,030 150 *P-110 685,362 2,013,221 150 *P-111 685,447 2,013,402 152 *P-112 685,531 2,013,584 106 *P-113W 685,614 2,013,765 150 Amendment 65 Page 241 of 369

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 *P-114 685,699 2,013,946 150 1b. D Series - Preliminary Subsurface Investigation - Plant and Auxiliary Dam Vicinity Coordinates Borehole Number North East Depth (feet)

(feet)

 *D-2 684,438 2,006,386 150 *D-3 684,116 2,006,767 150 *D-4 683,859 2,007,075 150 *D-5 683,859 2,007,575 140 *D-6 683,859 2,008,075 135 *D-7 683,859 2,008,575 150 *D-8 683,859 2,009,075 150 *D-9 683,859 2,009,575 150 *D-10 683,859 2,010,075 150 *D-11 For locations see Figure 2.5.1-11 20 *D-12 For locations see Figure 2.5.1-11 65 *D-13 For locations see Figure 2.5.1-11 65 *D-14 For locations see Figure 2.5.1-11 66 *D-15 For locations see Figure 2.5.1-11 66 *D-16 For locations see Figure 2.5.1-11 60 *D-17 For locations see Figure 2.5.1-11 60 *D-18 For locations see Figure 2.5.1-11 60 *D-19 For locations see Figure 2.5.1-11 24 *D-20 683,559 2,008,075 113 *D-21 683,559 2,008,575 147 *D-22 683,559 2,009,075 136 1c. BP Series - Plant Design Exploration Coordinates Borehole Number North East Depth (feet)

(feet)

 *BP-1 685,695 2,012,362 102 *BP-2 685,724 2,012,424 103 *BP-3 685,598 2,012,402 99 *BP-4 685,640 2,012,476 100 *BP-5 685,510 2,012,450 95 *BP-6 685,544 2,012,515 97 *BP-7 685,465 2,012,581 92 *BP-8 685,498 2,012,647 94 *BP-9 685,332 2,012,535 90 *BP-10 685,355 2,012,600 92 *BP-11 685,248 2,012,577 92 *BP-12 685,282 2,012,651 93 *BP-13 685,163 2,012,628 95 *BP-14 685,188 2,012,679 94 *BP-15 685,079 2,012,684 97 *BP-16 685,091 2,012,720 104 Amendment 65 Page 242 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BP-17 685,977 2,012,976 83 *BP-18 686,006 2,013,038 77 *BP-19 685,893 2,013,018 89 *BP-20 685,921 2,013,081 81 *BP-21 685,793 2,013,055 94 *BP-22 685,832 2,013,129 89 1c. BP Series - Plant Design Exploration (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BP-23 685,710 2,013,099 96 *BP-24 685,747 2,013,179 94 *BP-25 685,614 2,013,142 97 *BP-26 685,648 2,013,210 94 *BP-27 685,526 2,013,189 93 *BP-28 685,566 2,013,236 91 *BP-29 685,440 2,013,224 90 *BP-30 685,484 2,013,274 85 *BP-31 685,334 2,013,249 87 *BP-32 685,368 2,013,358 83 *BP-33 685,196 2,012,954 125 *BP-34 685,238 2,013,052 124 *BP-35 685,676 2,012,538 123 *BP-36 685,575 2,012,591 123 *BP-37 685,487 2,012,625 120 *BP-38 685,395 2,012,662 120 *BP-39 685,311 2,012,713 119 *BP-40 685,283 2,012,721 119 *BP-41 685,862 2,012,945 118 *BP-42 685,767 2,012,990 124 *BP-43 685,674 2,013,031 124 *BP-44 685,587 2,013,075 124 *BP-45 685,479 2,013,121 121 *BP-46 685,419 2,013,164 121 *BP-47 685,518 2,012,694 118 *BP-48 685,544 2,012,759 116 *BP-49 685,428 2,012,735 120 *BP-50 685,459 2,012,798 120 *BP-51 685,612 2,012,896 122 *BP-52 685,645 2,012,963 125 *BP-53 685,521 2,012,938 130 *BP-54 685,544 2,013,012 124 *BP-55 685,720 2,012,763 109 *BP-56 685,621 2,012,811 117 *BP-57 685,537 2,012,850 120 *BP-58 685,420 2,012,851 123 *BP-59 685,285 2,012,904 123 *BP-60 685,333 2,012,993 125 *BP-61 685,674 2,012,675 134 *BP-62 685,616 2,012,650 139 *BP-63 685,665 2,012,734 130 *BP-64 685,339 2,012,780 141 *BP-65 685,387 2,012,873 143 *BP-66 685,759 2,012,851 142 *BP-67 685,694 2,012,825 141 Amendment 65 Page 243 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BP-68 685,739 2,012,922 149 *BP-69 685,425 2,012,959 151 *BP-70 685,464 2,013,050 153 *BP-71 685,566 2,012,677 264 *BP-72 685,360 2,012,779 270 *BP-73 685,671 2,012,898 273 *BP-74 685,497 2,012,980 276 *BP-75 685,312 2,012,842 148 *BP-76 685,400 2,012,035 204 *BP-77 685,820 2,012,741 105 *BP-81 687,407 2,012,497 70 1c. BP Series - Plant Design Exploration (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BP-82 686,645 2,011,797 75 *BP-83 686,963 2,012,494 90 *BP-84 687,039 2,013,366 75 *BP-85 686,269 2,011,929 85 *BP-86 686,821 2,013,508 75 *BP-87 686,022 2,012,314 75 *BP-88 686,418 2,013,176 57 *BP-89 686,752 2,013,907 75 *BP-90 686,477 2,013,684 77 *BP-91 685,680 2,014,027 50 *BP-92 685,396 2,014,146 45 *BP-93 684,372 2,013,073 117 *BP-94 684,607 2,013,587 95 *BP-95 684,866 2,013,676 45 *BP-96 685,027 2,014,006 45 *BP-101 684,751 2,012,857 19 *BP-102 684,826 2,013,016 23 *BP-103 684,920 2,013,210 29 *BP-104 685,099 2,013,117 24 *BP-105 685,160 2,012,869 24 *BP-106 685,283 2,013,134 29 *BP-107 685,629 2,013,329 29 *BP-108 685,707 2,013,496 9 *BP-109 685,797 2,013,689 28 *BP-110 685,978 2,013,859 14 *BP-111 685,815 2,013,247 18 *BP-112 685,896 2,013,417 17 *BP-113 685,919 2,013,481 10 *BP-114 685,954 2,013,543 10 *BP-115 685,795 2,012,566 22 *BP-116 685,942 2,012,905 25 *BP-117 686,332 2,013,434 15 *BP-118 686,002 2,013,159 10 *BP-119 686,072 2,013,340 11 *BP-120 686,156 2,013,514 12 *BP-121 686,233 2,013,681 13 *BP-122 685,810 2,012,306 24 *BP-123 685,977 2,012,644 10 Amendment 65 Page 244 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BP-124 686,071 2,012,849 14 *BP-125 686,168 2,013,061 9 *BP-126 686,280 2,013,271 20 *BP-131 686,903 2,011,411 14 *BP-132 687,019 2,011,630 20 *BP-133 687,074 2,011,770 13 *BP-134 687,159 2,011,960 15 *BP-135 687,249 2,012,138 30 *BP-136 687,324 2,012,309 30 *BP-137 687,495 2,012,672 30 *BP-138 2,012,866 18 *BP-139 687,656 2,013,048 26 *BP-140 687,754 2,013,222 29 *BP-141 686,725 2,011,501 15 *BP-142 686,828 2,011,694 10 *BP-143 686,902 2,011,858 14 *BP-144 686,986 2,012,038 13 1c. BP Series - Plant Design Exploration (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BP-145 687,071 2,012,221 15 *BP-146 687,149 2,012,402 19 *BP-147 687,232 2,012,582 17 *BP-148 687,319 2,012,768 18 *BP-149 687,404 2,012,950 18 *BP-150 687,491 2,013,124 17 *BP-151 687,566 2,013,298 20 *BP-152 686,565 2,011,566 13 *BP-153 686,712 2,011,949 19 *BP-154 686,803 2,012,126 15 *BP-155 686,886 2,012,319 14 *BP-156 687,057 2,012,669 18 *BP-157 687,127 2,012,840 24 *BP-158 687,217 2,013,023 24 *BP-159 687,304 2,013,209 19 *BP-160 687,386 2,013,385 21 *BP-161 687,367 2,011,672 11 *BP-162 686,464 2,011,881 16 *BP-163 686,540 2,012,024 16 *BP-164 686,613 2,012,209 20 *BP-165 686,697 2,012,393 22 *BP-166 686,782 2,012,552 23 *BP-167 686,863 2,012,756 12 *BP-168 686,950 2,012,933 24 *BP-169 686,953 2,013,152 34 *BP-170 687,123 2,013,306 34 *BP-171 687,203 2,013,476 44 *BP-172 686,179 2,011,742 18 *BP-173 686,334 2,012,120 20 *BP-174 686,428 2,012,267 10 *BP-175 686,505 2,012,478 15 *BP-176 686,597 2,012,663 11 *BP-177 686,691 2,012,833 14 *BP-178 686,770 2,013,011 29 Amendment 65 Page 245 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BP-179 686,855 2,013,197 29 *BP-180 686,939 2,013,378 29 *BP-181 687,022 2,013,571 20 *BP-182 686,005 2,011,827 12 *BP-183 686,086 2,012,012 16 *BP-184 686,176 2,012,197 30 *BP-185 686,262 2,012,374 20 *BP-186 686,345 2,012,588 20 *BP-187 686,427 2,012,739 24 *BP-188 686,510 2,012,919 20 *BP-189 686,593 2,013,101 20 *BP-190 686,681 2,013,276 28 *BP-191 686,859 2,013,650 29 *BP-192 685,825 2,011,909 13 *BP-193 685,900 2,012,088 10 *BP-194 686,088 2,012,444 17 *BP-195 686,162 2,012,636 13 *BP-196 686,254 2,012,850 16 *BP-197 686,328 2,013,000 20 *BP-198 686,512 2,013,371 28 1c. BP Series - Plant Design Exploration (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BP-199 686,594 2,013,563 28 *BP-200 686,671 2,013,730 25 *BP-201 685,632 2,012,001 14 *BP-202 685,710 2,012,185 18 *BP-203 686,486 2,013,808 12 *BP-204 685,462 2,012,106 15 *BP-205 685,535 2,012,265 15 *BP-206 686,300 2,013,897 16 *BP-207 685,279 2,012,185 9 *BP-208 685,368 2,012,356 18 *BP-209 686,128 2,014,007 14 *BP-210 685,086 2,012,254 13 *BP-211 685,188 2,012,429 23 *BP-199 686,594 2,013,563 28 *BP-200 686,671 2,013,730 25 *BP-201 685,632 2,012,001 14 *BP-202 685,710 2,012,185 18 *BP-203 686,486 2,013,808 12 *BP-204 685,462 2,012,106 15 *BP-205 685,535 2,012,265 15 *BP-206 686,300 2,013,897 16 *BP-207 685,279 2,012,185 9 *BP-208 685,368 2,012,356 18 *BP-209 686,128 2,014,007 14 *BP-210 685,086 2,012,254 13 *BP-211 685,188 2,012,429 23 *BP-212 685,941 2,014,067 20 *BP-213 684,914 2,012,235 10 Amendment 65 Page 246 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BP-214 684,997 2,012,519 10 *BP-215 685,753 2,014,158 20 *BP-216 684,738 1,012,429 10 *BP-217 684,820 2,012,612 18 *BP-218 685,581 2,014,235 27 *BP-219 684,551 2,012,511 10 *BP-220 684,638 2,012,693 10 *BP-221 685,389 2,014,311 20 *BP-222 684,371 2,012,607 10 *BP-223 684,460 2,012,783 10 *BP-224 684,536 2,012,957 37 *BP-225 684,632 2,013,158 26 *BP-226 684,704 2,013,319 11 *BP-227 684,792 2,013,500 7 *BP-228 684,869 2,014,126 13 *BP-229 684,970 2,013,886 9 *BP-230 684,955 2,014,313 23 *BP-231 685,134 2,014,224 28 *BP-232 685,217 2,014,401 29 *BP-233 684,192 2,012,687 15 *BP-234 684,275 2,012,859 15 *BP-235 684,433 2,013,227 17 *BP-236 684,516 2,013,382 23 *BP-237 684,689 2,013,749 15 *BP-238 684,773 2,013,951 16 *BP-240 685,433 2,012,197 40 Amendment 65 Page 247 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1c. BP Series - Plant Design Exploration (Cont'd)Coordinates Borehole Depth Number North East (feet)(feet)

 *BP-241 685,453 2,012,241 40 *BP-242 685,347 2,012,247 35 *BP-243 685,360 2,012,282 35 *BP-244 685,799 2,013,374 38 *BP-245 685,824 2,013,414 37 *BP-246 685,713 2,013,418 41 *BP-247 685,751 2,013,446 38 *BP-248 684,953 2,013,082 58 *BP-249 684,995 2,013,133 63 *BP-250 684,544 2,013,302 61 *BP-251 686,960 2,012,990 50 *BP-252 686,979 2,013,142 50 *BP-253 685,617 2,013,438 42 *BP-254 685,638 2,013,499 37 *BP-255 685,248 2,012,275 30 *BP-256 685,265 2,012,347 30 1d. BC Series - Channels Explorations Coordinates Borehole Depth Number North East (feet)

(feet)BC-1 690,818 2,023,942 10 BC-2 690,628 2,023,871 30 BC-3 690,415 2,023,779 45 BC-4 690,748 2,024,356 20 BC-5 690,485 2,024,243 40 BC-6 690,236 2,024,138 45 BC-7 690,562 2,024,707 20 BC-8 690,331 2,024,596 25 BC-9 690,116 2,024,520 35 BC-10 689,988 2,024,901 30 BC-11 690,175 2,024,966 35 BC-12 690,018 2,024,903 35 BC-13 690,212 2,025,433 55 BC-14 690,048 2,025,340 65 BC-15 689,835 2,025,253 55 BC-16 690,109 2,025,820 40 BC-17 689,884 2,025,705 50 BC-18 689,665 2,025,615 71 BC-19 689,967 2,026,175 50 BC-20 689,743 2,026,081 35 BC-21 689,522 2,025,993 40 BC-22 689,813 2,026,566 45 BC-23 689,599 2,026,457 26 Amendment 65 Page 248 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1d. BC Series - Channels Explorations (Cont'd)Coordinates Borehole Depth Number North East (feet)(feet)BC-24 689,360 2,026,371 45 BC-25 689,659 2,026,934 50 BC-26 689,428 2,026,805 25 BC-27 689,221 2,026,748 35 BC-28 689,516 2,027,306 50 BC-29 689,292 2,027,199 20 BC-30 689,081 2,027,098 30 BC-31 689,364 2,027,661 30 BC-32 689,159 2,027,571 15 BC-33 688,923 2,027,481 30 BC-34 688,757 2,027,851 25 BC-35 688,629 2,028,226 15 BC-36 687,072 2,017,173 10 BC-37 686,851 2,017,261 10 BC-38 686,600 2,017,348 20 BC-39 687,145 2,017,507 35 BC-40 686,971 2,017,688 35 BC-41 686,748 2,017,762 35 BC-42 687,338 2,017,924 10 BC-43 687,118 2,018,016 15 BC-44 686,887 2,018,103 20

 *BC-45 687,372 2,007,956 20 *BC-46 687,414 2,008,125 35 *BC-47 687,421 2,008,332 30 *BC-48 687,524 2,008,514 15 BC-49 678,042 2,021,835 12 BC-50 678,182 2,022,096 22 BC-51 678,301 2,022,264 21 BC-52 677,698 2,022,057 12 BC-53 677,835 2,022,267 30 BC-54 677,945 2,022,472 15 BC-55 677,368 2,022,261 40 BC-56 677,493 2,022,467 38 BC-57 677,621 2,022,699 38 BC-58 677,031 2,022,499 29 BC-59 677,148 2,022,681 27 BC-60 677,276 2,022,876 36 BC-61 676,722 2,022,720 17 BC-62 676,806 2,022,885 30 BC-63 676,934 2,023,085 26 BC-64 676,355 2,022,915 13 BC-65 676,458 2,023,081 20 BC-66 676,572 2,023,306 31 BC-67 676,135 2,023,285 20 BC-68 676,244 2,023,509 38 BC-69 675,772 2,023,490 13 BC-70 675,896 2,023,708 28 Amendment 65 Page 249 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BC-71 683,816 2,012,211 40 *BC-72 683,904 2,012,391 50 *BC-73 683,989 2,012,570 60 *BC-74 684,056 2,012,748 64 BC-101 682,563 2,010,403 10 BC-102 682,782 2,010,316 20 1d BC Series - Channels Explorations (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)BC-103 682,586 2,010,774 17 BC-104 686,853 2,010,750 20 BC-105 682,715 2,011,175 25 BC-106 682,954 2,011,088 20 BC-107 682,793 2,011 582 15 BC-108 682,978 2,011,493 15 BC-109 682,937 2,011,957 15 BC-110 683,154 2,011,855 20 BC-111 683,098 2,012,316 10 BC-112 683,321 2,012,221 20 BC-113 683,416 2,012,389 20 BC-114 683,484 2,012,580 30

 *BC-115 684,556 2,013,813 55 *BC-116 683,801 2,012,659 35 *BC-117 683,649 2,012,257 25 *BC-118 684,478 2,013,641 55 *BC-119 684,396 2,013,465 60 *BC-120 684,298 2,013,257 65 *BC-121 684,236 2,013,120 65 *BC-122 684,137 2,012,924 60 BC-130 691,668 2,013,761 20 BC-131 691,625 2,014,025 35 BC-132 691,582 2,014,327 50 BC-133 691,517 2,014,602 30 *BC-140 682,480 2,009,982 22 *BC-141 682,681 2,009,908 9 *BC-142 682,416 2,009,621 10 *BC-143 686,522 2,012,591 35 *BC-144 686,062 2,011,593 35 *BC-145 685,669 2,010,783 36 *BC-146 684,799 2,013,174 20 *BC-147 684,435 2,012,376 20 *BC-148 684,056 2,011,570 25 *BC-149 684,199 2,010,620 35 *BC-150 686,343 2,012,226 20 *BC-151 686,173 2,011,859 14 *BC-152 685,934 2,011,314 19 *BC-153 685,837 2,011,135 17 *BC-154 685,740 2,010,471 19 *BC-155 684,684 2,012,922 21 *BC-156 684,555 2,012,645 17 *BC-157 684,315 2,012,108 18 *BC-158 684,179 2,011,839 19 Amendment 65 Page 250 of 369

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 *BC-159 683,987 2,011,190 24 *BC-160 684,091 2,010,898 21 *BC-161 684,302 2,010,350 15 *BC-162 684,407 2,010,049 11 *BC-165 687,407 2,008,244 65 *BC-166 687,369 2,008,037 50 *BC-170 685,840 2,010,230 16 *BC-171 685,930 2,010,310 15 1d. BP Series - Channels Explorations (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BC-172 686,020 2,010,490 40 *BC-173 686,070 2,010,670 19 *BC-174 686,180 2,010,850 13 *BC-175 686,260 2,011,100 60 *BC-176 686,255 2,011,360 55 *BC-177 686,240 2,011,570 49 *BC-178 686,180 2,012,300 50 *BC-179 684,405 2,010,210 30 *BC-180 684,130 2,010,530 20 *BC-181 683,990 2,010,695 39 *BC-182 683,870 2,010,830 35 *BC-183 683,740 2,010,980 33 *BC-184 683,600 2,011,200 34 *BC-185 683,510 2,011,480 40 *BC-186 683,510 2,011,780 23 *BC-187 683,600 2,012,040 36 *BC-188 683,880 2,012,850 11 *BC-189 683,970 2,013,030 15 *BC-190 684,150 2,013,270 29 *BC-191 684,185 2,013,710 65 *BC-192 684,010 2,013,800 35 *BC-193 683,900 2,013,910 40 BC-194 683,850 2,014,075 71 BC-195 683,840 2,014,370 18 BC-196 683,825 2,014,670 45 BC-197 683,830 2,014,970 35 BC-198 683,840 2,015,270 18 BC-199 683,840 2,015,570 21 BC-500 689,723 2,022,941 35 BC-501 689,351 2,022,609 30 BC-502 689,467 2,023,249 30 BC-503 689,266 2,023,102 35 BC-505 689,279 2,023,434 20 BC-521 687,753 2,024,730 65 BC-522 687,572 2,024,564 74 BC-523 687,658 2,025,189 65 BC-524 687,484 2,025,025 45 BC-525 687,300 2,024,870 55 Amendment 65 Page 251 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BC-526 687,396 2,025,495 44 BC-527 687,210 2,025,330 35 BC-529 687,108 2,025,788 26 BC-530 686,953 2,025,620 30 BC-531 686,747 2,025,458 40 BC-532 686,695 2,025,905 21 BC-533 686,480 2,025,730 40 BC-534 686,861 2,026,081 30 BC-535 686,580 2,026,376 35 BC-536 686,384 2,026,208 15 BC-537 686,226 2,025,998 30 BC-538 686,311 2,026,671 30 BC-539 686,112 2,026,510 20 BC-540 685,938 2,026,399 10 BC-541 686,048 2,026,949 27 BC-542 685,854 2,026,796 24 1e. BD Series - Exploration for Separating Dikes and Saddle Dams Coordinates Borehole Depth Number North East (feet)(feet)BD-1 685,345 2,014,831 58.0 BD-2 685,297 2,015,025 45.0 BD-3 685,250 2,015,220 43.0 BD-4 685,202 2,015,414 40.0 BD-5 685,154 2,015,608 45.0 BD-6 685,103 2,015,795 47.0 BD-7 685,063 2,016,003 91.0 BD-8 685,020 2,016,195 42.0 BD-9 684,968 2,016,389 41.0 BD-10 684,923 2,016,581 84.0 BD-11 685,081 2,014,457 31.0 BD-12 684,866 2,014,457 52.0 BD-13 684,705 2,014,383 64.0 BD-14 684,530 2,014,279 35.0 BD-15 684,351 2,014,189 74.0

 *BD-16 685,457 2,008,531 45.0 *BD-17 685,430 2,008,734 35.0 *BD-18 685,386 2,008,951 22.0 *BD-19 685,350 2,009,108 25.0 *BD-20 685,314 2,009,330 30.0 BD-21 682,268 2,012,810 40.0 BD-22 682,238 2,012,977 40.0 BD-23 682,426 2,013,132 40.0 BD-24 682,492 2,013,340 35.0 BD-25 682,557 2,013,531 41.0 BD-26 682,622 2,013,702 60.0 BD-27 682,678 2,013,858 51.5 BD-28 682,774 2,014,079 51.0 BD-29 682,843 2,014,262 51.0 BD-30 682,921 2,014,450 52.0 BD-31 683,888 2,017,726 37.0 Amendment 65 Page 252 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BD-32 683,734 2,017,836 43.5 BD-33 683,528 2,017,945 53.0 BD-34 683,423 2,018,011 64.9 BD-35 683,217 2,018,152 48.0 BD-36 683,039 2,018,264 57.0 BD-37 682,885 2,018,361 37.5 BD-38 681,751 2,018,445 40.5 BD-39 682,543 2,018,576 57.0 BD-40 682,321 2,018,719 40.5 BD-41 682,204 2,018,790 47.0 BD-42 681,989 2,018,912 41.0 BD-43 681,824 2,019,031 60.0 BD-44 681,269 2,019,536 50.0 BD-45 680,911 2,019,716 48.0 BD-46 680,543 2,019,898 54.0 BD-47 680,186 2,020,198 52.0 BD-56 678,445 2,021,114 30.0 BD-57 678,319 2,021,254 40.0 BD-58 678,124 2,021,392 59.0 BD-65 683,713 2,014,258 65.0 BD-66 683,526 2,014,338 69.0 1e. BD Series - Exploration for Separating Dikes and Saddle Dams (Contd)Coordinates Borehole Depth Number North East (feet)(feet)BD-67 683,320 2,014,395 50.0 BD-68 683,152 2,014,493 71.5 BD-101 682,471 2,004,934 16.0 BD-102 680,552 2,004,863 10.0 BD-103 680,341 2,004,837 13.0 BD-104 679,125 2,003,556 13.5 BD-105 670,055 2,004,135 8.0 BD-110 688,024 2,013,257 37.0 BD-111 688,260 2,013,358 37.0 BD-112 688,432 2,013,445 27.0 BD-113 688,706 2,013,570 36.5 BD-114 688,935 3,013,689 38.9 BD-115 689,129 2,013,809 74.0 BD-116 689,397 2,013,906 60.0 BD-117 689,617 2,014,011 50.0 BD-130 684,169 2,004,780 17.0 BD-131 681,899 2,019,704 9.0 BD-132 683,591 2,021,707 13.0 BD-133 674,954 2,018,497 14.0 BD-134 680,409 2,004,867 22.5 BD-135 680,146 2,004,810 30.0 BD-500 687,373 2,026,988 73.5 BD-501 687,240 2,027,084 53.0 BD-502 687,092 2,027,212 47.5 BD-503 686,901 2,027,378 57.0 BD-504 686,774 2,027,470 47.5 BD-505 686,488 2,027,726 37.0 BD-506 686,206 2,027,951 27.0 BD-507 685,850 2,028,210 22.0 Amendment 65 Page 253 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BD-510 684,920 2,028,970 27.0 BD-511 684,765 2,029,090 43.5 BD-512 684,558 2,029,221 33.5 BD-513 684,480 2,029,270 33.0 BD-514 684,275 2,029,467 40.0 BD-515 684,421 2,029,616 34.0 BD-516 684,167 2,029,294 40.0 BD-517 684,056 2,029,142 40.0 BD-518 684,276 2,029,759 52.0 BD-519 684,140 2,029,596 40.0 BD-520 684,006 2,029,448 44.0 BD-521 684,110 2,029,865 37.0 BD-522 683,995 2,029,695 44.1 BD-525 676,849 2,024,138 44.0 BD-526 676,675 2,024,037 55.0 BD-527 676,533 2,023,911 34.0 BD-528 676,640 2,024,210 53.8 BD-529 676,474 2,024,101 60.0 BD-530 676,312 2,023,984 57.0 BD-531 676,536 2,024,373 44.0 BD-532 676,203 2,024,161 60.0 BD-533 676,372 2,024,257 53.0 BD-534 676,266 2,024,429 40.0 BD-535 676,120 2,024,604 18.0 BD-536 676,001 2,024,861 21.0 BD-537 675,838 2,025,010 20.0 BD-538 675,675 2,025,062 21.0 BD-539 675,474 2,025,110 25.0 1f. BCT Series - Cooling Towers Foundations Explorations Coordinates Borehole Depth Number North East (feet)(feet)BCT-1 684,130 2,011,580 18.5 BCT-2 684,220 2,011,760 13.5 BCT-3 684,300 2,011,955 4.8 BCT-4 684,300 2,011,500 18.6 BCT-5 684,385 2,011,680 50.0 BCT-6 684,470 2,011,870 9.8 BCT-7 684,490 2,011,410 25.0 BCT-8 684,570 2,011,600 8.8 BCT-9 684,670 2,011,775 13.7 BCT-10 684,660 2,011,330 6.0 BCT-11 684,750 2,011,510 40.0 BCT-12 684,830 2,011,695 6.0 BCT-13 684,850 2,011,240 17.0 BCT-14 684,940 2,011,420 25.0 BCT-15 685,020 2,011,605 10.5 BCT-16 685,030 2,011,150 44.0 BCT-17 685,115 2,011,335 50.0 BCT-18 685,205 2,011,525 14.1 BCT-19 685,210 2,011,080 39.0 BCT-20 685,300 2,011,250 18.5 BCT-21 685,385 2,011,440 9.5 BCT-22 685,390 2,010,980 29.9 BCT-23 685,475 2,011,165 55.0 Amendment 65 Page 254 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BCT-24 685,560 2,011,350 9.1 BCT-25 685,510 2,014,505 50.0 BCT-26 685,595 2,014,690 9.6 BCT-27 685,680 2,014,430 33.5 BCT-28 685,765 2,014,610 9.7 BCT-29 685,855 2,014,345 45.0 BCT-30 685,945 2,014,525 9.0 BCT-31 686,040 2,014,255 13.5 BCT-32 686,120 2,014,440 13.5 BCT-33 686,200 2,014,165 20.0 BCT-34 686,310 2,014,250 4.7 BCT-35 686,410 2,014,080 22.5 BCT-36 686,495 2,014,260 13.9 BCT-37 685,590 2,013,990 55.0 BCT-38 686,685 2,014,170 14.6 BCT-39 686,860 2,014,090 7.4 BCT-40 686,940 2,014,270 26.2 BCT-41 686,770 2,014,350 15.0 BCT-42 686,580 2,014,440 15.0 BCT-43 686,395 2,014,530 8.0 BCT-44 686,210 2,014,625 10.5 BCT-45 686,020 2,014,700 11.5 BCT-46 685,850 2,014,790 33.5 BCT-47 685,680 2,014,870 28.5 BCT-48 685,495 2,014,955 23.5 BCT-49 685,435 2,014,770 9.9 BCT-50 685,325 2,014,590 30.0 BCT-51 684,040 2,011,390 4.9 BCT-52 684,210 2,011,315 4.6 BCT-53 684,395 2,011,230 5.0 BCT-54 684,575 2,011,150 8.8 BCT-55 684,765 2,011,060 33.6 1f. BCT Series - Cooling Towers Foundations Explorations (Contd)Coordinates Borehole Depth Number North East (feet)(feet)BCT-56 685,940 2,010,970 4.8 BCT-57 685,140 2,010,895 9.6 BCT-58 685,300 2,010,810 14.5 BCT-59 685,490 2,010,720 14.6 BCT-60 685,575 2,010,900 14.6 BCT-61 685,660 2,011,080 6.9 BCT-62 685,735 2,011,260 9.6 BCT-63 685,825 2,011,445 14.0 BCT-64 684,910 2,011,630 6.0 BCT-65 685,650 2,011,550 7.5 BCT-66 685,735 2,011,710 8.9 BCT-67 685,465 2,011,610 7.1 BCT-68 685,555 2,011,800 8.9 BCT-69 685,385 2,011,700 9.2 BCT-70 685,365 2,011,880 8.5 BCT-71 685,100 2,011,780 6.3 Amendment 65 Page 255 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BCT-72 685,180 2,011,965 4.5 BCT-73 684,915 2,011,875 5.0 BCT-74 685,000 2,012,045 4.9 BCT-75 684,735 2,011,960 13.5 BCT-76 684,855 2,012,115 3.8 BCT-77 684,550 2,012,216 13.0 BCT-78 684,640 2,012,215 4.9 BCT-79 684,380 2,012,120 6.0 BCT-80 684,460 2,012,295 6.9 1g. BX Series - Auxiliary Dam Exploration Coordinates Borehole Depth Number North East (feet)(feet)

 *BX-1 684,593 2,006,535 49.0 *BX-2 684,465 2,006,688 59.0 *BX-3 684,308 2,006,562 53.0 *BX-4 684,159 2,006,432 57.0 *BX-5 684,009 2,006,296 52.0 *BX-6 683,854 2,006,181 34.5 *BX-7 683,996 2,006,980 59.0 *BX-8 683,849 2,007,450 69.0 *BX-9 683,995 2,007,953 40.0 *BX-10 683,850 2,007,955 44.5 *BX-11 683,998 2,009,458 39.0 *BX-12 683,842 2,008,403 49.0 *BX-13 684,000 2,008,950 30.0 *BX-14 683,841 2,008,950 39.0 1g. BX Series - Auxiliary Dam Exploration (Contd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BX-15 683,833 2,009,439 54.0 *BX-16 683,850 2,009,819 59.0 *BX-17 684,564 2,006,224 59.0 BX-20 687,026 2,013,670 86.0 BX-21 687,178 2,013,864 83.0 BX-22 687,339 2,014,072 68.0 BX-23 687,481 2,014,255 55.0 BX-24 687,672 2,014,315 51.0 BX-25 687,599 2,014,407 38.0 BX-26 687,791 2,014,668 71.0 BX-27 687,937 2,014,856 32.0 Amendment 65 Page 256 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BX-28 688,079 2,015,053 55.0 BX-29 687,639 2,014,463 50.0 BX-30 688,234 2,015,246 61.0 BX-31 688,530 2,015,255 50.5 BX-32 688,425 2,015,485 54.0 BX-33 688,138 2,015,487 33.0 BX-34 687,510 2,014,106 78.0 BX-35 687,342 2,014,220 71.0 BX-36 687,806 2,014,489 41.5 BX-37 687,618 2,014,654 56.0 BX-38 688,305 2,015,350 54.0 BX-39 688,246 2,015,428 61.0 BX-39A 688,246 2,015,428 21.0

 *BX-40 684,200 2,006,980 38.5 *BX-41 684,115 2,006,795 87.0 *BX-42 684,050 2,006,730 86.0 *BX-43 683,970 2,006,680 75.0 *BX-44 683,940 2,006,865 75.0 *BX-45 683,850 2,006,810 75.0 *BX-46 683,870 2,006,700 70.0 *BX-47 683,670 2,006,615 65.0 *BX-48 683,470 2,006,665 11.5 *BX-49 683,340 2,006,790 12.5 *BX-50 683,280 2,006,980 45.0 *BX-51 683,220 2,007.170 23.0 *BX-52 683,175 2,007,360 20.0 *BX-53 683,125 2,007,550 45.0 *BX-54 683,085 2,007,770 9.0 *BX-55 683,010 2,007,940 9.0 *BX-56 682,970 2,008,130 40.0 *BX-57 682,895 2,008,320 14.0 *BX-58 682,720 2,008,450 25.5 *BX-59 684,055 2,006,863 75.0 *BX-60 684,010 2,006,935 75.0 *BX-61 683,980 2,006,800 75.0 *BX-62 684,370 2,006,458 70.0 *BX-63 684,255 2,006,625 68.0 *BX-64 684,295 2,006,430 75.0 *BX-65 684,175 2,006,565 75.0 *BX-66 684,220 2,006,385 80.0 *BX-67 684,120 2,006,570 73.0 Amendment 65 Page 257 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1h. BF Series - Exploration for Make up Water System Structures Coordinates Borehole Depth Number North East (feet)(feet)BF-1 661,067 2,011,275 35.5 BF-2 661,881 2,011,157 55.0 BF-3 661,946 2,003,020 81.0 BF-4 652,346 2,003,101 106.5 BF-5 652,083 2,003,066 71.0 BF-6 652,160 2,003,040 10.0 BF-7 652,280 2,030,040 10.0 BF-8 684,497 2,006,485 26.0 BF-9 684,067 2,006,139 27.0 BF-10 688,664 2,015,265 30.0 BF-11 688,294 2,015,571 30.0 BF-101 661,150 2,011,270 3.5 BF-102 661,240 2,011,260 0.5 BF-103 661,361 2,011,239 13.5 BF-104 661,491 2,011,221 13.0 BF-105 661,635 2,011,201 7.0 BF-106 661,777 2,011,179 18.0 BF-107 684,282 2,006,297 15.0 BF-108 684,160 2,006,197 15.5 BF-110 688,620 2,015,330 14.0 BF-111 688,475 2,015,436 16.5 BF-112 688,385 2,015,511 13.5 BF-113 683,265 2,008,700 10.5 BF-114 683,340 2,008,880 6.5 BF-115 683,430 2,009,060 55.0 BF-116 683,470 2,009,140 60.0 BF-117 683,505 2,009,250 65.0 BF-118 683,937 2,009,735 83.5 BF-119 683,610 2,009,420 9.0 BF-120 683,700 2,009,600 10.5 BF-121 652,595 2,002,720 24.0 BF-122 652,875 2,002,560 26.5 BF-123 653,174 2,002,473 25.0 BF-124 653,655 2,002,427 25.0 BF-125 654,039 2,002,484 9.5 BF-126 654,420 2,002,555 27.0 BF-127 654,755 2,002,775 10.0 BF-128 655,055 2,002,985 10.0 BF-129 655,338 2,003,183 4.5 BF-130 655,664 2,003,412 10.5 BF-131 655,866 2,003,554 8.5 BF-132 656,181 2,003,774 11.5 BF-133 656,540 2,004,026 12.5 BF-134 656,863 2,004,253 3.0 BF-135 657,184 2,004,402 7.0 BF-136 657,635 2,004,575 48.0 BF-137 657,945 2,004,694 22.0 BF-138 658,245 2,004,810 10.5 BF-139 658,585 2,004,940 7.5 BF-140 658,889 2,005,057 2.0 BF-141 659,105 2,005,160 15.0 BF-142 659,325 2,005,480 7.0 BF-143 659,379 2,005,900 27.5 Amendment 65 Page 258 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BF-144 659,417 2,006,225 10.5 BF-145 659,466 2,006,650 17.0 BF-146 659,504 2,006,981 35.0 BF-147 659,552 2,007,395 32.5 BF-148 659,577 2,007,610 8.0 1i. X,M, and A Series - Preliminary Subsurface Investigation, Main Dam Vicinity Coordinates Borehole Depth Number North East (feet)(feet)X-1 660,958 2,010,121 90.0 X-2 661,150 2,010,470 100.0 X-3 661,388 2,010,901 97.0 M-1 664,300 2,007,820 M-2 664,120 2,008,287 M-3 663,941 2,008,753 M-4 663,761 2,009,220 M-5 663,582 2,009,687 M-6 663,402 2,010,153 M-7 663,223 2,010,620 A-1 652,914 2,004,140 A-2 652,557 2,003,790 A-3 652,200 2,003,440 1j. BM Series - Main Dam Exploration Coordinates Borehole Depth Number North East (feet)(feet)

 *BM-1 660,972 2,009,204 104.0 *BM-2 660,852 2,009,178 130.0 *BM-3 660,718 2,009,148 148.6 *BM-4 660,527 2,009,143 121.0 *BM-5 660,442 2,009,288 81.0 *BM-6 660,328 2,009,406 48.5 *BM-7 660,178 2,009,603 45.8 *BM-8 660,898 2,009,970 96.0 *BM-9 660,956 2,010,108 74.0 *BM-10 661,070 2,010,288 71.0 *BM-11 661,138 2,010,463 75.0 *BM-12 661,190 2,010,554 90.0 *BM-13 661,233 2,010,663 101.0 *BM-14 661,319 2,010,804 95.0 *BM-15 661,417 2,010,980 78.5 *BM-16 661,541 2,011,166 93.5 *BM-17 661,299 2,010,421 40.3 *BM-18 661,363 2,010,581 39.0 *BM-19 661,484 2,010,781 56.5 *BM-20 660,950 2,010,340 36.0 *BM-21 661,017 2,010,552 71.0 *BM-22 661,096 2,010,733 39.3 Amendment 65 Page 259 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BM-23 661,212 2,010,867 39.0 *BM-24 661,237 2,011,010 30.0 *BM-25 661,618 2,010,983 98.2 *BM-26 661,730 2,011,141 87.0 *BM-27 661,845 2,011,293 57.0 *BM-28 661,214 2,010,116 133.0 *BM-29 661,075 2,009,952 98.0 *BM-30 661,008 2,009,769 48.0 *BM-31 661,327 2,010,400 120.5 *BM-32 661,178 2,010,548 120.0 1j. BM Series - Main Dam Exploration (Contd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BM-33 661,368 2,010,370 71.0 *BM-34 661,269 2,010,275 85.0 *BM-35 661,662 2,011,122 90.5 *BM-36 661,656 2,011,419 85.2 *BM-37 661,057 2,009,860 91.0 *BM-38 660,863 2,009,765 111.0 *BM-39 660,866 2,009,834 102.0 *BM-40 660,694 2,009,737 99.0 *BM-41 660,696 2,009,848 95.0 *BM-42 660,702 2,009,943 80.0 *BM-43 660,528 2,009,721 78.5 *BM-44 660,537 2,009,835 81.0 *BM-45 660,546 2,009,896 75.5 *BM-46 660,358 2,009,691 67.7 *BM-47 660,358 2,009,763 61.0 *BM-48 660,356 2,009,853 53.5 *BM-49 660,169 2,009,720 80.0 *BM-50 660,182 2,009,822 73.0 *BM-51 659,992 2,009,668 62.0 1k. BL Series - Exploration for Low Level Release System Pipeline Coordinates Borehole Depth Number North East (feet)

(feet)BL-1 661,065 2,009,817 4.0 BL-2 661,196 2,009,819 16.5 BL-3 661,290 2,009,804 18.5 BL-4 661,354 2,009,794 19.0 BL-5 661,443 2,009,780 20.0 BL-6 661,514 2,009,768 3.0 Amendment 65 Page 260 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1l. BA Series - Afterbay Dam Exploration Coordinates Borehole Depth Number North East (feet)(feet)BA-1 652,230 2,003,316 110.0 BA-2 652,350 2,003,440 90.0 BA-3 652,524 2,003,593 82.0 BA-4 652,397 2,003,699 80.0 BA-5 652,677 2,003,770 81.0 BA-6 652,535 2,003,839 71.0 BA-7 652,802 2,003,890 104.0 BA-8 652,937 2,004,038 124.0 BA-9 653,127 2,004,084 131.0 BA-10 653,312 2,004,146 141.0 BA-11 653,521 2,004,215 133.0 BA-12 653,684 2,004,248 130.0 BA-13 653,295 2,004,299 95.0 BA-14 653,093 2,004,327 63.0 BA-15 652,896 2,004,354 59.0 BA-16 652,707 2,004,388 91.0 BA-17 652,534 2,004,468 90.0 BA-18 653,319 2,004,597 104.0 BA-19 653,408 2,004,871 92.0 BA-20 652,210 2,003,180 60.0 BA-21 652,326 2,002,858 62.0 BA-22 652,384 2,002,743 96.0 BA-23 652,432 2,002,570 43.5 BA-24 652,503 2,002,419 64.0 BA-25 653,503 2,005,159 63.0 BA-26 653,578 2,005,431 79.0 1m. BW Series - Skimmer Wall Exploration Coordinates Borehole Depth Number North East (feet)(feet)BW-1 669,242 2,014,841 41.5 BW-2 669,077 2,015,039 41.5 BW-3 668,875 2,015,250 41.5 BW-4 668,681 2,015,465 47.0 BW-5 668,472 2,015,691 48.0 BW-6 668,192 2,015,977 57.0 BW-7 668,000 2,016,183 47.0 BW-8 667,682 2,016,532 48.0 Amendment 65 Page 261 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1n. TK Series - Exploration for Norfolk Southern Railroad (Main Line and Durham Branch Relocation)Coordinates Borehole Depth Number North East (feet)(feet)TK-1 661,365 2,005,222 15.0 TK-2 661,314 2,005,415 12.0 TK-3 661,034 2,006,479 17.0 TK-4 660,983 2,006,672 24.0 TK-5 660,936 2,006,867 34.0 TK-6 660,895 2,007,062 35.0 TK-7 660,863 2,007,260 29.4 TK-8 660,834 2,007,458 35.0 TK-9 660,805 2,007,656 28.5 TK-11 660,748 2,008,051 24.0 TK-12 660,720 2,008,249 38.0 TK-13 660,663 2,008,447 34.0 TK-14 660,663 2,008,645 11.0 TK-15 660,634 2,008,843 29.0 TK-16 660,606 2,009,041 20.0 TK-17 660,577 2,009,239 21.1 TK-18 660,549 2,009,437 29.0 TK-19 660,520 2,009,635 14.0 TK-20 660,267 2,011,417 20.0 TK-21 660,254 2,011,616 41.0 TK-22 660,258 2,011,791 17.0 TK-30A 662,317 1,994,473 10.0 TK-31A 663,035 1,994,691 12.0 TK-32A 663,178 1,994,735 17.0 TK-33A 664,033 1,995,013 17.0 TK-34A 664,169 1,995,076 15.0 TK-35A 664,902 1,995,593 13.0 TK-36A 665,010 1,995,696 14.0 TK-37 665,483 1,996,143 10.0 TK-38 665,591 1,996,246 6.1 TK-39A 666,536 1,997,140 19.0 TK-40A 666,681 1,997,277 22.0 TK-41A 666,826 1,997,414 17.0 TK-42A 666,971 1,997,522 18.0 TK-43A 667,719 1,998,214 14.0 TK-44A 667,888 1,998,322 14.0 TK-45 668,063 1,998,419 7.0 TK-46A 670,309 1,998,829 6.0 TK-47 670,508 1,998,848 14.5 TK-48 670,707 1,998,867 8.1 TK-49 670,906 1,998,886 7.2 TK-50A 672,001 1,998,990 8.0 TK-50AA 671,703 1,998,961 13.0 TK-51 672,399 1,999,028 8.9 TK-52 673,693 1,999,150 12.0 TK-53A 673,893 1,999,169 13.0 TK-54 674,042 2,000,183 11.0 TK-55A 675,983 1,999,367 12.0 TK-56 676,182 1,999,386 7.4 TK-57A 676,381 1,999,405 11.0 TK-58A 676,580 1,999,424 7.0 Amendment 65 Page 262 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TK-59A 676,779 1,999,443 11.0 TK-60A 676,978 1,999,462 14.0 TK-61 677,177 1,999,483 8.5 1n. TK Series - Exploration for Norfolk Southern Railroad (Main Line and Durham Branch Relocation) (Cont'd)Coordinates Borehole Depth Number North East (feet)(feet)TK-62A 679,016 2,000,610 10.0 TK-63A 679,164 2,000,744 10.0 TK-64A 680,286 2,001,657 11.5 TK-65A 681,525 2,002,013 15.0 TK-66A 681,724 2,002,034 8.5 TK-67 683,570 2,002,122 10.0 TK-68 683,768 2,002,143 15.2 TK-69A 683,165 2,002,164 13.1 TK-70 683,165 2,002,192 7.7 TK-71 684,101 2,002,530 8.8 TK-72 684,271 2,002,635 11.0 TK-73 684,433 2,002,753 9.5 TK-74 684,928 2,003,245 10.5 TK-75 685,046 2,003,406 14.5 TK-76 685,161 2,003,570 10.3 TK-77 685,276 2,003,733 7.0 TK-78A 685,391 2,003,897 12.0 TK-79 685,506 2,004,061 11.0 TK-80A 685,621 2,004,224 12.0 TK-81 685,708 2,004,347 7.0 TK-82 685,851 2,004,551 8.0 TK-83 686,868 2,005,737 9.0 TK-84 687,002 2,006,187 9.9 TK-85 687,117 2,006,351 10.0 TK-86A 687,232 2,006,514 16.0 TK-87 687,318 2,006,637 8.4 TK-88 688,037 2,007,659 9.0 TK-89A 688,152 2,007,823 14.5 TK-90 688,267 2,007,986 13.0 TK-91 688,382 2,008,150 14.5 TK-92 688,497 2,008,313 13.0 TK-93 689,576 2,009,460 6.2 1o. BR, RB, BHN, BG, SD, and SR Series - Exploration for Highway and Railroad Bridges and Embankments Coordinates Borehole Depth Number North East (feet)(feet)BR-1 658,736 2,007,964 43.0 BR-2 658,438 2,007,945 37.0 BR-3 658,530 2,007,983 38.0 BR-4 658,645 2,007,997 40.0 BR-101 659,162 2,007,967 7.5 Amendment 65 Page 263 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BR-102 658,968 2,008,003 19.5 BR-103 658,123 2,007,973 28.0 BR-104 662,614 2,005,436 10.0 BR-106 662,722 2,005,978 13.0 BR-107 662,288 2,006,317 15.0 RB-1 662,247 2,006,556 4.1 RB-2 659,964 2,001,678 18.7 RB-3 660,029 2,001,823 3.9 RB-4 659,616 2,001,386 13.6 1o. BR, RB, BHN, BG, SD, and SR Series - Exploration for Highway and Railroad Bridges and Embankments (Contd)Coordinates Borehole Depth Number North East (feet)(feet)RB-5 659,654 2,001,670 13.7 RB-6 659,710 2,011,915 9.1 BHN-1 660,511 2,011,098 15.0 BHN-2 660,409 2,011,109 25.0 BHN-3 660,306 2,011,119 25.0 BHN-4 660,207 2,011,105 25.0 BHN-5 660,108 2,011,091 20.0 BHN-6 660,561 2,011,106 10.5 BG-1 660,491 2,009,736 40.0 BG-2 660,486 2,009,767 48.5 BG-3 660,474 2,009,853 39.0 BG-4 660,469 2,009,884 23.0 BG-5 660,724 2,007,800 28.0 BG-6 660,730 2,007,761 44.0 BG-7 660,757 2,007,780 38.0 BG-8 660,763 2,007,740 59.1 BG-9 660,816 2,007,714 48.0 BG-10 660,822 2,007,714 57.1 BG-11 660,866 2,007,730 74.1 BG-12 660,872 2,007,685 76.3 BG-13 658,430 2,008,020 22.6 BG-14 658,530 2,008,020 14.0 BG-15 658,630 2,008,020 18.6 BG-16 658,730 2,008,020 19.5 BG-17 661,948 1,992,270 20.5 BG-18 661,924 1,992,336 30.0 BG-19 661,903 1,992,393 13.5 BG-20 661,880 1,992,459 25.0 SD-1 662,247 2,006,556 6.2 SD-2 662,454 2,006,339 14.8 SD-3 662,661 2,006,122 10.3 SD-4 662,868 2,005,905 10.0 SR-1A 659,615 2,008,000 20.0 SR-2A 659,815 2,007,999 12.5 SR-3A 660,014 2,007,985 13.8 Amendment 65 Page 264 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1p. FB Series - Exploration for Quarry Sites Coordinates Borehole Depth Number North East (feet)(feet)FB-1 654,925 2,002,025 76.0 FB-2 654,279 2,001,406 75.0 FB-3 655,222 2,001,589 90.0 FB-4 655,258 2,000,997 120.0 FB-11 671,441 2,018,130 58.0 FB-12 670,582 2,017,226 54.0 FB-13 671,069 2,017,801 52.0 FB-14 671,146 2,017,066 52.0 FB-15 671,738 2,017,443 77.0 FB-16 661,885 2,013,560 32.0 1p. FB Series - Exploration for Quarry Sites (Contd)Coordinates Borehole Depth Number North East (feet)(feet)FB-17 661,910 2,012,920 28.0 FB-18 662,260 2,013,250 42.0 FB-19 662,800 2,012,720 58.0 FB-20 662,800 2,011,850 45.5 FB-21 669.940 2,016,464 55.0 FB-22 670,408 2,016,228 60.0 FB-23 670,325 2,017,146 39.0 FB-24 670,835 2,017,146 56.5 FB-25 670,782 2,017,536 60.0 FB-26 671,056 2,017,383 80.0 FB-27 671,425 2,017,469 77.0 FB-28 671,524 2,017,864 59.0 FB-29 672,067 2,017,678 60.0 FB-30 672,150 2,017,126 60.0 FB-31 662,368 2,014,582 68.0 FB-32 663,107 2,014,593 90.0 FB-33 659,798 2,005,905 90.0 FB-34 659,847 2,006,969 82.8 FB-34A 659,847 2,006,969 42.0 FB-35 660,215 2,006,543 85.0 FB-35A 660,215 2,006,543 39.0 FB-36 660,328 2,006,045 60.0 FB-36A 660,328 2,006,045 19.0 FB-37 660,536 2,006,852 106.0 FB-37A 660,536 2,006,852 27.0 1q. BB Series - Borrow Areas Auger Borings Coordinates Borehole Depth Number North East (feet)(feet)

 *BB-1 687,395 2,016,599 50 Amendment 65 Page 265 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BB-2 687,204 2,016,401 56 *BB-3 686,802 2,015,995 51 *BB-4 686,804 2,016,594 61 *BB-5 682,603 2,006,800 44 *BB-6 682,599 2,007,399 41.5 *BB-7 682,018 2,006,782 34.7 *BB-8 682,002 2,007,400 61.0 *BB-101 687,575 2,016,197 32.0 *BB-102 687,593 2,016,393 30.0 *BB-103 687,584 2,016,604 25.0 *BB-104 687,595 2,016,808 19.0 *BB-105 687,390 2,016,014 23.0 *BB-106 687,396 2,016,202 26.0 *BB-107 687,379 2,016,383 28.0 *BB-108 687,397 2,016,393 18.0 *BB-109 687,203 2,016,010 31.0 *BB-110 687,209 2,016,193 17.0 *BB-111 687,200 2,016,600 23.0 *BB-112 687,195 2,016,792 20.0 *BB-113 687,005 2,016,007 14.0 *BB-114 686,992 2,016,193 19.0 *BB-115 686,990 2,016,409 28.0 *BB-116 686,995 2,016,591 28.0 1q. BB Series - Borrow Areas Auger Borings (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BB-117 687,014 2,016,805 18.0 *BB-118 686,798 2,016,193 15.0 *BB-119 686,810 2,016,377 23.0 *BB-120 686,806 2,016,792 22.0 *BB-121 686,570 2,016,009 23.0 *BB-122 686,638 2,016,199 28.5 *BB-123 686,590 2,016,389 30.0 *BB-124 686,624 2,016,610 23.0 *BB-125 686,599 2,016,592 18.0 *BB-127 684,047 2,008,384 30.0 *BB-137 684,600 2,007,800 18.0 *BB-143 684,600 2,009,128 11.0 *BB-153 684,194 2,008,393 14.0 *BB-155 683,006 2,006,618 16.0 *BB-156 682,988 2,006,809 14.0 *BB-157 683,009 2,007.003 6.0 *BB-158 682,991 2,007,200 8.0 *BB-159 683,000 2,007,400 18.0 *BB-160 683,000 2,007,605 11.0 *BB-161 682,800 2,006,606 17.0 *BB-162 682,809 2,006,795 10.0 *BB-163 682,794 2,007,006 7.5 *BB-164 682,803 2,007,203 7.0 *BB-165 682,782 2,007,406 24.0 *BB-166 682,794 2,007,597 9.0 *BB-167 682,597 2,006,597 15.0 *BB-168 682,597 2,006,997 15.0 *BB-169 682,591 2,007,207 16.0 *BB-170 682,591 2,007,597 25.0 Amendment 65 Page 266 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 *BB-171 682,391 2,006,597 16.0 *BB-172 682,409 2,006,805 11.0 *BB-173 682,397 2,007,005 10.0 *BB-174 682,401 2,007,202 15.0 *BB-175 682,400 2,007,392 22.0 *BB-176 682,400 2,007,596 23.0 *BB-177 682,200 2,006,606 10.0 *BB-178 682,210 2,006,800 23.0 *BB-179 682,191 2,007,005 10.0 *BB-180 682,218 2,007,206 14.0 *BB-181 682,205 2,007,400 15.0 *BB-182 682,218 2,007,597 17.5 *BB-183 682,027 2,006,599 12.0 *BB-184 681,997 2,007,009 15.0 *BB-185 682,003 2,007,207 15.0 *BB-186 681,981 2,007,600 18.0 *BB-187 661,085 2,010,920 11.0 *BB-188 660,941 2,011,059 8.5 *BB-189 660,795 2,011,200 5.0 *BB-803 660,711 2,012,787 11.0 *BB-806 663,233 2,006,200 18.5 *BB-807 663,205 2,006,576 23.0 *BB-809 665,400 2,009,600 9.5 *BB-811 664,975 2,009,600 10.0 *BB-812 664,473 2,005,628 18.0 *BB-813 664,649 2,005,877 27.5 *BB-813A 664,652 2,005,865 7.0 *BB-813B 664,641 2,005,869 1.0 *BB-813C 664,645 2,005,882 6.0 1q. BB Series - Borrow Areas Auger Borings (Cont'd)

Coordinates Borehole Depth Number North East (feet)(feet)

 *BB-813D 664,650 2,005,875 13.0 *BB-814 664,470 2,006,225 50.0 *BB-815 664,654 2,006,602 29.0 BB-835 658,975 2,011,775 16.5 *BB-858 665,851 2,006,007 8.0 *BB-859 665,847 2,006,646 10.0 *BB-860 665,850 2,007,197 11.0 BB-861A 655,300 2,007,200 0.5 BB-861B 655,300 2,007,200 2.0 *BB-862 665,101 2,006,660 9.0 *BB-863 664,489 2,007,247 10.0 *BB-864 664,497 2,007,747 12.0 *BB-865 664,003 2,006,246 5.0 *BB-866 663,998 2,007,253 19.0 *BB-867 664,001 2,007,747 7.0 *BB-868 663,610 2,007,278 16.0 *BB-869 663,394 2,006,250 12.0 BB-870 653,897 2,003,744 40.0 BB-871 653,580 2,004,180 45.0 BB-871A 653,580 2,004,180 30.0 BB-872 653,165 2,005,077 40.0 BB-873 653,535 2,004,145 21.0 BB-874 653,493 2,005,487 40.0 Amendment 65 Page 267 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BB-875 653,622 2,006,077 40.0 BB-876 653,778 2,016,856 40.0 BB-877 652,357 2,006,369 18.0 BB-878 652,384 2,007,019 20.0 BB-879 652,929 2,007,585 40.0 BB-880 653,714 2,008,192 40.0 BB-895 657,519 2,002,802 21.0 BB-896 657,430 2,003,841 3.0 BB-896A 657,430 2,003,841 4.5 BB-902 655,638 2,002,992 13.0 BB-903 655,184 2,002,938 9.5 BB-904 654,580 2,003,000 33.0 BB-906 654,641 2,006,925 34.0 BB-907 654,510 2,007,470 39.0 BB-908 654,137 2,008,134 39.0 BB-916 653,990 2,003,573 14.3 BB-917 653,001 2,003,771 14.0 BB-918 654,012 2,003,952 33.5 BB-919 653,841 2,003,572 7.2 BB-920 653,850 2,003,971 16.0 BB-921 653,853 2,003,967 25.5 BB-922 653,845 2,004,374 9.0 BB-924 653,741 2,003,775 10.2 BB-925 653,747 2,003,996 10.0 BB-926 653,721 2,004,181 35.0 BB-927 653,709 2,004,384 30.0 BB-928 653,575 2,003,575 14.3 BB-929 653,586 2,003,777 5.0 BB-930 653,586 2,003,981 14.0 BB-931 653,584 2,004,387 30.0 BB-932 653,392 2,003,769 16.0 BB-933 653,374 2,003,982 19.0 1r. TPM, TPY, and TPZ Series - Borrow Area Test Pits Coordinates Borehole Depth Number North East (feet)(feet)

 *TPM-1 664,330 2,005,625 8.0 *TPM-2 664,450 2,006,475 9.0 *TPM-3A 664,029 2,006,701 8.25 *TPM-4A 663,356 2,005,556 11.0 *TPM-5A 663,847 2,006,393 5.5 *TPM-6 662,976 2,006,291 6.0 *TPY-1 687,300 2,016,152 10.0 *TPY-2A 687,166 2,016,373 12.0 *TPY-3 686,861 2,016,159 7.0 *TPY-4A 686,526 2,016,183 6.5 *TPZ-1 682,700 2,006,835 9.0 *TPZ-2 682,700 2,007,492 5.7 *TPZ-3 682,200 2,006,858 7.5 *TPZ-4 682,195 2,007,500 10.0 Amendment 65 Page 268 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1s. TPA Series - Auxiliary Dam Test Pits Coordinates Borehole Depth Number North East (feet)(feet)

 *TPA-1 683,800 2,007,400 10.0 *TPA-2 683,800 2,009,325 7.0

2. LOGS OF BOREHOLES AND TEST PITS

a. P Series Boreholes APPENDIX 2.5B LABORATORY ANALYSES OF FOUNDATION MATERIALS FOR DAMS AND DIKE APPENDIX 2.5C PRECONSTRUCTION BORROW MATERIAL TESTING 2.5C.1 INDEX PROPERTIES OF BORROW MATERIAL This section of Appendix 2.5C discusses the index properties of the borrow materials proposed during the preconstruction stage for the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike. These index properties are for samples obtained from the test pits dug in the borrow areas.

The testing program for the evaluation of the static and dynamic engineering properties used in the embankment stability analysis is discussed in Section 2.5C.2 of this Appendix.BORROW AREA TEST PITS Fourteen test pits were excavated in proposed borrow areas and a total of two test trenches and six shallow borings were made in foundation soils for the Auxiliary Dam and Auxiliary Reservoir Separating Dike. Representative soil samples were obtained from each pit, four undisturbed block samples were obtained from the trenches, and six thin-wall Shelby tube samples were obtained from the borings.The test pit logs are given in Appendix 2.5A.TEST PIT COMPOSITE SAMPLES An approximate 300-lb. representative sample was obtained from each test pit. Each sample contained the proper proportion of the different types of soil observed in the pit.BORROW AREA COMPOSITE SAMPLES Selected composite test pit samples were mixed to prepare the composite sample for each borrow area. Selection was based on inspection of the test pits, examination of the test pit composite samples, the results of grain-size analyses and liquid and plastic limit tests Amendment 65 Page 269 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 performed on the composite test pit samples, and the index property of samples from borings previously made in the borrow areas.a) Main Dam Composite Sample (M)Four of the six test pit composite samples were used to prepare the representative composite sample for the main dam borrow area. The samples from TPM4A and TPM5A were not used because they were primarily cohesionless with very low plasticity and based on grain-size analysis, were unrepresentative of the soils encountered in the other four pits or in the borings made in the borrow area. It is noted that these exploratory test pits were made to establish the extent of the borrow area suitable for use as construction material.The percent passing the No. 200 sieve, the liquid limit, and the plasticity index obtained for each thoroughly mixed test pit composite sample are summarized below.Composite % Passing Liquid Plasticity Sample No. 200 Sieve Limit Index TPM1 45 29 7 TPM2 40 26 6 TPM3A 42 29 12 TPM4A 36 21 2 TPM5A 38 23 4 TPM6 42 28 4 Approximately 1000 lb. of the composite main dam borrow sample was prepared by mixing the test pit composite samples from TPM1, TPM2, TPM3A, and TPM6. Index property tests were performed on the composite samples. These tests consisted of grain-size analyses, liquid and plastic limit and specific gravity determinations, and standard compaction tests, ASTM D698-68T Method A. The results of these tests are given on page 2.5C.1-5 and summarized as follows.Amendment 65 Page 270 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Main Dam Composite Borrow Sample Red-brown silty clayey coarse to fine sand with trace of fine gravel Unified Soil Classification SC

 % Passing No. 200 Sieve 44 Liquid Limit 33 Plastic Limit 22 Plasticity Index 11 Standard Maximum Dry Unit Weight, lb./ft.3 121.8 Standard Optimum Water Content,% 12.4 Specific Gravity 2.72 b) Auxiliary Dam Composite Sample (Z)

All four of the test pit composite samples were used to prepare the representative composite sample for the auxiliary dam borrow area Z. The percent passing the No. 200 sieve, liquid limit and plasticity index for each thoroughly mixed test pit composite sample are summarized below.Composite % Passing Liquid Plasticity Sample No. 200 Sieve Limit Index TPZ1 47 28 6 TPZ2 79 34 11 TPZ3 80 24 4 TPZ4 82 29 7 Approximately 1000 lb. of the composite auxiliary dam borrow sample was prepared by mixing the test pit composite samples. Index property tests were performed on the composite sample.The results of these tests are given on page 2.5C.1-7 and summarized as follows.Auxiliary Dam Composite Borrow Sample (Z)Brown silty clay with some coarse to fine sand with trace of fine gravel Unified Soil Classification CL

 % Passing No. 200 Sieve 68 Liquid Limit 35 Plastic Limit 22 Plasticity Index 13 Standard Maximum Dry Unit Weight, lb./ft.3 114.1 Standard Optimum Water Content,% 15.5 Specific Gravity 2.73 c) Auxiliary Dam Composite Sample (Y)

All four of the test pit composite samples were used to prepare the representative composite sample for the auxiliary dam borrow area Y. The percent passing the No. 200 sieve, liquid and plasticity index obtained for each thoroughly mixed test pit composite sample are summarized below.Amendment 65 Page 271 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Composite % Passing Liquid Plasticity Sample No. 200 Sieve Limit Index TPY1 79 32 8 TPY2A 88 37 11 TPY3 67 29 9 TPY4A 86 33 8 Approximately 1000 lb. of the composite auxiliary dike borrow sample was prepared by mixing the test pit composite samples.Index property tests were performed on the composite sample. The results of these tests are given on page 2.5C.1-6 and summarized as follows.Auxiliary Dam Composite Borrow Sample (Y)Red brown silty clay with some coarse to fine sand Unified Soil Classification CL

 % Passing No. 200 Sieve 80 Liquid Limit 37 Plastic Limit 24 Plasticity Index 13 Standard Maximum Dry Unit Weight, lb./ft.3 114.8 Standard Optimum Water Content,% 15.4 Specific Gravity 2.76 d) Plasticity Index Comparison Plasticity index values of the composite borrow samples were estimated on the basis of visual and manual examination of the composite test pit soils and calculated as the average value of the soils obtained from the borings in each borrow area. These values and the values obtained in the laboratory are given below.

Composite Borrow Samples Plasticity Index Values Main Auxiliary Auxiliary Dam Dam (Z) Dam (Y)Estimated from composite test pit soils 13 11 12 Calculated average of samples from borings 13 15 11 WMAI laboratory tests 11 13 13 The plasticity index values from laboratory tests on the composite borrow area samples are in essential agreement with the estimated and calculated average values which give confirmation that the composite soils are representative of their respective borrow areas.TEST TRENCHES Two test trenches were excavated with a Case 580B backhoe for the purpose of obtaining undisturbed representative block samples of the foundation soils for the Auxiliary Dam. The Amendment 65 Page 272 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 trenches, identified as TPA1 and TPA2, were excavated in the foundation for the Auxiliary Dam.Each trench was inspected and logged.The trench logs are given in Appendix 2.5A.UNDISTURBED BLOCK SAMPLES Four undisturbed block samples were recovered from the test trenches; three from TPA1 and one from TPA2 at the auxiliary dam site. The block samples were cut in-situ to approximately 1-ft. cubes and sealed with wax in wooden boxes.SHELBY TUBE SAMPLES Six 3-in.-diameter thin-wall Shelby tube samples were taken from borings located 10 ft. to 20 ft.from the test trenches. The borings were made in groups of three near trenches TPA1 and TPA2. The borings in each group were approximately two ft. apart and each sample was taken from two ft. to four ft. below ground surface which is generally above the depths at which the block samples were taken.2.5C.2 LABORATORY TESTING PROGRAM FOR STATIC AND DYNAMIC ENGINEERING PROPERTIES INTRODUCTION This section presents results of laboratory investigation of the static and dynamic properties of the core material of the Main Dam (Material M), and the core material (Material Z) and foundation soils for the Auxiliary Dam and Auxiliary Reservoir Separating Dike. Results of these tests were utilized for the evaluation of the seismic stability of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike during the preconstruction stage of engineering and design.Subsequent changes in borrow areas and material properties criteria are discussed in Appendix 2.5F, "Report on Embankments". The scope of the testing program and typical test results are presented below. Additional preconstruction test results are presented in Appendix 2.5D.TEST PROGRAM a) General Tests were conducted on reconstituted specimens of composite materials M and Z, and undisturbed samples of foundation soils from the auxiliary dam area. Sampling locations are are described in Section 2.5.1.2.5, and index properties of these materials are given in Section 2.5C.1 of this Appendix.b) Scope The test program consisted of the following:

1) Static Tests a) index properties Amendment 65 Page 273 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 b) isotropically consolidated drained (CID) triaxial tests c) unconsolidated undrained (UU) triaxial tests

2) Dynamic Tests a) stress-controlled cyclic triaxial tests b) strain-controlled cyclic triaxial tests c) cyclic torsion tests c) Preparation of Specimens Material M - Static tests were performed on specimens compacted to 100 percent standard compaction at optimum water content (d = 121.8 lb./ft.3, = 12.4%). Dynamic tests were performed on specimens compacted at 100 percent standard compaction at optimum water content. A few dynamic tests were also conducted on specimens compacted at 95 percent standard compaction at optimum water content.

Material Z - Static tests were conducted on specimens compacted to 97 percent standard compaction at optimum water content (d = 11.6 lb./ft.3, = 15.5%). Dynamic tests were performed on specimens molded at 97 percent standard compaction and optimum water content. A few dynamic tests were also performed on specimens compacted at 95 percent and 100 percent standard compaction at optimum water content.d) Foundation Soils in the Auxiliary Dam Area Stress-controlled cyclic triaxial tests were conducted on undisturbed block and tube samples of foundation soils from the auxiliary dam area.TEST RESULTS a) Main Dam - Material M

1) Static Triaxial Tests -Results of static CID tests are summarized on pages 2.5C.2-14 and 2.5C.2-15. Results of static UU tests are presented on pages 2.5C.2-16 and 2.5C.2-17. These results are being utilized to determine parameters for static finite element analysis, namely, the modulus number K, modulus exponent n, and Poisson's ratio parameters G, F, and D (see Table 2.5C.2-1).

The incremental finite element static stress analysis is based on nonlinear, stress-dependent stress-strain behavior of soils. The following relationships are used to define the inelastic behavior.

1) Primary Loading Tangent modulus Et for primary loading (Kondner) (References 2.5C.2-1 and 2.5C.2-2).

Amendment 65 Page 274 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 1 sin 12 cos 2 sin where c and are the Mohr-Coulomb shear strength parameters, Ei is the initial tangent modulus, and Rf is the failure ratio or ratio between the compressive strength (1 - 3)f and the value of the asymptotic stress difference for the hyperbolic stress-strain curve (1 - 3)ult.The variation of the initial tangent modulus value with confining pressure is represented by an empirical equation (Janbu) (Reference 2.5C.2-3).in which the modulus number K and exponent n are both pure numbers and Pa is the value of atmospheric pressure expressed in appropriate units.

2) Volume Changes Kulhawy, Duncan, and Seed (Reference 2.5C.2-4) have developed a procedure for incorporating in the stress analysis the volume change characteristics of soil in terms of tangent Poisson's ratio. The initial Poisson's ratio may be expressed by:

The value of tangent Poisson's ratio may be expressed by 11 sin 12 cos 2 sin where G, F, and D are parameters.

2) Dynamic Tests (a) Stress controlled Cyclic Triaxial Tests - A total of 31 tests were conducted on specimens compacted at 100 percent standard compaction and 10 tests on specimens compacted at 95 percent standard compaction. The tests were performed by using effective confining pressures (3c) of 2000, 4000, 8000, and 12000 lb./ft.2 and initial consolidation ratios Kc = 1, 1.5, and 2. The specimens were subjected to different ratios of cyclic deviator stress to effective confining pressures ( ) and the number of cycles (N) causing five percent and ten percent strains and the point where the pore pressure becomes equal to the cell pressure were determined. The results of these tests are summarized in Table 2.5C.2-2. Pages 2.5C.2-18 through 2.5C.2-20 show the relationship between Amendment 65 Page 275 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the stress ratio(d/23c) and the number of cycles causing five percent strain for Kc = 1, 1.5, and 2, respectively.(b) Strain-controlled Cyclic Triaxial Tests - Four strain-controlled cyclic triaxial tests were performed on specimens compacted at 100 percent standard compaction. Results of these tests together with those from cyclic torsion tests were used for determining the dynamic shear modulus and damping for the material (see Table 2.5C.2-3).(c) Cyclic Torsion Tests - Cyclic torsion tests have been performed using confining pressures of 500, 1000, 2000, and 4000 lb./ft.2. As noted above, results of these tests have been utilized for determining the modulus and damping characteristics of the material (see Table 2.5C.2-4).b) Material Z

1) Static Tests -Results of the static CID and UU tests on specimens compacted to 97 percent standard compaction are presented on pages 2.5C.2-21 through 2.5C.2-24.

Parameters derived from the CID tests for use in static finite element analysis are given on Table 2.5C.2-5.

2) Dynamic Tests (a) Stress-controlled cyclic triaxial tests - A total of 24 cyclic triaxial tests were conducted on specimens compacted at 95 percent, 97 percent, and 100 percent standard compaction. Confining pressure of 1250, 2500, and 5000 lb./ft.2 were used and initial consolidation was carried out for Kc = 1, 1.5, and 2. As in the case of material M, the number of cycles causing five percent and ten percent strain and the point where the pore pressure is equal to the cell pressure were determined. Results of tests on material compacted at 97 percent are summarized on pages 2.5C.2-25 through 2.5C.2-27 and Table 2.5C.2-6.

(b) Strain-controlled cyclic triaxial tests -Four strain-controlled cyclic triaxial tests were performed. Results of these tests are presented on Table 2.5C.2-7. Data obtained from these tests together with those from the cyclic torsion tests are being utilized for determining the shear modulus and damping characteristics of the material.(c) Cyclic torsion tests - These series of cyclic torsion tests have been completed.Results of these tests are presented on Table 2.5C.2-8.c) Foundation Soils in the Auxiliary Dam Area Fourteen stress-controlled cyclic triaxial tests were conducted on undisturbed samples of foundation soils from the auxiliary dam area. Results of these tests are given on Table 2.5C.2-9.d) Seismic Wave Velocity Measurements at Auxiliary Dam and Auxiliary Dike

1) Introduction Amendment 65 Page 276 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Seismic wave velocity measurements were made at locations along the axes of the Auxiliary Dam and Auxiliary Reservoir Separating Dike during the period 5 to 8 February 1973. The main purpose of the measurements was to determine compression-wave (P-wave) velocities (Vp),shear-wave (S-wave) velocities (VS), and Rayleigh-wave (R-wave) velocities (VR)of in-situ residual soil; however, the seismic wave velocity of the transitional material and the upper portion of the weathered rock were also determined. Seismic wave velocities of weathered and fractured rock and sound bedrock are shown in Section 2.5.2. Methods used to measure seismic wave velocities are discussed below.

2) Methods Used to Measure Seismic Wave Velocities Two methods were used to measure seismic wave velocities: (1) pulse arrival measurements of compression-wave (P-wave) velocity (Vp) and shear-wave (S-wave) velocity (VS); and (2) steady-state vibration measurements of Rayleigh-wave (R-wave) velocities (VR).

Pulse arrival measurements were made using a Sprengnether VS 1200 seismograph and a three-component geophone; a sledge hammer impact was used as the energy source; and Electrotech vertical geophone, located adjacent to the impact station, was used to provide zero time.Pulse arrivals were recorded for both vertical and horizontal impacts; several records were made at each location to examine repeatability of each measurement. The three-component geophone records the propagated seismic waves in three planes at right angles.P-wave pulse arrivals are measured by the horizontal component of the geophone that is oriented along the line between the geophone and the impact station. S-wave pulse arrivals are measured by the two components of the geophone oriented perpendicular to the line between the geophone and impact station.To create a maximum of S-wave energy, horizontal impacts are oriented perpendicular to the direction of the measurement line; in addition, this minimizes the P-wave energy. By reversing the impact direction, the S-wave is reversed; and by comparing the two records which are symmetrical with respect to the time axis, accuracy of the interpretation is increased.The impact-to-receiver distance is measured to an accuracy of +/- 0.1 ft.; and the time of the first pulse arrival is scaled from the records to an accuracy +/- 1 millisecond (msec.). Velocities are calculated by dividing the impact-to-receiver distance by the time.Steady-state vibration measurements were made using a Heathkit audio generator (1 G-72), a Dyna Kit Mark III preamplifier, and a Goodman vibrator to generate R-waves. The velocities of the R-waves were measured using two Electrotech EV-17 vertical geophones and their response observed on a Tectronix R-5030 dual beam oscilloscope.R-wave velocities were measured by determining the frequency required to create in-phase response of two geophones spaced at a selected distance. The frequency is varied in increments of one hertz by the audio generator. Geophone response is displayed on the oscilloscope screen, and in-phase response is determined to an accuracy +/- 0.5 hz for frequency ranges of approximately 30 hz to 110 hz, and +/- 5 hz above frequencies 110 hz. The distance d between geophones is measured to an accuracy of +/- 0.1 ft.Amendment 65 Page 277 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The frequency of the vibrator is varied until in-phase response is obtained. At that frequency f, the distance d is a multiple of the wave length. The frequency is then progressively increased to f which corresponds to the next higher in-phase response. The R-wave velocity is equal to Vr =d (f' - f).

3) Description of Materials (a) Residual Soil - Residual soil includes materials which range from silty clay to silt to silty sand. As the residual soil resulted from differential weathering of the underlying bedrock, the gradation from soil to weathered rock is not clearly defined. Within the residual soil layer, there are 2-to 3-ft. thick layers of rock weathered to different degrees which are referred to as transitional material.

(b) Weathered Rock - Weathered rock includes medium hard to hard sandy to clayey siltstone, and medium to fine-grained silty sandstone. The classification used to define the weathered rock, for purposes of these seismic-wave velocity measurements, is that material which, when excavated with extreme difficulty with a backhoe, breaks down to medium hard rock fragments.

4) Results of Seismic-Wave Velocity Measurements (a) Auxiliary Dam - Seismic velocity measurements at the site of the auxiliary dam were made at the location shown on page 2.5C.2-28.

P-wave and S-wave velocity measurements were made for impact receiver spacings of: (1) 15 ft. and 20 ft. in the residual soil layer; (2) 10 ft. and 20 ft., and 12.7 ft. and 22 ft. in the transitional material; (3) 12 ft. and 26 ft. in the weathered sandstone. Measurement locations were selected based on their representative nature; the average materials. Results of measurements are presented on Tables 2.5C.2-9 through 2.5C.2-12. Range of measurements is very narrow, indicating an excellent accuracy. For the residual soil and transitional materials, we consider these measurements are representative of the material at a depth of one-half the impact-receiver spacing for P-wave and S-wave measurements in the residual soil, and a depth of one half the wave length for R-wave measurements. For the weathered rock, we consider that the measurements are representative of the material at a depth determined as above or at a depth equal to the thickness of the layer, whichever is smaller.The average P-wave and S-wave velocities of the residual soil are:Average Velocities, ft./sec.Impact-Receiver Spacing, ft. Vp Vs 15 1505 750 20 1420 705 R-wave velocity measurements for the residual soil are:Amendment 65 Page 278 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Wave Average Velocity, Corresponding VS Length,ft. ft./sec. Vs= Vr x 1.1, ft./sec.10.2 620 680 6.1 605 665 The average P-wave and S-wave velocities of the transitional material are:Average Velocities, ft./sec.Impact-Receiver Spacing, ft. Vp Vs 10 1785 1000 20 2385 1335 22 3060 1380 12.7 2395 1065 The average P-wave and S-wave velocities of the weathered rock are:Average Velocities, ft./sec.Impact-Receiver Vp Vs Spacing, ft.12 2400 1335 26 3250 1445 The average R-wave velocity measurements for the weathered rock are 1315 ft./sec. and 1570 ft./sec. The corresponding S-wave velocities for the weathered rock are:VS = 1315 x 1.1 = 1445 ft./sec.and VS = 1570 x 1.1 = 1725 ft./sec.(b) Auxiliary Reservoir Separating Dike - Seismic velocity measurements at the site of the Auxiliary Reservoir Separating Dike were made at the location shown on page 2.5C.2-29.P-wave and S-wave velocity measurements were made for impact-receiver spacings of 25 ft. for residual soil and 20 ft. for weathered rock. Measurement locations were selected based on their representative nature; the average values measured are believed to apply to the average materials. Average velocities for residual soil are VP = 1300 ft./sec. and VS = 715 ft./sec.average velocities for weathered rock are VP = 3590 ft./sec. and VS = 2055 ft./sec. Results of measurements are presented on Table 2.5.C.2-13. We consider these measurements representative of material at a depth of approximately one-half the impact-receiver spacing.CALCULATION OF K2,max ON THE BASIS OF SHEAR WAVE VELOCITIES Amendment 65 Page 279 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The equation relating K2,max to the shear wave velocity VS is 10 , 1/2 where VS = shear weight velocity, ft./sec.

 = unit weight of material, lb./ft.3 g = acceleration of gravity = 32.2 ft./sec.2 = mean effective pressure, lb./ft.2 The unit weights of the materials are selected as follows:

Residual Soil = 135 lb./ft.3 Transitional Material = 142.5 lb./ft.3 Weathered Rock = 150 lb./ft.3 The mean effective pressure is given by the equation:where v = vertical effective stress h = horizontal effective stress Selecting a horizontal to vertical effective stress ratio K = 0.6, the mean effective pressure is given by the equation:0.733 where d is the depth below ground surface

 / /

For residual soil = 0.733 x 135 d and = 9.95

 / /

For transitional material = 0.733 x 142.5 d and = 10.2

 / /

For weathered rock = 0.733 x 150 d and = 10.5 K2, max is given by the equations below:

 . /

For the residual soil K2,max = 10 4.22 10 /Amendment 65 Page 280 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2

 . /

For transitional material K2,max = 10 4.34 10 /

 . /

For weathered rock K2,max = 10 4.44 10 /The values of K2,max are given on Table 2.5C.2-14 for the shear wave velocities measured at the sites of the Auxiliary Dam and Auxiliary Reservoir Separating Dike.e) Material Properties Used For Dynamic Analyses of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike

1) Introduction This section presents the values of material properties which are used for the dynamic analyses of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike. These values are expected for the materials in the constructed dams and dike. Values for the Main Dam are presented on Table 2.5C.2-15 and those for the Auxiliary Dam and Auxiliary Reservoir Separating Dike are presented on Table 2.5C.2-16.

2) Basis for Selection of Material Properties for Dynamic Analyses of Main Dam a) Unit Weights (1) Core (Material M) - Based on laboratory tests.

(2) Fine Filter and Coarse Filter -Based on specified gradations, expected constructed relative density of 80 percent and several published (e.g., Burmister 1962) (Reference 2.5C.2-5) and unpublished (e.g., WMAI 1973) (Reference 2.5C.2-

13) data regarding unit weight of granular material.

(3) Rockfill Shell -Based on in-place unit weight measurements of a rockfill at Keban Dam, Turkey (Ebasco 1972) (Reference 2.5C.2-6) with properties similar to those of the proposed rockfill.(4) Weathered Rock - Based on laboratory tests.b) Ratio of Horizontal Effective Stress to Vertical Effective Stress Ko Values of ratios of horizontal effective stresses to vertical effective stresses of compacted and preconsolidated materials are available in several publications (e.g.,D'Appolonia et. al., 1969, Lacroix and Horn 1973) (References 2.5C.2-7 and 2.5C.2-8). Relatively high values are selected to account for compaction of the embankment materials and preconsolidation in the weathered rock.c) Poisson's Ratio Typical values of Poisson's ratio are available in several publications (e.g., Leonards 1962, Barkan 1962) (References 2.5C.2-9 and 2.5C.2-10).Amendment 65 Page 281 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 d) Shear Modulus Parameter K2,max (1) Core (Material M) - Based on cyclic torsion tests and cyclic triaxial tests. Tables 2.5C.2-17 and 2.5C.2-18, and pages 2.5C.2-30 and 2.5C.2-31 show the test results.(2) Fine Filter - Based on published data (Seed and Idriss 1970) (Reference 2.5C.2-11). Fine filter will be sand with expected constructed relative density of 80 percent; see page 2.5C-72.(3) Coarse Filter - Based on published data (Seed and Idriss 1970, Wong 1970)(References 2.5C.2-11 and 2.5C.2-12). Coarse filter will be gravel with sand having an expected constructed relative density of 80 percent.(4) Rockfill Shell - Because the maximum shear modulus of gravel is higher than that of sand, a maximum shear modulus for rockfill higher than that of gravel is selected by extrapolation of available data; see page 2.5C.2-32.(5) Weathered Rock - Based on seismic compression-wave velocity measurements in the weathered rock along the main dam axis; see page 2.5C.2-32.e) Damping Ratio (1) Core (Material M) - Based on cyclic torsion tests and cyclic triaxial tests. Tables 2.5C.2-17 through 2.5C.2-19 and page 2.5C.2-33 show the test results.(2) Fine Filter - Fine filter will be sand. Average damping ratio for sands (Seed and Idriss 1970) (Reference 2.5C.2-11) is used; see page 2.5C.2-34.(3) Coarse Filter - Coarse filter will be gravel with sand. Published data (Wong 1970) (Reference 2.5C.2-12) indicate that damping ratios of gravels and sands are similar. Average damping ratio for sands is used; see page 2.5C.2-34.(4) Rockfill Shell and Weathered Rock - Average damping ratio for sands is used; see page 2.5C.2-34.

3) Basis for Selection of Material Properties for Dynamic Analyses of Auxiliary Dam and Auxiliary Reservoir Separating Dike a) Unit Weights (1) Core (Material Z) - Based on laboratory tests.

(2) Filter - Based on specified gradation, expected constructed relative density of 80 percent, and several published (e.g., Burmister 1962) (Reference 2.5C.2-5) and unpublished (e.g., WMAI 1973) (Reference 2.5C.2-13) data regarding unit weight of granular material.(3) Random Rockfill -Based on in place unit weight measurements of a rockfill at Amos Dam, West Virginia (WMAI 1973) (Reference 2.5C.2-3), with properties Amendment 65 Page 282 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 similar to those of the proposed rockfill. Unit weight of siltstone and sandstone random rockfill is selected slightly smaller than that of the granitic rockfill of the Main Dam.(4) In-situ Soils - Based on laboratory tests.(5) Weathered Rock - Based on laboratory tests.b) Ratio of Horizontal Effective Stress to Vertical Effective Stress Ko Values of horizontal effective stress to vertical effective stress of compacted and preconsolidated materials are available in several publications (D'Appolonia et. al.1969, Lacroix and Horn 1973) (References 2.5C.2-7 and 2.5C.2-8). Relatively high values are selected to account for compaction of the embankment materials and preconsolidation in the weathered rock.c) Poisson's Ratio Typical values of Poisson's ratio are available in several publications (e.g., Leonards 1962, Barkan 1962) (References 2.5C.2-9 and 2.5C.2-10).d) Shear Modulus Parameter K2,max (1) Core (Material Z) - Based on cyclic torsion tests and cyclic triaxial tests. Tables 2.5C.2-20 and 2.5C.2-21, and pages 2.5C.2-35 and 2.5C.2-36 show the test results.(2) Filter -Based on published data (Seed and Idriss 1970, Wong 1970) (References 2.5.2-11 and 2.5C.2-12). Filter will be a gravelly sand with expected constructed relative density of 80 percent; see page 2.5C.2-32.(3) Random Rockfill -Based on seismic wave velocities measured at Amos Dam, West Virginia (WMMAI 1973) (Reference 2.5C.2-13), with properties similar to those of the proposed random rockfill. At Amos Dam, P-wave, S-wave, and R-wave velocities of the rockfill lead to K2,max of approximately 100; the value selected for the Auxiliary Dam and Auxiliary Reservoir Separating Dike random rockfills reflects the higher compaction specified for these random rockfills than for the Amos Dam rockfill; see page 2.5C.2-32.(4) In-situ Soil - Based on P-wave, S-wave, and R-wave velocities of the in-situ soil at several locations along the axes of the Auxiliary Dam and the Auxiliary Separating Reservoir Dike; see page 2.5C.2-32.(5) Weathered Rock - Based on the seismic velocities in the weathered rock along the axis of the Auxiliary Dam; see page 2.5C.2-32.e) Damping Ratio (1) Core (Material Z) - Based on laboratory torsion test and cyclic triaxial test data.Tables 2.5C.2-20 through 2.5C.2-22, and page 2.5C.2-37 show the test results.Amendment 65 Page 283 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 (2) Filter -Filter will be gravelly sand. Published data (Wong 1970) (Reference 2.5C.2-12) indicate that damping ratios of gravels and sands are similar.Average damping ratio for sands is used; see page 2.5C.2-34.(3) Random Rockfill, In-situ Soils, and Weathered Rock - Average damping ratio for sands is used; see page 2.5C.2-34.(4) Parametric studies -In the case of the core materials, the range of parametric variations was selected on the basis of laboratory test results. In the case of the filters, the range was selected on the basis of published data (Seed and Idress 1970, Wong 1970) (References 2.5C.2-11 and 2.5C.2-12). In the case of rockfill materials, the range was selected on the basis of material index properties. In the case of weathered rock the range was selected on the basis of seismic velocity measurements. For all materials the selection was made taking into account design and construction criteria. Table 2.5C.2-23 through 2.5C.2-26 show the material-property combinations used for the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike respectively.CONCLUSION Through the preconstruction stage of the SHNPP project, these results formed the basis for defining the static and dynamic properties of the respective materials and were used for the evaluation of seismic stability of the Main Dam, Auxiliary Dam, and the Auxiliary Reservoir Separating Dike.Static and dynamic properties utilized in the stability analyses are further discussed in Appendix 2.5D, together with the results of the analyses.Changes in borrow areas, and material properties criteria which became necessary after construction began are discussed in Appendix 2.5F, "Main and Auxiliary Dams and Auxiliary Reservoir Separating Dike Embankment Reports."2.5C.3 RESULTS OF TESTS ON SAMPLES FROM BORINGS IN BORROW AREAS The tests performed on samples from Borrow Areas Y and Z and Main Dam Borrow Area M are indicated on the matrices shown on pages 2.5C.3-2 through 2.5C.3-5. Pages 2.5C.3 6 through 2.5C.3 223 give the results of these tests. Tables 2.5C.3-1, 2 and 3 show the shrinkage factors for samples from Borrow Areas Y, Z and Main Dam Borrow Area M, respectively.Amendment 65 Page 284 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BORROW AREA Y - TESTING PROGRAM Boring Sample Grain Size Triaxial Shear Proctor Number Number Analysis Test Compaction Test BB101 S-1

• S-2
• S-6
• BB103 S-1
• S-2 * *
• S-4
• BB105 S-1 * *
• S-2
• BB107 S-1
• S-2
• BB109 S-1
• S-4
• BB110 S-2 * *
• BB111 S-1 * *
• BB113 S-1
• S-2 * *
• S-3 * *
• BB114 S-1 * *
• S-2
• BB116 S-1
• S-2
• S-4
• BB117 S-1 * *
• S-2 * *
• S-3 *
• S-4 * *
• BB119 S-1
• S-2 * *
• BB121 S-1
• S-2
• S-4
• BB124 S-1 *
• S-2 *
• S-4 *
• BB125 S-1 * *
• S-2 * *
• Amendment 65 Page 285 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 BORROW AREA Z - TESTING PROGRAM Boring Sample Grain Size Triaxial Shear Proctor Number Number Analysis Test Compaction Test BB156 S-1 * *

• BB156 S-2
• BB158 S-1 * *
• BB158 S-2
• BB159 S-1 *
• BB159 S-2 *
• BB159 S-3 * *
• BB159 S-4 *
• BB161 S-1
• BB161 S-4 * *
• BB163 S-2 * *
• BB166 S-1
• BB166 S-2
• BB167 S-1
• BB171 S-1 *
• BB171 S-2 *
• BB176 S-1 *
• BB176 S-2 *
• BB176 S-3 * *
• BB176 S-4 * *
• BB176 S-5 *
• BB177 S-1
• BB177 S-2
• BB178 S-1 BB178 S-2
• BB178 S-4
• BB180 S-1 * *
• BB180 S-2
• BB184 S-1
• BB184 S-2
• BB186 S-1
• BB186 S-2
• Amendment 65 Page 286 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 MAIN DAM BORROW AREA TESTING PROGRAM Boring Sample Grain Size Triaxial Proctor Number Number Analysis Shear Test Compaction Test BB806 S-1

• S-2 *
• S-3 *
• BB807 S-1 *
• S-2 * *
• S-4 * *
• BB812 S-3 *
• BB813 S-1 * *
• BB814 S-1 *
• S-2 *
• S-3 *
• S-4 *
• BB815 S-1 *
• S-3 & S-4 *
• BB863 S-1
• S-2
• BB864 S-1
• S-2 * *
• BB865 S-1 * *
• BB866 S-1 * *
• S-2
• S-4
• BB867 S-1
• BB868 S-1
• S-2
• BB869 S-1
• S-2
• S-3 * * *

REFERENCES:

APPENDIX 2.5C 2.5C.2-1 "Hyperbolic Stress-Strain Response-Cohesive Soils," R. L. Kondner, Journal of Soil Mechanics and Foundations Division, ASCE, Vol. 89, No. SM1, 1963.2.5C.2-2 "Hyperbolic Stress-Strain Formulation for Sands," R. L. Kondner, and J. S.Zelasko, Proceedings of Second Pan American Conference on Soil Mechanics and Foundation Engineering, Vol. 1, 1963.2.5C.2-3 "Soil Compressibility as Determined by Oedometer and Triaxial Tests," N. Janbu, European Conference on Soil Mechanics and Foundation Engineering at Wiesbaden, West Germany, Vol. 1, 1963.2.5C.2-4 "Finite Element Analyses of Stresses and Movements in Embankments During Construction," F. H. Kulhawy, J. M. Duncan and H. B. Seed, Report No. TE 69-4, Office of Research Services, University of California, Berkeley, 1969.Amendment 65 Page 287 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5C.2-5 "Physical, Stress-Strain, and Strength Responses of Granular Soils," D. M.Burmister, ASTM Spec. Tech. Pub. No. 322, 1962.2.5C.2-6 Ebasco, personal communication, 1972.2.5C.2-7 "Sand Compaction with Vibratory Rollers," D. J. D'Appolonia, R. V. Whitman, E.D'Appolonia, JSMFD, SM1, Paper 6366, January 1969.2.5C.2-8 "Direct Determination and Indirect Evaluation of Relative Density and Its Use on Earthwork Construction Projects," Yves Lacroix and H. M. Horn, Symp. on Relative Density of Cohesionless Soils, ASTM Spec. Tech. Pub. (in print), 1973.2.5C.2-9 "Dynamics of Bases and Foundations," D. D. Barkan, McGraw-Hill, New York, 1962 2.5C.2-10 "Foundation Engineering," G. A. Leonards, McGraw-Hill, New York, 1962.2.5C.2-11 "Soil Moduli and Damping Factors for Dynamic Response Analyses," H. Bolton Seed and I. M. Idriss, Eqk. Engr. Res. Ctr. Report EERC 70-10, Univ. of Calif.,Berkeley, December, 1970.2.5C.2-12 "Deformation Characteristics of Gravels and Gravelly Soils Under Cyclic Loading Conditions," R. T. Wong, Ph. D. Thesis, Univ. of Calif., Berkeley, 1970.2.5C.2-13 WMAI, file, Woodward-Moorhouse & Associates, Inc., 1973.APPENDIX 2.5D SEISMIC STABILITY ANALYSIS OF SEISMIC CATEGORY I DAMS AND DIKE 2.5D.0 PURPOSE AND CONCLUSION 2.5D.0.1 Introduction Appendix 2.5D describes the seismic stability analyses of the Seismic Category I dams and dike. The first seismic analysis of the SHNPP reservoir system was for the original cooling system, which included a 10,000-acre lake and afterbay; this analysis is described in the following paragraphs of Section 2.5D.0 and Sections 2.5D.1 through 2.5D.8. After the SHNPP reservoir system was redesigned for cooling tower operation, the Main Reservoir was reduced in size from 10,000 acres to approximately 4,000 acres (from NWL Elevation 250 ft. to NWL Elevation 220 ft.) and the Afterbay Reservoir was deleted. The seismic stability analysis of the reservoir following these changes is contained in Section 2.5D.9.Appendix 2.5D describes the seismic stability analyses of three loading conditions. Loading Condition I corresponds to normal operating conditions and the occurrence of the SSE for the Elevation 250 ft. lake. Condition II involves low water conditions and the occurrence for the OBE for the Elevation 250 ft. lake. Condition III corresponds to normal operating conditions and the occurrence of the SSE for the Elevation 220 ft. lake (the present design).The material properties and engineering properties used in the analysis of all three conditions are discussed in Sections 2.5D.10 through 2.5D.16. However, as a result of further Amendment 65 Page 288 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 investigation and data developed during construction, Borrow Area M (described in Section 2.5D.10) was not used. Instead, Borrow Area W was used for the Main Dam impervious core.A description of this material and a justification for its use is given in Appendix 2.5F.Also, there was not sufficient material for the Main Dam rockfill in the excavations for the spillway, spillway outlet, and spillway approach channel (described in Section 2.5D.12);additional Main Dam rockfill was obtained from a quarry located upstream of the dam, as described in Appendix 2.5F.The present dimensions of the dam and dike and the discussion of construction of the embankments are contained in Appendix 2.5F. The report on foundations and grouting is in Appendix 2.5E. Preconstruction investigations are described in Appendices 2.5A, 2.5B, 2.5C, 2.5G, and 2.5H. Specifications for the dams and dikes are contained in Appendix 2.5I.A location map showing the plant area, the Main Reservoir, the Main Dam, the Auxiliary Reservoir, the Auxiliary Dam, and the Auxiliary Separating Dike is presented in Figure 2.5D-1.2.5D.0.2 Category I Dams: Main Dam, Auxiliary Dam, and Auxiliary Separating Dike The Main Dam, Auxiliary Dam, and Auxiliary Separating Dike, which have heights generally smaller than one hundred feet, are normal earth and rockfill dams. The design of the Category I dams is in accord with the general practices of government agencies, such as the Army Corps of Engineers, the United States Bureau of Reclamation, and state departments of conservation.There are several aspects of the design and construction criteria of these Category I dams which make these dams much more conservative than most existing dams.a) The outside slopes of the dams are generally flatter than those commonly used. Dams higher than the Main Dam with similar rockfill shells of granitic rock often have outside slopes of 1.5 horizontal to 1 vertical (1.5:1). The Main Dam outside slopes are 2.0:1.Dams higher than the Auxiliary Dam and Auxiliary Reservoir Separating Dike with similar random rockfill shells built of compacted sedimentary rock often have outside slopes between 1.5:1 and 2.0:1. The Auxiliary Dam and Auxiliary Reservoir Separating Dike outside slopes are 2.5:1.b) The rockfill shells of the Category I dams are placed in layers and compacted, whereas common practice has been to construct rockfill by dumping with or without sluicing and no compaction. Compaction of the rockfill results in a material with greater strength and lower compressibility.c) At the site of the Category I dams, the foundation materials are excellent; they consist of rock or stiff residual soil. Unsatisfactory alluvial material has been removed.d) Most importantly, the construction of the dam is supervised and inspected by competent personnel. Inspection is an essential part of earth and rockfill dam construction. The purpose of the inspection is to assure that: (a) on the basis of observation and measurements of the soil and rock materials encountered during construction, the design criteria remain valid; and (b) compliance with the construction criteria is achieved.Amendment 65 Page 289 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.0.3 Conventional Static and Dynamic Stability Analyses of the Category I Dams Static and dynamic stability analyses of the Category I dams were made by Ebasco using conventional methods; see Section 2.5.6.5. The static stability of the Category I dams was investigated by using conservative static properties of the materials forming the dams and the foundation materials. Results obtained indicate that the static stability of the dams is ample.The dynamic stability of the dams was investigated using conventional methods. Results obtained indicate that the dynamic stability of the dams is ample during the SSE.2.5D.0.4 Finite Element Seismic Stability Analyses for the Category I Dams 2.5D.0.4.1 Procedure Because of the importance of the Category I dams, a finite element analysis was made to evaluate the seismic stability of the dams. The procedure which has been in development since the early 1960's has provided reasonable estimates of field behavior in a number of cases where dams have been subjected to strong earthquakes. The procedure has been used for some years for evaluating the seismic stability of existing dams and for the design of proposed dams. The procedure is described in Section 2.5D.3.2.5D.0.4.2 Design Criteria and Basic Data 2.5D.0.4.2.1 Loading Conditions The seismic stability analysis was made using three conditions of loading. Condition I corresponds to normal operating conditions and the occurrence of the safe shutdown earthquake (SSE). Condition II corresponds to other conditions and the occurrence of the operating basis earthquake (OBE). Condition III corresponds to a lower normal water surface elevation than Condition I and the occurrence of the SSE. Sections 2.5D.1 through 2.5D.8 present the analysis of seismic stability of Category I dams for Condition I and II. Section 2.5D.9 presents analysis of seismic stability of Category I dams for Condition III. Condition I and II are defined in Section 2.5D.4.5, and Condition III is defined in Section 2.5D.9.4.2.5D.0.4.2.2 Design Basis Earthquake The SSE originally used for the seismic stability analysis of the Category I dams and dike is shown on Figure 2.5D-7. The accelerogram selected for the analysis is an artificial accelerogram which represents a series of earthquakes applicable to the site and which has response spectra closely enveloping the smooth response spectra which define that SSE.In addition, the behavior of the dams and dike in response to an event prescribed by the Regulatory Guide 1.60 spectra was assessed and the details are given in Section 2.5D.18.2.5D.0.4.2.3 Material Properties The properties of the soil and rock materials which form the dams and the properties of the foundation materials were selected on the basis of laboratory test results, field measurements, and published and unpublished data. The material properties used in the basic set of analysis correspond to those which were expected to be obtained in the constructed dams. To account Amendment 65 Page 290 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 for possible variations, these material properties were varied within a conservative range and the analyses were repeated for a conservative combination of the material properties.2.5D.0.4.2.4 Geometry of the Dams and Compaction Criteria of the Materials The geometry of the cross sections of the dams is shown in Figures 2.5D-2 through 2.5D-6.The core, filters, and rockfill materials are placed in layers and compacted according to specified construction criteria.2.5D.0.4.2.4.1 Main Dam The core of the main dam is compacted to an average degree of standard compaction* of 100 percent at a water content within plus or minus two percent of optimum. The fine and coarse filters of the main dam are compacted to an average relative density** of 75 percent, except for the upstream coarse filter which is compacted to an average relative density of 80 percent above El. 220.2.5D.0.4.2.4.2 Auxiliary Dam The core of the auxiliary dam is compacted to an average degree of standard compaction of 97 percent at a water content within plus or minus two percent of optimum. The filters of the auxiliary dam are compacted to an average relative density of 75 percent below El. 220 and to an average relative density of 80 percent above El. 220.2.5D.0.4.2.4.3 Auxiliary Separating Dike The core of the auxiliary Separating dike is compacted to an average degree of standard compaction of 97 percent below El. 220 and 100 percent above El. 220.2.5D.0.5 Evaluation of the Seismic Stability of the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike 2.5D.0.5.1 Method The seismic stability of the dams was evaluated by comparing the shear stresses, f, required to cause 5 x 10-2 strain at any location within the dam to the shear stresses, d, induced by the SSE. The ratio, f/d, has been considered to represent a local factor of safety against the development of 5 x 10-2 strain. On the basis of correlations between the results of seismic stability evaluations by this procedure and the performance of dams which have been subjected to significant earthquake loading, it has been stated that a minimum value of the stress ratio, f/d, greater than approximately 1.1 provides an ample margin of safety for the dams.ASTM D 698-68T, Method A, B, C, or D Modified from ASTM D2-49-69 Amendment 65 Page 291 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.0.5.2 Results 2.5D.0.5.2.1 Main Dam For the expected constructed material properties, the minimum local factors of safety are approximately 1.8 in the core, 1.7 in the fine filters, 1.2 in the coarse filters, and 1.5 in the rockfill shell; see Section 2.5D.5.4.7, 2.5D.5.4.8, and 2.5D.5.5. The results due to an event prescribed by the Regulatory Guide 1.60 are approximately 1.7 in the core, 1.7 in the fine filters, 1.2 in the coarse filters, and 1.5 in the rockfill shell; see Section 2.5D.18. We conclude that the Main Dam will be stable and will maintain its integrity during the SSE.2.5D.0.5.2.2 Auxiliary Dam For the expected constructed material properties, the minimum factors of safety are approximately 1.4 in the core, 1.3 in the filter, 1.5 in the random rockfill shell, and 1.5 in the in-situ residual soil; see Section 2.5D.6.4.6, 2.5D.6.5.4, and 2.5D.6.6. The results due to an event prescribed by the Regulatory Guide 1.60 spectra are equal to the above safety factors; see Section 2.5D.18. We conclude that the Auxiliary Dam will be stable and will maintain its integrity during the SSE.2.5D.0.5.2.3 Auxiliary Reservoir Separating Dike For the expected constructed material properties, the minimum factors of safety are approx. 1.7 in the core, 1.9 in the random rockfill shell, and 1.8 in the in-situ residual soil; see Section 2.5D.7.4. The results due to an event prescribed by the Regulatory Guide 1.60 spectra are equal to the above safety factors; see Section 2.5D.18. We conclude that the Auxiliary Separating Dike will be stable and will maintain its integrity during the SSE.2.5D.1 INTRODUCTION This Appendix presents the results of seismic finite element stability analyses of Category I dams for the Shearon Harris Nuclear Power Plant. These dams are referred to as the Main Dam, the Auxiliary Dam, and the Auxiliary Reservoir Separating Dike; their locations are shown in Figure 2.5D.1. The plant is built and operated by Carolina Power & Light Company at the Shearon Harris site in Wake County, North Carolina. Ebasco Services, New York, New York, is the architect engineer.The evaluation of seismic stability is based on dynamic finite element analyses of the dams using equivalent linear, strain-dependent material properties.This Appendix is divided into eighteen sections. The Category I dams analyzed in the appendix are described in Section 2.5D.2. The procedure used in the evaluation of seismic stability is outlined in Section 2.5D.3. Characteristics of the safe shutdown earthquake (SSE) and operating basis earthquake (OBE) and loading conditions I and II for which the Category I dams are analyzed in this Appendix are defined in Section 2.5D.4. Evaluation of seismic stability of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike during the SSE (loading conditions Case I) are discussed in Section 2.5D.5, 2.5D.6, and 2.5D.7, respectively.Evaluation of seismic stability of the dams during the OBE (loading condition case II) are discussed in Section 2.5D.8. Evaluation of seismic stability of the dams for Condition III are discussed in Section 2.5D.9.Amendment 65 Page 292 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Sections 2.5D.10, 2.5D.11, 2.5D.12, and 2.5D.13 present the static and dynamic material properties of the core, filter, and rockfill materials, and of the residual soil and weathered rock.The method of evaluation of static stresses in the dams is described in Section 2.5D.14. The values of the expected constructed material properties for each zone of these dams and the parametric variations from these values are presented in Section 2.5D.15.The procedure used for the evaluation of seismic stability of the dams is described in Section 2.5D.16. A comparison of response by various alternate lumped mass analytical models is presented in Section 2.5D.17. An assessment of the behavior of the dams to an event prescribed by the Regulatory Guide 1.60 spectra is presented in Section 2.5D.18.2.5D.2 DESCRIPTION OF CATEGORY I DAMS 2.5D.2.1 General The seismic stability of three Category I dams is analyzed in this appendix. They are the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike which are located as shown in Figure 2.5D.1.2.5D.2.2 Main Dam The Main Dam is approximately 1300 ft.long with a maximum height of approximately 105 ft.Three cross sections and the longitudinal section of the Main Dam are shown in Figure 2.5D-2 and 2.5D-3. The Main Dam has a core of compacted silty clayey sand protected by two 8-ft.-thick transition filter zones and a rockfill shell on each side. The core is founded on suitable rock and the rockfill shell is founded on weathered rock. (Suitable rock and weathered rock are defined in Section 2.5D.13.1).The rock at the foundation of the Main Dam consists of granitic gneiss; see Section 2.5D.13.2.5D.2.3 Auxiliary Dam The Auxiliary Dam is approximately 3903 ft. long with a maximum structural height of approximately 72 ft. Three cross sections and the longitudinal section of the Auxiliary Dam are shown in Figure 2.5D.4 and 2.5D.5. The Auxiliary Dam has a core of compacted silty clay protected by a transition filter zone and a random rockfill shell on each side. The core is founded on suitable rock. In the central portion for approximately 1400 ft. the random rockfill shell is founded on weathered rock and in the remaining portions, the random rockfill is founded on a thin layer of stiff residual soil overlying weathered rock.The rock at the foundation of the Auxiliary Dam consists of Triassic sandstones, siltstones, claystones, and conglomerates. Residual soil consisting of silty clay and sandy clayey silt occurs as a thin layer of stiff soil overlying weathered rock; see Section 2.5D.13.2.5D.2.4Auxiliary Separating Dike The Auxiliary Separating Dike is approximately 1100 ft. long with a maximum height of approximately 53 ft. The maximum cross section and longitudinal section of the Auxiliary Separating Dike are shown in Figure 2.5D-6. The dike has a core of compacted silty clay protected by a random rockfill shell which is graded near the core, with the finer material placed Amendment 65 Page 293 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 adjacent to the core and the coarser particles placed to the outside. The core and rockfill shell are founded on a thin layer of stiff residual soil overlying weathered rock.The rock and residual soil at the foundation of the Auxiliary Dam are the same as those at the foundation of the Auxiliary Reservoir Separating Dike; see Section 2.5D.13.2.5D.3 PROCEDURE USED IN SEISMIC STABILITY EVALUATION The procedure used in evaluating the seismic stability of the Category I dams consists of the following steps:a) Determination of the response of the dam-foundation system to the rock accelerations, including the evaluation of the induced shear stresses at various locations throughout the dam and the foundation material.b) Representation of the irregular cycles of shear stresses induced in the dam-foundation system by an equivalent number of uniform cycles of shear stresses.c) Determination of the static stresses existing in the dam-foundation system (prior to the rock accelerations).d) Determination of the cyclic shear stresses required to cause strains greater than 5 x 10-2 (see Section 2.5D.5 and Section 2.5D.16) in the material for conditions representative of those existing in the dam foundation system by means of appropriate cyclic load tests on representative specimens of the materials or by correlation with data for similar materials.e) Evaluation of the seismic stability of the dams by comparing the shear stress required to cause strain greater than 5 x 10-2 with the equivalent shear stresses induced by the rock accelerations.This procedure was recently developed (Reference 2.5D-27 and 2.5D-28) and has provided reasonable estimates of field behavior in a number of cases (Reference 2.5D-28 and 2.5D-35).The procedure also has been used for evaluating the seismic stability of several existing dams and for the design of several proposed dams in California and other locations (Reference 2.5D-38). Most recently, it was utilized in the Virgil C. Summer Nuclear Station Near Columbia, South Carolina.Step (1) in this procedure involves the determination of the response of the dam-foundation system to postulated base rock accelerations. These postulated base rock accelerations are represented by the artificial accelerogram described in Section 2.5D.4. The response computation is performed using the finite element method of analysis (Reference 2.5D-9 and 2.5D-11). A recently developed computer program (Reference 2.5D-14) has been used to compute the response, including induced shear stresses, for the dams at the site. This program incorporates the use of strain-dependent modulus and damping ratio for each element in the finite element representation of the dam-foundation system.Typical response, including induced shear stresses, step (1), and a brief illustration of the procedure for representing the induced shear stresses by an equivalent number of stress cycles, step (2), are presented in Section 2.5D.16.Amendment 65 Page 294 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Determination of the static stresses, step (3), is described in Section 2.5D.14. Evaluations of the cyclic strength characteristics, step (4), of the material comprising each zone of the dams are presented in Sections 2.5D.10, 2.5D.11, 2.5D.12, and 2.5D.13.An example of the evaluation of the seismic stability of the dams, step (5), is described in Section 2.5D.16; the results for the three dams are presented in Section 2.5D.5, 2.5D.6, and 2.5D.7.2.5D.4 SAFE SHUTDOWN EARTHQUAKE AND LOADING CONDITIONS 2.5D.4.1 Safe Shutdown Earthquake The safe shutdown earthquake (SSE) is evaluated and defined in FSAR Sections 2.5.2 and 3.7.For the dynamic analysis of the Category I dams, the same SSE is postulated. In addition the behavior of the dams and dike in response to an event prescribed by Regulatory Guide 1.60 spectra was assessed, see Section 2.5D.18.A value of 0.15 gravity is assigned as the maximum horizontal ground acceleration with the corresponding maximum vertical acceleration as 0.10 g. The Category I dams are designed to remain stable assuming that the horizontal and vertical accelerations act simultaneously.The SSE is postulated as being a shock of low magnitude occurring close to the site with a maximum duration of ten seconds. The pattern of motion should be similar to the Golden Gate (San Francisco) earthquake of 1957 or the Helena, Montana earthquake of 1935. Records from these earthquakes indicate two to three cycles of strong motion. The SSE is defined by the smooth response spectra presented on FIGURE 3.7.1-5. Using the same criteria, Dames &Moore prepared the smooth response spectrum corresponding to a structural damping ratio of 0.07 presented in Figure 2.5D-7.2.5D.4.2 Artificial Accelerogram The dynamic analysis requires the use of accelerograms. Two methods can be used to select the accelerograms. One method consists of scaling to the proper acceleration (ie, 0.15 g) one or several actual accelerograms selected from existing earthquake strong motion records.However, the response spectra calculated from actual accelerograms are occasionally below the smooth response spectra defining the SSE and, at some periods, may have valleys which are significantly below the smooth response spectra. This may be alleviated by using a series of actual accelerograms for the design to assure that there is no period, within the range of interest, for which the response obtained by at least one of the accelerograms of the series is significantly smaller than that indicated by the smooth response spectra. Because of the great amount of analytical work involved, another method was used to select the accelerograms for the design of the power station.The method used for design of all Category I systems and components, including dams, was to generate a single artificial accelerogram, the response spectra of which closely envelope the smooth response spectra which define the SSE. This artificial accelerogram applicable to a damping ratio of 0.07 is presented in Figure 2.5D-8. Its duration of approximately ten seconds is equal to that of the SSE. The response-spectrum of the artificial accelerogram for 0.07 structural damping ratio is presented in Figure 2.5D-7. The accelerogram was specifically fine-tuned, so that its response spectrum for 0.07 damping has a maximum positive deviation of Amendment 65 Page 295 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 seven percent above the smooth response spectrum in the range of periods of interest; ie, 0.185 sec. to 1 sec. To validate this accelerogram, a high resolution response spectral analysis was performed. The period points were spaced from 0.0025 sec. in the short period range (0.2 sec.) to 0.0125 sec. in the long period range (1 sec.). Similar artificial accelerograms were developed for damping ratios of 0.05 and 0.10, and 0.15.Preliminary dynamic stress analysis of the Category I dams indicated that the calculated damping ratios ranged from 0.07 to 0.15; see Section 2.5D.16. Therefore, use of the artificial accelerogram presented in Figure 2.5D-8 which is applicable to a damping ratio of 0.07 for calculating the stresses induced by the SSE in the Category I dams is conservative. The conservative nature of this selection is shown in Section 2.5D.18, within the assessment of the dams and dike, to an event prescribed by the Regulatory Guide 1.60 spectra.2.5D.4.3 Comparison Between the Artificial Accelerogram and Actual Accelerograms of Earthquakes Applicable to the Site Envelope response spectra have been specified for design of all Category I structures for the SHNPP. These response spectra do not represent those of the earthquake motions that might be generated by any single shock, but are, however, representative of the maximum responses from a possible series of earthquakes, any one of which could be the SSE. Thus, the artificial accelerogram generated from the specified site smooth response spectra has the characteristics of the maximum motions of the series of earthquakes. As previously noted, this artificial accelerogram was utilized in the evaluation of the stresses in the Category I dams.However, the same accelerogram cannot be used in the determination of the cyclic strength, including the stress-strain characteristics of the soil and rock materials forming the dams.Theoretical analysis confirmed by laboratory testing and case histories (eg, Alaska earthquake.April 1964; San Fernando Valley earthquake, February 1971) (Reference 2.5D-35) show that cyclic strength characteristics of soil and rock are considerably affected by the number of strong motion loading cycles. The cyclic strength of soil and rock significantly decreases as the number of cycles increases because of progressive buildup of pore pressure; therefore, a conservative, but realistic, number of loading cycles are used for the determination of the cyclic strength of soil and rock.It is expected that the actual accelerogram of any single earthquake applicable to this site representative of the SSE would not have a duration of strong motion exceeding two to three seconds and would not contain any more than four to six high peaks* corresponding to two to three cycles of strong motion, as noted in Section 2.5. For this representative SSE, we have assumed that five cycles of strong motion would occur for determining the cyclic strength characteristics of the soil and rock.The comparison between the artificial accelerogram and the actual accelerogram of any single earthquake applicable to the site is summarized in the following table below.COMPARISON OF ARTIFICIAL ACCELEROGRAM AND ACTUAL ACCELEROGRAM Duration of Strong Equivalent Number of Cycles of Motions, Sec Strong Motions Artificial Accelerogram 10 20 High peaks are those with an intensity greater than two-thirds the intensity of the maximum acceleration.Amendment 65 Page 296 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Actual Accelerogram of any Single 2 to 3 3 to 5 Earthquake Applicable to the Site The stability of the Category I dams under the SSE was conservatively evaluated on the basis of the stress induced by the artificial accelerogram shown in Figure 2.5D-8 as well as a accelerogram prescribed by Regulatory Guide 1.60 spectra and of the cyclic strength and stress-strain characteristics of the soil and rock materials using five cycles of strong motion. To demonstrate the additional margin of conservatism built into the dams, the stability was also evaluated assuming ten cycles of strong motion, which is essentially twice the expected number of cycles of strong motion.2.5D.4.4 Operating Basis Earthquake The operating basis earthquake (OBE) is taken to have maximum horizontal and vertical accelerations equal to one-half the corresponding accelerations assigned to the SSE.2.5D.4.5 Loading Conditions For dynamic analyses, two loading conditions (condition I and II) comprising the design water surface elevations and applicable design earthquake are considered for each dam.2.5D.4.5.1 Main Dam Condition I corresponds to normal operating conditions in the Main Reservoir and the Afterbay Reservoir. The water surface elevations are shown in Figure 2.5D-2. The SSE is assumed to occur.Condition II corresponds to the normal operating condition in the Main Reservoir and a condition which would occur if the Afterbay Dam were to fail and the water surface elevation in the Afterbay Reservoir were to drop instantaneously to Elevation 165 ft. The water surface elevations are shown in Figure 2.5D-2. The OBE is assumed to occur.2.5D.4.5.2 Auxiliary Dam Condition I corresponds to a high water condition one foot above the Auxiliary Dam spillway elevation in the Auxiliary Reservoir, ie Elevation 253 ft., and normal operating condition in the Main Reservoir. The water surface elevations are shown in Figure 2.5D-4. The SSE is assumed to occur.Condition II corresponds to the normal operating condition in the Auxiliary Reservoir and a condition which would occur if the water surface elevation in the Main Reservoir were to drop instantaneously to Elevation 209 ft. The water surface elevations are shown in Figure 2.5D-4.The OBE is assumed to occur.Amendment 65 Page 297 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.4.5.3 Auxiliary Separating Dike Condition I corresponds to a high water condition at Elevation 253 ft. in the Auxiliary Reservoir on both sides of the dike. The water surface elevation is shown in Figure 2.5D-6. The SSE is assumed to occur.Condition II corresponds to a low water condition at Elevation 243 ft. in the Auxiliary Reservoir on both sides of the dike. The water surface elevation is shown in Figure 2.5D-6. The OBE is assumed to occur.Conditions I and II considered for each dam in Appendix 2.5D are summarized in the following table. Condition III for the Main Dam and Auxiliary Dam is summarized in Section 2.5D.9.LOADING CONDITIONS Dam Condition Water Surface Elevation, ft. Design Earthquake Main Dam Main Reservoir Afterbay Reservoir I 250 199 SSE II 250 165 OBE Auxiliary Dam Main Reservoir Afterbay Reservoir I 250 253 SSE II 209 250 OBE Auxiliary Reservoir Main Reservoir Afterbay Reservoir Separating Dike I 253 253 SSE II 243 243 OBE Dynamic analyses were made for Condition I for each dam. Because a difference of water surface elevation of a few feet is of no significance, the analyses were made assuming the water surface in the Main Reservoir and Auxiliary Reservoir were at the normal water level; i.e.Elevation 250 ft. The seismic stability evaluations for Condition I are reported in Sections 2.5D.5, 2.5D.6, and 2.5D.7. The seismic stability evaluations for Condition II are reported in Section 2.5D.8.2.5D.5 MAIN DAM, EVALUATION OF SEISMIC STABILITY 2.5D.5.1 Material Properties The material properties used in the analyses and evaluation of the Main Dam are presented in Section 2.5D.10 for the core material, in Section 2.5D.12 for the filters and rockfill and in Section 2.5D.13 for weathered rock. The expected constructed material properties given in these appendices were used for the "basic set" analysis. To demonstrate the added margin of conservatism, these expected constructed values were varied over a reasonable range to determine the influence of such variations on the stability of the dam, as presented in Section 2.5D.15.Amendment 65 Page 298 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.5.2 Static Stress Analysis The static stresses in the Main Dam were computed using the static finite element method of analysis described in Section 2.5D.14. These static stresses were used to determine the cyclic strength in each material of the dam.Typical results of the static analysis are presented in Section 2.5D.14 and 2.5D.16.2.5D.5.3 Dynamic Stress Analysis 2.5D.5.3.1 General The stresses induced within the Main Dam during the SSE were computed using the dynamic finite element method of analysis. Three cross sections of the Main Dam were analyzed in detail. They include the maximum cross section with a height of 105 ft. (M-105), a cross section with a height of 67 ft. (M-67), and a cross section with a height of 36 ft. (M-36).A total of eight cases were evaluated for the maximum cross section of the Main Dam. The case in which the expected constructed material properties were used was designated by Roman numeral IV with no alphabetical letter attached to it; thus, for the maximum section of the Main Dam, the case is designated M-105-IV. For the other seven cases, an alphabetical letter is used to distinguish among the various cases. The material properties used for all eight cases are summarized in Section 2.5D.15.The responses for all cases, except Case M-105-IV, are obtained using the finite element procedure. The response for Case M-105-IV are obtained by interpolation of the computed values for Cases M-105-IV A and M-105-IV C, which is justified because of the similarity of the response of the last two cases.For the other two cross sections of the Main Dam, Cases M-67-IV A and M-36-IV A were analyzed.For each case, the computed responses are utilized in evaluating the stability of the dam as illustrated in Section 2.5D.16.2.5D.5.3.2 Crest Acceleration The computed values of maximum crest accelerations together with the predominant period of the three cross sections of the Main Dam for all cases are presented in Table 2.5D-1. The maximum acceleration computed at the crest of the maximum cross section of the Main Dam varies from approximately 0.40 g. to 0.47 g. The maximum acceleration computed at the crest of the 67-ft.-high cross section and at the crest of the 36-ft.-high cross section are 0.43 g. and 0.45 g., respectively.The crest accelerations due to an event prescribed by the Regulatory Guide 1.60 spectra yielded maximum values differing only by 0.01 g.; namely; ranging from 0.41 g. to 0.48 g. at the crest of the maximum cross section and values of 0.43 g. and 0.44 g. for crest accelerations of the 67-ft.-high and 37 ft. high cross section, respectively.Amendment 65 Page 299 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.5.3.3 Induced Shear Stresses The stresses induced in the dam are computed for the entire duration of the artificial accelerogram. These stresses are then converted to equivalent uniform stress cycles using the procedure summarized in Section 2.5D.16. Typical time histories of computed shear stresses are also presented in Section 2.5D.16.2.5D.5.4 Seismic Stability Evaluation 2.5D.5.4.1 General The stability of the dam sections during the SSE has been evaluated using the procedure described in Section 2.5D.16. In addition, the behavior of the dams and dike to an event prescribed by the Regulatory Guide 1.60 spectra have been assessed in terms of the effects on the local factors of safety as reported in this Appendix. A detailed evaluation of the effects of the Regulatory Guide 1.60 spectra was performed for the expected constructed material properties cases. The methods and details of the assessment and evaluation are given in methods and details of the assessment and evaluation are given in Section 2.5D.18. The induced stresses, d, at any location within the dam are compared to the stresses, f, required to cause 5 x 10-2 strain at the location. As discussed in Section 2.5D.16, a criterion of 5 x 10-2 strain is used for evaluating the stability of the dam during the SSE. This criterion has been established on the basis of correlations between the results of seismic stability evaluation by the procedure used for the present studies and the performance of earth dams which have been subject to significant earthquake loading (Reference 2.5D-28 and 2.5D-35). Case histories of earth dams which have been subjected to earthquake loading show that if the strain at any location within the dam and its foundation is smaller than 5 x 10-2, the earthquake had no effect on the stability and integrity of the dam. It should not be concluded that the stability and integrity of the dam is impaired if the strain exceeds 5 x 10-2 at some locations within the dam and its foundation. The effect of strains exceeding 5 x 10-2 depend on the zone of the dam where they occur, and on the relative extent and location within a specific zone.The ratio, f/d, which has been considered to represent a local factor of safety against the development of 5 x 10-2 strain, is then computed at each location. On the basis of the explanation given in the preceding paragraph, a minimum value of this stress ratio greater than approximately 1.1 indicates an ample margin of safety. The variation of this stress ratio is assessed within each zone of the dam to investigate the potential behavior and stability of the dam during the SSE.A typical determination of the ratio, f/d, is shown in Figure 2.5D-9 along a horizontal plane at Elevation 192.5 ft. (for case M-105-IV A). The stresses induced by the ten-second duration artificial accelerogram and the stresses required to cause 5 x 10-2 strain in five cycles together with the values of the ratio, f/d, along this plane are shown in this figure. As can be noted, the value of this ratio (or of the local factor of safety) are well over unity along this plane in each zone of the dam.Similar determinations have been made along several other horizontal planes within the dam representing the induced stresses by five cycles as well as by ten cycles; see Section 2.5D.4.The computed local factors of safety along the plane at Elevation 192.5 ft. using five cycles and ten cycles are presented in Figure 2.5C-10. The local factor of safety is reduced by Amendment 65 Page 300 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 approximately 10% to 15% using the conservative assumption that the induced stresses are generated by ten cycles instead of five cycles.The minimum values of computed stress ration, f/d, for all cases considered for the maximum cross section of the Main Dam are presented in Tables 2.5D-2, 2.5D-3, 2.5D-4, and 2.5D-5 using five cycles and ten cycles. The minimum values of this ratio computed within the core are listed in Table 2.5D-2; those computed in the upstream fine filter and coarse filter are presented in Tables 2.5D-3 and 2.5D-4, respectively; and those computed in the upstream rockfill shell are listed in Table 2.5D-5.The data in these tables together with examination of Figure 2.5D-9 provide the means to assess the stability of the Main Dam during the SSE. The plots in Figure 2.5D-9 and Figures 2.5D-84 through 2.5D-88 indicate that the minimum local factor of safety is obtained over only a very small part of the plane within the zone of the dam under consideration. Within any zone, the average computed local factors of safety is somewhat higher (in the core) to significantly higher (in the filters and rockfill shell) than the minimum values listed in Tables 2.5D-2 through 2.5D-5. The data in Tables 2.5D-2 through 2.5D-5 indicate that the minimum value of f/d is everywhere greater than unity and in most parts of the dam it is well over unity. Thus, the stability and integrity of the dam is assured during the SSE.The data in Tables 2.5D-2 through 2.5D-5 are presented for the various cases considered for the maximum cross section of the Main Dam. The data can be used to assess the effects of variations in material properties on the local factor of safety (and thus stability) of the dam.These effects are summarized below for each zone of the dam.2.5D.5.4.2 Maximum Cross Section, Core The computed minimum values of f/d in the core for Case M-105-IV (the basic set using expected constructed properties) range from 1.79 to 2.02 based on five cycles and from 1.57 to 1.75 based on ten cycles. Thus, there is an ample margin against the development of 5 x 10-2 strain anywhere in the core.Cases M-105-IV A and M-105-IV C were evaluated to assess the influence of variations of approximately +40 percent and -20 percent in the modulus of the rockfill shell, respectively. The data in Table 2.5D-2 indicates that these variations have very little effect (a slight decrease) on the minimum local factor of safety in the core. The minimum local factor of safety decreased from 1.79 to 1.77 for both cases based on five cycles minimum, and from 1.57 to 1.53 and 1.52, respectively, for case M-105-IV A and M-105-IV C, respectively, based on ten cycles.Case M-105-IV B was evaluated to assess the influence of increasing the modulus of the core by approximately 20 percent and decreasing the damping in the core by 10 percent. The minimum local factor of safety in the core is decreased from 1.57 to 1.40 based on ten cycles.This constitutes a reduction in the value of the minimum factor of safety of the order of 10 percent.Case M-105-IV BC is a combination of cases M-105-IV B and M-1-5-IV C. The minimum local factor of safety based on five cycles is reduced to 1.58 ie, by approximately 12 percent (compared to 1.79 for case M-105-IV), but is still well over unity. Similarly, the minimum local factors of safety based on ten cycles decreased to 1.37 compared to 1.57 for case M-105-IV which is also well above unity.Amendment 65 Page 301 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Case M-105-IV D was evaluated to assess the influence of increasing the modulus in the rockfill shell by 40 percent and decreasing the damping in the core by 10 percent and in the filters and rockfill shell by 20 percent. These variations have practically no effect (compared to Case M-105-IV) on the minimum value of f/d. The minimum local factor of safety based on five cycles decreased from 1.79 to 1.77 and from 1.57 to 1.54 based on ten cycles.Case M-105-IV E was evaluated using what is believed to be the lower bound on the modulus values in the core, filters, and rockfill shell. The minimum computed local factor of safety is increased (compared to case M-105-IV) from 1.79 to 2.20 based on five cycles and from 1.57 to 1.92 based on ten cycles, i.e., an increase of approximately 22 percent.Thus, the minimum local factors of safety in the core, taking into account a conservative combination of material properties, is 1.58 based on five cycles (Case M-105-IV BC) and 1.37 based on ten cycles (Case M-105-IV BC).2.5D.5.4.3 Maximum Cross Section, Fine Filters The computed minimum values of f/d in the upstream fine filter are listed in Table 2.5D-3 for all cases. These values range from 1.73 to 3.30 based on five cycles and from 1.55 to 2.95 based on ten cycles for Case M-105-IV. As can be noted in Table 2.5D-3, the variations used in modulus and damping have little effect (ranging from a decrease of approximately 5 percent for Case M-105-IV B to an increase of approximately 10 percent for case M-105-IV E) on the minimum local factor of safety.The minimum values of f/d within the downstream fine filter are greater than the corresponding values listed in Table 2.5D-3 (see Figure 2.5D-87 and 2.5D-88) because the water surface elevation is approximately 50 ft. lower in this part of the dam.It should be noted that the values of f/d listed in Table 2.5D-3 are based on the use of filters placed at a relative density of 75 percent.The data in Table 2.5D-3 indicate that the minimum local factor of safety in the fine filters, taking into account a conservative combination of material properties, is 1.65 based on five cycles and 1.47 based on ten cycles.2.5D.5.4.4 Maximum Cross Section, Coarse Filters The computed minimum values of f/d in the upstream coarse filter are listed in Table 2.5D-4 for all cases. These values range from 1.22 to 2.06 for five cycles and from 1.20 to 1.84 for ten cycles for Case M-105-IV.Increasing the modulus value in the shell by approximately 40 percent (Case M-105-IV A) increases the minimum value of f/d (for Case M-105-IV) to values ranging between 1.46 and 2.86 for five cycles and 1.31 and 2.55 for ten cycles, ie, by approximately 10 percent to 40 percent. Decreasing the modulus value in the shell by approximately 20 percent (Case M-105-IV C) decreases the minimum value of f/d to values ranging between 1.27 and 1.82 for five cycles and between 1.14 and 1.63 for ten cycles, ie, by approximately 5 percent to 10 percent.An increase in the modulus and a decrease in the damping of the core together with an increase in the modulus of the rockfill shell (Case M-105-IV B) has essentially no effect at Elevation Amendment 65 Page 302 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 237.5 ft. At lower elevations, these changes result in an increase of the minimum local factor of safety ranging from approximately 10 percent to 30 percent. The minimum factor of safety varies between 1.51 and 2.73 based on five cycles and between 1.35 and 2.44 based on ten cycles.The computed values of minimum local factors of safety for Case M-105-IV BC (combination of Cases B and C) are 1.21 based on five cycles and 1.08 based on ten cycles, i.e., approximately 5 percent to 10 percent lower than for case M-105-IV.A decrease in the damping in the core by 10 percent and in the filters and rockfill shell by 20 percent together with an increase of the modulus in the shell by approximately 40 percent (Case M-105-IV D) has very little effect at Elevation 237.5 ft. At lower elevations, these changes result in an increase of the minimum local factor of safety ranging from approximately 10 percent to over 30 percent. The minimum local factors of safety range between 1.55 and 2.82 based on five cycles and between 1.38 and 2.51 based on ten cycles. Case M-105-IV E (using the lower bound on the modulus values) provides minimum values of f/d comparable to Case M-105-IV D. The minimum local factor of safety is 2.58 based on five cycles and 1.42 based on ten cycles.2.5D.5.4.5 Maximum Cross Section, Rockfill Shells The computed minimum values of f/d within the upstream rockfill shell are presented in Table 2.5D-5. For Case M-105-IV, these values range from 1.51 to 2.12 based on five cycles and from 1.35 to 1.90 based on ten cycles. As can be noted in Table 2.5D-5, increasing the modulus in the rockfill shell by 40 percent (Case M-105-IV A) lowers the least value of f/d to 1.3, while decreasing the modulus in the rockfill shell by 20 percent (Case M-105-IV C) raises the least values of f/d to 1.53.Case M-105-IV B (modulus increases by 20 percent and damping decreases by 10 percent in the core) provides values of minimum f/d comparable to Case M-105-IV A. The minimum local factors of safety range from 1.36 to 1.90 based on five cycles and from 1.21 to 1.70 based on ten cycles. Case M-105-IV BC provide values of minimum f/d comparable to the basic set (Case M-105-IV). The minimum local factors of safety range from 1.53 to 2.22 based on five cycles and from 1.37 to 1.98 based on ten cycles. Thus, increasing the modulus by 20 percent and decreasing the damping by 10 percent in the core, increasing the modulus in the rockfill shell by 40 percent, and decreasing the damping 20 percent in the filters and rockfill shell have little effect on the computed minimum values to f/d.Case M-105-IV D (similar to Case M-105-IV A, except that the damping is decreased by 10 percent in the core and 20 percent in the filters and rockfill shell) provides minimum values of f/d ranging from 1.24 to 1.84 based on five cycles and 1.10 to 1.64 based on ten cycles, i.e.,approximately 10 percent to 20 percent less than for Case M-105-IV.The use of the lower bound values on modulus (Case M-105-IV E) indicates minimum values of f/d approximately 10 percent greater than for Case M-105-IV. The minimum local factors of safety range from 1.63 to 2.50 based on five cycles and 1.46 and 2.24 based on ten cycles.Thus, the minimum local factors of safety in the rockfill shell taking into account a conservative combination of material properties is 1.24 based on five cycles and 1.10 based on ten cycles.Amendment 65 Page 303 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.5.4.6 Influence of Vertical Component As illustrated in Section 2.5D.16, the induced shear stresses are essentially unaffected by the vertical component of the SSE. Therefore, the minimum computed local factors of safety are essentially unaffected by the vertical component.2.5D.5.4.7 Summary for Maximum Cross Section The data presented in Tables 2.5D-2 through 2.5D-5 indicate that the minimum local factors of safety in each zone of the maximum cross section of the Main Dam for the expected constructed properties is as follows: (Note: See Section 2.5D.18 for safety factors from an assessment in accordance with Regulatory Guide 1.60 spectra)MAIN DAM, MAXIMUM CROSS SECTION MINIMUM LOCAL FACTORS OF SAFETY FOR EXPECTED CONSTRUCTED MATERIAL PROPERTIES No. of Cycles Zone 5 cycles 10 cycle Core 1.79 1.57 Fine Filters 1.73 1.55 Coarse Filters 1.22 1.20 Rockfill Shell 1.51 1.35 It should be noted that these minimum local factors of safety are obtained within a very small part of each zone. The values of f/d increase considerably in the other parts of each zone.The minimum local factors of safety in each zone of the maximum cross section of the Main Dam for conservative combinations of material properties are presented in Tables 2.5D-2 through 2.5D-5.2.5D.5.4.8 Other Cross Sections Two other sections (Case M-67-IV A and Case M-36-IV A) of the Main Dam were also analyzed.The minimum values of f/d computed within each zone of these two sections are as follows:(Note: See Section 2.5D.18 for safety factors from an assessment in accordance with Regulatory Guide 1.60 spectra)MAIN DAM, 67-FT. AND 36-FT. CROSS SECTIONS MINIMUM LOCAL FACTORS SAFETY Core (a) Filters (a) Rockfill (a)Case 5 cycles 10 Cycles 5 Cycles 10 cycles 5 cycles 10 cycles M-67-IV A 2.30 2.05 1.59 1.42 1.57 1.40 M-36-IV A 2.48 2.20 1.82 1.63 1.60 1.43 (a) minimum value occurs at Elevation 237.5 ft.(b) minimum value occurs at Elevation 222.5 ft.Amendment 65 Page 304 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 This data indicates that there is ample margin against the development of 5 x 10-2 strain everywhere within these two sections of the Main Dam. Comparison of this data with that corresponding to other cases studied for the maximum cross section of the Main Dam (Tables 2.5D-2 through 2.5D-5) indicate that there is an ample margin of safety within these two sections of the Main Dam taking into account a very conservative combination of material property variations and considering ten equivalent uniform cycles of shear stresses.2.5D.5.5 Conclusions for Main Dam 2.5D.5.5.1 Maximum Cross Section a) The minimum local factor of safety ranges from approximately 1.2 to 1.8 based on five cycles using the expected constructed values of material properties.b) Incorporating a conservative combination of material properties, the minimum local factor of safety ranges from approximately 1.2 to 1.6 based on five cycles.c) For the same conservative combination of material properties, and incorporating an added margin of conservatism by basing the evaluation on ten cycles of strong motion, the minimum local factors of safety range from approximately 1.1 to 1.5.d) The minimum local factor of safety is obtained within a very small part of each zone of the dam. The local factors of safety are considerably higher in other parts of each zone.2.5D.5.5.2 Other Cross Sections The minimum local factors of safety in other cross sections of the Main Dam are higher than those for the maximum cross section.2.5D.5.5.3 Stability of the Main Dam During the SSE There is an ample margin in all cross sections of the Main Dam against the development of 5 x 10-2 strain. Therefore, the Main Dam will be stable and will maintain its integrity during the SSE.2.5D.6 AUXILIARY DAM, EVALUATION OF SEISMIC STABILITY 2.5D.6.1 Material Properties The material properties used in the analyses and evaluation of the seismic stability of the Auxiliary Dam are presented in Section 2.5D.11 for the core material, Section 2.5D.12 for the filters and the random rockfill, and Section 2.5D.13 for the residual soil and weathered rock.The cases analyzed, and the range of dynamic properties used in these analyses, are presented in Section 2.5D.15.2.5D.6.2 Static Stress Analysis The static stresses in the Auxiliary Dam were obtained by the procedure used for the Main Dam.Amendment 65 Page 305 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.6.3 Dynamic Stress Analysis 2.5D.6.3.1 General Three cross sections of the Auxiliary Dam were analyzed in detail by the procedure used for the Main Dam.These cross sections include the maximum cross section of the Auxiliary Dam (designated A-63), a cross section near the west abutment (designated A-44), and a cross section near the east abutment (designated A-24).Five cases for the maximum cross section of the dam are evaluated in detail; these cases are designated A-63-IV, A-63-IV A, A-63-IV B, A-63-IV C, and A-63-IV AV. Three cases are evaluated for the cross section near the west abutment and one case for the cross section near the east abutment. These cases are designated A-44-IV A, A-44-IV B, A-44-IV C, and A-24-IV A, respectively.2.5D.6.3.2 Crest Accelerations The computed maximum crest accelerations together with the predominant periods of the three cross sections of the Auxiliary Dam for all cases are presented in Table 2.5D-1. The maximum acceleration computed at the crest of the maximum cross section of the Auxiliary Dam varies from approximately 0.39 g. to 0.51 g. The maximum acceleration computed at the crest of the cross section near the west abutment (A-44) varies from approximately 0.45 g. to 0.48 g.; that at the crest of the cross section near the east abutment (A-24) is approximately 0.45 g.The crest accelerations due to an event prescribed by the Regulatory Guide 1.60 spectra yielded maximum values differing only by 0.01 g.; namely, ranging from 0.40 g. to 0.51 g. at the crest of the maximum section. The respective accelerations at the crest of the section near the west abutment (A-44) varied from 0.44 g. to 0.47 g.; that near the east abutment (A-24) was approximately 0.43 g.2.5D.6.3.3 Induced Shear Stresses The shear stresses induced in the Auxiliary Dam were computed and utilized for the evaluation of seismic stability by employing the procedure outlined in Section 2.5D.16.2.5D.6.4 Seismic Stability Evaluation for the Maximum Cross Section 2.5D.6.4.1 General The stability of the cross sections of the Auxiliary Dam during the SSE has been evaluated by the procedure used for the Main Dam.The stresses, d, induced in five cycles along a horizontal plane at Elevation 215 ft. within the maximum cross section of the Auxiliary Dam (case A-63-IV A) and the stresses, f, required to cause 5 x 10-2 strain in five cycles are presented in Figure 2.5D-11. The compound values of the local factor of safety along this plane using five cycles and ten cycles are presented in Figure 2.5D-12. The computed values of the local factor of safety (the ratio f/d) along this plane are also presented in this figure. Similar evaluations were made along several other Amendment 65 Page 306 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 horizontal planes within the three cross sections of the Auxiliary Dam. The values of the ratio f/d were obtained from these evaluations and used to assess the potential behavior and the stability of the Auxiliary Dam.The minimum values of computed stress ratio, f/d, for all cases considered for the maximum cross section of the Auxiliary Dam are presented in Tables 2.5D-6, 2.5D-7, and 2.5D-8 for five cycles and ten cycles. The minimum values of this ratio within the core are listed in Table 2.5D-6; those for the filters are presented in Table 2.5D-7; and those computed in the random rockfill shell are listed in Table 2.5D-8.As in the Main Dam, the minimum local factor of safety is obtained over only a very small part of the plane within the zone of the Auxiliary Dam. The local factors of safety are considerably higher in the other parts of the core and random rockfill shell; see Figure 2.5D-11 and 2.5D-12.The computed minimum values of f/d in the core are listed in Table 2.5D-6 for all cases.The computed minimum values of f/d in the core for Case A-63-IV (the basic set comprising the expected constructed properties) range from 1.43 to 2.13 based on five cycles and from 1.26 to 1.92 based on ten cycles. These values are obtained for a core placed at 97 percent standard compaction.Case A-63-IV A, in which the modulus in the random rockfill shell is increased by 67 percent, provides minimum values of f/d that are comparable to Case A-63-IV.Case A-63-IV B, which incorporates an increase of 25 percent in the modulus and a decrease of 10 percent in the damping for the core, indicates lower values of the ratio f/d than Case A IV. The minimum local factor of safety for Case A-63-IV B is 1.22 based on five cycles and 1.08 based on ten cycles.Case A-63-IV C incorporates an increase in modulus of 25 percent in the core, 33 percent in the filter, and 67 percent in the random rockfill shell together with a decrease in damping of 10 percent in the core and 20 percent in the filters and random rockfill shell. This case represents the upper bound on modulus values and the lower bound on damping values. The minimum values of f/d for this case are comparable to those obtained for case A-63-IV B; the least value is 1.29 based on five cycles and 1.14 based on ten cycles.2.5D.6.4.3 Filters The computed minimum values of f/d in the filters are listed in Table 2.5D-7 for all cases.These values range from 1.33 to 1.60 based on five cycles and from 1.19 to 1.42 based on ten cycles for Case A-63-IV.The minimum local factors of safety are obtained for Case A-63-IV C; they are 1.12 based on five cycles and 1.00 based on ten cycles.2.5D.6.4.4 Random Rockfill Shells The computed minimum values of f/d in the random rockfill shell are listed in Table 2.5D-8 for all cases. These values range from 1.56 to 2.28 based on five cycles and from 1.38 to 2.10 based on ten cycles for Case A-63-IV.Amendment 65 Page 307 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The minimum local factors of safety are obtained for Case A-63-IV A; they are 1.19 based on five cycles and 1.08 based on ten cycles.2.5D.6.4.5 Influence of Vertical Component Case A-63-IV AV, using the same material properties as for Case A-63-IV A and applying the horizontal and vertical components of the base motion simultaneously, was evaluated to assess the influence of the vertical component. The shear stresses computed for Case A-63-IV AV are essentially equal to those computed for Case A-63-IV A. Therefore, the minimum local factors of safety are essentially unaffected by the vertical component.2.5D.6.4.6 Summary The data presented in Tables 2.5D-6 through 2.5D-8 indicate that the minimum local factors of safety in each zone of the maximum section of the Auxiliary Dam for the expected constructed properties is as follows:(Note: See Section 2.5D.18 for safety factors from an assessment of the Regulatory Guide 1.60 spectra.)AUXILIARY DAM, MAXIMUM CROSS SECTION MINIMUM LOCAL FACTORS OF SAFETY FOR EXPECTED CONSTRUCTED MATERIAL PROPERTIES No. of Cycles Zone 5 cycles 10 cycle Core 1.43 1.26 Filters 1.33 1.19 Random Rockfill Shell 1.56 1.38 It should be noted that these minimum local factors of safety are obtained within a very small part of each zone. The values of f/d increase considerably in the other parts of each zone.The minimum local factors of safety in each zone of the maximum cross section of the Auxiliary Dam for conservative combinations of material properties are presented in Tables 2.5D 6 through 2.5D-8.2.5D.6.5 Seismic Stability Evaluation for Cross Section A-44 2.5D.6.5.1 General Based on the results for the maximum cross section of the Auxiliary Dam, the basic set for Section A-44 was not computed, because local factors of safety would be higher than for the other cases for which computations were made. The material properties for the core, the filters, and the random rockfill assigned to Case A-63-IV A are used in all analyses for Section A-44.The expected inplace values of the material properties for the in-situ residual soil and weathered rock are used for Case A-44-IV A. Lower bound values for the moduli of the residual soils and weathered rock are assigned for Cases A-44 IV B and A-44-IV C (see Section 2.5D.15).Amendment 65 Page 308 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.6.5.2 Accelerations Along Top of Weathered Rock The maximum acceleration computed along the top of the weathered rock for Cases A-44-IV A and A-44-IV B ranges from 0.151 g. to 0.156 g. Thus, using the expected in-place modulus for the weathered rock, the acceleration values are essentially unmodified through this layer. For Case A-44-IV C (using the lower bound modulus for the weathered rock), the maximum acceleration along the top of the weathered rock ranges from 0.16 g. to 0.17 g. For the latter case, the acceleration values are slightly modified through the layer of weathered rock.2.5D.6.5.3 Acceleration Along Top of In-situ Residual Soils The maximum acceleration computed along the top of the layer of residual soils underlying the dam for Case A-44-IV A ranges from 0.21 g. to 0.23 g. For Case A-44-IV B, these values range from 0.24 g. to 0.28 g. and for Case A-44-IV C, they range from 0.25 g. to 0.29 g. Thus, the layer of residual soil amplifies the rock accelerations by approximately a factor of 1.4 to 2.0.2.5D.6.5.4 Seismic Stability Cross Section A-44 The minimum local factors of safety for Case A-44-IV A are as follows: (Note: see Section 2.5D.18 for safety factors from an assessment of the Regulatory Guide 1.60 spectra).AUXILIARY DAM, SECTION A-44 MINIMUM LOCAL FACTORS OF SAFETY FOR EXPECTED CONSTRUCTED MATERIAL PROPERTIES No. of Cycles Zone 5 cycles 10 cycles Core 1.94 1.72 Filters 1.37 1.22 Random Rockfill Shell 1.50 1.40 Residual Soil 1.50 1.30 The above values were proportioned from the other cases for the Auxiliary Dam.The shear stresses computed in the core for Case A-44-IV B are up to approximately 30 percent higher (i.e., in some parts, the stresses are essentially equal for the two cases) than for Case A-44-IV A: in the filters, they are up to approximately 20 percent higher; in the random rockfill shell they are up to approximately 10 percent higher; and in the residual soils, they are up to approximately 10 percent higher. The stresses computed in all zones for Case A-44-IV C are essentially equal (within 5 percent) to those computed for Case A-44-IV B.Therefore, the minimum value of f/d in all parts of Cross Section A-44 are well over unity for all cases, taking into account the location at which the minimum value is obtained and the change in the induced stress at the location.Amendment 65 Page 309 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.6.6 Seismic Stability Evaluation for Cross Section A-24 Case A-24-IV A was analyzed to assess the stability of this cross section. The minimum local factors of safety for this case are as follows:(Note: see Section 2.5D.18 for safety factors from an assessment of the Regulatory Guide 1.60 spectra.)AUXILIARY DAM, SECTION A-24 MINIMUM LOCAL FACTORS OF SAFETY FOR EXPECTED CONSTRUCTED MATERIAL PROPERTIES No. of Cycles Zone 5 cycles 10 cycle Core 2.25 1.99 Filters 1.59 1.42 Random Rockfill Shell 1.50 1.40 Residual Soil 1.70 1.50 2.5D.6.7 Conclusions There is an ample margin against the development of 5 x 10-2 strain in the Auxiliary Dam.Therefore, the Auxiliary Dam is stable and will maintain its integrity due the SSE.2.5D.7 AUXILIARY RESERVOIR SEPARATING DIKE, EVALUATION OF SEISMIC STABILITY 2.5D.7.1 Material Properties The material properties used in the analyses and evaluation of the Auxiliary Separating Dike are presented in Section 2.5D.11 for the core material, Section 2.5D.12 for the random rockfill, and Section 2.5D.13 for the in-situ residual soils and weathered rock. The cases analyzed, and the range of dynamic properties used in these analyses, are presented in Section 2.5D.15.2.5D.7.2 Static Stress Analysis The static stresses in the sections of the Auxiliary Separating Dike were obtained by the procedure used for the Main Dam.2.5D.7.3 Dynamic Stress Analysis 2.5D.7.3.1 General The maximum cross section of the Auxiliary Separating Dike is analyzed in detail by the procedure used for the Main Dam. Three cases are evaluated in detail; these cases are designated D-53-IV A, D-53-IV B, and D-53-IV C.Amendment 65 Page 310 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.7.3.2 Crest Accelerations The computed maximum crest accelerations together with the predominant period for the three cases of the Auxiliary Separating Dike are presented in Table 2.5D-1. The maximum acceleration computed at the crest of the maximum section of the Auxiliary Separating Dike varies from approximately 0.45 g to 0.52 g. The crest accelerations due to an event prescribed by the Regulatory Guide 1.60 spectra yielded the same range of values.2.5D.7.3.3 Induced Shear Stresses The stresses induced in the Auxiliary Separating Dike were obtained and utilized using the same procedures outlined for the Main Dam in Section 2.5D.16.2.5D.7.4 Seismic Stability Evaluation The stability of the maximum cross section of the Auxiliary Separating Dike during the SSE has been evaluated by the same procedure used for the Main Dam.The stresses, d, induced in five cycles along a horizontal plane at Elevation 217 ft. within the maximum cross section of the Auxiliary Separating Dike (case D-53-IV A) and the stresses, f, required to cause 5 x 10-2 strain in five cycles are presented in Figure 2.5D-13. The computed values of the local factor of safety along this plane using five cycles and ten cycles are presented in Figure 2.5D-14. The computed values of the local factor of safety (the ratio f/d) along this plane are also presented in Figure 2.5D-14. Similar evaluations were made along several other horizontal planes within the maximum section of the Auxiliary Separating Dike.The values of the ratio f/d are obtained from these evaluations and used to assess the potential behavior and the stability of the Auxiliary Separating Dike.The minimum values of computed stress ratio, f/d, for all cases considered for the maximum cross section of the Auxiliary Reservoir Separating Dike are presented in Tables 2.5D-9 and 2.5D-10 for five cycles and for ten cycles. The minimum values of this ratio computed within the core of the dam are listed in Table 2.5D-9, and those computed in the random rockfill shell are listed in Table 2.5D-10.Based on the results for the maximum section of the Auxiliary Dam, the maximum cross section of the Auxiliary Reservoir Separating Dike was not analyzed using the basic set of material properties (Case D-53-IV). The minimum values of f/d for this case would be higher than those computed for the other three cases.As in the Main Dam and Auxiliary Dam, the minimum local factor of safety is obtained over only a very small part of the plane within the zone of the dam under consideration. The local factors of safety are considerably higher in the other parts of the core and shell; see Figure 2.5D-13 and 2.5D-14.The minimum local factors of safety in each zone of the maximum cross section of the Auxiliary Separating Dike for the expected constructed properties is as follows: (Note: see Section 2.5D.18 for safety factors from an assessment of Regulatory Guide 1.60 spectra.)Amendment 65 Page 311 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 AUXILIARY RESERVOIR SEPARATING DIKE, MAXIMUM CROSS SECTION MINIMUM LOCAL FACTORS OF SAFETY FOR EXPECTED CONSTRUCTED MATERIAL PROPERTIES No. of Cycles Zone 5 cycles 10 cycle Core 1.7 1.5 Random Rockfill Shells 1.9 1.7 The above values were proportioned from the other cases for the Auxiliary Reservoir Separating Dike based on the results for the Auxiliary Dam.The minimum local factors of safety in each zone of the maximum cross section of the Auxiliary Reservoir Separating Dike for conservative combinations of material properties is presented in Tables 2.5D-9 and 2.5D-10. The minimum local factors of safety in the 5-ft. layer of residual soil underlying the Auxiliary Reservoir Separating Dike ranges from approximately 1.8 to 2.1 based on five cycles and from 1.7 to 1.9 based on ten cycles.2.5D.7.5 Conclusion There is an ample margin against the development of 5 x 10-2 strain in the Auxiliary Reservoir Separating Dike. Therefore, the Auxiliary Reservoir Separating Dike is stable and will maintain its integrity during the SSE.2.5D.8 SEISMIC STABILITY EVALUATION FOR LOADING CONDITION II 2.5D.8.1 General Loading Condition II pertains to the use of the accelerations associated with the operating basis earthquake (OBE) as input in the suitable rock underlying the dams. The water surface elevations in the reservoir considered for this condition are summarized in Section 2.5D.4.5.The OBE is considered to have maximum horizontal and vertical accelerations equal to one-half the corresponding values assigned to the SSE.The stresses induced in the dams for Condition II would be considerably lower than those presented in Section 2.5D.16 and in Section 2.5D.5, 2.5D.6, and 2.5D.7 during the SSE. The reduction in the induced stresses however, would not be as much as the reduction in the suitable rock accelerations because of the non-linear stress-strain characteristics of the soil and rock materials. Based on numerous analyses for other dams and soil deposits, it is concluded that the induced stresses for Condition II would be approximately 60 percent of those computed for Condition I.The stability of the dams for Condition II can then be assessed based on the results obtained for Condition I, as summarized below.2.5D.8.2 Main Dam The water level in the Main Reservoir is assumed to be at Elevation 250 ft. The water level in the Afterbay Reservoir is assumed to drop instantaneously to Elevation 165 ft. and the OBE is assumed to occur.Amendment 65 Page 312 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The stresses induced by the OBE would be approximately 60 percent of those computed during the SSE. The water level in the main reservoir is the same as is used for Condition I.Therefore, the effective normal stresses in the upstream rockfill shell and filters would be essentially the same as for Condition I. The stresses required to cause 5 x 10-2 strain in these zones would, therefore, be the same as those used for Condition I. Because the induced stresses are decreased, the values of f/d listed in Tables 2.5D-3, 2.5D-4, and 2.5D-5 would be approximately 1/0.6 = 1.7 times higher for Condition II than computed for Condition I.Because the water surface elevation in the Afterbay Reservoir is assumed to drop instantaneously to Elevation 165 ft., the phreatic line in the core would not change significantly.The effective normal stresses in the core would, therefore, be essentially the same as those used for Condition I. For the same reasons used for the upstream rockfill shell and filters, the values of f/d listed in Table 2.5D-2 would be approximately 1.7 times higher for Condition II than computed for Condition I.The downstream rockfill shell and filters have a high permeability and, therefore, would no longer be fully saturated during the assumed instantaneous drop in the water level in the Afterbay Reservoir. The effective normal stresses in these zones would, therefore, increase with a corresponding increase in the stresses required to cause 5 x 10-2 strain. Even with the conservative assumption that the effective normal stresses will be equal to those used for Condition I, the values of f/d would be 1.7 times higher for Condition II than computed for Condition I.2.5D.8.3 Auxiliary Dam Condition II corresponds to the water level in the Auxiliary Reservoir at Elevation 250 ft. and the water level in the Main Reservoir dropping instantaneously to Elevation 209 ft. The OBE is assumed to occur.For the same reasons described in the preceding section for the Main Dam, the values of f/d in the various zones of the Auxiliary Dam and underlying foundation layers would be at least 1.7 times higher for Condition II than presented in Section 2.5D.6 for Condition I.2.5D.8.4 Auxiliary Reservoir Separating Dike Condition II corresponds to a low water condition at Elevation 243 ft. in the Auxiliary Reservoir on both sides of the dike. The OBE is assumed to occur.The effective normal stresses in the Auxiliary Separating Dike are increased because of the lowering of the water level on both sides. Therefore, the values of f used in Section 2.5D.7 would also be increased. The induced stresses, d, would be approximately 60 percent of those computed for Condition I. Therefore, the values of f/d within the core, the random rockfill shell, and the underlying residual soil would be 1.7 times the corresponding values presented in Section 2.5D.7 for Condition I.2.5D.8.5 Conclusion Based on the above considerations, it is concluded that the Main Dam, the Auxiliary Dam and the Auxiliary Reservoir Separating Dike each has considerable margin against the development Amendment 65 Page 313 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 of 5 x 10-2 strain for loading Condition II. Therefore, the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike are stable and maintain their integrity during the OBE.2.5D.9 SEISMIC STABILITY EVALUATION FOR LOADING CONDITION III 2.5D.9.1 Introduction This section presents the results of seismic stability analyses of the Category I Main Dam and Auxiliary Dam for the Shearon Harris Nuclear Power Plant. The location of the dams is shown in Figure 2.5D-15.The analyses presented in this section are for loading Condition III, which corresponds to elimination of the Afterbay Reservoir, and a decrease in the design water surface elevation in the Main Reservoir from Elevation 250 ft. to Elevation 220 ft.The results of previous seismic stability analyses for the dams for loading Conditions I and II are presented in Section 2.5D.5 through 2.5D.8.This section does not include seismic stability evaluations of the Category I Auxiliary Reservoir Separating Dike (Figure 2.5D-15) because loading Condition III is identical to loading Condition I for the Auxiliary Reservoir Separating Dike. Therefore, the seismic stability evaluation of the Auxiliary Reservoir Separating Dike for loading Condition I presented in Section 2.5D.7, is applicable for loading Condition III.2.5D.9.2 Description of Dams The Main Dam and Auxiliary Dam are described in Section 2.5D.2. The maximum cross sections and longitudinal sections of these dams are shown on Figure 2.5D-16 and 2.5D-17.2.5d.9.3 Procedure Used In Seismic Stability Evaluation Detailed dynamic analyses and evaluations of the Main Dam and Auxiliary Dam for loading Condition I have previously been made. The procedures used and results obtained for loading Condition I are described in Section 2.5D.5 through 2.5D.7.The safe shutdown earthquake (SSE) in suitable rock beneath the dams is identical for loading Condition I and loading Condition III. The only difference between loading Condition III and loading Condition I is the change in the water surface elevations. This will result in a lower phreatic surface in the dams for loading Condition III. Because the dynamic material properties of the dams, described in previous sections, do not change, the peak shear stress d of the equivalent uniform shear-stress cycles, induced in the dams by the SSE for loading Condition III, would not be significantly different from the stresses computed for loading Condition I.The lower phreatic surface in the dams associated with loading Condition III would, however, result in significantly higher static normal effective stresses in portions of the dams as compared to loading Condition I. Therefore, the dynamic strength, i.e., cyclic stresses, f, required to cause 5 x 10-2 strain, would significantly increase in portions of the dams for loading Condition III. Thus, in portions of the dams, the values of the local factor of safety, f/d, with respect to the development of 5 x 10-2 strain, would be higher for loading Condition III than for loading Condition I.Amendment 65 Page 314 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Based on the preceding reasoning, the following procedures were used to assess the seismic stability of the Main Dam and Auxiliary Dam for loading Condition III.a) Values of peak shear stresses d of equivalent number N of uniform shear stress cycles, induced in the dams, are obtained from the results of dynamic analyses made for loading Condition I.b) Values of static normal effective stress in the dams are computed based on the change in position of the phreatic surface for loading Condition III as compared to loading Condition I. For this computation, a close approximation of the increase in normal effective stresses is obtained by multiplying the vertical distance between the phreatic surfaces for loading Conditions I and III by the differences in effective unit weight for the two loading conditions.c) Values of the cyclic shear stresses f required to cause 5 x 10-2 strain in five and ten cycles are computed using the values of normal effective stress obtained in step (b).Curves relating f to normal effective stress are presented in Sections 2.5D.4 through 2.5D.7 for each material in the dams.d) Values of the local factor of safety f/d, based on five and ten cycles, are computed from the data obtained in steps (a) and (c).2.5D.9.4 Design Earthquake and Loading Conditions For loading Condition III, the safe shutdown earthquake (SSE), is assumed to occur. The SSE is defined in Sections 2.5.2 and 3.7. The SSE has a maximum horizontal acceleration of 0.15 g., and a maximum vertical acceleration of 0.10 g.The following water surface elevations apply for loading Condition III:Main Dam - The Main Reservoir will be at the normal water level, i.e., Elevation 220 ft. The Afterbay Reservoir will be eliminated. The water surface elevation of the downstream side will be Elevation 165 ft.; see Figure 2.5D-16.Auxiliary Dam - The Auxiliary Reservoir will be at a high water condition one foot above the Auxiliary Dam spillway elevation; i.e., Elevation 253 ft. The Main Reservoir will be at the normal water level; i.e., Elevation 220 ft. The water surface elevations are shown on Figure 2.5D-17.Loading Condition III for each dam is summarized below:Amendment 65 Page 315 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 LOADING CONDITION III Dam Water Surface Elevation, ft. Design Earthquake Main Reservoir Downstream Side Main Dam 220 165 SSE Main Reservoir Auxiliary Reservoir Auxiliary Dam 220 253 SSE For comparison, loading Condition I in the Main Dam pertains to water surface elevations of Elevation 250 ft. in the Main Reservoir, and Elevation 199 ft. in the Afterbay Reservoir; loading Condition I in the Auxiliary Dam pertains to water surface elevations of Elevation 250 ft. in the Main Reservoir, and Elevation 253 ft. in the Auxiliary Reservoir.Because a difference in water surface elevation of a few feet is not significant, the seismic stability evaluation of the Auxiliary Dam for loading Condition III and for loading Condition I are made assuming the water surface in the Auxiliary Reservoir to be at the normal water level; ie, Elevation 250 ft.2.5D.9.5 Main Dam, Evaluation Of Seismic Stability 2.5D.9.5.1 General The procedures described in Section 2.5D.9.3 are used to assess the seismic stability of the Main Dam during loading Condition III. The maximum cross section, designated M-105 (Figure 2.5D-16), has been selected for analysis because the maximum cross section was previously found to be more critical than lower cross sections of the dam during loading Condition I.2.5D.9.5.2 Maximum Cross Section The case selected for analysis is M-105-IVA. As described in Section 2.5D.15, Case M-105-IVA is for the case of expected constructed material properties in the core and filters of the dam and weathered rock beneath the dam and upper bound values of shear modulus in the rockfill shells.Case M-105-IVA has been selected because the results of previous analyses of this case, for loading Condition I, have been illustrated and described in detail in Section 2.5D.5 (Figure 2.5D-9, 2.5D-10, and Section 2.5D.16).Typical results of the seismic stability evaluation for Case M-105-IVA, for loading Condition III, are shown on Figure 2.5D-18. For comparison, the results for loading Condition I are also shown on Figure 2.5D-18.The upper plot in Figure 2.5D-18 shows the static normal effective stresses along a typical horizontal plane. Because of the lower reservoir water levels, the static normal effective stresses in portions of the dam are higher for loading Condition III than for loading Condition I.Values of the cyclic shear stress f required to cause 5 x 10-2 strain in five cycles along the typical horizontal plane are shown in the middle part of Figure 2.5D-18. These values are obtained using the normal effective stresses shown in the upper part of the figure, and the Amendment 65 Page 316 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 relationships between normal effective stress and f defined for each material in the dam (Section 2.5D.10 for core; Section 2.5D.12 for coarse and fine filters and rockfill shells).Because of the higher static normal effective stresses, the values of f are higher in portions of the dam for loading Condition III than for loading Condition I. For loading Condition III, the downstream filters and rock shells are not submerged, except for approximately 2 ft. above weathered rock. The non-submerged materials are not subject to porewater pressure increases during the earthquake motions, and have values of f greater than submerged materials.Therefore, values of f are not computed in the downstream filters and rockfill shells.The middle part of Figure 2.5D-18 also shows the values of the peak shear stresses d of the five equivalent uniform shear-stress cycles induced by the earthquake. As discussed in Section 2.5D.9.3 values of d for loading Condition III do not differ significantly from those for loading Condition I; therefore, the values computed for loading Condition I are used for loading Condition III.The lower part of Figure 2.5D-18 shows the variation of the local factor of safety f/d against the development of 5 x 10-2 strain in five cycles along the typical plane. Figure 2.5D-18 shows that the local factor of safety is higher in portions of the dam for loading Condition III than for loading Condition I.Evaluations similar to those illustrated in Figure 2.5D-18 were made along several horizontal planes in the Main Dam. A summary of the computed minimum values of f/d obtained in the respective materials along five horizontal planes is presented in Table 2.5D-11. Minimum values obtained, based five cycles and ten cycles, are summarized. For comparison, values of f/d obtained for both loading Condition I and loading Condition III are shown in Table 2.5D-11.Values of f/d are not summarized in the downstream filters and rockfill shells for loading Condition III because these materials are not submerged, and are, therefore, stable for this loading condition.2.5D.9.5.2.1 Core The computed minimum values of f/d in the core of the Main Dam for Case M-105-IVA, loading Condition III, range from 2.32 to 2.62 based on five cycles and from 2.01 to 2.29 based on ten cycles. These values are obtained for the core constructed at 100 percent standard compaction.The corresponding minimum values of f/d in the core for loading Condition I range from 1.75 to 2.22 based on five cycles, and from 1.53 to 1.92 based on ten cycles; see Table 2.5D-11. Thus, the minimum values of the local factor of safety in the core are higher for loading Condition III than for loading Condition I.The cyclic shear stress f required to cause 5 x 10-2 strain in the core material was found to depend not only on the static normal effective stress, but also on the value , defined as the ratio of static shear stress to static normal effective stress; see Section 2.5D.10. The effect of on f in the core is shown in Figure 2.5D-30 and 2.5D-31.The local factors of safety in the core for loading Condition III have been computed assuming that values of for loading Condition III are the same as the values determined for loading Condition I. However, the values of could be slightly different for loading Condition III and, therefore, the minimum local factors of safety f/d could show a slight variation. The maximum Amendment 65 Page 317 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 effect that variations in could have on values of f/d for loading Condition III can be determined by making the very conservative assumption that equals zero throughout the core. Using this assumption, it is found that the minimum values of f/d in the core would be essentially unchanged on horizontal planes at Elevation 192.5 ft. and Elevation 207.5 ft., and would be reduced by approximately 10 percent on a plane at Elevation 177.5 ft. Thus, even with these very conservative assumptions, the minimum values of f/d for loading Condition III would be higher than the values for loading Condition I.2.5D.9.5.2.2 Upstream Fine Filters The computed minimum values of f/d in the upstream fine filter for Case M-105-IVA, loading Condition III, range from 3.83 to 4.61 based on five cycles, and from 3.46 to 4.16 based on ten cycles. These values are obtained for filters placed at a relative density of 75 percent.Corresponding minimum values of f/d in the upstream fine filter for loading Condition I range from 1.70 to 4.03 based on five cycles, and from 1.52 to 3.58 based on ten cycles; see Table 2.5D-11.2.5D.9.5.2.3 Upstream Coarse Filters The computed minimum values of f/d in the upstream coarse filter for case M-105-IVA, loading Condition III, range from 2.75 to 3.38 based on five cycles and from 2.46 to 3.04 based on ten cycles. These values are obtained for filters placed at a relative density of 80 percent above Elevation 220 ft., and a relative density of 75 percent below Elevation 220 ft. Corresponding minimum values of f/d in the upstream coarse filter for loading Condition I range from 1.46 to 2.86 based on five cycles, and from 1.31 to 2.55 based on ten cycles; see Table 2.5D-11.2.5D.9.5.2.4 Upstream Rockfill Shells The computed minimum values of f/d in the upstream rockfill shells for Case M-105-IVA, loading Condition III, range from 1.30 to 1.88 based on five cycles, and from 1.16 to 1.68 based on ten cycles. Corresponding minimum values of f/5 x 10 in the upstream rockfill shells for loading Conditions I range from 1.30 to 1.88 based on five cycles, and from 1.16 to 1.68 based on ten cycles; see Table 2.5D-11.The above summary shows that there is no difference in the minimum values of f/d in the upstream rockfill shells between loading Conditions I and III. The minimum values of f/d in the upstream rockfill shells occur near the slope where the static normal effective stresses are unchanged by the difference in reservoir water level for the two loading conditions; see Figure 2.5D 18.2.5D.9.5.2.5 Influence of Vertical Component Based on the analyses of the maximum section of the Main Dam for loading Condition I (Section 2.5D.16), the vertical component of the SSE has an insignificant effect on the local factors of safety for loading Condition III.2.5D.9.5.2.6 Summary of Maximum Cross Section The computed minimum local factors of safety obtained in each zone of the Main Dam for Case M-105-IVA, loading Condition I and loading Condition III, are summarized below:Amendment 65 Page 318 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 MAIN DAM, MAXIMUM CROSS SECTION MINIMUM LOCAL FACTORS OF SAFETY CASE M-105-IVA Loading Condition I Loading Condition II 5 Cycles 10 Cycles 5 Cycles 10 Cycles Core 1.75 1.53 2.32 2.01 Fine Filters 1.70 1.52 3.83 3.46 Coarse Filters 1.46 1.31 2.75 2.46 Rockfill Shells 1.30 1.16 1.30 1.16 These minimum local factors of safety are obtained in only a small portion of each zone. The values of f/d increase considerably in other portions of each zone.For Case M-105-IVA, the minimum local factors of safety for loading Condition III are equal to or greater than the values obtained for loading Condition I. For other cases applicable to the maximum cross section, described in this appendix, the minimum local factors of safety for loading Condition III similarly equal or exceed the values presented in this appendix for loading Condition I.Analyses of the maximum cross section of the Main Dam for loading Condition I were made for several conservative combinations of material properties (Section 2.5D.15). These analyses indicated ample margin against the development of 5 x 10-2 strain in all zones of the dam. For loading Condition III, the computed factors of safety are equal to or greater than the values for loading Condition I. Therefore, there is ample margin against the development of 5 x 10-2 strain in the maximum cross section for loading Condition III.2.5D.9.5.3 Other Cross Sections Analyses of the Main Dam for loading Condition I (Section 2.5D.5) showed that the minimum local factors of safety obtained for other cross sections were higher than those obtained for the maximum cross section. As is the case for the maximum cross section, other cross sections of the Main Dam have a lower phreatic surface for loading Condition III than for loading Condition I. Therefore, other cross sections of the Main Dam will have higher factors of safety against the development of 5 x 10-2 strain for loading Condition III than for loading Condition I.2.5D.9.5.4 Conclusion Based on the evaluations presented above, it is concluded that there is ample margin against the development of 5 x 10-2 strain in the Main Dam for loading Condition III. Therefore, the Main Dam will be stable and will maintain its integrity for loading Condition III.2.5D.9.6 Auxiliary Dam, Evaluation of Seismic Stability 2.5D.9.6.1 General The seismic stability of the Auxiliary Dam for loading Condition III has been evaluated using the procedures described in Section 2.5D.9.3. The cross section selected for analysis is the maximum cross section, designated A-63 (Figure 2.5D-17). This cross section was selected Amendment 65 Page 319 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 because it was previously found (Section 2.5D.6) to be more critical than lower cross sections for loading Condition I.2.5D.9.6.2 Maximum Cross Section Case A-63-IVA has been selected for analysis. As described in Section 2.5D.6, Case A-63-IVA corresponds to expected constructed material properties in the core and filters of the dam and upper bound values of shear modulus in the random rockfill shells. Case A-63-IVA has been selected because the results of previous analyses for this case for loading Condition I have been illustrated in Section 2.5D.6 (Figures 2.5D-11 and 2.5D-12).Values of equivalent uniform shear stress d induced by the SSE, were obtained from the previous analysis for loading Condition I (Section 2.5D.6). Values of cyclic shear stress f required to cause 5 x 10-2 strain were obtained using the computed values of normal effective stress (higher in portions of the dam for loading Condition III than for loading Condition I, and the relationships between f and normal effective stress defined in this appendix for each material in the dam (Section 2.5D.11 for core; Section 2.5D.12 for filter and random rockfill shells).Results of the seismic stability evaluation for Case A-63-IVA, loading Condition III, along a typical horizontal plane are shown on Figure 2.5D-19. For comparison, results obtained along this plane for loading Condition I are also shown on Figure 2.5D-19.The minimum values of the local factor of safety f/d against the development of 5 x 10-2 strain along five horizontal planes for Case A-63-IVA are summarized in Table 2.5D-12. The values obtained for loading Conditions I and III based on five cycles and on ten cycles are presented in this table.2.5D.9.6.2.1 Core The computed minimum values of the local factor of safety f/d in the core of the Auxiliary Dam for Case A-63-IVA, loading Condition III, range from 1.47 to 2.42 based on five cycles, and from 1.31 to 2.17 based on ten cycles. These values are obtained for the core constructed at 97 percent standard compaction.For loading Condition I, corresponding minimum values of f/d range from 1.46 to 2.24 based on five cycles, and from 1.29 to 2.00 based on ten cycles; see Table 2.5D-12.The minimum local factors of safety in the core of the Auxiliary Dam are slightly higher for loading Condition III than for loading Condition I. The phreatic surface in the core, corresponding to loading Condition III, is slightly lower than the phreatic surface corresponding to loading Condition I. Therefore, the normal effective stresses in the core, and the values of f, are only slightly higher for loading Condition III than for loading Condition I.2.5D.9.6.2.2 Filters Because the water level in the Auxiliary Reservoir is identical for loading Condition I and loading Condition III, the phreatic surface in the upstream filter of the Auxiliary Dam is essentially the same for the two loading conditions. Therefore, the minimum values of f/d are essentially the same for the two loading conditions. For Case A-63-IVA, the computed minimum values of f/d Amendment 65 Page 320 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 in the upstream filter range from 1.53 to 1.75 based on five cycles, and from 1.36 to 1.56 based on ten cycles; see Table 2.5D-12.The phreatic surface in the downstream filter for loading Condition III is essentially at the elevation of the Main Reservoir; i.e., Elevation 220 ft. for loading Condition III, and Elevation 250 ft. for loading Condition I. The computed minimum values of f/d in the downstream filter for Case A-63-IVA, loading Condition III, range from 2.23 to 2.34 based on five cycles, and from 2.03 to 2.12 based on ten cycles. The corresponding minimum values for loading Condition I range from 1.53 to 1.75 based on five cycles, and from 1.36 to 1.56 based on ten cycles; see Table 2.5D-12.The minimum values of f/d, summarized above, pertain to the submerged portion of the filters (below the phreatic surface) , and are for filters placed at a relative density of 75 percent below Elevation 220 ft. and 80 percent above Elevation 220 ft.2.5D.9.6.2.3 Random Rockfill Shells The phreatic surface in the upstream random rockfill shell is essentially the same for loading Conditions I and III; the phreatic surface in the downstream random rockfill shells is lower for loading Condition III than for loading Condition I.The computed minimum values of f/d in the upstream random rockfill shells for Case A-63-IVA, applicable to loading Condition I and loading Condition III, range from 1.19 to 1.70 based on five cycles, and from 1.08 to 1.57 based on ten cycles.In the downstream random rockfill shells, the computed minimum values of f/d for Case A IVA, loading Condition III, range from 1.64 to 1.66 based on five cycles, and from 1.40 to 1.43 based ten cycles. The minimum values of f/d for loading Condition I range from 1.19 to 1.70 based on five cycles, and from 1.08 to 1.57 based on ten cycles.2.5D.9.6.2.4 Influence of Vertical Component Based on the analyses of the maximum cross section of the Auxiliary Dam for loading Condition I (Section 2.5D.6), the vertical component of the SSE has an insignificant effect on the local factors of safety for loading Condition III.2.5D.9.6.2.5 Summary for Maximum Cross Section The computed minimum local factors of safety obtained in each zone of the Auxiliary Dam for Case A-63-IVA for loading Condition I and loading Condition III are summarized in the following table:AUXILIARY DAM, MAXIMUM CROSS SECTION MINIMUM LOCAL FACTORS OF SAFETY CASE A-63-IVA Loading Condition I Loading Condition III Zone 5 Cycles 10 Cycles 5 Cycles 10 Cycles Core 1.46 1.29 1.47 1.31 Filter-Upstream 1.53 1.36 1.53 1.36 Amendment 65 Page 321 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Filter-Downstream 1.53 1.36 2.23 2.03 Random Rockfill 1.19 1.08 1.19 1.08 Shells-Upstream Random Rockfill Shells- 1.19 1.08 1.64 1.40 Downstream These minimum local factors of safety are obtained within a small part of each zone. The values of f/d increase considerably in other parts of each zone.For Case A-63-IVA, the minimum local factors of safety for loading Condition III are equal to or greater than for loading Condition I. For other cases applicable to the maximum cross section, described in this appendix, the minimum local factors of safety for loading Condition III similarly equal or exceed the values obtained for loading Condition I.As described in Section 2.5D.6, analyses of the maximum cross section of the Auxiliary Dam for loading Condition I were made for several conservative combinations of material properties.These analyses indicated ample margin against the development of 5 x 10-2 strain in all zones of the dam. For loading Condition III, the factors of safety are equal to or greater than those for loading Condition I. Therefore, there is ample margin against the development of 5 x 10-2 strain in the maximum cross section during loading Condition III.2.5D.9.6.3 Other Cross Sections Analyses of the Auxiliary Dam during loading Condition I (Section 2.5D.6) show that the minimum local factors of safety obtained in the core filters, and random rockfill shells of other cross sections are comparable to or greater than the values obtained for the maximum cross section. Similar results would be obtained for loading Condition III.Other cross sections of the auxiliary dam, i.e., Section A-44 and A-24, described in this appendix, include a layer of in-situ residual soil beneath the random rockfill shells. Analyses of these cross sections for loading Condition I (Section 2.5D.6) indicate ample margin against the development of 5 x 10-2 strain in the in-situ residual soil. For loading Condition III, the static normal effective stresses in this layer beneath the upstream shells would be unchanged from the stress determined for loading Condition I. In the layer of in-situ residual soil beneath the downstream shells, the lower phreatic surface associated with loading Condition III results in higher local factors of safety in this layer than were determined for loading Condition I.2.5D.9.6.4 Conclusion There is ample margin against the development of 5 x 10-2 strain in the Auxiliary Dam during loading Condition III. Therefore, the Auxiliary Dam is stable and will maintain its integrity during loading Condition III.2.5D.10 PROPERTIES OF MATERIAL M 2.5D.10.1 Introduction Material M is a composite material obtained by mixing representative composite samples from test pits in borrow area M; see Figure 2.5D-20. A test program was undertaken to determine Amendment 65 Page 322 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 the physical properties and static and dynamic stress-strain characteristics of this material.Section 2.5D.10 is devoted to the discussion of test procedures, interpretation of test data, and the derivation of static and dynamic characteristics of material M.2.5D.10.2 Origin and Preparation of Material M 2.5D.10.2.1 Geology Soils in borrow area M are residual soils formed by weathering of the Triassic deposits consisting of sandstones, siltstones, claystones, and conglomerates. The depth and degree of weathering are variable. The in-situ residual soils used for preparing composite material M are of stiff consistency and extend to depths of approximately 5 ft. to 10 ft. (Section 2.5D.13.3).2.5D.10.2.2 Sampling Locations The location of borrow area M is shown in Figure 2.5D-20. Six test pits were excavated in this area to depths ranging from 5 ft. to 10 ft. and samples of different soils encountered were withdrawn for index property test. Table 2.5D-13 summarizes properties of soils taken from representative borings and test pits in this area. For boring logs, see Appendix 2.5A.2.5D.10.2.3 Preparation of Composite Material During excavation, the proportion of different soils in each pit was determined and a composite sample was prepared by mixing soils in the same proportion as for those in-situ. Approximately 300 lb. of material was obtained from each pit, thoroughly mixed, and a representative portion withdrawn for index property tests. The tests showed that composite samples from two test pits (TPM4A and TPM5A) were not representative of the soils in other pits or borings in the area and, therefore, were not used for preparing material M. Composite samples from the remaining four pits (TPM1, TPM2, TPM3A, and TPM6) were mixed thoroughly to obtain composite material M which was used in all further laboratory investigations.2.5D.10.3 Physical Properties 2.5D.10.3.1 General The following physical properties were determined:a) index properties, including grain-size distribution, Atterberg limits, and specific gravity; b) compaction characteristics; and c) permeability.2.5D.10.3.2 Index Properties Index properties of the material are summarized in Table 2.5D-14 and the grain-size distribution curve is given in Figure 2.5D-21. On the basis of these properties, the material is described as red-brown silty clayey coarse to fine sand, with a trace of fine gravel; it is classified as SC according to Unified Soil Classification System.Amendment 65 Page 323 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.10.3.3 Compaction Characteristics Standard and modified compaction tests were performed in accordance with ASTM D698-68T and ASTM D1557 Method A, respectively. The wet method was used; i.e., the method which consists of adjusting the water content of the specimen by wetting or drying. The dry method; i.e., the method which consists of air drying the specimen and adding water, was not used because it modifies the properties of the residual soil and compaction in the field will be done under conditions similar to those used for the wet method. Standard and modified compaction characteristics are summarized in Table 2.5D-14. The standard and modified compaction curves obtained from the wet method are shown on Figure 2.5D-21.2.5D.10.3.4 Permeability Two specimens were compacted at 100 percent standard compaction and optimum water content for permeability tests. The tests showed that the material is rather impervious with a permeability of 10-8 cm/sec.; see Table 2.5D-14.2.5D.10.4 Static Stress-Strain Characteristics 2.5D.10.4.1 Purpose The purpose of the static tests is to determine the stress-strain and Poisson's ratio parameters needed in a non-linear static finite element analysis. Two testing procedures have been utilized for determining these parameters in previous studies (Reference 2.5D-17). In one procedure, unconsolidated-undrained (UU) triaxial tests are conducted on partially saturated specimens. In the other procedure, isotropically consolidated-drained (CID) triaxial tests are conducted on saturated specimens. Both testing procedures have been used to determine the parameters for material M required in the static stress analysis.2.5D.10.4.2 UU Tests 2.5D.10.4.2.1 Test Procedure Specimens were molded at 100 percent standard compaction and optimum water content and cured. A specimen was placed inside a rubber membrane, enclosed in a mercury-filled chamber, and subjected to confining pressure in a triaxial cell, no drainage being allowed to occur. An axial deviator stress was next applied at a strain rate of 1 x 10-2/minute. The applied stress, axial strain, and volume change were recorded.2.5D.10.4.2.2 Results of UU Tests The stress-strain curves and strength envelope derived from the UU tests are given in Figure 2.5D-22 and 2.5D-23, respectively.2.5D.10.4.3 CID Tests 2.5D.10.4.3.1 Test Procedure Four specimens were molded at 100 percent standard compaction and optimum water content.They were cured, isotropically consolidated under different confining pressures applied in Amendment 65 Page 324 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 increments, and saturated under backpressure. After saturation, each specimen was tested in a triaxial cell at an average strain rate of approximately 4 x 10-5/minute, allowing complete drainage. The applied deviator stress, axial deformation, and volume change were measured.2.5D.10.4.3.2 Results of CID Tests The stress-strain curves and strength envelope derived from the CID tests are given in Figures 2.5D-24 and 2.5D-25, respectively. These results were used to determine static strength parameters and stress-strain properties of the material; see Section 2.5D.14.4.2.5D.10.4.4 Derivation of Parameters for Static Stress Analysis For purposes of incremental static stress analysis, soil behavior can be represented by stress-dependent hyperbolic relationships (Reference 2.5D-16). Thus, the elastic properties (Et and t) at any point on the stress-strain curve can be conveniently related to the initial elastic properties and stress conditions through a set of parameters; different relationships are defined for primary loading and unloading or reloading (Reference 2.5D-6).For primary loading, the initial tangent modulus Ei can be expressed as (1)Where 3 is the effective confining pressures, pa is the atmospheric pressure, and K and n are parameters.The tangent modulus Et is represented by the equation Et = 1 (2) where Rf is the ratio of the stress difference at failure (1-3)f to the asymptotic stress difference (1-3)ult and c and are the Mohr-Coulomb shear strength parameters.Similarly, the Poisson's ratio at a given point on the stress-strain curve, t, can be expressed in terms of the initial Poisson's ratio, i, stress conditions and three parameters, G, F, and D (Reference 2.5D-17).ui = G - F log (3) t = (4)Using the above equations and data from the CID and UU tests, values of K, n, Rf, G, F and D were determined; see Table 2.5D-15.Amendment 65 Page 325 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.10.5 Dynamic Stress-Strain Characteristics Laboratory investigations were carried out to determine the cyclic strength and dynamic properties of material M. The test program consisted of the following:a) stress-controlled cyclic triaxial tests for determining the cyclic strength characteristics and shear modulus and damping ratio at high strain; b) strain-controlled cyclic triaxial tests for determining shear modulus and damping ratio at intermediate levels of strain; and c) cyclic torsion tests for determining the shear modulus and damping ratio at low levels of strain.2.5D.10.5.1 Specimen Preparation Specimens were molded at 95 percent and 100 percent standard compaction and optimum water content with a Harvard kneading compactor. For cyclic triaxial tests, specimens approximately 2.0 in. diameter and 4.0 in. high were used. For cyclic torsion tests, specimens of 1.4 in. high and 2.5 in. diameter were used. The specimens were cured, saturated under backpressure, and subjected to initial consolidation. Specimens for strain-controlled cyclic triaxial test and cyclic torsion tests were consolidated isotropically (Kc = 1) while those for the cyclic stress-controlled tests were consolidated under three different initial consolidation ratios, Kc = 1, 1.5, and 2.2.5D.10.5.2 Stress-Controlled Cyclic Triaxial Tests Thirty-three tests were conducted on specimens molded at 100 percent standard compaction and ten tests on specimens molded at 95 percent standard compaction. The tests were performed using a Modular Testing System (MTS) which applies a sinusoidally varying cyclic deviator stress of peak amplitude, d, at a frequency of one hertz. The applied deviator stress, axial deformation, and pore water pressure were measured. Loading is continued either until the specimen undergoes excessive strain or the number of cycles is high (N 1000).2.5D.10.5.2.1 Results of Stress-Controlled Cyclic Triaxial Tests Results of stress-controlled cyclic triaxial tests are summarized in Tables 2.5D-16, 2.5D-17, and 2.5D-18. For specimens compacted at 100 percent standard compaction, the relationship between the superimposed cyclic stress ratio ( /2 ) and the number of cycles required to cause initial liquefaction* for Kc = 1 is given in Figure 2.5D-26. Figure 2.5D-27, 2.5D-28, and 2.5D-29 present the relationship between the stress ratios and the number of cycles N required to cause 5 x 10-2 strain for Kc = 1, 1.5 and 2, respectively. These data were utilized for determining the cyclic strength of material M. Table 2.5D-18 presents the results of shear modulus and damping ratio determinations for several of the stress-controlled cyclic triaxial tests.Defined as the number of cycles causing either +/-2.5 x 10-2 axial strain or a pore pressure equal to the initial effective confining pressures, whichever occurs first.Amendment 65 Page 326 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.10.5.3 Strain-Controlled Cyclic Triaxial Tests 2.5D.10.5.3.1 Test Procedure Four tests were conducted at intermediate levels of strain (10-3 to 10-4) on isotropically consolidated specimens by subjecting them to sinusoidally varying axial strains at a frequency of one hertz. The deformations, axial load, and pore pressures developed in the specimens were recorded. Tests were continued up to approximately 30 cycles.2.5D.10.5.3.2 Test Results Results of the strain-controlled cyclic triaxial tests are presented in Table 2.5D-19. To account for end effects and non-uniform strains within the specimens, the average axial strain was obtained by dividing the measured axial strain by a correction factor of 1.5 (Reference 2.5D-39).The shear modulus and damping ratio of the material was determined as outlined in Table 2.5D-19.2.5D.10.5.4 Cyclic Torsion Tests 2.5D.10.5.4.1 Test Procedure Specimens for cyclic torsion tests were molded at 100 percent standard compaction, saturated and isotropically consolidated under different confining pressures. The tests were conducted by using the equipment and procedure suggested by Reference 2.5D-9. After consolidation, a vibrating head was attached to the top of the specimen and a resonant frequency of the system and the corresponding amplitude were determined. Upon the completion of each cyclic stage, the vibration decay response of the system was recorded. The confining pressure on the sample was increased to a higher level and tests for shear modulus and damping were repeated.2.5D.10.5.4.2 Results of Cyclic Torsion Tests Results of the cyclic torsion tests are given in Table 2.5D-20. To account for non-uniform torsional strains in the specimens, the peripheral shear strains were multiplied by a factor of 0.7 to obtain the average strain. The shear modulus and damping ratios were determined for the strain range of 10-4 to 10-5.2.5D.10.5.5 Determination of Cyclic Strength of Material M The cyclic strength of material M was evaluated by analyzing results of stress controlled cyclic triaxial tests presented in Table 2.5D-16 and 2.5D-17 and Figure 2.5D-26 through 2.5D-29.Analyses of several case histories and laboratory investigations have shown (Reference 2.5D-28 and Reference 2.5D-29) that cyclic triaxial tests on isotropically consolidated specimens indicate higher dynamic strengths than the stresses required to cause failure or excessive strains in the field. Therefore, for predicting the field cyclic strength of the material from cyclic triaxial tests on isotropically consolidated samples, different correction factors are incorporated as described below. For isotropically consolidated specimens (Kc = 1), the correction factor, Cr, generally lies between 0.55 and 0.80. For anisotropically consolidated specimens (Kc = 1.5 or Amendment 65 Page 327 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.0), the data from seed et.al. (Reference 2.5D-28) indicates that correction factors are not required to predict the field cyclic strength from cyclic triaxial tests.2.5D.10.5.5.1 Isotropically Consolidated Specimens It has been shown that for clean sands, the correction factor Cr is a function of relative density (Reference 2.5D-30). Because material M is a silty clayey sand, it is difficult to associate a relative density with the material at a given degree of compaction. However, through a comparison of published data (Reference 2.5D-30) and test results given on Figure 2.5D-26, an equivalent relative density can be determined for the material with respect to the relative density of a sand having the same mean grain size. It was found that for confining pressures varying from 2 k/ft.2 to 12 k/ft.2, the behavior of material M at 100 percent standard compaction at optimum water content is comparable to that of a sand having a relative density of approximately 64 percent to 100 percent, with the higher values corresponding to lower confining pressures. The following correction factors Cr were thus obtained.CORRECTION FACTORS FOR MATERIAL M Confining Pressure Correction Factor

 , k/ft.2 CR 2 0.80 4 0.69 8 0.65 12 0.61 Using the above correction factors and the data in Figure 2.5D-27, the relationship was determined between cyclic shear stress causing 5 x 10-2 strain in five cycles and the effective normal stress for isotropically consolidated specimens (Kc = 1); see Figure 2.5D-30. Similarly, the relationship between the cyclic shear stress causing 5 x 10-2 strain in ten cycles and the normal effective stress is shown in Figure 2.5D-31.

2.5D.10.5.5.2 Anisotropically Consolidated Specimens Cyclic tests of anisotropically consolidated samples are made in order to simulate the condition within the constructed embankment where there are initial static shear stresses on the potential failure planes. The procedures described in Reference 2.5D-28 have been used to interpret these tests to determine the cyclic strength characteristics. These procedures utilize Mohr envelope relationships to find the static stresses and the superimposed cyclic shear stresses on the plane of failure in the specimen.The ratio of the initial static shear stress to the normal stress on the plane of failure is designated . For the constructed embankment, is determined for various points in the embankment by the static finite element analysis, as described in Section 2.5D.14. For the laboratory samples, depends primarily on the consolidation conditions (i.e., on the value of Kc) and also on the strength parameter, '. For values of Kc = 1.0., 1.5, and 2.0 (and ' = 30°) the corresponding values of a are equal to 0, 0.19, and 0.345, respectively.The cyclic strength characteristics determined from tests on anisotropically consolidated specimens are presented on Figure 2.5D-30 and 2.5D-31. Figure 2.5D-30 shows the relationship between the normal effective stress and the cyclic shear stress causing 5 x 10-2 Amendment 65 Page 328 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 strain in five cycles; Figure 2.5D-31 shows the similar relationship for ten cycles. As can be seen in these figures, the cyclic stresses causing 5 x 10-2 strain are nearly equal for = 0.19 and = 0.345. Therefore, a single line has been drawn representing the cyclic strength characteristics for a 0.12.2.5D.10.5.6 Dynamic Properties of Material M Four parameters were used to define the strain-dependent dynamic properties of material M.The parameters are the shear modulus G, the damping ratio , the ratio of horizontal to vertical effective stresses, , and the Poisson's ration .2.5D.10.5.6.1 Shear Modulus G Shear modulus was determined from stress-controlled and strain-controlled cyclic triaxial tests and cyclic torsion tests for high, intermediate, and low levels of strain, respectively. Data given in Table 2.5D-19 and 2.5D-20 were utilized to determine the relationship between shear modulus and the normal effective stress at very low strains (10-6); see Figure 2.5D-32. It has been shown that this relationship can be represented by an equation (Reference 2.5D-30 and 2.5D-10).

 , (5) where Gmax is the maximum shear modulus in k/ft.2, is the mean normal effective stress expressed in lb./ft.2, and K2, max is a parameter. As can be seen from Figure 2.5D-32, an average value of K2, max = 120 was selected for material M compacted at 100 percent standard compaction at optimum water content.

The variation of shear modulus with shear strain was also established on the basis of results of the above tests. The shear moduli are plotted versus shear strain in Figure 2.5D-33. In order to present the data at various confining pressure of 1000 lb./ft.2 in Figure 2.5D-33. The relationship between shear modulus and shear strain was established on the basis of the data in Figure 2.5D-33; the generalized relationship is presented in Figure 2.5D-34. Figure 2.5D-32 and 2.5D-34 define the shear modulus characteristics of material M for use in the dynamic analyses.2.5D.10.5.6.2 Damping Ratio Damping ratios were determined from stress-controlled and strain-controlled cyclic triaxial tests and cyclic torsion tests; see Tables 2.5D-18, 2.5D-19, and 2.5D-20. The variation of damping ratio with strain established from the same data is given in Figure 2.5D-35.2.5D.10.5.6.3 Ratio of Horizontal to Vertical Effective Stresses, o Values of o were selected on the basis of published data for compacted and preconsolidated material (Reference 2.5D-5 and 2.5D-18). A value, o = 0.6, is selected to account for compaction of material.Amendment 65 Page 329 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.10.5.6.4 Poisson's Ratio Typical values of Poisson's ratio for different materials are available in several publications (e.g.,References 2.5D-23 and 2.5D-1). A value of = 0.35 was selected on the basis of published data.2.5D.11 PROPERTIES OF MATERIAL Z 2.5D.11.1 Introduction Material Z is a composite material obtained by mixing representative composite samples from test pits in borrow area Z; see Figure 2.5D-36. A test program was undertaken to determine the physical properties and static and dynamic stress-strain characteristics of material Z. This test program was similar to that reported in Section 2.5D.10 for Material M. The presentation of test results for material Z will be similar to that for material M in Section 2.5D.10. The test procedures and methods for interpretation of test data are similar to those discussed in Section 2.5D.10 and will not be repeated in this section.2.5D.11.2 Origin and Preparation of Material Z 2.5D.11.2.1 Geology Soils in borrow area Z are residual soils formed from Triassic sandstones, siltstones, claystones, and conglomerates. The in-situ residual soils are of stiff consistency and extend to depths of about 5 ft. to 10 ft.2.5D.11.2.2 Sampling Locations The locations of borrow area Z test pits and sampled borings are shown in Figure 2.5D-36.Table 2.5D-21 summarizes properties of soils taken from representative borings and test pits in borrow area Z. For boring logs see Appendix 2.5A.2.5D.11.2.3 Preparation of Composite Material The same procedure was used to prepare material Z as was used for material M. Soil from all four test pits was used for the composite material.2.5D.11.3 Physical Properties 2.5D.11.3.1 Index Properties The index properties of Material Z are summarized in Table 2.5D-22. The grain-size distribution curve is presented in Figure 2.5D-37. On the basis of these index properties, material Z is described as brown silty clay with some coarse to fine sand and trace of fine gravel; it is classified as CL according to the Unified Soil Classification System.2.5D.11.3.2 Compaction Characteristics Standard and modified compaction test results are presented in Table 2.5D-22 and Figure 2.5D-37.Amendment 65 Page 330 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.11.3.3 Permeability Tests were performed on two specimens compacted at 100 percent standard compaction and optimum water content. The measured permeability was 2 x 10-8 cm/sec.; see Table 2.5D-22.2.5D.11.4 Static Stress-Strain Characteristics 2.5D.11.4.1 General The test program for material Z was similar to that reported in Section 2.5D.10 for material M.Unconsolidated undrained (UU) triaxial tests and isotropically consolidated drained (CID) triaxial tests were performed in this test program.2.5D.11.4.2 UU Tests The stress-strain curves and strength envelope derived from the UU tests are given in Figure 2.5D-38 and 2.5D-39, respectively.2.5D.11.4.3 CID Tests The stress-strain curves and strength envelope derived from the CID tests are given in Figure 2.5D-40 and 2.5D-41, respectively.2.5D.11.4.4 Parameters for Static Stress Analysis The parameters K, n, Rf, G, F, and D for the static stress analysis are presented in Table 2.5D-

23. These parameters were determined from the results of the UU and CID triaxial tests.

2.5D.11.5 Dynamic-Strain Characteristics 2.5D.11.5.1 General The test program for Material Z was similar to that reported in Section 2.5D.10 for material M.The test program consisted of (1) stress-controlled cyclic triaxial tests; (2) strain-controlled cyclic triaxial tests; and (3) cyclic torsion tests.2.5D.11.5.2 Stress-Controlled Cyclic Triaxial Tests Results of stress-controlled cyclic triaxial tests are summarized in Tables 2.5D-24, 2.5D-25, and 2.5D-26. For specimens compacted at 97 percent standard compaction and optimum water content, the relationship between the stress ratio ( d/23c) and the number of cycles required to cause initial liquefaction for Kc = 1.0 is shown in Figure 2.5D-42. The relationships for d/23c vs. number of cycles required to cause 5 x 10-2 strain for Kc = 1, 1.5, and 2 are shown in Figure 2.5D-43, 2.5D-44, and 2.5D-45. Similar relationships for specimens molded at 100 percent standard compaction and optimum water content are shown in Figure 2.5D-46 (initial liquefaction) and 2.5D-47 and 2.5D-48 (5 x 10-2 strain for Kc = 1 and 1.5). The results of four tests made on specimens molded at approximately 98 percent compaction and 2 percent above optimum water content are included in Tables 2.5D-24 and 2.5D-25. Moduli and damping ratios determined from three stress-controlled cyclic triaxial tests on specimens at 97 percent standard compaction are given in Table 2.5D-26.Amendment 65 Page 331 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.11.5.3 Strain-Controlled Cyclic Triaxial Tests Results of strain-controlled cyclic triaxial tests at 97 percent standard compaction are presented in Table 2.5D-27.2.5D.11.5.4 Cyclic Torsion Tests Results of cyclic torsion tests at 97 percent standard compaction and optimum water content are presented in Table 2.5D-28, and results at 100 percent standard compaction and optimum water content are presented in Table 2.5D-29. Table 2.5D-29 also includes results of tests conducted on specimens compacted at approximately 100 percent standard compaction and water contents 2 percent above optimum.2.5D.11.5.5 Cyclic Strength of Material Z Correction factors have been developed to predict the field cyclic strength of material Z from data obtained from cyclic triaxial tests at Kc = 1.0. Following the same procedure as outlined in Section 2.5D.10, it was found that the cyclic strength of material Z at 97 percent standard compaction and optimum water content is comparable to that of a clean sand at relative density of 66 percent to 100 percent. The cyclic strength of material Z at 100 percent standard compaction and optimum water content is comparable to that of a clean sand at relative density 80 percent to 100 percent. Therefore, the following correction factors were obtained.CORRECTION FACTORS FOR MATERIAL Z Correction Factor, Cr Confining 97 Percent Standard 100 Percent Standard Pressure Compaction at Optimum Compaction at Optimum

 , k/ft.2 Water Content Water Content 1.25 0.80 0.80 2.5 0.73 0.785 5.0 0.62 0.68 2.5D.11.5.5.1 Isotropically Consolidated Specimens Ninety-seven percent standard compaction and optimum water content: Using the above correction factors, and the test data presented in Figure 2.5D-43, the relationship between cyclic shear stress causing 5 x 10-2 strain in five cycles and the normal effective stress was determined; see Figure 2.5D-49. The similar relationship for 5 x 10-2 strain in ten cycles is presented in Figure 2.5D-50.

One-hundred percent standard compaction and optimum water content: Using the above correction factors, and the test data in Figure 2.5D-47, similar relationships were obtained for 100 percent standard compaction. The cyclic shear stresses cuasing 5 x 10-2 strain in five cycles and ten cycles are shown in Figure 2.5D-51 and 2.5D-52, respectively.Amendment 65 Page 332 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.11.5.5.2 Anisotropically Consolidated Specimens The test results from anisotropically consolidated specimens showed that the cyclic strength characteristics of material Z are not sensitive to different values of Kc (or different values of ).Therefore, single curves have been drawn defining the cyclic strength characteristics at 5 x 10-2 strain; see Figure 2.5D-49 (five cycles) and 2.5D-50 (ten ycles) for material Z at 97 percent standard compaction and optimum water content and Figure 2.5D-51 (five cycles) and 2.5D-52 (ten cycles) for material Z at 100 percent standard compaction and optimum water content.2.5D.11.5.5.3 Effect of Molding Water Content on Cyclic Strength Characteristics The effect of the molding water content on the cyclic strength of material Z was investigated by making stress-controlled cyclic triaxial tests on specimens compacted at 98 percent standard compaction and 2 percent above optimum water content. The test results are reported in Tables 2.5D-24 and 2.5D-25. The cyclic strengths determined by these tests (as indicated by the magnitude of cyclic stress and number of cycles required to cause 5 x 10-2 strain) are generally within the range of cyclic strength determined at 97 percent to 100 percent standard compaction and optimum water content, and, therefore, are within the expected variation of the material properties.2.5D.11.5.6 Dynamic Properties of Material Z 2.5D.11.5.6.1 Shear Modulus G The relationships between shear modulus, mean normal effective stress, and shear strain for material Z at 97 percent standard compaction and optimum water content are presented in Figures 2.5D-53, 2.5D-54, and 2.5D-55, based on the data in Tables 2.5D-26, 2.5D-27, and 2.5D-28. Figure 2.5D-53 describes the relationship between shear modulus and mean normal effective stress at very low strains (10-6). Figures 2.5D-54 and 2.5D-55 show the relationship between shear modulus and shear strain. The curves shown in Figures 2.5D-53 and 2.5D-55 define the modulus relationships for dynamic analysis. As can be seen in Figure 2.5D-53, an average value of K2 max equal to 100 has been selected. The cyclic torsion test results on material Z at 100 percent standard compaction, presented in Table 2.5D-29, indicate that the shear modulus values are somewhat lower or essentially comparable to those obtained at 97 percent standard compaction. Lower values of modulus would not be expected (Reference 2.5D-10 and Reference 2.5D-31) and, therefore, it has been assumed that the shear moduli at 97 percent and 100 percent compaction are essentially the same.2.5D.11.5.6.2 Damping Ratio Damping ratios were determined for specimens compacted at 97 percent standard compaction and optimum water content from stress-controlled and strain-controlled cyclic triaxial tests and cyclic torsion tests; see Tables 2.5D-26, 2.5D-27, and 2.5D-28. The variation of damping ratio with strain established from these data is presented in Figure 2.5D-56.For material Z at 100 percent standard compaction, damping ratios were determined by cyclic torsion tests; see Table 2.5D-29. These data indicate that the damping ratios at 100 percent standard compaction are essentially comparable to those at 97 percent compaction.Amendment 65 Page 333 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.11.5.6.3 Ratio of Horizontal to Vertical Effective Stresses, o Values of o were selected on the basis of published data for compacted and reconsolidated materials (Reference 2.5D-5 and Reference 2.5D-18). A value of o = 0.6, was selected to account for compaction of the material.2.5D.11.5.6.4 Poisson's Ratio, A value of = 0.35 was selected on the basis of published data (Reference 2.5D-23).2.5D.12 PROPERTIES OF FILTERS AND ROCKFILLS 2.5D.12.1 Introduction The static and dynamic properties of filters and rockfill materials for the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike are described in this section. These properties were determined on the basis of the material index properties, design and construction criteria, and published and unpublished data on the properties of similar materials.2.5D.12.2 Filter Materials 2.5D.12.2.1 Main Dam 2.5D.12.2.1.1 Proposed Design and Construction Criteria The Main Dam will have two filters designated as fine and coarse filters, each 8-ft.-thick, on either side of the core. The specified grain-size limits and average curves for the fine and coarse filters are shown in Figure 2.5D-57. On the basis of these curves, the fine filter will be a well-graded coarse to fine sand, SW, with a mean grain-size (D50) of 0.56 mm. and the coarse filter will be a well-graded sandy gravel GW with a mean grain-size (D50) of 9.0 mm. The fine and coarse filters of the Main Dam will be compacted to an average relative density of 75 percent, except for the upstream coarse filter which will be compacted to an average relative density of 80 percent above Elevation 220 ft.2.5D.12.2.1.2 Static Properties The static properties of filter materials were selected on the basis of published data on compacted granular materials (Reference 2.5D-2; Reference 2.5D-18; and Reference 2.5D-17).Unit weights and parameters for static stress analysis are given in Table 2.5D-30.2.5D.12.2.1.3 Dynamic Properties The dynamic properties of the filters, viz the shear modulus G, the damping ratio , the ratio of horizontal to vertical effective stress Ko, the Poisson's ratio , and the variation of shear modulus and damping ratio with strain, were determined on the basis of published data on the behavior of granular materials.Amendment 65 Page 334 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.12.2.1.3.1 Shear Modulus G The shear modulus G and the variation of shear modulus with strain were based on published data (Reference 2.5D-30) for the fine filter and on the data reported in References 2.5D-30 and 2.5D-42 for coarse filter; see Table 2.5D-31 and Figure 2.5D-58.2.5D.12.2.1.3.2 Damping Ratio, Work by Wong (Reference 2.5D-42) has shown that damping ratios for gravels are in the same range as sands. Therefore, the average curve for variation of damping with strain (Reference 2.5D-30) was used for both filters; see Figure 2.5D-59.2.5D.12.2.1.3.3 Ratio of Horizontal to Vertical Effective Stress, Ko Typical values of Ko for compacted granular materials are available in literature (Reference 2.5D-5 and Reference 2.5D-18). A value of 0.6 was selected to account for compaction and a relative density of 80 percent.2.5D.12.2.1.3.4 Poisson's Ratio, Typical values of Poisson's ratios for compacted sands are available in the literature (References 2.5D-23 and 2.5D-1). A value of = 0.35 was selected on the basis of the published data.2.5D.12.2.1.4 Dynamic Strength The cyclic strength of the filter materials for the main dam has been estimated based on published data for granular soils. Factors considered in the assessment of the cyclic strength characteristics include: (a) relative density; (b) grain size; (c) gradation; and (d) effect of initial static shear stresses.The fine and coarse filters of the main dam are compacted to an average relative density of 75 percent except for the upstream coarse filter which is compacted to an average relative density of 80 percent above Elevation 220 ft.The effect of mean grain size on the cyclic strength of granular soils has been studied by Lee and Fitton, Seed and Peaco*ck, and Wong (see References 2.5D-22, 2.5D-29, and 2.5D-42).The results by Lee and Fitton indicate a substantial increase in cyclic strength as D50 increases from about 0.1 to 4 millimeters (mm.), and a very large increase in strength as D50 increases above 4 mm.; see Figure 2.5D-60. Because of the small diameter of samples tested in comparison to the maximum grain sizes, the results of Lee and Fitton may not be representative for soils with mean grain size greater than 10 mm. The data presented by Seed and Peaco*ck (Reference 2.5D-29) indicate a similar increase in strength of the Lee and Fitton results over the grain size range considered (0.1 to 1 mm.). The results of Wong (Reference 2.5D-42) indicate only a slight increase in strength as the mean grain size increases from 0.6 to 10 mm., and a rapid increase in strength for mean grain sizes larger than 10 mm.; see Figure 2.5D-60.Because Wong's tests were made using large diameter (12 in.) samples, the results are probably more representative than Lee and Fitton's for soils having a mean grain size greater than 10 mm. Field evidence during the Alaskan earthquake of 1964 also consistently indicates Amendment 65 Page 335 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 that gravelly soils with large grain size have greater cyclic strengths than do sands (Reference 2.5D-32).The mean grain size of the fine filter is estimated to be 0.56 mm. (average) and of the coarse filter 9.0 mm. (average). Although the available data would indicate a slight to substantial increase in strength as the grain size increases, this effect has been conservatively ignored in assessing the cyclic strength characteristics of the filters. Rather, the cyclic strength of the filters has been assessed on the basis of data for Sacremento River sand (Reference 2.5D-20).The data on this particular sand (a uniformly graded fine sand with D50 = 0.2 mm.) has been utilized because of the comprehensive cyclic test data published, covering a wide range of confining pressures, relative densities, and cyclic stress and strain levels.The proposed filter materials will be well-graded materials, whereas Sacramento River sand is a uniformly graded sand. Very little data is available on the cyclic strength of uniformly graded sands or gravels as compared to well-graded sands or gravels. Lee and Fitton (Reference 2.5D-22) tested a well-graded fine to coarse silty sand and concluded that there was very little, if any, difference in the strength of the well-graded sand as compared to a uniformly graded sand of the same mean grain size. Wong (Reference 2.5D-42) tested a well-graded sandy gravel and found that the strengths were lower than uniformly graded gravels at low strains (5 x 10-2);however, the well-graded gravels were more resistant to the development of large strains; i.e.,10 x 10-2.Therefore, the published cyclic triaxial test data for Sacramento River sand provide a reasonable basis for assessing the cyclic strength characteristics of the filters. Figure 2.5D-61, prepared from this data, shows the cyclic shear stresses required to cause 5 x 10-2 strain in five and ten cycles for filter materials compacted to a relative density of 75 percent or 80 percent.The same published data also indicate that for a relative density of 85 percent, the cyclic shear stresses required to cause 5 x 10-2 strain are approximately 1.35 times greater than the cyclic shear stresses required to cause 5 x 10-2 strain for 75 percent relative density. The strength characteristics shown in Figure 2.5D-61, derived from tests on isotropically consolidated specimens (Kc = 1.0), do not include a correction factor. Because initial static shear stresses may be expected to occur in the filters (see Section 2.5D-14), the field strength will be increased and the resulting strength envelope will be close to that obtained directly from tests on isotropically consolidated samples (the effect of static shear stresses on increasing the cyclic shear strength of granular soils is described in References 2.5D-28 and 2.5D-21.2.5D.12.2.2 Auxiliary Dam 2.5D.12.2.2.1 Proposed Design and Construction Criteria The specified grain-size limits and an average distribution curve are shown in Figure 2.5D-62.The transition filter for the Auxiliary Dam will be a well-graded sand gravel mix with a mean grain size (D50) of approximately 2.3 mm., and will be compacted to an average relative density of 75 percent below Elevation 220 ft. and to an average relative density of 80 percent above Elevation 220 ft.2.5D.12.2.2.2 Static Properties Static properties of the transition filter were derived by using the same procedure as for the filters of the Main Dam. These values are summarized in Table 2.5D-30.Amendment 65 Page 336 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.12.2.23 Dynamic Properties Dynamic properties of the transition filter for the Auxiliary Dam were determined in the same manner as for the filters for the Main Dam. These properties are given in Table 2.5D-31 and Figures 2.5D-58 and 2.5D-59.2.5D.12.2.2.4 Dynamic Strength The dynamic strength envelope shown in Figure 2.5D-61 is also applicable to the transition filter for the Auxiliary Dam.2.5D.12.3 Rockfill Materials 2.5D.12.3.1 Main Dam 2.5D.12.3.1.1 Proposed Design and Construction Criteria Rockfill for the main dam was obtained from the spillway, spillway outlet and approach channel excavations. It consists of a minimum of 75 percent fresh granitic rock in a size range of 6 in. to 24 in. with a maximum particle size of 24 in. The mean grain size (D50) is estimated at approximately 240 mm.2.5D.12.3.1.2 Static Properties The unit weights for the rockfill shell were selected on the basis of the rockfill at Keban Dam, Turkey (Reference 2.5D-8), which has properties similar to that of the proposed rockfill. Other static properties were selected on the basis of published data (Reference 2.5D-17); see Table 2.5D-32.2.5D.12.3.1.3 Dynamic Properties 2.5D.12.3.1.3.1 Shear Modulus, G The shear modulus for the rockfill was selected from available data on sands and gravels (Reference 2.5D-42). Since the shear modulus of gravel is higher than that of sand, a maximum shear modulus for rockfill higher than that of gravel was selected; see Table 2.5D-33.The variation in shear modulus with strain is assumed to be similar to that for sands; see Figure 2.5D-58.2.5D.12.3.1.3.2 Damping Ratio, Because the damping ratios in granular soils are not very sensitive to grain size (Reference 2.5D-42) the average relationship for sands was used; see Figure 2.5D-59.2.5D.12.3.1.3.3 Ratio of Horizontal to Vertical Effective Stress, and Poisson's Ratio, Values were selected on the basis of published data (see References 2.5D-5, 2.5D-18, 2.5D-23, and 2.5D-1); see Section 2.5D.12.2.1.3 and Table 2.5D-33.Amendment 65 Page 337 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.12.3.1.4 Dynamic Strength The dynamic strength of the rockfill has been estimated in a similar manner as for the filters.Based on the data by Lee and Fitton (Reference 2.5D-22) and Wong (Reference 2.5D-42) presented in Figure 2.5D-60, the effect of mean grain sizes larger than 10 mm. results in a rapid increase in cyclic strength. The data by Wong is utilized because it is probably more representative of the behavior of large particle size material and results in lower cyclic strengths than the data by Lee and Fitton.In estimating the cyclic strength of the rockfill, the data in Figure 2.5D-60 by Wong has been extrapolated to mean grain size 75 mm. (approx 3 in.). The cyclic strength at 75 mm. is then compared with the strength of Sacramento River sand. By this comparison, the rockfill should be at least 65 percent stronger than Sacramento River sand. In Section 2.5D.12.2.1.4 the cyclic strength characteristics of Sacramento River sand were used to determine the cyclic strength of filter materials. Using the same procedure, the cyclic strength of compacted rockfill is estimated to be 1.65 times the strength of Sacramento River sand at 75 percent relative density. Figure 2.5D-63 presents the cyclic shear stresses required to cause 5 x 10-2 strain in five and ten cycles for rockfill at a relative density of 75 percent. As is the case in the filters, initial static shear stresses will occur in the rockfill. Therefore, the strength characteristics defined by Figure 2.5D-63 do not include a correction factor.It is estimated that the specifications will result in a rockfill with mean grain size of the order of 240 mm. (approximately 9 in.). The larger grain size should result in strengths greater than defined by Figure 2.5D-63 which is based on a mean grain size of 75 mm.In developing the cyclic strength characteristics, it has been assumed that pore pressures do not dissipate during the earthquake motions (i.e., the strength of the rockfill is for undrained conditions during cyclic loading). Because of the large particle size and high permeability of the rockfill, it is certain that any pore pressures would dissipate to some degree during the ground motions, and this will result in higher rockfill strengths.2.5D.12.3.2 Auxiliary Dam and Separating Dike 2.5D.12.3.2.1 Proposed Design and Construction Criteria The Auxiliary Dam and Auxiliary Separating Dike will have shell zones of random rockfill obtained from the excavations for the general plant area structures, channels, and the spillway.The rock will consist of weathered sandstone, siltstone, and claystone of which a minimum of 75 percent shall range in size from 1/4 in. to 24 in. The mean grain size (D50) is estimated at approximately 34 mm.2.5D.12.3.2.2 Static Properties The unit weight of random rockfill was based on in-place unit weight measurements of a rockfill at Amos Dam, West Virginia (Reference 2.5D-40), with properties similar to those of the proposed rockfill. The unit weight of siltstone and sandstone is generally slightly smaller than that of granitic rock used for the Main Dam rockfill.Other static properties were selected on the basis of published data used for selecting the properties for the rockfill for the Main Dam (Reference 2.5D-17); see Table 2.5D-32.Amendment 65 Page 338 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.12.3.2.3 Dynamic Properties The shear modulus for the random rockfill has been estimated on the basis of the seismic velocity measurements made in a similar rockfill at Amos Dam, West Virginia (Reference 2.5D-

40) and measurements made at Calaveras Rockfill Dam, California (Reference 2.5D-33).

At Amos Dam the rockfill zone consists of sandstone, siltstone, sand, and silt shales and up to five percent clay shales by volume. The rockfill was placed and spread by bulldozers with no additional compaction. As compacted, the rockfill has a maximum particle size of 8 in. with approximately 80 percent of the material being between 12 in. and 0.187 in. (No. 4 sieve). The in-place moist unit weight was determined as approximately 136 lb./ft.3. Data from P-wave, S-wave, and R wave velocity measurements indicated a possible range in the shear modulus parameter K2, max of from 60 to 150 for the random rockfill at Amos Dam.The rockfill at Calaveras Dam consists of material with a maximum particle size of 6 in., a few boulders of 12 in. size, and 25 percent fines. The rockfill has an estimated moist unit weight of 130 lb./ft.3. Shear wave velocity measurements in the upstream rockfill at depths ranging from 20 ft. to 60 ft. indicated that K2, max 90.On the basis of the above data, K2, max of 90 was selected for the random rockfill; see Table 2.5D-33. The variation of shear modulus with strain, and the variation of damping ratio with strain were selected using the same procedure as for the rockfill of the Main Dam; see Figure 2.5D-58 and 2.5D-59.The values for Poisson's ratio, , and ratio of horizontal to vertical effective stress, , were selected as described for the Main Dam rockfill.2.5D.12.3.2.4 Dynamic Strength The random rockfill for the Auxiliary Dam and Auxiliary Separating Dike consists of fragments of weathered sandstones, siltstones, and claystones well compacted under a vibratory roller. The softer fragments of siltstones and claystones tend to break down and the overall rockfill will be somewhat cohesive. At low confining pressures, it is reasonable to consider that the cyclic strength characteristics of the rockfill would be similar to the core material Z at 100 percent standard compaction which is derived from the same weathered rocks. At higher confining pressures, the grain-to-grain contacts will contribute a larger proportion of the strength and the material may exhibit a behavior intermediate between that of clean granular rockfill and a cohesive material such as core material Z.The estimated cyclic strength characteristics of the random rockfill presented in Figures 2.5D-64 and 2.5D-65 are based on the preceding considerations. At low confining pressures (up to approximately 1.5 k/ft.2), the results of tests on core material Z at 100 percent standard compaction (see Figures 2.5D-51 and 2.5D-52) have been used to define the cyclic stresses causing 5 x 10-2 strain in five and ten cycles. At higher confining pressures, a strength envelope intermediate between core material Z at 100 percent standard compaction and a clean granular rockfill have been estimated using the same procedure as described for the main dam rockfill, and using a relative density of 75 percent and a mean grain size of 30 mm. (approximately 1 1/4 in). (On the basis of specifications for random rockfill, the mean grain size of the random rockfill is estimated to be approximately 30 to 40 mm). Figure 2.5D-60 shows that for the mean grain Amendment 65 Page 339 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 size the cyclic strength would be approximately 1.3 times higher than that for Sacramento River sand.The strength characteristics derived from the random rockfill for five and ten cycles are shown in Figures 2.5D-64 and 2.5D-65, respectively.2.5D.13 PROPERTIES OF FOUNDATION MATERIALS 2.5D.13.1 Introduction The properties of the foundation materials used in the static and dynamic analyses for the Main Dam, Auxiliary Dam, and Auxiliary Separating Dike are summarized in this section. The foundation materials considered are in-situ residual soil and weathered rock. The selection of material properties used in the analyses was based on laboratory test results, field geophysical measurements, and published data on similar materials.The in-situ residual soil and weathered rock in the main dam foundation were derived from the granite and gneiss. The in-situ residual soil and weathered rock in the auxiliary dam and auxiliary dike foundations were derived from the local Triassic sedimentary rock consisting of claystone, siltstone, sandstone, and conglomerate. Differential weathering in all of these deposits has produced variable depths and degrees of weathering.In these analyses, suitable rock has been defined as material which has an average compressional wave velocity of 10,000 ft./sec. and requires blasting for excavation. Weathered rock is material which requires coring for sampling and can be excavated with a ripper on a Caterpillar D-8 tractor, but cannot be excavated by the blade of a Caterpillar D-8 dozer. In-situ residual soil occurs as a thin layer of stiff soil overlying weathered rock which can be sampled with a split spoon sampler and can be excavated with the blade of a Caterpillar D-8 dozer.2.5D.13.2 Properties of Weathered Rock 2.5D.13.2.1 Main Dam 2.5D.13.2.1.1 Static Properties Unit weights of rock samples from the Main Dam foundation area are given in Table 2.5B-3, Appendix 2.5B. The samples were taken from depths of approximately 22 ft.3 to 80 ft. and have unit weights ranging from 166 lb./ft.3 to 188 lb./ft.3. For weathered rock at depths of approximately 0 ft. to 20 ft., a slightly lower value of the unit weight equal to 150 lb./ft.3 was selected. Other parameters for static stress analysis were based on data reported in Reference 2.5D-17; see Table 2.5D-34.2.5D.13.2.1.2 Dynamic Properties 2.5D.13.2.1.2.1 Shear Modulus The shear modulus parameter K2 max was determined for the weathered rock at the Auxiliary Dam by field seismic measurements (see Section 2.5D.13.2.2.3.1 below). A value of K2 max =700 was obtained. Results of unit weight and compressional wave velocity measurements on rock at the Main and Auxiliary Dam areas are given in Tables 2.5B-3 and 2.5B-4, respectively.Amendment 65 Page 340 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 It is observed that for similar depths, the rock at the Main Dam has a higher unit weight and compression wave velocity then at the Auxiliary Dam. Therefore, it is conservative to use the same value of (K2 max = 700) for weathered rock at the Main Dam as was used for the Auxiliary Dam.2.5D.13.2.1.2.2 Variation of Shear Modulus and Damping Ratio with Shear Strain Because of the relatively high stiffness of weathered rock, the shear strains induced by the SSE are expected to be very low (of the order of 10-5). At this strain level, the reduction in shear modulus with strain is approximately 1 percent - 2 percent (Reference 2.5D-26) for typical rock and approximately 4 percent to 5 percent for sand (Reference 2.5D-30). Therefore to incorporate a measure of conservatism, the average relationship between shear modulus and shear strain for sands was used for weathered rock at the Main Dam. For the same strain level as above, the typical damping ratio of rock is low, of the order of 0.01 to 0.02 (Reference 2.5D-26). Work by Seed and Idriss (Reference 2.5D-30) and Hardin and Drenevich (Reference 2.5D-

10) has shown that, for a corresponding strain level, sand has a comparable average damping ratio. Therefore, the average relationship between damping ratio and shear strain for sand (Reference 2.5D-31) was used for weathered rock; see Figure 2.5D-59.

2.5D.13.2.1.2.3 Ratio of Horizontal to Vertical Effective Stress Values of earth pressure coefficient at rest for preconsolidated materials are available in the literature (Reference 2.5D-19). A value of = 0.6 was selected for weathered rock to account for preconsolidation effects.2.5D.13.2.1.2.4 Poisson's Ratio Typical values of Poisson's ratio are available in the literature (e.g., Reference 2.5D-19 and 2.5D-23). A value of 0.35 was selected for weathered rock.2.5D.13.2.2 Auxiliary Dam 2.5D.13.2.2.1 General In approximately 1400 ft. of the central portion of the Auxiliary Dam, the random rockfill shell is founded on weathered rock and in the remaining portions, the rockfill shell is founded on a thin layer of stiff in-situ residual soil overlying weathered rock. The core is founded on suitable rock.2.5D.13.2.2.2 Static Properties Unit weights of rock samples from depths of 18 ft. to 34 ft. from the Auxiliary Dam are given in Table 2.5B-4, Appendix 2.5B. The unit weights range between 150.7 and 177.6 lb./ft.3. For weathered rock at shallower depths a slightly lower value of unit weight equal to 150 lb./ft.3 was selected. Other parameters for static analysis were selected based on the data reported by Kulhawy in Reference 2.5D-17; see Table 2.5D-35.Amendment 65 Page 341 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.3.2.2.3 Dynamic Properties 2.5D.13.2.2.3.1 Shear Modulus The shear modulus parameter K2 max was determined by using the field shear wave velocity measurements and procedures reported in Appendix 2.5C. For an average shear wave velocity of 2500 ft./sec. at an average depth of 15.5 ft. and unit weight of 150 lb./ft.3 the value of K2 max is 700.2.5D.13.2.2.3.2 Variation of Shear Modulus and Damping Ratio with Shear Strain As in the case of weathered rock at the Main Dam, the variation in the shear modulus and damping ratio with shear strain can be represented by average relationships for sands.Therefore, the same curves were used; see Figures 2.5D-58 and 2.5D-59.2.5D.13.2.2.3.3 Ratio of Horizontal to Vertical Effective Stress A value equal to that for the weathered rock for the Main Dam was selected ( = 0.6); see Table 2.5D-35.2.5D.13.2.2.3.4 Poisson's Ratio Typical values of Poisson's ratio for sandstones and siltstones are available in literature (e.g.,Reference 2.5D-19). A value = 0.35 was selected for weathered rock.2.5D.13.2.3 Auxiliary Reservoir Separating Dike The core and random rockfill shell of the Auxiliary Reservoir Separating Dike is founded on stiff residual soil overlying weathered rock. The in-situ properties of the weathered rock were not required for the analyses of the maximum cross section of the Auxiliary Reservoir Separating Dike because the borings indicate that at this location, the weathered rock layer is only 2 ft. to 4 ft. thick and therefore, would not significantly influence the results of static and dynamic analyses.2.5D.13.3 Properties of In-situ Residual Soil in the Foundation of Auxiliary Dam and Auxiliary Reservoir Separating Dike 2.5D.13.3.1 General The properties of the in-situ residual soil in the foundation of the Auxiliary Dam and Separating Dike were investigated by laboratory tests and field geophysical measurements. The locations of test pits and borings where samples were obtained and the locations where geophysical measurements were made are shown in Figures 2.5D-66 and 2.5D-74.2.5D.13.3.2 Auxiliary Dam 2.5D.13.3.2.1 Static Properties The unit weight of the in-situ soil in the auxiliary dam area was selected on the basis of laboratory tests on undisturbed specimens; see Table 2.5D-36. A saturated unit weight of 134 Amendment 65 Page 342 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 lb./ft.3 was selected. Other parameters for static analysis were selected on the basis of material index properties, shear strength parameters (see Table 2.5D-36) and data published in Reference 2.5D-17.2.5D.13.3.2.2 Dynamic Properties 2.5D.13.3.2.2.1 Shear Modulus The value of the shear modulus parameter K2 max was determined on the basis of geophysical measurements made at the site (P2.5C.2-4 through 2.5C.2-7). Locations of geophysical measurements are shown in Figure 2.5D-66. Analysis of the measurements indicates that K2 max has a value ranging between 86 and 108 near Sta 14+60 and between 194 and 250 near Sta 34+60. It is considered that representative values are K2 max = 100 near Sta 14+60 and K2 max = 190 near Sta 34+60.2.5D.13.3.2.2.2 Variation of Shear Modulus and Damping Ratio with Shear Strain The residual soil shows variable degrees of weathering ranging from sandy silty material near the west abutment to very stiff transitional material resembling weathered rock near the east abutment. The variation of shear modulus and damping ratio with shear strain was characterized by the same relationships as for weathered rock; see Figures 2.5D-58 and 2.5D-59.2.5D.13.3.2.2.3 Ratio of Horizontal to Vertical Effective Stress Typical values of this ratio are available in literature (Reference 2.5D-5 and Reference 2.5D-18).A value of o= 0.6 was selected to account for the preconsolidation effects.2.5D.13.3.2.2.4 Poisson's Ratio Typical values are available in literature (e.g., References 2.5D-23 and 2.5D-1). A value of =0.30 was selected.2.5D.13.3.2.3 Cyclic Strength Characteristics The cyclic strength characteristics of the in-situ residual soil at the auxiliary dam were determined on the basis of a series of stress-controlled cyclic triaxial tests on undisturbed samples obtained from the site.2.5D.13.3.2.3.1 Sample Location The undisturbed samples were obtained from test trench TPA1 (near Sta 14+60) and TPA2 (near Sta 34+60); see Figure 2.5D-66.Undisturbed block samples, approximately 1 ft. x 1 ft. x 1 ft. cubes, were cut from the two test trenches. Undisturbed 3-in.-dia., thin-wall Shelby tube samples were also obtained from borings located 10 ft. to 20 ft. from the test trenches. A summary of the location, depth, and index properties of the samples obtained is presented in Table 2.5D-38.Amendment 65 Page 343 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.13.3.2.3.2 Sample Preparation Specimens for testing were obtained by careful trimming of the block and Shelby tube samples.The diameter of specimens tested was approximately 2 inches. Each specimen was placed in a triaxial cell, saturated, and consolidated; isotropic consolidation (Kc = 1.0) and anisotropic consolidation (Kc = 1.5, 1.7, and 2.0) were used.2.5D.13.3.2.3.3 Cyclic Testing Program Stress-controlled cyclic triaxial tests were performed on 15 specimens using the Modular Testing System (MTS). The applied cyclic deviator stress, axial strain, and pore waterpressure were measured. Loading was continued either until the specimen underwent excessive strain or until approximately 1000 cycles were reached. Following cyclic testing, the liquid and plastic limits and plasticity index of each specimen were determined.2.5D.13.3.2.3.4 Test Results A summary of the test results, including the cyclic stress ratio, /2 , and the number of cycles required to reach 2 x 10-2, 5 x 10-2, and 10 x 10-2 strain for each specimen is presented in Table 2.5D-38. As can be seen from the table, eight of the fifteen specimens strained less than 2 x 10-2 at 1000 cycles; nine specimens strained less than 5 x 10-2 at 1000 cycles. In order to assist in plotting and interpreting the data, the test results were extrapolated to estimate the number of cycles beyond the test duration (>1000 cycles) at which a given strain would occur; see Table 2.5D-38.2.5D.13.3.2.3.5 Determination of Cyclic Strength Characteristics The results summarized in Table 2.5D-38 are plotted in Figures 2.5D-67 through 2.5D-70.Figures 2.5D-67 and 2.5D-68 show the results for isotropically consolidated specimens (Kc =1.0) for a strain of 2 x 10-2 and 5 x 10-2, respectively. Figures 2.5D-69 and 2.5D-70 show similar results for anisotropically consolidated specimens (Kc = 1.5 1.7, and 2.0). Smooth curves have been drawn in these figures to define the relationship between the applied cyclic stress ratio and number of cycles. Because of the low strain attained by many of the specimens, some of the curves presented in Figures 2.5D-67 through 2.5D-70 had to be extrapolated from large numbers of cycles on the shape of similar curves determined for core materials M and Z (see Section 2.5D.10 and 2.5D.11). For a strain of 5 x 10-2 (Figures 2.5D-68 and 2.5D-70), curves are presented for 3c = 3 k./ft.2 only, as the specimens at lower confining pressures generally exhibited very small strains and extrapolation to larger strains and low numbers of cycles was not possible.The relationship between the cyclic shear stress required to cause 5 x 10-2 strain in five cycles and normal effective stress is presented in Figure 2.5D-71. The similar relationship for ten cycles is presented in Figure 2.5D-72. These curves were prepared from the test data shown in Figures 2.5D-68 and 2.5D-70 and using procedures described in Section 2.5D.10. The results of the tests on isotropically consolidated samples indicated that the cyclic strengths were comparable to a clean sand having a relative density of 100 percent; therefore, a correction factor, Cr, equal to 0.80 was applied (Reference 2.5D-30).Amendment 65 Page 344 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 In the evaluation of the potential for attaining 5 x 10-2 strain in the in-situ residual soil during the earthquake motions, the curves for = 0 in Figures 2.5D-71 and 2.5D-72 have been used, i.e.,higher cyclic strengths indicated for >0 have been ignored.2.5D.13.3.3 Auxiliary Reservoir Separating Dike 2.5D.13.3.3.1 Static Properties The in-situ residual soil in the Auxiliary Reservoir Separating Dike area is generally similar to that in the Auxiliary Dam area. Figure 2.5D-73 presents the index properties and compaction characteristics of a sample from test pit A. A saturated unit weight of 135 lb./ft.3 was used for analysis. Other parameters for static stress analysis were selected on the basis of material index properties and data published by Kulhawy (Reference 2.5D-17), see Table 2.5D-39.2.5D.13.3.3.2 Dynamic Properties 2.5D.13.3.3.2.1 Shear Modulus Seismic wave velocity measurements were made at locations shown in Figure 2.5D-74. The average shear wave velocity was 715 ft./sec. (Appendix 2.5C, p.2.5C.2-7). The velocity is considered to be representative of the material at depths one quarter to one half of the impact receiver spacing; K2 max values ranging between approximately 60 and 90 were obtained. In the analysis K2 max = 90 was used.2.5D.13.3.3.2.2 Variation of Shear Modulus and Damping Ratio with Shear Strain As in the case of residual soil at the Auxiliary Dam, average damping ratios for sands were used; see Figure 2.5D-59.2.5D.13.3.3.2.3 Ratio of Horizontal to Vertical Effective Stress Typical values are available in literature (e.g., Reference 2.5D-5 and 2.5D-18). A value of =0.60 was used; see Table 2.5D-39.2.5D.13.3.3.2.4 Poisson's Ratio Typical values are available in literature; (Reference 2.5D-1 and Reference 2.5D-23) (Leonards 1962, Barkan 1962). A value of = 0.30 was selected; see Table 2.5D-39.2.5D.13.3.3.3 Cyclic Strength Characteristics The in-situ residual soils at the Auxiliary Reservoir Separating Dike are similar to those at the Auxiliary Dam. Therefore, the cyclic strength characteristics defined in Figures 2.5D-71 and 2.5D-72 have also been used for the dike.Amendment 65 Page 345 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.14 STATIC STRESS ANALYSIS 2.5D.14.1 Introduction Static stress analyses were conducted to determine the normal and shear stresses in the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike. A knowledge of the initial (i.e.,prior to occurrence of earthquake) static effective stress conditions is required for the evaluation of the cyclic strength of materials in the dams and dike.2.5D.14.2 General Procedure An incremental finite element approach is used which simulates the construction of an embankment in a series of layers. The materials in various zones of the dam are assumed to have non-linear, stress-strain characteristics. The analysis is based on the procedures developed by Kulhawy (see Reference 2.5D-17) and Duncan and Chang (see Reference 2.5D-6). The dam is divided into several horizontal layers each represented by quadrilateral elements. During any increment, appropriate values of modulus E and Poisson's ratio are assigned to each element. After determining the stresses, E and are reevaluated for the average stress conditions during the new increment and compared with the assigned values. If a significant difference is obtained, a second analysis is made with adjusted values of E and until a reasonable correspondence is established between the input and computed values. This process is continued until the last layer is added.After the last layer is added, the water level is raised in one or two steps. The effect of seepage and buoyancy is taken into account. In the case of steady seepage, flownets are drawn and the computed seepage force is equally divided among the nodal points forming an element. The effect of buoyancy of stresses is evaluated for all elements below the water surfaces in the shell, core, and filters.2.5D.14.3 Selection of Material Properties 2.5D.14.3.1 Core Parameters for static stress analysis are derived from laboratory static triaxial tests on materials M and Z. The test results and discussion of parameters are presented in Sections 2.5D.10 and 2.5D.11. Consolidated undrained (UU) tests were made on partially saturated specimens; and, consolidated drained (CID) tests were made on saturated specimens.2.5D.14.3.2 Filters and Rockfills Parameters for these materials are based on index properties, design and construction specifications, and published data (Reference 2.5D-17) on dams built with similar materials; see Section 2.5D.12.2.5D.14.3.3 In-situ Soils and Weathered Rock The selection of parameters for static stress analysis was based on index properties, field and laboratory tests, and published data (Reference 2.5D-17); see Section 2.5D.13.Amendment 65 Page 346 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.14.4 Analysis of Dams 2.5D.14.4.1 Main Dam The maximum cross section (M-105) was analyzed. Foundation conditions for the maximum section consist of suitable rock below the core and a layer of weathered rock below filters and rockfill shells (see Figure 2.5D-2). The weathered rock layer is incorporated in the static (and dynamic) analyses.For the static analysis of the maximum section, the dam was divided into seven horizontal layers and the water level was raised in two steps. The finite element representation for the maximum cross section is shown in Figure 2.5D-75 and the material properties used for the analysis are given in Table 2.5D-40. Stresses were evaluated for two sets of material properties distinguished by the parameters used for the core; see Table 2.5D-41. Set M(2) in Table 2.5D-41 used the properties for the core obtained from CID test data, and Set M(3) used the properties obtained from UU test data. The normal stresses computed by the two analyses were not significantly different. The results of Set M(3) provided a more reasonable variation of the static shear stresses (Reference 2.5D-7). Set M(3) was, therefore, used for subsequent evaluations of the static stresses.Typical results of Set M(3) are presented in Figure 2.5D-76 and 2.5D-77 which show the vertical effective stresses and shear stresses on a horizontal plane at Elevation 192.5. As can be seen, a concentration of shear and normal stresses occurs in the filters. The stress concentration can be attributed to the sharp variation in the stiffnesses of the core, filters, and the rockfill within a relatively short horizontal distance. Redistribution of stresses can be expected to occur, and the actual stresses in the filters should be lower than computed while those in the shell should be higher than those computed (Reference 2.5D-7). On this basis, average curves were drawn, shown by the heavy lines in Figure 2.5D-76 and 2.5D-77, modifying the computed stresses, but still satisfying static equilibrium (i.e., the magnitude of the total shear and normal forces on the given plane does not change). The average curves in Figure 2.5D-76 and 2.5D-77 were then used to compute ratios of shear stress xy, to normal stress, . The values of = Txy/y are shown in Figure 2.5D-78.The two lower cross sections of the Main Dam (M-67 and M-36) have similar geometry and foundation conditions as the maximum section (M-105). Therefore, on a horizontal plane at a given distance below the crest of the dam, the static stresses in Section M-67 and M-36 would be similar to those in the maximum section, M-105. The results of the static finite element analysis of the maximum section were, therefore, utilized to determine the static stresses in Section M-67 and M-36.2.5D.14.4.2 Auxiliary Dam Two cross section were analyzed, viz the cross section at Sta 14+60 (A-63, maximum cross section) and the cross section at Sta 34+65 (A-44). The reasons for their selection were (a) variation in base geometry; and (b) difference in foundation conditions and relative size of the zones.Foundation conditions for the maximum section, A-63, consist of suitable rock below the core and a thin layer (approximately 3 ft. to 4 ft. thick) of weathered rock below the filters and rockfill shells (see Figure 2.5D-4). The thin weathered rock layer would not significantly affect the static Amendment 65 Page 347 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 (or dynamic) stresses in the embankment and is, therefore, neglected in the static (and the dynamic) analyses, i.e., the filters and rockfill are assumed to rest directly on suitable rock.Foundation conditions for cross section A-44 consist of suitable rock below the core and in-situ residual soil and weathered rock below filters and rockfill shells (see Figure 2.5D-5); these foundation layers are incorporated in the static (and dynamic) analyses.The maximum cross section A-63) was divided into six layers and the water level was raised to the design elevation in a single step. Material properties selected for the analysis are given in Table 2.5D-42. Five sets of material property combinations were used; see Table 2.5D-43.Sets A-II-8, and A II-9 were similar to A-II-5 and A-II-6, except that the Poisson's ratio parameter, G, equal to 0.30 was used in the random rockfill. Set A-II-7 incorporates both sets of core test parameters corresponding to the period of construction (UU data) and period of saturation (CID data) in a single analysis. Subsequent evaluation indicated that Set A-II-7 is not valid because the processes of consolidation and pore pressure dissipation cannot be taken into account during the saturation period. Results obtained from Set A-II-9, using properties from UU tests in the core appeared to provide most reasonable values of static stresses.One analysis was made for cross section A-44 using UU test data for the core and properties given in Table 2.5D-42 for other materials. The analysis of Section A-44 was also utilized as a basis for determining the static stresses in section A-24 which has a similar geometry and foundation.As in the case of the main dam, the vertical and shear stresses show sharp variations in the filter zones and rockfills. Average values were obtained for the evaluation of dynamic stability in the same manner as for the Main Dam.2.5D.14.4.3 Auxiliary Reservoir Separating Dike The maximum cross section was analyzed. The embankment is constructed on a layer of residual soil above a thin layer (approximately 2 ft. to 4 ft. thick) of weathered rock. As for Section A-63 of the Auxiliary Dam, the weathered rock is neglected in the static (and dynamic) analyses of the maximum cross section of the dike.The cross section analyzed had a 12-ft.-thick layer of foundation soil. Subsequent to the static analysis, the cross section design was modified so as to leave only a 5-ft.-thick layer of foundation soil, which was used in the dynamic analysis. The use of the thicker layer in the static analysis has a negligible influence on the evaluation in this case because (a) the normal stresses along vertical planes showed the same variation with depth in the foundation as in the dam, (b) the values of static shear stress in the foundation are not used in the evaluation of the cyclic strength of the foundation (see Section 2.5D.13), and (c) the foundation is relatively stiff, therefore, should not result in any significant arching and stress transfer in the embankment.The dike was assumed to be built in six horizontal lifts, and the water level was raised to design elevation in one step and two steps. Material properties used in the analysis are given in Table 2.5D-44 and the material property variations studied are given in Table 2.5D-45. Set D(3) in Table 2.5D-45, incorporating parameters from UU tests in the core, was used to obtain the static stresses for the evaluation.Amendment 65 Page 348 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The Auxiliary Reservoir Separating Dike does not contain filters; therefore, unlike the Main and Auxiliary Dams, sharp variations in stresses were not obtained in the case of the Auxiliary Reservoir Separating Dike.2.5D.15 PARAMETRIC STUDIES 2.5D.15.1 Introduction Parametric studies have been conducted to assess the effects of variations in material properties on the seismic response and, in particular, on the induced stresses in the various zones of the dam.The material properties required for a response computation are: unit weight, , shear modulus, G, damping ratio, , Poisson's ratio, , and the ratio of the horizontal to the vertical effective stresses, K .The shear modulus and the damping ratio for most soils are highly strain dependent. Therefore, the relationship between modulus and strain must also be known for a response computation.Such a relationship is most conveniently established through the use of a maximum shear modulus, Gmax, (corresponding to very low levels of strain, of the order of 10-6), and a reduction curve (e.g., Section 2.5D.10). Because the modulus of the foundation soils and the soils used in the construction of the dam are functions of the effective mean pressure, the maximum shear modulus for each soil type has been expressed by the parameter, K2, max (e.g., Section 2.5D.10).The relationship between damping and strain must also be known for a response computation.2.5D.15.2 Expected Constructed Values of Material Properties A basic set of material properties has been established for each zone of the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike and underlying foundations. The basic set represents values for the zone expected in accordance with the design and construction criteria. This basic set, therefore, incorporates the expected constructed values of material properties.To account for scatter in test data, some uncertainty in assigned values based on correlation with published data, and to incorporate an added margin of conservatism, the expected constructed values have been varied within a reasonable band. The bases for the variations and the combinations utilized in the analyses are presented in Sections 2.5D.15.3 and 2.5D.15.4.It may be noted that the parameters that have the largest effects on induced stresses are the shear modulus and the damping ratio. Therefore, the emphasis in the parametric studies has been made in varying these properties. The other material properties (, , and K ) were not varied because the probable range of variation is narrow. In addition, the effects of variations in these latter parameters are adequately incorporated by the variations in the shear modulus.Amendment 65 Page 349 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.15.3 Bases for Parametric Variations The basis for the parametric variations in each of the materials in the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike are discussed below. The combinations of these parametric variations used in the analyses are described in Section 2.5D.15.4.2.5D.15.3.1 Core Material M The shear modulus and damping relationships have been assigned based on a comprehensive laboratory testing program, described in Section 2.5D.10. An average value of K2 max, a shear modulus vs. shear strain reduction curve and a damping ratio vs. shear strain curve were selected based on the results of the test data. The average value of K2 max, selected is 120. For parametric studies, K2 max is varied from 90 to 140 which essentially covers the variation in the experimental data. The value of K2 max determines the values of maximum shear modulus, Gmax, at very low strains (10-6). The shear modulus at higher strains is determined by means of the modulus reduction curve, G/Gmax vs. (G is shear modulus at shear strain ). Thus, by varying the value of K2 max, the values of shear modulus G at any strain are varied in the same proportion. Therefore, the variation in K2 max effectively covers the variation in laboratory test data over the full range of shear strains.The average relationship selected for damping ratio in Section 2.5D.10 is close to the lower bound values of damping ratios determined by laboratory tests. To provide an additional margin of conservatism, the damping ratios are varied by 10 percent below the selected average values. Higher values of damping ratio are not incorporated as these would result in lower stresses.2.5D.15.3.2 Core Material Z Modulus and damping ratio relationships for material Z have been assigned based on a comprehensive laboratory testing program; see Section 2.5D.11. An average value of K2 max equal to 100 was selected. In parametric studies, K2 max is increased to 125 which encompasses all of the test data. Lower values of K2 max are not assigned because analyses of the Main Dam (with core material M) indicated that such lower values resulted in lower stresses in the core.Parametric variation in the damping ratio consisted of reducing the selected average damping curve by 10 percent. The resulting damping ratio curve is then essentially at or below the lowest values determined by laboratory tests.2.5D.15.3.3 Filters As described in Section 2.5D.12, average modulus and damping ratio relationships for the coarse and fine filters of the Main Dam and transition filter of the Auxiliary Dam were determined from published data on similar sandy and gravelly soils. For the coarse and fine filters, the values of K2 max selected for the basic analysis are close to upper bound values from published data for a relative density of approximately 75 percent to 80 percent. Therefore, higher values are not incorporated in the analysis. The values of K2 max have been reduced in one case (from 120 to 90 in the coarse filter and from 60 to 50 in the fine filter); see Table 2.5D-46. For the transition filter of the Auxiliary Dam, the value of K2 max is increased in one case from 90 to 120 Amendment 65 Page 350 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 which exceeds the upper bound values inferred from published data for a relative density of 75 percent to 80 percent.For all filter material average damping ratio values were reduced by 20% which is a reasonable variation for a single material. The results of the analysis showed that a simultaneous reduction in damping ratio in all materials resulted in relatively small changes in induced stresses; the effect of changes in modulus were more significant in affecting the stresses.2.5D.15.3.4 Rockfill An average of K2 max equal to 180 was selected for the rockfill of the Main Dam; see Section 2.5D.12. It is believed that this well-compacted granitic rockfill will be stiffer than the filter materials. Therefore, a lower bound value of K2 max equal to 150 has been assigned to the rockfill, as compared to the selected value of 120 for the coarse filter of the Main Dam. To cover the possibility of a higher modulus in the rockfill, an upper bound value of K2 max equal to 250 was assigned (a parametric variation of approximately 40% above the basic value). As was done for the filters, the assigned damping ratio in the rockfill was reduced by 20 percent in parametric studies.2.5D.15.3.5 Random Rockfill A value of K2 max equal to 90 was selected for the random rockfill of the Auxiliary Dam and Auxiliary Reservoir Separating Dike based on field seismic geophysical measurements in similar rockfills of dams; see Section 2.5D.12. The compacted random rockfill may be somewhat stiffer than the rockfills in which measurements were obtained; therefore, a value of K2 max equal to 150 (67 percent greater than the basic value) was used in the parametric studies. Damping ratio was varied as described for the rockfill of the Main Dam.2.5D.15.3.6 In-situ Residual Soil As described in Section 2.5D.13, values of K2 max in the residual soil at the Auxiliary Dam and Auxiliary Reservoir Separating Dike were determined based on geophysical measurements.The effect of increased modulus (plus 28 percent) and decreased damping ratio (minus 20 percent) was studied in parametric variations made for the Auxiliary Reservoir Separating Dike; see Table 2.5D-49.2.5D.15.3.7 In-situ Weathered Rock The values of K2 max assigned to the weathered rock at the Main Dam and Auxiliary Dam were assigned based on field geophysical measurements; see Section 2.5D.13. In order to study the effect of modulus changes in the rock and the response of the overlying materials, the value of K2 max was reduced from 700 to 250 in parametric variations for Section A-44 of the Auxiliary Dam; see Table 2.5D-48.Damping ratio was reduced by 20 percent below the basic value in one of the parametric variations made for the Main Dam; see Table 2.5D-46.Amendment 65 Page 351 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.15.4 Material Property Combinations Several combinations in material properties were used in the analyses of the Main Dam (maximum section M-105), Auxiliary Dam (maximum Section A-63 and A-44), and Auxiliary Reservoir Separating Dike (maximum Section D-53).The combinations are summarized in Tables 2.5D-46 through 2.5D-49. Emphasis was placed on selecting material property combinations which would tend to increase stresses in one or more zones above the stresses computed using the properties of the basic set.2.5D.15.4.1 Main Dam Material property combinations used in the analysis of the Main Dam are summarized in Table 2.5D-46. Eight sets of properties, including the basic set, were used. Sets A and C were used to study the effect of higher modulus and lower modulus, respectively, in the rockfill. Set BC was analyzed to assess the effect of lower modulus in the rockfill combined with higher modulus and reduced damping ratio in the core. Set B combined increased modulus and reduced damping ratio in the core with increased modulus in the rockfill shell; this set essentially constituted upper bound modulus values for all materials. In Set E, the modulus values for all embankment materials were decreased to the lower bound values. In Set D, the lower bound damping ratio was incorporated in all embankment materials. The effect of simultaneous application of horizontal and vertical accelerations was evaluated by set AV.2.5D.15.4.2 Auxiliary Dam, Maximum Section Material property combinations used in the analysis of the maximum section of the Auxiliary Dam are presented in Table 2.5D-47. Five sets of properties, including the basic set, were used.Using the results of the parametric studies for the Main Dam as a guide, sets A and B for the maximum section of the Auxiliary Dam incorporated increased moduli in the random rockfill and in the core, respectively. Set B also induced reduced damping ratio in the core. Set C combined increased moduli with decreased damping ratio in all materials. Therefore, set C represents the upper bound of modulus values and lower bound of damping values. The effect of simultaneous application of horizontal and vertical accelerations was studied in set AV.2.5D.15.4.3 Auxiliary Dam, Cross Section A-44 Table 2.5D-48 summarizes the material property combination used for section A-44 of the Auxiliary Dam. The analyses considered the effect of modulus changes in the random rockfill, in-situ residual soil, and weathered rock.Set A was used to evaluate the effects of an increase of the modulus of the rockfill to its upper bound value. Sets B and C were used to evaluate the effects of changes in the moduli of the in-situ residual soils and the weathered rock. The field geophysical measurements indicated a lower bound value of K2 max of 60 and 250 for the residual soils and weathered rock, respectively. Set B was used to assess the effects of lowering the modulus of the residual soils to its lower bound value. Set C was used to assess the effects of decreasing simultaneously the moduli of the residual soils and the weathered rock. In both sets B and C, the upper bound value of K2 max for the rockfill was used.Amendment 65 Page 352 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.15.4.4 Auxiliary Reservoir Separating Dike Maximum Section Table 2.5D-49 summarizes the parametric variations made for the maximum section of the Auxiliary Reservoir Separating Dike. Variations in moduli and damping ratio of the core material, in-situ residual soil, and random rockfill were made. Set A evaluated the effect of increased modulus in the rockfill while set B evaluated the effect of increased modulus and decreased damping ratio in the core. Set C combined increased moduli with decreased damping ratio in all materials. Therefore, set C constitutes the upper bound on modulus values and lower bound on damping ratio. Because of the very small thickness (2 ft. to 5 ft.) of residual soil beneath the Auxiliary Reservoir Separating Dike, no additional changes in the modulus value of this soil was deemed necessary.2.5D.16 DYNAMIC ANALYSES AND STABILITY EVALUATIONS 2.5D.16.1 Introduction The general procedure for computing the response and evaluating the stability of an earth dam during an earthquake is briefly outlined in Section 2.5D.3. This procedure has been used in analyzing the Main Dam, Auxiliary Dam, and Auxiliary Reservoir Separating Dike. The use of this procedure is illustrated in this section for one of the cases studied. The case presented is M-105-IVA for the maximum section of the main dam using the material property combination assigned to this case; see Section 2.5D.15.2.5D.16.2 Dynamic Analyses The response of each section of the dam during the SSE was computed using the dynamic finite element method. The finite element representation used for the maximum cross section of the Main Dam is shown in Figure 2.5D-79. The finite element representation was extended 55 ft. and 120 ft. beyond the toes of the dam (i.e., six times the thickness of the weathered rock).In accordance with the criteria presented by Idriss (Reference 2.5D-12), this extension is sufficient so that the artificial boundaries do not affect the response of the dam.An equivalent linear procedure is incorporated in the finite element solution so that strain-compatible values of shear modulus G, and damping ratio, , are used for each element. Thus, at the outset, the modulus and damping ratio values are estimated and the response is computed. The computed value of average strain in each element is then used to choose (using the applicable G vs. and vs. curves (see Section 2.5D.10 through 2.5D.12) new values of modulus and damping. This process is continued until a compatible solution is obtained. Normally, three to five iterations are required.The values of strain-compatible damping ratios obtained for this case in the various zones of the dam are as follows:DAMPING RATIOS FOR CASE M-105-IV A Zone Damping Ratio, Core 0.14 to 0.20 Filters 0.08 to 0.15 Rockfill Shells 0.04 to 0.09 Amendment 65 Page 353 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 It should be noted that the damping ratio for each element in the finite element representation is selected using the average strain induced in the element and the applicable vs. curve. The damping ratios for elements in the core of the dam are based on the curve presented in Section 2.5D.10 for material M; see Figure 2.5D-35. This curve is well below the values obtained from the controlled-strain cyclic triaxial tests. Therefore, in the range of interest, the curve used for material M is a conservative representation of the test data. Accordingly, the damping ratios used in the analysis are justified although they exceed 0.15 in some parts of the core.Comparable damping ratios are obtained for the other dams analyzed.The strain-compatible shear modulus obtained in each element for case M-105-IV A was selected using the average strain induced in the element and the applicable K2, max and the modulus reduction curve. The values of K2 obtained in each zone of the dam are approximately as follows:SHEAR MODULUS PARAMETER, K2, FOR CASE M-105-IV A Zone K2, max K2 Used in Analysis Core 120 20 to 43 Fine Filters 60 20 to 31 Coarse Filter 120 51 to 72 Rockfill Shells 250 135 to 200 The applicable value of K2 (based on average induced strain) was then multiplied by the square root of the effective mean stress at the centroid of the element to obtain the shear modulus for that element.The response computations provide a variety of response values including accelerations at each nodal point and stresses in each element for the entire duration of the SSE.Typical results of the response computations are presented in Figure 2.5D-80 and 2.5D-81.Figure 2.5D-80 shows the time history of computed crest accelerations as well as the rock accelerations. The computed maximum crest acceleration is approximately 0.43 g. (i.e., the rock acceleration is amplified by a factor of approximately 2.85). The time histories of computed shear stresses at six selected points within the dam are shown in Figure 2.5D-81. These stresses are computed at the centroid of the element and, therefore, represent the stresses at a point. Two of these time histories are for stresses computed within the core of the dam; two are for stresses within the upstream filters; and two are for stresses computed in the rockfill shells.As can be noted from Figure 2.5D-81, the stresses are essentially in phase everywhere throughout the dam.2.5D.16.5 Determination of Equivalent Uniform Peak Shear Stresses The stresses computed within any part of the dam (see Figure 2.5D-81) vary in amplitude throughout the duration of the earthquake. The stresses induced in the dam are compared with the cyclic strength data, which are generally obtained from laboratory tests using cyclic loads of uniform peak amplitudes. Therefore, it is necessary to convert the computed time history to an equivalent number of uniform stress cycles. This is most conveniently accomplished by appropriate weighting of the time history of computed stresses as illustrated in Figure 2.5D-82.Amendment 65 Page 354 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 The data in Figure 2.5D-82 illustrate the use of the procedure to convert the computed time history within a typical element in the core to twenty cycles of equivalent uniform stress. As can be noted, the ratio of the amplitude of equivalent uniform stress to the computed maximum stress is approximately 0.64. Similar computation for the time history of shear stress at other locations within the dam (including the core, filters, and rockfill shells) indicate that this ratio varies from approximately 0.64 to 0.66. An average value of 0.65 was, therefore, used to obtain the equivalent uniform peak shear stresses at all locations within the dam.The conversion illustrated in Figure 2.5D-82 was done to obtain the equivalent uniform stress applicable to twenty cycles taking into account the entire duration (10.24 sec.) of the artificial accelerogram. However, as discussed in Section 2.5D.4, the stability of the dams under the SSE is evaluated on the basis of the stresses induced by the artificial accelerogram using five cycles of strong motion for the determination of the cyclic strength of the soil and rock materials.The equivalent uniform stresses computed by the procedure shown in Figure 2.5D-82 were considered to be applicable to five cycles. The data for the example presented in this section is based on the use of five cycles.To provide an additional margin of conservatism, the stability was investigated using ten cycles.The results of the evaluations using five and ten cycles for all the cases considered are presented in Section 2.5D.5 for the Main Dam, in Section 2.5D.6 for the Auxiliary Dam, and in Section 2.5D.7 for the Auxiliary Reservoir Separating Dike.2.5D.16.4 Stress Induced During Ground Motions The stability of the dam is evaluated by comparing the stresses induced during the ground motions with the cyclic strength characteristics along selected planes within the dam. Typical variations of equivalent uniform stresses induced for five cycles during the ground motions along four planes are illustrated in Figure 2.5D-83. Similar plots have been prepared for other planes. Values of equivalent uniform stresses along horizontal planes at Elevation 237.5, 222.5, 207.5, 192.5, and 177.5 are shown in Figure 2.5D-84 through 2.5D-88.2.5D.16.5 Seismic Stability Evaluation The stresses required to cause 5 x 10-2 strain (which has been chosen as the criterion for evaluating the stability of the dams at this site (see Section 2.5D.5.4.1) along any plane within the dam are obtained by using the static stresses and the applicable cyclic strength characteristics. The static stresses are evaluated by a static finite element procedure described in Section 2.5D.15. The cyclic strength characteristics of the core materials are obtained from the results of appropriate laboratory tests as described in Section 2.5D.10 (for the core material of the Main Dam) and Section 2.5D.11 (for the core material in the Auxiliary Dam and Auxiliary Separating Dike). For the filters, rockfill, and random rockfill, the cyclic strength characteristics are selected on the basis of index properties, material type, and design and construction criteria as summarized in Section 2.5D.12.The evaluation of the stability of the Main Dam (using the dynamic material properties assigned to case M-105-IVA) along five horizontal planes is illustrated in Figure 2.5D-84 through 2.5D-88.The upper part of each figure shows the values of initial static vertical effective stress along the plane. The middle part shows the values of stresses, d, induced by the artificial accelerogram together with the stresses, f, required to cause 5 x 10-2 strain in five cycles. Values of the ratio f/d are presented in the lower part of the figure. This ratio has been considered as a local Amendment 65 Page 355 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 factor of safety against development of 5 x 10-2 strain. As can be noted from Fgiure 2.5D-84 through 2.5D-88, the values of this ratio are well over unity (i.e., there is ample safety against the development of 5 x 10-2 strain during the rock accelerations). The minimum local factors of safety against the development of 5 x 10-2 strain in the various zones of the Main Dam for case M-105-IV A are as follows:MINIMUM LOCAL FACTORS OF SAFETY FOR 5 x 10-2 STRAIN CASE M-105-IVA Rockfill Shell Filters Upstream Downstream Plane at Elevation Core Upstream Downstream Fine Coarse Fine Coarse 237.5 1.75 1.66

• 1.70 1.24 *
• 222.5 1.77 1.56
• 2.24 1.62 *
• 207.5 1.91 1.30
• 2.85 2.13 *
• 192.5 2.18 1.54 1.97 3.84 2.86 4.50 3.57 177.5 2.22 1.88 2.31 4.03 2.70 4.94 3.47
• Not applicable because above phreatic line 2.5D.16.6 Influence of Vertical Component The results presented previously in this section were obtained using only horizontal rock accelerations. The same case was also analyzed using simultaneous horizontal and vertical rock accelerations. The artificial accelerogram of the vertical component was identical to that of the horizontal component (see Figure 2.5D-80) scaled down to a maximum vertical acceleration of 0.10 g.

The response for the dam (case M-105-IV A) during the simultaneous application of the horizontal and vertical components of the artificial accelerogram was computed using the procedure outlined earlier. These response values were then compared with those obtained by using only the horizontal component to assess the influence of the vertical component. The computed shear stresses along two typical horizontal planes are shown in Figure 2.5D-89 for the two cases. As can be noted, the stresses induced during the rock acceleration are essentially unaffected by the vertical component.2.5D.16.7 Consideration of Tensile Stresses The response of the dam-foundation system was computed by the finite element method in which the continuum is idealized by elements in plane strain and all components of the stress tensor are incorporated. The strength of the dam and foundation materials are determined on the basis of triaxial tests by imposing on specimens cyclic stresses which simulate that induced by the earthquake. These tests provide cyclic strength parameters applicable to the field conditions. Therefore, any possibility of development of tensile stresses is accounted for by the method of analysis and the method of determination of the cyclic strength parameters; i.e.,tensile stresses are not neglected.Amendment 65 Page 356 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.17 COMPARISON OF RESPONSE BY VARIOUS ANALYTICAL MODELS 2.5D.17.1 Introduction The response of the Category I dams during the SSE has been obtained using the finite element procedure. Strain-dependent values of modulus and damping are used in every element. A Rayleigh expression is used to formulate the damping submatrix for each element; the damping matrix for the entire system is then assembled in the usual way by adding the appropriate components of the submatrices for all elements. The response of the system is then computed by a direct integration of the equations of motion, ie, modal analysis is not used in this procedure. In fact, all the modes of the system are automatically incorporated in the solution. Because of the use of a Rayleigh expression for each submatrix, some of the higher modes of the system may be overdamped.The comparative studies presented in this appendix have been conducted to assess the effects of the use of this procedure on response values and, in particular, the effects on induced shear stresses.2.5D.17.2 Case Studied An 80-ft. layer of dense sand and gravel (Figure 2.5D-90) has been analyzed using three analytical procedures. The layer is assumed to have an extensive horizontal extent; therefore, mathematical models appropriate for a semi-infinite system have been used in the analyses.The following material properties have been assigned to this layer: qm = 135 lb./ft.3, Ko = 0.45, and K2 max = 100. The average curves relating damping and modulus reduction to strain published for sands have been used to obtain the strain-dependent properties for this layer.The response of this layer has been computed utilizing the artificial accelerogram (Figure 2.5D-

8) that has been used in the analysis of the Category I dams.

The thickness of this layer and the material properties have been chosen so that the predominant period is within the range of interest.2.5D.17.3 Analytical Models In addition to the finite element model, the soil layer was represented by two other models as shown in Figure 2.5D-90. The mathematical formulation for each model is briefly described below.2.5D.17.3.1 Finite Element Model The finite element method is a numerical procedure by means of which the actual continuum is represented by an assemblage of elements interconnected at a finite number of nodal points.Details of the formulation of the general method are available in several recent publications (eg, References 2.5D-3, 2.5D-4, 2.5D-12, and 2.5D-36).In earthquake response evaluations, the following set of equations are solved:[M] + [C] + [K] {u} = {R(t)} (6)Amendment 65 Page 357 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 in which[M] = mass matrix for the assemblage of elements,[C] = damping matrix for the assemblage of elements,[K] = stiffness matrix for the assemblage of elements, u = nodal displacements vector (dots denote differentiation with respect to time), and{R(t)} = earthquake load vector.A detailed description for the formulation of [M], [K], and R(t) is available elsewhere.In order to permit the use of a damping ratio for each individual element, a variable damping solution was recently developed (Reference 2.5D-14). This solution has been used in the analysis of the Category I dams and for this comparative study.In a variable damping solution, a damping submatrix must be formulated for each individual element and then all element submatrices added, in the appropriate way, to obtain the damping matrix for the entire assemblage of elements. The following relationship is used for each element, q:[c]q = q [m]q + q [k]q (7) in which [c]q, [m]q, and [k]q are the damping, mass and stiffness submatrices for element q, respectively, and q and q are parameters that are functions of the damping ratio and stiffness characteristics of element q. The parameters q and q are obtained from:q = q 1 (8) q = q/1 (9)The value q, which represents the damping ratio for element q, is chosen based on the strain developed in the element. The parameter 1 is equal to the fundamental frequency of the system.The equations of motion (equation (6)) are readily solved by a direct numerical method, such as the step-by-step method (Reference 2.5D-37). If a linear variation of acceleration is assumed over the time increment of integration, t, then the unknown response values at the nodal points at time, t, can be expressed in terms of the known values at time, t-t, as follows:[u]t = [ ]-1 {R}t (10)[ ] = [K] = 6[M]/t2 = 3[C]/t (11)[]t = {R}t + [M] + [C] (12){A}t = {u} t-t = t-t + 2{ü}t-t (13)Amendment 65 Page 358 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2{B}t = {u}t-t + 2{ü}t-t + {ü}t-t (14) t = {u}t - {B}t (15){ü}t = {u}t - {A}t (16)The stresses and strains developed in each element can then be readily computed using the values of {u}t.2.5D.17.3.2 Lumped-Mass Model The lumped-mass solution is a numerical procedure by means of which a semi-infinite layer is represented by a series of sublayers. Each sublayer is then represented by a series of lumped masses interconnected by shear springs. The applicable equations of motion are identical in form to equation (6). The matrices [M], [K], and {R(t)} are readily formulated (e.g., Reference 2.5D-13). The solution of equation (6) is carried out using modal superposition. The displacements of the lumped masses are expressed in terms of the normal coordinates and mode shapes by:{u} = [] {X} (17)[] are the mode shapes of the system and {X} are the normal coordinates.The mode shapes and frequencies are determined from a solution of the eigen-value problem for the undamped free vibration equations of the system (i.e., for [C] = 0 and {R(t)} = 0 in equation (6):[K] {n} = n2 [M] {n} (18)Each column, {n} of the matrix [] represents the mode shape of the nth mode of vibration whose natural circular frequency is n. The normal coordinates for each mode n, are evaluated from a solution of the normal equations:n + 2nnn + Xn = (19) in which n = damping ratio for mode n Mn = {n}t [M] {n} and T denotes the transpose of the vector.The damping ratio for the layer is the weighted average of the strain-dependent damping ratio obtained for each sublayer. Generally, the same damping ratio is used for all modes, i.e., 1 =2 =... n.The lumped-mass solution can also be used to investigate directly the effects of increased damping in higher modes. The damping ratios for each mode are proportioned based on the frequencies of the mode, i.e., for mode i, the damping ratio i = 1 i/1.Amendment 65 Page 359 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.17.3.3 Wave Propagation Model The wave propagation solution is a numerical procedure whereby the soil layer is divided into a series of horizontal sublayers overlying a perfect elastic half space. Details of the formulation of this solution are available in recent publications (e.g., References 2.5D-25, 2.5D-26).The motions within any sublayer j are assumed to satisfy the damped wave equation:j = Gj + nj t (20) in which u = u(t,z) is the horizontal displacement, t is time, z is depth within the jth sublayer, and j, Gj, and nj are the mass density, the shear modulus, and viscosity, respectively, of the jth sublayer.The steady state solution to equation (20) is:u = Uj(z) exp(it) (21) where Uj(z) = Ej exp (ikjz) + Fj exp(-ikjz) (22) and kj = /Vj (23)The variable Vj is the complex shear wave velocity for the material of the jth sublayer. This velocity is related to j, Gj, and nj through the relation Vj2 = (Gj + inj)/j (24)To provide a solution where energy dissipation is independent of frequency, equation (24) is restated in the form Vj2 = Gj(1 + i2j)/j (25) where i is the fraction of critical damping for the the material of the jth sublayer (reference 2.5D-26), which is chosen based on the strain induced in the sublayer.The first term of equation (22) refers to a shear wave which propagates in the negative z-direction (upwards in Figure 2.5D-90, part d) with the complex amplitude Ej and the second term refers to a wave which propagates in the position z-direction (downwards in Figure 2.5D-90, part d) with the complex amplitude Fj.The determination of the displacements u for each sublayer during the applied base motion are most conveniently carried out using Fourier techniques (Reference 2.5D-25 and 2.5D-26).Once the displacements are determined, other response values (e.g., acceleration or shear stresses) are readily obtained.Amendment 65 Page 360 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.17.4 Response Evaluation The response of the 80-ft. layer of dense sand and gravel shown in Figure 2.5D-90, using the artificial accelerogram as input base motion (i.e.., at a depth of 80 ft.), has been evaluated by the three solutions described in the previous section.2.5D.17.4.1 Finite Element Solution The layer is represented by nine elements and twenty nodal points as shown in part b of Figure 2.5D-90. In order to simulate a semi-infinite system, nodal points 1 through 18 are fixed in the vertical direction and, therefore, are only permitted to move in the horizontal direction. Nodal points 19 and 20 are fixed to the base.The damping ratios obtained for each element (based on the induced strain in the element) are shown in Figure 2.5D-91. The maximum acceleration and the maximum shear stresses computed by this solution are presented in Figure 2.5D-92.The fundamental period of the layer computed by this procedure is approximately 0.46 sec.2.5D.17.4.2 Lumped-Mass Solution The layer is divided into nine sublayers, each of which is represented by two masses as shown in part c of Figure 2.5D-90. The weighted damping ratio equal to 0.12 is obtained for the entire layer. The same damping ratio is assumed for all frequencies. The maximum acceleration and the maximum shear stresses computed by this solution are presented in Figure 2.5D-92. The fundamental period of the layer computed by this procedure is approximately 0.46 sec.2.5D.17.4.3 Wave Propagation Solution The layer is represented by nine sublayers as shown in part d of Figure 2.5D-90. The damping ratios obtained for each sublayer are shown in Figure 2.5D-91. The maximum acceleration and the maximum shear stresses computed by this solution are presented in Figure 2.5D-92.The fundamental period of the layer computed by this procedure is approximately 0.52 sec.2.5D.17.4.4 Comparison of Damping Ratios The damping ratios used in the three solutions are shown in Figure 2.5D-91. As can be noted, the finite element and wave propagation solutions which permit the use of different damping ratios throughout the layer, give essentially equal values of damping ratios. Slightly greater damping ratios are obtained in the wave propagation solution. The weighted average damping ratio obtained in the lumped-mass solution is significantly greater than the individual damping ratios in the upper 20 ft. of the layer and somewhat smaller below a depth of 30 ft. It should be noted that the damping ratio obtained for each sublayer in the lumped-mass solution (based on the strain induced in the sublayer) is comparable to the corresponding value obtained in the other two solutions.Therefore, all three solutions provide strain-dependent damping ratios that are essentially equal in each sublayer. The lumped-mass solution, however, requires that one damping ratio be used Amendment 65 Page 361 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 for the entire layer. This results, for this case, in an overestimate of damping in the upper part of the layer and a slight underestimate of the damping in the lower part of the layer.2.5D.17.4.5 Comparison of Accelerations The maximum acceleration computed by the three solutions is shown in Figure 2.5D-92. The three solutions provided acceleration values that are quite comparable throughout the depth of the layer. The highest maximum surface acceleration is obtained by the lumped-mass solution.This value, however, is only approximately 9 percent higher than the minimum value, which is obtained by the wave propagation solution.2.5D.17.4.6 Comparison of Shear Stresses The maximum shear stresses computed by the three solutions are shown in Figure 2.5D-92. As can be noted, the three solutions provide essentially identical values of shear stresses in the upper 20 ft. Below this depth, the stresses computed by the wave propagation solution are slightly lower (approximately 10 percent) than those computed by the other two solutions.2.5D.17.5 Effects of Using Frequency Dependent Damping A direct assessment of the effects of using frequency dependent damping can be made using the lumped-mass procedure. The results shown in Figure 2.5D-92 are obtained by a lumped-mass solution (LM1) with constant damping for all modes. Another lumped-mass solution (LM2) has also been conducted for the 80-ft. layer shown in Figure 2.5-D-90. In solution LM2, the damping for each mode is increased in proportion to the frequency of the mode.For example, the damping ratios obtained in solutions LM1 and LM2 for the first five modes are as follows:Solution LM1 Solution LM2 Mode No. (n) n n n n 1 13.7 0.12 13.7 0.117 2 39.5 0.12 40.1 0.342 3 65.1 0.12 65.9 0.562 4 90.1 0.12 91.2 0.779 5 114.2 0.12 115.6 0.987 For each solution, 16 modes are incorporated in the analysis. The damping ratio for solution LM1 is 0.12 for all modes; for solution LM2, the damping ration for modes 6 to 16 were also proportioned by the ratio 1 n/1 as illustrated in the above listing.The computed maximum accelerations and maximum shear stresses for solutions LM1 and LM2 are presented in Figure 2.5D-93. The maximum surface acceleration computed by solution LM1 is approximately 20 percent greater than that computed by solution LM2. The shear stresses computed by solution LM1 are approximately 10 percent to 20 percent greater than those computed by solution LM2 in the upper 30 ft. of the layer. Below a depth of approximately 60 ft., the stresses computed by LM1 are approximately 5 percent lower.Amendment 65 Page 362 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 It would appear then that the use of frequency dependent damping solution results in an underestimate in the computed maximum surface acceleration and in the shear stresses computed within a shallow depth.2.5D.17.6 Discussion and Conclusion The finite element solution utilized in the response studies of the Category I dams incorporates the use of different damping ratios for each individual element. The solution is not carried out by modal analysis; a direct integration process is used to solve the equations of motions. The use of a Rayleigh expression to formulate the damping submatrix for each element probably introduces somewhat increased damping in the higher modes (all modes of the system are automatically included in this solution).The response values computed were compared with those obtained by two other solutions, ie, the wave propagation solution and the lumped-mass solution. The wave propagation solution, which also permits the use of different damping ratio for each individual sublayer, incorporates a damping term that is independent of frequency. The lumped-mass solution (LM1) also incorporates a damping term that is independent of frequency, but requires that a single damping ratio be used for the entire layer.The comparison of response values for the three solutions (Figure 2.5D-92) indicates that the finite element solution provides values which are slightly higher than those obtained by the wave propagation solution and slightly lower than those obtained by solution LM1. This comparison leads to the conclusion, that any possible increase of damping in the higher modes of the finite element solution does not affect the response significantly.The effects of using increased damping ratios in the higher modes has been investigated directly by the lumped-mass solution. A constant damping ratio is used for all modes in solution LM1, while the damping ratio is increased in the higher modes (based on the ratio of frequencies) in solution LM2. The results for the two solutions are summarized in Figure 2.5D-

93. These results indicate that the use of frequency dependent damping reduces the stresses up to 20% within shallow depths and yields slightly higher stresses at lower depths.

2.5D.18 BEHAVIOR DURING EVENT PRESCRIBED BY REGULATORY GUIDE 1.60 SPECTRA 2.5D.18.1 Introduction The seismic stability of the Main Dam, Auxiliary Dam and Auxiliary Separating Dike were evaluated in 1973 using an input motion an event prescribed by the smooth spectrum and accelerogram shown in Figure 2.5D-7. Details of this evaluation and the conclusions derived are described in Sections 2.5D.1 through 2.5D.8 and Section 2.5D.10 through 2.5D.17. At a meeting with the NRC in Bethesda on June 13, 1977, the staff requested that the behavior of the dams and dikes be assessed; taking into account an event prescribed by Regulatory Guide 1.60 spectra. This section presents the results of this assessment, which is based on direct comparison of the Regulatory Guide 1.60 spectra with the spectra calculated for the accelerogram used as input in 1973.Amendment 65 Page 363 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D.18.2 Procedure Used In lieu of detailed finite element analyses, a simplified procedure was utilized to make this assessment. This procedure consisted of the following steps:a) The spectral values for the accelerogram used as input in 1973 were calculated for 0.07, 0.10 and 0.15 spectral damping ratios and compared directly to the corresponding spectra obtained from Regulatory Guide 1.60 as shown in Figures 2.5D-94 through 2.5D-96. Note that Regulatory Guide 1.60 does not provide values for 0.15 damping; these values were obtained by extrapolation using the procedure suggested by Newmark-Blume-Kapur (Reference 2.5D-24) whereby a linear relationship is established between logarithm of damping and spectral values. The selection of 0.07, 0.10 and 0.15 damping ratio was based on the fact that the damping in the shells of the dam is of the order of 0.07, in the filters of the order of 0.10, and in the core is in the order of 0.15 (Section 2.5D.16).b) The ratio of stresses induced in various portions of the embankments by the events prescribed by Regulatory Guide 1.60 spectra and the accelerogram used as input in 1973 were then estimated by a weighting procedure, using the ratios of spectra derived from Figures 2.5D-94 through 2.5D-96 for various sections of the embankments. Lower range spectral values of the input accelerogram were utilized as a conservative estimate for proportioning the stresses induced in the embankment. For each section and for each case the spectral ratios were selected at the fundamental (first mode) period, TI.This selection was justified because the response of the embankment is controlled mainly by the fundamental mode and to a lesser extent by the second mode of vibration.A comparison of the old and new spectra at the higher modes would indicate a reduction in stress ratios if higher modes are included.c) The stresses induced by the earthquake ground motions were then estimated by using the ratios obtained in step (b). The ratios by which the induced stresses (as calculated in 1973) were multiplied to arrive at an estimate of the stresses induced by an event prescribed by Regulatory Guide 1.60 spectra are shown in Figures 2.5D-97, 2.5D-98, and 2.5D-99 for the cores, filters and shells, respectively.d) New local factors of safety were then calculated using the induced stresses estimated in step (c) and the cyclic strength data presented in Section 2.5D.10 through 2.5D.13.2.5D.18.3 Results 2.5D.18.3.1 Main Dam The minimum local factors of safety calculated in the maximum cross section of the Main Dam (M-105) for the expected constructed material properties are the following:Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 1.79 1.57 1.74 1.53 Amendment 65 Page 364 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Fine Filters 1.73 1.55 1.69 1.51 Coarse Filters 1.22 1.20 1.18 1.16 Rockfill Shell 1.51 1.35 1.45 1.30 where N is the number of cycles of stress applications.The minimum local factors of safety calculated in the Main Dam, 67-ft. and 36-ft. cross sections (cases M-67-IV A and M-36-IV A) are the following:67-ft. Section Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 2.30 2.05 2.37 2.11 Filters 1.59 1.42 1.62 1.45 Rockfill Shell 1.57 1.40 1.60 1.42 36-ft. Section Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 2.48 2.20 2.58 2.29 Filters 1.82 1.62 1.88 1.68 Rockfill Shell 1.60 1.43 1.61 1.44 Other cases of the main dam were also examined. The changes in the minimum local factors of safety for these cases were similar to those outlined above.2.5D.18.3.2 Auxiliary Dam The minimum local factors of safety calculated in the maximum cross section of the Auxiliary Dam (A-63) for the expected constructed material properties are the following:Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 1.43 1.26 1.41 1.24 Filters 1.33 1.19 1.29 1.16 Random Rockfill Shell 1.56 1.38 1.51 1.34 The minimum local factors of safety calculated in the Auxiliary Dam, 44-ft. and 24-ft. cross section (Cases A-44IV A and A-24IV A) are the following:44-ft. Section Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 1.94 1.72 2.00 1.78 Amendment 65 Page 365 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 Filters 1.37 1.22 1.40 1.25 Random Rockfill Shell 1.50 1.40 1.53 1.43 Residual Soils 1.50 1.30 1.53 1.33 24-ft. Section Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 2.25 1.99 2.34 2.07 Filters 1.59 1.42 1.64 1.46 Random Rockfill Shell 1.50 1.40 1.50 1.40 Residual Soils 1.70 1.50 1.70 1.50 Other cases of the Auxiliary Dam were also examined. The changes in the minimum local factors of safety for these cases were similar to those outlined above.2.5D.18.3.3 Auxiliary Separating Dike The minimum local factors of safety calculated in the maximum cross section of the Auxiliary Reservoir Separating Dike for the expected constructed material properties are the following:Estimated for Event Prescribed by 1973 Analyses Regulatory Guide 1.60 Spectra Zone N=5 N = 10 N=5 N = 10 Core 1.7 1.5 1.7 1.5 Random Rockfill Shells 1.9 1.7 1.9 1.7 2.5D.18.4 Conclusions The results of the assessment of the behavior of the Main Dam, Auxiliary Dam and Auxiliary Separating Dike during an event prescribed by Regulatory Guide 1.60 spectra lead to the following conclusions:a) The spectral values of the accelerogram, used as input motion of 19773, are somewhat larger (of the order of 5 to 15 percent) than the spectral values prescribed in Regulatory Guide 1.60 in the period range of approximately 0.1 to 0.2 sec. At a period of approximately 0.4 sec., the Regulatory Guide spectra are somewhat larger (of the order of 10 percent) for the lower damping ratios, but are essentially equal for damping ratios of the order of 0.15.b) The stresses induced in the dams and the dikes during the event prescribed by Regulatory Guide 1.60 spectra are estimated to differ by less than 4 percent from those calculated in 1973.c) The minimum local factors of safety in the dams and dikes estimated during the event prescribed by Regulatory Guide 1.60 spectra are comparable to those calculated in 1973.Amendment 65 Page 366 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 d) The Main Dam, Auxiliary Dam and Auxiliary Reservoir Separating Dike will be stable, will maintain their integrity during an event prescribed by Regulatory Guide 1.60 spectra and will have a peak (zero-period) acceleration of 0.15g.

REFERENCES:

APPENDIX 2.5D 2.5D-1 Barkan, D.D. (1962) Dynamics of Bases and Foundations, McGraw-Hill, New York.2.5D-2 Burmister, D.M. (1962) "Physical, Stress-Strain, and Strength Responses of Granular Soils" ASTM Spec. Tech. Pub. No. 322.2.5D-3 Clough, R.W. (1965) "The Finite Element Method in Structural Mechanics" Stress Analysis, O.C. Zienkiewicz and G.S. Holister, eds., John Wiley & Sons, Ltd., London (1966) Chapter 7.2.5D-4 Clough, R.W. and Chopra, A.A. (1966) "Earthquake Stress Analysis in Earth Dams" JEMD, ASCE, 92, No. EM2, Proc. Paper 4793, April, pp. 197-212.2.5D-5D D'Appolonia, D.J., Whitman, R.V., D' Appolonia, E. (1969) "Sand Compaction with Vibratory Rollers" JSMFD, ASCE, 95, No. SM1, Proc. Paper 6366, January, pp. 263-284.2.5D-6 Duncan, J.M. and Chang, C.Y. (1970) "Nonlinear Analysis of Stress and Strain in Soils" JSMFD, ASCE, 96, No. SM5, Proc. Paper 7613, September, pp. 1629-1653.2.5D-7 Duncan, J.M. (1973) personal communication.2.5D-8 Ebasco (1972) personal communication.2.5D-9 Hardin, B.O. (1970) "Suggested Method of Tests for Shear Modulus and Damping of Soils by the Resonent Column Method" ASTM, Spec. Tech. Pub. 479, pp. 516-529.2.5D-10 Hardin, B.O. and Drnevich, V.P. (1972) "Shear Modulus and Damping in Soils" Design Equations and Curves" JSMFD, ASCE, 98, No. SM7, Proc. Paper 9006, July, pp. 667-692.2.5D-11 Idriss, I.M. (1968) "Finite Element Analysis for the Seismic Response of Earth Banks" JSMFD, ASCE, 94, No. SM3, May.2.5D-12 Idriss, I.M. and Seed, H. Bolton (1967) "Response of Earth Banks During Earthquakes: JSMFD, ASCE, 93, No. SM3, Proc. Paper 5232, May, pp. 61-82.2.5D-13 Idriss, I.M. and Seed, H. Bolton (1968) "Seismic Response of Horizontal Soil Deposits" JSMFD, ASCE, 94, No. SM4, July.2.5D-14 Idriss, I.M., Lysmer, J., Hwang, R., and Seed, H. Bolton (1972) "Computer Programs for Evaluating the Seismic Response of Soil Structures by Variable Damping Finite Elements" Report, Eqk. Engr. Res. Ctr., Univ. of Calif., Berkely, August.Amendment 65 Page 367 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D-15 Kondner, R.L. (1963) "Hyperbolic Stress-Strain Response: Cohesive Soils" JSMFD, ASCE, 89, No. - SM1, Proc. Paper 3429, January, pp. 115.2.5D-16 Kondner, R.L. and Zelasko, J.S. (1963) "A Hyperbolic Stress-Strain Formulation for Sands" Proc. 2nd Pan Am Conf. on Soil Mech. and Fdn. Engr., Vol. 1.2.5D-17 Kulhawy, F.H., Duncan, J.M., and Seed, H. Bolton (1969) "Finite Element Analyses of Stresses and Movements in Embankments During Construction" Report No. TE 69-4, Off of Res. Serv., Univ of Calif. Berkeley.2.5D-18 Lacroix, Yves and Horn, H.M. (1973) "Direct Determination and Indirect Evaluation of Relative Density and Its Use on Earthwork Construction Projects", Evaluation of Relative Density and Its Role in Geotechnical Projects Involving Cohesionless Soils, ASTM STP 523 (in print) 2.5D-19 Lambe, T. William and Whitman, R.V. (1969) Soil Mechanics, John Wiley & Sons, New York.2.5D-20 Lee, K.L. and Seed, H. Bolton (1967a) "Dynamic Strength of Anisotropically Consolidated Sand" JSMFD, ASCE, 93, No. SM5, Proc. Paper 5451, September, pp.169-190.2.5D-21 Lee, K.L. and Seed, H. Bolton (1697b) "Cyclic Stress Conditions Causing Liquefaction of Sand" JSMFD, ASCE, 93, No. , Proc. Paper 5058, January, pp. 47.2.5D-22 Lee, K.L. and Fitton, J.A. (1969) "Factors Affecting the Dynamic Strength of Soil",Vibration Effects of Earthquakes on Soils and Foundation, ASTM, STP 450.2.5D-23 Leonards, G.A. (1962) Foundation Engineering, McGraw-Hill, New York.2.5D-24 Newmark, N.M., Blume, J.A., and Kapur, K.K. (1973) "Seismic Design Spectra for Nuclear Power Plants", Journal of the Power Division, ASCE, November, pp. 287-303.2.5D-25 Roesset, J.M. and Whitman, R.V. (1969) "Theoretical Background for Amplification Studies", Research Report No. R69-15, Soil Pub. No. 231, Dept. of Civil Engr., MIT.2.5D-26 Schanbel, P.B., Seed, H. Bolton, and Lysmer, J. (1971) "Modification of Seismograph Records for Effects of Local Soil Conditions", Eqk. Engr. Res. Ctr.,Report No. EERC 71-8, Univ. of Calif., Berkeley.2.5D-27 Seed, H. Bolton (1966) "A Method for Earthquake-Resistant Design of Earth Dams" JSMFD, ASCE, 92, No. SM1, Proc. Paper 4516, January, pp. 13-41.2.5D-28 Seed, H. Bolton, Lee, K.L., and Idriss, I.M. (1969) "Analysis of Sheffield Dam Failure" JSMFD, ASCE, 95, No. SM6, Proc. Paper 6906, November, pp. 1453-1490.2.5D-29 Seed, H. Bolton and Peaco*ck, W.H. (1970) "Applicability of Laboratory Test Procedures for Measuring Soil Liquefaction Characteristics Under Cyclic Loading" Eqk. Engr. Res. Report No. EERC 70-9, Univ. of Calif., Berkeley.Amendment 65 Page 368 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 2.5D-30 Seed, H. Bolton and Idriss, I.M. (November 1970) "A Simplified Procedure for Evaluating Soil Liquefaction Potential" Eqk. Engr. Res. Ctr. Report No. EERC 70-9, Univ. of Calif., Berkeley.2.5D-31 Seed, H. Bolton and Idriss, J.M. (December 1979) "Soil Moduli and Damping Factors for Dynamic Response Analyses" Eqk. Engr. Res. Ctr. Report No. EERC 70-10, Univ. of Calif., Berkeley.2.5D-32 Seed, H. Bolton (1973) personal communication.2.5D-33 Seed, H. Bolton (1973) personal communication.2.5D-34 Seed, H. Bolton (1973) personal communication.2.5D-35 Seed, H. Bolton, Lee, K.L., Idriss, I.M., and Makdisi, F. (1973) "Analysis of the Slides in the San Fernando Dams During the Earthquake of February 9, 1971" Report, Eqk.Engr. Res. Ctr. Univ. of Calif., Berkeley, March.2.5D-36 Wilson, E.L. (January 1968) "A Computer Program for the Dynamic Stress Analysis of Underground Structures" Strc. Engr. Lab Report No. 68-1, Univ. of Calif.,Berkeley.2.5D-37 Wilson, E.L. and Clough, R.W. (October 1962) "Dynamic Response by Step-by-Step Matrix Analysis" Proc. Symp. on the Use of Computers in Civil Engineering, Lisbon, Portugal.2.5D-38 WMAI (1973a) List of dams on which similar analysis was made.2.5D-39 WMAI (1973b) Files, correction factors for taking into account end effects of strains in triaxial tests.2.5D-40 WMAI (1973c) Seismic wave velocity measurements on rockfill at Amos Dam.2.5D-41 WMAI (1973d) Woodward-Moorhouse & Associates, Inc. Files.2.5D-42 Wong, R.T. (1970) "Deformation Characteristics of Gravels and Gravelly Soils Under Cyclic Loading Conditions" Ph.D Thesis, Univ. of Calif., Berkeley.Amendment 65 Page 369 of 369

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.1.2-1 MINIMUM DISTANCE FROM THE SHNPP TO THE EXCLUSION AREA BOUNDARY FOR EACH MAJOR COMPASS DIRECTION 2.1.3-1 DELETED BY AMENDMENT NO. 59 2.1.3-2 DELETED BY AMENDMENT NO. 59 2.1.3-3 DELETED BY AMENDMENT NO. 59 2.1.3-4 DELETED BY AMENDMENT NO. 59 2.1.3-5 DELETED BY AMENDMENT NO. 59 2.1.3-6

SUMMARY

OF TOTAL POPULATION DEMAND WITHIN THE 10-MILE EPZ 2.2.2-1 MINING AND QUARRIES WITHIN A TEN MILE RADIUS OF THE SHEARON HARRIS NUCLEAR POWER PLANT 2.2.2-2 AIRCRAFT OPERATIONS - RALEIGH DURHAM AIRPORT 2.2.3-1 DIMENSIONS OF PROPANE PLUME DOWNWIND OF 324 FT.3/SEC. SOURCE, STABILITY CATEGORY F, WIND SPEED 3.3 FT./SEC.2.2.3-2 DIMENSIONS OF PROPANE PLUME DOWNWIND OF 1000 FT.3/SEC. SOURCE, STABILITY CATEGORY F, WIND SPEED 3.3 FT./SEC 2.2.3-3 BLAST AND SEISMIC PARAMETERS FOR SHOCK WAVES FROM PROPANE DETONATIONS 2.2.3-4 HAZARDOUS MATERIALS MOST FREQUENTLY SHIPPED BY RAIL 2.2.3-5 EFFECTS OF INCREASE IN PIPELINE SIZE FROM SIX INCHES TO EIGHT INCHES 2.3.1-1 STATIONS REFERENCED FOR REGIONAL CLIMATOLOGY AND LOCAL METEOROLOGY 2.3.1-2 SITE AREA NUMBER OF THUNDERSTORM DAYS 2.3.1-3 NUMBER OF CLOUD-TO-GROUND FLASHES BY SEASON (PER SQ. KM.)2.3.1-4 EXTREME WINDS AND PRECIPITATION ASSOCIATED WITH HURRICANES RALEIGH-DURHAM AIRPORT (1950-1978) 2.3.1-5 SITE REGION METEOROLOGICAL EXTREMES (MONTH/YEAR OF OCCURRENCE)[DATA PERIOD]2.3.1-6 MEAN MONTHLY MAXIMUM MIXING DEPTHS (METERS ABOVE SURFACE) 2.3.1-7 FREQUENCY OF INVERSIONS BASED BELOW 500 FEET 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM) 2.3.2-2 WIND DIRECTION PERSISTENCE DATA HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS A)Amendment 65 Page 1 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.3.2-3 EXTREME WINDS AND PRECIPITATION ASSOCIATED WITH HURRICANES RALEIGH-DURHAM AIRPORT (1950-1978) 2.3.2-4 RALEIGH-DURHAM NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-5 GREENSBORO NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-6 CHARLOTTE NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-7 MONCURE NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-8 PINEHURST NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-9 ASHEBORO NORMAL PRECIPITATION (IN.) AND TEMPERATURE (F) 2.3.2-10 HARRIS ON-SITE DATA MEAN TEMPERATURE (JANUARY 1976 - DECEMBER 1978) 2.3.2-11 HARRIS ON-SITE DATA MAXIMUM - MINIMUM TEMPERATURES (JANUARY 1976 - DECEMBER 1978) 2.3.2-12 SITE REGION METEOROLOGICAL EXTREMES (MONTH/YEAR OF OCCURRENCE) [DATA PERIOD}2.3.2-13 DEWPOINT TEMPERATURE (F)AND ABSOLUTE HUMIDITY (g/m3) 2.3.2-14 HARRIS ONSITE DATA DEWPOINT TEMPERATURE (10 METER LEVEL) 2.3.2-15 CHARLOTTE RELATIVE HUMIDITY (PERCENT) 2.3.2-16 GREENSBORO RELATIVE HUMIDITY (PERCENT) 2.3.2-17 RALEIGH-DURHAM RELATIVE HUMDITY (PERCENT) 2.3.2-18 HARRIS ONSITE DATA PRECIPITATION (in.) (JANUARY 1976 - DECEMBER 1978) 2.3.2-19 SHNPP ONSITE EXTREME RAINFALL RATES 2.3.2-20 PRECIPITATION FREQUENCIES AND AMOUNTS (1951 - 1960: JANUARY, APRIL, JULY, OCTOBER) 2.3.2-21 SHNPP ONSITE HOURLY PRECIPITATION OCCURRENCE 2.3.2-22 SHNPP ONSITE FREQUENCY OF PASQUILL STABILITY CATEGORIES (PERCENT) 2.3.2-23 STABILITY CLASS PERSISTENCE DATA HARRIS 2.3.2-24 SEASONAL FREQUENCY OF PLUME LENGTHS (HOURS PER YEAR) 2.3.2-25 JOINT OCCURRENCE FREQUENCIES FOR 10M WIND DIRECTION AND 10M WIND SPEED BY PRECIPITATION RATE RANGES, INCLUDE LOWER END POINT, EXCLUDE UPPER END POINT 2.3.2-26 JOINT OCCURRENCE FREQUENCIES FOR 60M WIND DIRECTION AND 60M WIND SPEED BY PRECIPITATION RATE RANGES, INCLUDE LOWER END POINT, EXCLUDE UPPER END POINT 2.3.3-1 DELETED BY AMENDMENT NO. 51 Amendment 65 Page 2 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.3.3-2 DELETED BY AMENDMENT NO. 51 2.3.3-3 SHNPP OPERATIONAL SENSOR ELEVATIONS 2.3.3-4 CHANNEL ACCURACY 2.3.3-5 DELETED BY AMENDMENT NO. 51 2.3.3-5A DELETED BY AMENDMENT NO. 51 2.3.3-5B DELETED BY AMENDMENT NO. 51 2.3.3-6 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 4:00 PM 1/14/76 TO 11:00 PM 12/31/76 UPPER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-7 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 4:00 PM 1/14/76 TO 11:00 PM 12/31/76 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-8 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 1/01/77 TO 11:00 PM 12/31/77 UPPER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-9 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 1/1/77 TO 11:00 PM 12/31/77 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-10 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 1/1/78 TO 11:00 PM 12/31/78 UPPER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-11 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 1/1/78 TO 11:00 PM 12/31/78 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-12 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 4:00 PM 1/14/76 TO 11:00 PM 12/31/78 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-13 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 4:00 PM 1/14/76 TO 11:00 PM 12/31/78 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFF. TEMPERATURE HARRIS ON-SITE METEOROLOGICAL FACILITY 2.3.3-14 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 2/1/79 TO 11:00 PM 1/31/80 UPPER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFFERENTIAL TEMPERATURE H DIGAMET ON-SITE METEOROLOGICAL FACILITY 2.3.3-15 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND SPEED FOR THE PERIOD 12:00 AM 2/1/79 TO 11:00 PM 1/31/80 LOWER WIND LEVEL STABILITY CLASS A STABILITY CALCULATED FROM DIFFERENTIAL TEMPERATURE H DIGAMET ON-SITE METEOROLOGICAL FACILITY 2.3.3-16 JOINT OCCURRENCE FREQUENCIES FOR 10M WIND DIRECTION AND 10M WIND SPEED RANTES INCLUDE LOWER END POINT, EXCLUDE UPPER END POINT Amendment 65 Page 3 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.3.3-17 JOINT OCCURRENCE FREQUENCIES FOR 60M WIND DIRECTION AND 60M WIND SPEED RANGES INCLUDE LOWER END PINT, EXCLUDE UPPER END POINT 2.3.3-18 WIND INSTRUMENT HEIGHT 2.3.4-1 WIND VELOCITY GROUPINGS FROM THE JOINT WIND FREQUENCY DISTRIBUTION AND THE CORRESPONDING COMPUTATIONAL WIND VELOCITY 2.3.4-2 SHNPP SITE BOUNDARY DISTANCES (METERS) 2.3.4-3 SHNPP EXCLUSION AREA BOUNDARY DISTANCES (METERS) 2.3.4-4

SUMMARY

OF 0-2 HOUR SHNPP 5 AND 50 PERCENTILE VALUES BY DIRECTION OF SHORT TERM DIFFUSION ESTIMATES AT THE MIN EXCLUSION BOUNDARY AREA 2.3.4-5 SHNPP WORST 5 AND 50 PERCENTILE OF CUMULATIVE FREQUENCY DISTRIBUTION OF x/Q VALUES (SEC/m3) BASED ON T (59.85m - 11.03m) STABILITY DATA AND 12.46m WIND VELOCITY DATA (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-1 ANNUAL AVERAGE DILUTION FACTORS AT THE SHNPP MINIMUM EXCLUSION BOUNDARY (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-2 ANNUAL AVERAGE DILUTION FACTORS AT THE SHNPP ACTUAL SITE BOUNDARIES (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-3 ANNUAL AVERAGE DILUTION FACTORS AT THE SHNPP LOW POPULATION ZONE BOUNDARIES (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-4 ANNUAL AVERAGE DILUTION FACTORS FOR INCREMENTAL DISTANCES AT SHNPP (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-5 ANNUAL AVERAGE DILUTION FACTORS FOR INCREMENTAL DISTANCES AT SHNPP (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-6 ANNUAL AVERAGE DILUTION FACTORS FOR INCREMENTAL DISTANCES AT SHNPP (JANUARY 1976 THROUGH DECEMBER 1978) 2.3.5-7 ANNUAL AVERAGE DEPOSITION FACTORS FOR INCREMENTAL DISTANCES AT SHNPP (JANUARY 1976 THROUGH DECEMBER 1978) 2.4.1-1 ESTIMATED MONTHLY AVERAGE FLOWS OF BUCKHORN CREEK (CFS) (AVERAGE 1924 - 1981 =87.2) DRAINAGE AREA = 79.5 sq. mi.2.4.1-2 COMPARISON OF MONTHLY AVERAGE FLOW BETWEEN ESTIMATED AND ACTUAL FLOW OF BUCKHORN CREEK 2.4.1-3 ESTIMATED MONTHLY AVERAGE FLOW IN CAPE FEAR RIVER AT BUCKHORN DAM IN CFS DRAINAGE AREA = 3196 sq. mi.2.4.1-4 DEVELOPMENT OF WATER RESOURCES FOR CAPE FEAR RIVER BASIN 2.4.1-5 CAPE FEAR RIVER INDUSTRIAL WATER WITHDRAWALS DOWNSTREAM OF BUCKHORN DAM 2.4.1-6 CAPE FEAR RIVER MUNICIPAL WATER WITHDRAWALS DOWNSTREAM OF BUCKHORN DAM Amendment 65 Page 4 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.4.2-1 ESTIMATED MAXIMUM FLOOD PEAKS FOR BUCKHORN CREEK AT THE CAPE FEAR RIVER DRAINAGE AREA = 79.5 sq. mi.2.4.2-2 ESTIMATED AND MEASURED MAXIMUM FLOOD PEAKS FOR BUCKHORN CREEK AT USGS GAGE STATION NEAR CORINTH, N.C. (D.A. = 74.2 sq. mi.)2.4.2-3 MAXIMUM FLOOD FLOW OF THE CAPE FEAR RIVER AT BUCKHORN DAM 2.4.2-4 PLANT AREA WATER ACCUMULATION FOR DESIGN PMP CONDITIONS 2.4.3-1 PROBABLY MAXIMUM PRECIPITATION 2.4.3-2 TIME DISTRIBUTION OF PROBABLY MAXIMUM PRECIPITATION 2.4.3-3 SNYDER AND LOSS PARAMETERS BUCKHORN CREEK BASIN 2.4.5-1 WAVE RUNUP PARAMETERS FOR STRUCTURES PROTECTED BY RIPRAP 2.4.5-2 WAVE RUNUP PARAMETERS FOR PLANT ISLAND 2.4.11-1 DELETED BY AMENDMENT NO. 15 2.4.11-1A DELETED BY AMENDMENT NO. 15 2.4.11-2 NORMAL MONTHLY METEOROLOGICAL CONDITIONS AT SITE 2.4.11-3 WORST MONTHLY COINCIDENT METEOROLOGICAL CONDITIONS FOR PERIOD OF RECORD (1931 - 1970) FOR DETERMINATION OF CRITICAL MONTHLY EVAPORATION RATES 2.4.11-4 DELETED BY AMENDMENT NO. 15 2.4.11-5 DELETED BY AMENDMENT NO. 15 2.4.11-6 DELETED BY AMENDMENT NO. 15 2.4.11-7 DELETED BY AMENDMENT NO. 15 2.4.11-8 DELETED BY AMENDMENT NO. 15 2.4.11-9 DELETED BY AMENDMENT NO. 15 2.4.11-10 ALLOCATION OF MAIN RESERVOIR VOLUME 2.4.11-11 DELETED BY AMENDMENT NO. 15 2.4.11-12 CALCULATED MINIM FLOW OF BUCKHORN CREEK AT THE MAIN DAM (DA - 71.0 sq. mi.)2.4.11-13 DELETED BY AMENDMENT NO. 15 2.4.11-14 DELETED BY AMENDMENT NO. 46 2.4.11-15 DELETED BY AMENDMENT NO. 46 Amendment 65 Page 5 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.4.11-16 DELETED BY AMENDMENT NO. 15 2.4.11-17 DELETED BY AMENDMENT NO. 15 2.4.11-18 RESERVOIR ANALYSIS NORMAL OPERATION - CRITICAL PERIOD MAY 1980 - MAY 1982 2.4.11-19 RESERVOIR ANALYSIS NORMAL OPERATION - 100-YEAR DROUGHT 2.4.11-20 MONTHLY MAIN RESERVOIR RELEASE- AVERAGE AND 100-YEAR DROUGHT CONDITIONS 2.4.12-1 CONCENTRATION AND C/MPC IN THE CAPE FEAR RIVER WATER AT LILLINGTON FOLLOWING A TANK RUPTURE*2.4.13-1 PUBLIC WELLS WITHIN A 10-MILE RADIUS OF THE PLANT (AS REGISTERED WITH N.C. DIVISION OF ENVIRONMENTAL MANAGEMENT) 2.4.13-2 LOCATION OF SITE WELLS AND PIEZOMETERS 2.4.13-3 GROUNDWATER CONSUMPTION IN THOUSANDS OF GALLONS 2.4.13-4 ESTIMATED GROUNDWATER USE 2.4.13-5 GROUNDWATER LEVELS IN SITE WELLS 2.4.13-6 GROUNDWATER LEVELS IN SITE PIEZOMETERS 2.4.13-7 PERMEABILITY OF PLANT SITE MATERIALS BASED ON DOWN-HOLE PRESSURE TESTS 2.4.13-8 CHEMICAL QUALITY OF SITE GROUNDWATER 2.5.2-1 EARTHQUAKE LIST (SITE LOCATION 35.7N 79.0W) 2.5.2-2 COMPARISON OF SELECTED RESERVOIRS - DEPTHS AND VOLUMES 2.5.2-3 RESULTS OF SEISMIC REFRACTION SURVEY 2.5.2-4 AMBIENT VIBRATION MEASUREMENTS 2.5.4-1 RESULTS OF COMPRESSIONAL WAVE VELOCITY AND DENSITY OF ROCK MATERIAL - PLANT SITE 2.5.4-2 ROCK QUALITY DESIGNATION (ROD) PLANT FOUNDATION AREA 2.5.4-3 YOUNGS MODULUS (Et) AND ULTIMATE STRESS (qu) - ROCK MATERIAL 2.5.4-4 TRIAXIAL COMPRESSION TEST RESULTS 2.3.4-5 POISSONS RATIO () FOR ROCK MATERIAL (BASED ON UNCONFINED COMPRESSION TEST RESULTS) 2.5.4-6 SHOCKSCOPE TEST RESULTS 2.5.4-7 PIEZOMETRIC LEVEL FLUCTUATIONS AT THE SITE Amendment 65 Page 6 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.5.4-8 SPRING CONSTANTS FORMULAE FOR FOUNDATIONS 2.5.4-9 AVERAGE BUILDING SETTLEMENTS 2.5.6-1 MAIN DAM - MINIMUM FACTORS OF SAFETY 2.5.6-2 AUXILIARY DAM - MINIMUM FACTORS OF SAFETY 2.5.6-3 AUXILIARY SEPARATING DIKE - MINIMUM FACTORS OF SAFETY 2.5B-1 MAIN DAM AND SPILLWAY TESTING PROGRAM 2.5B-2 AUXILIARY DAM AND SEPARATING DIKE TESTING PROGRAM 2.5B-3 RESULTS OF COMPRESSIONAL WAVE VELOCITY AND DENSITY TESTS - MAIN DAM AND SPILLWAY 2.5B-4 RESULTS OF COMPRESSIONAL WAVE VELOCITY AND DENSITY TESTS - AUXILIARY DAM SITE 2.5C.2-1 MATERIAL M - RESULTS OF ISOTROPICALLY-CONSOLIDATED DRAINED (CID) TRIAXIAL TESTS 2.5C.2-2 MATERIAL M - RESULTS OF CYCLIC TRIAXIAL STRESS CONTROL TESTS 2.5C.2-3 MATERIAL M - RESULTS OF STRAIN-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5C.2-4 MATERIAL M - RESULTS OF CYCLIC TORSION TESTS 2.5C.2-5 MATERIAL Z - RESULTS OF ISOTROPICALLY-CONSOLIDATED DRAINED (CID) TRIAXIAL TESTS 2.5C.2-6 MATERIAL Z - RESULTS OF CYCLIC TRIAXIAL STRESS CONTROL TESTS 2.5C.2-7 MATERIAL Z - RESULTS OF STRAIN-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5C.2-8 MATERIAL Z - RESULTS OF CYCLIC TORSION TESTS 2.5C.2-9 AUXILIARY DAM AND DIKE - UNDISTURBED FOUNDATION SOILS - RESULTS OF CYCLIC TRIAXIAL (STRESS-CONTROL) TESTS 2.5C.2-10 AUXILIARY DAM,

SUMMARY

OF MEASUREMENTS IN RESIDUAL SOIL 2.5C.2-11 AUXILIARY DAM,

SUMMARY

OF MEASUREMENTS IN TRANSITIONAL MATERIAL 2.5C.2-12 AUXILIARY DAM,

SUMMARY

OF MEASUREMENTS IN WEATHERED ROCK 2.5C.2-13 AUXILIARY DIKE,

SUMMARY

OF MEASUREMENTS 2.5C.2-14 AUXILIARY DAM AND AUXILIARY DIKE, K2,MAX, BASED ON SHEAR WAVE VELOCITIES 2.5C.2-15 MAIN DAM, EXPECTED CONSTRUCTED MATERIAL PROPERTIES FOR DYNAMIC ANALYSIS 2.5C.2-16 AUXILIARY DAM AND AUXILIARY RESERVOIR SEPARATING DIKE, EXPECTED CONSTRUCTED MATERIAL PROPERTIES FOR DYNAMIC ANALYSES Amendment 65 Page 7 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.5C.2-17 MATERIAL M, SHEAR MODULUS AND DAMPING RATIO BASED ON CYCLIC TORSION TESTS 2.5C.2-18 MATERIAL M, SHEAR MODULUS AND DAMPING RATIO BASED ON STRAIN-CONTROLLED CYCLIC TRIAXIAL TEST RESULTS 2.5C.2-19 MATERIAL M, TYPICAL SHEAR MODULUS AND DAMPING RATIO BASED ON STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5C.2-20 MATERIAL Z, SHEAR MODULUS AND DAMPING RATIO BASED ON CYCLIC TORSION TESTS 2.5C.2-21 MATERIAL Z, SHEAR MODULUS AND DAMPING RATIO BASED ON STRAIN-CONTROLLED CYCLIC TRIAXIAL TEST RESULTS 2.5C.2-22 MATERIAL Z, TYPICAL SHEAR MODULUS AND DAMPING RATIO BASED ON STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5C.2-23 MAIN DAM, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES 2.5C.2-24 AUXILIARY DAM, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES - 1 2.5C.2-25 AUXILIARY DAM, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES - 2 2.5C.2-26 AUXILIARY DIKE, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES 2.5C.3-1 BORROW AREA Y SHRINKAGE FACTORS 2.5C.3-2 BORROW AREA Z SHRINKAGE FACTORS 2.5C.3-3 MAIN DAM BORROW AREA M SHRINKAGE FACTORS 2.5D-1 COMPUTED PREDOMINANT PERIODS AND MAXIMUM CREST ACCELERATIONS 2.5D-2 MAIN DAM, MAXIMUM CROSS SECTION , MINIMUM VALUES OF f/d IN CORE 2.5D-3 MAIN DAM, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN UPSTREAM FINE FILTER 2.5D-4 MAIN DAM, MAXIMUM CROSS SECTION MINIMUM VALUES OF f/d IN UPSTREAM COARSE FILTER 2.5D-5 MAIN DAM, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN UPSTREAM ROCKFILL SHELL 2.5D-6 AUXILIARY DAM, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN CORE 2.5D-7 AUXILIARY DAM, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN FILTER 2.5D-8 AUXILIARY DAM, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN RANDOM ROCKFILL 2.5D-9 AUXILIARY RESERVOIR SEPARATING DIKE, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d IN CORE 2.5D-10 AUXILIARY SEPARATING DIKE, MAXIMUM CROSS SECTION, MINIMUM VALUES OF f/d, IN RANDOM ROCKFILL Amendment 65 Page 8 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.5D-11 MAIN DAM, MAXIMUM CROSS SECTION MINIMUM VALUES OF f/d IN DIFFERENT ZONES, CASE M-105-IVA 2.5D-12 AUXILIARY DAM, MAXIMUM CROSS SECTION MINIMUM VALUES OF f/d IN DIFFERENT ZONES CASE A-63-IVA 2.5D-13 BORROW AREA M, INDEX PROPERTIES OF SOILS IN REPRESENTATIVE BORINGS AND COMPOSITE SAMPLES FROM TEST PITS 2.5D-14 PHYSICAL PROPERTIES OF MATERIAL M 2.5D-15 MATERIAL M - STATIC STRESS-STRAIN PARAMETERS 2.5D-17 MATERIAL M, RESULTS OF STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS ON ANISOTROPICALLY CONSOLIDATED SPECIMENS 2.5D-18 MATERIAL M, SHEAR MODULUS AND DAMPING RATIO BASED ON STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5D-19 MATERIAL M, SHEAR MODULUS AND DAMPING RATIO BASED ON STRAIN-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5D-20 MATERIAL M, SHEAR MODULUS AND DAMPING RATIO BASED ON CYCLIC TORSION TESTS 2.5D-21 BORROW AREA Z, INDEX PROPERTIES OF SOILS IN REPRESENTATIVE BORINGS AND COMPOSITE SAMPLES FROM TEST PITS 2.5D-22 PHYSICAL PROPERTIES OF MATERIAL Z 2.5D-23 MATERIAL Z - STATIC STRESS-STRAIN PARAMETERS 2.5D-24 MATERIAL Z, RESULTS OF STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS IN ISOTROPICALLY CONSOLIDATED SPECIMENS 2.5D-25 MATERIAL Z, RESULTS OF STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS ON ANISOTROPICALLY CONSOLIDATED SPECIMENS 2.5D-26 MATERIAL Z, SHEAR MODULUS AND DAMPING RATIO BASED ON STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5D-27 MATERIAL Z, SHEAR MODULUS AND DAMPING RATIO BASED ON STRAIN-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5D-28 MATERIAL Z, SHEAR MODULUS AND DAMPING RATIO BASED ON CYCLIC TORSION TESTS 2.5D-29 MATERIAL Z, RESULTS OF CYCLIC TORSION TESTS AT 100% STANDARD COMPACTION 2.5D-30 FILTER MATERIALS, STATIC PROPERTIES 2.5D-31 MAIN DAM, AUXILIARY DAM, DYNAMIC PROPERTIES FOR FILTERS 2.5D-32 MAIN DAM, AUXILIARY DAM, AUXILIARY RESERVOIR SEPARATING DIKE, STATIC PROPERTIES FOR ROCKFILLS Amendment 65 Page 9 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE TITLE 2.5D-33 MAIN DAM, AUXILIARY DAM, AUXILIARY RESERVOIR SEPARATING DIKE, DYNAMIC PROPERTIES FOR ROCKFILL MATERIAL 2.5D-34 MAIN DAM, STATIC AND DYNAMIC PROPERTIES OF WEATHERED ROCK 2.5D-35 AUXILIARY DAM, STATIC AND DYNAMIC PROPERTIES OF WEATHERED ROCK 2.5D-36 AUXILIARY DAM, STATIC AND DYNAMIC PROPERTIES OF IN-SITU RESIDUAL SOIL 2.5D-37 AUXILIARY DAM AND AUXILIARY RESERVOIR SEPARATING DIKE, IN SITU RESIDUAL SOIL K2 MAX BASED ON SHEAR-WAVE VELOCITY MEASUREMENTS 2.5D-38 AUXILIARY DAM, IN-SITU RESIDUAL SOILS, RESULTS OF STRESS-CONTROLLED CYCLIC TRIAXIAL TESTS 2.5D-39 AUXILIARY RESERVOIR SEPARATING DIKE, STATIC AND DYNAMIC PROPERTIES OF IN-SITU RESIDUAL SOIL 2.5D-40 MAIN DAM, MATERIAL PROPERTIES FOR STATIC STRESS ANALYSES 2.5D-41 MAIN DAM, MATERIAL PROPERTY COMBINATIONS FOR STATIC STRESS ANALYSES 2.5D-42 AUXILIARY DAM, MATERIAL PROPERTIES FOR STATIC STRESS ANALYSES 2.5D-43 AUXILIARY DAM, MAXIMUM SECTION, MATERIAL PROPERTY COMBINATIONS FOR STATIC STRESS ANALYSES 2.5D-44 AUXILIARY RESERVOIR SEPARATING DIKE, MATERIAL PROPERTIES FOR STATIC STRESS ANALYSES 2.5D-45 AUXILIARY RESERVOIR SEPARATING DIKE, MATERIAL PROPERTY COMBINATIONS FOR STATIC STRESS ANALYSES 2.5D-46 MAIN DAM, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES 2.5D-47 AUXILIARY DAM, MAXIMUM SECTION MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES 2.5D-48 AUXILIARY DAM, CROSS SECTION A-44 MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES 2.5D-49 AUXILIARY RESERVOIR SEPARATING DIKE, MATERIAL PROPERTY COMBINATIONS FOR PARAMETRIC STUDIES Amendment 65 Page 10 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.1.2-1 MINIMUM DISTANCE FROM THE SHNPP TO THE EXCLUSION AREA BOUNDARY FOR EACH MAJOR COMPASS DIRECTION Sector Distance (ft.)N 6980 NNE 7000 NE 7000 ENE 7000 E 7000 ESE 7000 SE 7000 SSE 7000 S 7200 SSW 7000 SW 7000 WSW 7000 W 7000 WNW 7000 NW 6660 NNW 6640 NOTE: Distances measured from center point of the originally planned four Units.Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.1.3-6

SUMMARY

OF TOTAL POPULATION DEMAND WITHIN THE 10-MILE EPZ Transit- Special Shadow External Sub-Zone Residents Dependent Transients Employees Facilities Schools Population Traffic Total A 134 4 401 519 44 0 0 0 1,102 B 1,257 42 289 0 0 0 0 0 1,588 C 2,086 69 70 0 3 0 0 0 2,228 D 346 11 224 0 0 0 0 0 581 E 45,269 1,504 1,230 1,228 261 8,889 0 0 58,376 F 22,342 743 703 789 44 7,936 0 0 32,552 G 21,463 713 824 582 407 5,002 0 0 28,991 H 3,868 128 80 0 0 0 0 0 4,076 I 963 32 0 0 0 0 0 0 995 J 1,126 37 0 57 137 0 0 0 1,357 K 688 23 440 247 0 0 0 0 1,398 L 815 27 2,767 45 0 0 0 0 3,654 M 1,753 58 2,306 0 0 285 0 0 4,402 N 851 28 2,108 0 0 0 0 0 2,987 1Shadow 0 0 0 0 0 1418 39,618 0 41,036 Total 102,961 3,419 11,442 3,467 896 23,530 39,618 0 185,323 NOTES:

1) Shadow Population has been reduced to 20%.

2) Special Facilities only includes medical facilities.

1 There are two schools in the Shadow that evacuate - Deep River Elementary School in Lee County and Lafayette Elementary School in Harnett County.County emergency plans call for these facilities to be evacuated because of their close proximity to the EPZ boundary.Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.2-1 MINING AND QUARRIES WITHIN A TEN MILE RADIUS OF THE SHEARON HARRIS NUCLEAR POWER PLANT Mines and Quarries (Figure 2.2.2-1) County Products A. Buckhorn Quarry Wake Inactive B. Holly Springs Quarry Wake Inactive C. Cherokee Brick Chatham Inactive D. Cherokee Brick Chatham Inactive E. Cherokee Brick Chatham Inactive F. Moncure Quarry Lee Crushed Stone G. Martin Marietta Aggregate Wake Granite Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.2-2 AIRCRAFT OPERATIONS - RALEIGH-DURHAM AIRPORT Landings Per Year Air Carrier General Aviation Air Taxi Military Total Actual 1976 30,826 147,861 9,365 9,568 197,620 1977 33,306 152,229 11,462 9,059 206,056 1978 34,145 154,476 13,153 7,470 209,244 1979 39,929 146,203 14,889 6,720 207,741 1980 40,225 130,079 24,382 7,487 202,173 1985 55,648 111,138 31,299 10,609 208,694 1990 124,113 83,041 67,113 8,683 283,055 1995 90,976 69,007 38,865 6,041 204,889 2000 152,817 67,325 71,434 5,103 296,679 Projected 2005 92,100 71,300 113,800 5,000 282,200 2010 113,200 77,600 134,300 5,000 330,100 2025 200,800 99,900 232,600 5,000 538,300 Source: Raleigh-Durham Airport Authority Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.3-1 DIMENSIONS OF PROPANE PLUME DOWNWIND OF 324 FT.3/SEC. SOURCE, STABILITY CATEGORY F, WIND SPEED 3.3 FT./SEC.Downwind Plume Dispersion Centerline Horizontal Distance From Plume Axis, ft. Vertical Distance From Plume Axis, ft.Distance (std deviation) ft. Concentration (ft.) Percent To Rich Limit To Lean Limit To Rich Limit To Lean Limit( 0.5 )yr y1 zr z1 250 11 6 47.5 21.6 26.2 11.6 14.3 500 20 10 15.6 25.4 37.2 12.7 18.6 750 27 14 8.8 15.2 39.8 7.85 20.6 830 30 15 7.0 0 40.5 0 20.2 920 33.6 16.8 5.55 -- 39.6 -- 19.9 1000 36 18 4.8 -- 36.6 -- 18.3 1425 47.5 23.8 2.8 -- 0 -- 0 1640 53.4 26.7 2.21 -- -- -- --3000 92 46 0.74 -- -- -- --Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.3-2 DIMENSIONS OF PROPANE PLUME DOWNWIND OF 1000 FT.3/SEC. SOURCE, STABILITY CATEGORY F, WIND SPEED 3.3 FT./SEC.Downwind Plume Dispersion Centerline Horizontal Distance From Plume Axis, ft. Vertical Distance From Plume Axis, ft.Distance (std deviation) ft. Concentration (ft.) Percent To Rich Limit To Lean Limit To Rich Limit To Lean Limit( 0.2 )yr y1 zr z1 750 27 5.4 66 47.2 68 9.44 13.6 1000 36 7.2 37.2 65.6 81.8 13.1 16.36 2000 72 14.4 9.34 54.6 112.0 10.92 22.4 2500 81 16.2 7.35 25.7 112.5 5.14 22.5 2600 83 16.75 7.0 0 112.5 0 22.5 3000 92 18.4 5.70 -- 109.5 -- 21.9 3500 104 20.8 4.50 -- 102.0 -- 20.4 4000 116 23.2 3.60 -- 82.5 -- 16.5 4750 131 26.2 2.80 -- 0 -- 0 5000 136 27.2 2.60 -- -- -- --Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.3-3 BLAST AND SEISMIC PARAMETERS FOR SHOCK WAVES FROM PROPANE DETONATIONS Yield (Ton of Distance From Peak Over- Peak Dynamic Peak Reflected Positive Phase Peak Acceleration Peak

• Peak TNT) Center of pressure PSI Pressure psi Pressure psi Duration (msec) Horizontal or Velocity (f/s) Displacement Detonation Vertical* (gs) (in.)

8.9 7500 0.10 .00024 0.20 209 .002 .0027 .0001 100 5000 0.5 .006 1.01 292 .023 .0306 .0025 119 2200 1.2 .034 2.48 310 .133 .177 .0157

• Based on a conservatively assumed seismic velocity of 1000 ft./sec. and a specific gravity of 1.5.

( )The peak dynamic pressure, Pd, is calculated from Pd =( )( )and the peak reflected pressure, Pr, is calculated from Pr = 2 ( )where p is the peak overpressure and Po the ambient pressure.Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.3-4 HAZARDOUS MATERIALS MOST FREQUENTLY SHIPPED BY RAIL(1)Vapor Pressure Toxicity Hazard Rating (2)Material Boiling Point (°C) Comments (mm Hg/°C) Acute Local: Inhalation Ammonia (Anhydrous) -33.35 10/25.7 3 Gaseous at atmospheric conditions and very toxic. Considered in control room habitability evaluation.Caustic Soda (Sodium Hydroxide) 1390 1/739 2 Not a threat to the control room ventilation system. (3)Liquid Propane Gas -42.1 - 1 Not a threat considering the toxicity and the location of the railroad (1.9 miles).Sulfuric Acid 330 1/145.8 2 Not a threat to the control room ventilation system. (3)Chlorine -34.5 4800/20 3, 4 Not a threat due to the probability of occurrence.Propane -42.1 760/-42.1 1 Not a threat considering the toxicity and the location of the railroad (1.9 miles).Ammonium Nitrate - 11/210 1 Not a threat to the control room ventilation system. (3)Gasoline (Aliphatic Hydrocarbons) 150-190 - 2 Not a threat to the control room ventilation system. (3)Phosphoric Acid - 0.0285/20 2 Not a threat to the control room ventilation system.Crude Oil (Mixture of hydrocarbons) 1 Not a threat. Low inhalation toxicity index.Methanol (Methyl Alcohol) 64.8 100/21.1 2 Not a threat considering the toxicity and the location of the railroad (1.9 miles).Petroleum Distillate (Mineral Spirits) 150-190 - 2 Not a threat considering the toxicity and the location of the railroad (1.9 miles)Vinyl Chloride -13.4 2600/25 3 Considered in the control room habitability evaluation.Butane -0.5 1520/18.8 2 Not a threat considering the toxicity and the location of the railroad (1.9 miles).Motor Fuel Anti-Knock Compound 198-202 1/38.4 3 Not a threat to the control room ventilation system. (3)(Lead Tetraethyl)Butadiene (Erythrene) -4.5 1840/21 2 Not a threat considering the toxicity and the location of the railroad (1.9 miles).Petroleum Naphtha (Petroleum Spirits) 40-80 - 2 Not a threat considering the toxicity and the location of the railroad (1.9 miles).(1) An Evaluation of Railroad Safety, Office of Technology Assessment, Congress of the United States, 1978.(2) Toxicity Hazard Rating Code:0 NONE: (a) No harm under any conditions; (b) Harmful only under unusual conditions or overwhelming dosage.1 SLIGHT: Causes readily reversible changes which disappear after end of exposure.2 MODERATE: May involve both irreversible and reversible changes not severe enough to cause death or permanent injury.3 HIGH: May cause death or permanent injury after very short exposure to small quantities.U UNKNOWN: No information on humans considered valid by authors.(3) Regulatory Guide 1.78 required (C.5.a) that consideration be given to those chemicals that are not gases at 100°F and normal atmospheric pressure but are liquids with vapor pressures in excess of 10 torr (10mm Hg at 0°C). The liquids excepted per this note do not fall in this category. They are not expected to be in the gaseous forms in quantities that can pose a threat.Amendment 65 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 (4) Reference SHNPP License Amendment No. 10 dated May 3, 1989 and Technical Specification change request, NLS-88-281, dated January 4, 1989, Chlorine Detection System.Amendment 65 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.2.3-5 EFFECTS OF INCREASE IN PIPELINE SIZE FROM SIX INCHES TO EIGHT INCHES

a. Detonable Clouds Proximity of Detonation to Peak Line Cloud Size [Volume TNT Equivalent Safety Related Overpressure Model Model Description Size (ft3)] [Length (ft)] [Tons] Structures (ft) [psi]

9 x 105 Atmospheric Dispersion Gaussian Cloud or Plume 6 100 5000 0.5 4750 8 1.595 x 107 158 4000 0.7 5750 Gravity Layer Gravity Slump, Atmospheric 6 1.5 x 107 118 2200 1.2 Formation/Dispersion Dispersion, Diffusion 4750 8 1.576 x 107 155 2500 <1.2 4750

b. Missiles. The energy of the postulated missile change from 45 to 70 foot-lbs. This is still below the energy required to penetrate safety related structures. Therefore the missile hazard resulting from the change from the six inch pipe to the eight inch pipe is acceptable.

c. Fire Hazard. There is no change to the analysis resulting from the change from six inch to the eight inch pipe.

d. Seismic. Analysis is bounded by the Gravity Layer Formation/Dispersion model for six inch pipe.

Amendment 65 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-1 Stations Referenced for Regional Climatology and Local Meteorology Climatological Distance from Direction from Region of North Station Elevation (ft.) Plant Site (mi.) Plant Site Carolina Raleigh-Durham 434 19 NNE Central Piedmont Moncure 202 7 W Central Piedmont Pinehurst 548 44 SW Southern Piedmont Asheboro 870 54 W Central Piedmont Greensboro 886 69 WNW Northern Piedmont Charlotte 736 117 WSW Southern Piedmont Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-2 SITE AREA NUMBER OF THUNDERSTORM DAYS Month Greensboro Charlotte Raleigh-Durham January

• 1
• February 1 1 1 March 2 2 2 April 3 3 4 May 7 6 6 June 9 8 7 July 11 10 11 August 9 7 8 September 3 3 4 October 1 1 1 November
• 1 1 December * *
• Annual 47 42 46 Period of Record (years) 49 38 33
• = Indicates less than .5 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-3 Number of Cloud-to-Ground Flashes by Season (per sq. km.)LOCATION Season Greensboro Charlotte Raleigh-Durham Winter (D, J, F) .18 .36 .18 Spring (M, A, M) 2.19 2.00 2.19 Summer (J, J, A) 5.28 4.55 4.74 Fall (S, O, N) .73 .91 1.10 Annual 8.56 7.65 8.38 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-4 EXTREME WINDS AND PRECIPITATION ASSOCIATED WITH HURRICANES RALEIGH-DURHAM AIRPORT (1950-1978)Maximum Precipitation 24-hr Precipitation Storm Date Maximum Winds (mph) (inches/hr) (inches)Able 31 Aug. 1952 ESE 30 G 40 1.22 3.52 Barbara 13 Aug. 1953 NE 20 G 28 Trace Trace Carol 30 Aug. 1954 N 18 Trace Trace Edna 10 Sept. 1954 N 20; NNE 16 G 25 Trace 0.01 Hazel 15 Oct. 1954 WNW 73 G 90; NW 48 G 62 1.55 4.04 Connie 11-12 Aug. 1955 E 35 G 46; NE 39 G 54; N 40 0.30; 0.25 0.68; 0.75 Diane 16-17 Aug. 1955 SE 38 G 44; ENE 32 G 53 0.48; 0.70 1.23; 4.12 Ione 19 Sept. 1955 NNE 30; G 49 0.18 0.86 Flossy 26 Sept. 1956 NNE 28; G 46; NNE 29 G 41 0.37 2.31 Helene 27 Sept. 1958 N 29 G 46 Trace 0.07 Gracie 30 Sept. 1959 SSE 25 G 37 0.64 0.78 Brenda 29 July 1960 N 24 0.47 2.60 Donna 11 Sept. 1960 N 29 G 35 0.31 1.48 Esther 20 Sept. 1961 N 17 0.15 0.15 Alma 28 Aug. 1962 NW 16 Trace Trace Ella 18-19 Oct. 1962 NE 22 G 32 0.00 0.00 Ginny 20-21 Oct. 1963 NNE 21 G 32; N 22 G 29 0.00 0.00 Cleo 31 Aug. 1964 NNW 15 1.12 2.95 Dora 13 Sept. 1964 NNE 25 G 38 0.31 2.36 Gladys 22 Sept. 1964 N 18 G 25 0.00 0.00 Isbell 16 Oct. 1964 NE 20 G 29 0.19 0.55 Alma 11 June 1966 NNE 23 G 32 Trace Trace Agnes 19-21 June 1972 14 E; 20 SE; 24 N .03; .29; .55 .03; 1.3; 1.59 Ginger 30 Sept. - 2 Oct. 1971 32 NNW; 29 N; 14 W .11; .34; .08 .61; 2.64; .33 Doria 26-28 Aug. 1971 20 E, 15 NNW, 9N .53; .08; Trace 1.23; .23; Trace Gladys 18-20 Oct. 1968 17 ESE; 15 S; 18 N .12; .59; .01 .63; l.84; .01 Doria 9 Sept. 1967 15 N .15 .89 Belle 8, 9 Aug. 1976 8 ENE; 13 NNW .09; 0.0 .18; 0.00 Eloise 22-26 Sept. 1975 16 E; 14 SSE .24; .32 .95; .77 16 SSE; 10 NW .72; .50 1.26; 1.02 10 WNW .30 .44 Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 NOTE: G indicates gusts to Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-5 SITE REGION METEOROLOGICAL EXTREMES (month/year of occurrence) [Data period]Charlotte Greensboro Raleigh-Durham Pinehurst Asheboro Moncure 12.48 in 13.26 in 12.94 in 13.88 in 13.79 in 12.55 in Maximum Monthly Precipitation (5/75) (9/47) (9/45) (7/59) (7/65) (7/73)(water equivalent)[1940-77] [1929 77] [1945-77] [1951-73] [1951-73] [1951-73]5.34 in 7.49 in 5.20 in 7.11 in 8.96 in 5.14 in Maximum 24 hour Precipitation (10/76) (9/47) (8/55) (10/54) (8/66) (8/67)(water equivalent)[1940-77] [1929-77] [1945-77] [1951-73] [1951-73] [1951-73]Trace .13 in .23 in Minimum Monthly Precipitation (10/53) (9/39) (4/76) --- --- ---(water equivalent)[1940-77] [1929-77] [1945-77]19.3 in 22.9 in 14.4 in 16.0 in 18.5 in 14.0 in Maximum Monthly Snowfall (inches) (3/60) (1/66) (1/55) (12/58) (3/60) (3/60)[1940-77] [1929-70] [1945-77] [1951-73] [1951-73] [1951-73]12.0 in 14.3 in 9.3 in Maximum 24 hour Snowfall (inches) (2/69) (12/30) (3/69) --- --- ---[1940-77] [1929-70] [1945-77]104°F 102°F 105°F 106°F 103°F* 107°F*Maximum Temperature (°F) (9/54) (7/77) (7/52) (8/54) (7/52) (7/52)[1940-77] [1929-70] [1945-77] [1931-73] [1951-73] [1951-73]

 -5°F -8°F -9°F +3°F* -8°F -4°F Minimum Temperature (°F) (1/85) (1/85) (1/85) (12/62) 1/85 (1/66)

[1940-85] [1929-85] [1945-85] [1951-73] [1951-85] [1951-73]

• On earlier dates Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-6 MEAN MONTHLY MAXIMUM MIXING DEPTHS (METERS ABOVE SURFACE)Greensboro Month Depth (m)January 390 February 650 March 1130 April 1180 May 1530 June 1790 July 1490 August 1420 September 1370 October 1020 November 840 December 580 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.1-7 FREQUENCY OF INVERSIONS BASED BELOW 500 FEET Percent Frequency of Inversion Occurrence at Specific Times and All Times Season 0300 GMT 1500 GMT 0000 GMT 1200 GMT All Times Winter 73 15 58 72 43 Spring 70 3 13 66 32 Summer 78 1 11 6 33 Fall 74 4 52 74 40 NOTE: 1. 0300 and 1500 GMT observations for the period 6/55 - 5/57.

2. 0000 and 1200 GMT observations for the period 6/57 - 5/59.

3. Observations made at Greensboro, N.C.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)A STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.000690 0.000354 0.000000 0.000000 0.000000 0.000000 0.001044 NNE 0.000458 0.000194 0.000000 0.000000 0.000000 0.000000 0.000652 NE 0.000295 0.000148 0.000000 0.000000 0.000000 0.000000 0.000444 ENE 0.000272 0.000171 0.000000 0.000000 0.000000 0.000000 0.000444 E 0.000423 0.000308 0.000000 0.000000 0.000000 0.000000 0.000731 ESE 0.000254 0.000137 0.000000 0.000000 0.000000 0.000000 0.000391 SE 0.000266 0.000126 0.000000 0.000000 0.000000 0.000000 0.000391 SSE 0.000217 0.000148 0.000000 0.000000 0.000000 0.000000 0.000365 S 0.000442 0.000263 0.000000 0.000000 0.000000 0.000000 0.000705 SSW 0.000445 0.000285 0.000000 0.000000 0.000000 0.000000 0.000731 SW 0.000897 0.000434 0.000000 0.000000 0.000000 0.000000 0.001331 WSW 0.000651 0.000445 0.000000 0.000000 0.000000 0.000000 0.001096 W 0.000775 0.000400 0.000000 0.000000 0.000000 0.000000 0.001174 WNW 0.000573 0.000263 0.000000 0.000000 0.000000 0.000000 0.000835 NW 0.000395 0.000205 0.000000 0.000000 0.000000 0.000000 0.000600 NNW 0.000527 0.000308 0.000000 0.000000 0.000000 0.000000 0.000835 TOTAL 0.0007581 0.004190 0.000000 0.000000 0.000000 0.000000 RELATIVE FREQUENCY OF OCCURRENCE OF A STABILITY = 0.011770 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH A STABILITY = 0.006622 Amendment 61 Page 1 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)B STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.001245 0.001781 0.001199 0.000000 0.000000 0.000000 0.004224 NNE 0.000853 0.001164 0.001039 0.000000 0.000000 0.000000 0.003056 NE 0.001054 0.001393 0.001085 0.000000 0.000000 0.000000 0.003532 ENE 0.000533 0.000890 0.000605 0.000000 0.000000 0.000000 0.002029 E 0.000594 0.000993 0.000833 0.000000 0.000000 0.000000 0.002420 ESE 0.000517 0.000788 0.000616 0.000000 0.000000 0.000000 0.001922 SE 0.000557 0.000970 0.000628 0.000000 0.000000 0.000000 0.002155 SSE 0.000576 0.000833 0.000582 0.000000 0.000000 0.000000 0.001991 S 0.001040 0.001792 0.001735 0.000000 0.000000 0.000000 0.004568 SSW 0.001387 0.001564 0.001450 0.000000 0.000000 0.000000 0.004401 SW 0.001902 0.002740 0.002443 0.000000 0.000000 0.000000 0.007085 WSW 0.001415 0.002055 0.001747 0.000000 0.000000 0.000000 0.005217 W 0.001568 0.002169 0.002158 0.000000 0.000000 0.000000 0.005895 WNW 0.001421 0.001678 0.001313 0.000000 0.000000 0.000000 0.004413 NW 0.000973 0.001370 0.000913 0.000000 0.000000 0.000000 0.003257 NNW 0.000883 0.001119 0.000902 0.000000 0.000000 0.000000 0.002984 TOTAL 0.016520 0.023301 0.019328 0.000000 0.000000 0.000000 RELATIVE FREQUENCY OF OCCURRENCE OF B STABILITY = 0.059149 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH B STABILITY = 0.009168 Amendment 61 Page 2 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)C STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.000836 0.002614 0.006256 0.001027 0.000057 0.000011 0.010803 NNE 0.000450 0.001439 0.004327 0.000696 0.000011 0.000011 0.006935 NE 0.000392 0.001564 0.005058 0.000913 0.000057 0.000000 0.007984 ENE 0.000399 0.001244 0.002923 0.000434 0.000000 0.000000 0.005000 E 0.000379 0.001279 0.003048 0.000502 0.000023 0.000000 0.005231 ESE 0.000340 0.001005 0.001712 0.000091 0.000000 0.000000 0.003149 SE 0.000307 0.001187 0.002021 0.000148 0.000000 0.000000 0.003664 SSE 0.000328 0.001085 0.002101 0.000263 0.000000 0.000000 0.003776 S 0.000578 0.002397 0.006165 0.000719 0.000034 0.000000 0.009894 SSW 0.000905 0.003334 0.006565 0.001153 0.000034 0.000000 0.011991 SW 0.001361 0.004806 0.009259 0.001347 0.000011 0.000000 0.016785 WSW 0.001003 0.003208 0.005309 0.000674 0.000011 0.000000 0.010205 W 0.000800 0.002923 0.005594 0.001005 0.000023 0.000000 0.010344 WNW 0.000638 0.002215 0.004338 0.000742 0.000046 0.000000 0.007979 NW 0.000551 0.001827 0.004236 0.000879 0.000023 0.000000 0.007515 NNW 0.000572 0.001724 0.003539 0.000639 0.000011 0.000000 0.006486 TOTAL 0.009841 0.033850 0.072449 0.011234 0.000342 0.000023 RELATIVE FREQUENCY OF OCCURRENCE OF C STABILITY = 0.127740 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH C STABILITY = 0.006976 Amendment 61 Page 3 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)D STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.001991 0.006039 0.018175 0.014716 0.001747 0.000320 0.042988 NNE 0.001685 0.005309 0.016177 0.014316 0.002066 0.000285 0.039839 NE 0.001945 0.006085 0.016668 0.011576 0.000925 0.000080 0.037279 ENE 0.001629 0.005046 0.010720 0.005994 0.000251 0.000046 0.023686 E 0.001416 0.004886 0.010686 0.003904 0.003423 0.000068 0.021304 ESE 0.001234 0.003825 0.007090 0.002580 0.000274 0.000068 0.015071 SE 0.001022 0.003539 0.008608 0.002580 0.000308 0.000034 0.016092 SSE 0.001313 0.003608 0.007615 0.003699 0.000377 0.000023 0.016634 S 0.002016 0.007078 0.019168 0.011188 0.001153 0.000091 0.040695 SSW 0.002026 0.006667 0.019648 0.013540 0.001427 0.000171 0.043480 SW 0.001947 0.006553 0.015709 0.011085 0.001267 0.000160 0.036722 WSW 0.001517 0.004053 0.005434 0.004121 0.000422 0.000080 0.015628 W 0.001496 0.004281 0.005537 0.008220 0.001553 0.000126 0.021212 WNW 0.001226 0.003128 0.005434 0.011816 0.002375 0.000217 0.024196 NW 0.001090 0.003140 0.006987 0.011519 0.002295 0.000251 0.025281 NNW 0.001153 0.003573 0.009031 0.009373 0.001336 0.000228 0.024694 TOTAL 0.024705 0.076811 0.182688 0.140230 0.018118 0.002249 RELATIVE FREQUENCY OF OCCURRENCE OF D STABILITY = 0.444801 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH D STABILITY = 0.017661 Amendment 61 Page 4 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)E STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.000000 0.003939 0.006873 0.000000 0.000000 0.000000 0.010811 NNE 0.000000 0.003265 0.003356 0.000000 0.000000 0.000000 0.006622 NE 0.000000 0.003722 0.003322 0.000000 0.000000 0.000000 0.007044 ENE 0.000000 0.002980 0.002318 0.000000 0.000000 0.000000 0.005297 E 0.000000 0.004156 0.004624 0.000000 0.000000 0.000000 0.008779 ESE 0.000000 0.003025 0.002763 0.000000 0.000000 0.000000 0.005788 SE 0.000000 0.003345 0.002340 0.000000 0.000000 0.000000 0.005685 SSE 0.000000 0.003311 0.003356 0.000000 0.000000 0.000000 0.006667 S 0.000000 0.008665 0.009944 0.000000 0.000000 0.000000 0.018609 SSW 0.000000 0.008186 0.009658 0.000000 0.000000 0.000100 0.017844 SW 0.000000 0.006633 0.005948 0.000000 0.000000 0.000100 0.012581 WSW 0.000000 0.002500 0.001895 0.000000 0.000000 0.000000 0.004395 W 0.000000 0.002375 0.004395 0.000000 0.000000 0.000000 0.006770 WNW 0.000000 0.001792 0.004978 0.000000 0.000000 0.000000 0.006770 NW 0.000000 0.001918 0.004921 0.000000 0.000000 0.000000 0.006839 NNW 0.000000 0.001998 0.005069 0.000000 0.000000 0.000000 0.007067 TOTAL 0.000000 0.061809 0.075760 0.000000 0.000000 0.000000 RELATIVE FREQUENCY OF OCCURRENCE OF E STABILITY = 0.137570 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH E STABILITY = 0.000000 Amendment 61 Page 5 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)F STABILITY SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.008955 0.009430 0.000000 0.000000 0.000000 0.000000 0.018385 NNE 0.006378 0.005594 0.000000 0.000000 0.000000 0.000000 0.011972 NE 0.005050 0.005137 0.000000 0.000000 0.000000 0.000000 0.010187 ENE 0.004837 0.004384 0.000000 0.000000 0.000000 0.000000 0.009221 E 0.005642 0.006051 0.000000 0.000000 0.000000 0.000000 0.011693 ESE 0.005481 0.004966 0.000000 0.000000 0.000000 0.000000 0.010447 SE 0.004368 0.003756 0.000000 0.000000 0.000000 0.000000 0.008124 SSE 0.004762 0.004738 0.000000 0.000000 0.000000 0.000000 0.009499 S 0.011355 0.012124 0.000000 0.000000 0.000000 0.000000 0.023479 SSW 0.014542 0.015127 0.000000 0.000000 0.000000 0.000100 0.029669 SW 0.012224 0.012296 0.000000 0.000000 0.000000 0.000100 0.024520 WSW 0.006704 0.005137 0.000000 0.000000 0.000000 0.000000 0.011842 W 0.006311 0.005754 0.000000 0.000000 0.000000 0.000000 0.012065 WNW 0.004924 0.004464 0.000000 0.000000 0.000000 0.000000 0.009388 NW 0.004409 0.004532 0.000000 0.000000 0.000000 0.000000 0.008942 NNW 0.004764 0.004772 0.000000 0.000000 0.000000 0.000000 0.009537 TOTAL 0.110706 0.108263 0.000000 0.000000 0.000000 0.000000 RELATIVE FREQUENCY OF OCCURRENCE OF F STABILITY = 0.218970 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE WITH F STABILITY = 0.084494 Amendment 61 Page 6 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-1 WIND DISTRIBUTION BY PASQUILL STABILITY CLASSES (STAR PROGRAM)ALL STABILITIES SPEED (KTS)DIRECTION 0-3 4-6 7 - 10 11 - 16 17 - 21 GREATER TOTAL THAN 21 N 0.013149 0.024157 0.032503 0.015743 0.001804 0.000331 0.087688 NNE 0.009875 0.016965 0.024900 0.015013 0.002078 0.000297 0.069127 NE 0.009486 0.018050 0.026133 0.012490 0.000982 0.000080 0.067219 ENE 0.008073 0.014716 0.016565 0.006428 0.000251 0.000046 0.046079 E 0.008703 0.017673 0.019191 0.004407 0.000365 0.000068 0.050408 ESE 0.007822 0.013746 0.012181 0.002671 0.000274 0.000068 0.036763 SE 0.006990 0.012924 0.013597 0.002729 0.000308 0.000034 0.036581 SSE 0.007458 0.013723 0.013654 0.003962 0.000377 0.000023 0.039196 S 0.015870 0.032320 0.037013 0.011907 0.001187 0.000091 0.098389 SSW 0.018655 0.035163 0.037321 0.014693 0.001461 0.000171 0.107465 SW 0.018068 0.033462 0.033359 0.012433 0.001279 0.000160 0.098760 WSW 0.011126 0.017399 0.014385 0.004795 0.000434 0.000080 0.048218 W 0.010639 0.017901 0.017684 0.009225 0.001575 0.000126 0.057150 WNW 0.008538 0.013540 0.016063 0.012558 0.002420 0.000217 0.053336 NW 0.007277 0.012992 0.017056 0.012398 0.002318 0.000251 0.052292 NNW 0.007625 0.013494 0.018620 0.010012 0.001347 0.000228 0.051327 TOTAL 0.169353 0.308224 0.350226 0.151463 0.018461 0.002272 RELATIVE FREQUENCY OF OCCURRENCE = 1.000000 RELATIVE FREQUENCY OF CALMS DISTRIBUTED ABOVE = 0.124920 Amendment 61 Page 7 of 7

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS A NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 43 18 11 6 4 NNE 24 13 4 7 4 1 NE 20 10 8 2 6 ENE 13 3 3 3 1 E 6 4 2 2 ESE 11 4 1 SE 13 1 1 SSE 9 7 1 1 S 19 4 2 SSW 29 12 8 12 4 SW 32 26 16 5 9 WSW 33 14 14 6 7 W 29 11 2 2 2 WNW 36 11 11 4 8 NW 44 21 8 6 4 NNW 31 15 6 3 9 1 AVERAGE 1.0 2.0 3.0 4.0 5.6 8.5 0.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 9 0 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 49

• See last page of Table 2.3.2-2.

Amendment 61 Page 1 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS C NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 67 15 5 2 1 NNE 43 7 2 1 NE 39 4 2 ENE 33 8 4 E 27 7 1 ESE 14 2 4 SE 15 2 3 SSE 24 6 1 1 S 35 11 1 SSW 80 15 1 2 SW 80 14 7 3 WSW 72 9 9 1 1 W 46 10 1 5 1 WNW 67 7 5 1 NW 64 13 4 1 NNW 66 8 3 3 AVERAGE 1.0 2.0 3.0 4.0 5.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 5 0 0 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 17

• See last page of Table 2.3.2-2.

Amendment 61 Page 2 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS B NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 45 11 5 NNE 41 4 1 1 2 NE 37 7 6 ENE 21 6 1 E 19 3 1 ESE 15 4 1 SE 14 SSE 10 2 S 22 6 2 1 SSW 39 13 3 2 1 SW 56 16 6 WSW 62 16 3 1 2 W 31 7 2 WNW 53 16 4 2 NW 48 19 4 2 2 NNW 50 8 1 AVERAGE 1.0 2.0 3.0 4.0 5.1 0.0 0.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 6 0 0 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 18

• See last page of Table 2.3.2-2.

Amendment 61 Page 3 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS D NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 145 61 24 13 18 10 4 1 NNE 134 54 21 11 23 4 3 3 2 1 NE 124 46 17 14 5 5 1 1 ENE 101 31 12 7 11 2 1 E 83 34 11 2 7 2 ESE 85 20 11 7 4 SE 82 28 11 5 6 1 SSE 98 31 15 11 10 2 S 132 48 15 5 7 1 SSW 203 65 26 11 14 2 1 SW 219 71 33 14 16 5 WSW 160 61 31 16 15 2 W 141 37 12 4 5 1 WNW 138 44 15 11 8 2 2 NW 137 37 13 9 14 2 NNW 148 52 20 15 14 3 1 AVERAGE 1.0 2.0 3.0 4.0 5.5 8.8 11.6 14.7 17.5 20.0 24.0 28.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 16 18 20 24 28 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 88

• See last page of Table 2.3.2-2.

Amendment 61 Page 4 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS E NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 153 35 20 10 14 3 NNE 127 51 24 4 13 5 NE 124 31 17 5 10 4 1 1 ENE 123 22 8 5 5 1 2 E 99 21 9 4 9 2 ESE 100 20 5 2 5 SE 109 27 9 3 1 SSE 147 31 18 10 10 3 S 176 59 39 16 23 1 1 1 SSW 226 66 41 22 24 11 1 SW 235 41 22 11 13 2 WSW 126 27 24 7 8 1 W 116 25 10 3 3 1 WNW 120 29 12 3 4 1 NW 117 34 10 8 11 2 1 NNW 127 33 16 5 6 4 AVERAGE 1.0 2.0 3.0 4.0 5.6 8.5 11.5 15.0 18.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 15 18 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 103

• See last page of Table 2.3.2-2.

Amendment 61 Page 5 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS F NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 113 28 14 6 6 1 NNE 111 19 9 5 1 NE 99 20 5 4 3 ENE 68 19 10 3 1 E 82 12 5 1 2 ESE 97 15 3 1 1 SE 82 10 1 4 SSE 111 24 5 3 4 S 121 45 14 9 5 1 SSW 127 41 15 14 8 1 1 SW 105 29 16 8 3 1 WSW 90 19 6 5 3 W 75 12 4 4 WNW 70 14 3 1 NW 82 7 3 1 NNW 95 20 9 4 1 AVERAGE 1.0 2.0 3.0 4.0 5.4 9.0 11.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 11 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 51

• See last page of Table 2.3.2-2.

Amendment 61 Page 6 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS G NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 198 55 28 11 11 2 1 NNE 206 44 37 17 4 NE 218 53 16 15 3 ENE 200 35 27 6 2 E 190 31 18 5 1 ESE 165 27 8 3 2 SE 140 17 10 3 SSE 128 20 6 3 2 S 150 30 12 3 3 1 SSW 163 27 15 7 2 SW 135 31 19 6 3 WSW 161 15 6 4 6 W 131 12 6 1 1 1 WNW 126 9 4 1 2 NW 125 16 11 3 1 NNW 171 21 17 1 6 1 AVERAGE 1.0 2.0 3.0 4.0 5.7 8.6 11.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 11 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 67

• See last page of Table 2.3.2-2.

Amendment 61 Page 7 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978

SUMMARY

NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 396 168 101 50 84 30 14 2 3 1 NNE 440 153 95 66 62 20 7 3 4 1 1 NE 463 130 65 41 47 19 7 1 3 1 ENE 399 110 60 31 39 6 7 1 E 375 89 58 25 22 11 1 1 ESE 361 87 43 27 22 2 SE 358 88 36 19 20 3 1 SSE 394 99 53 35 41 10 2 S 434 149 86 54 62 13 4 2 1 SSW 451 182 94 74 93 31 9 5 2 2 SW 445 158 111 61 93 27 8 2 WSW 410 123 77 47 62 20 8 2 1 1 W 393 97 56 25 25 7 2 WNW 369 83 61 30 38 17 6 1 2 NW 366 90 68 43 54 8 8 2 1 NNW 421 107 84 45 55 16 5 3 1 1 AVERAGE 1.0 2.0 3.0 4.0 5.7 8.8 11.7 14.9 17.7 20.8 24.0 29.6 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 16 19 22 24 30 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 371

• See last page of Table 2.3.2-2.

Amendment 61 Page 8 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS A NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 38 22 5 8 3 NNE 39 11 7 5 8 1 NE 15 10 6 3 6 ENE 12 6 4 2 E 15 3 2 1 ESE 12 4 3 1 SE 6 2 1 SSE 9 3 4 S 23 6 4 1 SSW 31 13 6 9 6 SW 40 18 22 6 8 1 WSW 40 17 8 7 6 W 29 13 6 3 WNW 37 16 9 7 8 NW 33 14 10 8 2 NNW 23 16 6 3 8 2 AVERAGE 1.0 2.0 3.0 4.0 5.4 8.2 0.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 9 0 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 13

• See last page of Table 2.3.2-2.

Amendment 61 Page 9 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS B NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 39 11 4 1 NNE 48 6 3 2 1 NE 32 8 3 1 1 ENE 21 5 1 E 20 4 1 ESE 15 4 1 SE 9 2 SSE 20 2 1 S 18 7 1 1 1 SSW 41 11 4 1 3 SW 66 12 5 4 WSW 47 14 3 2 2 W 32 4 5 WNW 49 18 6 3 3 NW 44 9 4 3 1 NNW 42 4 2 AVERAGE 1.0 2.0 3.0 4.0 5.3 0.0 0.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 6 0 0 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 4

• See last page of Table 2.3.2-2.

Amendment 61 Page 10 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS C NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 62 15 3 1 NNE 53 6 3 NE 42 6 2 1 ENE 35 7 E 29 4 1 ESE 19 2 4 SE 11 4 2 SSE 22 6 4 1 S 37 9 5 1 1 SSW 84 12 2 3 SW 80 15 3 4 WSW 69 14 8 3 2 W 51 10 2 1 WNW 55 6 5 2 NW 62 14 4 NNW 65 10 2 1 AVERAGE 1.0 2.0 3.0 4.0 5.0 0.0 11.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 5 0 11 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 7

• See last page of Table 2.3.2-2.

Amendment 61 Page 11 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS D NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 126 51 18 14 24 7 6 1 NNE 119 54 31 12 27 7 3 1 1 NE 118 34 19 7 16 4 1 1 1 1 ENE 107 36 25 5 8 2 E 77 28 13 7 6 2 ESE 79 25 12 3 3 3 SE 89 28 6 7 5 1 SSE 88 34 15 7 14 1 2 S 122 49 19 11 4 1 SSW 181 65 25 16 17 1 1 SW 182 71 41 19 23 2 1 WSW 161 60 22 18 12 3 W 139 31 19 4 4 WNW 128 44 15 5 6 2 1 1 NW 126 33 11 20 10 4 NNW 126 44 23 6 12 3 1 1 AVERAGE 1.0 2.0 3.0 4.0 5.5 8.6 11.3 14.5 17.7 0.0 24.0 34.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 15 19 0 24 34 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 101

• See last page of Table 2.3.2-2.

Amendment 61 Page 12 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS E NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 124 34 9 12 15 4 NNE 97 36 16 13 16 5 NE 110 34 19 11 11 5 2 1 ENE 74 20 13 10 9 3 1 E 90 26 9 4 9 2 ESE 71 19 11 7 5 SE 80 26 5 4 3 SSE 104 36 22 7 17 2 S 144 51 37 21 28 5 1 SSW 189 71 34 24 30 15 2 1 SW 201 44 21 16 13 3 WSW 131 43 21 5 13 1 W 111 27 6 6 5 1 WNW 100 32 14 10 2 1 NW 82 26 13 6 15 2 1 NNW 118 28 16 5 6 4 AVERAGE 1.0 2.0 3.0 4.0 5.6 8.6 11.7 15.3 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 16 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 32

• See last page of Table 2.3.2-2.

Amendment 61 Page 13 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS F NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 79 19 13 8 2 1 NNE 66 23 4 5 3 NE 67 15 9 3 3 ENE 50 18 5 3 1 E 54 17 11 2 2 2 ESE 64 15 7 3 SE 53 15 1 1 2 SSE 73 23 8 4 1 S 88 36 11 11 10 1 SSW 106 52 16 13 16 4 SW 102 44 11 18 7 5 1 WSW 100 34 10 6 7 W 73 18 7 4 5 WNW 68 9 9 3 1 NW 77 9 4 1 NNW 70 18 7 4 2 AVERAGE 1.0 2.0 3.0 4.0 5.6 8.3 11.0 0.0 0.0 0.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 9 11 0 0 0 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 16

• See last page of Table 2.3.2-2.

Amendment 61 Page 14 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978 STABILITY CLASS G NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 65 19 18 7 6 NNE 68 26 11 7 7 2 NE 56 26 12 5 7 2 ENE 68 22 13 4 9 1 E 44 28 11 3 6 1 ESE 41 25 15 4 6 SE 68 26 11 5 2 SSE 56 32 16 10 12 1 S 82 36 17 16 12 1 SSW 95 53 26 22 22 6 SW 113 56 26 18 16 3 WSW 69 50 37 28 27 9 2 W 97 29 22 11 5 1 WNW 92 31 16 8 9 1 NW 90 32 15 11 3 1 NNW 70 36 16 9 8 1 AVERAGE 1.0 2.0 3.0 4.0 5.6 8.5 11.6 0.0 0.0 20.0 0.0 0.0 DURATION HOURS MAXIMUM 1 2 3 4 7 10 12 0 0 20 0 0 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 38

• See last page of Table 2.3.2-2.

Amendment 61 Page 15 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978

SUMMARY

NUMBER OF OCCURRENCES - WIND DIRECTION PERSISTENCE (HOURS)LOWER 1 2 3 4 5-7 8 - 10 11 - 13 14 - 16 17 - 19 20 - 22 23 - 25 > 25 LEVEL WIND DIRECTION N 197 113 67 50 79 24 15 5 3 1 NNE 199 123 69 36 76 37 8 6 2 1 NE 225 100 52 34 52 24 10 2 6 1 ENE 195 84 55 23 48 12 6 2 1 E 135 83 46 23 35 13 4 2 ESE 169 76 55 29 33 4 3 SE 184 83 30 28 27 6 1 SSE 181 97 60 34 73 13 4 1 1 S 231 137 85 60 93 15 13 3 SSW 252 149 94 72 137 44 17 11 2 2 1 2 SW 332 155 111 66 108 47 15 4 4 WSW 241 156 102 60 99 38 17 2 1 W 319 122 61 33 48 9 2 2 1 WNW 248 92 66 38 52 23 10 2 2 1 NW 249 97 54 54 62 15 5 3 2 NNW 235 114 87 36 50 21 5 4 1 AVERAGE 1.0 2.0 3.0 4.0 5.7 8.7 11.6 14.8 17.7 20.7 23.5 32.2 DURATION HOURS MAXIMUM 1 2 3 4 7 10 13 16 19 22 24 37 HOURS NUMBER HOURS OF MISSING WIND DIRECTIONS: 374

• See last page of Table 2.3.2-2.

Amendment 61 Page 16 of 17

Shearon Harris Nuclear Power Plant UFSAR Chapter: 2 TABLE 2.3.2-2 (continued)WIND DIRECTION PERSISTENCE DATA*HARRIS ON-SITE METEOROLOGICAL FACILITY JANUARY 14, 1976 TO DECEMBER 31, 1978

• PERSISTENCE IS DEFINED AS A DELTA T EXISTING WITHIN A DEFINED WIND DIRECTION SECTOR AND IS NOT CONSIDERED TO BE INTERRUPTED IF IT DEPARTS FROM THAT DELTA T VALUE FOR UP TO 1 HOUR AND THEN RETURNS, OR IF THERE IS ONE HOUR OF MISSING DATA FOLLOWED BY A CONTINUED DELTA T VALUE. TWO OR MORE CONSECUTIVE HOURS OF LOST DATA ARE NOT INCLUDED IN THE PERSISTENCE DETERMINATION BUT ARE INDICATED AS "MISSING WIND DIRECTIONS]]