ML23335A067

Updated Final Safety Analysis Report (Fsar), Amendment 30, Public Use Redacted

ML23335A067
Person / Time
Site:	Browns Ferry
Issue date:	10/31/2023
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
Green K
References
EPID L-2023-LRO-0070
Download: ML23335A067 (1)

Text

{{#Wiki_filter:TENNESSEE VALLEY AUTHORITY Browns Ferry Nuclear Plant UPDATED FINAL SAFETY ANALYSIS REPORT (FSAR) Amendment 30 PUBLIC USE October, 2023

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Effective Page Listing 1 through 38 30 Table of Contents 1 through 8 30 Chapter 1 1.0-i through 1.0-v 29 1.1-1 and 1.1-2 28 1.1-3 27 1.2-1 through 1.2-7 25 1.2-8 28 1.2-9 through 1.2-13 25 Figures 1.2-1 through 1.2-3 16 1.3-1 through 1.3-4 19 Table 1.3-1, Sheet 1 28 Table 1.3-1, Sheets 2 through 13 25 Table 1.3-2, Sheets 1 through 5 25 Table 1.3-2, Sheet 6 through 11 28 Figure 1.3-1 16 Figure 1.3-2 28 1.4-1 16 1.4-2 17 Table 1.4-1 21 Table 1.4-2A, Sheets 1 through 6 21 Table 1.4-2B, Sheets 1 through 3 21 1.5-1 through 1.5-13 19 1.6-1 30 1.6-2 27 1.6-3 and 1.6-4 27 1.6-5 28 1.6-6 27 1.6-7 through 1.6-16 27 1.6-17 30 1.6-18 through 1.6-24 27 Figure 1.6-1 23 1

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 1.6-2 28 Figure 1.6-3, Sheet 1 30 Figure 1.6-3, Sheet 2 30 Figure 1.6-4 30 Figure 1.6-5 30 Figure 1.6-6 28 Figure 1.6-7 27 Figure 1.6-8, Sheet 1 28 Figure 1.6-8, Sheet 2 28 Figure 1.6-8, Sheet 3 28 Figure 1.6-9 21 Figure 1.6-10 28 Figure 1.6-11 28 Figure 1.6-12 29 Figure 1.6-13 30 Figure 1.6-14 27 Figure 1.6-15 29 Figures 1.6-16 through 1.6-17 25 Figure 1.6-18 30 Figure 1.6-19 25 Figures 1.6-20 through 1.6-22 (Deleted Sheets) 21 Figure 1.6-23 29 Figure 1.6-24 29 Figure 1.6-25 24 Figure 1.6-26 28 Figure 1.6-27 30 Figure 1.6-28 28 Figure 1.6-29, Sheets 1 and 2 28 Figure 1.6-29, Sheet 3 29 Figure 1.6-30 28 1.7-1 and 1.7-2 17 Table 1.7-1, Sheets 1 through 8 17 Tables 1.7-2 and 1.7-3 17 Table 1.7-4, Sheets 1 through 3 17 Table 1.7-5 17 1.8-1 21 1.9-1 17 1.10-1 16 1.11-1 and 1.11-2 16 Table 1.11-1, Sheets 1 and 2 16 2

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table 1.11-2, Sheets 1 and 2 16 Table 1.11-3, Sheets 1 and 2 16 Table 1.11-4 16 Table 1.11-5, Sheets 1 through 3 16 Table 1.11-6 16 1.12-1 28 Chapter 2 2.0-i through 2.0-xi 29 2.1-1 16 2.2-1 30 2.2-2 and 2.2-3 27 Table 2.2-1 (Deleted Sheet) 18 Table 2.2-2, Sheets 1 through 3 18 Table 2.2-3 18 Table 2.2-4, Sheets 1 through 7 18 Table 2.2-5, Sheets 1 through 7 18 Table 2.2-6 18 Table 2.2-7, Sheets 1 through 5 18 Table 2.2-8 18 Table 2.2-9, Sheets 1 and 2 18 Table 2.2-10, Sheets 1 and 2 18 Table 2.2-11 (Deleted Sheet) 18 Figures 2.2-1 and 2.2-2 18 Figure 2.2-3 (Deleted Sheet) 16 Figure 2.2-4 29 2.3-1 through 2.3-7 25 2.3-8 and 2.3-9 30 2.3-10 through 2.3-13 25 Tables 2.3-1 through 2.3-42 16 Table 2.3-43, Sheets 1 and 2 16 Table 2.3-44, Sheets 1 and 2 16 Tables 2.3-45 through 2.3-49 16 Figures 2.3-1 through 2.3-40 16 2.4-1 25 2.4-2 through 2.4-9 28 2.4-10 through 2.4-13 25 Tables 2.4-1 through 2.4-3 (Deleted Sheets) 19 Table 2.4-4 19 Table 2.4-5 (Deleted Sheet) 19 3

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table 2.4-6, Sheets 1 and 2 19 Figure 2.4-1 16 Figures 2.4-1a through 2.4-1e 16 Figures 2.4-2 through 2.4-4 16 2.4A-i through 2.4A-iv 25 2.4A-1 through 2.4A-11 25 2.4A-12 and 2.4A-13 30 Table 2.4A-1, Sheets 1 and 2 25 Table 2.4A-2, Sheets 1 and 2 25 Table 2.4A-3, Sheets 1 and 2 25 Table 2.4A-4 (deleted) 25 Table 2.4A Max. Possible Flood - Historical, Retire in Place 16 Figures 2.4A-1 through 4 25 Figures 2.4A-5 Sheets 1 through 14 25 Figures 2.4A-6 through 2.4A-16 25 Figures 2.4A-17 through 2.4A-20 (deleted) 25 Figure 2.4A-21 16 Figures 2.4A-22 and 2.4A-22a 24 Figure 2.4A-22B 16 Figures 2.4A-23 through 2.4A-24 16 Figure 2.4A-25 24 Figure 2.4A-26, Sheet 1 and Sheet 2 24 2.5-1 through 2.5-17 26 Table 2.5-1, Sheets 1 through 7 16 Figure 2.5-S1 28 Figures 2.5-S2 through 2.5-S5 16 Figures 2.5-1 through 2.5-5 16 Figures 2.5-5a through 2.5-5l 16 Figures 2.5-5m through 2.5-5aj (Deleted Sheet) 21 Figures 2.5-6 through 2.5-16 16 Figure 2.5-17 25 Figure 2.5-18 (Deleted Sheet) 16 Figure 2.5-19 16 2.6-1 through 2.6-2 17 Chapter 3 3.0-i through 3.0-v 29 3.1-1 16 3.1-2 17 3.2-1 through 3.2-6 28 4

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Tables 3.2-1 through 3.2-5 (Deleted Sheets) 16 Figure 3.2-1 (Deleted Sheet) 21 3.3-1 through 3.3-5 26 3.3-6 28 3.3-7 through 3.3-9 26 3.3-10 and 3.3-11 30 3.3-12 30 3.3-13 and 3.3-14 28 3.3-15 26 3.3-16 28 3.3-17 26 3.3-18 28 Table 3.3-1 28 Figures 3.3-1 through 3.3-6 16 Figure 3.3-7 17 Figure 3.3-8 (Deleted Sheet) 17 Figures 3.3-9 and 3.3-10 16 Figure 3.3-11 (Deleted Sheet) 16 3.4-1 through 3.4-10 27 3.4-11 30 3.4-12 through 3.4-34 27 Figures 3.4-1 through 3.4-6 16 Figure 3.4-7 22 Figure 3.4-8a, Sheet 1 30 Figure 3.4-8a, Sheet 2 29 Figure 3.4-8a, Sheet 3 28 Figure 3.4-8a, Sheet 4 29 Figure 3.4-8a, Sheet 5 28 Figure 3.4-8b 22 Figure 3.4-8c 27 Figure 3.4-8d 22 Figure 3.4-8e 30 Figure 3.4-8f 21 Figure 3.4-8g 25 Figure 3.4-8h 28 Figure 3.4-9 16 Figure 3.4-9a 16 Figure 3.4-10 (Deleted Sheet) 16 Figure 3.4-11 (Deleted Sheet) 22 Figure 3.4-12 16 5

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 3.4-13 20 3.5-1 16 3.5-2 and 3.5-3 17 Figure 3.5-1 16 3.6-1 through 3.6-5 28 3.6-6 and 3.6-7 28 3.6-8 through 3.6-10 30 Table 3.6-1 28 Figures 3.6-1 through 3.6-13 (Deleted Sheet) 16 3.7-1 26 3.7-2 28 3.7-3 26 3.7-4 and 3.7-5 28 3.7-6 26 3.7-7 and 3.7-8 28 3.7-9 through 3.7-12 29 3.7-13 through 3.7-19 28 3.7-20 29 3.7-21 28 3.7-22 and 3.7-23 29 Figure 3.7-1 29 Figures 3.7-2 and 3.7-3 28 3.8-1 through 3.8-2 27 3.8-3 28 3.8-4 and 3.8-5 27 3.8-6 28 3.8-7 27 3.8-8 28 3.8-9 and 3.8-10 27 Figure 3.8-1 28 Figure 3.8-2 27 Figure 3.8-3 28 Figure 3.8-4 (Deleted Sheet) 22 Figure 3.8-5 27 Figure 3.8-6 27 Figure 3.8-7 (Deleted Sheet) 25 Figure 3.8-8 27 6

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Chapter 4 4.1-1 through 4.1-3 16 4.2-1 through 4.2-3 27 4.2-4 29 4.2-5 through 4.2-11 27 4.2-12 28 4.2-13 through 4.2-16 27 4.2-17 28 Tables 4.2-1 through 4.2-3 26 Figure 4.2-1 29 Figure 4.2-2 16 Figure 4.2-3 28 Figure 4.2-4 28 4.3-1 28 4.3-2 26 4.3-3 28 4.3-4 through 4.3-6 26 4.3-7 28 4.3-8 26 Tables 4.3-1a and 4.3-1b 28 Figure 4.3-1 28 Figure 4.3-2a, Sheet 1 30 Figure 4.3-2a, Sheet 2 30 Figure 4.3-2a, Sheet 3 30 Figure 4.3-2b (Deleted Sheet) 16 Figures 4.3-3 and 4.3-4 16 4.4-1 through 4.4-8 26 4.4-9 and 4.4-10 28 4.4-11 and 4.4-12 26 Table 4.4-1 (Deleted Sheet) 22 Table 4.4-1A 22 Figures 4.4-1 and 4.4-2 17 Figure 4.4-3 16 Figures 4.4-4 and 4.4-5 (Deleted Sheets) 16 Figures 4.4-6 and 4.4-7 17 Figure 4.4-8 25 4.5-1 through 4.5-3 21 Figure 4.5-1 23 Figure 4.5-2 24 7

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 4.5-3 23 4.6-1 27 4.6-2 28 4.6-3 and 4.6-4 27 4.6-5 28 4.6-6 27 4.6-7 28 4.6-8 and 4.6-9 27 Figure 4.6-1 16 4.7-1 through 4.7-3 28 4.7-4 27 Table 4.7-1 28 Figure 4.7-1a 25 Figure 4.7-1b 28 Figure 4.7-1c 27 Figure 4.7-1d 29 Figure 4.7-1e 27 Figure 4.7-1f 30 Figures 4.7-2a through 4.7-2h (Deleted Sheet) 22 4.8-1 through 4.8-3 26 4.8-4 and 4.8-5 28 4.8-6 through 4.8-8 26 4.8-9 and 4.8-10 28 Tables 4.8-1 28 Figure 4.8-1 16 4.9-1 through 4.9-3 30 Table 4.9-1 23 Figure 4.9-1 30 Figure 4.9-2 30 Figure 4.9-3 30 Figure 4.9-4 (Deleted Sheet) 16 Figure 4.9-4a through 4.9-4d (Deleted Sheet) 22 Figure 4.9-5 29 Figure 4.9-6 30 Figure 4.9-7 29 Figure 4.9-8 30 Figure 4.9-9 30 Figure 4.9-10 29 4.10-1 through 4.10-7 27 Figure 4.10-1 17 8

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 4.10-2 (Deleted Sheet) 16 Figure 4.10-3 16 4.11-1 through 4.11-3 26 Figure 4.11-1 16 4.12-1 and 4.12-20 28 Chapter 5 5.0-i through 5.0-v 29 5.1-1 29 5.2-1 through 5.2-5 27 5.2-2 27 5.2-3 29 5.2-4 27 5.2-5 27 5.2-6 28 5.2-7 29 5.2-8 through 5.2-14 27 5.2-15 29 5.2-16 30 5.2-17 through 5.2-19 27 5.2-20 28 5.2-21 30 5.2-22 through 5.2-24 27 5.2-25 and 5.2-26 28 5.2-27 through 5.2-37 27 5.2-38 28 5.2-39 through 5.2-40 27 5.2-41 29 5.2-42 28 5.2-43 and 5.2-44 27 5.2-45 29 5.2-46 27 5.2-47 through 5.2-50 28 5.2-51 through 5.2-53 29 Table 5.2-1, Sheet 1 29 Table 5.2-1, Sheet 2 28 Table 5.2-2, Sheet 1 29 Table 5.2-2, Sheet 2 29 Table 5.2-2, Sheet 3 29 Table 5.2-2, Sheet 4 29 9

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table 5.2-2, Sheets 5 through 7 29 Table 5.2-2, Sheet 8 30 Table 5.2-2, Sheet 9 29 Table 5.2-2, Sheet 10 through 13 29 Tables 5.2-3 and 5.2-4 (Deleted Sheets) 27 Figure 5.2-1a 16 Figures 5.2-1b and 5.2-1c (Deleted Sheet) 16 Figure 5.2-2 (Deleted Sheet) 16 Figure 5.2-2a, Sheet 1 27 Figure 5.2-2a, Sheet 2 27 Figure 5.2-2a, Sheet 3 28 Figure 5.2-2b 22 Figure 5.2-2c 27 Figure 5.2-2d 17 Figure 5.2-2e 28 Figure 5.2-2f 22 Figure 5.2-2g 27 Figures 5.2-3 through 5.2-4a 16 Figure 5.2-5 16 Figure 5.2-5b through 5.2-5e (deleted) 28 Figure 5.2-6a, Sheet 1 30 Figure 5.2-6a, Sheet 2 30 Figure 5.2-6a, Sheet 3 29 Figure 5.2-6a, Sheets 4 and 5 30 Figure 5.2-6a, Sheet 6 29 Figure 5.2-6a, Sheet 7 24 Figure 5.2-6b 25 Figure 5.2-6c 26 Figure 5.2-6d 25 Figures 5.2-6e through 5.2-6g 16 Figure 5.2-7, Sheet 1 27 Figure 5.2-7, Sheets 2 through 3 26 Figure 5.2-8, Sheets 1 through 3 22 Figures 5.2-9 and 5.2-10 (Deleted Sheet) 16 Figure 5.2-11 (Deleted Sheet) 18 Figure 5.2-12 (Deleted Sheet) 16 Figure 5.2-13 28 Figure 5.2-13a 28 Figure 5.2-14 28 Figure 5.2-14a 28 10

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 5.2-15 28 Figure 5.2-15a 28 Figure 5.2-16 28 Figure 5.2-16a 28 Figure 5.2-17 28 Figure 5.2-17a 28 Figure 5.2-18 (Deleted Sheet) 17 Figures 5.2-19 through 5.2-21 16 Figure 5.2-22, Sheets 1 through 3 (Deleted Sheet) 16 5.3-1 through 5.3-15 27 5.3-16 28 5.3-17 and 5.3-18 27 5.3-19 28 5.3-20 through 5.3-28 27 Figures 5.3-1 and 5.3-2 16 Figures 5.3-2a through 5.3-2d 16 Figure 5.3-3a 27 Figure 5.3-3b 28 Figure 5.3-3c 27 Figure 5.3-3d 28 Figure 5.3-4 (Deleted Sheet) 16 Figure 5.3-5 16 Figures 5.3-6 through 5.3-8 (Deleted Sheet) 16 Figure 5.3-9 23 Figure 5.3-10 16 Chapter 6 6.0-i through 6.0-v 29 6.1-1 21 6.2-1 and 6.2-2 22 6.3-1 and 6.3-2 17 Table 6.3-1 28 Figure 6.3-1 (Deleted Sheet) 16 6.4-1 27 6.4-2 and 6.4-3 28 6.4-4 through 6.4-12 27 Table 6.4-1 (Deleted Sheet) 16 Figure 6.4-1 29 Figure 6.4-2 28 Figure 6.4-3 30 11

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 6.4-4 28 Figure 6.4-5 29 Figure 6.4-6 22 6.5-1 through 6.5-2 27 6.5-3 and 6.5-4 28 6.5-5 through 6.5-6 27 6.5-7 and 6.5-8 28 6.5-9 27 6.5-10 28 6.5-11 27 6.5-12 through 6.5-13 29 6.5-14 27 6.5-15 through 6.5-20 28 6.5-21 29 Tables 6.5-1 through 6.5-2 28 Table 6.5-3 26 Table 6.5-4 28 Table 6.5-5 (Deleted Sheet) 26 Table 6.5-6 (Deleted Sheet) 26 Table 6.5-7 (Deleted Sheet) 26 Figures 6.5-1 through 6.5-43 (deleted) 24 Figure 6.5-44 28 6.6-1 through 6.6-4 25 Chapter 7 7.0-i through 7.0-xi 29 7.1-1 17 7.1-2 16 7.1-3 through 7.1-5 17 7.1-6 through 7.1-10 28 Table 7.1-1 28 Figure 7.1-1 16 7.2-1 through 7.2-10 27 7.2-11 28 7.2-12 through 7.2-21 25 Tables 7.2-1 and 7.2-1a 22 Table 7.2-2 (Deleted Sheet) 22 Figure 7.2-1 22 Figure 7.2-2 21 Figure 7.2-3 25 12

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figures 7.2-3a through 7.2-3l (Deleted Sheet) 22 Figure 7.2-4 22 Figure 7.2-5 16 Figure 7.2-6 22 Figure 7.2-7 (Deleted Sheet) 16 Figure 7.2-7a 16 Figure 7.2-7b 17 Figure 7.2-7c 25 Figure 7.2-7d 22 Figure 7.2-8 25 Figure 7.2-9 21 Figure 7.2-10 22 Figures 7.2-11 and 7.2-12 16 Figure 7.2-13 (Deleted Sheet) 16 7.3-1 through 7.3-4 27 7.3-5 and 7.3-6 29 7.3-7 through 7.3-12 27 7.3-13 28 7.3-14 and 7.3-15 27 7.3-16 and 7.5-17 29 7.3-18 through 7.3-22 27 7.3-23 29 7.3-24 through 7.3-29 27 Table 7.3-1 (Deleted Sheet) 26 Table 7.3-2, Sheets 1 and 2 26 Figure 7.3-1 (Deleted Sheet) 19 Figure 7.3-1, Sheet 1 30 Figure 7.3-1, Sheet 2 30 Figure 7.3-1, Sheet 3 30 Figures 7.3-2a through 7.3-2L (Deleted Sheet) 22 7.4-1 through 7.4-18 26 7.4-19 28 7.4-20 through 7.4-31 26 Table 7.4-1 (Deleted Sheet) 24 Tables 7.4-2 through 7.4-4 24 Figure 7.4-1a (Deleted Sheet) 18 Figure 7.4-1b, Sheet 1 30 Figure 7.4-1b, Sheet 2 30 Figure 7.4-1b, Sheet 3 30 Figure 7.4-2a through 7.4-2h (Deleted Sheet) 22 13

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 7.4-3 (Deleted Sheet) 22 Figure 7.4-4 (Deleted Sheet) 19 Figure 7.4-5a 28 Figures 7.4-5b and 7.4-5c (Deleted Sheet) 22 Figure 7.4-5d 28 Figures 7.4-5e through 7.4-5h (Deleted Sheet) 22 Figure 7.4-5i 28 Figure 7.4-5l 28 Figure 7.4-5m 27 Figure 7.4-6a, Sheet 1 30 Figure 7.4-6a, Sheet 2 30 Figure 7.4-6a, Sheet 3 30 Figure 7.4-6b, Sheet 1 29 Figure 7.4-6b, Sheet 2 26 Figure 7.4-6b, Sheet 3 27 Figure 7.4-6b, Sheet 4 29 Figure 7.4-6b, Sheet 5 28 Figure 7.4-7a (Deleted Sheet) 22 Figure 7.4-7b 28 Figures 7.4-7c through 7.4-7h (Deleted Sheet) 22 Figure 7.4-7i 28 Figures 7.4-7j through 7.4-7n (Deleted Sheet) 22 Figure 7.4-7p 28 Figure 7.4-8 (Deleted Sheet) 16 Figures 7.4-8a through 7.4-8d (Deleted Sheet) 22 Figure 7.4-9 (Deleted Sheet) 22 7.5-1 through 7.5-4 25 7.5-5 28 7.5-6 through 7.5-14 25 7.5-15 30 7.5-16 29 7.5-17 and 7.5-18 25 7.5-19 28 7.5-20 through 7.5-23 25 7.5-24 30 Tables 7.5-1 and 7.5-2 22 Table 7.5-3 28 Table 7.5-4a 29 Table 7.5-4b 28 Table 7.5-4c (Deleted Sheet) 22 14

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 7.5-1 (Deleted Sheet) 16 Figure 7.5-1a 24 Figure 7.5-1b 28 Figure 7.5-1c 22 Figure 7.5-2 23 Figure 7.5-3a 16 Figures 7.5-3b and 7.5-3c (Deleted Sheet) 22 Figure 7.5-4 16 Figure 7.5-5 (Deleted Sheet) 16 Figure 7.5-6 16 Figure 7.5-7 19 Figures 7.5-8 through 7.5-10 16 Figure 7.5-11 (Deleted Sheet) 16 Figures 7.5-11a 29 Figure 7.5-11b 23 Figure 7.5-11c 28 Figure 7.5-12 16 Figure 7.5-13 17 Figures 7.5-14a and 7.5-14b 17 Figure 7.5-14c 22 Figures 7.5-15 and 7.5-16 (Deleted Sheets) 22 Figure 7.5-17 (Deleted Sheet) 16 Figure 7.5-17a 16 Figure 7.5-17b 22 Figures 7.5-18 through 7.5-22 16 Figures 7.5-23a 29 Figure 7.5-23b 29 Figures 7.5-23c and 7.5-23d 30 Figure 7.5-24 (Deleted Sheet) 16 Figures 7.5-24a through 7.5-24f (Deleted Sheet) 22 Figures 7.5-25 and 7.5-26 16 Figures 7.5-27 and 7.5-28 (Deleted Sheets) 22 7.6-1 through 7.6-5 30 Table 7.6-1, Sheets 1 and 2 30 Figure 7.6-1 16 Figure 7.6-2 (Deleted Sheet) 16 7.7-1 through 7.7-7 25 7.7-8 28 7.7-9 25 7.7-10 29 15

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 7.7-11 through 7.7-14 25 Table 7.7-1, Sheets 1 and 2 25 Figures 7.7-1a through 7.7-1d (Deleted Sheet) 22 Figures 7.7-1e and 7.7-1f (Deleted Sheet) 17 Figure 7.7-2 22 Figure 7.7-3 17 Figure 7.7-4 16 Figure 7.7-5 (Deleted Sheet) 18 Figure 7.7-6 (Deleted Sheet) 16 Figures 7.7-6a and 7.7-6b (Deleted Sheet) 22 Appendix 7.7A (Deleted Sheet) 16 Appendix 7.7B (Deleted Sheet) 16 7.7B-i and 7.7B-ii (Deleted Sheets) 16 7.8-1 through 7.8-8 25 Table 7.8-1, Sheets 1 and 2 18 Table 7.8-2, Sheets 1 and 2 18 Figure 7.8-1, Sheet 1 29 Figure 7.8-1, Sheet 2 29 Figure 7.8-1, Sheet 3 29 Figure 7.8-1, Sheet 4 30 Figure 7.8-1, Sheet 5 24 Figure 7.8-1, Sheet 6 30 Figure 7.8-2 (Deleted Sheet) 17 Figure 7.8-3 24 7.9-1 through 7.9-2 25 7.9-3 28 7.9-4 25 7.9-5 28 7.9-6 and 7.9-7 25 7.9-8 and 7.9-9 28 7.9-10 25 Figure 7.9-1 (Deleted Sheet) 16 Figure 7.9-2 (Deleted Sheet) 22 Figure 7.9-3 (Deleted Sheet) 16 Figure 7.9-4a through 7.9-4e (Deleted Sheet) 22 Figure 7.9-4f (Deleted Sheet) 21 7.10-1 through 7.10-9 23 Figure 7.10-1 (Deleted Sheet) 16 Figure 7.10-2 29 Figures 7.10-3 and 7.10-4 29 16

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 7.10-5 29 Figure 7.10-6 and 7.10-7 29 Figure 7.10-8 23 7.11-1 through 7.11-7 22 Figure 7.11-1 (Deleted Sheet) 22 Figure 7.11-2 22 7.12-1 through 7.12-12 24 Table 7.12-1 16 Table 7.12-2 (Deleted Sheet) 16 Figure 7.12-1 (Deleted Sheet) 16 Figure 7.12-2a, Sheet 1 28 Figure 7.12-2a, Sheet 2 29 Figure 7.12-2a, Sheets 3 and 4 29 Figure 7.12-2a, Sheet 5 25 Figure 7.12-2a, Sheet 6 28 Figure 7.12-2a, Sheet 7 23 Figure 7.12-2b, Sheet 1 (Deleted Sheet) 19 Figure 7.12-2b, Sheet 2 29 Figure 7.12-2b, Sheet 3 (Deleted Sheet) 19 Figure 7.12-2b, Sheets 4 through 5 30 Figure 7.12-2b, Sheet 6 30 7.13-1 and 7.13-2 21 7.13-3 30 7.13-4 21 Table 7.13-1 (Deleted Sheet) 18 Table 7.13-2, Sheets 1 through 7 18 Figure 7.13-1 (Deleted Sheet) 16 7.14-1 16 7.14-2 17 7.15-1 25 7.16-1 through 7.16-9 23 Tables 7.16-1 and 7.16-2 (Deleted Sheets) 16 Figure 7.16-1 (Deleted Sheet) 16 7.17-1 (Deleted Sheet) 16 Tables 7.17-1 and 7.17-2 (Deleted Sheet) 16 Figures 7.17-1 through 7.17-9d (Deleted Sheet) 16 7.18-1 through 7.18-5 27 7.19-1 24 7.19-2 28 7.19-3 29 17

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 7.19-4 24 7.19-5 28 7.20-1 through 7.20-4 22 Table 7.20-1, Sheets 1, 2, and 3 22 Chapter 8 8.0-i through vii 29 8.1-1 29 8.1-2 through 8.1-4 26 8.2-1 29 8.2-2 27 8.3-1 and 8.3-2 26 8.3-3 30 8.3-4 29 8.3-5 26 Figure 8.3-1 (Deleted Sheet) 16 Figure 8.3-2 29 Figure 8.3-2a 17 Figure 8.3-3 16 Figures 8.3-4 through 8.3-6a 30 Figure 8.3-7 (Deleted Sheet) 16 Figures 8.3-7a through 8.3-17 (Deleted Sheet) 17 8.4-1 through 27 8.4-5 through 8.4-7 30 8.4-8 through 8.4-13 27 8.4-14 30 8.4-15 through 8.4-24 27 Table 8.4-1, Sheets 1 through 8 26 Table 8.4-1, Sheet 9 29 Table 8.4-1, Sheets 10 through 19 26 Tables 8.4-2 through 8.4-14 (Deleted Sheets) 26 Figure 8.4-1a 29 Figure 8.4-1b 29 Figure 8.4-2 30 Figure 8.4-3 28 Figure 8.4-4 16 8.5-1 27 8.5-2 30 8.5-3 through 8.5-9 27 8.5-10 28 18

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 8.5-11 30 8.5-12 through 8.5-14 27 8.5-15 29 8.5-16 and 8.5-17 27 8.5-18 29 8.5-19 and 8.5-20 27 Table 8.5-1 21 Tables 8.5-2 and 8.5-2a (Deleted Sheets) 21 Tables 8.5-3 and 8.5-4 (Deleted Sheets) 21 Tables 8.5-5 and 8.5-6 21 Tables 8.5-7 through 8.5-9 (Deleted Sheets) 21 Figure 8.5-1 28 Figure 8.5-2, Sheet 1 25 Figure 8.5-2, Sheet 2 24 Figure 8.5-2, Sheet 3 27 Figure 8.5-2, Sheet 4 26 Figure 8.5-2, Sheet 5 24 Figure 8.5-2, Sheet 6 26 Figure 8.5-3a 29 Figure 8.5-3b 29 Figure 8.5-4a 30 Figure 8.5-4b 30 Figure 8.5-4c 30 Figure 8.5-4d 29 Figure 8.5-4e 29 Figure 8.5-4f 30 Figure 8.5-4g 29 Figure 8.5-4h 29 Figure 8.5-5 29 Figure 8.5-6 29 Figure 8.5-7 (Deleted Sheet) 16 Figure 8.5-7a 28 Figure 8.5-7b 29 Figure 8.5-7c 29 Figure 8.5-7d 27 Figure 8.5-7e 28 Figure 8.5-7f 27 Figure 8.5-8 (Deleted Sheet) 16 Figure 8.5-8a 29 Figure 8.5-8b 28 19

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 8.5-8c 27 Figure 8.5-8d 27 Figure 8.5-8e 27 Figure 8.5-8f 28 Figure 8.5-9a 30 Figure 8.5-9b 25 Figure 8.5-9c 30 Figure 8.5-9d 25 Figure 8.5-10 25 Figure 8.5-11 27 Figure 8.5-11a 27 Figure 8.5-11b (Deleted Sheet) 16 Figure 8.5-11c 30 Figure 8.5-11d 30 Figure 8.5-12a 30 Figure 8.5-12b 29 Figure 8.5-12c 29 Figure 8.5-13a 30 Figure 8.5-13b 29 Figure 8.5-13c 29 Figure 8.5-13d 24 Figure 8.5-13e 30 Figure 8.5-14 (Deleted Sheet) 16 Figure 8.5-14a 22 Figure 8.5-14b 16 Figure 8.5-14c 22 Figure 8.5-15 22 Figures 8.5-16a through 8.5-16c 22 Figure 8.5-17 22 Figure 8.5-18 25 Figures 8.5-19 through 8.5-21 (Deleted Sheets) 16 Figure 8.5-22 16 Figure 8.5-23 (Deleted Sheet) 16 Figure 8.5-24 16 Figure 8.5-25 30 Figure 8.5-26 29 Figure 8.5-27 30 Figure 8.5-28 30 8.6-1 through 8.6-3 20 8.6-4 28 20

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 8.6-5 20 Table 8.6-1 30 Table 8.6-3 (Deleted Sheet) 22 Figures 8.6-1a and 8.6-1b 22 Figure 8.6-1c 16 Figure 8.6-1d 28 Figure 8.6-1e 21 Figure 8.6-1f` 23 Figure 8.6-2a 17 Figure 8.6-2b 30 Figure 8.6-2c 30 Figure 8.6-3 25 Figure 8.6-4 (Deleted Sheet) 16 Figure 8.6-5 22 Figure 8.6-6 30 8.7-1 through 8.7-4 22 Figure 8.7-1 28 Figure 8.7-2 (Deleted Sheet) 16 Figure 8.7-3 30 Figure 8.7-4a 30 Figure 8.7-4b, Sheet 1 28 Figure 8.7-4b, Sheet 2 30 Figure 8.7-4c, Sheet 1 27 Figure 8.7-4c, Sheet 2 26 Figure 8.7-4c, Sheet 3 30 Figure 8.7-4d 30 8.8-1 through 8.8-4 25 Table 8.8-1 and 8.8-2 (Deleted Sheet) 16 Table 8.8-3 17 Figure 8.8-1 (Deleted Sheet) 16 Figure 8.8-2 (Deleted Sheet) 17 Figure 8.8-3 (Deleted Sheet) 16 8.9-1 through 8.9-12 27 Table 8.9-1 (Deleted Sheet) 16 Figure 8.9-1 (Deleted Sheet) 16 8.10-1 through 8.10-2 28 Chapter 9 9.0-i through 9.0-iv 29 9.1-1 18 21

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 9.2-1 and 9.2-2 26 9.2-3 29 9.2-4 through 9.2-7 26 Tables 9.2-1 and 9.2-2 (Deleted Sheets) 18 Table 9.2-3 28 Table 9.2-4, Sheet 1 29 Table 9.2-4, Sheet 2 29 Figures 9.2-1a and 9.2-1b (Deleted Sheet) 19 Figure 9.2-2 (Deleted Sheet) 16 Figure 9.2-3a 29 Figure 9.2-3b 28 Figure 9.2-3c 30 Figure 9.2-3d 29 Figure 9.2-3e 23 Figure 9.2-3f 29 Figures 9.2-3g 22 Figures 9.2-3h and 9.2-3i 16 Figure 9.2-3j 30 Figure 9.2-3k 27 Figure 9.2-3l 27 Figure 9.2-3m 23 Figure 9.2-3n 27 Figure 9.2-3o 29 Figure 9.2-3p 29 Figure 9.2-3q 29 Figure 9.2-3r and 9.2-3s 16 Figure 9.2-3t 25 Figure 9.2-4 (Deleted Sheet) 22 Figures 9.2-4a through 9.2-4f (Deleted Sheet) 16 9.3-1 through 9.3-3 25 9.3-4 and 9.3-5 28 9.3-6 25 Figures 9.3-1a and 9.3-1b (Deleted Sheet) 17 Figures 9.3-2a and 9.3-2b (Deleted Sheet) 17 9.4-1 (Deleted Sheet) 16 9.5-1 through 9.5-11 26 Table 9.5-1, Sheets 1 and 2 26 Table 9.5-2 26 Table 9.5-3, Sheets 1 and 2 26 Table 9.5-4, Sheets 1 and 2 26 22

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table 9.5-4, Sheet 3 28 Table 9.5-5, Sheets 1 through 5 26 Tables 9.5-6 and 9.5-7 26 Figure 9.5-1, Sheet 1 27 Figure 9.5-1, Sheet 2 27 Figure 9.5-1, Sheet 3 27 Figure 9.5-1, Sheet 4 26 Figure 9.5-1, Sheet 5 26 Figure 9.5-1, Sheet 6 27 Figure 9.5-2 26 Figure 9.5-3 26 Figure 9.5-4 27 Chapter 10 10.0-i through 10.0-x 29 10.1-1 16 10.2-1 through 10.2-3 21 Figures 10.2-1a and 10.2-1b (Deleted Sheets) 19 Figure 10.2-2 16 10.3-1 through 10.3-3 27 10.3-4 through 10.3-7 30 Figure 10.3-1 28 Figure 10.3-2 19 Figure 10.3-3 24 10.4-1 through 10.4-2 23 10.4-3 30 10.4-4 29 10.4-5 through 10.4-7 23 Table 10.4-1 30 Figures 10.4-1 through 10.4-6 (Deleted Sheets) 16 Figure 10.4-7 16 10.5-1 through 10.5-7 27 Table 10.5-1 28 Figure 10.5-1a 27 Figure 10.5-1b, Sheet 1 27 Figure 10.5-1b, Sheet 2 26 Figure 10.5-1b, Sheet 3 27 Figure 10.5-1b, Sheet 4 27 Figure 10.5-1c 27 Figure 10.5-1d 27 Figure 10.5-2, Sheet 1 20 23

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 10.5-2, Sheet 2 30 Figure 10.5-2, Sheet 3 21 Figure 10.5-2, Sheet 4 17 Figures 10.5-3 and 10.5-4 (Deleted Sheets) 16 Figures 10.5-4a through 10.5-4d (Deleted Sheet) 22 10.6-1 26 10.6-2 28 10.6-3 and 10.6-4 26 Table 10.6-1 26 Table 10.6-2 28 Tables 10.6-3a and 10.6-3b 28 Figure 10.6-1a 29 Figure 10.6-1b 30 Figure 10.6-1c 29 Figures 10.6-2a through 10.6-2d (Deleted Sheet) 22 10.7-1 27 10.7-2 28 Figure 10.7-1a, Sheet 1 25 Figure 10.7-1a, Sheet 2 26 Figure 10.7-1a, Sheet 3 30 Figure 10.7-1b, Sheets 1 and 2 27 Figure 10.7-1b, Sheet 3 25 Figure 10.7-1b, Sheet 4 30 Figure 10.7-1b, Sheet 5 26 Figure 10.7-1b, Sheet 6 30 Figure 10.7-1b, Sheet 7 27 Figure 10.7-2, Sheet 1 27 Figure 10.7-2, Sheet 2 25 Figure 10.7-2, Sheet 3 25 Figure 10.7-2, Sheet 4 30 Figure 10.7-2, Sheet 5 25 Figure 10.7-2, Sheet 6 29 10.8-1 and 10.8-2 19 10.9-1 30 10.9-2 30 10.9-3 27 10.9-4 30 Figure 10.9-1a, Sheet 1 30 Figure 10.9-1a, Sheet 2 30 Figure 10.9-1a, Sheet 3 30 24

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 10.9-1b 23 Figure 10.9-2a 30 Figure 10.9-2b 28 Figure 10.9-2c 28 Figure 10.9-2d 30 Figure 10.9-2e 30 Figures 10.9-3 and 10.9-4 (Deleted Sheet) 22 10.10-1 through 10.10-4 27 Figure 10.10-1a 30 Figure 10.10-1b 27 Figure 10.10-1c 28 Figure 10.10-1d 29 Figure 10.10-2 29 Figure 10.10-3, Sheet 1 16 Figure 10.10-3, Sheet 2 28 Figure 10.10-3, Sheet 3 29 Figure 10.10-3, Sheet 4 26 Figure 10.10-4 (Deleted Sheet) 16 Figures 10.10-4a and 10.10-4b (Deleted Sheet) 22 10.11-1 through 10.11-3 27 10.11-4 30 10.11-5 through 10.11-7 27 Table 10.11-1 27 Figure 10.11-1 (Deleted Sheet) 16 Figures 10.11-1a and 10.11-1b (Deleted Sheet) 16 Figures 10.11-2 through 10.11-12 (Deleted Sheet) 16 10.12-1 through 10.12-7 22 10.12-8 28 10.12-9 and 10.12-10 22 Figure 10.12-1 28 Figure 10.12-2 (Deleted Sheet) 16 Figure 10.12-2a 25 Figure 10.12-2b 27 Figure 10.12-2c 23 Figure 10.12-3 30 Figure 10.12-4 25 Figure 10.12-5 28 Figure 10.12-6 21 Figure 10.12-7 28 Figure 10.12-8 28 25

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 10.12-9 30 10.13-1 and 10.13-2 17 Figure 10.13-1, Sheets 1 and 2 (Deleted Sheet) 17 10.14-1 29 10.14-2 29 10.14-3 27 10.14-4 29 10.14-5 27 Figure 10.14-1, Sheet 1 25 Figure 10.14-1, Sheet 2 30 Figure 10.14-1, Sheet 3 22 Figure 10.14-2a 25 Figure 10.14-2b 28 Figure 10.14-3 16 Figure 10.14-4, Sheet 1 28 Figure 10.14-4, Sheet 2 28 Figure 10.14-4, Sheet 3 28 Figure 10.14-4, Sheet 4 28 Figure 10.14-4, Sheet 5 27 Figure 10.14-4, Sheet 6 27 10.15-1 19 10.16-1 through 10.16-4 24 10.16-5 29 10.17-1 through 10.17-4 19 Table 10.17-1, Sheets 1 through 3 18 Table 10.17-2 18 Figure 10.17-1a 28 Figure 10.17-1b 28 Figure 10.17-1c, Sheet 1 22 Figure 10.17-1c, Sheet 2 27 Figure 10.17-1d 25 Figure 10.17-1e 29 Figure 10.17-1f 24 Figure 10.17-2 16 10.18-1 27 10.18-2 29 10.18-3 through 10.18-6 27 10.18-7 28 10.18-8 27 Figures 10.18-1 22 26

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 10.18-2 19 Figure 10.18-3 (Deleted Sheet) 16 Figure 10.18-4 22 Figures 10.18-5a and 10.18-5b (Deleted Sheet) 16 Figure 10.18-6 (Deleted Sheet) 16 10.19-1 27 10.20-1 29 10.20-2 22 10.21-1 through 10.21-4 23 Table 10.21-1 16 Figure 10.21-1 16 Figure 10.21-2 24 Figure 10.21-3 17 Figure 10.21-4 30 Figure 10.21-5 22 Figure 10.21-6 25 10.22-1 and 10.22-2 22 10.23-1 through 10.23-4 23 10.24-1 and 10.24-2 28 10.25-1 and 10.25-2 28 Chapter 11 11.0-i and 11.0-iv 29 11.1-1 28 Figure 11.1-1a 30 Figure 11.1-1b 29 Figure 11.1-1c 30 Figure 11.1-1d 30 Figure 11.1-1e 30 Figure 11.1-1f 30 11.2-1 and 11.2-2 29 11.2-3 28 11.2-4 23 11.3-1 28 11.3-2 29 11.4-1 30 11.4-2 25 11.5-1 28 11.6-1 and 11.6-2 30 11.6-3 28 27

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 11.6-4 and 11.6-5 30 11.6-6 and 11.6-7 27 Table 11.6-1 (deleted) 28 Figure 11.6-1 30 Figure 11.6-2 26 Figure 11.6-3, Sheet 1 30 Figure 11.6-3, Sheet 2 26 Figure 11.6-3, Sheet 3 26 Figure 11.6-3, Sheet 4 30 Figure 11.6-3, Sheet 5 30 Figure 11.6-4 30 Figure 11.6-5 29 Figure 11.6-6 28 11.7-1 25 11.7-2 28 11.7-3 25 Figure 11.7-1 25 Figure 11.7-2 27 Figure 11.7-3 27 11.8-1 28 11.8-2 29 11.8-3 and 11.8-4 28 Figure 11.8-1, Sheet 1 28 Figure 11.8-1, Sheet 2 27 Figure 11.8-1, Sheet 3 30 Figure 11.8-1, Sheet 4 27 Figure 11.8-1, Sheet 5 30 Figure 11.8-1, Sheet 6 27 11.9-1 through 11.9-5 25 Table 11.9-1 17 Figure 11.9-1a 29 Figure 11.9-1b, Sheet 1 28 Figure 11.9-1b, Sheets 2 and 3 28 Figure 11.9-2 30 Figure 11.9-3 (Deleted Sheet) 22 Figure 11.9-4 29 Figure 11.9-5 29 28

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Chapter 12 12.0-i through 10.0-vi 29 12.1-1 28 12.2-1 through 12.2-20 27 12.2-21 28 12.2-22 and 12.2-20 27 12.2-21 28 12.2-22 and 12.2-23 27 12.2-26 through 12.2-28 27 12.2-29 29 12.2-30 26 12.2-31 through 12.2-43 27 12.2-44 29 12.2-45 30 12.2-46 27 12.2-47 30 12.2-48 through 12.2-49 27 12.2-50 29 12.2-51 through 12.-62 27 12.2-63 through 12.2-75 29 12.2-76 through 12.2-89 27 12.2-90 28 Tables 12.2-1 through 12.2-12 27 Tables 12.2-13 through 12.2-15 (Deleted Sheets) 27 Table 12.2-16 27 Tables 12.2-16.1 through 12.2-16.3 (Deleted Sheets) 27 Table 12.2-16.4 27 Table 12.2-17 27 Tables 12.2-17.1A through 12.2-17.1C 27 Tables 12.2-18 through 12.2-22 (Deleted Sheets) 27 Table 12.2-23 27 Tables 12.2-24 and 12.2-25 29 Tables 12.2-26 through 12.2-30, Sheet 2 27 Tables 12.2-31 and 12.2-32 (Deleted Sheets) 27 Tables 12.2-33 through 12.2-39, Sheet 2 27 Tables 12.2-40 through 12.2-46, Sheet 2 27 Figure 12.2-1 (Deleted Sheet) 17 Figure 12.2-2 23 Figure 12.2-2a (Deleted Sheet) 16 Figures 12.2-2b through 12.2-2d 16 29

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 12.2-3 22 Figure 12.2-4 16 Figure 12.2-5 24 Figures 12.2-6 through 12.2-8 16 Figures 12.2-9 through 12.2-20 (Deleted Sheet) 16 Figures 12.2-21 and 12.2-22a 16 Figures 12.2-22b and 12.2-22c 28 Figure 12.2-22d (Deleted Sheet) 21 Figure 12.2-23 22 Figure 12.2-24, Sheet 1 24 Figure 12.2-24, Sheet 2 23 Figure 12.2-24, Sheet 3 16 Figure 12.2-25, Sheet 1 23 Figure 12.2-25, Sheets 2 and 3 16 Figure 12.2-26 16 Figure 12.2-27 (Deleted Sheet) 16 Figures 12.2-27A through 12.2-27C 16 Figures 12.2-28 through 12.2-39 (Deleted Sheet) 16 Figures 12.2-40 through 12.2-42 16 Figure 12.2-43 27 Figure 12.2-44 16 Figure 12.2-45 27 Figures 12.2-46 and 12.2-47 16 Figures 12.2-48 and 12.2-49 30 Figures 12.2-50 and 12.2-51 16 Figure 12.2-52 17 Figures 12.2-53 through 12.2-56 16 Figures 12.2-57 through 12.2-59 (Deleted Sheet) 16 Figure 12.2-60 23 Figure 12.2-61 17 Figures 12.2-62 and 12.2-63 16 Figures 12.2-64 28 Figure 12.2-65 19 Figures 12.2-66 and 12.2-67 18 Figure 12.2-68 (Deleted Sheet) 21 Figure 12.2-69 26 Figure 12.2-70 26 Figures 12.2-71a through 12.2-71c 16 Figure 12.2-72a 26 Figure 12.2-72b 30 30

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 12.2-72c 17 Figure 12.2-73 25 Figure 12.2-74 26 Figures 12.2-75a through 12.2-75b, Sheet 3 16 Figure 12.2-76 19 Figure 12.2-77 (Deleted Sheet) 19 Figures 12.2-77a and 12.2-77b 19 Figure 12.2-78 (Deleted Sheet) 19 Figure 12.2-79 16 Figure 12.2-80 21 Figure 12.2-81 16 Figure 12.2-82 17 Figure 12.2-83 16 Figure 12.2-84 22 Figure 12.2-85 25 12.3-1 28 12.3-2 through 12.3-4 18 Table 12.3-1 17 Chapter 13 13.0-i through 13.0-iii 26 13.1-1 and 13.1-2 16 13.2-1 through 13.2-5 17 Figures 13.2-1 through 13.2-8 (Deleted Sheets) 16 13.3-1 through 13.3-2 24 13.4-1 through 13.4-57 18 Figure 13.4-1 16 13.5-1 through 13.5-6 22 13.5-7 and 13.5-8 30 13.5-9 through 13.5-16 22 13.5-17 and 13.5-18 30 13.5-19 through 13.5-74 22 Tables 13.5-1 through 13.5-6 16 Figure 13.5-1, Sheets 1 and 2 16 Figure 13.5-2, Sheets 1 through 3 16 13.6-1 through 13.6-5 27 Figure 13.6-1, Sheets 1 and 2 (Deleted Sheet) 17 13.7-1 (Deleted Sheet) 17 13.8-1 and 13.8-2 23 13.9-1 and 13.9-2 19 31

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number 13.10-1 through 13.10-3 26 13.10-4 28 13.10-5 through 13.10-10 26 Chapter 14 14.0-i through 14.0-ix 29 14.1-1 29 14.2-1 16 14.3-1 21 14.4-1 and 14.4-2 28 14.4-3 through 14.4-7 21 Table 14.4-1, Sheets 1 and 2 16 Table 14.4-2 17 Figures 14.4-1 and 14.4-2 16 14.5-1 29 14.5-2 through 14.5-5 28 14.5-6 through 14.5-10 29 14.5-11 through 14.5-16 28 14.5-17 and 14.5-18 29 14.5-19 through 14.5-20 28 14.5-21 29 14.5-22 through 14.5-26 28 14.5-27 29 14.5-28 through 14.5-36 28 Table 14.5-1 28 Table 14.5-2 29 Figures 14.5-1 through 14.5-4 (Deleted Sheet) 19 Figures 14.5-5 (deleted) 28 Figures 14.5-5a through 14.5-5c 28 Figure 14.5-6 (deleted) 28 Figure 14.5-6a 28 Figure 14.5-6b 28 Figure 14.5-6c 28 Figure 14.5-7a (deleted) 28 Figure 14.5-7b (deleted) 28 Figure 14.5-8 (deleted) 28 Figure 14.5-9 (deleted) 28 Figure 14.5-9a 28 Figure 14.5-9b 28 Figure 14.5-9c 28 32

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 14.5-10a (deleted) 28 Figure 14.5-10b (deleted) 28 Figure 14.5-11 (deleted) 28 Figure 14.5-11a 28 Figure 14.5-11b 28 Figure 14.5-11c 28 Figure 14.5-11d 28 Figure 14.5-12a 28 Figure 14.5-12b 28 Figure 14.5-12c 28 Figure 14.5-13a (deleted) 28 Figure 14.5-13b (deleted) 28 Figure 14.5-14a (deleted) 28 Figure 14.5-14b (deleted) 28 Figure 14.5-15a 28 Figure 14.5-15b 28 Figure 14.5-16a (deleted) 28 Figure 14.5-16b (deleted) 28 Figure 14.5-17a (deleted) 28 Figure 14.5-17b (deleted) 28 Figure 14.5-17c (deleted) 28 Figure 14.5-18a (deleted) 28 Figure 14.5-18b (deleted) 28 Figure 14.5-19a (deleted) 28 Figure 14.5-19b (deleted) 28 Figure 14.5-20a (deleted) 28 Figure 14.5-20b (deleted) 28 Figure 14.5-21a (deleted) 28 Figure 14.5-21b (deleted) 28 Figure 14.5-22a (deleted) 26 Figure 14.5-22b (deleted) 26 Figure 14.5-22c (deleted) 28 Figure 14.5-22d (deleted) 28 Figure 14.5-22e (deleted) 28 Figure 14.5-22f (deleted) 28 Figures 14.5-23a through 14.5-23b (deleted) 28 Figure 14.5-24a (deleted) 26 Figure 14.5-24b (deleted) 28 Figures 14.5-24c through 14.5-24f 28 Figure 14.5-25a (deleted) 28 33

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figure 14.5-25b (deleted) 26 Figures 14.5-25c through 14.5-25f 28 Figure 14.5-26a (deleted) 28 Figure 14.5-26b (deleted) 28 Figure 14.5-27a (deleted) 28 Figure 14.5-27b (deleted) 28 Figure 14.5-28 (deleted) 28 Figure 14.5-28a 28 Figure 14.5-28b 28 Figure 14.5-28c 28 Figure 14.5-29 (deleted) 28 Figure 14.5-30 (deleted) 28 Figure 14.5-30a 28 Figure 14.5-30b 28 Figure 14.5-30c 28 Figure 14.5-31 (deleted) 28 Figure 14.5-31a 28 Figure 14.5-31b 28 Figure 14.5-31c 28 14.6-1 28 14.6-2 27 14.6-3 and 14.6-4 28 14.6-5 through 14.6-8 27 14.6-9 28 14.6-10 29 14.6-11 30 14.6-12 and 14.6-13 27 14.6-14 28 14.6-15 30 14.6-16 27 14.6-17 28 14.6-18 through 14.6-23 27 14.6-24 28 14.6-25 27 14.6-26 and 14.6-27 28 14.6-28 27 14.6-29 30 Table 14.6-1, Sheets 1 and 2 30 Table 14.6-2 (Deleted Sheet) 27 Tables 14.6-3 through 14.6-6 (deleted) 28 34

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table 14.6-7 27 Table 14.6-8, Sheet 1 30 Table 14.6-8, Sheet 2 30 Table 14.6-9 (Deleted Sheet) 27 Figures 14.6-1 through 14.6-6 (deleted) 28 Figure 14.6-7 and 14.6-8 17 Figure 14.6-9 and 14.6-10 (deleted) 28 Figures 14.6-11 and 14.6-12 (Deleted Sheet) 26 Figure 14.6-13 (Deleted Sheet) 16 Figures 14.6-14 through 14.6-18 (Deleted Sheet) 26 14.7-1 16 14.8-1 through 14.8-26 18 Tables 14.8-1 through 14.8-6 16 Figure 14.8-1 16 14.9-1 (deleted) 28 Tables 14.9-1 and 14.9-2, Sheets 1 through 4 28 14.10-1 through 14.10-18 (deleted) 28 Figures 14.10-1 through 14.10-20 28 14.11-1 through 14.11-38 (deleted) 28 Table 14.11-1 through 14.11-9 (deleted) 28 Table 14.11-2 (deleted) 28 Table 14.11-3 through 14.11-8, Sheet 2 28 Tables 14.11-9 28 Figures 14.11-1 through 14.11-10 28 Figure 14.11-11 through Figure 14.11-18 (deleted) 28 14.12-1 through 14.12-3 28 14.12-4 29 14.12-5 29 14.12-6 29 14.12-7 through 14.12-16 28 14.12-17 29 14.12-18 29 14.12-19 29 14.12-20 14.12-21 Tables 14.12-1 through 14.12-5 28 Table 14.12-6 29 Figure 14.12-1 through 14.12-14 28 14.13-1 29 Table 14.13-1 29 35

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Appendix A A.0-i and A.0-ii 26 A.0-1 through A.0-46 18 Tables A.0-1 through A.0-7, Sheets 1 through 3 16 Tables A.0-8 and A.0-9 16 Appendix B B.0-1 (Submitted Under Separate Cover) 17 Appendix C C.0.i through C.0-iii 29 C.0-1 through C.0-5 27 C.0-6 28 C.0-7 through C.0-17 27 C.0-18 28 C.0-19 through C.0-25 27 C.0-26 through C.0-30 28 C.0-31 27 C.0-32 28 C.0-33 27 C.0-34 and C.0-35 28 C.0-36 through C.0-44 27 C.0-45 and C.0-46 30 C.0-47 through C.0-49 27 C.0-50 30 C.0-51 through C.0-57 27 C.0-58 30 Tables C.2-1 and C.2-2, Sheets 1 through 3 27 Tables C.2-3 and C.2-4 27 Table C.3-1A, Sheets 1 and 2 27 Table C.3-1B, Sheets 3 and 4 27 Table C.3-1C, Sheets 5 and 6 27 Table C.3-1A, 1B, 1C, Sheets 7 and 8 27 Table C.3-2, Sheets 1 through 5 27 Table C.4-1, Sheets 1 through 3 27 Table C.4-1, Sheet 4 28 Table C.4-1, Sheets 5 through 8 27 Table C.4-1, Sheet 9 28 Table C.4-2, Sheets 1 through 25 27 36

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Table C.4-2, Sheet 26 29 Table C.4-2, Sheets 27 through 37 27 Table C.5-1 27 Appendix D D.0-i and D.0-ii 16 D.0-1 and D.0-2 21 Figures D.0-1 and D.0-2 (Deleted Sheets) 16 Appendix E E.0-i 16 E.0-1 17 Appendix F F.0-i and F.0-ii 29 F.0-1 through F.0-5 27 F.0-6 29 F.0-7 and F.0-8 27 F.0-9 and F.0-10 30 F.0-11 through F.0-16 27 F.0-17 29 F.0-18 through F.0-22 27 F.0-23 28 F.0-24 and F.0-25 27 F.0-26 29 F.0-27 through F.0-29 27 Appendix G G.0-i through G.0-v 26 G.0-1 through G.0-45 16 Table G.0-1, Sheets 1 and 2 16 Matrices 1 and 2, Sheets 1 and 2 16 Matrices 3, State A through F 16 Figures G.0-1 through G.0-27 16 Figures G.0-28a and G.0-28b 16 Figures G.0-29 through G.0-30b 16 Figures G.0-31 through G.0-35b 16 Figures G.0-36a through G.0-36e 16 Figures G.0-37 through G.0-41c 16 Figures G.0-42a and G.0-42b 16 37

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figures G.0-43 through G.0-44c 16 Figures G.0-45 and G.0-46 16 Appendix H H.0-i (Deleted Sheet) 16 Appendix I I.0-i and I.0-ii 26 I.0-1 through I.0-57 16 Appendix J J.0-i 16 J.0-1 17 J.0-2 16 Appendix K K.0-i 16 K.0-1 and K.0-2 16 Appendix L L.0-i 16 L.0-1 and L.0-2 16 Appendix M M.0-i through M.0-iii 25 M.0-1 26 M.0-2 28 M.0-3 26 M.0-4 28 M.0-5 26 M.0-6 28 M.0-7 through M.0-12 26 M.0-13 28 M.0-14 through 20 26 M.0-21 and M.0-22 28 M.0-23 through M.0-27 26 M.0-28 28 Tables M.0-1 and Table M.0-2 (Deleted Sheets) 16 Figure M.0-1 16 Figure M.0-2 17 38

BFN-30 EFFECTIVE PAGE LISTING Page Number Amendment Number Figures M.0-3 through M.0-6 16 Figure M.0-7 17 Appendix N N.0-i 30 N.0-ii 30 Pages N.1-1 through N.1-83 30 Pages N.2-1 through N.2-106 30 Pages N.3-1 through N.3-85 30 Appendix O O.0-i and O.0-ii 30 O.1-1 through 0.1-7 27 O.0-8 30 O.0-9 through O.0-11 27 O.1-12 29 O.1-13 through 0.1-15 27 O.2-1 30 O.2-2 and O.2-3 27 O.3-1 through O.3-8 27 O.4-1 and O.4-2 27 O.4-3 and O.4-4 30 O.5-1 27 39

BFN-30

1.0 INTRODUCTION

AND

SUMMARY

Project Identification 1.1 Definitions 1.2 Methods of Technical Presentation 1.3 Classification of BWR Systems, Criteria, 1.4 and Requirements for Safety Evaluation Principal Design Criteria 1.5 Plant Description 1.6 Comparison of Principal Design Characteristics 1.7 Summary of Radiation Effects 1.8 Plant Management 1.9 Quality Assurance Program 1.10 Identification-Resolution of Construction Permit 1.11 Concern-Summary General Conclusions 1.12 2.0 SITE Summary Description 2.1 Site Description 2.2 Meteorology 2.3 Hydrology, Water Quality, and Aquatic Biology 2.4 Geology and Seismology 2.5 Environmental Radiological Monitoring Program 2.6 3.0 REACTOR Summary Description 3.1 Fuel Mechanical Design 3.2 Reactor Vessel Internals Mechanical Design 3.3 Reactivity Control Mechanical Design 3.4 Control Rod Drive Housing Supports 3.5 Nuclear Design 3.6 Thermal and Hydraulic Design 3.7 Standby Liquid Control System 3.8 4.0 REACTOR COOLANT SYSTEM Summary Description 4.1 Reactor Vessel and Appurtenances Mechanical Design 4.2 Reactor Recirculation System 4.3 Nuclear System Pressure Relief System 4.4 Main Steam Line Flow Restrictor 4.5 Main Steam Isolation Valves 4.6 1

BFN-30 Reactor Core Isolation Cooling System 4.7 Residual Heat Removal System (RHRS) 4.8 Reactor Water Cleanup System 4.9 Nuclear System Leakage Rate Limits 4.10 Main Steam Lines, Feedwater Piping, and Drains 4.11 Inservice Inspection and Testing 4.12 5.0 CONTAINMENT Summary Description 5.1 Primary Containment System 5.2 Secondary Containment System 5.3 6.0 EMERGENCY CORE COOLING SYSTEMS Safety Objective 6.1 Safety Design Basis 6.2 Summary Description-Emergency Core Cooling Systems 6.3 Description 6.4 Safety Evaluation 6.5 Inspection and Testing 6.6 7.0 CONTROL AND INSTRUMENTATION Summary Description 7.1 Reactor Protection System 7.2 Primary Containment Isolation System 7.3 Emergency Core Cooling Control and Instrumentation 7.4 Neutron Monitoring System 7.5 Refueling Interlocks 7.6 Reactor Manual Control System 7.7 Reactor Vessel Instrumentation 7.8 Recirculation Flow Control System 7.9 Feedwater Control System 7.10 Pressure Regulator and Turbine-Generator Control 7.11 Process Radiation Monitoring 7.12 Area Radiation Monitoring System 7.13 Drywell Leak Detection Radiation Monitoring System 7.14 Health Physics Laboratory Radiation Monitoring Equipment 7.15 Process Computer System 7.16 Deleted 7.17 Backup Control System 7.18 Anticipated Transient Without Scram 7.19 Instrument Setpoint Methodology 7.20 2

BFN-30 8.0 ELECTRICAL POWER SYSTEMS Summary Description 8.1 Generators 8.2 Transmission System 8.3 Normal Auxiliary Power System 8.4 Standby AC Power Supply and Distribution 8.5 250-V DC Power Supply and Distribution 8.6 120-V AC Power Supply and Distribution 8.7 Auxiliary DC Power Supply and Distribution 8.8 Safety Systems Independence Criteria and Bases for Electrical Cable Installation 8.9 Station Blackout 8.10 9.0 RADIOACTIVE WASTE CONTROL SYSTEMS Summary Description 9.1 Liquid Radwaste System 9.2 Solid Radwaste System 9.3 Deleted 9.4 Gaseous Radwaste System (Modified) 9.5 10.0 AUXILIARY SYSTEMS Summary Description 10.1 New Fuel Storage 10.2 Spent Fuel Storage 10.3 Tools and Servicing Equipment 10.4 Fuel Pool Cooling and Cleanup System 10.5 Reactor Building Closed Cooling Water System 10.6 Raw Cooling Water System 10.7 Raw Service Water System 10.8 RHR Service Water System 10.9 Emergency Equipment Cooling Water System 10.10 Fire Protection Systems 10.11 Heating, Ventilating and Air-Conditioning Systems 10.12 Demineralized Water System 10.13 Control and Service Air Systems 10.14 Potable Water and Sanitary Systems 10.15 Equipment and Floor Drainage Systems 10.16 Process Sampling Systems 10.17 Plant Communications System 10.18 Lighting System 10.19 Auxiliary Boiler System 10.20 Postaccident Sampling System 10.21 Auxiliary Decay Heat Removal System 10.22 Hydrogen Water Chemistry System (HWC) 10.23 3

BFN-30 Supplemental Diesel Generator System 10.24 Emergency High Pressure Makeup (EHPM) System 10.25 11.0 POWER CONVERSION SYSTEMS Summary Description 11.1 Turbine-Generator 11.2 Main Condenser System 11.3 Main Condenser Gas Removal and Turbine Sealing Systems 11.4 Turbine Bypass System 11.5 Condenser Circulating Water System 11.6 Condensate Filter-Demineralizer System 11.7 Condensate and Reactor Feedwater Systems 11.8 Condensate Storage and Transfer Systems 11.9 12.0 STRUCTURES AND SHIELDING Summary Description 12.1 Principal Structures and Foundations 12.2 Shielding and Radiation Protection 12.3 13.0 CONDUCT OF OPERATIONS Organizational Structure for the Conduct of Operations 13.1 Organization and Responsibility 13.2 Training Programs 13.3 Preoperational Test Program 13.4 Startup and Power Test Program 13.5 Normal Operations 13.6 Deleted 13.7 Operational Review and Audits 13.8 Refueling Operations 13.9 Refueling Test Program 13.10 14.0 PLANT SAFETY ANALYSIS Analytical Objective 14.1 Unacceptable Safety Results for Abnormal Operational Transients 14.2 Unacceptable Safety Results for Accidents 14.3 Approach to Safety Analysis 14.4 Analyses of Abnormal Operational Transients - Uprated 14.5 Analysis of Design Basis Accidents - Uprated 14.6 Conclusions 14.7 Analytical Methods 14.8 4

BFN-30 Deleted 14.9 Deleted 14.10 Deleted 14.11 Analysis of the Primary Containment Response 14.12 Anticipated Transients Without Scram with Instability (ATWSI) 14.13 APPENDIX A. CONFORMANCE TO AEC PROPOSED GENERAL DESIGN CRITERIA Summary Description A.1 Criterion Conformance A.2 APPENDIX B. TECHNICAL SPECIFICATIONS AND TECHNICAL REQUIREMENTS MANUAL APPENDIX C. STRUCTURAL QUALIFICATION OF SUBSYSTEMS AND COMPONENTS Scope C.1 Loading Conditions, Definitions, and Overview C.2 Piping and Pipe Supports C.3 Major Components C.4 Primary Containment System and Penetrations C.5 Equipment C.6 Heating, Ventilation, and Air Conditioning (HVAC) Ductwork and Supports C.7 Control of Heavy Loads C.8 References C.9 APPENDIX D. QUALITY ASSURANCE PLAN FOR THE BROWNS FERRY NUCLEAR PLANT Quality Assurance During Design and Construction D.1 General Electric Quality System for BWR Nuclear Steam Supply Projects D.2 Quality Assurance Program for Station Operation D.3 5

BFN-30 APPENDIX E. SITE GASEOUS RELEASE RATE LIMIT CALCULATION Site Gaseous Release Rate Limits E.1 APPENDIX F. UNIT SHARING AND INTERACTIONS Introduction F.1 Scope F.2 References F.3 Criteria F.4 List of Shared Features F.5 Description of Shared Conventional Systems F.6 Description of Shared Class I Seismic Features, Structures, Safeguards Systems and Supporting Auxiliary Systems F.7 APPENDIX G. PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS Analytical Objective G.1 Basis for Selecting Operational Requirements G.2 Basis for Selection of Surveillance Test Frequencies for Nuclear Safety Systems and Engineered Safeguards G.3 Method of Analysis G.4 Analysis and Results G.5 Conclusion G.6 APPENDIX H. CORE THERMAL DESIGN Deleted APPENDIX I. IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERNS Summary Description I.1 Areas Specified in the Browns Ferry AEC-ACRS Construction Permit Reports I.2 Areas Specified in the AEC-Staff Construction Permit-Safety Evaluation Reports I.3 Areas Specified in Other Related AEC-ACRS Construction Permit and Operating License Reports I.4 Areas Specified in Other Related AEC-Staff Construction Permit or Operating License Safety Evaluation Reports I.5 6

BFN-30 Summary Conclusions I.6 References I.7 APPENDIX J. REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 1 Introduction J.1 Design and Fabrication Requirements - Summary J.2 APPENDIX K. REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 2 Introduction K.1 Design and Fabrication Requirements - Summary K.2 APPENDIX L. REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 3 Introduction L.1 Design and Fabrication Requirements - Summary L.2 APPENDIX M. REPORT ON PIPE FAILURES OUTSIDE CONTAINMENT IN THE BROWNS FERRY NUCLEAR PLANT Introduction M.1 Pressure and Analyses M.2 Piping Design Philosophy M.3 Breaks Postulated and Loading Effects Considered M.4 Pipe Rupture Loads M.5 Pipe Break Assumptions, Analysis, and Break Locations M.6 Structural Analysis M.7 Effects on Safety-Related Components and Structures M.8 Additional Work M.9 Summary and Conclusions M.10 References M.11 7

BFN-30 APPENDIX N. RELOAD LICENSING REPORT Browns Ferry, Unit 1 Cycle 15 Reload Safety Report N.1-1 Browns Ferry Unit 2 Cycle 23 Reload Safety Analysis N.2-1 Browns Ferry Unit 3 Cycle 21 Reload Safety Analysis N.3-1 APPENDIX O. AGING MANAGEMENT PROGRAMS Aging Management Programs O.1 Plant Specific Aging Management Programs O.2 Time-Limited Aging Management Programs O.3 10 CFR 54.37(b) New Identified Systems, Structures, and Components O.4 References O.5 8

BFN-29

1.0 INTRODUCTION

AND

SUMMARY

TABLE OF CONTENTS 1.1 Project Identification .................................................................................................................. 1.1-1 1.1.1 Identification and Qualification of Contractors ............................................................ 1.1-1 1.1.2 Licensing Basis Documents ....................................................................................... 1.1-3 1.2 Definitions.................................................................................................................................. 1.2-1 1.3 Methods of Technical Presentation ........................................................................................... 1.3-1 1.3.1 Purpose ..................................................................................................................... 1.3-1 1.3.2 Radioactive Material Barrier Concept ........................................................................ 1.3-1 1.3.3 Organization of Contents ........................................................................................... 1.3-1 1.3.4 Format Organization of Sections................................................................................ 1.3-2 1.3.5 Power Level Basis for Analysis of Abnormal Operational Transients and Accidents ............................................................................................................ 1.3-3 1.4 Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation ................... 1.4-1 1.4.1 Introduction ................................................................................................................ 1.4-1 1.4.2 Classification Basis .................................................................................................... 1.4-1 1.4.3 Use of the Classification Plan .................................................................................... 1.4-2 1.5 Principal Design Criteria ............................................................................................................ 1.5-1 1.5.1 Principal Design Criteria Classification-by-Classification ........................................... 1.5-1 1.5.2 Principal Design Criteria, System-By-System ............................................................ 1.5-7 1.6 Plant Description ....................................................................................................................... 1.6-1 1.6.1 General ...................................................................................................................... 1.6-1 1.6.2 Nuclear Safety Systems and Engineered Safeguards ............................................... 1.6-9 1.6.3 Special Safety Systems ............................................................................................. 1.6-15 1.6.4 Process Control and Instrumentation ......................................................................... 1.6-15 1.6.5 Auxiliary Systems....................................................................................................... 1.6-19 1.6.6 Shielding .................................................................................................................... 1.6-22 1.6.7 Implementation of Loading Criteria ............................................................................ 1.6-22 1.7 Comparison of Principal Design Characteristics ........................................................................ 1.7-1 1.7.1 Nuclear System Design Characteristics ..................................................................... 1.7-1 1.7.2 Power Conversion Systems Design Characteristics .................................................. 1.7-1 1.7.3 Electrical Power Systems Design Characteristics ...................................................... 1.7-1 1.7.4 Containment Design Characteristics .......................................................................... 1.7-1 1.7.5 Structural Design Characteristics ............................................................................... 1.7-2 1.7.6 Discussion of Core Design Improvement ................................................................... 1.7-2 1.0-i

BFN-29

1.0 INTRODUCTION

AND

SUMMARY

TABLE OF CONTENTS (Cont'd) 1.8 Summary of Radiation Effects ................................................................................................... 1.8-1 1.8.1 Normal Operation ....................................................................................................... 1.8-1 1.8.2 Abnormal Operational Transients .............................................................................. 1.8-1 1.8.3 Accidents ................................................................................................................... 1.8-1 1.9 Plant Management .................................................................................................................... 1.9-1 1.10 Quality Assurance Program ....................................................................................................... 1.10-1 1.11 Identification-Resolution of Construction Permit Concern - Summary ....................................... 1.11-1 1.11.1 General ...................................................................................................................... 1.11-1 1.12 General Conclusions ................................................................................................................. 1.12-1 1.0-ii

BFN-29 INTRODUCTION AND

SUMMARY

LIST OF TABLES Table Title 1.3-1 List of FSAR Engineering Drawings 1.3-2 Engineering Drawings Cross-Reference List 1.4-1 BWR Safety Engineering Concept for Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.4-2A Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.4-2B Classification of BWR Systems, Criteria, and Requirements for Safety Evaluation 1.7-1 Comparison of Nuclear System Design Characteristics 1.7-2 Comparison of Power Conversion Systems Design Characteristics 1.7-3 Comparison of Electrical Power System Design Characteristics 1.7-4 Comparison of Containment Design Characteristics 1.7-5 Comparison of Containment Design Characteristics 1.11-1 Browns Ferry Nuclear Plant Topical Reports Submitted to the AEC in Support of Docket 1.11-2 Browns Ferry Nuclear Plant AEC-ACRS Concerns - Resolutions 1.11-3 Browns Ferry Nuclear Plant AEC-Staff Concerns - Resolutions Units 1 and 2 1.11-4 AEC-Staff Concerns - Resolutions Unit 3 1.11-5 AEC-ACRS Concerns On Other Dockets - Resolutions 1.11-6 AEC-ACRS Concerns On Other Dockets - Capability for Resolution 1.0-iii

BFN-29 INTRODUCTION AND

SUMMARY

LIST OF ILLUSTRATIONS Figure Title 1.2-1 Relationship Between Safety Action and Protective Action 1.2-2 Relationship Between Protective Functions and Protective Actions 1.2-3 Relationships Between Different Types of Systems, Actions, and Objectives 1.3-1 Piping and Instrument Symbols 1.3-2 General Symbols Flow Diagram 1.6-1 Equipment Plans - Roof 1.6-2 Equipment Plans - Elevations 664 and 639 1.6-3 sht 1 Equipment Plans - Elevations 621.25 and 617 (Unit 1) 1.6-3 sht 2 Equipment Plans - Elevations 621.25 and 617 (Unit 2) 1.6-4 Equipment Plans - Elevations 606 and 604 1.6-5 Equipment Plans - Elevations 593 and 586 1.6-6 Equipment Plans - Elevations 565 and 557 1.6-7 Equipment Plans - Elevations 541.5 and 519 1.6-8 sht 1 Equipment - Transverse Section (Unit 1) 1.6-8 sht 2 Equipment - Transverse Section (Unit 2) 1.6-8 sht 3 Equipment - Transverse Section (Unit 3) 1.6-9 Equipment - Longitudinal Section 1.6-10 Equipment - Longitudinal Sections 1.6-11 Equipment Plans - Roof and Elevations 664 and 639 1.6-12 Equipment Plans - Elevations 621.25, 617, 606 and 604 1.6-13 Equipment Plans - Elevations 593, 586, 565, and 557 1.6-14 Equipment Plans - Elevations 541.5 and 519 1.6-15 Equipment Plans - Roof Elevations 664 and 639 1.6-16 Equipment Plans - Roof Elevations 621.25 and 617 1.6-17 Equipment Plans - Elevations 557, 565, 541.5 and 519 1.6-18 Equipment Refueling Floor Laydown Space 1.6-19 Equipment Refueling Floor Laydown Space 1.6-20 Deleted 1.6-21 Deleted 1.6-22 Deleted 1.6-23 General Arrangement Plan - Elevations 595, 580, and 578 1.6-24 General Arrangement Plans - Elevations 565 and 546 1.0-iv

BFN-29 INTRODUCTION AND

SUMMARY

LIST OF ILLUSTRATIONS (Cont'd) Figure Title 1.6-25 General Arrangement - Sections 1.6-26 Equipment - Plans and Sections 1.6-27 Equipment - Plan 1.6-28 Reactor Heat Balance - 3952 MWt 1.6-29 sht 1 Turbine-Generator Heat Balance - Rated Power (Unit 2) 1.6-29 sht 2 Turbine-Generator Heat Balance - Rated Power (Unit 3) 1.6-29 sht 3 Turbine-Generator Heat Balance - Rated Power (Unit 1) 1.6-30 General Plant Systems Flow Diagram 1.0-v

BFN-28 INTRODUCTION AND

SUMMARY

1.1 PROJECT IDENTIFICATION This Final Safety Analysis Report is in support of the application of the Tennessee Valley Authority (TVA), herein designated as the applicant, for facility operating licenses for a three-unit nuclear power plant located at the Browns Ferry site in Limestone County, Alabama, for initial power levels up to 3293 MWt each, under Section 104(b) of the Atomic Energy Act of 1954, as amended, and the regulations of the Atomic Energy Commission set forth in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50). The FSAR is now maintained up-to-date and used as a complete and accurate description of the Browns Ferry Nuclear Plant as constructed and as modified since. The facility is designated as the Browns Ferry Nuclear Plant, hereinafter referred to as the plant. Commercial operation of each unit began on the following dates: unit one on August 1, 1974, unit two on March 1, 1975, and unit three on March 1, 1977. Browns Ferry Nuclear Plant, Units 1, 2, and 3 were subsequently uprated by five percent from 3293 MWt to 3458 MWt. In August 2017, license amendments were issued for all three units for a core thermal power uprate from 3458 MWt to 3952 MWt. As used throughout this document, Atomic Energy Commission (AEC) is equivalent to the Nuclear Regulatory Commission (NRC) formed under the Energy Reorganization Act of 1974. 1.1.1 Identification and Qualification of Contractors Irrespective of any contractual responsibilities with any suppliers, the Tennessee Valley Authority is the sole applicant for the facility licenses and as owner and applicant, is responsible for the design, construction, and operation of the plant. 1.1.1.1 Applicant The TVA power system is one of the largest in the United States. TVA is primarily a wholesaler of power, operating generating plants, and transmission facilities, but no retail distribution systems. The TVA transmission system contains over 17,000 miles of lines. TVA supplies power over an area of about 80,000 square miles in parts of seven southeastern states, containing more than 2.3 million residential, farm, commercial, and industrial customers. 1.1-1

BFN-28 TVA has pioneered in erecting large generating units. Examples are the 1,150-megawatt unit at the Paradise Steam Plant; the 1,300-megawatt units at the Cumberland Steam Plant; and the two 1,170-megawatt units at the Sequoyah Nuclear Plant; and two 1,170-megawatt units at the Watts Bar Nuclear Plant. A total of over 67 individual steam generating units have been designed, constructed, and placed in operation by TVA in the past 35 years. Much of TVA's experience has been gained from early and continuing participation in nuclear power studies. In 1946, TVA participated in the Daniels power pile study at Oak Ridge and the work of the Parker Committee, which surveyed prospects of nuclear power application. In 1953, TVA started developing a nuclear power staff and began a more detailed study of possible uses of nuclear power on its own system. In 1960, TVA agreed to operate the Experimental Gas-Cooled Reactor for AEC at Oak Ridge, and developed a technical and operating staff. Many of these trained and experienced people were assigned to TVA engineering and operating organizations that have been or are directly involved in the planning, design, construction, and operation of the Browns Ferry Nuclear Plant. 1.1.1.2 Engineer-Constructor TVA acts as its own engineer-constructor. Since 1949, TVA has designed and constructed a number of projects including twelve major coal-fired steam plants, consisting of 63 individual generating units. TVA has an experienced, competent nuclear plant design organization, including a large number of engineers with many years of steam plant experience. TVA also has a similarly experienced construction organization which has had extensive experience in the construction of large steam plants. A comprehensive quality assurance program has been developed to assure that the plant has been designed and constructed and will be operated to adequate standards of quality. 1.1.1.3 Nuclear Steam Supply System Supplier General Electric Company was awarded a contract to design, fabricate, and deliver the nuclear steam supply system and nuclear fuel for the plant, as well as to provide technical direction for installation and startup of this equipment. General Electric (GE) has been engaged in the development, design, construction, and operation of boiling water reactors since 1955. Operating boiling water reactors designed and built by General Electric include the Vallecitos Boiling Water Reactor, Dresden Unit 1, Humboldt Bay, Big Rock Point, KRB (Germany), KAHL (Germany), JPDR (Japan), SENN (Italy), Oyster Creek Unit 1, and Dresden Unit 2. Among the domestic reactors of General Electric design are Millstone Point Unit 1, Dresden Unit 3, Quad-Cities Units 1 and 2, Monticello Unit 1, Vermont Yankee Unit 1, Peach Bottom Units 2 and 3, Pilgrim, Hatch Units 1 and 2, Brunswick Units 1 and 2, 1.1-2

BFN-27 Cooper, Duane Arnold, and Fitzpatrick. Thus, General Electric has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation, startup, and support of the normal operation of the reactor. 1.1.1.4 Turbine-Generator Supplier The applicant awarded a contract to General Electric to design, fabricate, and deliver the turbine generators for the plant as well as to provide technical assistance for installation and startup of this equipment. General Electric has a long history in the application of turbine generators in nuclear power stations going back to the inception of nuclear facilities for the production of electrical power and has furnished the turbine-generator units for most of its BWR nuclear steam supply contracted stations. General Electric was competent to design, fabricate and deliver the turbine-generator units and to provide technical assistance for the installation and startup of this equipment. 1.1.2 Licensing Basis Documents The following documents are typical documents submitted periodically to NRC. Implementation of changes to these documents without NRC approval may be controlled by regulation or the plant license. The following list provides references for the review and approval requirements for the listed documents. Document Regulation Or Requirement Updated Final Safety 10 CFR 50.71(e) Analysis Report (UFSAR) Technical Requirements 10 CFR 50.59 Manual and Tech. Spec. Bases Organizational Topical Report 10 CFR 50.54(a)(3) Quality Assurance Plan 10 CFR 50.54(a)(3) Offsite Dose Calculation Manual (ODCM) Tech. Spec., Section 5.5.1 Physical Security Plan 10 CFR 50.54(p) Radiological Emergency Plan (REP) 10 CFR 50.54(q) Core Operating Limits Report (COLR) Tech. Spec., Section 5.6.5 1.1-3

BFN-25 1.2 DEFINITIONS The following definitions apply to the terms used in the Safety Analysis Report.

1. Radioactive Material Barrier - A radioactive material barrier includes the systems, structures, or equipment that, together, physically prevent the uncontrolled release of radioactive materials. The four barriers are identified as follows:

a. Reactor Fuel Barrier - The uranium dioxide fuel is sealed in a zirconium cladding tube.

b. Nuclear System Process Barrier - The nuclear system process barrier includes the systems of vessels, pipes, pumps, tubes, and similar process equipment that contain the steam, water, gases, and radioactive materials coming from, going to, or in communication with the reactor core. The actual boundaries of the nuclear system process barrier depend upon the status of plant operation.

For example, process system isolation valves, when closed, form part of the barrier. The steam-jet ejector offgas path forms a planned process opening in the barrier during power operation. Because the nuclear system process barrier is designed to be divided by isolation valve action into two major sections under certain conditions, this barrier is considered in two parts as follows: (1) Nuclear system primary barrier - This barrier includes the reactor vessel and attached piping out to and including the second isolation valve in each attached pipe. In various codes and standards used in the industry, this barrier is sometimes referred to as the "primary system pressure boundary," (2) Nuclear system secondary barrier - This barrier is that portion of the nuclear system process barrier not included in the nuclear system primary barrier.

c. Primary Containment - The primary containment is defined as the drywell in which the reactor vessel is located, the pressure suppression chamber, and process line reinforcements out to the outermost containment isolation valve outside valve outside the containment wall.

Portions of the nuclear system process barrier may become part of the primary containment, depending upon the location of a postulated failure. For example, a closed main steam isolation valve is part of the 1.2-1

BFN-25 primary containment barrier when the postulated failure of the main steam line is inside the primary containment.

d. Secondary Containment - The secondary containment is the reactor building, which completely encloses the primary containment. The reactor building ventilation system and the standby gas treatment system constitute controlled process openings in this barrier.

2. Radioactive Material Barrier Damage - Radioactive material barrier damage is defined as an unplanned, undesirable breach in a barrier, except that the operation of a main steam relief valve does not constitute barrier damage.

3. Nuclear System - The nuclear system generally includes those systems most closely associated with the reactor vessel which are designed to contain or be in communication with the water and steam coming from or going to the reactor core. The nuclear system includes the following:

Reactor vessel Reactor vessel internals Main steam lines from reactor vessel to the isolation valves outside the primary containment Neutron monitoring system Reactor recirculation system Control rod drive system Residual heat removal system Reactor core isolation cooling system Core standby cooling systems Reactor water cleanup system Feedwater system piping between the reactor vessel and the first valve outside the primary containment.

4. Safety - The word "safety," when used to modify such words as objective, design basis, action, and system, indicates that the objective, design basis, action, or system is related to concerns considered to be of primary safety significance, as opposed to the plant mission to generate electrical power.

Thus, the word "safety" is used to identify aspects of the plant which are considered to be of primary importance with respect to safety.

5. Power Generation - The phrase "power generation," when used to modify such words as objective, design basis, action, and system, indicates that the objective, design basis, action, or system is related to the mission of the plant - to generate electrical power - as opposed to concerns considered to be of primary safety importance. Thus, the phrase "power generation" is used to identify aspects of the plant which are not considered to be primary importance with respect to safety.

1.2-2

BFN-25

6. Operational - The adjective "operational," along with its noun and verb forms, is used in reference to the working or functioning of the plant, in contrast to the design of the plant.

7. Scram - Scram refers to the rapid insertion of control rods. A scram is initiated either automatically in response to the detection of undesirable conditions or manually by the control room operator.

8. Limiting Safety System Setting (LSSS) - The limiting safety system setting is a setting on instrumentation which initiates the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

9. Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

10. Safety Limit - The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of the unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

11. Normal Operation - Normal operation is normal plant operation under planned conditions in absence of significant abnormalities. Operations subsequent to an incident (transient, accident, or special event) are not considered planned operations until the actions taken in the plant are identical to those which would be used had the incident not occurred. The established planned operations can be considered as a chronological sequence: refueling outage, achieving criticality, heatup, power operation, achieving shutdown, cooldown, and refueling outage.

The following planned operations are identified:

a. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage.

1.2-3

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b. Achieving Criticality - Achieving criticality includes all the plant actions which are normally accomplished in bringing the plant from a condition in which all control rods are fully inserted to a condition in which nuclear criticality is achieved and maintained.

c. Heatup - Heatup begins where achieving criticality ends and includes all plant actions which are normally accomplished in approaching nuclear system rated temperature and pressure by using nuclear power (reactor critical). Heatup extends through warmup and synchronization of the turbine generator.

d. Power Operation - Power operation begins where heatup ends and includes continued operation of the plant at power levels in excess of heatup power.

e. Achieving Shutdown - Achieving shutdown begins where power operation ends and includes all plant actions normally accomplished in achieving nuclear shutdown (more than one rod subcritical) following power operation.

f. Cooldown - Cooldown begins where achieving shutdown ends and includes all plant actions normally accomplished in the continued removal of decay heat and the reduction of nuclear system temperature and pressure.

12. Incident - An incident is any event--abnormal operational transient, accident, special event, or other event, not considered as part of planned operation.

13. Abnormal Operational Transient - An abnormal operational transient includes the events following a single equipment malfunction or a single operator error that is reasonably expected during the course of plant operations. Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors initiating the events in this category.

14. Abnormal Occurrence - Abnormal occurrence refers to the occurrence of any plant condition that:

a. Causes any abnormal operational transient, or

b. Violates a limiting condition for operation as established in the technical specifications, or

c. Exceeds a limiting safety system setting as established in the technical specifications, or 1.2-4

BFN-25

d. Causes any uncontrolled or unplanned release of radioactive material from the site.

15. Accident - An accident is a single event, not reasonably expected during the course of plant operations, that has been hypothesized for analysis purposes or postulated from unlikely but possible situations, and that causes or threatens a rupture of a radioactive material barrier. A pipe rupture qualifies as an accident; a fuel cladding defect does not.

16. Design Basis Accident - A design basis accident is a hypothesized accident the characteristics and consequences of which are utilized in the design of those systems and components pertinent to the preservation of radioactive material barriers and the restriction of radioactive material release from the barriers. The potential radiation exposures resulting from a design basis accident are greater than any similar accident postulated from the same general accident assumptions. For example, the consequences of a complete severance of a recirculation loop line are more severe than those resulting from any other single pipeline failure inside the primary containment.

17. Special Event - A special event that neither qualifies as an abnormal operational transient nor an accident but that is postulated to demonstrate some special capability of the plant or its systems.

18. Safety Action - A safety action is an ultimate action in the plant that is essential to the avoidance of specified conditions considered to be of primary safety significance. The specified conditions are those that are most directly related to the ultimate limits on the integrity of the radioactive material barriers or the release of radioactive material. There are safety actions associated with planned operation, abnormal operational transients, accidents, and special events. Safety actions include such actions as the indication to the operator of the values of certain process variables, reactor scram, core standby cooling, and reactor shutdown from outside the control room. See Figures 1.2-1 and 1.2-3 and Tables 1.4-2A and 1.4-2B.

19. Power Generation Action - A power generation action is an action in the plant that is essential to the avoidance of specified conditions considered to be of primary significance to the plant mission--the generation of electrical power.

The specified conditions are those that are directly related to the following:

a. The ability to carry out the plant mission--the generation of electrical power--through planned operation,

b. The avoidance of conditions that would limit the ability of the plant to generate electrical power, and 1.2-5

BFN-25

c. The avoidance of conditions that would prevent or hinder the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

There are power generation actions associated with planned operation, abnormal operational transients, accidents, and special events. See Figure 1.2-3.

20. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.

21. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

22. Safety System - A safety system is any system, group of systems, component, or group of components the actions of which are essential to accomplishing a safety action. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

23. Process Safety System - A process safety system is a safety system the actions of which are essential to a safety action required during planned operation. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

24. Nuclear Safety System - A nuclear safety system is a safety system the actions of which are essential to a safety action required in response to an abnormal operational transient. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

25. Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

26. Protection System - Protection system is a generic term that may be applied to nuclear safety systems and engineered safeguards. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

27. Special Safety System - A special safety system is a safety system the actions of which are essential to a safety action required in response to a special event. See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

28. Power Generation System - A power generation system is any system the actions of which are not essential to a safety action, but which are essential to a power generation action. Power generation systems are provided for any of the following purposes:

a. To carry out the mission of the plant--generate electrical power--through planned operation, 1.2-6

BFN-25

b. To avoid conditions which would limit the ability of the plant to generate electrical power, and

c. To facilitate and expedite the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

See Figure 1.2-3 and Table 1.4-2A and 1.4-2B.

29. Safety Objective - A safety objective describes in functional terms the purpose of a system or component as it relates to conditions considered to be of primary significance to the protection of the public. This relationship is stated in terms of radioactive material barriers or radioactive material release.

The only systems that have objectives are safety systems. See Figure 1.2-3.

30. Power Generation Objective - A power generation objective describes in functional terms the purpose of a system or component as it relates to the mission of the plant. This includes objectives that are specifically established so the plant can fulfill the following purposes:

a. The generation of electrical power through planned operation,

b. The avoidance of conditions that would limit the ability of the plant to generate electrical power, and

c. The avoidance of conditions that would prevent or hinder the return to conditions permitting the use of the plant to generate electrical power following an abnormal operational transient, accident, or special event.

See Figure 1.2-3. A system or piece of equipment has a power generation objective if it is a power generation system. A safety system can have a power generation objective, in addition to a safety objective, if parts of the system are intended to function for power generation purposes.

31. Analytical Objective - An analytical objective describes the purpose or intent of a portion of the Safety Analysis Report presenting an analysis.

32. Safety Design Basis - The safety design basis for a safety system states in functional terms the unique design requirements which establish the limits within which the safety objective shall be met. A power generation system may have a safety design basis which states in functional terms the unique design requirements that ensure that neither planned operation nor operational failure by the system results in conditions for which plant safety actions would be inadequate.

1.2-7

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33. Power Generation Design Basis - The power generation design basis for a power generation system states in functional terms the unique design requirements that establish the limits within the power generation objective shall be met. A safety system may have a power generation design basis which states in functional terms the unique design requirements which establish the limits within which the power generation objective for the system shall be met.

34. Safety Evaluation - A safety evaluation is an evaluation that shows how the system satisfies the safety design basis. A safety evaluation is performed for those systems having a safety design basis. Safety evaluations form the bases for the technical specifications and establish why specific safety limitations are imposed.

35. Power Generation Evaluation - A power generation evaluation is an evaluation that shows how the system satisfies some or all of the power generation design bases. Because power generation evaluations are not directly pertinent to public safety, they are generally not included. However, where a system or component has both safety and power generation objectives, a power generation evaluation can be used to clarify the safety versus power generation capabilities.

36. Operational Nuclear Safety Requirements - An operational nuclear safety requirement is a limitation or restriction on either the value of a process variable or the operability of a plant system. Such operational nuclear safety requirements must be observed in the operation (not necessarily at power) of the plant to satisfy specified operational nuclear safety criteria. The aggregate of all operational nuclear safety requirements defines an operational framework within which actual plant operations must remain.

37. Rated Power - Rated power refers to operation at a reactor thermal power of 3952 MWt. Rated power is also termed 100 percent power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power.

38. Design Power - Design power refers to the power level used in safety and licensing analyses which support operation at rated power. Power corresponds to 3952 MWt. For radiological dose analyses provided in Section 14.6, design power has been assumed to be 3952 MWt.

1.2-8

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39. Single Failure - A single failure is a failure that can be ascribed to a single causal event. Single failures are considered in the design of certain systems and are presumed in the evaluations of incidents to investigate the ability of the plant to respond in the required manner under degraded conditions. The nature of single causal event to be presumed depends on the risk of the event being evaluated. Reasonably expected single failures are presumed as the cause of abnormal operational transients. Single failures of passive equipment are assumed sometimes to be the causes of accidents. Safety actions essential in response to abnormal operational transients and accidents must be carried out in spite of single failures in active equipment.

In any case, a single failure includes the multiple effects resulting from the single causal event.

40. Operable - Operability - A system, subsystem, division, component, or device shall be Operable or have operability when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

41. Operating - A system or component is operating when it is performing its intended functions in its required manner.

42. Operating Cycle - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit.

43. Deleted.

44. Mode - A Mode shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning, specified as follows, with fuel in the reactor vessel.

1.2-9

BFN-25 MODE TITLE REACTOR MODE AVERAGE REACTOR SWITCH POSITION COOLANT TEMPERATURE (F) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot NA Standby 3 Hot Shutdown(a) Shutdown 212 4 Cold Shutdown(a) Shutdown 212 5 Refueling(b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned. (b) One or more reactor vessel head closure bolts less than fully tensioned.

45. Deleted.

46. Deleted.

47. Deleted.

48. Deleted.

49. Deleted.

50. Deleted.

51. Place in Isolated Condition - Place in isolated condition means conduct an uninterrupted normal isolation of the reactor from the main (turbine) condenser including the closure of the main steam isolation valves.

52. Deleted.

53. Deleted.

54. Deleted.

55. Refueling Outage - Refueling outage is a period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and 1.2-10

BFN-25 surveillance, a refueling outage shall mean a regular scheduled refueling outage.

56. Core Alteration - Core Alteration shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be Core Alterations:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of Core Alterations shall not preclude completion of movement of a component to a safe position.

57. Risk - Risk is the product of the probability of an event and the adverse consequences of the event.

58. Reliability - Reliability is the probability that an item will perform its specified function without failure for a specified time period in a specified environment.

59. Unreliability - Unreliability is the probability that a component or system will fail to perform its specified action for a specified time period in a specified environment. (The sum of reliability and unreliability equals unity.)

60. Availability - Availability is the probability that a system will be functional at any randomly selected instant.

61. Unavailability - Unavailability is the probability that component or system will not be functional at any randomly selected instant. (The sum of availability and unavailability equals unity.)

62. Repair Rate - The repair rate is the number of repairs completed per unit time.

63. Failure Rate - The failure rate is the number of failures per unit time.

64. Test Duration - The test duration is the elapsed time between test initiation and test termination.

65. Test Interval - The test interval is the elapsed time between the initiation of identical tests.

1.2-11

BFN-25

66. Active Component - A device characterized by an expected significant change of state or discernible mechanical motion in response to an imposed design basis load demand upon the system. Examples are: switch, relay, valve not remaining in a stationary position, pressure switch, turbine, transistor, motor, damper, pump, and analog meter.

67. Passive Component - A device characterized by an expected negligible change of state or negligible mechanical motion in response to an imposed design basis load demand upon the system. Examples are: cable, piping, valve in stationary position, resistor, capacitor, fluid filter, indicator lamp, cabinet, and case.

68. Operating Basis Earthquake - That earthquake which produces the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional.

69. Design Basis Earthquake - That earthquake which produces the vibratory ground motion for which those features of the nuclear power plant necessary to shut down the reactor and maintain the plant in a safe condition without undue risk to the health and safety of the public are designed to remain functional.

70. Deleted.

71. Deleted.

72. Probable Maximum Flood - The Probable Maximum Flood (PMF) is the hypothetical flood (peak discharge, volume, and hydrograph shape) that is considered to be the most severe reasonable possible, based on comprehensive hydrometerological application of probable maximum precipitation, and other hydrologic factors favorable for maximum flood runoff, such as sequential storms and snowmelt. The PMF design level at the Browns Ferry site is 572.5 feet.

The term Maximum Possible Flood (MPF) has also been used in Browns Ferry design documents, however the preferred term for all Browns Ferry design is PMF. (See also Appendix 2.4.A, Probable Maximum Flood).

73. Emergency Core Cooling Systems (ECCS) are defined as:

a. High Pressure Coolant Injection System (HPCI),

b. Automatic Depressurization System,

c. Core Spray System, and 1.2-12

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d. Low Pressure Coolant Injection System (LPCI) (an operating mode of the Residual Heat Removal System).

The term Core Standby Cooling Systems (CSCS) has also been used in the FSAR, design documents, and plant procedures to describe the same systems. The terms ECCS and CSCS may be used interchangeably. 1.2-13

r-----7 I SAFETY I I SYSTEM B I IL _ _ _ _ ...JI r-----7 ,-----7 I SAFETY I PROTECTIVE I SAFETY I I SYSTEM A I ACTION 8 I SYSTEM C I I _____ ...JI L IL _____ ...JI PROTECTIVE PROTECTIVE ACT ION A ACTION C SAFETY ACTION CONCEPT I 1 STANDBY 7 A-C I I POWER I IL __ T__ SYSTEM ...JI PROVIDE ,-----7 A-C 1 - Low- - 7 I CORE I POWER 1 PRESSURE 1 I SPRAY I 1 COOLANT 1 I SYSTEM I L __ T__ ...J __ T__ ILINJECTION...JI PUMP WATER PUMP WATER TO CORE TO CORE CORE STBY

       ~ - - - - COOLING (LOW------~

PRESSURE) EXAMPLE AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Relationship between Safety Action end Protective Action FIGURE 1 .2-1

SAFETY SYSTEM PROTECTIVE PROTECTIVE FUNCTION A FUNCTION B INTRA-

     '-----..i-    SYSTEM   i..i---_,

ACTIONS PROTECTIVE ACTION CONCEPT HIGH PRESSURE COOLANT INJECTION SYSTEM HIGH REACTOR DRYWELL VESSEL LOW PRESSURE WATER LEVEL START SYSTEM OPEN VALVES START TURBINE PUMP PUMP WATER TO CORE EXAMPLE AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Relationship between Protective Functions and Protective Actions FIGURE 1 . 2-2

0111: SAP'ETT' DIR MISStON POw:R CENERA TJON J TI SAFETY SYSTtNS PIIOlECTlVE ACTJ(lfS SAfETY M'.:rtONS SAFETY 08JtCTIV[S NOT[S: I. ONLY TIO SYSTEMS C-, [A01 TYl'E Alt£ SHCIIN. TM(R[ WAY IE MOit( TNAM THIS IUIIIU Of SYSU:YS JN ANY CATEGOllr,

2. THERE MAY BE CASES IHERE THE SYSTEM LEVEL AC1'JON IS IDENJICAL TO THE ULTIMATE ACTION [N TH£ PLAHT IN SUCH A CA$[ THE JNTEIIICDIATE SYSTEM LEY!l ACflON NCEO NOTH IDENTJP'I[O.

AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT f INAL SAFETY ANALYSIS REPORT Re lot fon1hlp1 bet***n Offhr1nt Typ** of Sy11tems, Actions, and Object i ve1 F !CURE 1 . 2-J

BFN-19 1.3 METHODS OF TECHNICAL PRESENTATION 1.3.1 Purpose The original purpose of this Final Safety Analysis Report (FSAR) was to provide the technical information required by Section 50.34 of 10 CFR 50 to establish a basis for evaluation of the plant with respect to the issuance of facility operating licenses. The FSAR is to serve as the one document which will provide a complete and up-to-date description of the Browns Ferry Nuclear Plant as constructed and as modified since. In accordance with 10 CFR 50.71(e), as amended, the FSAR will be kept up-to-date through the issuance of amendments. An effective page listing will be placed just ahead of the FSAR Table of Contents. This listing will ensure that each copy in use within TVA is maintained in the most up-to-date condition possible. Except where otherwise specified, the information in this report is applicable to all three units of the Browns Ferry Nuclear Plant. 1.3.2 Radioactive Material Barrier Concept Because the safety aspects of this report pertain to the relationship between plant behavior under a variety of circumstances and the radiological effects on persons off site, the report is oriented to the radioactive material barriers. This orientation facilitates evaluation of the radiological effects of the plant on the environment. Thus, the presentation of technical information is considerably different from that which would be expected in an operational manual, maintenance manual, or nuclear engineer's handbook. The overriding consideration that determines the depth of detailed technical information presented about a system or component is the relationship of the system or component to the radioactive material barriers. Systems that must operate to preserve or limit the damage to the radioactive material barriers are described in the greatest detail. Systems that have little relationship to the radioactive material barriers are described only with as much detail as necessary to establish their functional role in the plant. 1.3.3 Organization of Contents The Final Safety Analysis Report is organized into 14 major chapters, each of which consists of a number of sections. A system for classifying the various aspects of the BWR with respect to safety is given in section 1.4 (fourth section in Chapter 1). This classification system is fundamental to assessing the adequacy of the plant with respect to the relative importances of different safety concerns. The principal architectural and engineering criteria, which define the broad frame of reference within which the plant is designed, are set forth in section 1.5. Section 1.6 presents a brief description of the plant in which the nuclear safety systems and engineered 1.3-1

BFN-19 safeguards are separated from the other plant systems, so that those systems essential to safety are clearly identified. Chapters 2 through 13 present detailed information about the design and operation of the plant. The nuclear safety systems and engineered safeguards are integrated into these sections according to system function (core standby cooling, control), system type (electrical, mechanical), or according to their relationship to a particular radioactive material barrier. Chapter 3 (Reactor) describes plant components and presents design details that are most pertinent to the fuel barrier. Chapter 4 (Reactor Coolant System) describes plant components and systems that are most pertinent to the nuclear system process barrier. Chapter 5 describes the primary and secondary containments. Thus Chapters 3, 4, and 5 are arranged according to the four radioactive material barriers. The remainder of the chapters group system information according to plant function (radioactive waste control, core standby cooling, power conversion, control) or system type (electrical, structures). Chapter 14 (Plant Safety Analysis) provides an overall safety evaluation of the plant which demonstrates both the adequacy of equipment designed to protect the radioactive material barriers and the ability of the safeguard features to mitigate the consequences of situations in which one or more radioactive material barriers are assumed damaged. 1.3.4 Format Organization of Sections Sections are numerically identified by representing their order of appearance in a chapter by two numbers separated by a decimal point; e.g., 3.4 is the fourth section in Chapter 3. Sections are further divided into subsections by numbers separated by decimal points (3.4.1, 3.4.1.1, etc.). Pages within each section are consecutively numbered (3.4-1, 3.4-2, etc.). Tables are identified by the section number followed by a decimal point and the number of the table according to its order of mention in the text; e.g., Table 7.5-3 is the third table of section 7.5. Drawings, pictures, sketches, curves, graphs, and engineering diagrams are identified as Figures and are numbered in the same manner as tables. Figures 1.3-1 and 1.3-2 defines the meanings of piping and instrumentation symbols used in the figures of this report. Table 1.3-1 provides a list of all design and plant system figures appearing in the FSAR with the FSAR figure number cross-referenced to the engineering drawing number. 1.3-2

BFN-19 The general organization of a section describing a system or component is as follows: Objective Design Basis Description Evaluation Inspection and Testing To clearly distinguish the safety versus operational aspects of a system, the objective, design basis, and evaluation titles are modified by the word "safety" or "power generation", according to the definitions given in Section 1.2. Systems that have safety objectives are safety systems. A safety evaluation is included only when the system has a safety design basis; the evaluation shows how the system satisfies the safety design basis. A power generation evaluation is included only when needed to clarify the safety versus power generation aspects of a system that has both safety and power generation functions. A nuclear safety operational analysis of the plant was performed to systematically identify the operational limitations or restrictions which were to be observed with regard to certain process variables and certain plant systems to satisfy specified nuclear safety operational criteria during the initial operational fuel cycle. The method used for this analysis is described in Appendix G. Subsequent nuclear safety operational analyses have been and will continue to be generated for the plant as they are necessitated by plant modifications. Sections presenting information on topics other than plant systems or components are arranged individually according to the subject matter so that the relationship between the subject and public safety is emphasized. Within each section of the text, applicable supporting technical material is referenced. References are cited either at the bottom of a page or at the end of a subsection. Most of the references are cited as a particular technical basis for BWR plant design and analysis, but some are specifically applicable to the Browns Ferry Nuclear Plant. The references in this category generally provide a full development and analysis of some aspect of GE BWR plant technology. These special references are incorporated by reference into the safety analysis report, thereby becoming part of the license application. 1.3.5 Power Level Basis for Analysis of Abnormal Operational Transients and Accidents For those abnormal operational transients and accidents for which high power operation increases the severity of the results, the analyses assume plant operation at design power as an initial condition. For those events for which an initial condition 1.3-3

BFN-19 of low or intermediate power level operation renders the most severe results, the analyses presented in this report represent the most severe case within the operating spectrum. 1.3-4

BFN-28 Table 1.3-1 Sheet 1 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 1.3-1 Piping and Instrument Symbols 104R900 1.3-2 General Symbols Flow Diagram 0-47E800-2 1.6-1 Equipment Plans - Roof 47E200-1 1.6-2 Equipment Plans - Elevations 664 and 639 0-47E200-2 1.6-3 Equipment Plans - Elevations 621.25 and 617 Sheet 1 1-47E200-3 Sheet 2 2-47E200-3 1.6-4 Equipment Plans - Elevations 606 and 604 0-47E200-4 1.6-5 Equipment Plans - Elevations 593 and 586 0-47E200-5 1.6-6 Equipment Plans - Elevations 565 and 557 0-47E200-6 1.6-7 Equipment Plans - Elevations 541.5 and 519 0-47E200-7 1.6-8 Equipment - Transverse Section Sheet 1 1-47E200-8 Sheet 2 2-47E200-8 Sheet 3 3-47E200-8 1.6-9 Equipment - Longitudinal Section 0-47E200-9 1.6-10 Equipment - Longitudinal Sections 0-47E200-10 1.6-11 Equipment Plans - Roof and Elevations 664 and 639 3-47E200-11 1.6-12 Equipment Plans - Elevations 621.25, 617, 606 and 604 0-47E200-12 1.6-13 Equipment Plans - Elevations 593, 586, 565, and 557 3-47E200-13 1.6-14 Equipment Plans - Elevations 541.5 and 519 3-47E200-14 1.6-15 Equipment Plans - Roof Elevations 664 and 639 0-47E200-15 1.6-16 Equipment Plans - Roof Elevations 621.25 and 617 0-47E200-16 1.6-17 Equipment Plans - Elevations 557, 565, 541.5 and 519 0-47E200-17 1.6-18 Equipment Refueling Floor Laydown Space 0-47E200-18 1.6-19 Equipment Refueling Floor Laydown Space 0-47E200-19 1.6-23 General Arrangement Plan - Elevations 595, 580, and 578 0-47W215-1 1.6-24 General Arrangement Plans - Elevations 565 and 546 0-47W215-2 1.6-25 General Arrangement - Sections 0-47E215-3 1.6-26 Equipment - Plans and Sections 0-47W220-1 1.6-27 Equipment - Plan 3-47E220-2 1.6-29 Turbine Generator Heat Balance Values Sheet 1 2-47K1110-13 Sheet 2 3-47K1110-13 Sheet 3 1-47K1110-32 1.6-30 General Plant Systems Flow Diagram 0-47E800-1 2.2-4 Location of Principal Plant Structures 0-10E201-7

BFN-25 Table 1.3-1 (Contd) Sheet 2 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 2.4A Figure 25 Channel Relocation West of Cooling Towers, Typical Sections 10H237 2.4A Figure 26 Channel Relocation West of Cooling Towers Sheet 1 0-10H243 Sheet 2 10H242 2.5-S1 Foundation Investigation Grid Low Level Radwaste Storage 0-10E151-7 2.5-17 Mechanical Instruments and Controls 0-47E600-121 2.5-19 Rock Excavation 41N703 3.4-8a CRD Hydraulic System - Mechanical Control Diagram Sheet 1 2-47E610-85-1 Sheet 2 2-47E2610-85-2 Sheet 3 3-47E610-85-1 Sheet 4 3-47E3610-85-2 Sheet 5 1-47E610-85-1 3.4-8b Control Rod Hydraulic System - Flow Diagram 0-47E820-1 3.4-8c Control Rod Drive Hydraulic System - Flow Diagram 2-47E820-2 3.4-8d CRD Hydraulic System - Mechanical Control Diagram 2-47E2610-85-5 3.4-8e Control Rod Drive Hydraulic System - Flow Diagram 3-47E820-2 3.4-8f Control Rod Drive Hydraulic System - Mechanical Control Diagram 3-47E610-85-5 3.4-8g CRD Hydraulic System - Mechanical Control Diagram 1-47E1610-85-2 3.4-8h CRD Hydraulic System - Mechanical Control Diagram 1-47E1610-85-5 3.8-1 Standby Liquid Control System Flow Diagram 2-47E854-1 3.8-2 Standby Liquid Control System Mechanical Control Diagram 2-47E610-63-1 3.8-3 Standby Liquid Control System - Flow Diagram 1-47E854-1 3.8-5 Standby Liquid Control System - Flow Diagram 3-47E854-1 3.8-6 Standby Liquid Control System - Mechanical Control Diagram 3-47E610-63-1 3.8-8 Standby Liquid Control System - Mechanical Control Diagram 1-47E610-63-1 4.2-1 Reactor Vessel 2-104R935-1 4.2-3 Reactor Vessel 3-104R935-1 4.2-4 Reactor Vessel 1-104R935-1 4.3-2a Nuclear Boiler Flow Diagram Sheet 1 1-47E817-1 Sheet 2 2-47E817-1 Sheet 3 3-47E817-1 4.4-6 T-Quencher for Safety/Relief Discharge 2-47W401-5 4.4-7 Mechanical Main Steam Relief Valve Vent Piping 3-47E401-5 4.4-8 Mechanical Main Steam Relief Valve Vent Piping 1-47W401-5 4.5-1 Primary Steam Piping 1-729E229-1 4.5-2 Primary Steam Piping 2-729E229-4 4.5-3 Primary Steam Piping 3-729E229-4 4.7-1a Reactor Core Isolation Cooling System Flow Diagram 2-47E813-1 4.7-1b Reactor Core Isolation Cooling System, Mechanical Control Diagram 2-47E610-71-1

BFN-25 Table 1.3-1 (Contd) Sheet 3 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 4.7-1c Reactor Core Isolation Cooling System - Flow Diagram 3-47E813-1 4.7-1d Reactor Core Isolation Cooling System - Mechanical Control Diagram 3-47E610-71-1 4.7-1e Reactor Core Isolation Cooling System - Mechanical Control Diagram 1-47E610-71-1 4.7-1f Reactor Core Isolation Cooling System - Flow Diagram 1-47E813-1 4.9-1 Reactor Water Cleanup System Flow Diagram 2-47E810-1 4.9-2 Reactor Water Cleanup Demineralizer Flow Diagram 2-47E837-1 4.9-3 Reactor Water Cleanup System, Mechanical Control Diagram 2-47E610-69-1 4.9-5 Reactor Water Cleanup System - Flow Diagram 3-47E810-1 4.9-6 Reactor Water Cleanup Demineralizer - Flow Diagram 3-47E837-1 4.9-7 Reactor Water Cleanup System - Mechanical Control Diagram 3-47E610-69-1 4.9-8 Reactor Water Cleanup System - Flow Diagram 1-47E810-1 4.9-9 Reactor Water Cleanup Demineralizer - Flow Diagram 1-47E837-1 4.9-10 Reactor Water Cleanup System - Mechanical Control Diagram 1-47E610-69-1 5.2-2a Primary Containment System, Mechanical Control Diagram Sheet 1 1-47E610-64-1 Sheet 2 2-47E610-64-1 Sheet 3 3-47E610-64-1 5.2-2b Primary Containment System, Mechanical Control Diagram 2-47E610-64-2 5.2-2c Primary Containment System, Mechanical Control Diagram 2-47E610-64-3 5.2-2d Primary Containment System - Mechanical Control Diagram 3-47E610-64-2 5.2-2e Primary Containment System - Mechanical Control Diagram 3-47E610-64-3 5.2-2f Primary Containment System - Mechanical Control Diagram 1-47E610-64-2 5.2-2g Primary Containment System - Mechanical Control Diagram 1-47E610-64-3 5.2-6a Containment Inerting System, Mechanical Control Diagram Sheet 1 1-47E610-76-1 Sheet 2 2-47E610-76-4 Sheet 3 2-47E610-76-1 Sheet 4 3-47E610-76-1 Sheet 5 3-47E610-76-4 Sheet 6 0-47E610-76-1 Sheet 7 1-47E1610-76-3 5.2-6b Primary Containment Cooling Temperature Monitoring System - Mechanical Control Diagram 1-47E610-80-1 5.2-6c Primary Containment Cooling Temperature Monitoring System, Mechanical Control Diagram 2-47E610-80-1 5.2-6d Primary Containment Cooling Temperature Monitoring System - Mechanical Control Diagram 3-47E610-80-1 5.2-7 Containment Atmosphere Dilution System, Flow Diagram Sheet 1 1-47E862-1 Sheet 2 2-47E862-1 Sheet 3 3-47E862-1 5.2-8 Containment Atmosphere Dilution System, Mechanical Control Diagram Sheet 1 1-47E610-84-1 Sheet 2 2-47E610-84-1 Sheet 3 3-47E610-84-1

BFN-25 Table 1.3-1 (Contd) Sheet 4 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 5.3-3a Heating and Ventilation Air Flow Diagram 1-47E865-1 5.3-3b Heating and Ventilating - Standby Gas Treatment System, Flow Diagram 0-47E865-11 5.3-3c Heating and Ventilating Air Flow, Flow Diagram 2-47E2865-12 5.3-3d Heating and Ventilation - Flow Diagram 3-47E865-12 5.3-9 Standby Gas Treatment System, Mechanical Control Diagram 0-47E610-65-1 6.4-1 HPCI System - Mechanical Control Diagram 2-47E610-73-1 6.4-2 Core Spray System, Flow Diagram 2-47E814-1 6.4-3 HPCI System - Mechanical Control Diagram 3-47E610-73-1 6.4-4 Core Spray System - Flow Diagram 3-47E814-1 6.4-5 HPCI System - Mechanical Control Diagram 1-47E610-73-1 6.4-6 Core Spray System, Flow Diagram 1-47E814-1 7.2-1 Reactor Protection System, Single Line 2-45E702-4 7.2-2 Reactor Protection System, Auxiliary Instrument Room Panel 2-791E167RF 7.2-3 Reactor Protection System, Single Line 1-45E701-3 7.2-7a Reactor Protection System Instrument Engineering Diagram 2-729E136-3 7.2-7b Reactor Protection System Instrument Engineering Diagram (Unit 3) 3-729E136-3 7.2-7c Reactor Protection System - Single Line 3-45E703-3 7.2-7d Reactor Protection System Instrument Engineering Diagram (Unit 1) 1-729E136-3 7.2-8 Reactor Protection System, Auxiliary Instrument Room Panel 1-791E167 7.2-9 Reactor Protection System Auxiliary Instrument Room Panel 3-791E167 7.3-1 Nuclear Boiler Flow Diagram Sheet 1 1-47E817-1 Sheet 2 2-47E817-1 Sheet 3 3-47E817-1 7.4-1b High Pressure Coolant Injection System, Flow Diagram Sheet 1 2-47E812-1 Sheet 2 3-47E812-1 Sheet 3 1-47E812-1 7.4-5a ECCS Preferred Pump Logic - Mechanical Control Diagram 2-47E610-75-3 7.4-5d Pre-ACD and Common ACD Signal - Mechanical Control Diagram 2-47E610-75-2 7.4-5i ECCS Preferred Pump - Mechanical Control Diagram 1-47E610-75-3 7.4-5l Pre-ACD and Com ACD Signal - Mechanical Control Diagram 1-47E610-75-2 7.4-5m ECCS Preferred Pump Logic - Mechanical Control Diagram 3-47E610-75-3 7.4-6a Residual Heat Removal System, Flow Diagram Sheet 1 2-47E811-1 Sheet 2 1-47E811-1 Sheet 3 3-47E811-1 7.4-6b Residual Heat Removal System, Mechanical Control Diagram Sheet 1 2-47E2610-74-1 Sheet 2 2-47E2610-74-2 Sheet 3 1-47E610-74-1A

BFN-25 Table 1.3-1 (Contd) Sheet 5 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. Sheet 4 3-47E610-74-1 Sheet 5 3-47E610-74-2 7.4-7b ECCS Preferred Pump Logic - Mechanical Control Diagram 2-47E610-74-3 7.4-7i Pre-ACD and Com ACD Signal - Mechanical Control Diagram 3-47E610-75-2 7.4-7p ECCS Preferred Pump - Residual Heat Removal System - Mechanical Control Diagram 1-47E610-74-3 7.5-1a Startup Range Neutron Monitoring System, Instrument Engineering Diagram 2-105E1512-1 7.5-1b Startup Range Neutron Monitoring System, Instrument Engineering Diagram 3-105E1512-1 7.5-1c Startup Range Neutron Monitoring System, Instrument Engineering Diagram 1-105E1512RE-1 7.5-2 SRM/IRM Neutron Monitoring Unit 729E946-1 7.5-11a Power Range Neutron Monitoring System, Instrument Engineering Diagram 2-105E1512-2 7.5-11b Power Range Neutron Monitoring System, Instrument Engineering Diagram 3-105E1512-2 7.5-11c Startup Range Neutron Monitoring System - Instrument Engineering Diagram 1-105E1512RE-2 7.5-13 Power Range Neutron Monitoring Unit 0-729E989-1 7.5-23a Neutron Monitoring System Physical Arrangement 0-729E761-1 7.5-23b Neutron Monitoring System Physical Arrangement 0-729E761-2 7.5-23c Neutron Monitoring System Physical Arrangement (Unit 1) 1-729E761-1 7.5-23d Neutron Monitoring System Physical Arrangement (Unit 1) 1-729E761-2 7.8-1 Sheet 1 - Reactor Feedwater System Mechanical Control Diagram 2-47E610-3-1 Sheet 2 - Reactor Water Recirculation System Mechanical Control Diagram 2-47E610-68-1 Sheet 3 - Reactor Feedwater System - Mechanical Control Diagram 3-47E610-3-1 Sheet 4 - Reactor Water Recirculation System - Mechanical Control Diagram 3-47E610-68-1 Sheet 5 - Reactor Feedwater System - Mechanical Control Diagram 1-47E610-3-1 Sheet 6 - Reactor Water Recirculation System - Mechanical Control Diagram 1-47E610-68-1 7.8-3 Reactor Vessel Temperature Monitoring System Physical Arrangement 0-47E600-91 7.10-2 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-1 7.10-3 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-2 7.10-4 Feedwater Control System, Mechanical Control Diagram 2-47E610-46-3 7.10-5 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-1 7.10-6 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-2 7.10-7 Feedwater Control System - Mechanical Control Diagram 3-47E610-46-3 7.10-8 Feedwater Control System - Mechanical Control Diagram 1-47E610-46-1 7.12-2a Radiation Monitoring System Mechanical Control Diagram Sheet 1 2-47E610-90-1 Sheet 2 2-47E610-90-2 Sheet 3 0-47E610-90-4 Sheet 4 0-47E610-90-20 Sheet 5 1-47E610-90-1 Sheet 6 3-47E610-90-1 Sheet 7 0-47E610-90-21 7.12-2b Radiation Monitoring System Mechanical Control Diagram Sheet 2 0-47E610-90-2

BFN-25 Table 1.3-1 (Contd) Sheet 6 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. Sheet 4 2-47E610-90-3 Sheet 5 1-47E610-90-3 Sheet 6 3-47E610-90-3 8.3-2 Electrical Equipment General Arrangement Plan 0-75W200 8.3-2a Browns Ferry Nuclear Arrangement of Transmission Lines LC48417-1 8.3-4 Gen. 1 and 500-kV Switchyard - Main Single Line 0-45E1504 8.3-5 Generator 2 and 500-kv Switchyard - Main Single Line 0-45E1505 8.3-6 500-kV Switchyard - Main Single Line 0-45E1506 8.3-6a 161-kV Switchyard - Main Single Line 0-45E506 8.4-1a Normal Auxiliary Power System Key Diagram 0-15E500-2 8.4-1b Standby Auxiliary Power System Key Diagram 0-15E500-1 8.4-2 Normal and Standby Auxiliary Power System Key Diagram 3-15E500-3 8.4-3 4160-V Shutdown Auxiliary Power Schematic Diagram 0-45E765-1 8.5-1 Diesel Generator Panel - One-Line Diagram 0-731E747-1 8.5-2 Sheet 1 - Diesel Starting Air System Diesel Generator A Flow and Control Diagram 0-47E861-1 Sheet 2 - Diesel Starting Air System Diesel Generator A Flow Diagram 0-47E861-1A Sheet 3 - Cooling System and Lubricating Oil System Standby Diesel Generator A Flow Diagram 0-47E861-5 Sheet 4 - Diesel Starting Air System Diesel Generator 3A- Flow Diagram 3-47E861-1 Sheet 5 - Diesel Starting Air System Diesel Generator 3A- Flow Diagram 3-47E861-1A Sheet 6 - Cooling System and Lubricating Oil System Diesel Generator 3A - Flow Diagram 3-47E861-5 8.5-3a Fuel Oil System - Flow Diagram 0-47E840-1 8.5-3b Fuel Oil System - Flow Diagram 0-47E840-3 8.5-4a 4160-V Shutdown Board A - Single Line 0-45E724-1 8.5-4b 4160-V Shutdown Board 3EA - Single Line 3-45E724-6 8.5-4c 4160-V Shutdown Board B - Single Line 0-45E724-2 8.5-4d 4160-V Shutdown Board C - Single Line 0-45E724-3 8.5-4e 4160-V Shutdown Board D - Single Line 0-45E724-4 8.5-4f 4160-V Shutdown Board 3EB - Single Line 3-45E724-7 8.5-4g 4160-V Shutdown Board 3EC - Single Line 3-45E724-8 8.5-4h 4160-V Shutdown Board 3ED - Single Line 3-45E724-9 8.5-5 480-V Shutdown Board 2A - Single Line 2-45E749-3 8.5-6 480-V Shutdown Board 2B - Single Line 2-45E749-4 8.5-7a 480-V Reactor MOV Board 2A - Single Line 2-45E751-1 8.5-7b 480-V Reactor MOV Board 2A - Single Line 2-45E751-2 8.5-7c 480-V Reactor MOV Board 1A - Single Line 1-45E751-1 8.5-7d 480-V Reactor MOV Board 1A - Single Line 1-45E751-2 8.5-7e 480-V Reactor MOV Board 3A - Single Line 3-45E751-1 8.5-7f 480-V Reactor MOV Board 3A - Single Line 3-45E751-2 8.5-8a 480-V Reactor MOV Board 2B - Single Line 2-45E751-3 8.5-8b 480-V Reactor MOV Board 2B - Single Line 2-45E751-4

BFN-25 Table 1.3-1 (Contd) Sheet 7 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 8.5-8c 480-V Reactor MOV Board 1B - Single Line 1-45E751-3 8.5-8d 480-V Reactor MOV Board 1B - Single Line 1-45E751-4 8.5-8e 480-V Reactor MOV Board 3B - Single Line 3-45E751-3 8.5-8f 480-V Reactor MOV Board 3B - Single Line 3-45E751-4 8.5-9a 480-V Reactor MOV Board 2C - Single Line 2-45E751-5 8.5-9b 480-V Reactor MOV Board 2C - Single Line 2-45E751-6 8.5-9c 480-V Reactor MOV Board 3C - Single Line 3-45E751-5 8.5-9d 480-V Reactor MOV Board 3C - Single Line 3-45E751-6 8.5-10 480-V Reactor MOV Board 2D - Single Line 2-45E751-8 8.5-11 480-V Reactor MOV Board 2E - Single Line 2-45E751-11 8.5-11a 480-V Reactor MOV Board 3D - Single Line 3-45E751-9 8.5-11c 480-V Control Bay Vent Board A - Single Line 0-45E736-1 8.5-11d 480-V Control Bay Vent Board B - Single Line 0-45E736-2 8.5-12a 480-V Diesel Auxiliary Board A - Single Line 0-45E732-1 8.5-12b 480-V Diesel Auxiliary Board A - Single Line 0-45E732-2 8.5-12c 480-V Diesel Auxiliary Board 3EA - Single Line 3-45E732-5 8.5-13a 480-V Diesel Auxiliary Board B - Single Line 0-45E732-3 8.5-13b 480-V Diesel Auxiliary Board B - Single Line 0-45E732-4 8.5-13c 480-V Diesel Auxiliary Board 3EB - Single Line 3-45E732-6 8.5-13d 480-V Standby Gas Treatment Bd - Single Line 0-45E733-1 8.5-13e 480-V Diesel Auxiliary Board B - Single Line 0-45E732-8 8.5-25 480-V Shutdown Board 1A - Single Line 1-45E749-1 8.5-26 480-V Shutdown Board 1B - Single Line 1-45E749-2 8.5-27 480-V Shutdown Board 3A - Single Line 3-45E749-5 8.5-28 480-V Shutdown Board 3B - Single Line 3-45E749-6 8.6-1a Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-1 8.6-1b Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-2 8.6-1c Instrument and Controls DC and AC Power Systems Key Diagram (Safeguards Information - Located in Drawing Control) 0-45W710-3 8.6-1d Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-4 8.6-1e Instrument and Controls DC and AC Power Systems Key Diagram 0-45E710-5 8.6-1f Instrument and Controls DC and AC Power System Key Diagram 0-45E710-7 8.6-2b Shutdown Boards 250-V Battery and Chargers - Single Line 0-45E709-1 8.6-2c Shutdown Boards 250-V Battery and Chargers - Single Line 3-45E709-2 8.6-3 Engineered Safeguards and RCIC 250-V DC System Separations Scheme Block Diagram 0-731E744-1 8.6-5 DC Board 9-9 One-Line Diagram 0-731E723-1 8.6-6 Control Room DC Board - Single Line 0-55E715-2 8.7-1 Instrumentation and Controls AC Power System, One-Line Diagram 0-731E752-1 8.7-3 Unit Preferred AC Power System, One-Line Diagram 0-731E751-1 8.7-4a AC Board 9-9 One-Line Diagram 0-731E753-1 8.7-4b AC Board 9-9 Preferred and Nonpreferred Loads One-Line Diagram Sheet 1 0-731E753-2

BFN-25 Table 1.3-1 (Contd) Sheet 8 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. Sheet 2 2-731E753-2 8.7-4c AC Board 9-9 Instrumentation and Control One-Line Diagram Sheet 1 1-731E753-3 Sheet 2 2-731E753-3 Sheet 3 3-731E753-3 8.7-4d Control Room DC Board - Single Line 0-55E715-1 9.2-3a Radwaste System - Flow Diagram 0-47E830-1 9.2-3b Radwaste System - Flow Diagram 0-47E830-2 9.2-3c Radwaste System - Flow Diagram 0-47E830-3 9.2-3d Radwaste System - Flow Diagram 0-47E830-4 9.2-3e Radwaste System - Flow Diagram 0-47E830-5 9.2-3f Radwaste System - Flow Diagram 0-47E830-6 9.2-3g Radwaste System - Flow Diagram 0-47E830-7 9.2-3h Radwaste System - Flow Diagram 0-47E830-8 9.2-3i Radwaste System - Flow Diagram 0-47E830-9 9.2-3j Radwaste System - Mechanical Control Diagram 0-47E610-77-1 9.2-3k Radwaste System - Mechanical Control Diagram 0-47E610-77-2 9.2-3l Radwaste System - Mechanical Control Diagram 0-47E610-77-3 9.2-3m Radwaste System - Mechanical Control Diagram 0-47E610-77-4 9.2-3n Radwaste System - Mechanical Control Diagram 0-47E610-77-5 9.2-3o Radwaste System - Mechanical Control Diagram 0-47E610-77-6 9.2-3p Radwaste System - Mechanical Control Diagram 0-47E610-77-7 9.2-3q Radwaste System - Mechanical Control Diagram 0-47E610-77-8 9.2-3r Radwaste System - Mechanical Control Diagram 0-47E610-77-9 9.2-3s Radwaste System - Mechanical Control Diagram 0-47E610-77-10 9.2-3t Radwaste System - Mechanical Control Diagram 0-47E610-77-11 9.5-1 Offgas System Flow Diagram Sheet 1 2-47E809-2 Sheet 2 2-47E809-3 Sheet 3 3-47E809-2 Sheet 4 3-47E809-3 Sheet 5 1-47E809-2 Sheet 6 1-47E809-3 9.5-2 Offgas System - Flow Diagram 3-47E809-4 9.5-3 Offgas System Flow Diagram 2-47E809-4 9.5-4 Offgas System Flow Diagram 1-47E809-4 10.3-1 High Density Spent Fuel Rack - Unit 1 1-C5445E-102 10.3-2 High Density Spent Fuel Rack - Unit 2 C5445E-103 10-3-3 High Density Spent Fuel Rack - Unit 3 3-C5445E-104 10.5-1a Fuel Pool Cooling System, Flow Diagram 2-47E855-1 10.5-1b Fuel Pool Cooling and Demineralizer System, Mechanical Control Diagram

BFN-25 Table 1.3-1 (Contd) Sheet 9 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. Sheet 1 2-47E610-78-1 Sheet 2 0-47E610-78-1 Sheet 3 1-47E610-78-1 Sheet 4 3-47E610-78-1 10.5-1c Fuel Pool Cooling System - Flow Diagram 1-47E855-1 10.5-1d Fuel Pool Cooling System - Flow Diagram 3-47E855-1 10.5-2 Fuel Pool Filter Demineralizer, Flow Diagram Sheet 1 2-47E832-1 Sheet 2 0-47E832-1 Sheet 3 1-47E832-1 Sheet 4 3-47E832-1 10.6-1a Reactor Building Closed Cooling Water System, Flow Diagram 1-47E822-1 10.6-1b Reactor Building Closed Cooling Water System, Flow Diagram 2-47E822-1 10.6-1c Reactor Building Closed Cooling Water System - Flow Diagram 3-47E822-1 10.7-1a Raw Cooling Water, Flow Diagram Sheet 1 1-47E844-1 Sheet 2 2-47E844-1 Sheet 3 3-47E844-1 10.7-1b Raw Cooling Water, Flow Diagram Sheet 1 1-47E844-2 Sheet 2 2-47E844-2 Sheet 3 0-47E844-3 Sheet 4 3-47E844-2 Sheet 5 2-47E844-3 Sheet 6 3-47E844-3 Sheet 7 1-47E844-3 10.7-2 Raw Cooling Water System, Mechanical Control Diagram Sheet 1 2-47E610-24-1C Sheet 2 2-47E610-24-1D Sheet 3 0-47E610-24-3 Sheet 4 3-47E610-24-1 Sheet 5 3-47E610-24-2 Sheet 6 1-47E610-24-1 10.9-1a RHR Service Water System, Flow Diagram Sheet 1 1-47E858-1 Sheet 2 2-47E858-1 Sheet 3 3-47E858-1 10.9-1b Raw, Potable, Demineralized Residual Heat Removal, Emergency 0-17W300-5 Equipment Cooling Water and Compressed Air 10.9-2a RHR Service Water System, Mechanical Control Diagram 2-47E610-23-1 10.9-2b RHR Service Water System, Mechanical Control Diagram 0-47E610-23-2 10.9-2c RHR Service Water System, Mechanical Control Diagram 0-47E610-23-3

BFN-25 Table 1.3-1 (Contd) Sheet 10 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 10.9-2d RHR Service Water System - Mechanical Control Diagram 3-47E610-23-1 10.9-2e RHR Service Water System - Mechanical Control Diagram 1-47E610-23-1 10.10-1a Emergency Equipment Cooling Water, Flow Diagram 1-47E859-1 10.10-1b Emergency Equipment Cooling Water, Flow Diagram 3-47E859-2 10.10-1c Emergency Equipment Cooling Water, Flow Diagram 2-47E859-1 10.10-1d Emergency Equipment Cooling Water - Flow Diagram 3-47E859-1 10.10-2 Emergency Equipment Cooling Water System, Mechanical Control Diagram 1-47E610-67-1 10.10-3 Emergency Equipment Cooling Water System, Mechanical Control Diagram Sheet 1 3-47E610-67-3 Sheet 2 2-47E610-67-2 Sheet 3 3-47E610-67-2 Sheet 4 0-47E610-67-2 10.12-1 Heating and Ventilating Air Flow, Flow Diagram 2-47E865-3 10.12-2a Ventilating and Air Conditioning Air Flow, Flow Diagram 0-47E865-4 10.12-2b Ventilation and Air Conditioning Air Flow, Flow Diagram 2-47E2865-4 10.12-2c Ventilating and Air Conditioning Air Flow, Flow Diagram 3-47E865-4 10.12-3 Heating and Air Conditioning Hot and Chilled Water, Flow Diagram 0-47E866-3 10.12-4 Heating and Ventilating Air Flow, Flow Diagram 0-47E865-6 10.12-5 Heating, Ventilating, and Air Conditioning Air Flow 0-47E865-8 10.12-6 Heating, Ventilating, and Air Conditioning Air Flow 3-47E865-8 10.12-7 Heating and Ventilating Air Flow System - Flow Diagram 1-47E865-3 10.12-8 Heating and Ventilating Air Flow - Flow Diagram 3-47E865-3 10.12-9 Flow Diagram Air Conditioning Chilled Water 0-47E866-9 10.14-1 Control Air System, Mechanical Control Diagram Sheet 1 1-47E610-32-1 Sheet 2 2-47E610-32-1 Sheet 3 3-47E610-32-1 10.14-2a Compressed Air, Station Service, Flow Diagram 0-47E845-1 10.14-2b Compressed Air, Station Service, Flow Diagram 0-47E845-2 10.14-4 Control Air System, Mechanical Control Diagram Sheet 1 - Control Air System - Mechanical Control Diagram 2-47E610-32-2 Sheet 2 - Control Air System - Flow Diagram 2-47E2847-5 Sheet 3 - Control Air System - Mechanical Control Diagram 3-47E610-32-2 Sheet 4 - Control Air System - Mechanical I & C - Flow Diagram 3-47E3847-5 Sheet 5 - Control Air System - Mechanical I & C - Flow Diagram 1-47E1847-6 Sheet 6 - Control Air System - Mechanical Control Diagram 1-47E610-32-2 10.17-1a Sampling and Water Quality Systems, Mechanical Control Diagram 2-47E610-43-1 10.17-1b Sampling and Water Quality Systems, Mechanical Control Diagram 2-47E610-43-2 10.17-1c Sampling and Water Quality Systems, Mechanical Control Diagram Sheet 1 0-47E610-43-3 Sheet 2 2-47E610-43-3

BFN-25 Table 1.3-1 (Contd) Sheet 11 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 10.17-1d Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-1 10.17-1e Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-2 10.17-1f Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-3 10.21-1 Flow Diagram - PASS 2-47E867-3 10.21-2 Mechanical Control Diagram - PASS 2-47E610-43-5 10.21-3 Sampling and Water Quality System - Flow Diagram 3-47E867-3 10.21-4 Sampling and Water Quality System - Mechanical Control Diagram 3-47E610-43-5 10.21-5 Sampling and Water Quality System - Flow Diagram 1-47E867-3 10.21-6 Sampling and Water Quality System - Control Diagram 1-47E610-43-5 11.1-1a Main Steam - Flow Diagram 2-47E801-1 11.1-1b Main Steam - Flow Diagram 2-47E801-2 11.1-1c Main Steam - Flow Diagram 3-47E801-1 11.1-1d Main Steam - Flow Diagram 3-47E801-2 11.1-1e Main Steam - Mechanical Flow Diagram 1-47E801-1 11-1-1f Main Steam - Flow Diagram 1-47E801-2 11.6-1 Condenser Circulating Water - Flow Diagram 2-47E831-1 11.6-2 Condenser Circulating Water - Flow Diagram 1-47E831-2 11.6-3 Condenser Circulating Water - Flow Diagram Sheet 1 1-47E831-3 Sheet 2 2-47E831-3 Sheet 3 3-47E831-3 Sheet 4 2-47E831-2 Sheet 5 3-47E831-2 11.6-4 Condenser Circulating Water - Flow Diagram 3-47E831-1 11.6-5 Condenser Circulating Water - Flow Diagram 1-47E831-1 11.6-6 Condenser Circulating Water - Flow Diagram 0-47E831-5 11.7-1 Condensate Demineralizers - Flow Diagram 2-47E833-1 11.7-2 Condensate Demineralizers - Flow Diagram 3-47E833-1 11.7-3 Condensate Demineralizers - Flow Diagram 1-47E833-1 11.8-1 Reactor Feedwater - Flow Diagram Sheet 1 2-47E803-1 Sheet 2 2-47E803-5 Sheet 3 3-47E803-1 Sheet 4 - Mechanical RPV Level Sensing Lines Instruments and Controls 3-47E803-5 Sheet 5 - Reactor Feedwater - Flow Diagram Mechanical RPV Level Sensing Lines Instruments 1-47E803-1 and Controls Sheet 6 - Reactor Feedwater - Flow Diagram Mechanical RPV Level Sensing Lines Instruments 1-47E803-5 and Controls 11.9-1a Condensate - Flow Diagram 2-47E804-1 11.9-1b Condensate Storage and Supply System - Flow Diagram Sheet 1 1-47E818-1 Sheet 2 2-47E818-1 Sheet 3 3-47E818-1

BFN-25 Table 1.3-1 (Contd) Sheet 12 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 11.9-2 Condensate and Demineralized Water Storage Systems - Mechanical Control Diagram 0-47E610-2-2 11.9-4 Condensate - Flow Diagram 3-47E804-1 11.9-5 Condensate - Flow Diagram 1-47E804-1 12.2-2 Reactor Building General Plans and Sections - Units 1 and 2 41N700 12.2-3 Reactor Building General Plans and Sections - Units 1 and 2 41N701 12.2-4 Reactor Building General Plans and Sections - Units 1 and 2 0-41E702 12.2-5 Reactor Building General Plans and Sections - Unit 3 - Sheet 1 3-41E1000 12.2-6 Reactor Building General Plans and Sections - Unit 3 - Sheet 2 3-41N1001 12.2-21 Reactor Building Structural Steel - Typical Cross Section 0-48E408 12.2-22a Reactor Building Crane Runway - Plans and Sections 0-48E402 12.2-22b Reactor Building Crane General Arrangement 0-PA-2422 12.2-22c Reactor Building Crane - General Arrangement 0-44N220 12.2-23 Reactor Building Sacrificial Shield Wall - Plans and Elevation 0-48E445 12.2-24 Sheet 1 - Structural Steel Drywell Floor Framing - El. 563 ft 1/2 in. 1-48E442-1 Sheet 2 - Structural Steel Drywell Floor Framing - El. 563 ft 1/2 in. 2-48N442 Sheet 3 - Reactor Building Drywell Floor Framing - El. 563 ft 1/2 in. 3-48E442-1 12.2-25 Sheet 1 - Structural Steel Drywell Floor Framing - El. 584 ft 9-1/2 in. 1-48E443-1 Sheet 2 - Reactor Building Drywell Floor Framing - El. 584 ft 9-1/2 in. 2-48E443 Sheet 3 - Reactor Building Drywell Floor Framing - El. 584 ft 9-1/2 in. 3-48N443 12.2-42 Turbine Building, General Plans and Sections, Sheet 1 0-41E200 12.2-43 Turbine Building, General Plans and Sections, Sheet 2 0-41E201 12.2-44 Turbine Building, General Plans and Sections, Sheet 3 41N202 12.2-45 Turbine Building, General Plans and Sections, Sheet 4 0-41E203 12.2-46 Turbine Building, General Plans and Sections, Sheet 5 41N203-1 12.2-47 Turbine Building, Concrete Floor Live Loading, Sheet 1 41N600 12.2-48 Turbine Building, Concrete Floor Live Loading, Sheet 2 0-41E601 12.2-49 Turbine Building, Concrete Floor Live Loading, Sheet 3 41N602 12.2-50 Turbine Building, Structural Steel, Typical Cross Section, 0-48E320 Units 1 and 2 12.2-51 Turbine Building, Structural Steel, Typical Cross Section, Unit 3 48N321 12.2-52 Reinforced Concrete Chimney, Location Plan and Details 0-10E300 12.2-60 Reinforced Concrete Chimney, Foundation Outline 0-10N302 12.2-61 Reinforced Concrete Chimney Interior Above El. 568 0-10N312 12.2-62 Miscellaneous Steel Offgas Stack Exhaust Vent 0-18E211 12.2-64 Concrete, General Plans and Sections, Sheet 1 0-41E569 12.2-65 Concrete, General Plans and Sections, Sheet 2 41N570 12.2-66 Architectual Floor Plans 0-46W426 12.2-67 Architectual Elevations 0-46E425 12.2-69 Concrete - General Outline Features, Sheet 1 0-31E201 12.2-70 Concrete - General Outline Features, Sheet 2 0-31N202 12.2-71a Concrete General Arrangement 0-31E200

BFN-25 Table 1.3-1 (Contd) Sheet 13 List of FSAR Engineering Drawings Engineering Figure Title Drawing No. 12.2-72a Cooling Tower System, Concrete General Arrangement 0-31E400-1 12.2-72b Cooling Tower System, Concrete General Arrangement 0-31N400-2 12.2-72c Water Supply Cooling Tower System, Concrete General Arrangement 0-31N400-3 12.2-73 Diffuser Pipes from Discharge Conduits 0-31N327 12.2-74 Intake Channel Gate Structure 0-31N410-2 12.2-75b Cooling Tower System - Concrete Arrangement Sheet 1 0-31E418-1 Sheet 2 0-31E418-2 Sheet 3 0-31E418-4 12.2-76 Diesel Generator Building, and Standby Gas Treatment Building, General Plans and Sections 0-41E572 12.2-80 Equipment Access Lock Doors 0-44E235-1 12.2-81 Diesel Generator Building, Concrete General Plans and Sections 3-41E595 12.2-82 Concrete - Offgas Treatment Building, General Plans and Sections 0-10E400 12.2-83 Concrete Pipe Plan, Slab and Walls Outline 0-41E598-1 12.2-84 Gate Structure No. 2, Sheet 1 31E420-1 12.2-85 Gate Structure No. 2, Sheet 2 31E420-2

BFN-25 Table 1.3-2 Sheet 1 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-10E151-7 2.5-S1 0-10E201-7 2.2-4 0-10E300 12.2-52 0-10E400 12.2-82 0-10H243 2.4a, Figure 26 Sheet 1 0-10N302 12.2-60 0-10N312 12.2-61 0-15E500-1 8.4-1b 0-15E500-2 8.4-1a 0-17W300-5 10.9-1b 0-18E211 12.2-62 0-31E200 12.2-71a 0-31E201 12.2-69 0-31E400-1 12.2-72a 0-31E418-1 12.2-75b Sheet 1 0-31E418-2 12.2-75b Sheet 2 0-31E418-4 12.2-75b Sheet 3 0-31N202 12.2-70 0-31N327 12.2-73 0-31N400-2 12.2-72b 0-31N400-3 12.2-72c 0-31N410-2 12.2-74 0-41E200 12.2-42 0-41E201 12.2-43 0-41E203 12.2-45 0-41E569 12.2-64 0-41E572 12.2-76 0-41E598-1 12.2-83 0-41E601 12.2-48 0-41E702 12.2-4 0-44N220 12.2-22c 0-44E235-1 12.2-80 0-45E506 8.3-6a 0-45E709-1 8.6-2b 0-45E710-1 8.6-1a 0-45E710-2 8.6-1b 0-45E710-4 8.6-1d 0-45E710-5 8.6-1e 0-45E710-7 8.6-1f 0-45E724-1 8.5-4a 0-45E724-2 8.5-4c 0-45E724-3 8.5-4d 0-45E724-4 8.5-4e

BFN-28 Table 1.3-2 (Contd) Sheet 2 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-45E1504 8.3-4 0-45E1505 8.3-5 0-45E1506 8.3-6 0-45E732-1 8.5-12a 0-45E732-2 8.5-12b 0-45E732-3 8.5-13a 0-45E732-4 8.5-13b 0-45E732-8 8.5-13e 0-45E733-1 8.5-13d 0-45E736-1 8.5-11c 0-45E736-2 8.5-11d 0-45E765-1 8.4-3 0-45W710-3 8.6-1c 0-46E425 12.2-67 0-46W426 12.2-66 0-47E200-2 1.6-2 0-47E200-4 1.6-4 0-47E200-5 1.6-5 0-47E200-6 1.6-6 0-47E200-7 1.6-7 0-47E200-9 1.6-9 0-47E200-10 1.6-10 0-47E200-12 1.6-12 0-47E200-15 1.6-15 0-47E200-16 1.6-16 0-47E200-17 1.6-17 0-47E200-18 1.6-18 0-47E200-19 1.6-19 0-47E215-3 1.6-25 0-47E600-121 2.5-17 0-47E600-91 7.8-3 0-47E610-2-2 11.9-2 0-47E610-23-2 10.9-2b 0-47E610-23-3 10.9-2c 0-47E610-24-3 10.7-2 Sheet 3 0-47E610-43-3 10.17-1c Sheet 1 0-47E610-65-1 5.3-9 0-47E610-67-2 10.10-3 Sheet 4 0-47E610-76-1 5.2-6a Sheet 6 0-47E610-77-1 9.2-3j 0-47E610-77-10 9.2-3s 0-47E610-77-11 9.2-3t 0-47E610-77-2 9.2-3k 0-47E610-77-3 9.2-3l

BFN-28 Table 1.3-2 (Contd) Sheet 3 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-47E610-77-4 9.2-3m 0-47E610-77-5 9.2-3n 0-47E610-77-6 9.2-3o 0-47E610-77-7 9.2-3p 0-47E610-77-8 9.2-3q 0-47E610-77-9 9.2-3r 0-47E610-78-1 10.5-1b Sheet 2 0-47E610-90-2 7.12-2b Sheet 2 0-47E610-90-4 7.12-2a Sheet 3 0-47E610-90-20 7.12-2a Sheet 4 0-47E610-90-21 7.12-2a, Sheet 7 0-47E800-1 1.6-30 0-47E800-2 1.3-2 0-47E820-1 3.4-8b 0-47E830-1 9.2-3a 0-47E830-2 9.2-3b 0-47E830-3 9.2-3c 0-47E830-4 9.2-3d 0-47E830-5 9.2-3e 0-47E830-6 9.2-3f 0-47E830-7 9.2-3g 0-47E830-8 9.2-3h 0-47E830-9 9.2-3i 1-47E831-1 11.6-5 0-47E831-5 11.6-6 0-47E832-1 10.5-2 Sheet 2 0-47E840-1 8.5-3a 0-47E840-3 8.5-3b 0-47E844-3 10.7-1b Sheet 3 0-47E845-1 10.14-2a 0-47E845-2 10.14-2b 0-47E861-1 8.5-2 Sheet 1 0-47E861-1A 8.5-2 Sheet 2 0-47E861-5 8.5-2 Sheet 3 0-47E865-11 5.3-3b 0-47E865-4 10.12-2a 0-47E865-6 10.12-4 0-47E865-8 10.12-5 0-47E866-3 10.12-3 0-47E866-9 10.12-9 0-47W215-1 1.6-23 0-47W215-2 1.6-24 0-47W220-1 1.6-26 0-48E320 12.2-50

BFN-28 Table 1.3-2 (Contd) Sheet 4 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 0-48E402 12.2-22a 0-48E408 12.2-21 0-48E445 12.2-23 0-55E715-1 8.7-4d 0-55E715-2 8.6-6 0-729E286-1 7.9-23 0-729E761-1 7.5-23a 0-729E761-2 7.5-23b 0-729E989-1 7.5-13 0-731E723-1 8.6-5 0-731E744-1 8.6-3 0-731E747-1 8.5-1 0-731E751-1 8.7-3 0-731E752-1 8.7-1 0-731E753-1 8.7-4a 0-731E753-2 8.7-4b Sheet 1 0-75W200 8.3-2 0-PA-2422 12.2-22b 1-45E701-3 7.2-3 1-45E749-1 8.5-25 1-45E749-2 8.5-26 1-45E751-1 8.5-7c 1-45E751-2 8.5-7d 1-45E751-3 8.5-8c 1-45E751-4 8.5-8d 1-47E1847-6 10.14-4 Sheet 5 1-47E200-3 1.6-3 Sheet 1 1-47E200-8 1.6-8 Sheet 1 1-47W401-5 4.4-8 1-47E610-3-1 7.8-1, Sheet 5 1-47E610-23-1 10.9-2e 1-47E610-24-1 10.7-2 Sheet 6 1-47E610-32-1 10.14-1 Sheet 1 1-47E610-32-2 10.14-4 Sheet 6 1-47E610-43-5 10.21-6 1-47E610-46-1 7.10-8 1-47E610-63-1 3.8-8 1-47E610-64-1 5.2-2a Sheet 1 1-47E610-64-2 5.2-2f 1-47E610-64-3 5.2-2g 1-47E610-67-1 10.10-2 1-47E610-68-1 7.8-1, Sheet 6 1-47E610-69-1 4.9-10 1-47E610-71-1 4.7-1e

BFN-28 Table 1.3-2 (Contd) Sheet 5 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 1-47E610-73-1 6.4-5 1-47E610-74-1A 7.4-6b Sheet 3 1-47E610-75-2 7.4-5l 1-47E610-75-3 7.4-5i 1-47E610-74-3 7.4-7p 1-47E610-76-1 5.2-6a Sheet 1 1-47E610-78-1 10.5-1b Sheet 3 1-47E610-80-1 5.2-6b 1-47E610-84-1 5.2-8 Sheet 1 1-47E610-85-1 3.4-8a, Sheet 5 1-47E610-90-1 7.12-2a Sheet 5 1-47E610-90-3 7.12-2b Sheet 5 1-47E801-1 11.1-1e 1-47E801-2 11-1-1f 1-47E803-1 11-8-1 Sheet 5 1-47E803-5 11.8-1 Sheet 6 1-47E804-1 11.9-5 1-47E809-2 9.5-1 Sheet 5 1-47E809-3 9.5-1 Sheet 6 1-47E809-4 9.5-4 1-47E810-1 4.9-8 1-47E811-1 7.4-6a Sheet 2 1-47E812-1 7.4-1b, Sheet 3 1-47E813-1 4.7-1f 1-47E814-1 6.4-6 1-47E817-1 4.3-2a Sheet 1 1-47E818-1 11.9-1b Sheet 1 1-47E822-1 10.6-1a 1-47E831-1 11.6-5 1-47E831-2 11.6-2 1-47E831-3 11.6-3 Sheet 1 1-47E832-1 10.5-2 Sheet 3 1-47E833-1 11.7-3 1-47E837-1 4.9-9 1-47E844-1 10.7-1a Sheet 1 1-47E844-2 10.7-1b Sheet 1 1-47E844-3 10.7-1b Sheet 7 1-47E854-1 3.8-3 1-47E855-1 10.5-1c 1-47E858-1 10.9-1a Sheet 1 1-47E859-1 10.10-1a 1-47E862-1 5.2-7 Sheet 1 1-47E865-1 5.3-3a 1-47E865-3 10.12-7

BFN-28 Table 1.3-2 (Contd) Sheet 6 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 1-47E867-3 10.21-5 1-47E1610-85-2 3.4-8g 1-47E1610-85-5 3.4-8h 1-47E1610-76-3 5.2-6a, Sheet 7 1-47K1110-32 1.6-29 Sheet 3 1-48E442-1 12.2-24 Sheet 1 1-48E443-1 12.2-25 Sheet 1 1-104R935-1 4.2-4 1-105E1512RE-1 7.5-1c 1-105E1512RE-2 7.5-11c 1-729E136-3 7.2-7d 1-729E229-1 4.5-1 1-729E761-1 7.5-23c 1-729E761-2 7.5-23d 1-729-E761-1 7.5.23c 1-729E761-2 7.5.23d 1-731E753-3 8.7-4c Sheet 1 1-791E167 7.2-8 1-C5445E-102 10.3-1 2-104R935-1 4.2-1 2-105E1512-1 7.5-1a 2-105E1512-2 7.5-11a 2-45E702-4 7.2-1 2-45E749-3 8.5-5 2-45E749-4 8.5-6 2-45E751-1 8.5-7a 2-45E751-11 8.5-11 2-45E751-2 8.5-7b 2-45E751-3 8.5-8a 2-45E751-4 8.5-8b 2-45E751-5 8.5-9a 2-45E751-6 8.5-9b 2-45E751-8 8.5-10 2-47E200-3 1.6-3 Sheet 2 2-47E200-8 1.6-8 Sheet 2 2-47E610-23-1 10.9-2a 2-47E610-24-1C 10.7-2 Sheet 1 2-47E610-24-1D 10.7-2 Sheet 2 2-47E610-3-1 7.8-1 Sheet 1 2-47E610-32-1 10.14-1 Sheet 2 2-47E610-32-2 10.14-4 Sheet 1 2-47E610-43-1 10.17-1a 2-47E610-43-2 10.17-1b 2-47E610-43-3 10.17-1c Sheet 2

BFN-28 Table 1.3-2 (Contd) Sheet 7 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 2-47E610-43-5 10.21-2 2-47E610-46-1 7.10-2 2-47E610-46-2 7.10-3 2-47E610-46-3 7.10-4 2-47E610-63-1 3.8-2 2-47E610-64-1 5.2-2a Sheet 2 2-47E610-64-2 5.2-2b 2-47E610-64-3 5.2-2c 2-47E610-67-2 10.10-3 Sheet 2 2-47E610-68-1 7.8-1 Sheet 2 2-47E610-69-1 4.9-3 2-47E610-71-1 4.7-1b 2-47E610-73-1 6.4-1 2-47E610-74-3 7.4-7b 2-47E610-75-2 7.4-5d 2-47E610-75-3 7.4-5a 2-47E610-76-1 5.2-6a Sheet 3 2-47E610-76-4 5.2-6a Sheet 2 2-47E610-78-1 10.5-1b Sheet 1 2-47E610-80-1 5.2-6c 2-47E610-84-1 5.2-8 Sheet 2 2-47E610-85-1 3.4-8a Sheet 1 2-47E610-90-1 7.12-2a Sheet 1 2-47E610-90-2 7.12-2a Sheet 2 2-47E610-90-3 7.12-2b Sheet 4 2-47E801-1 11.1-1a 2-47E801-2 11.1-1b 2-47E803-1 11.8-1 Sheet 1 2-47E803-5 11.8-1 Sheet 2 2-47E804-1 11.9-1a 2-47E809-2 9.5-1 Sheet 1 2-47E809-3 9.5-1 Sheet 2 2-47E809-4 9.5-3 2-47E810-1 4.9-1 2-47E811-1 7.4-6a Sheet 1 2-47E812-1 7.4-1b Sheet 1 2-47E813-1 4.7-1a 2-47E814-1 6.4-2 2-47E817-1 4.3-2a Sheet 2 2-47E818-1 11.9-1b Sheet 2 2-47E820-2 3.4-8c 2-47E822-1 10.6-1b 2-47E831-1 11.6-1 2-47E831-2 11.6-3 Sheet 4

BFN-28 Table 1.3-2 (Contd) Sheet 8 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 2-47E831-3 11.6-3 Sheet 2 2-47E832-1 10.5-2 Sheet 1 2-47E833-1 11.7-1 2-47E837-1 4.9-2 2-47E844-1 10.7-1a Sheet 2 2-47E844-2 10.7-1b Sheet 2 2-47E844-3 10.7-1b Sheet 5 2-47E854-1 3.8-1 2-47E855-1 10.5-1a 2-47E858-1 10.9-1a Sheet 2 2-47E859-1 10.10-1C 2-47E862-1 5.2-7 Sheet 2 2-47E865-3 10.12-1 2-47E867-3 10.21-1 2-47E2610-74-1 7.4-6b Sheet 1 2-47E2610-74-2 7.4-6b Sheet 2 2-47E2610-85-2 3.4-8a Sheet 2 2-47E2610-85-5 3.4-8d 2-47E2847-5 10.14-4 Sheet 2 2-47E2865-12 5.3-3c 2-47E2865-4 10.12-2b 2-47K1110-13 1.6-29 Sheet 1 2-47W401-5 4.4-6 2-48E443 12.2-25 Sheet 2 2-48N442 12.2-24 Sheet 2 2-729E136-3 7.2-7a 2-729E229-4 4.5-2 2-731E753-2 8.7-4b Sheet 2 2-731E753-3 8.7-4c Sheet 2 2-791E167RF 7.2-2 3-C5445E-104 10.3-3 3-104R935-1 4.2-3 3-105E1512-1 7.5-1b 3-105E1512-2 7.5-11b 3-15E500-3 8.4-2 3-41E1000 12.2-5 3-41E595 12.2-81 3-41N1001 12.2-6 3-45E703-3 7.2-7c 3-45E709-2 8.6-2c 3-45E724-6 8.5-4b 3-45E724-7 8.5-4f 3-45E724-8 8.5-4g 3-45E724-9 8.5-4h

BFN-28 Table 1.3-2 (Contd) Sheet 9 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-45E732-5 8.5-12c 3-45E732-6 8.5-13c 3-45E749-5 8.5-27 3-45E749-6 8.5-28 3-45E751-1 8.5-7e 3-45E751-2 8.5-7f 3-45E751-3 8.5-8e 3-45E751-4 8.5-8f 3-45E751-5 8.5-9c 3-45E751-6 8.5-9d 3-45E751-9 8.5-11a 3-47E200-11 1.6-11 3-47E200-13 1.6-13 3-47E200-14 1.6-14 3-47E200-8 1.6-8 Sheet 3 3-47E220-2 1.6-27 3-47E401-5 4.4-7 3-47E610-23-1 10.9-2d 3-47E610-24-1 10.7-2 Sheet 4 3-47E610-24-2 10.7-2 Sheet 5 3-47E610-3-1 7.8-1 Sheet 3 3-47E610-32-1 10.14-1 Sheet 3 3-47E610-32-2 10.14-4 Sheet 3 3-47E610-43-1 10.17-1d 3-47E610-43-2 10.17-1e 3-47E610-43-3 10.17-1f 3-47E610-43-5 10.21-4 3-47E610-46-1 7.10-5 3-47E610-46-2 7.10-6 3-47E610-46-3 7.10-7 3-47E610-63-1 3.8-6 3-47E610-64-1 5.2-2a Sheet 3 3-47E610-64-2 5.2-2d 3-47E610-64-3 5.2-2e 3-47E610-67-2 10.10-3 Sheet 3 3-47E610-67-3 10.10-3 Sheet 1 3-47E610-68-1 7.8-1 Sheet 4 3-47E610-69-1 4.9-7 3-47E610-71-1 4.7-1d 3-47E610-73-1 6.4-3 3-47E610-74-1 7.4-6b Sheet 4 3-47E610-74-2 7.4-6b Sheet 5 3-47E610-75-2 7.4-7i 3-47E610-75-3 7.4-5m

BFN-28 Table 1.3-2 (Contd) Sheet 10 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-47E610-76-1 5.2-6a Sheet 4 3-47E610-76-4 5.2-6a Sheet 5 3-47E610-78-1 10.5-1b Sheet 4 3-47E610-80-1 5.2-6d 3-47E610-84-1 5.2-8 Sheet 3 3-47E610-85-1 3.4-8a Sheet 3 3-47E610-85-5 3.4-8f 3-47E610-90-1 7.12-2a Sheet 6 3-47E610-90-3 7.12-2b Sheet 6 3-47E801-1 11.1-1c 3-47E801-2 11.1-1d 3-47E803-1 11.8-1 Sheet 3 3-47E803-5 11.8-1 Sheet 4 3-47E804-1 11.9-4 3-47E809-2 9.5-1 Sheet 3 3-47E809-3 9.5-1 Sheet 4 3-47E809-4 9.5-2 3-47E810-1 4.9-5 3-47E811-1 7.4-6a Sheet 3 3-47E812-1 7.4-1b Sheet 2 3-47E813-1 4.7-1c 3-47E814-1 6.4-4 3-47E817-1 4.3-2a Sheet 3 3-47E818-1 11.9-1b Sheet 3 3-47E820-2 3.4-8e 3-47E822-1 10.6-1c 3-47E831-1 11.6-4 3-47E831-2 11.6-3 Sheet 5 3-47E831-3 11.6-3 Sheet 3 3-47E832-1 10.5-2 Sheet 4 3-47E833-1 11.7-2 3-47E837-1 4.9-6 3-47E844-1 10.7-1a Sheet 3 3-47E844-2 10.7-1b Sheet 4 3-47E844-3 10.7-1b Sheet 6 3-47E854-1 3.8-5 3-47E855-1 10.5-1d 3-47E858-1 10.9-1a Sheet 3 3-47E859-1 10.10-1d 3-47E859-2 10.10-1b 3-47E861-1 8.5-2 Sheet 4 3-47E861-1a 8.5-2 Sheet 5 3-47E861-5 8.5-2 Sheet 6 3-47E862-1 5.2-7 Sheet 3

BFN-28 Table 1.3-2 (Contd) Sheet 11 Engineering Drawing Cross-Reference List Engineering Drawing No. Figure 3-47E865-12 5.3-3d 3-47E865-3 10.12-8 3-47E865-4 10.12-2c 3-47E865-8 10.12-6 3-47E867-3 10.21-3 3-47E3610-85-2 3.4-8a, Sheet 4 3-47E3847-5 10.14-4 Sheet 4 3-47K1110-13 1.6-29 Sheet 2 3-48E442-1 12.2-24 Sheet 3 3-48N443 12.2-25 Sheet 3 3-729E136-3 7.2-7b 3-729E229-4 4.5-3 3-731E753-3 8.7-4c Sheet 3 3-791E167 7.2-9 104R900 1.3-1 10H237 2.4A Fig 25 10H242 2.4A Fig 26, Sh 2 31E420-1 12.2-84 31E420-2 12.2-85 41N202 12.2-44 41N203-1 12.2-46 41N570 12.2-65 41N600 12.2-47 41N602 12.2-49 41N700 12.2-2 41N701 12.2-3 41N703 2.5-19 47E200-1 1.6-1 48N321 12.2-51 729E946-1 7.5-2 C5445E-103 10.3-2 LC48417-1 8.3-2a

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AMENDMENT 16 BROWNS FERRY NVCLEAR PLANT

l NA.L SAFETY 4NA.L vs IS REPORT FI GURE 1.3-1

600~ Z:-0083Lt,-Q 11'1 L9 SYMBOLS FLOW DIAGRAM DRAWING INDEX THROTTLING VALVE (GATE) ---5-- INSTRUMENT POT 47E800-SERIES ......... FLOW DIAGRAM GENERAL PLANT SYSTEt.tS A SYt.eOLS 47E801-SERIES ......... FLOW DIAGRAM MAIN STEAM H 47E802-SERIES ......... FLOW DIAGRAM EXTRACTION STEAM FILTER 47E803-SERIES ......... FLOW DIAGRAM REACTOR FEEDWATER THROTTLING VALVE (GLOBE) 47E804-SERIES ......... FLOW DIAGRAM CONDENSATE J-WAY BLOCK VALVE 47E805-SERIES ......... FLOW DIAGRAM HEATER DRAINS, VENTS A MISC PIPING ADJUSTABLE FIXED 47E807-SERIES ......... FLOW DIAGRAM TURBINE DRAINS A MISC PIPING LOUVERS LOUVERS 47E808-SERIES ......... FLOW DIAGRAM HYDROGEN WATER CHEMISTRY SYSTEM

                        -t><l-   GATE VALVE (NORMALLY OPEN) 47E809-SERIES ......... FLOW DIAGRAM OFF-GAS SYSTEM Cat.eINATION AIR FILTER AND                                                                                              47E810-SERIES ......... FLOW DIAGRAM REACTOR WATER CLEANUP SYSTEM 47E811-SERIES ......... FLOW DIAGRAM RESIDUAL HEAT REWlVAL SYSTEM GATE VALVE (NORMALLY CLOSED)
                                                                   ---t><llill><l-BACKFLOW PREVENTER ASSEMBLY; CCMMON BODY DOUBLE CHECK VALVE TYPE WITH SHUTOFF PRESSURE REGULATOR f--f8)=)-)----JJ                    PRIMARY CONTAI ...ENT PENETRATION OR WALL AND FLOOR SLEEVES 47E812-SERIES ......... FLOW DIAGRAM HIGH PRESSURE COOLANT INJECTION SYSTEM 47E813-SERIES ......... FLOW DIAGRAM REACTOR CORE ISOLATION COOLING SYSTEM 47E814-SERIES ......... FLOW DIAGRAM CORE SPRAY SYSTEM                                       -

GATE VALVE (NORMALLY CLOSED) WITH ORIFICE THROUGH WEDGE TEST ANNUALLY VALVE ON EACH SIDE-REQUIRES 3 UNIDS

                                                                                                                                                --rnm-    BREAKDOWN ORIFICE NOTE, CONTAI ...ENT PENETRATIONS DENOTED BY X-NUMBERS IE. X-29 47E815-SERIES ......... FLOW DIAGRAM AUXILIARY BOILER SYSTEM 47E816-SERIES ......... FLOW DIAGRAM LUBRICATING OIL SYSTEM 47E817-SERIES ......... FLOW DIAGRAM REACTOR WATER RECIRCN, DRAINS, VENTS A BLOIDOWN SYSTEMS 47E818-SERIES ......... FLOW DIAGRAM CONDENSATE STORAGE AND SUPPLY SYSTEM
                        -[:x:(]- GLOBE VALVE (NORMALLY OPEN)             -               BACKFLOW PREVENTER ASSEMBLY; DIFFERENTIAL PRESSURE (DP) TYPE, CO~N BODY DOUBLE                   -§- STRAIGHTENING VANES                                                              TUBING TO PIPING TRANSITION 47E819-SERIES ......... FLOI DIAGRAM EMERGENCY HIGH PRESSURE MAKEUP PUMP SYSTEM 47E820-SERIES ......... FLOW DIAGRAM CONTROL ROD DRIVE HYDRAULIC SYSTEM
                                                                   ~

CHECK VALVE (WITH DP RELIEF VALVE) AND ,,. 1/2" .500

                                                                                                                                                                                                                             ,,.     (FRACTIONAL SIZE DENOTES PIPE AND 47E821-SERIES ......... FLOW DIAGRAM CHEMICAL CLEANING SYSTEM 47E822-SERIES ......... FLOW DIAGRAM REACTOR BUILDING CLOSED COOLING WATER SYSTEM a

SHUTOFF VALVE ON EACH SIDE-REQUIRES 3 UNIDS DECIMAL SIZE DENOTES TUBING) 47E825-SERIES ......... FLOW DIAGRAM BREATHING AIR SYSTEM GLOBE VALVE (NORMALLY CLOSED) 47E830-SERIES .....*... FLOW DIAGRAM RADWASTE SYSTEM TEST ANNUALLY DESUPERHEATER 47E831-SERIES ......... FLOW DIAGRAM CONDENSATE CIRCULATING WATER SYSTEM 8 47E832-SERIES ......... FLOW DIAGRAM FUEL POOL FILTER/DEMINERALIZING SYSTEM

                        --l<e>I- BALANCING VALVE                                         BACKFLOW PREVENTER ASSEMBLY; SEPARATE BODY DEVICE MOUNTED ON MAIN CONTROL ROOM PANEL 47EBJJ-SERIES ......... FLOW DIAGRAM CONDENSATE DEMINtRALIZERS 47E834-SERIES ......... FLOW DIAGRAM MAKEUP WATER TREAT~NT SYSTEM                             G TEST ANNUALLY DOUBLE CHECK VALVES WITH SHUTOFF VALVE ON EACH SIDE-REQUIRES 4 UNIDS                                ---l I BLIND FLANGE                                                                (SEE NOTE 3)                                 47E835-SERIES ......... FLOW DIAGRAM POTABLE WATER DISTRIBUTION SYSTEM 47E836-SERIES ......... FLOW DIAGRAM RAW SERVICE WATER & FIRE PROTECTION SYSTEM 47E837-SERIES ......... FLOW DIAGRAM REACTOR WATER CLEANUP DEMINERALIZER SLIDE OR BLAST GATE VALVE                                                                                                                                                                                                                         47E838-SERIES ......... FLOW DIAGRAM RAW WATER CHLORINATION SYSTEM
                                                                                                                                                 ---INf-  FLEXIBLE CONNECTION CAPILLARY TUBING 47E839-SERIES ......... FLOW DIAGRAM HYPOCHLORITE SYSTEM 47E840-SERIES ......... FLOW DIAGRAM FUEL OIL SYSTEM (SEE NOTE 3)                                   47E841-SERIES ......... FLOW DIAGRAM GLAND SEAL WATER SYSTEM 47E842-SERIES ......... FLOW DIAGRAM INSULATING OIL SYSTEM THREE WAY VALVE                                                                                                           MAIN PROCESS LINES                                                                                                      47E84J-SERIES ......... FLOW DIAGRAM CO2 STORAGE, FIRE PROT A PURGING SYSTEM 47E844-SERIES .....*... FLOW DIAGRAM RAW COOLING WATER SYSTEM FOUR WAY VALVE CHECK VALVE DIAPHRAGM OPERATOR ELECTRIC MOTOR OPERATOR (FLOW DIRECTION}

AUXILIARY PROCESS LINES

                                                                                                                                                                                                         ,HC I'"  PNEUMATIC TUBING

{SEE NOTE 3) NITROGEN PIPE 47E845-SERIES ......... FLOW DIAGRAM COMPRESSED AIR-STATION SERVICE SYSTEM 47E846-SERIES .....*... FLOW DIAGRAM DEMINERALIZER BACKWASH AIR SYSTEM 47E847-SERIES ......... FLOW DIAGRAM CONTROL AIR SYSTEM 47E848-SERIES .....*... FLOW DIAGRAM VACUUM PRIMING SYSTEM 47E849-SERIES ......... FLOW DIAGRAM HYDROGEN SYSTEM FOR GENERATOR COOLING 47E850-SERIES .....*... FLOW DIAGRAM FIRE PROTECTION ARAI SERVICE WATER SYSTEM 47E851-SERIES ......... FLOW DIAGRAM DRAINAGE-TURBINE, DIESEL GENERATOR, OFFICE AND SERVICE BUILDINGS AND STACK OR TUBING 47E852-SERIES ......... FLOW DIAGRAM DRAINAGE-REACTOR BLDG CLOSED DRAIN (CRI} (SEE NOTE 3) 47E85J-SERIES ......... FLOW DIAGRAM AUXILIARY BOILER FEEDWATER SECONDARY TREATMENT STOP CHECK VALVE 47E854-SERIES ......... FLOW DIAGRAM STANDBY LIQUID CONTROL SYTEM 47E855-SERIES ......... FLOW DIAGRAM FUEL POOL COOLING SYSTEM 47E856-SERIES ......... FLOW DIAGRAM DEMINERALIZED WATER SYSTEM ANGLE VALVE (NORMALLY OPEN) VALVE POSITIONER OPEN DRAIN (DRW)

                                                                                                                                                                                                    ~                              CONSTANT HEAD POT                               47E857-SERIES ......... FLOW DIAGRAM CONDENSER TUBE CLEANING SYSTEM 47E858-SERIES ......... FLOW DIAGRAM RHR SERVICE WATER SYSTEM 47E859-SERIES ......... FLOW DIAGRAM EMERGENCY EQPT COOLING WATER SYSTEM 47E860-SERIES .....*... FLOW DIAGRAM CONTAINMENT INERTING SYSTEM 47E861-SERIES ......... FLOW DIAGRAM DIESEL STARTING AIR SYSTEM 47E862-SERIES .....*... FLOW DIAGRAM CONTAI ...ENT ATMOSPHERE DILUTION SYSTEM F

ANGLE VALVE (NORMALLY CLOSED) HEAT TRACING 47E865-SERIES .....*... FLOW DIAGRAM HVAC AIR FLOW SYSTEM EXPLOSIVE OPERATOR 47E866-SERIES ......... FLOW DIAGRAM BUILDING HEATING A AIR CONDITIONING HOODED SAMPLE STATION 47E870-SERIES ......... FLOW DIAGRAM WARM WATER WASTE HEAT DEMONSTRATION 47E873-SERIES ......... FLOW DIAGRAM AUXILIARY DECAY HEAT REWVAL SYSTEM ANGLE GLOBE VALVE 47E885-SERIES ......... FLOW DIAGRAM GENERATOR COOLING SYSTEMS CYLINDER OPERATOR FAN WITH MOTOR DAMPER MANUAL WITH LOCKING QUADRANT RELIEF VALVE ANGLE RELIEF VALVE ....!"-x--L SOLENOID OPERATOR HAND CONTROLLED VALVE r-,,,<----f FIRE DAMPER

                                                                                                                                                                                                    ._I~>          ,I' CONTROL AIR SUPPLY. PRESSURE TO BE SET AT MANUFACTURER'S REQUIREMENT FOR THE DEVICE BEING SUPPLIED NOTES:

1 . DELETED

2. DELETED J. SEE OS E18.3.3 FOR ADDITIONAL INSTRUMENT NOMENCLATURE.

4. INFORMATION ON THIS DRAWING ORIGINALLY DEPICTED ON 471801-1.

FLOAT OPERATED VALVE lg GRAB SAMPLE

                                                                                                                                          ~
                                                                                                                                          -     V- --

1-------:

AIR FLOW IN DUCT. AIR VOLUME {CFMj SHOULD BE USED TO INDICATE DI~ECT ONAL y ATt.<<JS VENT/EMERGENCY RELIEF

5. GRILLE WILL BE DEFINED AS HAVING FIXED DIFFUSERS (NOT ADJUSTABLE t.tANUALLY OR BY MOTOR).

6. REGISTER WILL BE DEFINED AS HAVING ADJUSTABLE DIFFUSERS (MANUALLY OPERATED).

FLOW WITH AND/OR WITHOUT CFM NOTED 7. THIS DRAWING ESTABLISHES SYMBOLS TO BE USED ON THE BROWNS FERRY I ~I I CAD RESTORED DRAWINGS. SYMBOLIC REPRESENTATION DOES NOT DEPICT NEEDLE VALVE SINGLE PASS HEAT EXCHANGER EXACT SCALE. E

8. USE OF SYMBOLS NOT DEPICTED ON THIS DRAWING SHALL FOLLOW SIGHT FLOW INDICATOR (SEE NOTE J)

H,,,-1----, BACK DRAFT DAMPER (IN DUCTWORK) THE GUIDELINES ADDRESSED IN DES 7.01-GENERAL DRAFTING PRACTICES. PINCH VALVE 9. DELETED

                                                                                                                                      ~
                                                                                                                                                                                                    ~

BACK DRAFT DAMPER DOUBLE PASS (AT DOOR-GRILLE, WALL AL TITU E VALVE OPENING, NOT IN DUCT- HEAT EXCHANGER WORK, ETC.) ORIFICE OR NOZZLE TYPE FLOW ELEMENT . DIAPHRAGM VALVE BUTTERFLY VALVE (NORMALLY OPEN) FE WITH TRANSMITTER (SEE NOTE J) DOOR/WALL TRANSFER GRILLE FLAME ARRESTOR RETURN OR EXHAUST BUTTERFLY VALVE (NORMALLY CLOSED) GRILLE PRESSURE INDICATOR (SEE NOTE 3) ',----{}-----<, CONDENSING POT MUD VALVE FIRE HOSE COMBINATION ---J---1~ SUPPLY GRILLE REVERSE CURRENT VALVE PLUG VALVE PRESS. RESTRICTING ANGLE VALVE SPRINKLER/NOZZLE/ORIFICE

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                                                                              ~          END SHUTOFF
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ANGLE CHECK VALVE TUBE UNION OR CONNECTOR

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                                                                                                                                                       ,    CLOSED IN PROCESS LINE AMENDMENT                               28 JJ_          PRESSURE REDUCING REGULATOR WITH EXTERNAL PRESSURE TAP                 -0]--

J CAP OR PLUG (SCREWED OR WELDED) IMPULSE TRAP WITH STRAINER (Qr FAN (VENTILATION) GENERAL BROWNS FERRY NUCLEAR PLANT SELF CONTAINED PRESSURE

                                                                             -          SINGLE STRAINER A/C UNIT                                                                                                                                           FINAL SAFETY ANALYSIS REPORT REGULATING VALVE
                                                                             -          TWIN STRAINER GENERAL SYMBOLS                                A GATE VALVE WITH GLAND SEAL H~                                         BUCKET TRAP PUMP                                                                                                                                                                      FLOW DIAGRAM FIGURE 1 . 3-2 8                                        7                                    6                                                       5                          t                            4                                                               3

BFN-16 1.4 CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION 1.4.1 Introduction To fully evaluate the many aspects of the design and operation of the boiling water reactor plant, it is necessary to classify the various systems, criteria, design bases, and operating requirements in light of specified personnel (including the public) hazard considerations. A system has been developed which allows classification of any BWR aspect-criterion, system, design basis, or operating requirement-relative to either personnel hazard or the plant mission (the generation of electrical power). Table 1.4-1 illustrates the concept used in the classification process. The concept applies to the total plant: design and operation. A major distinction is made between those BWR aspects which are most pertinent to personnel hazard and those which are most pertinent to the plant mission-the generation of electrical power. Those aspects most pertinent to personnel hazard would appear under the "safety consideration" side (left side) of the table, and the aspects most pertinent to the plant mission would appear under the "power generation" side (right). All plant components contribute in some measure to safety, but those classified under "power generation" considerations are considerably less important to safety than those items classified under "safety" considerations. Therefore, the right and left sides of the table represent a major difference in importance to safety. Down the left side of Table 1.4-1 are listed the various types of plant operation, including events resulting in transients and accidents. An allowance is made for a special event in the left column to enable the classification of criteria, systems, and operational requirements not otherwise classifiable. The left-hand column is actually a gross probability scale. Planned operation is certain, abnormal operational transients are reasonably expected, and accidents are very improbable. Any special events would have to be fitted into the probability scale as appropriate. The left-hand column might ultimately develop into a quantified probability scale. The rectangular spaces formed under the safety considerations heading and the power generation heading represent potential classification categories for BWR criteria, systems, and operational requirements. This classification concept, when applied, allows an accurate distinction between the importances of the various aspects of BWR design and operation. 1.4.2 Classification Basis Tables 1.4-2A and 1.4-2B present the basis for classifying various BWR items. The format of the tables is similar to that used in Table 1.4-1, which presented the classification concept. A list of unacceptable results is given within each 1.4-1

BFN-17 classification category. The unacceptable results represent a set of master criteria, from which the design and operation of the BWR can be consistently evaluated. The only unacceptable results listed for the power generation consideration (Table 1.4-2B) are those that are more restrictive than those for the safety consideration (Table 1.4-2A). In the various columns inside each classification category, generic labels are assigned to the specific elements which appear or would appear, if listed, in the column. A generic label is given only to facilitate discussion and identification of a group of elements united by their common classification. Beneath the generic names are listed some of the more illustrative BWR items which can be classified in the different columns. Some of the listed items are the limits and restrictions found in the technical specifications. Technical specifications are limited to those concerns that are only on Table 1.4-2A. Classification analyses have been performed to establish the essentiality of the various BWR systems to the avoidance or prevention of the listed unacceptable results. Such analyses consider any applicable criteria requiring redundancy or specified levels of functional reliability in the avoidance of unacceptable results. Once a system is classified, it is evaluated with reference to the criteria applicable to the group in which it performs an essential action. A classification analysis is not the same as a plant safety analysis. A classification analysis takes no credit whatever for the system under study; whereas, a plant safety analysis represents the true response of the whole plant to an event under specified analytical assumptions. 1.4.3 Use of the Classification Plan Because Tables 1.4-2A and B permits the classification of any BWR criterion, system, or operational requirement into one or more of the classification categories, the plan facilitates a plantwide safety overview. The plan explains the reasons for the differences in the designs of apparently similar systems by relating the actions of the systems to specified unacceptable results. With the design complete, the classification plan is used to establish operational requirements and procedures whose differences are consistent with the different importances of unacceptable results. It should be noted that a system may be classified in several categories. This occurs because classification is the result of a functional analysis of the plant. When classified in more than one category, a system must satisfy all of the requirements for each category with regard to its contributions to the various safety actions within each of the categories. 1.4-2

BFN-21 TABLE 1.4-1 BWR SAFETY ENGINEERING CONCEPT FOR CLASSIFICATION OF BWR SYSTEMS, CRITERIA AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION Type of Operation Safety Considerations Power Generation Considerations or Event

1. Planned Operation In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems, and operational requirements power generation, criteria, plant actions, systems and pertinent to safety during planned operation. This space operational requirements pertinent to the production of represents the aspects of the BWR which must be considered to electrical power during planned operation. Process assure that the BWR operator can operate the plant within systems and normal operational procedures would be specified safety limitations. Certain process indicators, process classified here.

variable limits and limits on the release of radioactive material would be classified here.

2. Abnormal Operational In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for Transients criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to abnormal operational transients. operational requirements pertinent to the ability to produce Certain protection systems, safety limits, and limiting safety electrical power as that ability is affected by abnormal system settings would be classified here. operational transients. Certain systems not used for planned operation would be classified here.

3. Accidents In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to accidents. Engineered operational requirements pertinent to the ability to produce safeguards would be classified here. electrical power as that ability is affected by accidents.

Design considerations and post-accident procedures provided to enable the plant to be used for power generation after an accident would be classified here.

4. Special Event In this category are classified the unacceptable safety results, In this category are classified the unacceptable results for criteria, plant actions, systems and operational requirements power generation, criteria, plant actions, systems and pertinent to safety in regard to the stated special event. Safety operational requirements pertinent to the ability to produce systems provided especially for the special event would be electrical power as that ability is affected by the stated classified here. special event. Systems and procedures provided to enable the plant to be returned to power operation following the special event would be classified here.

BFN-21 Table 1.4-2A (Sheet 1) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

1. Planned 1-1 The release of Nuclear Safety Design Safety Action- Safety Systems- Operational Nuclear Safety Operation radioactive material Criteria-Type S-1 Type S-1 Type S-1 Requirements-Type S-1 to the environs to such an extent that the Nuclear Safety Process Safety Action Process Safety Systems Operational Nuclear Safety limits of 10CFR20 Operational Criteria- (A Category of Safety (A Category of Safety Limits-Type S-1 are exceeded. Type S-1 Action) Systems)

Process Safety Indication of Process Indicators Technical Specifications-1-2 Fuel failure to such an Design Criteria Variables Type S-1 extent that were the freed fission products Process Safety Rod Worth Monitoring Rod Worth Minimizer Process Safety Limits to the environs via the Operational Criteria Program of Process Computer normal discharge paths Rod Pattern Control Limiting Conditions for for radioactive material, Various Industry Radwaste Systems for Operation for Indicators limits of 10CFR20 Codes Control of Process would be exceeded. Variables Process Radiation Monitors Radioactive Material Radwaste Criteria Release Limits Control Rod Control Control Rod Drive System 1-3 Nuclear System stress Loading Criteria Refueling Block in excess of that (Normal Conditions) Reactor Manual Rod Pattern Limits allowed for planned Control Rod Control Control System operation by applicable industry codes. Refueling Block Limiting Conditions for Refueling Interlocks Operation for Radwaste Systems Core Shutdown Control 1-4 The existence of a plant Reactor Protection System condition not considered Radwaste (Manual Scram) Nuclear System Leakage by plant safety analysis. Limits Isolation Condensate Storage System Neutron Monitoring System

BFN-21 Table 1.4-2A (Sheet 2) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

2. Abnormal 2.1 The release of radioactive Nuclear Safety Design Safety Action-Type S-2 Safety Systems-Type S-2 Operational Nuclear Operational material to the environs Criteria-Type S-2 Safety Requirements-Transients to such an extent that Scram Protection System (Generic Type S-2 the limits of 10CFR20 Nuclear Safety Term) are exceeded. Operational Criteria- Pressure Relief Type S-2 Nuclear Safety Systems Operational Nuclear Safety Core Cooling (A Category of Protection Limits-Type S-2 Systems) 2.2 Any fuel failure calculated as a result of the transient. Various Industry Codes Containment Reactor Protection System Technical Specifications-Cooling (RHRS) (Scram) Type S-2 IEEE-279 Primary Containment Control Rod Drive System (Scram) Safety Limits Secondary Availability Goals Containment Neutron Monitoring System (IRM, APRM) Limiting Safety System Loading Criteria Settings (Upset Conditions) Pressure Relief System 2.3 Nuclear system stress in Limiting Conditions for excess of that allowed for Single Failure Reactor Vessel Isolation Operation for Protection transients by applicable Criterion Control System Systems industry codes.

Testability Criteria High Pressure Coolant Surveillance Requirements Injection System for Protection Systems D-C Power System Standby A-C Power

BFN-21 Table 1.4-2A (Sheet 3) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

3. Accidents 3-1 Radioactive Nuclear Safety Design Safety Action-Type S-3 Safety Systems-Type S-3 Operational Nuclear Safety material release Criteria-Type S-3 Requirement-Type S-3 to such an extent that the guideline Scram Protection Systems (Generic values of 10CFR50.67 Nuclear Safety Term) Operational Nuclear Safety would be exceeded. Operational Criteria - Core Cooling Limits-Type S-3 Type S-3 Engineered Safeguards 3-2 Fuel cladding temperatures Containment Reactor Protection System Technical Specifications-in excess of 2200F for Control Rod Drive System Type S-3 Pipe Breaks Neutron Monitoring System 3-3 Nuclear system Various Industry Codes Pressure Relief System (Main pressure in excess Containment Cooling Steam Relief Valves) Limiting Safety System of that allowed for IEEE-279 Reactor Vessel Isolation Settings accidents by applicable Control System industry codes. Availability Goals Stop Control Rod Primary Containment Isolation Limiting Conditions for Ejection Control System Operation for Protection Primary Containment Systems Secondary Containment 3-4 Containment stresses Loading Criteria Limit Reactivity Main Steam Line Isolation sufficient to produce (Emergency and Insertion Rate Valves Surveillance Requirements containment failure Faulted Conditions) Main Steam Line Flow Restrictor for Nuclear System when containment is Pressure Relief High Pressure Coolant Injection required. System Automatic Depressurization 3-5 Overexposure to Single Failure Criteria Reactor Vessel System radiation of operating Isolation Low Pressure Coolant Injection personnel in the Testability Criteria Core Spray System control room. Primary Containment RHRS (Containment Cooling)

Isolation Control Rod Velocity Limiter

BFN-21 Table 1.4-2A (Sheet 4) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

3. Accidents 3-5 Overexposure Testability Criteria Secondary Containment Control Rod Drive Housing Surveillance requirements (Cont.) to radiation Isolation Supports for nuclear systems of operating Standby Gas Treatment System personnel in Standby A-C Power System the control Treatment of Fission D-C Power System room. Products Main Steam Line Radiation Monitoring System Restriction of Coolant Reactor Building Ventilator Loss Rate Radiation Monitoring System RHR Service Water System Control Room Isolation 3-6 Peak enthalpy of fuel in excess of 280 cal/gm for the control rod drop accident.

BFN-21 Table 1.4-2A (Sheet 5) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

4. Special Event 4-1 The inability to Nuclear Safety Design Safety Action - Type S-4 Safety Systems-Type S-4 Operational Nuclear Safety Loss of bring the reactor Criteria-Type S-4 Requirements-Type S-4 Habitability to the shutdown Special Safety Action Special Safety Systems of the Control condition by manip-Room ulation of the local controls and Nuclear Safety Operational Nuclear Safety equipment which Operational Criteria- Shutdown From Outside Local Controls Outside Limits-Type S-4 are available out- Type S-4 Control Room Control Room side the control room. Special Safety Design Technical Specifications-Criteria Cooldown from Outside Local Indicators Outside Type S-4 Control Room Control Room 4-2 The inability to bring the reactor Special Safety Limiting Conditions for to the cold shut- Operational Criteria Condensate Storage System Operation for Special Safety down condition from Systems outside the control room. Reactor Core Isolation Surveillance Requirements for Cooling System Special Safety Systems Pressure Relief System Reactor Protection System Control Rod Drive System RHR (containment Cooling)

BFN-21 Table 1.4-2A (Sheet 6) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION SAFETY CONSIDERATIONS Types of Requirements to Types of Actions Required be Observed in Operation Type of Operation Types of Applicable to Avoid Unacceptable Types of Systems Required of Plant to Avoid or Event Unacceptable Safety Results Criteria Results to Carry Out Action Unacceptable Results

5. Special Event- 5-1 The inability to Nuclear Safety Design Safety Action-Type S-5 Safety Systems-Type S-5 Operational Nuclear Safety Inability to shut down the Criteria-Type S-5 Requirements-Type S-5 Shut Down reactor independent Special Safety Action Special Safety Systems Reactor With of control rods Operational Nuclear Safety Control Rods Limits-Type S-5 5-2 The inability to Nuclear Safety maintain the Operational Criteria Shutdown Without Control Standby Liquid Control reactor in the Type S-5 Rods System shutdown condition Technical Specifications-independent of Special Safety Design Maintain Shutdown During RWCU Isolation Type S-5 control rods Criteria Reactor Cooldown Limiting Conditions for Operation for Special Safety Systems Special Safety Operational Criteria Surveillance Requirements for Special Safety Systems

BFN-21 Table 1.4-2B (Sheet 1) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results

1. Planned Operation 1-1 Inability to generate Power Generator Design Power Generation Action- Power Generator Systems- Operational Power Generator electrical power Criteria-Type PG-1 Type PG-1 Type PG-1 Requirements-Type PG-1 1-2 Fuel Failure Power Generator Operational Power Generator Operational Criteria - Limits-Type PG-1 Type PG-1 1-3 Inability to Perform Process Action Process Systems Routine Maintenance (A Category of Power (A Category of Power with Plant at Power Generation Action) Generator Systems)

Process Design Criteria Normal Operating Procedures 1-4 Inability to Optimize Fuel Performance Process Operational Indications of Process Indicators Maintenance Procedures Criteria Variables 1-5 Inability to Respond Process Operations Process Computer System Calibration Procedures to Changes in Power Demand Fuel Performance Recirculation Flow Control Refueling Procedures Calculations System 1-6 Inability to Shut Down Power Level Control Reactor with Control Reactor Manual Control Rods in the Normal Consideration of Exhaust System Manner Steam Control Rod Drive System Feedwater System Turbine-Generator Main Condenser

BFN-21 Table 1.4-2B (Sheet 2) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Types of Requirements Power Generation (Where Required to Avoid Required to Avoid to be Observed in More Restrictive Than Unacceptable Results Unacceptable Results Operation of Plant Type of Operation Unacceptable Safety Types of Applicable (Where Not Required (Where Not Required to Avoid Unacceptable or Event Results) Criteria as a Safety Action) as a Safety Action) Results

2. Abnormal 2-1 Fuel Failure Power Generation Design Power Generation Action Power Generation Systems Operational Power Generation Operational Criteria-Type PG-2 Type PG-2 Type PG-2 Requirements-Type PG-2 Transients 2-2 The Lifting of Main Steam Power Generation Operational Power Generation Relief Valves Operational Criteria Limits-Type PG-2 Type PG-2 2-3 Conditions Requiring the Rod Block Reactor Manual Control Opening of the Reactor System (Rod Block)

Vessel for Inspection or Pressure Relief Repair Pressure Relief System Normal Operating Procedures Refueling Block 2-4 Inability to Return to Scram Refueling Interlocks Post Transient Recovery Power Operation Procedures 2-5 Inadvertent Criticality Core Cooling Reactor Protection Refueling Restrictions During Refueling System (RPS) Electro Hydraulic Control (EHC) System Reactor Core Isolation Cooling (RCIC)

BFN-21 Table 1.4-2B (Sheet 3) BROWNS FERRY NUCLEAR PLANT CLASSIFICATION OF BWR SYSTEMS, CRITERIA, AND REQUIREMENTS FOR SAFETY EVALUATION ACTUAL PLANT DESIGN AND OPERATION POWER GENERATION CONSIDERATIONS Unacceptable Results for Types of Actions Types of Systems Power Generation Required to Avoid Required to Avoid Types of Requirements (Where More Restrictive Unacceptable Results Unacceptable Results to be Observed in Operation Type of Operation Than Unacceptable Types of Applicable (Where Not Required (Where Not Required of Plant to Avoid or Event Safety Results) Criteria as a Safety Action) as a Safety Action) Unacceptable Results

3. Accidents 3-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems - Operational Power Generation to Power Operation Criteria-Type PG-3 Type PG-3 Type PG-3 Requirements-Type PG-3 Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-3 Type PG-3 Post Accident Recovery Procedures

4. Special Event 4-1 Inability to Return Power Generation Design Power Generation Action- Power Generation Systems- Operational Power Generation Loss of to Power Operation Criteria-Type PG-4 Type PG-4 Type PG-4 Requirements-Type PG-4 Habitability of the Control Power Generation Operational Power Generation Room Operational Criteria - Limits-Type PG-4 Type PG-4 Post Event Recovery Procedures

5. Special Event 5-1 Inability to Return Power Generation Design Power Generation Action - Power Generation Systems- Operational Power Generation Inability to to Power Operation Criteria-Type PG-5 Type PG-5 Type PG-5 Requirements-Type PG-5 Shut Down Reactor With Control Rods Power Generation Operational Power Generation Operational Criteria - Limits-Type PG-5 Type PG-5 Post Event Recovery Procedures

BFN-19 1.5 PRINCIPAL DESIGN CRITERIA There are two ways of considering principal design criteria. One way is to consider the criteria on a system-by-system (or system group) basis. The second way is to consider criteria classification-by-classification as given in Tables 1.4-2 A and B. In the classification-by-classification approach, the criteria must be stated in sufficient detail to allow placement of each criterion into one classification category. Thus, there may be closely related criteria pertaining to any given system in each classification category. This is a natural outgrowth of the functional (unacceptable result) approach to classification. The actual design of a system must reflect all of the criteria that pertain to it; thus, the less restrictive (but more important) criteria pertaining to the system in the classification approach will be masked by the more restrictive (and less important) criteria. Safety analysis requires the information gained in the classification-by-classification approach to criteria, but system description is more easily understood through the system-by-system method. Both approaches to criteria are given in this section; both are useful. 1.5.1 Principal Design Criteria Classification-By-Classification The principal architectural and engineering criteria for the design and construction of the plant are summarized below. The criteria are grouped according to the classification plan given in Tables 1.4-2 A and B. Some of the more general criteria are so broad that they are applicable, at least in part, to more than one classification. In these very general cases, all of the affected classifications are indicated. Specific design bases and design features are detailed in other sections of this report. Criteria pertaining to operation of the plant are given in Appendix G. 1.5.1.1 General Criteria Applicable Classifications Criteria PG-1,S-1,S-2,S-3 1. The plant shall be designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner. The plant design shall be in accordance with applicable codes and regulations. S-1,S-2,S-3 2. The plant shall be designed in such a way that the release of radioactive materials to the environment is limited so that the limits and guideline values of 1.5-1

BFN-19 applicable regulations pertaining to the release of radioactive materials are not exceeded. S-1,S-2,S-3,S-4 3. The reactor core and reactivity control system shall be designed so that control rod action shall be capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion. S-1,S-2,S-3 4. Adequate strength and stiffness with appropriate safety factors shall be provided so that a hazardous release of radioactive material shall not occur. 1.5.1.2 Power Generation Design Criteria, Type PG-1 (Planned Operation)

1. The nuclear system shall employ a General Electric boiling water reactor to produce steam for direct use in a turbine generator.

2. The fuel cladding shall be designed to retain integrity as a radioactive material barrier for the design power range.

3. The fuel cladding shall be designed to accommodate, without loss of integrity, the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

4. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions from plant shutdown to design power. The capacity of such systems shall be adequate to prevent fuel clad damage.

5. (Deleted).

6. It shall be possible to manually control the reactor power level.

7. Control of the nuclear system shall be possible from a single location.

8. Nuclear system process controls shall be arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

9. Fuel handling and storage facilities shall be designed to maintain adequate shielding and cooling for spent fuel.

1.5-2

BFN-19

10. Interlocks or other automatic equipment shall be provided as a backup to procedural controls to avoid conditions requiring the functioning of nuclear safety systems or engineered safeguards.

1.5.1.3 Power Generation Design Criteria, Type PG-2 (Abnormal Operational Transients)

1. The fuel cladding, in conjunction with other plant systems, shall be designed to retain integrity throughout any abnormal operational transient.

2. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for any abnormal operational transient. The capacity of such systems shall be adequate to prevent fuel clad damage.

3. Heat removal systems shall be provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems shall be adequate to prevent fuel clad damage.

4. Standby electrical power sources shall be provided to allow removal of decay heat under circumstances where normal auxiliary power is not available.

5. Fuel handling and storage facilities shall be designed to prevent inadvertent criticality.

1.5.1.4 Nuclear Safety Design Criteria, Type S-1 (Planned Operation)

1. The Plant shall be designed so that fuel failure during planned operation is limited to such an extent that, were the freed fission products released to the environs via the normal discharge paths for radioactive materials, the limits of 10 CFR 20 would not be exceeded.

2. The reactor core shall be designed so that its nuclear characteristics exhibit no tendency toward a divergent power transient.

3. The nuclear system shall be so designed that there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.

4. Gaseous, liquid, and solid waste disposal facilities shall be so designed that the discharge and offsite shipment of radioactive effluents can be made in accordance with applicable regulations.

1.5-3

BFN-19

5. The design shall provide means by which plant operations personnel can be informed whenever limits on the release of radioactive material are exceeded.

6. Sufficient indications shall be provided to allow determination that the reactor is operating within the envelope of conditions considered by plant safety analysis.

7. Radiation shielding shall be provided and access control patterns shall be established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal plant operation.

1.5.1.5 Nuclear Safety Design Criteria, Type S-2 (Abnormal Operational Transients)

1. The plant shall be so designed that fuel failure as a result of any abnormal operational transient is limited to such an extent that, were the freed fission products released to the environs via the normal discharge paths for radioactive materials, the limits of 10 CFR 20 would not be exceeded.

2. Those portions of the nuclear system which form part of the nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients.

3. Nuclear safety systems shall act to assure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients.

4. Where positive, precise action is immediately required in response to abnormal operational transients, such action shall be automatic and shall require no decision or manipulation of controls by plant operations personnel.

5. Essential safety actions shall be carried out by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE-279 is applicable, single failures of passive electrical components will be considered, as well as single failure of active components in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.

6. The design of nuclear safety systems shall include allowances for environmental phenomena at the site.

7. Provision shall be made for control of active components of nuclear safety systems from the control room.

1.5-4

BFN-19

8. Nuclear safety systems shall be designed to permit demonstration of their functional performance requirements.

9. Standby electrical power sources shall be provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

10. Standby electrical power sources shall have sufficient capacity to power all nuclear safety systems requiring electrical power.

1.5.1.6 Nuclear Safety Design Criteria, Type S-3 (Accidents)

1. Those portions of the nuclear system which form part of the nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following accidents. For accidents in which one breach in the nuclear system process barrier is postulated, such breach shall not cause additional breaches in the nuclear system process barrier.

2. Engineered safeguards shall act to assure that no damage to the nuclear system process barrier results from internal pressures caused by an accident.

3. Where positive, precise action is immediately required in response to accidents, such action shall be automatic and shall require no decision or manipulation of controls by plant operations personnel.

4. Essential safety actions shall be carried out by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE-279 is applicable, single failures of passive electrical components will be considered, as well as single failure of active components in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.

5. Features of the plant which are essential to the mitigation of accident consequences shall be designed so that they can be fabricated and erected to quality standards which reflect the importance of the safety action to be performed.

6. The design of engineered safeguards shall include allowances for environmental phenomena at the site.

7. Provision shall be made for control of active components of engineered safeguards from the control room.

1.5-5

BFN-19

8. Engineered safeguards shall be designed to permit demonstration of their functional performance requirements.

9. A primary containment shall be provided that completely encloses the reactor vessel.

10. The primary containment shall be designed to retain integrity as a radioactive material barrier during and following accidents that release radioactive material into the primary containment volume.

11. It shall be possible to test primary containment integrity and leak tightness at periodic intervals.

12. A secondary containment shall be provided that completely encloses both the primary containment and fuel storage areas.

13. The secondary containment shall be designed to act as a radioactive material barrier under the same conditions that require the primary containment to act as a radioactive material barrier.

14. The secondary containment shall be designed to act as a radioactive material barrier, if required, whenever the primary containment is open for expected operational purposes.

15. The primary and secondary containments, in conjunction with other engineered safeguards, shall act to prevent the radiological effects of accidents resulting in the release of radioactive material to the containment volumes from exceeding the guideline values of applicable regulations.

16. Provisions shall be made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment.

17. Piping that penetrates the primary containment structure, and which could serve as a path for the uncontrolled release of radioactive material to the environs, shall be automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.

18. Core Standby Cooling Systems shall be provided to prevent excessive fuel clad temperatures as a result of a loss-of-coolant accident.

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19. The Core Standby Cooling Systems shall provide for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.

20. The Core Standby Cooling Systems shall be diverse, reliable and redundant.

21. Operation of the Core Standby Cooling Systems shall be initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the plant.

22. Standby electrical power sources shall have sufficient capacity to power all engineered safeguards requiring electrical power.

23. The control room shall be shielded against radiation so that occupancy under accident conditions is possible.

1.5.1.7 Nuclear Safety Design Criteria, Type S-4 (Special Event In the event that the control room becomes inaccessible, it shall be possible to bring the reactor from power range operation to cold shutdown (Mode 4) by manipulation of the local controls and equipment which are available outside the control room. 1.5.1.8 Nuclear Safety Design Criteria, Type S-5 (Special Event) Backup reactor shutdown capability shall be provided independent of normal reactivity control provisions. This backup system shall have the capability to shut down the reactor from any normal operating condition, and subsequently to maintain the shutdown condition. 1.5.2 Principal Design Criteria, System-By-System The principal architectural and engineering criteria for design are summarized below on a system-by-system or system group basis. The system-by-system presentation facilitates the understanding of the actual design of any one system, but significant distinctions in the importance to safety of different criteria pertaining to a system cannot be made clear, as they are in the classification-by-classification presentation. To make consistent judgments regarding plant safety, the classification-by- classification approach to criteria must be used. In the system-by-system presentation of criteria, only the most restrictive of any related criteria are stated for a system. Where the most restrictive criterion is one which is classified as a power generation consideration in Table 1.4-2B, less 1.5-7

BFN-19 restrictive, but more important, safety criteria may be hidden (not stated) in the system-by-system presentation. 1.5.2.1 General Criteria

1. The plant shall be designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner. The plant design shall be in accordance with applicable codes and regulations.

2. The plant shall be designed in such a way that the release of radioactive materials to the environment is limited, so that the limits and guideline values of applicable regulations pertaining to the release of radioactive materials are not exceeded.

1.5.2.2 Nuclear System Criteria

1. The nuclear system shall employ a General Electric boiling water reactor to produce steam for direct use in a turbine-generator.

2. The fuel cladding shall be designed to retain integrity as a radioactive material barrier for the design power range and for any abnormal operational transient.

3. Those portions of the nuclear system which form part of the nuclear system process barrier shall be designed to retain integrity as a radioactive material barrier following abnormal operational transients and accidents. For accidents in which one breach in the nuclear system process barrier is postulated, such breach shall not cause additional breaches in the nuclear system process barrier.

4. The fuel cladding shall be designed to accommodate, without loss of integrity, the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

5. Heat removal systems shall be provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions from plant shutdown to design power, and for any abnormal operational transient. The capacity of such systems shall be adequate to prevent fuel clad damage.

6. Heat removal systems shall be provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems shall be adequate to prevent fuel clad damage.

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7. The reactor core and reactivity control system shall be designed so that control rod action shall be capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.

8. The nuclear system shall be so designed that there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.

9. The reactor core shall be so designed that its nuclear characteristics exhibit no tendency toward a divergent power transient.

1.5.2.3 Power Conversion Systems Criteria

1. Appropriate power conversion systems shall be provided to efficiently convert the heat energy of the steam produced in the reactor vessel to mechanical energy for turning a generator to produce electrical power.

2. Means shall be provided for furnishing makeup (feedwater) to the reactor vessel to allow continued operation.

1.5.2.4 Electrical Power Systems Criteria

1. A generator capable of efficiently producing electric power shall be provided.

2. Electrical power for protection systems and engineered safeguards shall be available from two offsite sources so that no single failure in the facility can result in loss of offsite power.

1.5.2.5 Radioactive Waste Disposal Criteria

1. Gaseous, liquid, and solid waste disposal facilities shall be designed so that the discharge and offsite shipment of radioactive effluents can be made in accordance with applicable regulations.

2. The design shall provide means by which plant operations personnel can be informed whenever operational limits on the release of radioactive material are exceeded.

1.5.2.6 Nuclear Safety Systems and Engineered Safeguards Criteria 1.5.2.6.1 General

1. Nuclear safety systems shall act in response to abnormal operational transients to limit fuel damage such that, were the freed fission products 1.5-9

BFN-19 released to the environs via the normal discharge paths for radioactive material, the limits of 10 CFR 20 would not be exceeded.

2. Nuclear safety systems and engineered safeguards shall act to assure that no damage to the nuclear system process barrier results from internal pressures caused by abnormal operational transients or accidents.

3. Where positive, precise action is immediately required in response to accidents, such action shall be automatic and shall require no decision or manipulation of controls by plant operations personnel.

4. Essential safety actions shall be carried out by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE-279 is applicable, single failures of passive electrical components will be considered, as well as single failure of active components in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.

5. Features of the plant which are essential to the mitigation of accident consequences shall be designed so that they can be fabricated and erected to quality standards which reflect the importance of the safety function to be performed.

6. The design of nuclear safety systems and engineered safeguards shall include allowances for environmental phenomena at the site (e.g., weather extremes and proximity to other high energy systems). Furthermore, electrical equipment in these systems shall be capable of performing their safety function as required under environmental conditions associated with all normal, abnormal, and plant accident operation.

7. Provision shall be made for control of active components of nuclear safety systems and engineered safeguards from the control room.

8. Nuclear safety systems and engineered safeguards shall be designed to permit demonstration of their functional performance requirements.

1.5.2.6.2 Containment and Isolation Criteria

1. A primary containment shall be provided that completely encloses the reactor vessel.

2. The primary containment shall be designed to retain integrity as a radioactive material barrier during and following accidents that release radioactive material into the primary containment volume.

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3. It shall be possible to test primary containment integrity and leak tightness at periodic intervals.

4. A secondary containment shall be provided that completely encloses both the primary containment and fuel storage areas.

5. The secondary containment shall be designed to act as a radioactive material barrier under the same conditions that require the primary containment to act as a radioactive material barrier.

6. The secondary containment shall be designed to act as a radioactive material barrier, if required, whenever the primary containment is open for expected operational purposes.

7. The primary and secondary containments, in conjunction with other engineered safeguards, shall act to prevent the radiological effects of accidents resulting in the release of radioactive material to the containment volumes from exceeding the guideline values of applicable regulations.

8. Provisions shall be made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment.

9. Piping that penetrates the primary containment structure, and could serve as a path for the uncontrolled release of radioactive material to the environs, shall be automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.

1.5.2.6.3 Core Standby Cooling Criteria

1. Core Standby Cooling Systems shall be provided to prevent excessive fuel clad temperatures as a result of a loss-of-coolant accident.

2. The Core Standby Cooling Systems shall provide for continuity of core cooling over the complete range of postulated break sizes in the nuclear system process barrier.

3. The Core Standby Cooling Systems shall be diverse, reliable, and redundant.

4. Operation of the Core Standby Cooling systems shall be initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the plant.

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BFN-19 1.5.2.6.4 Standby Power Criteria

1. Standby electrical power sources shall be provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

2. Standby electrical power sources shall have sufficient capacity to power all engineered safeguards requiring electrical power.

1.5.2.7 Reactivity Control Criteria

1. Backup reactor shutdown capability shall be provided independent of normal reactivity control provisions. This backup system shall have the capability to shut down the reactor from any operating condition, and subsequently to maintain the shutdown condition.

2. In the event that the control room is inaccessible, it shall be possible to bring the reactor from power range operation to cold shutdown (Mode 4) by manipulation of the local controls and equipment which are available outside the control room.

1.5.2.8 Process Control Systems Criteria 1.5.2.8.1 Nuclear System Process Control Criteria

1. It shall be possible to manually control the reactor power level.

2. Control of the nuclear system shall be possible from a single location.

3. Nuclear system process controls shall be arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

4. Interlocks or other automatic equipment shall be provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear safety systems or engineered safeguards.

1.5.2.8.2 Deleted 1.5.2.8.3 Electrical Power Systems Process Control Criteria Controls shall be provided in the electrical power systems to protect against faults and to increase the reliability of incoming and outgoing power. 1.5-12

BFN-19 1.5.2.9 Auxiliary Systems Criteria

1. Fuel handling and storage facilities shall be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel.

2. Means shall be provided to remove heat from process systems that is generated through operation of the plant.

3. Fire detection and protection systems capable of protecting the plant against all types of fires shall be provided.

4. Means shall be provided to adequately heat, ventilate, and air-condition plant buildings for personnel comfort and equipment protection.

5. Means shall be provided to furnish other auxiliary services as required for safe and efficient operation of the plant.

1.5.2.10 Shielding and Access Control Criteria

1. Radiation shielding shall be provided and access control patterns shall be established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal plant operation.

2. The control room shall be shielded against radiation so that occupancy under accident conditions is possible.

1.5.2.11 Structural Loading Criteria Adequate strength and stiffness, with appropriate safety factors, shall be provided so that a hazardous release of radioactive material shall not occur. Details of implementation are given in Chapter 12 and Appendix C. 1.5-13

BFN-30 1.6 PLANT DESCRIPTION 1.6.1 General 1.6.1.1 Site and Environs 1.6.1.1.1 Location and Size of Site The site contains approximately 880 acres and is located on the north shore of Wheeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. It is approximately 30 miles west of Huntsville, Alabama. 1.6.1.1.2 Site Ownership The plant is located on property owned by the United States and in the custody of TVA. 1.6.1.1.3 Activities at the Site Activities at the site are those performed by TVA in operating the three-unit nuclear plant to produce electric power. 1.6.1.1.4 Access to the Site (See Figure 2.2-4) The three-unit plant, including the intake and discharge canals, is enclosed by a security fence. Primary access to the plant area is by way of an access road through a security gate. 1.6.1.1.5 Description of the Environs (See Table 2.2-6) The Browns Ferry site is located in an area where the land is used primarily for agriculture. Population densities are low, with a projected population of 33,340 within ten miles for the year 2020. There are no population centers of significance within ten miles of the plant. The low population zone is determined to be seven miles. 1.6.1.1.6 Geology The site is underlain by massive formations of nearly horizontal bedrock. Historically, this region has been one of little structural deformation, and major folds and faults are entirely absent. 1.6-1

BFN-27 1.6.1.1.7 Seismology There has been no known major seismic activity originating in or near the site area. The major seismic activity experienced at the site has been caused by distant major earthquakes. 1.6.1.1.8 Hydrology Groundwater movement in the area is from the plant site to the Tennessee River. A thick mantle of residuum in the site area retards the movement of shallow groundwater. 1.6.1.1.9 Regional and Site Meteorology The meteorology of the Browns Ferry site provides generally favorable atmospheric conditions for dispersion of plant emissions. The immediate terrain is flat and slightly undulating, with scattered 400- to 600-foot foothills. Thus, local entrapment or accumulation of emissions should not occur. 1.6.1.1.10 Design Bases Dependent Upon the Site and Environs

a. Offgas Systems The plant offgas systems are designed to maintain gaseous waste releases to the environment, during normal operation, at levels which assure that concentrations at the site boundary will be within the limits of 10 CFR 20. The effects of releases at or beyond the site boundary resulting from the design basis accidents will be within the reference values of 10 CFR 50.67.

b. Liquid Waste Effluents The plant Liquid Radwaste System is designed to maintain liquid waste releases to the environment at levels which comply with the plant's National Pollutant Discharge Elimination System (NPDES) permit limitations and assure that concentrations at the site boundary will be within the limits of 10 CFR 20.

c. Wind Loading Design A structural design capable of withstanding loadings resulting from a 100-mph sustained wind, except for Low Level Radioactive Waste Storage Facility (LLRWSF) which uses a wind of 95 mph, is considered appropriate. All Class I structures and equipment that are required to support and maintain safe shut down of all the units as a result of a tornado design basis event are designed to maintain their integrity when subjected to loading resulting from a 300-mph 1.6-2

BFN-28 tornado. The LLRWSF is designed for a 290 mph rotational speed at 150 feet radius and 70 mph translational speed. For the 600 foot tall reinforced concrete chimney, only the bottom 280 feet of the chimney is designed for the 300-mph tornado. The top 320 feet is designed only for the 100-mph sustained wind. See Section 12 for greater detail of design wind and tornado loadings.

d. Seismic Design The design of all Class I structures is based on a ground motion due to an acceleration of 0.10g (Operating Basis Earthquake). In addition, the design is such that the plant can be safely shut down during a ground acceleration of 0.20g (Design Basis Earthquake).

e. Flooding Plant grade is established at 565 feet above mean sea level. The probable maximum flood at Browns Ferry would reach El. 572.5, plus wind wave runup produced by a coincidental 45 MPH sustained wind speed.

f. Loss of Normal Heat Sinks (Downstream Dam Failure)

Failure of the downstream Wheeler Dam was assumed as part of the basis for the UHS. The assumed failure results in a consequential failure of the Wilson Dam further downstream. If t h e Wheeler Dam downstream from the plant site were to fail, Wheeler Reservoir would pot ent ial ly drain down to a minimum elevation of 529 feet at the plant intake due to a high point in the Tennessee River basin downstream of BFN at approximately river mile 291.8. A pool of water approximately 1,000 feet wide and 7 miles long, containing a volume of about 69.6X10 6 cubic feet of water, would be available at the plant intake. The resultant pool elevation is more than sufficient to maintain adequate flow and NPSH to the RHRSW p umps, which supply shutdown cooling water to the three units. No credit is taken for normal effluent flow rates from the upstream Guntersville Dam. However, an evaluation of historical data for the upstream Guntersville Dam found that the daily flow rate exceeded 1,319 cubic feet per second (cfs) for greater than 99 percent of the historical time period and that the average daily flow rate was 33,500 cfs. Within the historical data, the lowest recorded daily flow rate was found to be 100 cfs, which is above the minimum required 1.6-3

BFN-28 RHRSW flow of 80 cfs for one unit responding to an accident and two units in shutdown. Therefore, inflow contribution to the UHS from the upstream Guntersville Dam at the lowest recorded daily average flow rate combined with the dependable watershed drainages upstream of BFN will provide at least 100 cfs inflow to the UHS. The inflow to the UHS from the upstream sources provides additional assurance that resultant pool elevation is maintained at least 529 feet with more than the required 80 cfs of cooling water available.

g. Environmental Radiation Monitoring System The availability of past wind direction and persistence data and river flow records, along with knowing the location of population centers, has aided in the selection of monitoring locations and frequency of sampling.

1.6.1.2 Facility Arrangement The facility arrangement is shown in Figure 2.2-4. Plan and elevation views of the major buildings are shown in Figures 1.6-1 through 1.6-27. 1.6.1.3 Nuclear System Each nuclear system includes a single-cycle, forced-circulation, General Electric boiling water reactor producing steam for direct use in a steam turbine. A typical heat balance showing the major parameters of the nuclear system for the rated power condition is shown in Figure 1.6-28 (3952 MWt). 1.6.1.3.1 Reactor Core and Control Rods The fuel for the reactor core consists of uranium dioxide pellets made from slightly enriched uranium. These pellets are contained in sealed Zircaloy-2 tubes. These fuel rods are assembled into individual fuel bundles. The detailed description of fuel in the reactor core is given in Section 3.2 of the FSAR. The description of the core for each unit is given in the current reload licensing document for that unit as described in FSAR Appendix N. 1.6-4

BFN-27 1.6.1.3.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structure, the steam separators and dryers, the jet pumps, the control rod guide tubes, distribution lines for the feedwater, core spray, and standby liquid control, the incore instrumentation, and other components. The main connections to the vessel include the steam lines, the coolant recirculation lines, feedwater lines, control rod drive housings, and core standby cooling lines. Each reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure is 1050 psia (uprated) in the steam space above the separators. The vessel is fabricated of carbon steel and is clad internally (except for the top head) with weld overlay. The reactor core is cooled by demineralized water which enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers, located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steam lines. Each steam line is provided with two isolation valves in series--one on each side of the primary containment barrier. 1.6.1.3.3 Reactor Recirculation System The Reactor Recirculation System pumps reactor coolant through the core to remove the energy generated in the fuel. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. Each loop has one motor-driven recirculation pump. Recirculation pump speed can be varied to allow control of reactor power level through the effects of coolant flow rate on moderator void content. For Unit 2 only, the two recirculation loops have a cross-connect line with one normally closed valve and one normally open valve to prevent pressure buildup between the valves. 1.6.1.3.4 Residual Heat Removal System The Residual Heat Removal System (RHRS) is a system of pumps, heat exchangers, and piping that fulfills the following functions.

a. Removal of decay heat during and after plant shutdown.

b. Injection of water into the reactor vessel following a loss-of-coolant accident rapidly enough to reflood the core and prevent excessive fuel clad temperatures independent of other core cooling systems. This is discussed in paragraph 1.6.2 (Nuclear Safety Systems and Engineered Safeguards).

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BFN-28

c. Removal of heat from the primary containment following a loss-of-coolant accident to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the water inside the primary containment. The redundancy of the equipment provided for containment cooling is further extended by a separate part of the RHRS which sprays cooling water into the drywell and pressure suppression pool.

d. Provide standby cooling.

e. Provide assistance for fuel pool cooling when required.

1.6.1.3.5 Reactor Water Cleanup System A Reactor Water Cleanup System, which includes a demineralizer arrangement, is provided to clean up the reactor cooling water, to reduce the amounts of activated corrosion products in the water, and to remove reactor coolant from the nuclear system under controlled conditions. 1.6.1.3.6 Reactor Core Isolation Cooling System The Reactor Core Isolation Cooling System (RCICS) provides makeup water to the reactor vessel whenever the vessel is isolated. The RCICS uses a steam-driven, turbine-pump unit and operates automatically to maintain adequate reactor vessel water level. 1.6.1.4 Power Conversion Systems The Power Conversion Systems use the steam produced in the reactor vessel to produce electrical power. Figure 1.6-29, Sheets 1, 2, and 3, shows the turbine generator heat balance for rated power conditions. Figure 1.6-30 is a flow diagram for general plant systems. 1.6.1.4.1 Turbine Generator The turbine is an 1,800-rpm tandem-compound, six-flow, nonreheat unit. It has a double-flow, high-pressure cylinder and three double-flow low-pressure cylinders. Steam from the high-pressure cylinder passes through moisture separators before entering the low-pressure units. The turbine has five extraction stages for reactor feedwater 1.6-6

BFN-27 system heating. Turbine controls include an electric-hydraulic speed governor, overspeed protection, steam admission control valves, emergency stop valves, combined intermediate stop-intercept valves, bypass valves, and pressure regulators. The electrical generator is direct-driven, hydrogen-cooled with liquid-cooled stator, and is equipped with an automatic voltage regulator bus-fed from auxiliary transformers. 1.6.1.4.2 Turbine Bypass System The Turbine Bypass System is provided to pass steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the load permitted to pass to the turbine generator (such as during generator synchronization or following sudden load changes). 1.6.1.4.3 Main Condenser Three deaerating, single-pass, single-pressure, radial-flow-type surface condensers provide the primary heat sinks for each turbine-generator. Each condenser is located beneath one of the low-pressure turbines with the tubes oriented transverse to the turbine-generator axis. Baffling in the hotwell is arranged to ensure a minimum of 1.5 minutes retention time for the condensate. 1.6.1.4.4 Main Condenser Gas Removal and Turbine Sealing Systems Two 100-percent capacity steam jet air ejectors are provided for each unit to remove air and noncondensables from the main condensers during normal operation. A mechanical vacuum pump is provided for startup operation. The Turbine Sealing System is provided to prevent steam leakage and air inleakage at the turbine seals. 1.6.1.4.5 Condenser Circulating Water System Seven mechanical-draft cooling towers are provided to dissipate waste heat to the atmosphere. Water is pumped through the main condenser to an open channel going to the towers of the circulating water pumps for each unit. Water is pumped to each cooling tower by lift pumps. The system is designed for open and helper modes of operation. In the open mode, water is drawn into the circulating water pumping station forebay from Wheeler reservoir, pumped through the main condenser, and discharged back into the reservoir through a diffuser discharge system consisting of perforated metal pipes which extend across the reservoir channel to diffuse the warmer water from 1.6-7

BFN-27 the plant. In the helper mode, the water is pumped from the reservoir, through the plant, and into an open channel going to the cooling towers where it is pumped through the towers and is returned to the reservoir through the diffusers. 1.6.1.4.6 Condensate Filter/Demineralizer System This full-flow system removes dissolved and suspended solids from the condensate, providing high-quality water for the nuclear system. It consists of filter/demineralizer vessels containing filter elements which are coated with a mixture of powdered cation and anion exchange resins. These resins perform both the filtration and deionization functions. 1.6.1.4.7 Condensate and Reactor Feedwater Systems The Condensate and Reactor Feedwater Systems take suction from the main condensers and deliver demineralized water to the reactor vessel at an elevated temperature and pressure. Three vertical, centrifugal, motor-driven condensate pumps; three horizontal, centrifugal, motor-driven condensate booster pumps; and three horizontal, centrifugal, single-stage reactor feedwater pumps with variable-speed steam turbines are provided for these systems. Feedwater is controlled by varying the speed of the reactor feedwater-pump turbine-drives. Five stages of feedwater heating are provided for each of the three feedwater streams. All heaters are of the two-pass, U-tube type. 1.6.1.5 Electrical Power Systems Each generator produces electrical power at 22-kV. This 22-kV generator output is transmitted through isolated-phase buses to a bank of three single-phase main power transformers, where the voltage is stepped up to 500-kV and transmitted to the 500-kV switchyard. The 500-kV switchyard connects the plant to the TVA 500-kV system. The plant has generator breakers so that startup and shutdown are from the 500-kV system. The 161-kV system is also available to provide plant startup and shutdown power. 1.6.1.6 Radioactive Waste Systems The Radioactive Waste Systems are designed to control the release of plant-produced radioactive material to within the limits specified in the ODCM and NPDES permits. The methods employed for the controlled release of those contaminants are dependent primarily upon the state of the material: liquid, solid, or gaseous. 1.6-8

BFN-27 1.6.1.6.1 Liquid Radwaste System The Liquid Radioactive Waste Control System collects, treats, stores, and disposes of all radioactive liquid wastes. These wastes are collected in sumps and drain tanks at various locations throughout the plant and then transferred to the appropriate collection tanks in the Radwaste Building for treatment, storage, and disposal. Wastes to be discharged from the system are processed on a batch basis, with each batch being processed by such method or methods appropriate for the quality and quantity of materials determined to be present. Processed liquid wastes may be returned to the condensate system or discharged to the environs through the circulating water discharge canal. The liquid wastes in the discharge canal are diluted with condenser effluent circulating water to achieve a permissible concentration at the site boundary. Batches of low-conductivity liquid waste are processed through a filter and a waste demineralizer. Demineralizer effluent is sent to a waste sample tank. Depending upon the conductivity and level of radioactivity, the liquid may then be discharged to the circulating-water discharge canal or the cooling tower blowdown line, transferred to condensate storage tanks, returned for further processing through the waste demineralizer. High-conductivity liquids are processed through a filter and are collected in a floor drain sample tank. If the concentration after dilution is less than or equal to the applicable limits, the filtered liquid may be discharged. An alternate method of processing low and high conductivity liquid is the use of vendor supplied skid mounted equipment, interconnected to the permanent Radwaste System. Depending on effluent quality and plant needs, the water can be sent to either the waste sample tank or floor drain sample tank. Processing from the waste sample tank or floor drain sample tank is identical as described above. Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with minimum personnel exposure. For example, tanks and processing equipment which will contain significant radiation sources are located behind shielding; and sumps, pumps, instruments, and valves are located in controlled access rooms or spaces. Processing equipment is selected and designed to require a minimum of maintenance. Protection against accidental discharge of liquid radioactive waste is provided by valving redundance, instrumentation for detection with alarms of abnormal conditions, procedural controls, interlocks, and radiation monitor controlled valves 1.6-9

BFN-27 1.6.1.6.2 Solid Radwaste System With the Solid Radwaste System, solid radioactive wastes are collected, processed, and packaged for storage. Generally, these wastes are stored onsite until the short half-lived activities are insignificant. Solid wastes from equipment originating in the nuclear system are stored for radioactive decay in the fuel storage pool and prepared for reprocessing or offsite storage. Examples of these wastes are spent fuel, spent control rods, incore ion chambers, etc. Process solid wastes are collected, dewatered, and loaded in shielded containers for storage and shipping. Examples of these solid wastes are spent demineralizer resins and filter aid. Wastes such as paper, rags, and used clothing are placed into containers for storage and shipment. 1.6.1.6.3 Gaseous Radwaste System The Gaseous Radwaste System collects, processes, and delivers to the plant stack, for elevated release to the atmosphere, gases from each main condenser air ejector, startup vacuum pump, condensate drain tank vent, and steam packing exhauster. Gases from each main condenser air ejector are passed through a preheater, a catalytic recombiner, a condenser, a moisture separator, and a dehumidification coil. The gases then enter a decay pipe which provides a retention time of approximately 6 hours, during which N-16 and 0-19 decay to negligible levels. The gases are then passed through a cooler-condenser, a moisture separator, a reheater, a prefilter, six charcoal beds, an afterfilter, and mixed with dilution air, after which they are exhausted to the stack. The charcoal beds provide about 9.7 hours retention for krypton isotopes and 7.3 days retention for xenon isotopes. Gland seal and startup vacuum-pump gases are held up for approximately 1 3/4 minutes, to allow sufficient decay of N-16 and 0-19, and then passed directly to the stack for release. 1.6.2 Nuclear Safety Systems and Engineered Safeguards 1.6.2.1 Reactor Protection System The Reactor Protection System initiates a rapid, automatic shutdown (scram) of the reactor. This action is taken in time to prevent excessive fuel cladding damage and any nuclear system process barrier damage following abnormal operational transients. The Reactor Protection System overrides all operator actions and process controls. 1.6-10

BFN-27 1.6.2.2 Neutron Monitoring System Although not all of the Neutron Monitoring System qualifies as a nuclear safety system, those portions that provide high neutron flux signals to the Reactor Protection System do. The intermediate range monitors (IRM) and average power range monitors (APRM), which monitor neutron flux via incore detectors, signal the Reactor Protection System to scram in time to prevent excessive fuel cladding damage as a result of overpower transients. 1.6.2.3 Control Rod Drive System When a scram is initiated by the Reactor Protection System, it is the Control Rod Drive System that inserts the negative reactivity necessary to shut down the reactor. Each control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high pressure water from an accumulator for each rod forces each control rod rapidly into the core. 1.6.2.4 Nuclear System Pressure Relief System A pressure relief system consisting of relief valves mounted on the main steam lines is provided to prevent excessive pressure inside the nuclear system following either abnormal operational transients or accidents. 1.6.2.5 [Deleted] 1.6.2.6 Primary Containment The design employs a pressure suppression primary containment which houses the reactor vessel, the reactor coolant recirculating loops, and other branch connections of the Reactor Primary System. The pressure suppression system consists of a drywell, a pressure suppression chamber which stores a large volume of water, connecting vents between the drywell and the pressure suppression chamber, isolation valves, containment cooling systems, and other service equipment. In the event of a process system piping failure within the drywell, reactor water and steam would be released into the drywell air space. The resulting increased drywell pressure would then force a mixture of air, drywell atmosphere, steam, and water through the vents into the pool of water in the pressure suppression chamber. The steam would condense in the pressure suppression pool, resulting in a rapid pressure reduction in the drywell. Air that was transferred to the pressure suppression chamber pressurizes the pressure suppression chamber, and is subsequently vented back to the drywell to equalize the pressure between the two vessels. Cooling systems are provided to remove heat from the reactor core, the drywell, and from the water in the pressure suppression chamber, and thus provide continuous cooling of the primary containment under accident conditions. 1.6-11

BFN-27 Appropriate isolation valves are actuated during this period to ensure containment of radioactive material, which might otherwise be released from the reactor containment during the course of the accident. 1.6.2.7 Primary Containment and Reactor Vessel Isolation Control System The Primary Containment and Reactor Vessel Isolation Control System automatically initiates closure of isolation valves to close off all potential leakage paths for radioactive material to the environs. This action is taken upon indication of a potential breach in the nuclear system process barrier. 1.6.2.8 Secondary Containment The secondary containment substructure consists of poured-in-place, reinforced concrete exterior walls that extend up to the refueling floor. The refueling room floor is also constructed of reinforced, poured-in-place concrete. The superstructure of the secondary containment above the refueling floor is a structural steel frame which supports metal roof decking, foamwall-stepped fascia panels, and insulated metal siding panels. The secondary containment structure completely encloses the primary containment drywells, fuel storage and handling facilities, and essentially all of the Core Standby Cooling Systems for the three units. During normal operation and when isolated, the secondary containment is maintained at a negative pressure relative to the building exterior. Excessive pressure differentials are relieved by blowout panels in the metal siding. 1.6.2.9 Main Steam Line Isolation Valves Although all pipelines that both penetrate the primary containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steam lines, because of their large size, are given special isolation consideration. Two automatic isolation valves, each powered by both air pressure and spring force, are provided in each main steam line. These valves fulfill the following objectives:

a. Prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting from either a major leak from the steam piping outside the primary containment or a malfunction of the pressure control system resulting in excessive steam flow from the reactor vessel, and

b. Limit the release of radioactive materials by closing the primary containment barrier in case of a major leak from the nuclear system inside the primary containment.

1.6-12

BFN-27 1.6.2.10 Main Steam Line Flow Restrictors A venturi-type flow restrictor is installed in each steam line close to the reactor vessel. These devices limit the loss of coolant from the reactor vessel before the main steam line isolation valves are closed in case of a main steam line break outside the primary containment. 1.6.2.11 Core Standby Cooling Systems A number of standby cooling systems are provided to prevent excessive fuel clad temperatures in the event of a breach in the nuclear system process barrier that results in a loss of reactor coolant. The four Core Standby Cooling Systems are:

1. High Pressure Coolant Injection System (HPCI),

2. Automatic Depressurization System,

3. Core Spray System, and

4. Low Pressure Coolant Injection System (an operating mode of the Residual Heat Removal System) (LPCI).

1.6.2.11.1 High Pressure Coolant Injection System The HPCI System provides and maintains an adequate coolant inventory inside the reactor vessel to prevent fuel clad melting as a result of postulated small breaks in the nuclear system process barrier. A high-pressure system is needed for such breaks because the reactor vessel depressurizes slowly, preventing low-pressure systems from injecting coolant. The HPCI includes a turbine-pump powered by reactor steam. The system is designed to accomplish its function on a short-term basis without reliance on plant auxiliary power supplies other than the DC power supply. 1.6.2.11.2 Automatic Depressurization System The Automatic Depressurization System acts to rapidly reduce reactor vessel pressure in a loss-of-coolant accident situation in which the HPCI fails to automatically maintain reactor vessel water level. The depressurization provided by the system enables the low pressure standby cooling systems to deliver cooling water to the reactor vessel. The Automatic Depressurization System uses some of the main steam relief valves which are part of the nuclear system pressure relief system. The automatic main steam relief valves are arranged to open upon conditions indicating both that a break in the nuclear system process barrier has occurred and that the HPCI System is not delivering sufficient cooling water to the 1.6-13

BFN-27 reactor vessel to maintain the water level above a preselected value. The Automatic Depressurization System will not be automatically activated unless either the core spray or LPCI system is operating. 1.6.2.11.3 Core Spray System The Core Spray System consists of two independent pump loops that deliver cooling water to spray spargers over the core. The system is actuated by conditions indicating that a breach exists in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water onto the core and preventing excessive fuel clad temperatures following a loss-of-coolant accident. 1.6.2.11.4 Low Pressure Coolant Injection Low Pressure Coolant Injection is an operating mode of the Residual Heat Removal System (RHR) but is discussed here because the LPCI mode acts as an engineered safeguard in conjunction with the other standby cooling systems. LPCI uses the pump loops of the RHR to inject cooling water at low pressure into the reactor recirculation loops. LPCI is actuated by conditions indicating a breach in the nuclear system process barrier, but water is delivered to the core only after reactor vessel pressure is reduced. LPCI operation, together with the core shroud and jet pump arrangement, provides the capability of core reflooding, following a loss-of-coolant accident, in time to prevent excessive fuel clad temperatures. 1.6.2.12 Residual Heat Removal System (Containment Cooling) The containment cooling subsystem is placed in operation to limit the temperature of the water in the pressure suppression pool following a design basis loss-of-coolant accident. In the containment cooling mode of operation, the RHR main system pumps take suction from the pressure suppression pool and pump the water through the RHR heat exchangers, where cooling takes place by transferring heat to the RHR service water system. The fluid is then discharged back to the pressure suppression pool. Another portion of the RHR is provided to spray water into the primary containment as an augmented means of removing energy from the containment following a loss-of-coolant accident. This capability is placed into service as required by manual operator action. 1.6.2.13 Control Rod Velocity Limiter A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its control rod drive. 1.6-14

BFN-27 The rate of reactivity insertion resulting from a rod drop accident is limited by this action. The limiters are passive components. 1.6.2.14 Control Rod Drive Housing Supports Control rod drive housing supports are located underneath the reactor vessel near the control rod housings. The supports limit the travel of a control rod in the event that a control rod housing is ruptured. The supports prevent a nuclear excursion as a result of a housing failure, thus protecting the fuel barrier. 1.6.2.15 Standby Gas Treatment System (SGTS) The system provides a means of removing radioactive material from the secondary containment by filtration and exhausting to the atmosphere through the plant stack in the event of accidental release. Three trains, any 2 of which can provide 100 percent design flow, consisting of a moisture separator, heater, particulate and charcoal filters, and blower, are provided. The results of laboratory carbon sample analysis shall show 90 percent radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989. The blowers are powered from independent, safety-related power supplies. The SGTS is a Class I system. 1.6.2.16 Standby AC Power Supply The standby AC power supply consists of eight diesel generator sets. The diesel generators are sized so that they can supply all necessary power requirements for one unit under design basis accident conditions, plus necessary loads for safe shutdown of the other two units. The diesel generators are specified to start up and reach rated speed within ten seconds. The diesel generator system is arranged with eight independent 4160-V load buses, each connected to one diesel generator. 1.6.2.17 DC Power Supply Eleven 250-V batteries, associated chargers, and distribution systems (3 unit batteries, 3 station batteries, and 5 batteries supplying control power for the 4160-V and 480-V shutdown boards) are provided for the plant. The various safety-related loads derive normal power from the batteries or their associated battery charger through distribution boards. 1.6.2.18 RHR Service Water System The RHR Service Water System is a Class I system that consists of four pairs of pumps located on the intake structure for pumping raw river water to the heat 1.6-15

BFN-27 exchangers in the RHR System and four additional pumps for supplying water to the Emergency Equipment Cooling Water System. 1.6.2.19 Emergency Equipment Cooling Water System This Class I system distributes cooling water supplied by the RHR Service Water System to essential equipment during normal and accident conditions. 1.6.2.20 Deleted 1.6.2.21 Reactor Building Ventilation Radiation Monitoring System The Reactor Building Ventilation Radiation Monitoring System consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the Reactor Building. Upon detection of high radiation, the Reactor Building is automatically isolated and the Standby Gas Treatment System is started. 1.6.3 Special Safety Systems 1.6.3.1 Standby Liquid Control System Although not intended to provide prompt reactor shutdown, the Standby Liquid Control System provides a redundant, independent, and different way from the control rods to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition (Mode 4). The SLC system is also required to supply sodium pentaborate solution for post-LOCA events that involve fuel damage to maintain the suppression pool pH at or above 7.0 for 30 days. The sodium pentaborate solution is credited as a buffering agent to offset the post-LOCA production of acids. 1.6.3.2 Plant Equipment Outside the Control Room Sufficient local controls are provided to allow the plant to be shut down from outside the control room. The plant design does not preclude bringing the plant to the cold shutdown condition (Mode 4) from outside the control room. 1.6-16

BFN-30 1.6.4 Process Control and Instrumentation 1.6.4.1 Nuclear System Process Control and Instrumentation 1.6.4.1.1 Reactor Manual Control System The Reactor Manual Control System provides the means by which control rods are manipulated from the control room for gross power control. Only one control rod can be manipulated at a time. The Reactor Manual Control System includes the controls that restrict control rod movement (rod block) under certain conditions as a backup to procedural controls. 1.6.4.1.2 Recirculation Flow Control System The Recirculation Flow Control System controls the speed of the reactor recirculation pumps. Adjusting the pump speed changes the coolant flow rate through the core. This effects changes in core power level. 1.6.4.1.3 Neutron Monitoring System The Neutron Monitoring System is a system of incore neutron detectors and out of core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level, for the entire range of flux conditions that may exist in the core. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux level indications during reactor startup and low power operation. The local power range monitors (LPRM) and average power range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. Rod block monitors (RBM) are provided to prevent rod withdrawal when reactor power should not be increased at the existing reactor conditions. The Traversing Incore Probe System (TIPS) provides a means for calibrating the LPRM portion of the neutron monitoring sensors. 1.6.4.1.4 Refueling Interlocks A system of interlocks that restricts the movements of refueling equipment and control rods when the reactor is in the refuel mode (Mode 5) is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling bridge, the refueling bridge hoists, the fuel grapple, control rods, and the service/inspection platform hoist when the hoist is capable of handling fuel. 1.6-17

BFN-27 1.6.4.1.5 Reactor Vessel Instrumentation In addition to instrumentation provided for the Nuclear Safety Systems and engineered safeguards, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. The instrumentation provided monitors reactor vessel pressure, water level, surface temperature, internal differential pressures and coolant flow rates, and top head flange leakage. 1.6.4.1.6 Process Computer System An online process computer is provided to monitor and log process variables, and to make certain analytical computations. The rodworth minimizer function of the computer prevents rod withdrawal/insertion under low power conditions, if the rod to be withdrawn/inserted is not in accordance with a preplanned pattern. The effect of the rod block is to limit the reactivity worth of the control rods by enforcing adherence to the preplanned rod pattern during startup or shutdown. 1.6.4.2 Power Conversion Systems Process Control and Instrumentation 1.6.4.2.1 Pressure Regulator and Turbine Generator Control The pressure regulation function of the turbine control system maintains control of turbine control valves to regulate pressure at the turbine inlet and therefore the pressure of the entire nuclear system. The turbine control system is an electrohydraulic control (EHC) system with an integral pressure regulation function. When not in pressure control mode, the EHC system maintains a fixed load or speed of the turbine. In addition, the EHC system provides overspeed protection for large load rejections. 1.6.4.2.2 Feedwater Control System The three element controller is used to regulate the feedwater system so that proper water level is maintained in the reactor vessel. The controller uses main steam flow rate, reactor vessel water level, and feedwater flow rate signals. The feedwater control signal is used to control the speed of the steam turbine driven feedwater pumps. 1.6.4.3 Electrical Power Systems Process Control and Instrumentation Each generator neutral is grounded through a distribution transformer and a secondary loading resistor. Each generator is equipped with a shaftdriven alternator exciter, an exciter field circuit breaker, rectifiers, and voltage regulating equipment. Current transformers are provided on the generator main and neutral terminals for relaying and metering. 1.6-18

BFN-27 Highspeed relays provide protection for the generator stator windings against faults. Incoming power is received from the 500-kV and 161-kV systems. The TVA 161-kV network receives power via the 161-kV switchyard. Two 161-kV lines terminate at separate buses which are connected by a circuit breaker. Two common station service transformers are energized from these buses. Normally, the switchyard will be operated with the breaker closed and both transformers energized. Disconnect switches are provided to permit either incoming line to be isolated from the switchyard and both transformers supplied from the remaining line. Output from the generators is fed into the TVA system by seven 500-kV lines via the 500-kV switchyard. The switchyard has a main and transfer zigzag bus arrangement. The two main bus sections are physically separated, and the transfer bus sections are separated from the main bus section by sectionalizing disconnect switches. Normally, the main and transfer bus sections are tied together through their respective disconnect switches. 1.6.4.4 Radiation Monitoring and Control 1.6.4.4.1 Process Radiation Monitoring Radiation monitors are provided on various lines to monitor either for radioactive materials released to the environs via process liquids and gases or for process system malfunctions. The following monitors are provided: Main Stack Radiation Monitors, Air Ejector Offgas Radiation Monitor, Raw Cooling Water System Discharge Radiation Monitor, Reactor Building Closed Cooling Water System Radiation Monitor, Liquid Radwaste System Radiation Monitor, RHR Service Water System Radiation Monitors, and Plant Ventilation Exhaust Radiation Monitors. 1.6.4.4.2 Area Radiation Monitors A number of radiation monitors are provided to monitor for abnormal radiation at various locations in the Reactor Building, Turbine Building, and Radwaste Building. These monitors annunciate alarms when abnormal radiation levels are detected. 1.6-19

BFN-27 1.6.4.4.3 Site Environs Radiation Monitors Radiation monitoring stations are provided to monitor the effects from natural and plant radiation sources. The stations employ appropriate devices to collect samples as well as measure direct radiation effects which can be used to determine changes in environmental radioactivity levels. 1.6.4.4.4 Liquid Radwaste System Control Liquid wastes to be discharged are handled on a batch basis, with protection against accidental discharge provided by procedural controls. Instrumentation with alarms to detect abnormal concentration and terminate release of liquid waste is provided. 1.6.4.4.5 Solid Radwaste Control The Solid Radwaste System collects, processes, stores, and prepares solid radioactive waste materials for offsite shipment. Wastes are handled on a batch basis, and radiation levels of the various batches are determined by the operating personnel. 1.6.4.4.6 Gaseous Radwaste System Control The Gaseous Radwaste System is continuously monitored by a radiation monitor located downstream of the recombiner system water separator, a monitor located downstream of the charcoal/absorbers but upstream of the afterfilters, and the main stack radiation monitor. Each of these monitors alarms on high radiation level. In addition, a high level signal from the monitor downstream of the air ejectors automatically isolates the Gaseous Radwaste System by closing a valve in the line between the after-filters and the stack. This action causes an increase in condenser back pressure. Hydrogen concentration in the gas downstream of the recombiners is continuously monitored. Although an explosion is not likely, temperature and pressure instrumentation in the line upstream of the decay pipe, in response to an explosion, causes valves downstream of the air ejectors to automatically isolate. These actions stop the supply of hydrogencontaining gas, and minimize release of radioactivity from a damaged filter. A main steamline high radiation condition will automatically close a valve between the main condensers and the mechanical vacuum pump. In addition, the mechanical vacuum pump is stopped. 1.6-20

BFN-27 1.6.5 Auxiliary Systems 1.6.5.1 Normal Auxiliary AC Power System The normal power source for unit auxiliaries is the 20.7- to 4.16-kV unit station service transformers. This source is connected to each unit generator's output leads. The startup power source for unit auxiliaries is the 500 kV system, with backup from the 161 kV switchyard through the common station service transformers. 1.6.5.2 Reactor Building Closed Cooling Water System The Reactor Building Closed Cooling Water System (RBCCWS) provides cooling water to designated auxiliary plant equipment located in the primary and secondary containments. The cooling water is available to the nuclear system auxiliaries under normal and accident conditions. 1.6.5.3 Raw Water Systems The Raw Cooling Water System is provided to remove heat from turbine associated equipment and accessories located in and adjacent to the Turbine Building, from the Reactor Building Closed Cooling Water System heat exchangers, and from other reactor associated equipment. The Raw Cooling Water System pumps are located in the Turbine Building and are supplied with river water from the condenser circulating water conduits. Three pumps are provided for each unit, with one spare provided for Units 1 and 2 and two spares for Unit 3. A Raw Service Water System, consisting of four pumps, supplies river water from the condenser circulating water conduits for yard watering, cooling for miscellaneous plant equipment requiring small quantities of cooling water, washdown services in unlimited access areas, and provides a means of pressurizing the raw water Fire Protection System. The Raw Service Water System also serves as a charging source for the RHR Service Water and Emergency Equipment Cooling Water Systems. 1.6.5.4 Fire Protection Systems A high pressure, raw water Fire Protection System provides water for fixed water spray, water sprinkler, aqueous film forming foam, and water fog systems, and to fire hoses and hydrants located throughout plant buildings and the surrounding yard. Fixed CO2, halon, and portable fire extinguishers furnish protection for hazards where use of water is not desirable. Fire detection, annunciation, and initiation systems are installed in selected areas of the Reactor Building, Control Building, 1.6-21

BFN-27 intake pumping station, cable tunnel to intake pumping station, Diesel Generator Buildings, and Turbine Building. 1.6.5.5 Heating, Ventilating, and Air Conditioning Systems Heating, Ventilating, and Air Conditioning Systems are provided for the Reactor Building, Turbine Building, Radwaste Building, and Control Building. The design of these systems varies; but in all cases, they maintain the indoor environment necessary for equipment protection and personnel comfort. In areas where significant airborne activity is expected, these systems limit the spread of contamination and filter the exhaust air before discharge. 1.6.5.6 New and Spent Fuel Storage A dry vault in the Reactor Building is provided for storage of new fuel. The new fuel is normally transferred directly to the spent fuel storage pool upon receipt. Fuel transfer during refueling is conducted underwater. Irradiated (spent) fuel is stored underwater in the Reactor Building until prepared for shipment from the site. 1.6.5.7 Fuel Pool Cooling and Cleanup System A Fuel Pool Cooling and Cleanup System is provided to remove decay heat from spent fuel stored in the fuel pool and to maintain a specified water temperature, purity, clarity, and level. 1.6.5.8 Control and Service Air Systems Clean, dry, control air is provided to pneumatically operated instruments and controls throughout the plant and yard. Each reactor unit has a drywell control air system that provides control air for the equipment inside its drywell. Service air outlets are provided throughout the plant. 1.6.5.9 Demineralized Water System A makeup demineralized water unit is used to furnish a supply of high purity water for makeup of the primary coolant systems, the Reactor Building Closed Cooling Water Systems, the pressure suppression chambers, and the Standby Liquid Control Systems. The water is also used for radioactive decontamination work and preoperational cleaning of reactor and piping systems. 1.6.5.10 Potable Water and Sanitary Systems These systems provide potable water from a nearby municipal water system for use in the plant plumbing systems and sewage treatment in a 65,000 gallon per day biological treatment system. 1.6-22

BFN-27 1.6.5.11 Equipment and Floor Drainage System Radioactive drainage from equipment leaks and from areas which may contain radioactive materials is collected and routed to shielded sumps. This waste is then pumped to drain collection tanks in the Radwaste Building, where it is treated and returned for reuse in the plant or discharged to the river. Nonradioactive drainage is collected in drain sumps and discharged to the condenser circulating water discharge tunnels. 1.6.5.12 Process Sampling Systems These systems provide samples of process liquids and gases to obtain data from which the performance of the plant, items of equipment, and systems may be determined. Sampling is continuous or periodic as appropriate. These systems will function at all times and under all operating conditions. 1.6.5.13 Communications Systems An extensive, private telephone system, along with a paging system, sound powered telephone systems, and closed circuit television systems, provides complete communications throughout the plant. 1.6.6 Shielding Plant shielding allows personnel access to the plant to perform maintenance and carry out operational duties, with personnel exposures limited to the values given in Table 12.3-1. 1.6.7 Implementation of Loading Criteria When correctly installed in a suitable facility, structures, and equipment are designed to substantially resist mechanical damage due to loads produced by mechanical and thermal forces. For the purpose of categorizing mechanical strength designs for these loads, the following definitions are established.

a. Class I This class includes those structures, equipment, and components whose failure or malfunction might cause, or increase the severity of, an accident 1.6-23

BFN-27 which would endanger the public health and safety. This category includes those structures, equipment, and components required for safe shutdown and isolation of the reactor.

b. Class II This class includes those structures, equipment, and components which are important to reactor operation, but are not essential for preventing an accident which would endanger the public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade the integrity of any item designated Class I.

The loading conditions may be divided into four categories: (1) normal, (2) upset, (3) emergency, and (4) faulted conditions. These categories are generically described, and their meaning is expanded in quantitative, probabilistic language in Appendix C. The purpose of this expansion is to clarify the classification of any hypothesized accident or sequence of loading events. Event probability is used to establish meaningful and adequate safety factors for structural design so that the appropriate structural safety margins are applied. 1.6-24

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-1

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-2

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-3, Sheet 1

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-3, Sheet 2

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-4

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-5

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-6

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-7

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-8, Sheet 1

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-8, Sheet 2

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-8, Sheet 3

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-9

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-10

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-11

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-12

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-13

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-14

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-15

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-16

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-17

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-18

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-19

         

        

  

         

        

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SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-23

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-24

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-25

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-26

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 1.6-27

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BFN-17 1.7 COMPARISON OF PRINCIPAL DESIGN CHARACTERISTICS This section highlights the principal design features of the plant and provides a comparison of the major features with other boiling water reactor facilities. The design of this facility is based upon proven technology attained during the development, design, construction and operation of boiling water reactors of similar or identical types. The data, performance characteristics, and other information presented herein are historical and represent the plant as originally designed. The parameter values presented in Tables 1.7-1 to 1.7-5 for the various nuclear plants are the values used in the design of these plants. Since the various owner-utilities were not contacted, no guarantee is given that these parameter values are current. The information contained in this section is, therefore, maintained for historical purpose only. More updated information can be found in the specific chapters dealing with specific topics. 1.7.1 Nuclear System Design Characteristics Table 1.7-1 summarizes the design and operating characteristics for Browns Ferry Nuclear Plant. The same characteristics are presented for the nuclear system of Duane Arnold Energy Center, Cooper Nuclear Station, Vermont Yankee Nuclear Power Station, and Hatch Nuclear Plant Unit 1. 1.7.2 Power Conversion Systems Design Characteristics Table 1.7-2 presents a summary of the power conversion systems design characteristics for the plant and compares these with Duane Arnold Energy Center, Cooper Nuclear Station, Vermont Yankee Nuclear Power Station, and Hatch Nuclear Plant Unit 1. 1.7.3 Electrical Power Systems Design Characteristics Table 1.7-3 is a summary and comparison of the electrical power systems design characteristics of the plant and the same four similar facilities. 1.7.4 Containment Design Characteristics Table 1.7-4 summarizes the design characteristics for the primary and secondary containments of the Browns Ferry Nuclear Plant. Design characteristics are also presented for the primary and secondary containment systems employed for Hatch Unit 1, Vermont Yankee Nuclear Power Station, Cooper Station, and Duane Arnold 1.7-1

BFN-17 Energy Center. In addition, data is given for the type, construction, and height of the elevated release point for the above plants. 1.7.5 Structural Design Characteristics Table 1.7-5 is a summary and comparison of the seismic and wind design factors considered in the structural design of Browns Ferry Nuclear Plant and the above similar plants. 1.7.6 Discussion of Core Design Improvement Numerous improvements have been made to the core design of Browns Ferry subsequent to receipt of the operating license for each of the three units. A general description of reload fuel designs presently used in Browns Ferry is given in Chapter 3. The specific fuel types loaded in each unit along with analytical results of the cycle-specific reload core design and licensing analyses are given in the applicable Supplemental Reload Licensing Report (SRLR). The current SRLR for each BFN unit is included in Appendix N of the FSAR. 1.7-2

BFN-17 TABLE 1.7-1 (Sheet 1) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD THERMAL AND HYDRAULIC DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Rated Power, MWt 3293 2436 1593 2381 1593 Design Power, MWt 3440 2537 1665 2500 1670 Steam Flow Rate, lb/hr 13.37 x 106 10.03 x 106 6.43 x 106 9.81 x 106 6.847 x 106 Core Coolant Flow Rate, lb/hr 102.5 x 106 75.5 x 106 48.5 x 106 74.5 x 106 48.5 x 106 Feedwater Flow Rate, lb/hr 13.315 x 106 10.445 x 106 6.43 x 106 9.81 x 106 6.77 x 106 Feedwater Temperature, F 378.4 387.4 372 367 420 System Pressure, Nominal in Steam Dome, psia 1020 1020 1020 1020 1020 Average Power Density, kW/liter 49.69/49.46/ 51.2 50.8 51.2 50.9 49.2 Maximum Thermal Output, kW/ft 18.5 (7x7)/13.4 18.3 18.37 18.5 18.5 (8x8) Average Thermal Output, kW/ft 7.050 (7x7)/ 7.114 7.1 7.079 7.079 5.59 (8x8) Average Heat Flux, Btu/hr-ft2 148937/142007/ 164,734 163,900 164,500 163,933 143635 Maximum UO2 Temperature, F 4430 4430 4430 4430 4430 Average Volumetric Fuel Temperature, F 1210 1210 1210 1210 1210 Average Fuel Rod Surface Temperature, F 560 560 560 560 560 Minimum Critical Power Ratio (MCPR)(1) >1.07 >1.9 >1.9 >1.9 >1.9 Coolant Enthalpy at Core Inlet, Btu/lb 521.3 526.2 522.9 520.1 525.6 Core Maximum Exit Voids Within Assemblies 79 79 79 79 79 Core Average Exit Quality, % Steam 13.2 13.9 13.6 13.2 14.3

BFN-17 TABLE 1.7-1 (Sheet 2) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD THERMAL AND HYDRAULIC DESIGN (Cont-d) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Design Power Peaking Factors Transverse Peaking Factor 1.4 1.4 1.4 1.4 1.405 Local Peaking Factor 1.24 1.24 1.24 1.24 1.24 Axial Peaking Factor 1.5 1.5 1.5 1.5 1.5 Total Peaking Factor 2.63 2.6 2.6 2.6 2.6 NUCLEAR DESIGN (First Core) Water/UO2 Volume Ratio (Cold) 2.43 Type I 2.41 2.41 2.41 2.41 2.53 Type II & III Reactivity with Strongest Control Rod <0.99 <0.99 <0.99 <0.99 <0.99 Out, keff Moderator Temperature Coefficient At 68F, k/k - F Water -3.5 x 10-5 -3.5 x 10-5 -5.0 x 10-5 -3.5 x 10-5 -3.5 x 10-5 Hot, no voids, k/k - F Water -11.6 x 10-5 -11.6 x 10-5 -17.0 x 10-5 -11.6 x 10-5 -11.6 x 10-5 Moderator Void Coefficient Hot, no voids, k/k - % Void -8.7 x 10-4 -8.7 x 10-4 -1.0 x 10-3 -8.7 x 10-4 -8.7 x 10-4 At Rated Output, k/k - % Void -1.05 x 10-3 -1.05 x 10-3 -1.5 x 10-3 -1.05 x 10-3 -1.05 x 10-3 Fuel Temperature Doppler Coefficient At 68F, k/k - F Fuel -0.9 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 Hot, No Void, k/k - F Fuel -1.0 x 10-5 -1.2 x 10-5 -1.2 x 10-5 -1.2 x 10-5 -1.2 x 10-5 At Rated Output, k/k - F Fuel -0.9 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.3 x 10-5 Initial Average U-235 Enrichment, W/O 2.19% 2.30% 2.50% 2.15% 2.25% Fuel Average Discharge Exposure, MWD/Ton 19,000 19,000 19,000 19,000 18,350 (6) Nuclear Design (Reload Core) See applicable Nuclear Design Reports.

BFN-17 TABLE 1.7-1 (Sheet 3) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Fuel Assembly Number of Fuel Assemblies 764 560 368 548 368 Fuel Rod Array 7 x 7 or 8 x 8 7x7 7x7 7x7 7x7 Overall Dimensions, inches 175.98 175.98 175.98 175.98 175.98 Weight of UO2 per Assembly, pounds See applicable Undished - Undished - 487.4 Undished - Nuclear Design 490.35 487.4 490.35 (6) Reports Dished - Dished - 483.42 483.42 Weight of Fuel Assembly, pounds 681 Undished - Undished - 682 Undished - 681.48 682 681.48 Dished - Dished - 674.55 674.55 Fuel Rods Number per Fuel Assembly 49 or 64* 49 49 49 49 (mixed cores) 1.483 Outside Diameter, inch 0.563 0.563 0.563 0.563 0.563 Clad Thickness, inch 0.032 0.032 0.032 0.032 0.032 Gap - Pellet to Clad, inch 0.006/0.009 0.006 0.006 0.006 0.006 Length of Gas Plenum, inches 16/9.48 16 16 16 16 Clad Material Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Cladding Process Free standing Free standing Free Standing Free Standing Free Standing loaded tubes loaded tubes loaded tubes loaded tubes loaded tubes

• Two different 8 x 8 fuel bundle arrangements are used. One uses 63 fuel rods and 1 water rod; the other uses 62 fuel rods and 2 water rods.

BFN-17 TABLE 1.7-1 (Sheet 4) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Fuel Pellets Material Uranium Dioxide Uranium Dioxide Uranium Dioxide Uranium Dioxide Uranium Dioxide Density, % of theoretical 94% 93% 93% 93% 93% Diameter, inch 0.410 0.487 0.487 0.487 0.487 Length, inch 0.410 0.75 0.75 0.75 0.75 Fuel Channel Overall Dimension, inches (length) 166.906 166.906 166.906 166.096 166.906 Thickness, inch 0.080 0.080 0.080 0.080 0.080 Cross-Section Dimensions, inches 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 5.438 x 5.438 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core Assembly Fuel Weight as UO2, pounds 361,837 272,849 179,370 267,095 179,298 Zirconium Weight, pounds 140,397 96,370 63,300 94,305 63,300 (Zr.2 + Zr.4 Spacers) Core Diameter (equivalent), inches 187.1 160.2 129.9 158.5 129.9 Core Height (Active Fuel), inches 144 - 150 144 144 144 144 Reactor Control System Method of Variation of Reactor Power Movable Control Movable Control Movable Control Movable Control Movable Control Rods and Variable Rods and Variable Rods and Variable Rods and Variable Rods and Variable Coolant Pumping Coolant Pumping Coolant Pumping Coolant Pumping Coolant Pumping Number of Movable Control Rods 185 137 89 137 89 Shape of Movable Control Rods Cruciform Cruciform Cruciform Cruciform Cruciform Pitch of Movable Control Rods 12.0 12.0 12.0 12.0 12.0

BFN-17 TABLE 1.7-1 (Sheet 5) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE MECHANICAL DESIGN (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Reactor Control System (Cont'd) Control Material in Movable Rods B4C granules B4C granules B4C granules B4C granules B4C granules Compacted Compacted Compacted Compacted Compacted in SS Tubes in SS Tubes in SS Tubes in SS Tubes in SS Tubes Type of Control Rod Drives Bottom Entry, Bottom Entry, Bottom Entry, Bottom Entry, Bottom Entry, Locking Piston Locking Piston Locking Piston Locking Piston Locking Piston Supplementary Reactivity Control Grandolinia 156 Burnable Poison Flat, boron-stainless steel control curtains In-Core Neutron Instrumentation Number of In-Core Neutron Detectors (Fixed) 172 124 80 124 80 Number of In-Core Detector Assemblies 43 31 20 31 20 Number of Detectors Per Assembly 4 4 4 4 4 Number of Flux Mapping Neutron Detectors 5 4 3 4 3 Range (and Number) of Detectors Source Range Monitor Source to Source to Source to Source to Source to 0.001% power 0.001% power 0.001% power 0.001% power 0.001% power (4) (4) (4) (4) (4) Intermediate Range Monitor 0.0001% to 10% 0.0001% to 10% 0.0001% to 10% 0.0001% to 10% 0.0001% to 10% power (8) power (8) power (8) power (8) power (8) Local Power Range Monitor 5% to 125% 5% to 125% 5% to 125% 5% to 125% 5% to 125% power (172) power (124) power (80) power (124) power (80) Average Power Range Monitor 2.5% to 125% 2.5% to 125% 2.5% to 125% 2.5% to 125% 2.5% to 125% power (U1-6; U2-4; power (6) power (6) power (6) power (6) U3-6)

BFN-17 TABLE 1.7-1 (Sheet 6) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD REACTOR VESSEL DESIGN UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER Material Carbon Steel/Clad Stainless Steel (ASME SA-336 & SA-302B) Design pressure, psia 1265 1265 1265 1265 1265 Design Temperature, F 575 575 575 575 575 Inside Diameter ft-in. 20 - 11 18 - 2 17 - 2 18 - 2 15 - 3 Inside Height, ft-in. 73 1/2 69 - 4 63 - 1.5 69 - 4 66 - 4 Side Thickness (including clad) 6.313 5.531 5.187 5.531 5.625 Minimum Clad Thickness, inches 1/8 1/8 1/8 1/8 1/8 REACTOR COOLANT RECIRCULATION DESIGN Number of Recirculation Loops 2 2 2 2 2 Design Pressure Inlet Leg. psig 1148 1148 1175 1148 1148 Outlet Leg. psig 1326 1274 1274 1274 1268 CORE MECHANICAL DESIGN Design Temperature, F 562 562 562 562 562 Pipe Diameter Max. inches 28 28 28 28 22 Pipe Material 304/316 304/316 304/316 304/316 304/316 Recirculation Pump flow Rate, GPM 45,200 45,200 32,500 45,200 27,100 Number of Jet Pumps in Reactor 20 20 20 20 16 MAIN STEAM LINES Number of Steam Lines 4 4 4 4 4 Design Pressure, psig 1146 1146 1146 1146 1146 Design Temperature, F 563 563 563 563 563 Pipe Diameter, inches 26 24 20 24 20 Pipe Material Carbon Steel (ASTM A155 KC70 or ASTM A106 Grade B)

BFN-17 TABLE 1.7-1 (Sheet 7) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD CORE STANDBY COOLING SYSTEMS UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER (These systems are sized on design power) Core Spray System Number of Loops 2 2 2 2 2 Flow Rate (gpm) 6250 at 4625 at 3000 at 4500 at 3020 at 105 psid 120 psid 136 psid 115 psid 127 psid High Pressure Coolant Injection system (No.) 1 1 1 1 1 Number of Loops 1 1 1 1 1 Flow Rate (gpm) 5000 4250 4250 4220 2980 Automatic Depressurization system (No.) 1 1 1 1 1 Low Pressure Coolant Injection (No.) 1 1 1 1 1 Number of Pumps 4 4 4 4 4 Flow Rate (gpm/pump) 10,800 gpm 7700 at 4800 at 7000 at 4800 at (1 pump per loop) 20 psid 20 psid 20 psid 20 psid 20,000 gpm (2 pumps per loop) AUXILIARY SYSTEMS Residual Heat Removal System Reactor Shutdown Cooling (number of pumps) 4 4 4 4 4 Flow Rate (gpm/pump)(2) 10,000 7,700 7,000 7,700 4,800 Capacity (Btu/hr/heat exchanger)(3) 70 x 106 32 x 106 57.5 x 106 70 x 106 35 x 106 Number of heat exchangers 4 2 2 2 2 Primary Containment Cooling Flow rate (gpm)(4) 32,000 30,800 28,000 30,800 19,200

BFN-17 TABLE 1.7-1 (Sheet 8) COMPARISON OF NUCLEAR SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) (Parameters are related to Rated Power Output for a single plant unless otherwise noted) BROWNS FERRY VERMONT COOPER DUANE ARNOLD AUXILIARY SYSTEMS (Cont'd) UNITS 1/2/3 HATCH UNIT 1 YANKEE STATION ENERGY CENTER RHR Service Water System Flow Rate (gpm/pump) 4,500 8,000 2,700 8,000 2,500 Number of pumps 12(5) 4 4 4 4 Reactor Core Isolation Cooling System Flow Rate (gpm) 616 at 400 at 400 416 at 416 1120 psid 1120 psid 1120 psid Fuel Pool Cooling and Cleanup system 6 6 6 6 6 Capacity (BTU/hr) 8.8 x 10 3.3 x 10 2.37 x 10 3.4 x 10 2.37 x 10 (1) The operating MCPR limits are subject to change from one cycle to the next and also from one part of the current cycle to the next. The appropriate value for MCPR may be obtained by consulting the applicable current Reload Licensing Amendment. (2) Capacity during reactor flooding mode with three of four pumps running. (3) Capacity during post-accident cooling mode with 165F shell side inlet temperature, maximum service water temperature, and 1 RHR pump and 1 RHR service water pump in operation. (4) The existing design requires 16,000 gpm (2 pumps, 1 loop) to ensure torus water temperature is maintained within acceptable limits for following all postulated events. (5) For all three units. (6) See Appendix N

BFN-17 TABLE 1.7-2 COMPARISON OF POWER CONVERSION SYSTEMS DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) Browns Ferry Duane Arnold TURBINE GENERATOR Each Unit Hatch Unit 1 Vermont Yankee Cooper Station Energy Center Design Power, MWt 3440 2537 1665 2487 1670 Design Power, MWe 1152 849 564 836 597 Generator Speed, rpm 1800 1800 1800 1800 1800 6 6 6 6 6 Design Steam Flow, lb/hr 14.035 x 10 10.48 x 10 6.423 x 10 10.049 x 10 6.696 x 10 Turbine Inlet Pressure, psia 965 970 950 970 950 TURBINE BYPASS SYSTEM Capacity, percent of turbine design steam flow 25 25 100 25 25 MAIN CONDENSER 6 6 6 6 6 Heat removal capacity, Btu/hr 7,770 x 10 5,800 x 10 3,500 x 10 5,367 x 10 3,681 x 10 CIRCULATING WATER SYSTEM Number of Pumps 3 3 3 4 2 or more Flow Rate gpm/pump 220,000 185,000 117,000 162,500 130,000 or less CONDENSATE AND FEEDWATER SYSTEMS 6 6 6 6 6 Design Flow Rate, lb/hr 13.845 x 10 10.096 x 10 6.4 x 10 9.773 x 10 7.146 x 10 Number Condensate Pumps 3 3 2 3 2 Number Condensate Booster Pumps 3 - --- Number Feedwater Pumps 3 2 2 2 2 Condensate Pump Drive AC power AC power AC power AC power AC power Condensate Booster Pump Drive AC power - - - - Feedwater Pump Drive Turbine Turbine AC power Turbine AC power

BFN-17 TABLE 1.7-3 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) BROWNS FERRY VERMONT COOPER DUANE ARNOLD TRANSMISSION SYSTEM NUCLEAR PLANT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Outgoing lines (number-rating) 7-500kV 2-230kV 2-345kV 4-345kV 2-345kV NORMAL AUXILIARY AC POWER Incoming lines (number-rating) 2-161kV 2-230kV 2-345kV 1-115kV 2-345kV 1-230kV 1-69kV 3-161kV 1-115kV 1-4160kV Auxiliary transformers 3 1 1 1 2 Startup transformers 2 2 1 2 1 STANDBY AC POWER SUPPLY Number diesel generators 8 3 2 4 2 Number of 4160V Shutdown buses 8 3 2 2 2 Number of 480V Shutdown buses 6 4-660V 3 3 3 DC POWER SUPPLY Number of 125V or 250V batteries* 6 2 2 2 2 Number of 125V or 250V buses* 6 4 4 4 2

• 3 of the 6 250V systems are qualified

BFN-17 TABLE 1.7-4 Sheet 1 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) BROWNS FERRY VERMONT COOPER Duane Arnold PRIMARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION Energy Center Type Pressure Pressure Pressure Pressure Pressure Suppression Suppression Suppression Suppression Suppression Construction Drywell Light bulb Light bulb Light bulb Light bulb Light bulb shape; steel shape; steel shape; steel shape; steel shape; steel vessel vessel vessel vessel vessel Pressure Suppression Chamber Torus; steel Torus; steel Torus; steel Torus; steel Torus; steel vessel vessel vessel vessel vessel Pressure Suppression Chamber 56 56 56 56 56 Internal Design Pressure (psig) Pressure Suppression chamber - 2 2 2 2 2 External Design Pressure (psig) Drywell-Internal Design Pressure (psig) 56 56 56 56 56 Drywell-External Design Pressure (psig) 2 2 2 2 2 Drywell Free Volume (ft3) 159,000 146,400 134,000 145,430 130,930 Pressure Suppression chamber 119,000 101,410 99,000 109,810 94,630 Free Volume (ft3), minimum Pressure Suppression Pool Water 128,700 86,660 78,000 87,660 61,500 Volume (ft3), maximum Submergence of Vent Pipe Below 4 4 4 4 4 Pressure Pool Surface (ft), nominal Design Temperature of Drywell (F) 281 281 281 281 281 Design Temperature of Pressure 281 281 281 281 281 Suppression Chamber (F)

• Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-4 (Cont'd) Sheet 2 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) BROWNS FERRY VERMONT COOPER DUANE ARNOLD PRIMARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Downcomer Vent Pressure Loss Factor 4.1 6.21 6.21 6.21 6.21 Break Area/total Vent Area 0.017 0.019 0.019 0.019 0.019 Calculated Maximum Pressure After Blow- 49.6 45 35 46 45 down Drywell (psig) Pressure Suppression chamber (psig) 27 28 22 28 29 Initial Pressure Suppression Pool 40 50 35 50 50 Temperature Rise (F) Leakage Rate (% Free Volume/Day 0.5 0.5 0.5 0.5 0.5 at 56 psig and 281F SECONDARY CONTAINMENT Type Controlled Leakage, Controlled Leakage, Controlled Leakage, Controlled Leakage, Controlled Leakage, Elevated Release Elevated Release Elevated Release Elevated Release Elevated Release Construction Reinforced Reinforced Reinforced Reinforced Reinforced Concrete Concrete Concrete Concrete Concrete Upper Levels Steel Super- Steel Super- Steel Super- Steel Super- Steel Super-structure and structure and structure and structure and structure and Siding Siding Siding Siding Siding Roof Steel Decking Steel Steel Steel Steel with Builtup Sheeting Sheeting Sheeting Sheeting Composition Roof Internal Design Pressure (psig) +7 to -5 in. H2O 0.25 0.25 0.25 0.25 Design Inleakage Rate (% Free 100 100 100 100 100 Volume/Day at 0.25 inches H2O)

• Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-4 (Cont'd) Sheet 3 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) BROWNS FERRY VERMONT COOPER Duane Arnold SECONDARY CONTAINMENT* EACH UNIT HATCH UNIT 1 YANKEE STATION Energy Center ELEVATED RELEASE POINT Type Stack Stack Stack Stack Stack Construction Reinforced Steel Steel Steel Steel Concrete Height (above ground) 600 feet 150 meters 318 feet 100 meters 100 meters

• Where applicable, containment parameters are based on design power.

BFN-17 TABLE 1.7-5 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Data in this table has not been updated to reflect the power uprate at Browns Ferry) BROWNS FERRY VERMONT COOPER DUANE ARNOLD SEISMIC DESIGN NUCLEAR PLANT HATCH UNIT 1 YANKEE STATION ENERGY CENTER Operating Basis Earthquake (horizontal g) 0.10 0.08 0.07 0.10 0.06 Design Basis Earthquake (horizontal g) 0.20 0.15 0.14 0.20 0.12 WIND DESIGN Maximum sustained (mph) 100 105 80 100 105 Tornadoes (mph) 300 300 300 300 300

BFN-21 1.8

SUMMARY

OF RADIATION EFFECTS 1.8.1 Normal Operation The gaseous and liquid radioactive waste systems are designed so that dose to any offsite person will not exceed that permitted within the limits specified in the Offsite Dose Calculation Manual (ODCM), applicable limits in the plant technical specifications, and technical requirements manual. The expectancy, based on operating experience, is that dose to any off-site person from gaseous waste discharge will not average more than a small fraction of the permissible dose, and that concentrations of liquid waste at the point of discharge will average less than 1% of the concentrations permitted by 10 CFR 20. Both effects are only a small fraction of the effect of natural background radiation. 1.8.2 Abnormal Operational Transients A design objective is to avoid fuel damage as a result of abnormal operational transients. Analyses of these events, which are described in the "Plant Safety Analysis", show that abnormal operational transients do not result in any significant increase of radioactive material release to the environs over that experienced during normal operation. 1.8.3 Accidents The ability of the plant to withstand the consequences of accidents without posing a hazard to the health and safety of the public is evaluated by analyzing a variety of postulated accidents. The calculated consequences of the design basis accidents, which result in the greatest potential off-site radiation exposures, are presented in Chapter 14. These doses are substantially below the guideline doses given in 10 CFR 50.67. 1.8-1

BFN-17 1.9 PLANT MANAGEMENT The summary of information originally presented in this section was superfluous and has been deleted. The Browns Ferry Nuclear Plant Organization is presented in Tennessee Valley Authority Topical Report, TVA-NPOD89-A Nuclear Power Organization Description. 1.9-1

BFN-16 1.10 QUALITY ASSURANCE PROGRAM Appendix D describes the comprehensive quality assurance plan that meets the requirements of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." 1.10-1

BFN-16 1.11 IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERN -

SUMMARY

1.11.1 General The information presented in this section and FSAR Appendix I had direct applicability to the licensing efforts expended by TVA at the time of the preparation and submittal of the FSAR. Section 1.11 and Appendix I are retained in the FSAR for historical and traceability reasons only. The design of the General Electric boiling water reactors for this station is based upon proven technological concepts developed during the development, design, and operation of numerous similar reactors. The AEC in reviewing the Browns Ferry docket at the Construction Permit stage identified several areas where further R&D efforts were required to more definitely assure safe operation of this station. Also, both the AEC Staff and the Advisory Committee for Reactor Safeguards had, in their review of this and more recent reactor projects, identified several additional technical areas for which further detailed support information had to be obtained. All of these development efforts thus were of three general types: (1) those which pertain to the broad category of water-cooled reactors; (2) those which pertain specifically to boiling water reactors; and (3) those which have been noted particularly for a facility during the construction or operating permit licensing activities by the AEC Staff and ACRS reviews. Appendix I of this FSAR provides a complete, comprehensive examination and discussion of each of these concern areas, indicating the resolution accomplished or planned at the time of FSAR preparation. A summary conclusion of this analysis is provided in this subsection by Tables 1.11-1 through 1.11-6. The concerns have been subdivided as follows:

a. Areas specified in the Browns Ferry AEC-ACRS Construction Permit Reports.

(See Table 1.11-2.)

b. Areas specified in the Browns Ferry AEC Staff Construction Permit Safety Evaluation Reports. (See Tables 1.11-3 and 1.11-4.)

c. Areas specified in other related AEC-ACRS construction permit and operating license reports. (See Table 1.11-5.)

d. Areas specified in other somewhat related AEC Staff construction permit and operating license safety evaluation reports. (See Table 1.11-6.)

e. Additional information available and supplied by General Electric on BWR.

(See Table 1.11-1.) 1.11-1

BFN-16 The scope of many of the areas of technology for items in a, b, and c above is discussed in Appendix I in detail and as part of an official response1 by General Electric to the various ACRS concern subjects. General Electric has submitted many topical reports to the AEC in support of this application and those of other facilities. Table 1.11-1 provides a list of all topical reports submitted to the AEC on behalf of TVA while trying to obtain an operating license for Browns Ferry. Topical reports submitted to the NRC (and the former AEC) after receipt of the operating license are not included in this list but can be found from a search of TVA-NRC correspondence. 1 Bray, A. P., et al., "The General Electric Company, Analytical and Experimental programs for Resolution of ACRS Safety Concerns," APED-5608, April 1968. 1.11-2

BFN-16 TABLE 1.11-1 (Sheet 1) BROWNS FERRY NUCLEAR PLANT TOPICAL REPORTS SUBMITTED TO THE AEC IN SUPPORT OF DOCKET GE Report No. Title

1. APED 5286 Design Basis for Critical Heat Flux in Boiling Water Reactors (September 1966)

2. APED 5446 Control Rod Velocity Limiter (March 1967)

3. APED 5449 Control Rod Worth Minimizer (March 1967)

4. APED 5450 Design Provisions for In Service Inspection (April 1967)

5. APED 5453 Vibration Analysis and Testing of Reactor Internals (April 1967)

6. APED 5555 Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RDB144A (November 1967)

7. TR67SL211 An Analysis of Turbine Missiles Resulting from Last Stage Wheel Failure (October 1967)

8. APED 5608 General Electric Company Analytical and Experimental Program for Resolution of ACRS Safety Concerns (April 1968) (Not Class I)

9. APED 5455 The Mechanical Effects of Reactivity Transients (January 1968)

10. APED 5528 Nuclear Excursion Technology (August 1967)

11. APED 5448 Analysis Methods of Hypothetical Super Prompt Critical Reactivity Transients in Large Power Reactors (April 1968)

12. APED 5458 Effectiveness of Core Standby Cooling Systems for General Electric Boiling Water Reactors (March 1968)

13. APED 5640 Xenon Considerations in Design of Large Boiling Water Reactors (June 1968)

14. APED 5454 Metal Water Reactions Effects on Core Cooling and Containment (March 1968)

15. APED 5460 Design and Performance of General Electric Boiling Water Reactor Jet Pumps (September 1968)

16. APED 5654 Considerations Pertaining to Containment Inerting (August 1968)

17. APED 5696 Tornado Protection for the Spent Fuel Storage Pool (November 1968)

18. APED 5706 In Core Neutron Monitoring System for General Electric Boiling Water Reactors, Rev. 1 (April 1969)

19. APED 5703 Design and Analysis of Control Rod Drive Reactor Vessel Penetrations (November 1968)

20. SPED 5698 Summary of Results Obtained From a Typical Startup and Power Test Program for a General Electric Boiling Water Reactor (February 1969)

21. APED 5750 Design and Performance of General Electric Boiling Water Reactor Main Steam Line Isolation Valves (March 1969)

22. APED 5756 Analytical Methods for Evaluating the Radiological Aspects of the General Electric Boiling Water Reactor (March 1969)

23. APED 5652 Stability and Dynamic Performance of the General Electric Boiling Water Reactor (April 1969)

BFN-16 TABLE 1.11-1 (Cont'd) (Sheet 2) BROWNS FERRY NUCLEAR PLANT TOPICAL REPORTS SUBMITTED TO THE AEC IN SUPPORT OF DOCKET GE Report No. Title

24. APED 5736 Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards (April 1969)

25. APED 5447 Depressurization Performance of the General Electric Boiling Water Reactor High Pressure Coolant Injection System (June 1969)

26. NEDO 10017 Field Testing Requirements for Fuel, Curtains and Control Rods (June 1969)

27. NEDO 10029 An Analytical Study on Battle Fracture of GE-BWR Vessel Subject to the Design Basis Accident (July 1969)

28. NEDO 10045 Consequences of a Steam Line Break for a General Electric Boiling Water Reactor (October 1969)

29. NEDO 10173 Current State of Knowledge High Performance BWR Zircaloy-Clad UO Fuel (May 2

1970)

30. NEDO 10139 Compliance of Protection Systems to Industry Criteria: General Electric BWR Nuclear Steam Supply System (June 1970)

31. NEDO 10179 Effects of Cladding Temperature and Material on ECCS Performance (June 1970)

32. NEDO 10174 Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor (May 1970)

33. NEDO 10189 An Analysis of Functional Common-Mode Failures in GE BWR Protection and Control Instrumentation (July 1970)

34. NEDO 10208 Effects of Fuel Rod Failure on ECCS Performance (August 1970)

35. NEDO 10320 The General Electric Pressure Suppression Containment Analytical Model (April 1971)

36. NEDO 10320 The General Electric Pressure Suppression Containment Analytical Supplement 1 Model(May 1971)

37. NEDO 10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors (April 1971)

38. NEDO 10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for Supplement 1 General Electric Boiling Water Reactors (April 1971)

39. NEDO 10349 Analysis of Anticipated Transients Without Scram (March 1971)

BFN-16 TABLE 1.11-2 (Sheet 1) BROWNS FERRY NUCLEAR PLANT AEC-ACRS CONCERNS - RESOLUTIONS Identification Section No. AEC-ACRS Concern Browns Ferry Resolutions 1.2.2 Effects of Fuel Failure on CSCS Topical Report (GE-APED-5608) Performance Topical Report (NEDO-10208 August 1970) 1.2.3 Effects of Fuel Bundle Flow Blockage Topical Report (GE-APED-5608) Topical Report (NEDO-10174 July 1970) 1.2.4 Verification of Fuel Damage Limit Topical Report (GE-APED-5608) Dresden 2/3 - Amendment 14/15 Topical Report (NEDO-10173 May 1970) 1.2.6 Effects of Cladding Temperature and Topical Report (GE-APED-5608) Materials on CSCS Performance Topical Report (GE-APED-5458) Topical Report (NEDO-10179 June 1970) 1.2.5 Quality Assurance and Inspection of FSAR (Incorporated in Design - the Reactor Primary System Section 4 and Appendix D) 1.2.7 Control Rod Block Monitor Design FSAR (Incorporated in Design - Sections 1, 7 and Appendix G) Dresden 2/3 - Amendments 17/18 and 19/20 Brunswick 1/2 - Supplement 5 1.2.8 Station Startup Program Topical Report (GE-APED-5698) FSAR (Incorporated in Design - Section 13) 1.2.9 Main Steam Line Isolation Valve FSAR (Incorporated in Design - Testing Under Simulated Accident Section 4) Conditions Topical Report (GE-APED-5750) Topical Report (GE-NEDO-10045) Topical Report (GE-APED-5608) 1.2.10 Performance Testing of the Plant FSAR (Incorporated in Design - Standby Diesel Generator System Section 8) General Motors Report 1.2.11 Formulation of an In-Service FSAR (Incorporated in Design - Inspection Program Section 4) Technical Specifications - Sections 3/4) 1.2.12 Diversification of CSCS Initiation FSAR (Incorporated in Design - Signals Sections 6 and 7) 1.2.13 Control Systems for Emergency Power FSAR (Incorporated in Design - Section 8) 1.2.14 Misorientation of Fuel Assemblies FSAR (Incorporated in Design - Section 3) 1.2.15 Concern of Dr. Stephen H. Hanauer - FSAR (Incorporated in Design - emergency power and Core Standby Sections 6, 8 and 14, Appendix I) Cooling Systems

BFN-16 TABLE 1.11-2 (Cont'd) (Sheet 2) BROWNS FERRY NUCLEAR PLANT AEC-ACRS CONCERNS - RESOLUTIONS Identification Section No. AEC-ACRS Concern Browns Ferry Resolutions 1.2.16 Fuel Clad Disintegration Limitations FSAR (Incorporated in Design - Section 6) Topical Report (GE-APED-5608) Dresden 2/3 - Amendment 7/8 1.2.17 General concerns with regard to reactors FSAR (Incorporated in Design - of high power density and all large Appendix I) water-cooled power reactors

BFN-16 TABLE 1.11-3 (Sheet 1) BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 2 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.2.1 Units 1 and 2 ACRS concerns FSAR Appendix I, subsection 1.2 1.3.2.2 Core Spray Cooling Effectiveness FSAR (Incorporated in Design Section 6) Topical Report (GE-APED-5458) 1.3.2.3a Reliability of CSCS Injection Valves FSAR (Incorporated in Design - Section 4, 6 and 7) 1.3.2.3b Diversification of the CSCS Initiation FSAR (Incorporated in Design - Signals Sections 6 and 7) 1.3.2.3c Sequencing of CSCS FSAR (Incorporated in Design - Sections 6 and 8) 1.3.2.3d Core Spray Cooling Effectiveness (See 1.3.2.2 above) 1.3.2.3e Performance Testing of the Standby FSAR (Incorporated in Design - Diesel Generator System Section 8) General Motors Report 1.3.2.3f Fuel Failure Modes FSAR (Incorporated in Design - Sections 6, 7, 14 and Appendix A) Topical Report (GE-APED-5652) Topical Report (GE-APED-5756) Topical Report (GE-APED-5448) Topical Report (GE-APED-5528) Topical Report (GE-APED-5455) Topical Report (GE-APED-5608) Topical Report (GE-APED-5458) Topical Report (NEDO 10208 August 1970) Topical Report (NEDO 10174 July 1970) Topical Report (NEDO 10173 May 1970) Topical Report (NEDO 10179 June 1970) Dresden 2/3 Amendment 7/8 1.3.2.4 Control Rod Block Monitor Design FSAR (Incorporated in Design - Sections 1, 7 and Appendix G) Dresden 2/3 Amendments 17/18 and 19/20 Brunswick 1/2 Supplement 5 1.3.2.5 Core Cooling FSAR (Incorporated in Design - Section 6) Topical Report (GE-APED-5458) 1.3.2.6 Control Rod Worth Minimizer Topical Report (GE-APED-5449) FSAR (Incorporated in Design - Section 7) 1.3.2.7 Control Rod Velocity Limiter Topical Report (GE-APED-5446) FSAR (Incorporated in Design - Section 3)

BFN-16 TABLE 1.11-3 (Cont'd) (Sheet 2) BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 2 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.2.8 In-Core Nuclear Instrumentation Topical Report (GE-APED-5456) Topical Report (GE-APED-5706) FSAR (Incorporated in Design - Section 7) 1.3.2.9 Jet Pump Development Topical Report (GE-APED-5460) 1.3.2.10.1 Core Analytical Models FSAR (Incorporated in Design - Sections 6, 7 and 14) Topical Report (GE-APED-5652) Topical Report (GE-APED-5756) Topical Report (GE-APED-5448) Topical Report (GE-APED-5528) Topical Report (GE-APED-5455) Dresden 2/3 Amendment 10/11 1.3.2.10.2 Fuel Failure Modes (See 1.3.2.3f above) 1.3.2.10.3 Electrical Load Control Using FSAR (Incorporated in Design - Variable Speed Reactor Coolant Section 7) Recirculation System Pumps Startup Test Results Oyster Creek No. 1 Nine Mile Point No. 1 Dresden No. 2 Millstone No. 1 1.3.2.10.4 Diversification of the CSCS Initiation (See 1.3.2.3b above) Signals 1.3.2.10.5 Main Steam Line Isolation Valve FSAR (Incorporated in Design - Testing Under Simulated Accident Section 4) Conditions Topical Report (GE-APED-5750) Topical Report (GE-NEDO-10045) Topical Report (GE-APED-5608) 1.3.2.10.6 Performance Testing of the Station (See 1.3.2.3e above) Standby Diesel Generator System

BFN-16 TABLE 1.11-4 BROWNS FERRY NUCLEAR PLANT UNIT 3 AEC-STAFF CONCERNS - RESOLUTIONS Identification Section No. AEC-Staff Concern Resolutions 1.3.3.1 Performance Testing of the Standby (See 1.3.2.3e above, Table 1.11-3) Diesel Generator System 1.3.3.2 Reactor Building Basement Corner FSAR (Incorporated in Design - Room Flooding Section 4) 1.3.3.3 Automatic Pressure Relief System FSAR (Incorporated in Design - Initiation Interlock Sections 6 and 7) 1.3.3.4 Criterion 35 Intent FSAR (Incorporated in Design - Section 4 Appendix A) 1.3.3.5 RPV-Stub Tube Design FSAR (Incorporated in Design - Section 4) Topical Report (GE-APED-5703) 1.3.3.6 Requirements for Further Technical (See Table 1.11-3) Information from Unit 1 and 2 C.P. 1.3.3.7 CSCS Thermal Effects on The Reactor Topical Report (GE-NEDO-10029) Vessel and Internals FSAR (Incorporated in Design - Sections 3 and 4) 1.3.3.8 Depressurization Performance of HPCIS FSAR (Incorporated in Design - Section 6) Topical Report (GE-APED-5608) Topical Report (GE-APED-5447) 1.3.3.9 Electrical Equipment Inside FSAR (Incorporated in Design - Containment Section 7) 1.3.3.10 Primary System Leakage Detection FSAR (Incorporated in Design - Sections 4 and 10)

BFN-16 TABLE 1.11-5 (Sheet 1) AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions_ 1.4.2 Ring Header Leakage Design FSAR (Incorporated in Design - Sections 4, 5 and 6) 1.4.3 CSCS Thermal Effects on The Reactor Topical Report (GE-NEDO-10029) Vessel and Internals FSAR (Incorporated in Design - Sections 3 and 4) 1.4.4 Effects of Blowdown Forces on Reactor FSAR (Incorporated in Design - Primary System Components Sections 3, 4 and Appendix C) 1.4.5 Separation of Control and Protection FSAR (Incorporated in Design - System Functions Sections 6, 7 and Appendix A) 1.4.6 Instrumentation For Prompt Detection FSAR (Incorporated in Design - of Gross Fuel Failures Section 7) Brunswick 1/2 - Supplements 3 and 4 1.4.7 Design of Piping Systems to Withstand FSAR (Incorporated in Design - Earthquake Forces Section 12 and Appendix C) Dresden 2/3 - Amendment 13/14 1.4.8 LPCIS - Logic Control System Design FSAR (Incorporated in Design - Section 6) 1.4.9 Reevaluation of Main Steam Line Break Topical Report (GE-APED-5608) Accident Topical Report (NEDO-10045) FSAR (Incorporated in Design - Section 14) 1.4.10 Depressurization Performance of HPCIS FSAR (Incorporated in Design - Section 6) Topical Report (GE-APED-5608) Topical Report (GE-APED-5447) 1.4.11 AEC General Design Criteria No. 35 FSAR (Incorporated in Design - Intent Design Conformance Section 4) 1.4.12 Automatic Pressure Relief System FSAR (Incorporated in Design - Initiation Interlock Sections 6 and 7) 1.4.13 Scram Reliability Study FSAR (Incorporated in Design - Sections 3 and 7) Study Results (To be Available Early 1970) Brunswick 1/2, Supplement 6 1.4.14 Design Basis of Engineered Safety Topical Report (GE-APED-5756) Features FSAR (Examined Capability of Design - Section 14)

BFN-16 TABLE 1.11-5 (Cont') (Sheet 2) AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions 1.4.15 Hydrogen Generation Study Topical Report (GE-APED-5454) Topical Report (GE-APED-5654) Brunswick 1/2, Supplement 4 Study Results (To be Available Middle 1970) 1.4.16 Primary Containment Inerting Topical Report (GE-APED-5454) Topical Report (GE-APED-5654) FSAR (Incorporate in Design - Sections 5 and 6) Dresden 2 - ACRS Letter, 9/10/69) 1.4.17 Seismic Design and Analysis Models FSAR (Re-Confirmation of Design - Section 12 and Appendix C) Dresden 2 - Re-Confirmation Information (submitted October 1969) 1.4.18 Automatic Pressure Relief System FSAR (Incorporated in Design - Single Component Failure Capability Sections 6 and 8) Manual Operation 1.4.19 Matters of Current Regulatory Staff Applicant Discussion (a) Standby Gas Treatment System FSAR (Incorporated in Design - Electrical and Physical Separation Sections 5, 7, and 8) (b) Official, Issued Technical Specifi- Proposed Technical Specifications cations - License Appendix A Appendix B 1.4.20 Flow Reference Scram FSAR (Incorporated in Design - Section 7) 1.4.21 Future Items of Consideration for Incorporation .... (a) Radiolytic Decomposition of Topical Report (GE-APED-5454) Cooling Water Topical Report (GE-APED-5654) Brunswick 1/2, Supplement 4 Study Results (To be Available Middle 1970) (b) Development of Instrumentation FSAR (Justified Design - Sections Vibration and Loose Parts 3, 4 and Appendix C) Detection (c) Consequences of Water Contamination FSAR (Incorporated in Design - Structural Material - LOCA Section 14) 1.4.22 Diesel Generator Synchronization FSAR (Incorporated in Design - Considerations Sections 6, 7, and 8)

BFN-16 TABLE 1.11-5 (Cont'd) (Sheet 3) AEC ACRS CONCERNS ON OTHER DOCKETS - RESOLUTIONS Identification Section No. AEC-ACRS Concern _ Resolutions 1.4.23 Development of Instrumentation FSAR (Justified Design - Section 4) Primary Containment Leakage Technication Specification - Detection System Increased Appendix B Sections 3 and 4) Sensitivity Studies 1.4.24 Development of Instrumentation - FSAR (Justified Design - Sections 3, Vibration and Loose Parts 4, and Appendix C) Detection Studies 1.4.25 CSCS - Leakage Detection, Protection, FSAR (Justified in Design - Sections and Isolation Capability 4, 10 and Appendix A) Brunswick 1/2 - Supplement 4, C/R 6.4 1.4.26 Main Steam Lines - Standards For FSAR paragraph 1.4.2.6 Fabrication, Q/C and Inspection

BFN-16 TABLE 1.11-6 AEC ACRS CONCERNS ON OTHER DOCKETS - CAPABILITY FOR RESOLUTION Identification Section No. AEC-Staff Concern Capability for Resolution 1.5.2 Tornado and Missile Protection FSAR (Incorporated in Design - GE BWR-Spent Fuel Storage Pool Sections 2, 10, and 12) Topical Report (GE-APED-5696) 1.5.3 BWR System Stability Analysis FSAR (Incorporated in Design - Section 7) Topical Report (GE-APED-5652) Topical Report (GE-APED-5640) Peach Bottom 2/3 - Amendment 2

BFN-28 1.12 GENERAL CONCLUSIONS Based on the favorable plant site characteristics, on the design of the plant herein analyzed, on the criteria, principles, and design requirements pertinent to safety, on the calculated potential consequences of routine and accidental release of radioactive material to the environs, on the results of research and development programs, and on the technical competence of the applicant and his contractors, there is reasonable assurance that the Browns Ferry Nuclear Plant can be operated without endangering the health and safety of the public. NRC has accepted these conclusions and has issued operating licenses for all three units of the Browns Ferry Nuclear Plant. The dates of issuance for each unit's operating license are as follows: UNIT ONE - June 26, 1973 UNIT TWO - June 28, 1974 UNIT THREE - July 2, 1976 A subsequent license amendment for a five percent core thermal power uprate, from 3293 MWt to 3458 MWt, was issued for Units 2 and 3 on September 8, 1998, and for Unit 1 on March 6, 2007. On August 14, 2017, license amendments were issued for all three units for a core thermal power uprate from 3458 to 3952 MWt. Operating data for each unit will be provided to the NRC. This operating data shall comply with the operating data (for each calendar month) as described in Generic Letter 97-02 Revised Contents of the Monthly Operating Report. This data will be provided by the last day of the month following the end of each calendar quarter. This operating data may be provided by the use of an Industry Database (e.g., the Industrys Consolidated Data Entry Program (CDE) or other reports to prevent any gaps in the monthly operating statistics and shutdown experience provided to the NRC. 1.12-1

BFN-29 CHAPTER 2 TABLE OF CONTENTS Page No. 2.1 Summary Description ................................................................................................................ 2.1-1 2.2 Site Description ......................................................................................................................... 2.2-1 2.2.1 Location ..................................................................................................................... 2.2-1 2.2.2 Population .................................................................................................................. 2.2-1 2.2.3 Land Use.................................................................................................................... 2.2-2 2.3 Meteorology ............................................................................................................................... 2.3-1 2.3.1 General ...................................................................................................................... 2.3-1 2.3.2 Climatology ................................................................................................................ 2.3-1 2.3.3 Atmospheric Stability ................................................................................................. 2.3-1 2.3.4 Wind ........................................................................................................................... 2.3-2 2.3.5 Temperature and Precipitation ................................................................................... 2.3-5 2.3.6 Storms ....................................................................................................................... 2.3-6 2.3.7 Onsite Meteorological Measurement Program ........................................................... 2.3-7 2.3.8 Conclusions ............................................................................................................... 2.3-10 2.4 Hydrology, Water Quality, and Aquatic Biology ......................................................................... 2.4-1 2.4.1 General ...................................................................................................................... 2.4-1 2.4.2 Hydrology ................................................................................................................... 2.4-1 2.4.3 Water Quality ............................................................................................................. 2.4-5 2.4.4 Water Use .................................................................................................................. 2.4-7 2.4.5 Aquatic Biota .............................................................................................................. 2.4-9 2.4.6 Monitoring Programs .................................................................................................. 2.4-9 2.4.7 Conclusions ............................................................................................................... 2.4-10 Appendix 2.4A Probable Maximum Flood .......................................................................................... 2.4A-i 2.5 Geology and Seismology........................................................................................................... 2.5-1 2.5.1 General ...................................................................................................................... 2.5-1 2.5.2 Geology ..................................................................................................................... 2.5-1 2.5.3 Seismology ................................................................................................................ 2.5-9 2.5.4 Conclusions ............................................................................................................... 2.5-11 2.5.5 Seismic Instrumentation Program .............................................................................. 2.5-11 2.5.6 References................................................................................................................. 2.5-16 2.6 Environmental Radiological Monitoring Program ....................................................................... 2.6-1 2.6.1 General ...................................................................................................................... 2.6-1 2.6.2 Monitoring Program ................................................................................................... 2.6-1 2.0-i

BFN-29 LIST OF TABLES Table Title 2.2-1 (Deleted) 2.2-2 1970 and 1980 Population of all Incorporated Places in 1980 Within 60-Mile Radius of Browns Ferry Site 2.2-3 Cities and Towns Within 60 Miles Having 2,500 or More Residents in 1990 2.2-4 Population Distribution Within 10 Miles of the Site: 1970 - 2020 2.2-5 60-Mile Population Distribution Within 60 Miles of Site: 1970 - 2020 2.2-6 Population Density at Various Distances from the Site: 1970 - 2020 2.2-7 Peak Hour Recreation Visitation Within 10 Miles of the Site: 1986 - 2020 2.2-8 Listing of 1993 School Enrollments and Industrial Employment Within 10 Miles of the Site 2.2-9 Browns Ferry Nuclear Plant Statistical Data for Nearby Counties 2.2-10 Hazardous River Traffic that Passes Browns Ferry Nuclear Plant: 1983-1993 2.2-11 (Deleted) 2.3-1 Joint Percentage Frequencies of Wind Speed by Stability Class 2.3-2 Joint Percentage Frequencies of Wind Speed by Stability Class 2.3-3 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class A 2.3-4 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class B 2.3-5 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class C 2.3-6 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class D 2.3-7 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class E 2.3-8 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class F 2.3-9 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class G 2.3-10 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class 2.3-11 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - January (77, 78, 79) 2.3-12 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - February (77, 78, 79) 2.0-ii

BFN-29 LIST OF TABLES (Cont'd) Table Title 2.3-13 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - March (77, 78, 79) 2.3-14 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - April (77, 78, 79) 2.3-15 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - May (77, 78, 79) 2.3-16 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - June (77, 78, 79) 2.3-17 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - July (77, 78, 79) 2.3-18 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - August (77, 78, 79) 2.3-19 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - September (77, 78, 79) 2.3-20 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - October (77, 78, 79) 2.3-21 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - November (77, 78, 79) 2.3-22 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - December (77, 78, 79) 2.3-23 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class A 2.3-24 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class B 2.3-25 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class C 2.3-26 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class D 2.3-27 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class E 2.3-28 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class F 2.3-29 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class G 2.3-30 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class 2.0-iii

BFN-29 LIST OF TABLES (Cont'd) Table Title 2.3-31 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - January (77, 78, 79) 2.3-32 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - February (77, 78, 79) 2.3-33 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - March (77, 78, 79) 2.3-34 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - April (77, 78, 79) 2.3-35 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - May (77, 78, 79) 2.3-36 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - June (77, 78, 79) 2.3-37 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - July (77, 78, 79) 2.3-38 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - August (77, 78, 79) 2.3-39 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - September (77, 78, 79) 2.3-40 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - October (77, 78, 79) 2.3-41 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - November (77, 78, 79) 2.3-42 Joint Percentage Frequencies of Wind Speed by Direction - Disregarding Stability Class - December (77, 78, 79) 2.3-43 Wind Direction Persistence Data - Disregarding Stability - 10 M Level 2.3-44 Wind Direction Persistence Data - Disregarding Stability - 93 M Level 2.3-45 Temperature Data 1879 - 1958 2.3-46 Temperature Data - January 1, 1977 - December 31, 1979 2.3-47 Precipitation Data - Huntsville, Alabama 2.3-48 Precipitation Data - Browns Ferry Nuclear Plant - 1/1/77 - 12/31/79 2.3-49 Snowfall Data 2.0-iv

BFN-29 LIST OF TABLES (Cont'd) Table Title 2.4-1 (Deleted) 2.4-2 (Deleted) 2.4-3 (Deleted) 2.4-4 Public and Industrial Surface Water Supplies Withdrawn From The 61 Mile Reach of the Tennessee River Between Decatur, Alabama and TVA Colbert Steam Plant 2.4-5 (Deleted) 2.4-6 Private Water Wells Within Two Miles of Browns Ferry Nuclear Plant Stack, June 1995 2.4A Table 1 - Facts about TVA Dams and Reservoirs Table 2 - Unit Hydrograph Data Table 3 - Probable Maximum Flood Rainfall and precipitation Excess Table 4 - (deleted) 2.5-1 Browns Ferry Nuclear Plant - Historical Earthquake Listing 2.0-v

BFN-29 LIST OF ILLUSTRATIONS Figure Title 2.2-1 Browns Ferry Site Location Map Mile Radius 2.2-2 Browns Ferry Site Location Map Mile Radius 2.2-3 Deleted 2.2-4 Location of Principal Plant Structures 2.3-1 Wind Rose Plot Foot Level - Stability Class A 2.3-2 Wind Rose Plot Foot Level - Stability Class B 2.3-3 Wind Rose Plot Foot Level - Stability Class C 2.3-4 Wind Rose Plot Foot Level - Stability Class D 2.3-5 Wind Rose Plot Foot Level - Stability Class E 2.3-6 Wind Rose Plot Foot Level - Stability Class F 2.3-7 Wind Rose Plot Foot Level - Stability Class G 2.3-8 Wind Rose Plot Foot Level - All Stability Classes 1/1/77 - 12/31/79 2.3-9 Wind Rose Plot Foot Level - All Stability Classes January (77, 78, 79) 2.3-10 Wind Rose Plot Foot Level - All Stability Classes February (77, 78, 79) 2.3-11 Wind Rose Plot Foot Level - All Stability Classes March (77, 78, 79) 2.3-12 Wind Rose Plot Foot Level - All Stability Classes April (77, 78, 79) 2.3-13 Wind Rose Plot Foot Level - All Stability Classes May (77, 78, 79) 2.3-14 Wind Rose Plot Foot Level - All Stability Classes June (77, 78, 79) 2.3-15 Wind Rose Plot Foot Level - All Stability Classes July (77, 78, 79) 2.3-16 Wind Rose Plot Foot Level - All Stability Classes August (77, 78, 79) 2.3-17 Wind Rose Plot Foot Level - All Stability Classes September (77, 78, 79) 2.3-18 Wind Rose Plot Foot Level - All Stability Classes October (77, 78, 79) 2.3-19 Wind Rose Plot Foot Level - All Stability Classes November (77, 78, 79) 2.3-20 Wind Rose Plot Foot Level - All Stability Classes December (77, 78, 79) 2.0-vi

BFN-29 LIST OF ILLUSTRATIONS (Cont'd) Figure Title 2.3-21 Wind Rose Plot - 300-Foot Level - Stability Class A 2.3-22 Wind Rose Plot - 300-Foot Level - Stability Class B 2.3-23 Wind Rose Plot - 300-Foot Level - Stability Class C 2.3-24 Wind Rose Plot - 300-Foot Level - Stability Class D 2.3-25 Wind Rose Plot - 300-Foot Level - Stability Class E 2.3-26 Wind Rose Plot - 300-Foot Level - Stability Class F 2.3-27 Wind Rose Plot - 300-Foot Level - Stability Class G 2.3-28 Wind Rose Plot - 300-Foot Level - All Stability Classes 1/1/77 - 12/31/79 2.3-29 Wind Rose Plot - 300-Foot Level - All Stability Classes January (77 - 79) 2.3-30 Wind Rose Plot - 300-Foot Level - All Stability Classes February (77 - 79) 2.3-31 Wind Rose Plot - 300-Foot Level - All Stability Classes March (77 - 79) 2.3-32 Wind Rose Plot - 300-Foot Level - All Stability Classes April (77 - 79) 2.3-33 Wind Rose Plot - 300-Foot Level - All Stability Classes May (77 - 79) 2.3-34 Wind Rose Plot - 300-Foot Level - All Stability Classes June (77 - 79) 2.3-35 Wind Rose Plot - 300-Foot Level - All Stability Classes July (77 - 79) 2.3-36 Wind Rose Plot - 300-Foot Level - All Stability Classes August (77 - 79) 2.3-37 Wind Rose Plot - 300-Foot Level - All Stability Classes September (77 - 79) 2.3-38 Wind Rose Plot - 300-Foot Level - All Stability Classes October (77 - 79) 2.3-39 Wind Rose Plot - 300-Foot Level - All Stability Classes November (77 - 79) 2.3-40 Wind Rose Plot - 300-Foot Level - All Stability Classes December (77 - 79) 2.4-1 Cross Sections of Wheeler Reservoir at Browns Ferry Site 2.4-1a Location Plan for Cross Sections of Wheeler Reservoir at Browns Ferry Site 2.4-1b Cross Sections of Wheeler Reservoir at Browns Ferry Site - Sheet 1 2.4-1c Cross Sections of Wheeler Reservoir at Browns Ferry Site - Sheet 2 2.4.1d Cross Sections of Wheeler Reservoir at Browns Ferry Site - Sheet 3 2.0-vii

BFN-29 LIST OF ILLUSTRATIONS (Cont'd) Figure Title 2.4-1e Cross Sections of Wheeler Reservoir at Browns Ferry Site - Sheet 4 2.4-2 Channel Profile - Wheeler Reservoir 2.4-3 Privately owned Groundwater wells - 1 and 2 mile Radius from the Stack 2.4-4 Ground Water Supplies Within 20 Mile Radius of Plant Site 2.4A Figure 1 - TVA River Operations Dam Location Map Figure 2 - Seasonal Operating Curve, Cherokee Figure 3 - Seasonal Operating Curve, Guntersville Figure 4 - Browns Ferry Nuclear Plant Hydrologic Model Unit Areas Figure 5 - Sheet 1 of 14 - Unit Hydrographs, Areas 1 - 5 Sheet 2 of 14 - Unit Hydrographs, Areas 6 - 9 Sheet 3 of 14 - Unit Hydrographs, Areas 10 - 13 Sheet 4 of 14 - Unit Hydrographs, Areas 14 - 18 Sheet 5 of 14 - Unit Hydrographs, Areas 19 - 22 Sheet 6 of 14 - Unit Hydrographs, Areas 23 - 27 Sheet 7 of 14 - Unit Hydrographs, Areas 33, 34, 36 Sheet 8 of 14 - Unit Hydrographs, Areas 35, 37 Sheet 9 of 14 - Unit Hydrographs, Areas 38, 39, 41, 42 Sheet 10 of 14 - Unit Hydrographs, Areas 40, 43, 44A, 44B Sheet 11 of 14 - Unit Hydrographs, Areas 45, 46, 47A, 47B Sheet 12 of 14 - Unit Hydrographs, Areas 48 - 50 Sheet 13 of 14 - Unit Hydrographs, Areas 51 - 52, 54 - 56, 58, 63 - 65 Sheet 14 of 14 - Unit Hydrographs, Areas 53, 57, 59 - 62 Figure 6 - Hydrologic Model Verification - 1973 Flood, Observed Elevations vs. Computed Elevations Figure 7 - Hydrologic Model Verification - 1973 Flood, Observed Flows vs. Computed Flows Figure 8 - Hydrologic Model Verification - 2004 Flood, Observed Elevations vs. Computed Elevations Figure 9 - Hydrologic Model Verification - 2004 Flood, Observed Flows vs. Computed Flows Figure 10 - Hydrologic Model Verification - Steady State Profiles Figure 11 - Hydrologic Model Verification - Guntersville Tailwater Curve 2.0-viii

BFN-29 LIST OF ILLUSTRATIONS (Cont'd) Figure Title Figure 12 - Guntersville Hydro Plant, General Plan Elevation and Sections Figure 13 - Probable Maximum Precipitation Isohyets for 21,400 Sq. Mi. Event Figure 14 - Probable Maximum Precipitation Isohyets for 16,170 Sq. Mi. Event Figure 15 - 21,400 Sq. Mi. Downstream Centered March Flood Event for BFN, Design Basis PMF Discharge Figure 16 - 21,400 Sq. Mi. Downstream Centered March Flood Event for BFN, Design Basis PMF Elevation Figure 17 - (Deleted) Figure 18 - (Deleted) Figure 19 - (Deleted) Figure 20 - (Deleted) Figure 21 - Watershed, Unnamed Tributary Northwest of Plant Figure 22 - General Plan Figure 22a - Plant Topography Figure 22b - Probable Maximum Precipitation Point Rainfall Figure 23 - One Hour Unit Hydrograph for Unnamed Steam Northwest of Plant Figure 24 - Probable Maximum Flood for Unnamed Steam Northwest of Plant Figure 25 - Channel Relocation West of Cooling Towers - Typical Sections Figure 26, Sheet 1 - Channel Relocation West of Cooling Towers Figure 26, Sheet 2 - Channel Relocation West of Cooling Towers 2.5-S1 Foundation Investigation Grid Low Level Radwaste Storage 2.5-S2 Foundation Investigations Grid Low Level Radwaste Storage - Soils Investigation 2.5-S3 Generalized Cross Section LLRW and A&B Structures 2.5-S4 Generalized Cross Section LLRW O&OOE B&C Structures 2.5-S5 Core Hole Correlation 2.5-1 Location and Summary of Exploratory Drilling 2.5-2 Geologic Sections 2.5-3 Geologic Sections 2.5.4 Drill Hole Locations and Core Drill Summary 2.0-ix

BFN-29 2.5-5 Graphic Logs of Core Drill Holes 2.5-5a Miscellaneous Foundation Investigations 1972 - 1980 2.5-5b Access Highway Bridge Drill Layout 2.5-5c Access Highway Bridge Hole BF-1 2.5-5d Access Highway Bridge Hole BF-2 2.5-5e Access Highway Bridge Hole BF-3 2.5-5f Access Highway Bridge Hole BF-4 2.5-5g Access Highway Bridge Hole BF-5 2.0-x

BFN-29 LIST OF ILLUSTRATIONS (Cont'd) Figure Title 2.5-5h Access Highway Bridge Hole BF-6 2.5-5i Access Highway Bridge Hole BF-7 2.5-5j Access Highway Bridge Hole BF-8 2.5-5k Access Highway Bridge Hole BF-9 2.5-5l Access Highway Bridge Hole BF-10 2.5-5m through 2.5-5aj (Deleted) 2.5-6 Historical Earthquake Map 2.5-7 Hodified Mercalli Intensity Scale of 1931 (Abridged) 2.5-8 Site Design Spectrum - Operational Basis Earthquake - Horizontal 2.5-9 Site Design Spectrum - Design Basis Earthquake - Horizontal 2.5-10 Comparison of Site Design Spectrum and El Centro Spectrum - 5 Percent Damping 2.5-11 Site Spectrum - Compatible Time History 2.5-12 Comparison Of Site Spectrum and Spectrum Of Acceleration Time History - 0.5 Percent Damping 2.5-13 Comparison Of Site Spectrum and Spectrum Of Acceleration Time History - 1.0 Percent Damping 2.5-14 Comparison Of Site Spectrum and Spectrum Of Acceleration Time History - 2.0 Percent Damping 2.5-15 Comparison Of Site Spectrum and Spectrum Of Acceleration Time History - 5.0 Percent Damping 2.5-16 Comparison Of Site Spectrum and Spectrum Of Acceleration Time History - 3.0 Percent Damping 2.5-17 Mechanical Instruments And Controls 2.5-18 (Deleted) 2.5-19 Rock Excavation 2.0-xi

BFN-16 2.0 SITE 2.1 Summary Description The "Site Description" section gives the size, geographical location, and certain features of the Browns Ferry site with an analysis of the population distribution and land use in the areas adjacent to the site. "Meteorology" presents the climate and weather of the Browns Ferry site. Included in this section are results of onsite weather measurements, which are the basis for determining diffusion and transport properties of the atmosphere. The sections "Hydrology, Water Quality, and Marine Biology" present data on the streamflow, temperature, and aquatic life of the Tennessee River at the site. Also included are studies of subsurface waterflow and uses of water in the plant area. Details of the geological formations underlying plant structures and the general site area are given in "Geology and Seismology." The seismic history of the area is presented with an analysis of the earthquake hazard at the plant site. "Environmental Monitoring Program" outlines the program for monitoring the site environs for plant effluents. The information provided in these sections was prepared by various groups within the TVA organization that have been involved for many years with the study of environmental sciences and site selection for locating and operating conventional steam plants and hydro stations. 2.1-1

BFN-30 2.2 SITE DESCRIPTION The information contained in this section is considered historical with the exception of Section 2.2.3 and Table 2.2-10, which are periodically updated. Estimated populations for the surrounding counties within a 10-mile radius are reviewed and updated if determined necessary for state and local emergency planning purposes. 2.2.1 Location The site is located on the north shore of Wheeler Lake at river mile 294 in Limestone County in north Alabama. The site is approximately 10 miles southwest of Athens, Alabama, and 10 miles northwest of the center of Decatur, Alabama. Figures 2.2-1 and 2.2-2 show the site location. The plant site and adjoining areas are shown in Figure 2.2-4 which is considered historical and is not being updated. The site contains approximately 880 acres which are owned by the United States and are in the custody of TVA. The site has been developed to accommodate three units. 2.2.2 Population 2.2.2.1 Resident Population The populations of the various towns and cities within 60 miles of the site are shown in Table 2.2-2. Only 21 towns or cities within a 60-mile radius of the site had population over 2,500 in 1990. Most of the smaller communities showed only small changes in population between 1980 and 1990. The largest center of population is about 30 miles from the site and the nearest city with a population of 25,000 or greater is Decatur. The projected growth of the larger centers of population is shown in Table 2.2-3. The greatest change in population is projected for Huntsville and Decatur. Population distributions from the site for various directions and distances for the years 1970, 1980, 1986, 1990, 2000, 2010, and 2020 are shown in Tables 2.2-4 and 2.2-5. Within a 4-mile radius of the site, the 1990 population was 1425 persons for a density of about 45 persons per square mile, with a slight increase expected through 2020, as shown in Table 2.2-6. The population within a 10-mile radius of the site is expected to increase from 26,740 in 1970 to 33,340 by 2020 with a corresponding increase in population density from 101 to 126 persons per square mile. There are only three towns within a radius of 20 miles (Athens and Decatur, 10 miles to the northeast and south, south-southeast, respectively and Moulton, 18 miles to the southwest), having a 1980 population greater than 1,800 persons. The population of Athens is expected to increase from 14,360 in 1970 to 18,600 in 2020. The population in Decatur is 2.2-1

BFN-27 expected to increase from 38,044 to 54,000 in the same time period. Within a 60-mile radius the largest city is Huntsville, located approximately 30 miles due east from the site. The population of Huntsville is expected to increase from its 1970 level of 139,282 to 177,100 by 2020. 2.2.2.2 Transient Population Transient population consists of visitors to recreation sites, students in schools, and employees at industrial facilities. Recreation--Estimated and projected peak hour visitation to recreation facilities within 10 miles of the plant are contained in Table 2.2-7. The visitation is based on the maximum capacity of facilities plus some overflow. Capacities are based on an inventory done in 1989. There are no recreation facilities beyond 10 miles which are large enough to cause significant variations in the total population within any annular segment. Schools--Eleven schools are located within 10 miles of the Browns Ferry Nuclear Plant. In 1993 these schools served 4,350 students, distributed as shown in Table 2.2-8. Industries--Industries located in the 5 to 10 mile area are shown in Table 2.2-8. Employment in 1993 ranges from 80 to 1,700 for the day shift. 2.2.3 Land Use Few centers of population exist within a 60-mile radius of the site. The dominant character of the land is small, scattered villages and homes in an agricultural area. Employment in the counties near the site is shown in Table 2.2-9. In 1990 agriculture employment was a higher percent of the labor force in Lawrence and Limestone counties than Morgan and Madison Counties where there was a greater concentration of manufacturing employment. The area immediately surrounding the site is primarily agricultural with industrial areas concentrated along the Tennessee River primarily at the large centers of population. The closest industrial area is adjacent to Decatur. 3M Company, the NUCOR Steel Company, Worthington Steel Company, United Launch Alliance, and the Ascend Performance Materials, LLC, Plant are the largest industries in Morgan County and are approximately 4 to 7 air miles from the site. Browns Ferry, GM, and Steel Case are the largest industries in Limestone County. GM is located approximately 10 miles from the plant and Steel Case is located approximately 9 miles from the plant. The largest industrial complex is at the Redstone Arsenal, which is located approximately 25 miles east of the site. This is the NASA center for research and development and is the principal single economic force in the area. The remaining industrial area is located in the quad-cities area. 2.2-2

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 BFN-27 The nearest site boundary is approximately 4,000 feet northeast of the reactor building. The nearest house is approximately 5,400 feet north-northwest of the center of the site. There are no rai lroads or principal highways penetrating the site. The closest railroad tracks are those of the Louisville and Nashville railroad approximately 8 miles east of the site running in the north-south direction, and those of the Southern railroad about 6 miles south of the site runn ing in the east-west direction, as shown in Figure 2.2-2. The nearest principal highways are U.S. 72, about 6 miles north of the site, and State Highway 20, about 4.5 miles south of the site. The Browns Ferry Nuclear Plant is located on a 9-foot deep navigable channel on Wheeler Reservoir. Table 2.2-10 shows the total amount of certain hazardous materials shipped past the Browns Ferry Nuclear Plant from 1983 to 1993 on a yearly basis. The product listed as gasoline on the table is actually RU250. The nearest missile testing facility is located on Redstone Arsenal about 25 miles east of the plant. The current range of any tested missile is approximately 8 kilometers (about 5 miles) which is well within the maximum range of any tested missile which is approximately 15 ki lometers (about 9 miles). Adeq uate safety precautions are exercised at all times during all testing to prevent impacts outside the test range. There are no airports within five miles of the site. The Athens-Decatur Airport is about 10 miles east of the plant. The nearest commercial airport is located in Huntsville about 25 miles from the site. The Athens-Decatur field may serve up to 10 jets, 10 turboprops, and 100 light (under 150 horsepower) aircraft operations on a busy day. 2.2-3

BFN-18 Table 2.2-1 (Deleted by Amendment 7)

BFN-18 Table 2.2-2 (SHEET 1) Historical Information 1970 and 1980 POPULATION OF ALL INCORPORATED PLACES IN 1980 WITHIN 60-MILE RADIUS OF BROWNS FERRY SITE Population Town County Direction Miles 1970 1980 0-10 Trinity Morgan S 7.0 881 1,328 Hillsboro Lawrence SW 6.2 222 278 Total 1,103 1,606 10-20 Athens Limestone NE 10.0 14,360 14,558 Elkmont Limestone NNE 16.6 394 429 Mooresville Limestone ESE 13.8 72 58 Decatur Morgan SE 10.0 38,044 42,002 Flint City Morgan SE & SSE 15.3 404 673 Priceville Morgan SE 18.0 --- 966** Moulton Lawrence SW & SSW 18.9 2,470 3,197 Courtland Lawrence W & WSW 11.5 547 456 Town Creek Lawrence W 17.0 1,203 1,201 Rogersville Lauderdale NW 13.0 950 1,224 Total 58,444 64,764 20-30 Lester Limestone N 20.0 70 117 Minor Hill Giles (T) N 24.0 315 564 Elkton Giles (T) NNE 27.0 341 540 Ardmore Giles (T) and Limestone NE 25.5 1,362 1,931 Huntsville Madison E 29.5 139,282 142,513 Madison Madison E 20.0 3,086 4,057 Triana Madison ESE 23.8 228 285 Hartselle Morgan SSE 21.0 7,355 8,858 Somerville Morgan SE 23.4 180 140 Falkville Morgan SE 26.0 946 1,310 Leighton Colbert W 24.0 1,231 1,218 Killen Lauderdale WNW 26.7 683 747 Lexington Lauderdale NW 23.0 278 884 Anderson Lauderdale NNW 28.0 --- 405** Total 155,357 163,569

BFN-18 Table 2.2-2 (Continued) (SHEET 2) Historical Information Population Town County* Direction Miles 1970 1980 30-40 Pulaski Giles (T) N 34.0 6,989 7,184 Lawrenceburg Lawrence (T) NNW 39.0 8,889 10,184 Owens Cross-roads Madison ESE 37.5 767 804 Eva Morgan SE 32.0 146 185 South Vinemont Cullman SSE 33.2 480 615 Cullman Cullman SSE 40.0 12,601 13,084 Addison Winston S 35.5 692 746 Muscle Shoals Colbert W 31.0 6,907 8,911 Sheffield Colbert W 33.4 3,115 11,903 Tuscumbia Colbert W 34.0 8,828 9,137 Littleville Colbert WSW 34.3 858 1,262 Russellville Franklin WSW 38.3 7,814 8,195 Florence Lauderdale W & WNW 33.0 34,031 37,029 St. Florian Lauderdale WNW 32.0 --- 305** Loretto Lawrence (T) NW 31.8 1,375 1,612 St. Joseph Lawrence (T) NW 32.3 637 897 Iron City Lawrence (T) NW 34.0 504 482 Total 104,633 112,535 40-50 Lynnville Giles (T) N 46.4 327 383 Cornersville Marshall (T) NNE 46.5 655 722 Fayetteville Lincoln (T) NE 43.0 7,030 7,559 Petersburg Lincoln (T)

               & Marshall(T)     NNE                  49.2       463         681 Etheridge      Lawrence (T)      NNW                   45.0        ---        548**

Gurley Madison E 41.0 647 735 New Hope Madison ESE 41.8 1,300 1,546 Paint Rock Jackson E 44.1 226 221 Woodville Jackson E 47.4 322 609 Grant Marshall ESE 49.2 382 632 Arab Marshall SE 43.2 4,399 5,967 Union Grove Marshall ESE 44.2 118 127 Bailytown Cullman SE 42.0 --- 396** West Point Cullman SSE 43.0 --- 177** Fairview Cullman SE 40.2 313 450 Holly Pond Cullman SE 46.2 325 493

BFN-18 Table 2.2-2 (Cont'd) (SHEET 3) Historical Information Population Town County* Direction Miles 1970 1980 Hanceville Cullman SSE 47.8 2,027 2,220 Good Hope Cullman SSE 44.4 840 1,442 Arley Winston S 45.0 164 276 Double Springs Winston SSW 42.8 957 1,057 Haleyville Winston SW 45.0 4,190 5,306 Phil Campbell Franklin SW 42.5 1,230 1,549 Bear Creek Marion SW 46.2 336 353 Cherokee Colbert W 49.2 1,484 1,589 Collinwood Wayne (T) NW 48.0 922 1,064 Total 28,657 36,102 50-60 Lewisburg Marshall (T) NNE 54.0 7,207 8,706 Mount Pleasant Maury (T) N 57.2 3,530 3,375 Huntland Franklin (T) ENE 52.8 849 983 Lynchburg Moore (T) NE 56.9 538 668 Guntersville Marshall ESE 51.5 6,491 9,041 Albertville Marshall ESE 59.5 9,963 12,039 Garden City Cullman SSE 52.0 745 655 Blountsville Blount SE 52.1 1,254 1,509 Cleveland Blount SSE 58.0 413 487 Nectar Blount SSE 57.5 --- 367** Rosa Blount SE 60.0 --- 204** Hayden Blount SSE 59.0 195 268 Snead Blount SE 59.5 347 667 Hodges Franklin WSW 54.0 207 250 Vina Franklin WSW 59.0 366 346 Hackleburg Marion SW 51.6 726 883 Brilliant Marion SW 60.0 726 871 Douglas Marshall SE 57.5 --- 116** Nauvoo Walker SSW 54.5 265 259 Lynn Winston SSW 52.6 286 554 Waterloo Lauderdale WNW 56.3 262 260 Waynesboro Wayne (T) NW 56.0 1,983 2,109 Total 36,353 44,617

*All counties are in Alabama except those Tennessee counties noted by (T).

• • Place incorporated since 1970.

Source: U.S. Bureau of the Census. U.S. Census of Population: 1980. Number of Inhabitants.

BFN-18 Table 2.2-3 HISTORICAL INFORMATION BROWNS FERRY NUCLEAR PLANT CITIES AND TOWNS WITHIN 60 MILES HAVING 2,500 OR MORE RESIDENTS IN 1990 Alabama County 19701 19801 19862 19901 20002 20102 20202 Albertville Marshall 9,963 12,039 13,000 14,507 15,400 16,100 16,900 Arab Marshall 4,399 5,967 6,600 6,321 6,700 7,000 7,400 Athens Limestone 14,360 14,558 15,400 16,901 17,500 18,000 18,600 Cullman Cullman 12,601 13,084 13,600 13,367 13,700 14,100 14,600 Decatur Morgan 38,044 42,002 44,700 48,761 50,100 51,900 54,000 Florence Lauderdale 34,031 37,029 37,600 36,426 36,500 37,000 38,200 Guntersville Marshall 6,491 7,041 7,500 7,038 7,500 7,800 8,200 Haleyville Winston 4,190 5,306 5,300 4,452 4,500 4,600 4,700 Hartselle Morgan 7,355 8,858 9,700 10,795 11,200 11,600 12,100 Huntsville Madison 139,282 142,513 156,600 159,789 166,000 170,800 177,100 Madison Madison 3,086 4,057 5,600 14,904 15,500 15,900 16,500 Moulton Lawrence 2,470 3,197 3,300 3,248 3,200 3,200 3,200 Muscle Shoals Colbert 6,907 8,911 8,700 9,611 9,500 9,700 9,900 Russellville Franklin 7,814 8,195 8,200 7,812 7,700 7,700 7,900 Sheffield Colbert 13,115 11,903 11,900 10,380 10,300 10,400 10,700 Tuscumbia Colbert 8,828 9,137 9,000 8,413 8,400 8,500 8,700 Tennessee Fayetteville Lincoln 7,030 7,559 7,600 6,921 7,200 7,400 7,700 Lawrenceburg Lawrence 8,889 10,184 10,400 10,412 10,900 11,200 11,500 Lewisburg Marshall 7,207 8,760 9,400 8,351 9,000 9,500 9,800 Mount Pleasant Maury 3,530 3,375 3,400 4,278 4,500 4,700 4,900 Pulaski Giles 6,989 7,184 7,200 7,895 7,900 8,000 8,100

1. U.S. Bureau of the Census, Census of Population: 1970, 1980, and 1990, Number of Inhabitants, Alabama and Tennessee.

2. TVA estimates.

BFN-18 Table 2.2-4 (Sheet 1) Historical Information 1970 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE Total Population 0-1 1-2 2-3 3-4 4-5 5-10 N 855 --- 10 55 35 85 670 NNE 1,055 --- 5 15 65 55 915 NE 4,335 --- 5 25 45 80 4,180 ENE 1,485 --- 15 50 40 70 1,310 E 1,025 --- --- 30 10 40 945 ESE 170 --- --- 5 --- --- 165 SE 10,400 --- --- --- --- 20 10,380 SSE 1,680 --- --- --- --- 50 1,630 S 1,395 --- --- 20 35 90 1,250 SSW 1,155 --- --- 60 75 175 845 SW 830 --- --- 20 35 90 685 WSW 480 --- --- 35 15 135 295 W 685 --- --- 25 5 30 625 WNW 130 --- --- --- 25 55 50 NW 350 --- --- --- --- 5 345 NNW 710 --- 5 35 25 20 625 Total 26,740 --- 40 375 410 1,000 24,915

BFN-18 Table 2.2-4 (Sheet 2) Historical Information 1980 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE Total Population 0-1 1-2 2-3 3-4 4-5 5-10 N 820 --- 10 50 35 80 645 NNE 1,010 --- 5 15 60 55 875 NE 5,175 --- 5 30 50 80 5,010 ENE 2,060 --- 20 70 55 100 1,815 E 1,435 --- --- 45 15 55 1,320 ESE 235 --- --- 5 --- --- 230 SE 11,245 --- --- --- --- 20 11,225 SSE 2,520 --- --- --- --- 50 2,470 S 1,455 --- --- 25 35 90 1,305 SSW 1,200 --- --- 60 80 170 890 SW 860 --- --- 25 40 95 700 WSW 495 --- --- 35 15 140 305 W 685 --- --- 25 5 30 625 WNW 130 --- --- --- 25 55 50 NW 345 --- --- --- --- 5 340 NNW 685 --- 5 35 20 20 605 Total 30,355 --- 45 420 435 1,045 28,410

BFN-18 TABLE 2.2-4 (Sheet 3) Historical Information BROWNS FERRY NUCLEAR PLANT 1986 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE 0-1 1-2 2-3 3-4 4-5 5-10 TOTAL N 34 0 0 73 118 1135 1360 NNE 0 11 20 137 48 1341 1557 NE 0 3 20 39 109 4259 4430 ENE 0 34 76 90 39 1615 1854 E 0 0 20 6 28 1408 1462 ESE 0 0 11 0 0 106 117 SE 0 0 0 0 0 6960 6960 SSE 0 0 0 0 10 2304 2314 S 0 0 14 17 66 1405 1502 SSW 0 0 46 55 258 1307 1666 SW 0 0 0 20 152 640 812 WSW 0 0 14 26 100 354 494 W 0 115 20 11 23 116 285 WNW 0 0 6 11 37 171 225 NW 0 0 67 8 53 1059 1187 NNW 0 168 162 56 70 1179 1635 TOTAL 34 331 476 549 1111 25359 27860

BFN-18 TABLE 2.2-4 (Sheet 4) Historical Information BROWNS FERRY NUCLEAR PLANT 1990 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE 0-1 1-2 2-3 3-4 4-5 5-10 TOTAL N 38 0 0 82 133 1338 1592 NNE 0 12 23 155 54 1516 1760 NE 0 3 23 44 123 4560 4753 ENE 0 38 86 102 44 1750 2019 E 0 0 23 7 32 1586 1647 ESE 0 0 12 0 0 123 135 SE 0 0 0 0 0 8435 8435 SSE 0 0 0 0 0 2483 2483 S 0 0 10 12 46 1525 1593 SSW 0 0 33 40 188 1345 1606 SW 0 0 0 15 111 847 973 WSW 0 0 10 19 73 244 346 W 0 84 15 8 17 79 203 WNW 0 0 4 8 27 36 75 NW 0 0 76 9 60 826 970 NNW 0 189 183 63 79 1381 1895 TOTAL 38 328 496 563 987 28073 30485

BFN-18 TABLE 2.2-4 (Sheet 5) Historical Information BROWNS FERRY NUCLEAR PLANT 2000 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE 0-1 1-2 2-3 3-4 4-5 5-10 TOTAL N 40 0 0 85 137 1382 1644 NNE 0 13 23 160 56 1565 1817 NE 0 3 23 45 127 4708 4908 ENE 0 40 89 105 45 1807 2085 E 0 0 23 7 33 1638 1701 ESE 0 0 13 0 0 127 140 SE 0 0 0 0 0 8777 8777 SSE 0 0 0 0 0 2584 2584 S 0 0 10 12 45 1520 1587 SSW 0 0 32 39 184 1315 1570 SW 0 0 0 15 108 828 951 WSW 0 0 10 19 71 238 337 W 0 82 15 8 17 76 198 WNW 0 0 4 8 26 37 75 NW 0 0 78 9 62 853 1002 NNW 0 196 189 65 82 1426 1957 TOTAL 40 334 508 576 993 28881 31332

BFN-18 TABLE 2.2-4 (Sheet 6) Historical Information BROWNS FERRY NUCLEAR PLANT 2010 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE 0-1 1-2 2-3 3-4 4-5 5-10 TOTAL N 41 0 0 87 141 1421 1691 NNE 0 13 24 164 58 1610 1869 NE 0 4 24 47 131 4843 5048 ENE 0 41 91 108 47 1858 2145 E 0 0 24 7 34 1684 1749 ESE 0 0 13 0 0 131 144 SE 0 0 0 0 0 9081 9081 SSE 0 0 0 0 0 2673 2673 S 0 0 10 12 45 1532 1598 SSW 0 0 32 39 183 1307 1560 SW 0 0 0 15 108 822 944 WSW 0 0 10 18 71 236 335 W 0 82 15 8 17 76 197 WNW 0 0 4 8 26 38 76 NW 0 0 80 10 63 877 1030 NNW 0 201 194 67 84 1467 2013 TOTAL 41 340 520 589 1006 29657 32153

BFN-18 TABLE 2.2-4 (Sheet 7) Historical Information BROWNS FERRY NUCLEAR PLANT 2020 POPULATION DISTRIBUTION WITHIN 10 MILES OF THE SITE 0-1 1-2 2-3 3-4 4-5 5-10 TOTAL N 42 0 0 91 147 1475 1755 NNE 0 14 25 170 60 1672 1940 NE 0 4 25 49 136 5028 5241 ENE 0 42 95 112 49 1930 2227 E 0 0 25 7 35 1749 1816 ESE 0 0 14 0 0 136 149 SE 0 0 0 0 0 9452 9452 SSE 0 0 0 0 0 2782 2782 S 0 0 10 12 46 1574 1641 SSW 0 0 33 40 186 1332 1591 SW 0 0 0 15 110 838 963 WSW 0 0 10 19 72 241 342 W 0 83 15 8 17 78 201 WNW 0 0 4 8 27 39 78 NW 0 0 83 10 66 911 1070 NNW 0 209 202 70 87 1524 2091 TOTAL 42 352 539 610 1036 30760 33339

BFN-18 Table 2.2-5 (Sheet 1) Historical Information BROWNS FERRY 1970 POPULATION DISTRIBUTION WITHIN 60 MILES OF THE SITE Total Pop 0-10 10-20 20-30 30-40 40-50 50-60 N 41,285 855 1,515 2,615 10,660 3,690 21,950 NNE 13,500 1,055 2,990 2,230 3,125 3,420 680 NE 47,980 4,335 14,180 6,625 5,385 12,625 4,830 ENE 39,115 1,485 4,990 9,615 13,860 5,425 3,740 E 159,800 1,025 1,910 73,405 75,125 4,610 3,725 ESE 40,045 170 1,880 2,535 7,465 9,575 18,420 SE 76,230 10,400 30,945 4,680 6,230 13,850 10,125 SSE 60,505 1,680 6,250 11,630 15,175 18,945 6,825 S 25,535 1,395 3,805 1,800 4,475 3,730 10,330 SSW 21,100 1,155 5,895 1,270 1,490 2,535 8,755 SW 23,825 830 2,970 2,280 2,725 10,675 4,345 WSW 25,685 480 3,060 3,005 11,545 3,755 3,840 W 53,150 685 2,960 6,830 35,070 4,785 2,820 WNW 57,415 130 885 9,300 39,875 5,545 1,680 NW 24,970 350 4,345 5,215 5,485 3,260 6,315 NNW 26,670 710 2,090 2,440 12,350 7,360 1,720 Total 736,810 26,740 90,670 145,475 250,040 113,785 110,100

BFN-18 Table 2.2-5 (Sheet 2) Historical Information BROWNS FERRY 1980 POPULATION DISTRIBUTION WITHIN 60 MILES OF THE SITE Total Pop. 0-10 10-20 20-30 30-40 40-50 50-60 N 46,435 820 1,635 2,605 11,460 5,970 23,945 NNE 12,515 1,010 3,900 2,385 3,255 1,265 700 NE 55,670 5,175 17,740 8,065 6,440 13,330 4,920 ENE 46,970 2,060 6,065 12,225 17,105 5,670 3,845 E 168,185 1,435 2,145 77,390 77,680 4,735 4,800 ESE 48,460 235 2,240 2,665 8,265 12,095 22,960 SE 91,230 11,245 34,400 5,045 6,645 18,935 14,960 SSE 71,140 2,520 9,925 12,855 18,595 22,305 4,940 S 40,510 1,455 4,010 1,900 5,565 9,660 17,920 SSW 23,055 1,220 6,145 1,305 1,840 2,700 9,865 SW 23,075 860 3,140 2,265 3,150 8,760 4,900 WSW 28,390 495 3,115 3,080 13,650 3,770 4,280 W 59,880 685 2,940 7,750 40,615 4,985 2,905 WNW 63,690 130 905 10,500 44,435 6,025 1,695 NW 26,735 345 4,975 6,055 5,905 3,620 5,835 NNW 32,100 685 2,290 2,475 15,260 8,155 3,235 Total 838,040 30,355 105,570 158,565 279,865 131,980 131,705

BFN-18 TABLE 2.2-5 (Sheet 3) Historical Information BROWNS FERRY NUCLEAR PLANT 1986 60-MILE POPULATION DISTRIBUTION 0-10 10-20 20-30 30-40 40-50 50-60 TOTAL N 1360 3050 3241 11152 4028 12052 34883 NNE 1557 2680 4420 3714 4362 15419 32152 NE 4430 13573 7600 7231 14001 4606 51441 ENE 1854 10211 14149 16807 8139 5603 56763 E 1462 2360 88100 75567 7899 8881 184269 ESE 117 1972 4357 12995 11469 27572 58482 SE 6960 31383 7856 8331 21430 10150 86110 SSE 2314 18602 13747 21140 19700 10945 86448 S 1502 5186 2493 6344 4899 13509 33933 SSW 1666 6657 1134 2332 4429 9901 26119 SW 812 5028 2456 2938 17158 4881 33273 WSW 494 2761 2844 14925 2644 3306 26974 W 285 4170 10452 31575 5607 1678 53767 WNW 225 2190 7360 51760 6844 3313 71692 NW 1187 4350 6957 6368 4204 5401 28467 NNW 1635 2465 2819 15156 9731 4096 35902 TOTAL 27860 116638 179985 288335 146544 141313 900675

BFN-18 TABLE 2.2-5 (Sheet 4) Historical Information BROWNS FERRY NUCLEAR PLANT 1990 60-MILE POPULATION DISTRIBUTION 0-10 10-20 20-30 30-40 40-50 50-60 TOTAL N 1591 2887 3172 11844 4190 13165 36850 NNE 1760 2756 4551 3827 4236 12820 29950 NE 4753 13356 7431 7554 13972 4304 51370 ENE 2019 9232 30963 18076 7866 5086 73243 E 1647 15764 48934 90006 6583 9228 172161 ESE 135 2466 5115 26482 11164 24667 70029 SE 8435 22086 7344 8313 21440 13146 80764 SSE 2483 30555 13272 21617 20244 12218 100389 S 1592 4758 2533 6218 5550 18001 38652 SSW 1606 6369 1160 2570 4703 10937 27346 SW 971 5719 2470 3460 13138 4422 30179 WSW 346 3461 2895 13472 4772 4570 29516 W 203 3761 10877 29784 5845 2533 53003 WNW 76 1970 7936 49335 5307 3493 68117 NW 970 4145 7270 6059 4242 5044 27730 NNW 1896 2332 2958 15381 10031 4023 36620 TOTAL 30483 131617 158882 313999 143281 147657 925919

BFN-18 TABLE 2.2-5 (Sheet 5) Historical Information BROWNS FERRY NUCLEAR PLANT 2000 60-MILE POPULATION DISTRIBUTION 0-10 10-20 20-30 30-40 40-50 50-60 TOTAL N 1643 2981 3196 11933 4249 13976 37979 NNE 1817 2846 4621 3882 4441 13730 31338 NE 4908 13791 7681 7861 14539 4393 53173 ENE 2085 9533 32224 18816 8182 5404 76244 E 1701 16390 50939 93693 6788 9386 178897 ESE 140 2549 5323 27594 11760 26153 73520 SE 8777 22981 7642 8618 22346 13553 83916 SSE 2584 31794 13802 22132 20718 12446 103475 S 1587 4762 2610 6332 5658 18430 39380 SSW 1570 6226 1143 2596 4756 11180 27470 SW 949 5590 2414 3454 13338 4578 30324 WSW 339 3383 2849 13329 4720 4584 29202 W 198 3687 10801 29581 5813 2547 52627 WNW 76 1960 7951 49491 5368 3570 68416 NW 1002 4158 7301 6275 4530 5394 28659 NNW 1957 2357 3051 16033 10460 4209 38068 TOTAL 31331 134988 163549 321621 147664 153533 952687

BFN-18 TABLE 2.2-5 (Sheet 6) Historical Information BROWNS FERRY NUCLEAR PLANT 2010 60-MILE POPULATION DISTRIBUTION 0-10 10-20 20-30 30-40 40-50 50-60 TOTAL N 1690 3067 3213 11992 4292 14585 38838 NNE 1869 2927 4679 3926 4613 14480 32494 NE 5048 14186 7887 8121 15036 4473 54750 ENE 2145 9806 33156 19367 8428 5646 78548 E 1749 16865 52414 96405 6970 9577 183980 ESE 144 2622 5498 28444 12246 27407 76362 SE 9081 23776 7906 8910 23206 13954 86834 SSE 2673 32894 14275 22774 21316 12728 106660 S 1598 4808 2687 6488 5799 18872 40251 SSW 1559 6185 1142 2641 4841 11430 27799 SW 942 5554 2398 3481 13600 4726 30701 WSW 336 3361 2848 13353 4724 4630 29252 W 197 3676 10926 29930 5882 2582 53193 WNW 76 1977 8059 50177 5472 3671 69433 NW 1030 4215 7407 6465 4768 5683 29568 NNW 2013 2398 3132 16556 10805 4328 39233 TOTAL 32152 138316 167628 329031 151997 158771 977895

BFN-18 TABLE 2.2-5 (Sheet 7) Historical Information BROWNS FERRY NUCLEAR PLANT 2020 60-MILE POPULATION DISTRIBUTION 0-10 10-20 20-30 30-40 40-50 50-60 TOTAL N 1755 3184 3277 12228 4384 15115 39942 NNE 1941 3039 4800 4012 4766 15065 33623 NE 5241 14728 8165 8375 15482 4593 56585 ENE 2227 10181 34388 20078 8721 5855 81449 E 1816 17494 54360 99985 7220 9885 190758 ESE 149 2723 5716 29533 12789 28730 79641 SE 9452 24747 8229 9269 24197 14501 90395 SSE 2782 34237 14856 23606 22094 13174 110750 S 1640 4942 2788 6710 6000 19541 41621 SSW 1590 6306 1168 2721 4990 11826 28601 SW 961 5662 2445 3564 14036 4918 31587 WSW 343 3426 2910 13618 4813 4748 29857 W 201 3755 11242 30796 6054 2656 54703 WNW 78 2032 8306 51728 5647 3787 71578 NW 1070 4344 7635 6661 4948 5900 30559 NNW 2090 2477 3222 17035 11119 4483 40426 TOTAL 33337 143277 173506 339920 157260 164776 1012076

BFN-18 TABLE 2.2-6 Historical Information BROWNS FERRY NUCLEAR PLANT POPULATION DENSITY AT VARIOUS DISTANCES FROM THE SITE DISTANCE 1970 1980 1986 1990 2000 2010 2020 FROM REACTOR LAND AREA -------------------------------------- ---------------------------------------- --------------------------------------- -------------------------------------- ------------------------------------- ------------------------------------- ------------------------------------- BUILDING (SQUARE AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE AVERAGE (MILES) MILES) POPULATION DENSITY POPULATION DENSITY POPULATION DENSITY POPULATION DENSITY POPULATION DENSITY POPULATION DENSITY POPULATION DENSITY 0-1 0.2 --- --- --- --- 34 170.0 38 190.0 40 200.0 41 205.0 42 211.5 1-2 3.5 40 11.4 45 12.9 331 94.6 328 93.7 334 95.4 340 97.1 352 100.5 2-3 11.3 375 33.2 420 37.2 476 42.1 496 43.9 508 45.0 520 46.0 539 47.7 3-4 16.8 410 24.4 435 25.9 549 32.7 563 33.5 576 34.3 589 35.1 610 36.3 4-5 23.2 1,000 43.1 1,045 45.0 1,111 47.9 987 42.5 993 42.8 1,006 43.4 1,036 44.6 0-5 55.0 1,825 33.2 1,945 35.4 2,501 45.5 2,412 43.9 2,451 44.6 2,496 45.4 2,579 46.9 5-10 210.5 24,915 118.4 28,410 135.0 25,359 120.5 28,073 133.4 28,881 137.2 29,657 140.9 30,760 146.1 0-10 265.5 26,740 100.7 30,355 114.3 27,860 104.9 30,483 114.8 31,332 118.0 32,153 121.1 33,339 125.6 10-20 903.0 90,670 100.4 105,570 116.9 116,638 129.2 131,617 145.8 134,988 149.5 138,316 153.2 143,277 158.7 20-30 1,560 145,475 94.5 158,565 103.0 179,985 116.9 158,882 101.8 163,549 104.8 167,628 107.5 173,506 111.2 30-40 2,180 250,040 114.7 279,865 128.4 288,335 132.3 313,999 144.0 321,621 147.5 329,031 150.9 339,920 155.9 40-50 2,790 113,785 40.8 131,980 47.3 146,544 52.5 143,281 51.4 147,664 52.9 151,997 54.5 164,776 59.1 50-60 3,400 110,100 32.4 131,705 38.7 141,313 41.6 147,657 43.4 153,533 45.2 158,771 46.7 157,260 46.3 0-60 11,078 736,810 66.5 838,040 75.6 900,675 81.3 925,919 83.6 163,549 14.8 977,895 88.3 1,012,076 91.4

BFN-18 TABLE 2.2-7 (Sheet 1) Historical Information 1986 PEAK HOUR RECREATION VISITATION WITHIN 10 MILES OF THE SITE Mile(s) From Site Total 0-1 1-2 2-3 3-4 4-5 5-10 N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 0 0 0 0 0 0 0 E 0 0 0 0 0 0 0 ESE 1,538 0 0 563 0 0 975 SE 285 0 0 0 0 0 285 SSE 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 WSW 568 0 0 568 0 0 0 W 441 0 0 441 0 0 0 WNW 305 0 0 0 0 0 305 NW 922 0 0 0 0 0 922 NNW 0 0 0 0 0 0 0 TOTAL 4,059 0 0 1,572 0 0 2,487

BFN-18 TABLE 2.2-7 (Sheet 2) Historical Information BROWNS FERRY NUCLEAR PLANT 1990 ESTIMATED PEAK HOUR RECREATION VISITATION WITHIN 10 MILES OF THE SITE Mile(s) From Site Total 0-1 1-2 2-3 3-4 4-5 5-10 N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 0 0 0 0 0 0 0 E 0 0 0 0 0 0 0 ESE 420 0 0 420 0 0 0 SE 1,009 0 0 0 0 0 1,009 SSE 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 WSW 454 0 0 454 0 0 0 W 353 0 0 353 0 0 0 WNW 244 0 0 0 0 0 244 NW 1,232 0 0 0 0 0 1,232 NNW 0 0 0 0 0 0 0 TOTAL 3,712 0 0 1,227 0 0 2,485

BFN-18 TABLE 2.2-7 (Sheet 3) Historical Information BROWNS FERRY NUCLEAR PLANT 2000 ESTIMATED PEAK HOUR RECREATION VISITATION WITHIN 10 MILES OF THE SITE Mile(s) From Site Total 0-1 1-2 2-3 3-4 4-5 5-10 N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 0 0 0 0 0 0 0 E 0 0 0 0 0 0 0 ESE 432 0 0 432 0 0 0 SE 1,038 0 0 0 0 0 1,038 SSE 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 WSW 467 0 0 467 0 0 0 W 363 0 0 363 0 0 0 WNW 251 0 0 0 0 0 251 NW 1,267 0 0 0 0 0 1,267 NNW 0 0 0 0 0 0 0 TOTAL 3,818 0 0 1,262 0 0 2,556

BFN-18 TABLE 2.2-7 (Sheet 4) Historical Information BROWNS FERRY NUCLEAR PLANT 2010 ESTIMATED PEAK HOUR RECREATION VISITATION WITHIN 10 MILES OF THE SITE Mile(s) From Site Total 0-1 1-2 2-3 3-4 4-5 5-10 N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 0 0 0 0 0 0 0 E 0 0 0 0 0 0 0 ESE 443 0 0 443 0 0 0 SE 1,065 0 0 0 0 0 1,065 SSE 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 WSW 479 0 0 479 0 0 0 W 372 0 0 372 0 0 0 WNW 257 0 0 0 0 0 257 NW 1,299 0 0 0 0 0 1,299 NNW 0 0 0 0 0 0 0 TOTAL 3,915 0 0 1,294 0 0 2,621

BFN-18 TABLE 2.2-7 (Sheet 5) Historical Information BROWNS FERRY NUCLEAR PLANT 2020 ESTIMATED PEAK HOUR RECREATION VISITATION WITHIN 10 MILES OF THE SITE Mile(s) From Site Total 0-1 1-2 2-3 3-4 4-5 5-10 N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 0 0 0 0 0 0 0 E 0 0 0 0 0 0 0 ESE 458 0 0 458 0 0 0 SE 1,102 0 0 0 0 0 1,102 SSE 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 WSW 496 0 0 496 0 0 0 W 385 0 0 385 0 0 0 WNW 266 0 0 0 0 0 266 NW 1,344 0 0 0 0 0 1,344 NNW 0 0 0 0 0 0 0 TOTAL 4,051 0 0 1,339 0 0 2,712

BFN-18 TABLE 2.2-8 Historical Information Browns Ferry Nuclear Plant Listing of 1993 School Enrollments and Industrial Employment Within Ten Miles of the Site Distance from School Site (Mi) Enrollment Alternative School1 8.4 71 Clements High 7.7 963 Hubbard Elementary 6.0 237 James L. Cowart Elementary2 9.8 348 Leon Sheffield Elementary3 9.1 272 Reid Elementary 5.7 147 Tanner High4 8.4 775 Tennessee Valley School 6.6 122 West Decatur Elementary 9.2 261 West Lawn Elementary 9.0 203 West Morgan Elementary 8.5 950 Industry Employment5 American Fructose 6.4 200 Amoco 6.0 450 Diakan 5.5 100 Freuhauf 7.4 150 General Motors Plant 10.0 1,717 Hispan 5.5 100 Monsanto 7.4 800 Polysar 7.7 80 Prestolite 7.4 300 Steel Case Plant 9.3 375 3M Plant 5.5 450 Whittaker 7.5 150 1 Formerly Tanner Primary facility. 2 Name change - formerly West Athens Elementary. 3 Name change - formerly Lakeview Elementary. 4 Includes grades K-12. 5 Day shift.

BFN-18 TABLE 2.2-9 (Sheet 1) Historical Information BROWNS FERRY NUCLEAR PLANT STATISTICAL DATA FOR NEARBY COUNTIES 1990 EMPLOYMENT BY COUNTY(1) LAWRENCE LIMESTONE MADISON MORGAN TOTAL EMPLOYMENT 10,488 26,899 162,673 52,401 WAGE AND SALARY 8,138 22,920 147,430 44,778 PROPRIETORS 2,350 3,979 15,243 7,623 FARM 1,232 1,198 1,106 1,363 NONFARM 1,118 2,781 14,137 6,260 TOTAL EMPLOYMENT BY INDUSTRY: FARM 1,492 1,591 1,630 1,575 NONFARM 8,996 25,308 161,043 50,826 PRIVATE 7,065 17,402 123,082 43,353 AG.SERV.,FOR.,FISH., AND OTHER 161 165 708 321 MINING (D) (L) 131 140 CONSTRUCTION 898 1,461 7,252 4,329 MANUFACTURING 2,387 6,923 34,270 13,666 TRANSPORTATION AND PUBLIC UTILITIES 176 371 3,705 1,803 WHOLESALE TRADE (D) 670 5,268 1,929 RETAIL TRADE 1,020 3,643 23,097 8,605 FINANCE, INSURANCE, AND REAL ESTATE 157 532 7,093 2,642 SERVICES 2,054 3,634 41,558 9,918 GOVERNMENT AND GOVERNMENT ENTERPRISES 1,931 7,906 37,961 7,473 FEDERAL, CIVILIAN 141 3,861 17,211 310 MILITARY 316 542 5,795 1,045 STATE AND LOCAL 1,474 3,503 14,955 6,118

1. BY PLACE OF WORK D- WITHHELD TO AVOID DISCLOSURE OF INFORMATION ABOUT INDIVIDUAL FIRMS L- LESS THAN 10 SOURCE: U.S. DEPARTMENT OF COMMERCE, BUREAU OF ECONOMIC ANALYSIS, TABLE CA25 FOR ALABAMA

BFN-18 TABLE 2.2-9 (Continued) (Sheet 2) Historical Information BROWNS FERRY NUCLEAR PLANT STATISTICAL DATA FOR NEARBY COUNTIES MORGAN MADISON LIMESTONE LAWRENCE Agricultural Use--1987 1/ Total farmland (acres) 159,757 235,478 223,190 188,365 Number of farms 1,243 977 1,090 1,123 Cropland harvested 44,763 112,625 101,416 77,593 Value of Products Sold - 1987 (dollars) 2/ (All Farms) Crops, including nursery and greenhouse crops 6,429,600 23,088,000 23,250,000 16,453,000 Poultry and poultry products 27,676,000 5,271,000 993,000 27,328,000 Dairy products 3,865,000 2,349,000 2,639,000 936,000 Other livestock 8,878,000 11,356,000 9,333,000 15,577,000 Total 46,848,600 42,064,000 36,215,000 60,294,000 Livestock and Poultry on Farms - 1987 (number) 1/ Cattle and calves 35,979 25,471 25,636 27,776 Milk cows 2,235 1,164 1,417 570 Sheep and lambs (D) 207 (D) 466 Hogs and pigs 2,706 7,989 5,727 6,835 Chickens (3 months old and older) 280,865 (D) (D) 1,156,297 1/ Source = 1987 Census of Agriculture 2/ Source = Alabama Agricultural Statistics Service 1987 (D) - Withheld to avoid disclosing data for individual farms

BFN-18 TABLE 2.2-10 (Sheet 1) HAZARDOUS RIVER TRAFFIC THAT PASSES BROWNS FERRY NUCLEAR PLANT 1983-1993 (TONS) U.S. Army Corps of Engineers Data Old New Commodity Commodity Codes(1) Codes(1) CODE COMMODITY 1983 1984 1985 1986 1987 1988 1989 1990 1990 1991 1992 1993 2871 Nitrogenous Fert. 38,712 47,301 71,826 28,507 59,722 30,155 35,961 41,707 NA NA NA NA 56211 Amon Nitrate Fert. NA NA NA NA NA NA NA NA 11,742 4,456 4,693 4,666 56216 Urea Fert. NA NA NA NA NA NA NA NA 15,165 46,696 33,706 25,871 56219 Min., Chem. NA NA NA NA NA NA NA NA 14,800 16,903 6,917 13,216 Fert., Nitrogenous 2911 Gasoline 0 1,283 0 3,287(2) 0 4,900 0 0 NA NA NA NA 33411 Gasoline NA NA NA NA NA NA NA NA 0 0 0 0 2914 Distillate Fuel Oil 25,545 0 0 7,374 29,037 3,385 8,054 16,746 NA NA NA NA 33419 Oth Lt. Oils NA NA NA NA NA NA NA NA 16,746 74,983 7,249 0 From Petroleum 2915 Residual Fuel Oil 113,737 147,018 153,938 329,305 295,813 332,553 292,931 230,255 NA NA NA NA 33440 Fuel Oils, NEC NA NA NA NA NA NA NA NA 230,255 133,642 235,508 215,461 2917 Naphtha 38,295 63,870 79,690 110,349 115,263 128,011 143,450 87,586 NA NA NA NA 33429 Oth Medium Oils NA NA NA NA NA NA NA NA 87,586 8,022 8,887 5,982 2811 Crude Pds. from 94,594 46,241 175,255 105,100 98,894 222,789 167,272 335,539 NA NA NA NA Coal Tar, Petro 33521 Tar NA NA NA NA NA NA NA NA 18,158 21,878 3,079 1,400 33525 Oils & Other Prod NA NA NA NA NA NA NA NA 8,626 2,878 3,654 5,965 Dist of Coal 51124 Xylene, Pure NA NA NA NA NA NA NA NA 245,962 137,597 190,582 276,854 51125 Styrene NA NA NA NA NA NA NA NA 58,624 95,008 109,472 164,200 51127 Cumene NA NA NA NA NA NA NA NA 0 0 1,408 0

BFN-18 TABLE 2.2-10 (Sheet 2) HAZARDOUS RIVER TRAFFIC THAT PASSES BROWNS FERRY NUCLEAR PLANT 1983-1993 (TONS) U.S. Army Corps of Engineers Data Old New Commodity Commodity Codes(1) Codes(1) CODE COMMODITY 1983 1984 1985 1986 1987 1988 1989 1990 1990 1991 1992 1993 51133 Tetreclorethlene NA NA NA NA NA NA NA NA 4169 1399 0 0 2813 Alcohols 0 0 0 1,400 33,780 24,173 22,352 16,478 NA NA NA NA 51215 Ethylal alchol NA NA NA NA NA NA NA NA 16,478 11,127 9,487 8,511 51216 Ethyl Alcohol NA NA NA NA NA NA NA NA 0 5,563 0 0 Denatured 2817 Benzene & Toluene 0 0 16,968 0 0 2,779 0 13,812 NA NA NA NA 33522 Benzole NA NA NA NA NA NA NA NA 12,412 42,220 0 6,128 51122 Benzene NA NA NA NA NA NA NA NA 1,400 0 0 6,731 2819 Basic Chemicals 237,821 341,032 331,028 342,246 384,343 458,767 398,525 283,910 NA NA NA NA 51372 Esters of NA NA NA NA NA NA NA NA 11,861 32,440 41,530 30,570 Acetic Acid 51483 Acrylonitrile NA NA NA NA NA NA NA NA 211,094 302,644 330,668 349,172 51484 Nitrile Function NA NA NA NA NA NA NA NA 2,649 67,159 68,236 67,333 Compounds, NEC 52210 Carbon, NEC NA NA NA NA NA NA NA NA 0 0 2,869 2,947 52224 Chlorine NA NA NA NA NA NA NA NA 46,200 34,100 38,500 36,300 52349 Oth. Sulfates/ Alum NA NA NA NA NA NA NA NA 9,406 3,265 1,423 1,756 52264 Potass.Hydroxide NA NA NA NA NA NA NA NA 2,700 0 0 0 TOTAL 548,704 646,745 828,705 927,568 1,016,852 1,207,512 1,068,545 1,026,033 1,026,033 1,041,980 1,097,868 1,223,063 (1) In 1990, more detailed and specific commodity codes became available. The double entries for that year indicate the commodities under both the old and new system. (2) The actual product was RU250.

BFN-18 Table 2.2-11 (Deleted by Amendment 12)

10 laeeee 0 . llftNS FERRY NUCL!AR PLA.MT FlMAL:SAPETY ANALYSIS REPORT C. lroWfll Ferry ... lAalCion Map.

                      ,      a,MileRaclu FIGURE 2.2*1 AMENDMENT 18                  HISTORICAL

BROWNS PERRY MU'CLEAR Pi.JANT flM~ SAPETY ANALYSIS REPORT Browm Fany Site Location Mllp - 10 Mlle R*dlus FIGURE2.2-2 AMENDMENT 18 HISTQRICAL

BFN-16 Figure 2.2-3 has been deleted.

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 2.2-4

BFN-25 2.3 METEOROLOGY 2.3.1 General The Browns Ferry site is adjacent to the Tennessee River (Wheeler Lake) which flows northwest at this location. There are no local physiographical features to cause significant climatological anomalies at the site, as the immediate terrain is flat or slightly undulating, with scattered 400- to 600-foot foothills and ridges located 20 to 25 miles to the east through south and southwest. Wheeler Lake adjoins the site and averages 1 to 1-1/2 miles in width. Normally, discontinuities in ambient thermal structure from differential surface heating between land and water should not cause detectable lake breeze circulation at the site area.1 Limited air mass modification may occur within the lower few hundred feet, particularly with southeast winds, when the over-water trajectory may approach 10 miles. 2.3.2 Climatology The site is in a temperate latitude about 300 miles north of the Gulf of Mexico. The area is dominated in winter and spring by alternating cool, dry continental air from the north and warm, moist maritime air from the south. During this period, migratory cyclonic disturbances with numerous thundershowers and thunderstorms pass through the area, with frequent precipitation. Storms, including tornadoes, reach strongest intensity in March and April when maximum air mass contrast generally occurs. Persistent and unstable maritime air in the summer results in frequent thundershower and thunderstorm activity. Stagnating anticyclones sometimes dominate the area in the fall, with extended periods of low wind speeds and poor dispersion conditions. Climatological appraisal of the Browns Ferry site is based primarily upon data collected at (1) the plant site during the three-year period, January 1, 1977 through December 31, 1979, (2) Volunteer Weather Observation Stations in Decatur, Alabama, about 13 miles southeast of the Browns Ferry site, for the 80-year period, 1879-1958, and (3) The National Weather Service Station near Huntsville, Alabama, about 20 miles east of the site, for the 13-year period, 1968-1980. Climatological features affecting the atmospheric dispersion of plant emissions are discussed later in this subsection. 2.3.3 Atmospheric Stability Temperature data from the Browns Ferry meteorological tower were used to determine atmospheric stability. The Pasquill seven category (A-G) classification 2.3-1

BFN-25 scheme, based on temperature change with height, was used. Table 2.3-1 gives the percent occurrences of the Pasquill stabilities classified by lapse rates between 33 and 150 feet (10 and 45 meters), and based on wind speeds at 33 feet (10 meters). Table 2.3-2 gives the percent occurrences of the stabilities classified by lapse rates between 150 and 300 feet (45 and 90 meters), and based on wind speeds at 300 feet (93 meters). For the lower layer (Table 2.3-1), Classes D and E each occurred about 32 percent of the time. Unstable conditions (Classes A, B, and C) occurred least often, only about l5 percent of the time. In the upper layer (Table 2.3-2), Classes D and E combined occurred about 88 percent of the time. Unstable classes were much less frequent, occurring less than one percent of the time. During the three-year period of record, low-level inversion conditions (temperature increases with height) occurred during about 38 percent of the total hours. This compares well with an inversion frequency of about 37 percent obtained from a study by Hosler2 of two years of data from selected NWS stations. Seasonally, the greatest occurrence of inversion conditions is during the fall. 2.3.4 Wind Wind data from the Browns Ferry meteorological tower were used to represent airflow at the site. Tables 2.3-3 through 2.3-9 are joint percentage frequency distributions of wind speed by wind direction for Pasquill stability classes A-G, respectively. These are classified by lapse rates between 33 and 150 feet and based on wind data at 33 feet. Table 2.3-10 is a distribution of wind speed by wind direction based on wind data at 33 feet, disregarding stability class. Tables 2.3-11 through 2.3-22 are monthly distributions of wind data at 33 feet, based on the three-year period. A corresponding set of tables for lapse rates between 150 and 300 feet, and wind data at 300 feet, is given in Tables 2.3-23 through 2.3-42. 2.3.4.1 Wind Direction Lower Level From Tables 2.3-3 through 2.3-10, it can be seen that the highest frequency of winds at the 33-foot level was generally from the southeast sector. The only exception was under Class G stability conditions when the highest frequency was from the north-northeast sector. The monthly distributions (Tables 2.3-11 through 2.3-22) show a similar high frequency of winds from the southeast sector. Exceptions are the distributions for 2.3-2

BFN-25 January, February, and September, which reveal a high frequency of winds from the north-northwest, north-northeast, and north-northeast sectors, respectively. Annual and monthly wind direction patterns at the 33-foot level are shown in wind rose plots, Figures 2.3-1 through 2.3-20. Upper Level At the 300-foot wind sensor level, the distributions (Tables 2.3-23 through 2.3-30) reveal that the maximum frequency of winds is generally from the southeast sector. Under stabilities A, B, and C, the highest frequencies of winds were from the west or southwest sector, but these combined frequencies occurred less than one percent of the total time. For Class G stability conditions, the highest frequency was from the south-southeast sector. The monthly distributions (Tables 2.3-31 through 2.3-42) show a high frequency of winds from the southeast and south-southeast sectors. Exceptions are the distributions for January and February, which reveal a maximum frequency of winds from the northwest and north sectors, respectively. Annual and monthly wind direction patterns at the 300-foot level are given in wind rose plots, Figures 2.3-21 through 2.3-40. 2.3.4.2 Wind Direction Persistence Persistent wind is defined in this analysis as a continuous wind from one of the 22-1/2 degree sectors (e.g., north-northeast). The persistence is not considered to be interrupted if the wind departs from the sector for one hour and then returns, or if there are up to two hours of missing data followed by a continuation of the same directional persistence. Tables 2.3-43 and 2.3-44 are summaries of the wind direction persistence durations at the 33 and 300-foot levels, respectively. From these tables, it can be seen that the wind directions which were the most persistent were from the southeast and south-southeast sectors. At the lower level, about 20 percent of the persistence cases were equal to 6 hours or more, about 4 percent were equal to 12 hours or more, and about 0.2 percent were equal to 24 hours or more. The four highest persistence durations at this level were 36, 33, 32, and 32 hours with southeast, southeast, south, and north-northwest winds, respectively. At the upper level, about 22 percent of the persistence cases were equal to 6 hours or more, about 4 percent were equal to 12 hours or more, and about 0.4 percent 2.3-3

BFN-25 were equal to 24 hours or more. The four highest persistence durations at the 300-foot level were 38, 36, 33, and 32 hours with south-southeast, southeast, southeast and south winds, respectively. 2.3.4.3 Wind Speed Lower Level Average wind speeds at the 33-foot level under different stability conditions (Tables 2.3-3 through 2.3-9) ranged from a low of 3.2 mph under G stability to a high of 7.2 mph under B stability. The highest occurrence of wind speeds was in the 3.5-5.4 mph range under unstable and neutral conditions (Classes A, B, C, and D), and in the 1.5-3.4 mph range under stable conditions (Classes E, F, and G). The overall mean wind speed, disregarding stability, was 5.7 mph (Table 2.3-10). The highest occurrence of wind speeds for this distribution was in the 3.5-5.4 mph range with winds primarily from the southeast sector. From the monthly distributions (Tables 2.3-11 through 2.3-22), it is observed that the highest average wind speeds occurred in the winter and early spring (December through March) and the lowest values occurred in late spring to late fall (May through October). The highest occurrence of wind speeds was in the 7.5-12.4 mph range for January; in the 1.5-3.4 mph range for May, June, and November; and in the 3.5-5.4 mph range for the remaining months. Annual and monthly wind speed patterns at the 33-foot level are also shown in wind rose plots, Figures 2.3-1 through 2.3-20. Upper Level Average wind speeds at the 300-foot level, under different stability conditions (Tables 2.3-23 through 2.3-29), ranged from a low of 9.5 mph under A stability to a high of 12.1 mph under D and F stabilities. The highest occurrence of wind speeds was in the 5.5-7.4 mph range under A stability, in the 7.5-12.4 mph range under C, D, and G stabilities, and in the 12.5-18.4 mph range under B, E, and F stabilities. The overall mean wind speed, disregarding stability, was 12.0 mph (Table 2.3-30). The highest occurrence of wind speeds for this distribution was in the 7.5-12.4 mph range. From the monthly distributions (Tables 2.3-31 through 2.3-42), it is observed that the highest average wind speeds occurred in the winter and early spring (December through March) and the lowest values occurred in the late spring and summer (May through August). The highest occurrence of wind speeds was in the 7.5-12.4 mph range for May through September, and in the 12.5-18.4 mph range for the remaining months. 2.3-4

BFN-25 Annual and monthly wind speed patterns at the 300-foot level are also shown in wind rose plots, Figures 2.3-21 through 2.3-40. A detailed discussion of meteorological diffusion evaluation methods is presented in Section 14.8.5.2. 2.3.5 Temperature and Precipitation 2.3.5.1 Temperature The climate at the Browns Ferry site is interchangeably continental and maritime in winter and spring, predominantly maritime in summer, and generally continental in fall. The mean annual temperature at Decatur3, Alabama, during 1879-1958, was 62.0°F. In a typical year at Decatur, there are about 70 days with maximum temperatures equal to or greater than 90°F and about 57 days with minimum temperatures equal to or less than 32°F. The most extreme daily temperatures recorded during this 80-year period occurred in June 1914 (108°F) and in February 1899 (-12°F). Temperature statistics for Decatur (period of record 1879-1958) are given in Table 2.3-45. Table 2.3-46 gives temperature statistics for the Browns Ferry meteorological facility for January 1, 1977 - December 31, 1979. Because of the longer period of record, the Decatur data are considered more representative of normal temperatures in the area. 2.3.5.2 Precipitation Much of the annual precipitation at the Browns Ferry site results from migratory storms in the winter and early spring (December through April).4 Most of the remaining precipitation is in June and July when air mass thundershower activity is common. October usually has the lowest precipitation. The maximum 1-hour rainfall which may be representative for the Browns Ferry site area was 2.12 inches, recorded in Moulton,5 20 miles southwest of the site, for the 11-year period, 1940-50. The maximum 1-hour rainfall for a 100-year frequency is 3.3 inches.6 The site underground storm drainage system is designed for a maximum rainfall of 4 inches per hour. Precipitation statistics from Huntsville for the 13-year period, 1968-1980, and normals for 1941-1970 are given in Table 2.3-47. These data are considered more representative of the normal than precipitation data from the Browns Ferry 2.3-5

BFN-25 meteorological facility for January 1, 1977 - December 31, 1979, given in Table 2.3-48. 2.3.5.3 Snowfall Snow does not often occur at the Browns Ferry site and seldom accumulates on the ground for more than a few days. Decatur3 snowfall data in Table 2.3-49 are considered representative of the site area. The maximum 24-hour snowfall7 reported was 17.1 inches in December 1963; next highest were 10.1 and 10.0 inches in January 1940 and February 1895, respectively. 2.3.6 Storms 2.3.6.1 Thunderstorms During 1968-1980, there were about 57 days annually on which thunderstorms were reported in the Huntsville area.4 Thunderstorms occurred most frequently in July, August, June, and May. November and December had the smallest number of thunderstorms, with an average of one thunderstorm day each month. Windstorms (often associated with thunderstorms) may occur several times a year, particularly in winter, spring, and summer, with winds occasionally exceeding 40 mph. In 1964, 95 mph winds, with rain and hail, were reported at the Redstone Arsenal, 25 miles east-southeast of the site. Also in April 1958 and July 1963, winds were reported in excess of 70 mph in the Huntsville area. 2.3.6.2 Tornadoes There were four tornadoes reported in Limestone County during the 50-year period 1916-65.8, 9 In the adjacent counties, Morgan and Madison, 18 tornadoes were reported during the same period. The bordering Southern Tennessee counties, Giles and Lincoln, reported 13 and 5 tornadoes, respectively, during the 55-year period 1916-70.10 Tornado data compiled by the severe local storms (SELS) unit of the National Weather Service11 for the period 1955-67 were used for evaluating the tornado probability for the Browns Ferry site. In the 1-degree latitude/longitude square containing the site (about 3,930 mi2), there were 31 tornadoes reported during the 13-year period, or about 2.38 tornadoes per year. Thom's value12 for the mean tornado path area (2.82 mi2) was used in calculating an annual point probability for the site of 1.71 x 10-3; this is equivalent to a mean recurrence interval of about 600 years. The National Severe Storms Forecast Center in Kansas City, Missouri calculated the tornado return probability for the Browns Ferry site based on tornado occurrences 2.3-6

BFN-25 within a 30 nautical mile (nm) radius during 1950-1986.13 A circle of 30 nm radius has an area comparable to a one degree latitude-longitude square. Based on 48 tornado occurrences having path size estimates in the 37-year period, the return probability is 6.979x10-4 and the mean return interval is 1,433 years. The annual tornado occurrence in the 30 nm radius circle is 1.81 (based on 67 tornadoes reported). 2.3.7 Onsite Meteorological Measurement Program 2.3.7.1 Siting and Description of Instruments Collection of onsite meteorological data at the Browns Ferry Nuclear Plant commenced in February 1967 from a meteorological tower located about 0.5 mile north-northeast of the reactor building and about 25 feet above plant grade. This facility was moved in early 1970 to a new location approximately 0.7 mile north-northwest of the reactor building and about 10 feet above plant grade. In March 1973, the facility was moved to its present location about 0.5 mile east-southeast of the reactor building and about 30 feet above plant grade. The permanent meteorological facility consists of a 91-meter (300-foot) instrumented tower for wind and temperature measurements, a separate 10-meter (33-foot) tower for dewpoint measurements, a ground based instrument for rainfall measurements, and a data collection system in an instrument building (Environmental Data Station or EDS). The data collected include: wind speeds and directions at the 33-, 150-, and 300-foot levels (wind data collection at 150 feet began on April 23, 1976); temperatures at the 33-, 150-, and 300-foot levels (temperature data collection at four feet ended on May 24, 1979); and dewpoint temperatures at the 33-foot level (dewpoint data collection at 150 and 300 feet ended on March 6, 1978 and the 4-foot dewpoint data collection ended on November 15, 1978). The dewpoint sensor was moved to a separate tower on September 27, 1994. More exact heights for wind and temperature sensors are given in the EDS Manual.14 Rainfall is monitored from a rain gage located about 70 feet from the tower. The meteorological sensors are connected to the data collection and recording equipment in the EDS. A system of lightning and surge protection circuitry with proper grounding is included in the facility design. Instrument Description A description of the meteorological sensors follows. More detailed sensor specifications are included in the EDS manual. Replacement sensors, which may be of a different manufacturer or model, will satisfy Regulatory Guide 1.23 (Revision 1) specifications.15 2.3-7

BFN-30 SENSOR HEIGHT (feet) DESCRIPTION Wind Direction 33, 150, Ultrasonic Wind Sensor and Wind Speed and 300 Temperature 33, 150 Platinum Wire Resistance and 300 Detector (RTD) with aspirated radiation shield Dewpoint 33 Capacitive Humidity Sensor Rainfall 4 Tipping bucket rain gauge 2.3.7.2 Data Acquisition System The data acquisition system is located at the Environmental Data Station (EDS) and consists of meteorological sensors and various interface devices. These devices send meteorological data to the plant and to offsite archival systems. System Accuracies The meteorological data collection system is designed and replacement components are chosen to meet or exceed specifications for accuracy identified in NRC Regulatory Guide 1.23 (Revision 1). The meteorological data collection system satisfies the Regulatory Guide 1.23 accuracy requirements. A detailed listing of error sources for each parameter is included in the EDS manual. 2.3.7.3 Data Recording and Display The data acquisition is under control of the computer program. The output of each meteorological sensor is scanned periodically, scaled, and the data values are stored. Meteorological sensor outputs (except rainfall) are measured every five seconds (720 per hour). Rainfall is measured continuously as it occurs. Software data processing routines within the computer accumulate output and perform data 2.3-8

BFN-30 calculations to generate 15-minute and hourly averages of wind speed and temperature, 15-minute and hourly vector wind speed and direction, 15-minute and hourly total precipitation, hourly average dewpoint, and hourly horizontal wind direction sigmas. Prior to April 1987, a prevailing wind direction calculation method was used. Subsequently, vector wind speed and direction have been calculated along with arithmetic average wind speed. Prior to October 20, 2010, temperature and dewpoint were measured every minute (60 per hour). Selected data each 15 minutes and all data each hour are stored for remote data access. Data sent to the plant control room every 15 minutes includes 33-, 150-, and 300-foot wind direction, average wind speed, and temperature values. Hourly data sent includes all of these plus the precipitation amount. Meteorological data sent to the plant process computer includes 10, 46, and 91 meter wind direction, wind speed, and temperature values and atmospheric stability classification. Meteorological data sent to the CECC computer system by the plant process computer system includes calculated fifteen minute values for 10, 46, and 91 meter wind direction, wind speed, temperature values, and atmospheric stability classification to be used for Radiological Dose Assessment. This data is available from the CECC computer system to other TVA and the State emergency centers in support of the Radiological Emergency Plan, including the technical support center at Browns Ferry. In addition, automated delivery of meteorological data to the NRC is available through the CECC computer. Meteorological data are also sent from the EDS to offsite computer for validation. Data are sent from the EDS to an offsite computer for validation, reporting, and archiving. 2.3.7.4 Equipment Servicing, Maintenance, and Calibration The meteorological equipment at the EDS is serviced by either engineering aides, instrument technicians, or engineers. Maintenance and calibrations are performed by either instrument technicians, electrical engineering associates, or electrical engineers. Most equipment is calibrated or replaced at least every six months of service. The methods for maintaining a calibrated status for the components of the meteorological data collection system (sensors, recorders, electronics, DVM, data logger, etc.) include field checks, field calibration, and/or replacement intervals for individual components, on the basis of operational history of the component type. Procedures and processes such as appropriate maintenance processes (procedures, work order/work request documents, etc.) are used to calibrate and maintain meteorological and station equipment. 2.3-9

BFN-25 2.3.7.5 Operational Meteorological Program The operational phase of the meteorological program includes those procedures and responsibilities related to activities beginning with the initial fuel loading and continuing through the life of the plant. The meteorological data collection program is continuous without major interruptions. The meteorological program has been developed to be consistent with the guidance given in NRC Regulatory Guide 1.23 (Revision 0) and the reporting procedures in Regulatory Guide 1.21 (Revision 1).18 The basic objective is to maintain data collection performance to assure at least 90% joint recoverability and availability of data needed for assessing the relative concentrations and doses resulting from accidental or routine releases. The restoration of the data collection capability of the meteorological facility in the event of equipment failure or malfunction will be accomplished by replacement or repair of affected equipment. A stock of spare parts and equipment is maintained to minimize and shorten the periods of outages. Equipment malfunctions or outages are detected by personnel during routine or special checks. Equipment outages that affect the data transmitted to the plant can be detected by review of data displays in the reactor control room. Also, checks of data availability to the emergency centers are performed each work day. When an outage of one or more of the critical data items occurs, the appropriate maintenance personnel will be notified. In the event that the onsite meteorological facility is rendered inoperable, or there is an outage of the communications of data access systems; there is no fully representative offsite source of meteorological data for identification of atmospheric dispersion conditions. Therefore, TVA has prepared procedures to provide for missing or garbled data. These procedures incorporate available onsite data (for a partial loss of data), offsite data, and conditional climatology. The CECC meteorologist will apply the appropriate procedures. 2.3.8 Conclusions The meteorology of the Browns Ferry site provides generally favorable atmospheric conditions for transport and dispersion of plant emissions (see Section 14 and Appendix E). There are no physiographical features in the area to cause local entrapment or accumulation of emissions, particularly during extended periods of anticyclonic circulation or atmospheric stagnation. Limited air mass modification may occur within the lower few hundred feet, particularly with southeast winds when the over-water trajectory may approach 10 miles. Evaluation of the site meteorological information collected since preparation of the Design and Analysis Report confirms earlier judgment that the protective features for provision of routine atmospheric discharge of radioactive material will be adequate. 2.3-10

BFN-25 The Browns Ferry site is located in an area occasionally traversed by cyclonic storms. Wind speeds in excess of 40 mph are occasionally reported, but wind speeds in excess of 75 mph are rare. The estimated probability of a tornado occurrence at the Browns Ferry site in any one year is 6.979 x 10-4, or about one occurrence in 1,433 years should be expected. In spite of the low probability, the plant is designed to withstand tornado forces. (See section 12 for design wind and tornado loadings.) Because of the suitable physiographical features and adequate atmospheric diffusion conditions, it is anticipated that routine emission rates for atmospheric release of radioactive material will be favorable as compared with calculated maximum permissible emission rates (see Appendix E). 2.3-11

BFN-25 REFERENCES

1. Discussion with staff meteorologists. Dr. W. Johnson, W. F. Hilsmeier, and W. M. Culkowski, Atmospheric Turbulence and Diffusion Laboratory, Oak Ridge, Tennessee, April 6, 1966.

2. "Low-Level Inversion Frequency in the Contiguous U.S.," Hosler, Charles R.,

Monthly Weather Review, 89, No. 9, 1961, pp. 319-39.

3. The Climate of Decatur, Alabama (1879-1958), Long, Arthur R., U.S. Weather Bureau State Climatologist for Alabama, Weather Bureau Office, Montgomery, Alabama, July 1959.

4. "Local Climatological Data," Annual Summary with Comparative Data, Huntsville, Alabama, NOAA, National Climatic Center, Asheville, NC, 1980.

5. Maximum Station Precipitation for 1, 2, 3, 6, 12, and 24 Hours, Alabama, Technical Paper No. 15, U.S. Weather Bureau, December 1955.

6. Rainfall Frequency Atlas of the United States, Technical Paper No. 40, U.S.

Weather Bureau.

7. "Summary of Monthly Aerological Records." Huntsville, Alabama (WBAN 03856), Office of Navy Representative, National Weather Records Center.

8. Tornado Occurrences in the United States, Technical Paper No. 20, U.S.

Department of Commerce, Weather Bureau, September 1952.

9. Monthly Weather Review, Vol. 78 (1950) through Vol. 93 (1965), U.S.

Department of Commerce, Weather Bureau.

10. Climatology of Tennessee, NWS Office for State Climatology, NOAA, Environmental Data Service, NCC, Asheville, NC.

11. Severe Local Storm Occurrences, 1955-1967, ESSA, Technical Memorandum WBTM 12, U.S. Department of Commerce, September 1969.

12. "Tornado Probabilities," Thom, H.C.S., Monthly Weather Review, October-December 1963.

13. Tornado data for the Browns Ferry Nuclear Plant site prepared by the National Severe Storms Forecast Center, Kansas City, Missouri, November 1987.

2.3-12

BFN-25

14. Browns Ferry Nuclear Plant Environmental Data Station Manual, U.S.

Tennessee Valley Authority

15. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.23, Revision 1, Meteorological Monitoring Programs for Nuclear Power Plants, Washington, D.C., March 2007.

16. Deleted

17. Deleted

18. Regulatory Guide 1.21, Revision 1, "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," U.S.

Atomic Energy Commission, Washington, D.C., June 1974. 2.3-13

BFN-16 TABLE 2.3-1 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY STABILITY CLASS JAN 1, 77 - DEC 31, 79 WIND SPEED STABILITY CLASS (MPH) A B C D E F G CALM 0.0 0.0 0.0 0.00 0.01 0.01 0.02 0.6- 1.4 0.0 0.0 0.0 0.05 0.68 0.50 0.57 1.5- 3.4 0.09 0.34 0.50 5.77 10.41 5.48 4.45 3.5- 5.4 2.29 2.11 1.52 8.77 9.77 4.03 2.13 5.5- 7.4 1.19 1.00 1.04 6.02 5.55 1.59 0.57 7.5-12.4 0.90 1.63 1.22 8.09 4.50 0.90 0.17 12.5-18.4 0.42 0.47 0.33 3.36 0.70 0.04 0.0 18.5-24.4 0.02 0.06 0.06 0.46 0.07 0.01 0.0

  >=24.5               0.0             0.0            0.0                         0.03      0.0    0.0  0.0 TOTAL                4.91            5.61           4.67                      32.55      31.69  12.56 7.91 TOTAL HOURS OF VALID STABILITY OBSERVATIONS                                        25935 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED STABILITY OBSERVATIONS              25577 TOTAL HOURS OF OBSERVATIONS                                                        26280 JOINT RECOVERABILITY PERCENTAGE                                                     97.3 METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL

BFN-16 TABLE 2.3-2 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY STABILITY CLASS JAN 1, 77 - DEC 31, 79 WIND SPEED STABILITY CLASS (MPH) A B C D E F G CALM 0.0 0.0 0.0 0.0 0.00 0.00 0.0 0.6- 1.4 0.0 0.0 0.0 0.04 0.05 0.01 0.01 1.5- 3.4 0.0 0.0 0.04 2.22 1.25 0.29 0.08 3.5- 5.4 0.0 0.01 0.08 5.98 2.73 0.65 0.18 5.5- 7.4 0.01 0.01 0.11 6.98 3.34 0.83 0.29 7.5-12.4 0.0 0.02 0.16 17.73 9.76 2.64 0.90 12.5-18.4 0.0 0.03 0.14 15.61 9.95 3.19 0.58 18.5-24.4 0.0 0.01 0.02 6.30 3.07 0.90 0.18

  >=24.5               0.0             0.0            0.05                        2.48      0.82  0.0  0.0 TOTAL                0.01            0.08           0.60                      57.34      30.97  8.51 2.22 TOTAL HOURS OF VALID STABILITY OBSERVATIONS                                      25729 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED STABILITY OBSERVATIONS            25347 TOTAL HOURS OF OBSERVATIONS                                                      26280 JOINT RECOVERABILITY PERCENTAGE                                                   96.4 METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL

BFN-16 TABLE 2.3-3 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS A (DELTA T< = -1.9 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.0 0.0 0.04 0.12 0.05 0.0 0.0 0.21 NNE 0.0 0.0 0.0 0.05 0.19 0.10 0.0 0.0 0.34 NE 0.0 0.0 0.0 0.04 0.06 0.0 0.0 0.0 0.10 ENE 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.01 E 0.0 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.01 ESE 0.0 0.01 0.11 0.17 0.02 0.0 0.0 0.0 0.31 SE 0.0 0.03 1.11 0.40 0.02 0.0 0.0 0.0 1.56 SSE 0.0 0.04 0.52 0.10 0.02 0.0 0.0 0.0 0.68 S 0.0 0.01 0.38 0.11 0.04 0.0 0.0 0.0 0.54 SSW 0.0 0.0 0.04 0.05 0.01 0.0 0.0 0.0 0.10 SW 0.0 0.0 0.05 0.04 0.0 0.0 0.0 0.0 0.09 WSW 0.0 0.0 0.04 0.07 0.04 0.0 0.0 0.0 0.15 W 0.0 0.0 0.01 0.05 0.05 0.01 0.0 0.0 0.12 WNW 0.0 0.0 0.02 0.03 0.09 0.06 0.0 0.0 0.20 NW 0.0 0.0 0.0 0.02 0.17 0.11 0.0 0.0 0.30 NNW 0.0 0.0 0.01 0.01 0.06 0.09 0.02 0.0 0.19 SUBTOTAL 0.0 0.09 2.29 1.19 0.90 0.42 0.02 0.0 4.91 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS A 1262 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS A 1259 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 6.8 MPH

BFN-16 TABLE 2.3-4 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS B (-1.9< DELTA-T< = -1.7 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.0 0.05 0.09 0.30 0.04 0.01 0.0 0.49 NNE 0.0 0.0 0.05 0.07 0.27 0.05 0.0 0.0 0.44 NE 0.0 0.0 0.04 0.02 0.09 0.01 0.0 0.0 0.16 ENE 0.0 0.01 0.01 0.01 0.01 0.0 0.0 0.0 0.04 E 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.03 ESE 0.0 0.02 0.10 0.04 0.0 0.0 0.0 0.0 0.16 SE 0.0 0.13 0.64 0.09 0.02 0.0 0.0 0.0 0.88 SSE 0.0 0.09 0.31 0.02 0.01 0.0 0.0 0.0 0.43 S 0.0 0.05 0.42 0.07 0.02 0.0 0.0 0.0 0.56 SSW 0.0 0.02 0.07 0.01 0.0 0.0 0.0 0.0 0.10 SW 0.0 0.0 0.17 0.02 0.0 0.0 0.0 0.0 0.19 WSW 0.0 0.0 0.11 0.13 0.05 0.01 0.0 0.0 0.30 W 0.0 0.02 0.04 0.17 0.17 0.03 0.0 0.0 0.43 WNW 0.0 0.0 0.07 0.11 0.23 0.08 0.04 0.0 0.53 NW 0.0 0.0 0.01 0.07 0.27 0.13 0.01 0.0 0.49 NNW 0.0 0.0 0.0 0.07 0.19 0.12 0.0 0.0 0.38 SUBTOTAL 0.0 0.34 2.11 1.00 1.63 0.47 0.06 0.0 5.61 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS B 1445 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS B 1440 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 7.2 MPH

BFN-16 TABLE 2.3-5 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS C (-1.7< DELTA-T< =-1.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.01 0.08 0.11 0.21 0.02 0.0 0.0 0.43 NNE 0.0 0.01 0.07 0.09 0.17 0.02 0.0 0.0 0.36 NE 0.0 0.0 0.03 0.08 0.05 0.0 0.0 0.0 0.16 ENE 0.0 0.0 0.02 0.02 0.0 0.0 0.0 0.0 0.04 E 0.0 0.0 0.03 0.02 0.0 0.0 0.0 0.0 0.05 ESE 0.0 0.01 0.05 0.02 0.0 0.0 0.0 0.0 0.08 SE 0.0 0.17 0.29 0.09 0.01 0.0 0.0 0.0 0.56 SSE 0.0 0.12 0.17 0.04 0.01 0.0 0.0 0.0 0.34 S 0.0 0.11 0.25 0.04 0.02 0.0 0.0 0.0 0.42 SSW 0.0 0.03 0.06 0.01 0.0 0.0 0.0 0.0 0.10 SW 0.0 0.03 0.12 0.03 0.01 0.0 0.0 0.0 0.19 WSW 0.0 0.0 0.11 0.07 0.07 0.0 0.0 0.0 0.25 W 0.0 0.0 0.05 0.12 0.10 0.02 0.01 0.0 0.30 WNW 0.0 0.01 0.12 0.13 0.17 0.07 0.04 0.0 0.54 NW 0.0 0.0 0.05 0.09 0.22 0.10 0.01 0.0 0.47 NNW 0.0 0.0 0.02 0.08 0.18 0.10 0.0 0.0 0.38 SUBTOTAL 0.0 0.50 1.52 1.04 1.22 0.33 0.06 0.0 4.67 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS C 1202 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS C 1197 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 7.0 MPH

BFN-16 TABLE 2.3-6 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS D (-1.5< DELTA-T< =-0.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.19 0.41 0.53 1.00 0.37 0.01 0.0 2.51 NNE 0.01 0.20 0.56 0.58 1.18 0.18 0.01 0.0 2.72 NE 0.01 0.12 0.38 0.43 0.52 0.01 0.0 0.0 1.47 ENE 0.0 0.26 0.23 0.15 0.05 0.01 0.0 0.0 0.70 E 0.0 0.20 0.31 0.17 0.05 0.0 0.0 0.0 0.73 ESE 0.0 0.24 0.51 0.30 0.08 0.0 0.0 0.0 1.13 SE 0.02 1.16 1.31 0.83 0.26 0.0 0.0 0.0 3.58 SSE 0.01 0.99 0.99 0.26 0.11 0.02 0.0 0.0 2.38 S 0.0 0.92 1.17 0.34 0.17 0.0 0.0 0.0 2.60 SSW 0.0 0.45 0.29 0.08 0.04 0.0 0.0 0.0 0.86 SW 0.0 0.24 0.29 0.09 0.02 0.01 0.0 0.0 0.65 WSW 0.0 0.32 0.70 0.29 0.33 0.11 0.0 0.0 1.75 W 0.0 0.18 0.55 0.62 0.63 0.22 0.03 0.0 2.23 WNW 0.0 0.13 0.39 0.42 1.10 0.82 0.22 0.01 3.09 NW 0.0 0.04 0.28 0.38 1.01 0.87 0.14 0.02 2.74 NNW 0.0 0.13 0.40 0.55 1.54 0.74 0.05 0.0 3.41 SUBTOTAL 0.05 5.77 8.77 6.02 8.09 3.36 0.46 0.03 32.55 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS D 8438 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS D 8341 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 7.1 MPH

BFN-16 TABLE 2.3-7 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS E (-0.5< DELTA-T< = 1.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.04 0.47 0.54 0.43 0.41 0.05 0.01 0.0 1.95 NNE 0.05 0.61 0.74 0.55 0.47 0.04 0.0 0.0 2.46 NE 0.05 0.57 0.63 0.42 0.27 0.02 0.0 0.0 1.96 ENE 0.05 0.71 0.45 0.17 0.08 0.02 0.0 0.0 1.48 E 0.04 0.61 0.74 0.16 0.07 0.0 0.0 0.0 1.62 ESE 0.03 0.76 1.01 0.53 0.16 0.01 0.0 0.0 2.50 SE 0.11 2.04 1.75 0.92 0.55 0.02 0.0 0.0 5.39 SSE 0.07 1.16 0.78 0.48 0.33 0.04 0.0 0.0 2.86 S 0.05 1.03 0.74 0.44 0.63 0.14 0.01 0.0 3.04 SSW 0.02 0.52 0.14 0.08 0.06 0.01 0.0 0.0 0.83 SW 0.04 0.30 0.07 0.02 0.03 0.0 0.0 0.0 0.46 WSW 0.01 0.53 0.60 0.14 0.11 0.04 0.0 0.0 1.43 W 0.02 0.37 0.77 0.42 0.27 0.04 0.0 0.0 1.89 WNW 0.03 0.15 0.13 0.11 0.22 0.09 0.02 0.0 0.75 NW 0.02 0.17 0.20 0.14 0.25 0.09 0.02 0.0 0.89 NNW 0.05 0.41 0.48 0.54 0.59 0.09 0.01 0.0 2.17 SUBTOTAL 0.68 10.41 9.77 5.55 4.50 0.70 0.07 0.0 31.68 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS E 8264 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS E 8098 TOTAL HOURS CALM 3 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.0 MPH

BFN-16 TABLE 2.3-8 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS F (1.5< DELTA-T< = 4.0 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.05 0.36 0.52 0.28 0.06 0.0 0.0 0.0 1.27 NNE 0.05 0.51 0.66 0.34 0.11 0.0 0.0 0.0 1.67 NE 0.07 0.34 0.27 0.18 0.01 0.0 0.0 0.0 0.87 ENE 0.03 0.53 0.33 0.05 0.0 0.0 0.0 0.0 0.94 E 0.01 0.59 0.52 0.03 0.0 0.0 0.0 0.0 1.15 ESE 0.0 0.52 0.22 0.0 0.0 0.0 0.0 0.0 0.74 SE 0.09 0.97 0.48 0.17 0.13 0.01 0.0 0.0 1.85 SSE 0.05 0.54 0.34 0.17 0.25 0.02 0.01 0.0 1.38 S 0.03 0.29 0.18 0.20 0.27 0.01 0.0 0.0 0.98 SSW 0.03 0.13 0.03 0.0 0.01 0.0 0.0 0.0 0.20 SW 0.0 0.09 0.03 0.0 0.0 0.0 0.0 0.0 0.12 WSW 0.0 0.09 0.07 0.0 0.0 0.0 0.0 0.0 0.16 W 0.02 0.09 0.06 0.0 0.01 0.0 0.0 0.0 0.18 WNW 0.01 0.08 0.01 0.0 0.0 0.0 0.0 0.0 0.10 NW 0.01 0.08 0.04 0.01 0.0 0.0 0.0 0.0 0.14 NNW 0.05 0.27 0.27 0.16 0.05 0.0 0.0 0.0 0.80 SUBTOTAL 0.50 5.48 4.03 1.59 0.90 0.04 0.01 0.0 12.55 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS F 3268 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS F 3223 TOTAL HOURS CALM 2 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 4.0 MPH

BFN-16 TABLE 2.3-9 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS G (DELTA T > 4.0 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.07 0.76 0.32 0.02 0.0 0.0 0.0 0.0 1.17 NNE 0.05 0.83 0.51 0.18 0.02 0.0 0.0 0.0 1.59 NE 0.04 0.34 0.12 0.02 0.0 0.0 0.0 0.0 0.52 ENE 0.04 0.48 0.18 0.02 0.0 0.0 0.0 0.0 0.72 E 0.02 0.52 0.34 0.0 0.0 0.0 0.0 0.0 0.88 ESE 0.01 0.18 0.01 0.0 0.0 0.0 0.0 0.0 0.20 SE 0.08 0.43 0.09 0.04 0.03 0.0 0.0 0.0 0.67 SSE 0.03 0.44 0.31 0.16 0.08 0.0 0.0 0.0 1.02 S 0.05 0.09 0.12 0.10 0.04 0.0 0.0 0.0 0.40 SSW 0.05 0.05 0.01 0.0 0.0 0.0 0.0 0.0 0.11 SW 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.0 0.01 WSW 0.02 0.02 0.0 0.0 0.0 0.0 0.0 0.0 0.04 W 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.0 0.02 WNW 0.01 0.02 0.0 0.0 0.0 0.0 0.0 0.0 0.03 NW 0.04 0.04 0.0 0.0 0.0 0.0 0.0 0.0 0.08 NNW 0.05 0.23 0.12 0.03 0.0 0.0 0.0 0.0 0.43 SUBTOTAL 0.57 4.45 2.13 0.57 0.17 0.0 0.0 0.0 7.89 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25935 TOTAL HOURS OF STABILITY CLASS G 2056 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS G 2019 TOTAL HOURS CALM 4 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 10.03 AND 45.30 METERS WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 3.2 MPH

BFN-16 TABLE 2.3-10 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.16 1.82 1.90 1.50 2.10 0.53 0.03 0.0 8.04 NNE 0.17 2.19 2.61 1.86 2.40 0.39 0.01 0.0 9.63 NE 0.17 1.36 1.46 1.19 1.01 0.05 0.0 0.0 5.24 ENE 0.13 1.99 1.22 0.42 0.15 0.02 0.0 0.0 3.93 E 0.07 1.93 1.94 0.38 0.12 0.0 0.0 0.0 4.44 ESE 0.05 1.74 2.01 1.07 0.27 0.01 0.0 0.0 5.15 SE 0.31 4.92 5.70 2.55 1.02 0.04 0.0 0.0 14.54 SSE 0.16 3.37 3.49 1.22 0.79 0.07 0.01 0.0 9.11 S 0.14 2.50 3.27 1.30 1.19 0.15 0.02 0.0 8.57 SSW 0.10 1.20 0.65 0.23 0.12 0.01 0.0 0.0 2.31 SW 0.04 0.69 0.73 0.22 0.07 0.01 0.0 0.0 1.76 WSW 0.03 0.97 1.60 0.69 0.60 0.15 0.0 0.0 4.04 W 0.04 0.66 1.47 1.38 1.22 0.32 0.05 0.0 5.14 WNW 0.04 0.39 0.74 0.78 1.81 1.12 0.33 0.02 5.23 NW 0.07 0.33 0.57 0.71 1.91 1.29 0.17 0.02 5.07 NNW 0.15 1.05 1.31 1.44 2.58 1.13 0.09 0.01 7.76 SUBTOTAL 1.83 27.11 30.67 16.94 17.36 5.29 0.71 0.05 99.96 TOTAL HOURS OF VALID WIND OBSERVATIONS 25810 TOTAL HOURS OF OBSERVATIONS 26280 RECOVERABILITY PERCENTAGE 98.2 TOTAL HOURS CALM 10 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.7 MPH

BFN-16 TABLE 2.3-11 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JANUARY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 1.34 1.79 1.74 1.84 0.25 0.10 0.0 7.06 NNE 0.05 1.99 2.09 1.39 2.24 0.15 0.0 0.0 7.91 NE 0.0 1.25 1.44 1.20 1.15 0.0 0.0 0.0 5.04 ENE 0.0 1.29 1.29 0.45 0.0 0.0 0.0 0.0 3.03 E 0.05 1.39 2.84 0.50 0.15 0.0 0.0 0.0 4.93 ESE 0.05 0.80 2.29 0.80 0.10 0.0 0.0 0.0 4.04 SE 0.25 1.99 4.08 2.19 1.89 0.0 0.0 0.0 10.40 SSE 0.0 1.94 2.24 0.95 0.90 0.05 0.0 0.0 6.08 S 0.05 0.50 0.70 0.55 0.40 0.0 0.0 0.0 2.20 SSW 0.05 0.30 0.25 0.05 0.05 0.0 0.0 0.0 0.70 SW 0.0 0.10 0.15 0.05 0.0 0.0 0.0 0.0 0.30 WSW 0.0 0.60 0.75 0.10 0.10 0.15 0.0 0.0 1.70 W 0.0 0.80 0.85 1.54 1.69 1.05 0.15 0.0 6.08 WNW 0.0 0.50 0.75 0.85 3.29 3.64 1.99 0.10 11.12 NW 0.05 0.35 1.10 0.90 4.33 5.48 0.30 0.05 12.56 NNW 0.0 0.85 1.15 1.89 8.47 3.69 0.80 0.05 16.90 SUBTOTAL 0.55 15.99 23.76 15.15 26.60 14.46 3.34 0.20 100.05 TOTAL HOURS OF VALID WIND OBSERVATIONS 2008 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 90.0 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 7.9 MPH

BFN-16 TABLE 2.3-12 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS FEBRUARY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.05 1.59 1.93 2.03 3.77 1.34 0.0 0.0 10.71 NNE 0.05 1.54 2.18 2.43 4.96 1.14 0.05 0.0 12.35 NE 0.10 0.84 1.64 1.44 1.39 0.10 0.0 0.0 5.51 ENE 0.0 1.34 0.89 0.74 0.25 0.0 0.0 0.0 3.22 E 0.15 1.54 1.19 0.60 0.25 0.0 0.0 0.0 3.73 ESE 0.10 1.39 1.14 0.69 0.05 0.0 0.0 0.0 3.37 SE 0.20 3.87 3.82 1.14 0.35 0.15 0.0 0.0 9.53 SSE 0.05 3.42 4.17 1.59 0.94 0.05 0.0 0.0 10.22 S 0.10 1.64 2.38 1.14 1.29 0.0 0.0 0.0 6.55 SSW 0.15 0.35 0.50 0.20 0.30 0.0 0.0 0.0 1.50 SW 0.0 0.35 0.64 0.40 0.15 0.10 0.0 0.0 1.64 WSW 0.05 0.45 0.64 0.30 1.34 0.64 0.0 0.0 3.42 W 0.0 0.40 0.74 0.64 1.39 0.60 0.0 0.0 3.77 WNW 0.0 0.30 0.79 0.69 2.58 1.88 0.30 0.0 6.54 NW 0.0 0.20 0.60 0.55 2.83 1.93 0.64 0.0 6.75 NNW 0.0 1.04 2.18 2.43 3.57 1.93 0.05 0.0 11.20 SUBTOTAL 1.00 20.26 25.43 17.01 25.41 9.86 1.04 0.0 100.01 TOTAL HOURS OF VALID WIND OBSERVATIONS 2016 TOTAL HOURS OF OBSERVATIONS 2016 RECOVERABILITY PERCENTAGE 100.0 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 6.8 MPH

BFN-16 TABLE 2.3-13 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS MARCH (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.09 0.68 1.49 0.95 1.35 1.31 0.0 0.0 5.87 NNE 0.05 0.95 1.89 1.26 1.85 0.41 0.0 0.0 6.41 NE 0.05 0.77 1.40 1.94 0.95 0.0 0.0 0.0 5.11 ENE 0.18 0.77 0.81 0.27 0.0 0.0 0.0 0.0 2.03 E 0.0 0.90 1.53 0.27 0.0 0.0 0.0 0.0 2.70 ESE 0.09 2.07 2.03 0.99 0.54 0.0 0.0 0.0 5.72 SE 0.05 4.64 4.96 2.39 2.52 0.18 0.05 0.0 14.79 SSE 0.14 3.83 3.92 2.88 1.76 0.23 0.09 0.0 12.85 S 0.09 1.67 3.70 2.70 3.42 0.14 0.09 0.0 11.81 SSW 0.14 0.50 0.45 0.32 0.27 0.0 0.0 0.0 1.68 SW 0.05 0.41 0.59 0.18 0.18 0.0 0.0 0.0 1.41 WSW 0.0 0.54 1.13 0.63 1.31 0.27 0.0 0.0 3.88 W 0.05 0.27 0.90 1.35 1.98 0.90 0.09 0.0 5.54 WNW 0.0 0.27 0.41 0.81 2.66 1.85 0.14 0.05 6.19 NW 0.0 0.09 0.63 0.90 2.79 1.44 0.32 0.05 6.22 NNW 0.0 0.68 1.13 1.35 2.97 1.71 0.05 0.0 7.89 SUBTOTAL 0.98 19.04 26.97 19.19 24.55 8.44 0.83 0.10 100.10 TOTAL HOURS OF VALID WIND OBSERVATIONS 2219 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.4 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 6.7 MPH

BFN-16 TABLE 2.3-14 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS APRIL (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.14 1.44 1.26 0.79 1.26 0.51 0.14 0.0 5.54 NNE 0.09 2.37 1.72 1.63 1.40 0.47 0.0 0.0 7.68 NE 0.23 1.44 1.49 1.82 1.44 0.05 0.0 0.0 6.47 ENE 0.14 2.00 1.16 0.51 0.19 0.05 0.0 0.0 4.05 E 0.05 1.44 1.12 0.19 0.0 0.0 0.0 0.0 2.80 ESE 0.14 1.86 1.44 1.02 0.33 0.0 0.0 0.0 4.79 SE 0.70 4.42 5.26 2.42 1.40 0.0 0.0 0.0 14.20 SSE 0.14 4.93 4.75 0.70 0.93 0.0 0.0 0.0 11.45 S 0.33 2.93 4.10 1.86 2.09 0.19 0.0 0.0 11.50 SSW 0.09 1.16 0.84 0.61 0.33 0.0 0.0 0.0 3.03 SW 0.0 0.51 0.23 0.28 0.05 0.0 0.0 0.0 1.07 WSW 0.09 0.93 1.16 0.74 1.12 0.05 0.0 0.0 4.09 W 0.0 0.56 1.63 1.30 0.88 0.19 0.14 0.0 4.70 WNW 0.05 0.28 0.42 0.79 2.28 1.58 0.56 0.0 5.96 NW 0.14 0.09 0.42 0.74 1.91 1.40 0.14 0.0 4.84 NNW 0.14 0.88 1.07 1.40 2.65 1.68 0.0 0.0 7.82 SUBTOTAL 2.47 27.24 28.07 16.80 18.26 6.17 0.98 0.0 99.99 TOTAL HOURS OF VALID WIND OBSERVATIONS 2148 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.4 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.8 MPH

BFN-16 TABLE 2.3-15 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS MAY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.32 2.43 1.31 1.04 1.85 0.18 0.0 0.0 7.13 NNE 0.23 2.30 3.02 1.49 1.98 0.18 0.0 0.0 9.20 NE 0.23 1.76 1.49 0.81 1.04 0.0 0.0 0.0 5.33 ENE 0.32 3.16 1.44 0.09 0.14 0.0 0.0 0.0 5.15 E 0.09 2.75 1.58 0.41 0.14 0.0 0.0 0.0 4.97 ESE 0.09 1.85 2.98 1.17 0.18 0.0 0.0 0.0 6.27 SE 0.23 5.55 6.40 1.80 0.23 0.0 0.0 0.0 14.21 SSE 0.18 3.88 3.11 0.90 0.50 0.0 0.0 0.0 8.57 S 0.18 3.07 3.34 1.26 0.72 0.05 0.0 0.0 8.62 SSW 0.0 1.71 1.26 0.41 0.0 0.0 0.0 0.0 3.38 SW 0.05 1.04 1.31 0.32 0.23 0.0 0.0 0.0 2.95 WSW 0.0 1.13 2.30 1.35 0.54 0.05 0.0 0.0 5.37 W 0.0 0.59 1.49 1.58 1.67 0.32 0.14 0.0 5.79 WNW 0.18 0.50 0.41 0.90 1.53 0.27 0.36 0.0 4.15 NW 0.09 0.41 0.59 1.04 1.35 0.41 0.05 0.0 3.94 NNW 0.32 1.22 0.81 0.41 1.44 0.86 0.05 0.0 5.11 SUBTOTAL 2.51 33.35 32.84 14.98 13.54 2.32 0.60 0.0 100.14 TOTAL HOURS OF VALID WIND OBSERVATIONS 2218 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.4 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.0 MPH

BFN-16 TABLE 2.3-16 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JUNE (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.28 2.75 1.58 1.35 2.56 0.47 0.0 0.05 9.04 NNE 0.37 2.28 1.68 1.07 2.23 0.65 0.0 0.0 8.28 NE 0.42 1.40 0.93 0.42 0.42 0.19 0.0 0.0 3.78 ENE 0.33 2.09 0.74 0.09 0.0 0.0 0.0 0.0 3.25 E 0.09 1.58 1.58 0.05 0.0 0.0 0.0 0.0 3.30 ESE 0.0 2.05 1.96 0.14 0.0 0.0 0.0 0.0 4.15 SE 0.28 7.64 5.31 1.30 0.14 0.0 0.0 0.0 14.67 SSE 0.19 3.40 2.37 0.37 0.0 0.0 0.0 0.0 6.33 S 0.19 3.82 3.26 0.70 0.05 0.0 0.0 0.0 8.02 SSW 0.23 2.65 0.79 0.05 0.0 0.0 0.0 0.0 3.72 SW 0.05 1.72 1.30 0.14 0.0 0.0 0.0 0.0 3.21 WSW 0.0 2.47 3.21 1.49 0.51 0.0 0.0 0.0 7.68 W 0.0 1.82 3.35 2.42 1.72 0.0 0.0 0.0 9.31 WNW 0.09 0.33 0.98 0.93 2.14 0.23 0.0 0.0 4.70 NW 0.0 0.37 0.65 1.16 1.91 0.19 0.0 0.0 4.28 NNW 0.23 1.77 1.30 2.05 0.70 0.09 0.0 0.0 6.14 SUBTOTAL 2.75 38.14 30.99 13.73 12.38 1.82 0.0 0.05 99.86 TOTAL HOURS OF VALID WIND OBSERVATIONS 2148 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.4 TOTAL HOURS CALM 3 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 4.6 MPH

BFN-16 TABLE 2.3-17 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JULY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.13 1.70 2.11 1.03 0.72 0.04 0.0 0.0 5.73 NNE 0.13 2.69 3.14 1.48 2.24 0.09 0.04 0.0 9.81 NE 0.09 1.30 1.26 0.99 0.85 0.0 0.0 0.0 4.49 ENE 0.04 2.11 1.57 0.22 0.04 0.0 0.0 0.0 3.98 E 0.0 1.26 2.47 0.49 0.36 0.0 0.0 0.0 4.58 ESE 0.0 1.03 2.24 1.61 0.18 0.09 0.0 0.0 5.15 SE 0.22 4.53 6.32 3.81 1.08 0.09 0.0 0.0 16.05 SSE 0.13 3.36 3.63 0.22 0.13 0.0 0.0 0.0 7.47 S 0.13 4.39 5.25 0.94 0.27 0.04 0.0 0.0 11.02 SSW 0.0 2.65 1.17 0.04 0.13 0.0 0.0 0.0 3.99 SW 0.09 1.57 1.79 0.22 0.0 0.0 0.0 0.0 3.67 WSW 0.0 2.11 4.08 1.08 0.67 0.0 0.0 0.0 7.94 W 0.0 1.48 2.74 2.20 0.90 0.0 0.0 0.0 7.32 WNW 0.04 0.31 1.39 0.76 1.26 0.27 0.0 0.0 4.03 NW 0.04 0.13 0.40 0.36 0.90 0.22 0.0 0.0 2.05 NNW 0.18 0.54 0.45 0.81 0.58 0.09 0.0 0.0 2.65 SUBTOTAL 1.22 31.16 40.01 16.26 10.31 0.93 0.04 0.0 99.93 TOTAL HOURS OF VALID WIND OBSERVATIONS 2230 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.9 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 4.7 MPH

BFN-16 TABLE 2.3-18 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS AUGUST (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.18 2.34 1.80 1.53 1.21 0.09 0.0 0.0 7.15 NNE 0.36 2.97 3.42 2.20 0.90 0.13 0.04 0.0 10.02 NE 0.27 2.07 1.35 0.63 0.31 0.0 0.0 0.0 4.63 ENE 0.04 2.43 1.26 0.40 0.04 0.0 0.0 0.0 4.17 E 0.13 2.56 1.80 0.13 0.0 0.0 0.0 0.0 4.62 ESE 0.0 2.02 2.07 0.90 0.13 0.0 0.0 0.0 5.12 SE 0.18 8.00 10.88 2.74 0.09 0.0 0.0 0.0 21.89 SSE 0.18 4.09 4.00 0.36 0.04 0.0 0.0 0.0 8.67 S 0.22 4.32 4.59 0.72 0.45 0.0 0.0 0.0 10.30 SSW 0.04 2.61 0.94 0.18 0.0 0.0 0.0 0.0 3.77 SW 0.0 0.99 0.85 0.18 0.04 0.0 0.0 0.0 2.06 WSW 0.09 1.35 2.07 0.58 0.04 0.0 0.0 0.0 4.13 W 0.04 0.76 1.71 1.71 0.58 0.0 0.0 0.0 4.80 WNW 0.0 0.40 0.72 0.99 0.45 0.04 0.0 0.0 2.60 NW 0.0 0.54 0.36 0.36 0.63 0.09 0.0 0.0 1.98 NNW 0.31 0.90 1.21 0.90 0.58 0.09 0.0 0.0 3.99 SUBTOTAL 2.04 38.35 39.03 14.51 5.49 0.44 0.04 0.0 99.90 TOTAL HOURS OF VALID WIND OBSERVATIONS 2224 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.6 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 4.1 MPH

BFN-16 TABLE 2.3-19 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEPT. (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.15 2.64 2.74 2.79 2.79 0.40 0.05 0.0 11.56 NNE 0.10 2.99 4.98 2.94 3.33 0.75 0.0 0.0 15.09 NE 0.10 1.49 2.49 1.89 1.49 0.20 0.0 0.0 7.66 ENE 0.20 3.04 2.59 1.29 0.45 0.20 0.05 0.0 7.82 E 0.0 2.24 3.93 1.00 0.40 0.0 0.0 0.0 7.57 ESE 0.05 2.39 3.68 1.79 0.25 0.0 0.0 0.0 8.16 SE 0.35 3.88 6.57 3.14 0.30 0.0 0.0 0.0 14.24 SSE 0.10 1.54 3.14 0.45 0.20 0.20 0.0 0.0 5.63 S 0.05 1.69 3.04 0.30 0.0 0.0 0.0 0.0 5.08 SSW 0.05 0.45 0.05 0.25 0.0 0.0 0.0 0.0 0.80 SW 0.05 0.40 0.80 0.15 0.0 0.0 0.0 0.0 1.40 WSW 0.0 0.35 0.60 0.0 0.20 0.0 0.0 0.0 1.15 W 0.0 0.25 1.19 0.90 0.30 0.05 0.0 0.0 2.69 WNW 0.0 0.30 0.65 0.55 0.35 0.20 0.0 0.0 2.05 NW 0.10 0.30 0.50 0.75 1.05 0.15 0.0 0.05 2.90 NNW 0.15 1.74 1.74 1.34 0.95 0.15 0.0 0.05 6.12 SUBTOTAL 1.45 25.69 38.69 19.53 12.06 2.30 0.10 0.10 99.92 TOTAL HOURS OF VALID WIND OBSERVATIONS 2009 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 93.0 TOTAL HOURS CALM 3 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.1 MPH

BFN-16 TABLE 2.3-20 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS OCTOBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.18 2.38 3.64 1.75 2.43 0.90 0.0 0.0 11.28 NNE 0.09 2.79 3.14 1.80 1.93 0.18 0.0 0.0 9.93 NE 0.27 1.44 0.99 0.54 0.67 0.0 0.0 0.0 3.91 ENE 0.13 2.61 0.76 0.18 0.09 0.0 0.0 0.0 3.77 E 0.18 3.19 2.70 0.13 0.0 0.0 0.0 0.0 6.20 ESE 0.0 2.34 2.25 1.21 0.67 0.0 0.0 0.0 6.47 SE 0.09 5.03 4.99 1.89 1.48 0.04 0.0 0.0 13.52 SSE 0.31 2.92 3.37 0.72 0.31 0.04 0.0 0.0 7.67 S 0.0 2.29 3.32 1.44 0.31 0.0 0.0 0.0 7.36 SSW 0.27 0.63 0.22 0.27 0.0 0.0 0.0 0.0 1.39 SW 0.09 0.49 0.36 0.22 0.04 0.0 0.0 0.0 1.20 WSW 0.0 0.49 1.03 0.81 0.49 0.0 0.0 0.0 2.82 W 0.13 0.31 1.08 1.03 1.26 0.04 0.0 0.0 3.85 WNW 0.04 0.58 0.76 0.76 1.71 0.45 0.0 0.0 4.30 NW 0.18 0.63 0.90 0.81 2.07 1.26 0.0 0.0 5.85 NNW 0.09 1.66 2.02 2.07 3.32 1.21 0.0 0.0 10.37 SUBTOTAL 2.05 29.78 31.53 15.63 16.78 4.12 0.0 0.0 99.89 TOTAL HOURS OF VALID WIND OBSERVATIONS 2226 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.7 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.3 MPH

BFN-16 TABLE 2.3-21 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS NOVEMBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.33 1.73 2.10 2.19 2.89 0.23 0.0 0.0 9.47 NNE 0.37 2.38 2.71 3.22 3.55 0.42 0.0 0.0 12.65 NE 0.09 1.26 2.10 1.87 1.73 0.09 0.0 0.0 7.14 ENE 0.14 1.96 1.26 0.61 0.61 0.05 0.0 0.0 4.63 E 0.0 2.89 1.35 0.51 0.14 0.0 0.0 0.0 4.89 ESE 0.0 1.35 0.84 1.77 0.65 0.0 0.0 0.0 4.61 SE 0.47 5.74 3.73 2.61 1.12 0.0 0.0 0.0 13.67 SSE 0.28 3.55 3.03 2.43 1.63 0.05 0.0 0.0 10.97 S 0.19 1.31 1.77 1.31 1.63 0.37 0.05 0.0 6.63 SSW 0.09 0.37 0.42 0.0 0.05 0.0 0.0 0.0 0.93 SW 0.09 0.42 0.14 0.23 0.09 0.0 0.0 0.0 0.97 WSW 0.09 0.51 1.03 0.75 0.37 0.19 0.0 0.0 2.94 W 0.23 0.56 1.03 0.79 0.70 0.51 0.09 0.0 3.91 WNW 0.05 0.65 0.93 0.79 1.03 0.89 0.14 0.0 4.48 NW 0.09 0.42 0.23 0.47 1.35 1.45 0.14 0.0 4.15 NNW 0.33 1.17 1.68 1.31 2.71 0.61 0.0 0.0 7.81 SUBTOTAL 2.84 26.27 24.35 20.86 20.25 4.86 0.42 0.0 99.85 TOTAL HOURS OF VALID WIND OBSERVATIONS 2143 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.2 TOTAL HOURS CALM 3 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 5.8 MPH

BFN-16 TABLE 2.3-22 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS DECEMBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.09 0.90 1.13 1.04 2.75 0.63 0.14 0.0 6.68 NNE 0.09 1.04 1.40 1.49 2.48 0.23 0.0 0.0 6.73 NE 0.14 1.31 1.08 0.86 0.77 0.0 0.0 0.0 4.16 ENE 0.0 1.04 0.90 0.27 0.0 0.0 0.0 0.0 2.21 E 0.14 1.35 1.35 0.36 0.09 0.0 0.0 0.0 3.29 ESE 0.09 1.62 1.31 0.68 0.09 0.0 0.0 0.0 3.79 SE 0.68 3.42 5.67 5.00 1.58 0.0 0.0 0.0 16.35 SSE 0.18 3.38 4.05 3.02 2.16 0.23 0.05 0.0 13.07 S 0.14 2.07 3.38 2.52 3.47 1.04 0.09 0.0 12.71 SSW 0.09 0.86 0.81 0.41 0.36 0.09 0.0 0.0 2.62 SW 0.05 0.14 0.50 0.23 0.0 0.0 0.0 0.0 0.92 WSW 0.0 0.59 0.95 0.36 0.50 0.50 0.0 0.0 2.90 W 0.0 0.09 0.81 1.04 1.53 0.27 0.0 0.0 3.74 WNW 0.05 0.27 0.68 0.50 2.52 2.30 0.59 0.05 6.96 NW 0.09 0.41 0.54 0.50 2.03 1.85 0.50 0.05 5.97 NNW 0.05 0.23 1.08 1.44 3.51 1.67 0.14 0.0 8.12 SUBTOTAL 1.88 18.72 25.64 19.72 23.84 8.81 1.51 0.10 100.22 TOTAL HOURS OF VALID WIND OBSERVATIONS 2221 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.5 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL MEAN WIND SPEED = 6.8 MPH

BFN-16 TABLE 2.3-23 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS A (DELTA T< = -1.9 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NNE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ENE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 E 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ESE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SSE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 S 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SSW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SW 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.01 WSW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 W 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 WNW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NNW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SUBTOTAL 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.01 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS A 6 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS A 6 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 9.5 MPH

BFN-16 TABLE 2.3-24 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS B (-1.9< DELTA-T< = -1.7 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NNE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ENE 0.0 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.01 E 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ESE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SE 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 SSE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 S 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SSW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SW 0.0 0.0 0.0 0.01 0.01 0.0 0.0 0.0 0.02 WSW 0.0 0.0 0.0 0.0 0.0 0.01 0.0 0.0 0.01 W 0.0 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.03 WNW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NNW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SUBTOTAL 0.0 0.0 0.01 0.01 0.02 0.03 0.01 0.0 0.08 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS B 30 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS B 30 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 10.4 MPH

BFN-16 TABLE 2.3-25 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS C (-1.7< DELTA-T < = -1.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NNE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 NE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ENE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 E 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ESE 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.03 SE 0.0 0.01 0.03 0.02 0.01 0.0 0.0 0.0 0.07 SSE 0.0 0.01 0.01 0.01 0.0 0.0 0.0 0.0 0.03 S 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SSW 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 SW 0.0 0.01 0.02 0.04 0.04 0.02 0.0 0.0 0.13 WSW 0.0 0.01 0.01 0.03 0.04 0.02 0.0 0.0 0.11 W 0.0 0.0 0.0 0.0 0.03 0.03 0.02 0.04 0.12 WNW 0.0 0.0 0.0 0.01 0.02 0.03 0.0 0.01 0.07 NW 0.0 0.0 0.0 0.0 0.0 0.03 0.0 0.0 0.03 NNW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SUBTOTAL 0.0 0.04 0.08 0.11 0.16 0.14 0.02 0.05 0.60 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS C 164 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS C 164 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 11.0 MPH

BFN-16 TABLE 2.3-26 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS D (-1.5< DELTA-T< = -0.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.01 0.11 0.30 0.48 1.54 1.51 0.38 0.05 4.38 NNE 0.0 0.09 0.22 0.43 1.65 1.42 0.15 0.04 4.00 NE 0.0 0.06 0.21 0.41 1.18 0.48 0.04 0.02 2.40 ENE 0.01 0.09 0.21 0.21 0.42 0.15 0.05 0.02 1.16 E 0.0 0.11 0.24 0.20 0.25 0.12 0.02 0.0 0.94 ESE 0.0 0.12 0.28 0.32 0.81 0.64 0.20 0.06 2.43 SE 0.01 0.22 0.71 0.75 1.42 1.39 0.81 0.32 5.63 SSE 0.0 0.20 0.69 0.56 1.39 1.34 0.58 0.28 5.04 S 0.0 0.25 0.49 0.48 1.33 1.24 0.71 0.31 4.81 SSW 0.0 0.20 0.44 0.39 0.86 0.82 0.39 0.16 3.26 SW 0.0 0.27 0.43 0.37 0.68 0.65 0.26 0.10 2.76 WSW 0.0 0.15 0.53 0.44 0.59 0.53 0.24 0.13 2.61 W 0.0 0.13 0.50 0.60 1.25 0.65 0.35 0.30 3.78 WNW 0.0 0.07 0.27 0.49 1.63 1.27 0.67 0.28 4.68 NW 0.01 0.07 0.28 0.54 1.42 1.70 0.86 0.28 5.16 NNW 0.0 0.08 0.18 0.31 1.31 1.70 0.59 0.13 4.30 SUBTOTAL 0.04 2.22 5.98 6.98 17.73 15.61 6.30 2.48 57.34 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS D 14807 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS D 14557 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.1 MPH

BFN-16 TABLE 2.3-27 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS E (-0.5< DELTA-T< = 1.5 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.05 0.07 0.13 0.45 0.77 0.15 0.0 1.62 NNE 0.0 0.04 0.07 0.12 0.67 0.88 0.17 0.0 1.95 NE 0.0 0.04 0.10 0.15 0.77 0.87 0.26 0.0 2.19 ENE 0.0 0.04 0.08 0.12 0.45 0.39 0.11 0.02 1.21 E 0.0 0.09 0.16 0.19 0.47 0.17 0.05 0.01 1.14 ESE 0.01 0.07 0.17 0.21 0.63 0.70 0.17 0.03 1.99 SE 0.0 0.09 0.43 0.39 1.35 1.50 0.71 0.37 4.84 SSE 0.0 0.17 0.42 0.51 1.14 1.01 0.57 0.25 4.07 S 0.01 0.09 0.34 0.45 0.89 0.95 0.27 0.09 3.09 SSW 0.0 0.09 0.14 0.20 0.59 0.61 0.20 0.01 1.84 SW 0.01 0.07 0.14 0.17 0.56 0.64 0.10 0.03 1.72 WSW 0.01 0.08 0.11 0.14 0.42 0.34 0.07 0.01 1.18 W 0.0 0.07 0.13 0.18 0.39 0.30 0.04 0.0 1.11 WNW 0.01 0.07 0.15 0.15 0.35 0.20 0.04 0.0 0.97 NW 0.0 0.09 0.15 0.14 0.33 0.27 0.07 0.0 1.05 NNW 0.0 0.10 0.07 0.09 0.30 0.35 0.09 0.0 1.00 SUBTOTAL 0.05 1.25 2.73 3.34 9.76 9.95 3.07 0.82 30.97 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS E 7965 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS E 7855 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.0 MPH

BFN-16 TABLE 2.3-28 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS F (1.5< DELTA-T< = 4.0 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.02 0.02 0.02 0.09 0.18 0.06 0.0 0.39 NNE 0.0 0.0 0.04 0.02 0.11 0.36 0.17 0.0 0.70 NE 0.0 0.01 0.02 0.04 0.17 0.40 0.24 0.0 0.88 ENE 0.0 0.01 0.02 0.03 0.22 0.34 0.16 0.0 0.78 E 0.0 0.01 0.06 0.05 0.24 0.11 0.0 0.0 0.47 ESE 0.0 0.04 0.06 0.10 0.29 0.27 0.05 0.0 0.81 SE 0.0 0.02 0.08 0.14 0.39 0.33 0.03 0.0 0.99 SSE 0.0 0.02 0.06 0.09 0.30 0.21 0.02 0.0 0.70 S 0.0 0.04 0.05 0.08 0.17 0.17 0.01 0.0 0.52 SSW 0.0 0.02 0.06 0.06 0.19 0.29 0.10 0.0 0.72 SW 0.0 0.02 0.03 0.05 0.18 0.24 0.06 0.0 0.58 WSW 0.0 0.01 0.03 0.06 0.09 0.14 0.0 0.0 0.33 W 0.01 0.0 0.02 0.02 0.09 0.04 0.0 0.0 0.18 WNW 0.0 0.02 0.03 0.03 0.04 0.04 0.0 0.0 0.16 NW 0.0 0.02 0.05 0.02 0.04 0.0 0.0 0.0 0.13 NNW 0.0 0.03 0.02 0.02 0.03 0.07 0.0 0.0 0.17 SUBTOTAL 0.01 0.29 0.65 0.83 2.64 3.19 0.90 0.0 8.51 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS F 2183 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS F 2167 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.1 MPH

BFN-16 TABLE 2.3-29 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS G (DELTA T> 4.0 C/100 M) JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.01 0.01 0.0 0.0 0.02 0.02 0.01 0.0 0.07 NNE 0.0 0.0 0.0 0.0 0.05 0.06 0.02 0.0 0.13 NE 0.0 0.01 0.0 0.01 0.04 0.10 0.05 0.0 0.21 ENE 0.0 0.0 0.01 0.01 0.07 0.08 0.03 0.0 0.20 E 0.0 0.01 0.01 0.02 0.05 0.03 0.01 0.0 0.13 ESE 0.0 0.01 0.02 0.02 0.06 0.04 0.01 0.0 0.16 SE 0.0 0.0 0.01 0.05 0.12 0.03 0.0 0.0 0.21 SSE 0.0 0.0 0.03 0.04 0.18 0.06 0.0 0.0 0.31 S 0.0 0.01 0.01 0.06 0.12 0.04 0.0 0.0 0.24 SSW 0.0 0.0 0.01 0.03 0.08 0.05 0.01 0.0 0.18 SW 0.0 0.0 0.02 0.03 0.08 0.05 0.04 0.0 0.22 WSW 0.0 0.0 0.01 0.01 0.02 0.02 0.0 0.0 0.06 W 0.0 0.0 0.01 0.0 0.01 0.0 0.0 0.0 0.02 WNW 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.0 0.03 NW 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.03 NNW 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 SUBTOTAL 0.01 0.08 0.18 0.29 0.90 0.58 0.18 0.0 2.22 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 25729 TOTAL HOURS OF STABILITY CLASS G 574 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS G 568 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 45.30 AND 89.59 METERS WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 11.0 MPH

BFN-16 TABLE 2.3-30 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JAN 1, 77 - DEC 31, 79 WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.02 0.19 0.38 0.64 2.07 2.47 0.61 0.06 6.44 NNE 0.0 0.13 0.33 0.60 2.46 2.69 0.50 0.04 6.75 NE 0.0 0.12 0.35 0.64 2.16 1.85 0.58 0.02 5.72 ENE 0.02 0.14 0.32 0.36 1.15 0.95 0.34 0.04 3.32 E 0.0 0.22 0.47 0.45 0.99 0.43 0.08 0.01 2.65 ESE 0.01 0.23 0.53 0.66 1.79 1.63 0.42 0.09 5.36 SE 0.02 0.36 1.26 1.36 3.25 3.20 1.54 0.69 11.68 SSE 0.01 0.38 1.20 1.22 2.97 2.59 1.16 0.59 10.12 S 0.02 0.40 0.90 1.05 2.53 2.40 1.03 0.43 8.76 SSW 0.0 0.31 0.65 0.69 1.73 1.77 0.73 0.19 6.07 SW 0.02 0.38 0.66 0.69 1.55 1.62 0.50 0.14 5.56 WSW 0.01 0.26 0.69 0.68 1.15 1.05 0.36 0.17 4.37 W 0.02 0.20 0.66 0.81 1.76 1.04 0.42 0.35 5.26 WNW 0.01 0.17 0.46 0.69 2.03 1.54 0.76 0.30 5.96 NW 0.02 0.19 0.49 0.70 1.80 2.01 0.96 0.28 6.45 NNW 0.01 0.22 0.28 0.41 1.66 2.13 0.70 0.13 5.54 SUBTOTAL 0.19 3.90 9.63 11.65 31.05 29.37 10.69 3.53 100.01 TOTAL HOURS OF VALID WIND OBSERVATIONS 25784 TOTAL HOURS OF OBSERVATIONS 26280 RECOVERABILITY PERCENTAGE 98.1 TOTAL HOURS CALM 2 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.0 MPH

BFN-16 TABLE 2.3-31 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JANUARY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.34 0.58 0.87 1.69 1.74 0.10 0.0 5.32 NNE 0.0 0.15 0.63 1.21 2.90 1.21 0.0 0.0 6.10 NE 0.0 0.05 0.39 0.92 2.32 0.73 0.0 0.0 4.41 ENE 0.0 0.05 0.58 0.44 1.16 0.48 0.0 0.0 2.71 E 0.0 0.19 0.48 0.63 0.87 0.05 0.0 0.0 2.22 ESE 0.0 0.10 0.29 0.92 1.84 1.89 0.34 0.05 5.43 SE 0.0 0.10 0.34 1.16 2.51 2.80 2.61 1.06 10.58 SSE 0.0 0.15 0.29 0.77 1.60 2.27 0.92 0.29 6.29 S 0.0 0.19 0.29 0.58 1.89 1.64 0.58 0.10 5.27 SSW 0.0 0.19 0.24 0.15 0.48 0.44 0.05 0.0 1.55 SW 0.0 0.15 0.19 0.29 0.77 0.58 0.05 0.15 2.18 WSW 0.0 0.24 0.29 0.34 0.63 0.34 0.39 0.44 2.67 W 0.0 0.19 0.29 0.44 1.60 1.55 0.97 1.55 6.59 WNW 0.0 0.10 0.44 0.29 2.66 4.55 1.93 1.50 11.47 NW 0.05 0.19 0.73 0.87 2.13 5.95 3.53 0.77 14.22 NNW 0.0 0.34 0.15 0.73 3.34 5.90 1.74 0.87 13.07 SUBTOTAL 0.05 2.72 6.20 10.61 28.39 32.12 13.21 6.78 100.08 TOTAL HOURS OF VALID WIND OBSERVATIONS 2068 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 92.7 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 13.4 MPH

BFN-16 TABLE 2.3-32 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS FEBRUARY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.15 0.51 1.02 3.63 3.78 1.07 0.0 10.16 NNE 0.0 0.20 0.20 0.66 4.19 4.55 0.26 0.05 10.11 NE 0.0 0.15 0.51 0.77 2.91 1.53 0.05 0.05 5.97 ENE 0.0 0.26 0.36 0.56 1.33 1.12 0.10 0.0 3.73 E 0.0 0.05 0.61 0.56 1.43 0.66 0.10 0.0 3.41 ESE 0.0 0.15 0.51 0.46 1.18 1.18 0.0 0.05 3.53 SE 0.0 0.36 0.66 1.07 1.38 1.99 0.56 0.15 6.17 SSE 0.0 0.05 0.61 0.41 1.84 2.15 1.12 1.12 7.30 S 0.05 0.31 0.46 0.46 1.79 1.58 1.23 0.66 6.54 SSW 0.0 0.15 0.46 0.56 1.64 1.48 0.87 0.31 5.47 SW 0.0 0.26 0.36 0.41 1.23 1.99 0.92 0.41 5.58 WSW 0.0 0.20 0.36 0.31 0.41 0.92 1.02 0.82 4.04 W 0.0 0.10 0.20 0.31 0.92 0.92 0.87 0.41 3.73 WNW 0.0 0.0 0.36 0.41 1.64 1.89 1.53 0.46 6.29 NW 0.05 0.20 0.72 0.61 2.10 3.17 1.69 0.77 9.31 NNW 0.0 0.20 0.26 0.61 3.22 2.71 1.43 0.15 8.58 SUBTOTAL 0.10 2.79 7.15 9.19 30.84 31.62 12.82 5.41 99.92 TOTAL HOURS OF VALID WIND OBSERVATIONS 1956 TOTAL HOURS OF OBSERVATIONS 2016 RECOVERABILITY PERCENTAGE 97.0 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 13.1 MPH

BFN-16 TABLE 2.3-33 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS MARCH (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.14 0.23 0.41 1.04 1.99 1.27 0.05 5.13 NNE 0.0 0.14 0.23 0.27 2.08 2.35 0.41 0.0 5.48 NE 0.05 0.09 0.18 0.54 1.90 1.90 0.50 0.0 5.16 ENE 0.0 0.14 0.23 0.09 1.00 0.81 0.23 0.0 2.50 E 0.0 0.05 0.18 0.23 0.68 0.23 0.0 0.0 1.37 ESE 0.0 0.14 0.27 0.36 1.40 1.36 0.59 0.32 4.44 SE 0.0 0.05 0.59 1.09 2.63 2.81 1.54 2.94 11.65 SSE 0.0 0.14 0.72 0.81 2.81 2.49 1.49 2.49 10.95 S 0.0 0.14 0.27 0.72 1.63 4.03 2.04 0.86 9.69 SSW 0.0 0.14 0.36 0.32 1.67 3.62 1.95 0.32 8.38 SW 0.0 0.05 0.59 0.45 1.18 1.36 0.41 0.32 4.36 WSW 0.0 0.27 0.45 0.50 0.91 2.08 0.77 0.27 5.25 W 0.0 0.0 0.54 0.41 1.67 1.40 0.91 0.45 5.38 WNW 0.0 0.09 0.27 0.41 1.72 2.90 1.22 0.23 6.84 NW 0.0 0.23 0.23 0.77 2.13 3.44 0.91 0.45 8.16 NNW 0.0 0.32 0.09 0.23 1.18 2.44 1.04 0.05 5.35 SUBTOTAL 0.05 2.13 5.43 7.61 25.63 35.21 15.28 8.75 100.09 TOTAL HOURS OF VALID WIND OBSERVATIONS 2209 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.0 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 14.6 MPH

BFN-16 TABLE 2.3-34 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS APRIL (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.05 0.19 0.19 0.47 1.40 1.45 0.28 0.19 4.22 NNE 0.0 0.14 0.09 0.47 1.31 1.87 0.47 0.0 4.35 NE 0.0 0.19 0.37 0.70 2.80 2.19 0.75 0.05 7.05 ENE 0.0 0.05 0.23 0.19 1.59 1.31 0.42 0.05 3.84 E 0.0 0.09 0.47 0.65 1.17 0.37 0.09 0.0 2.84 ESE 0.0 0.37 0.75 1.07 1.73 0.56 0.61 0.05 5.14 SE 0.0 0.47 1.96 1.35 2.80 2.47 2.33 0.79 12.17 SSE 0.0 0.47 1.45 0.93 2.29 2.01 1.59 0.84 9.58 S 0.05 0.61 0.98 0.89 1.49 1.96 1.63 0.33 7.94 SSW 0.0 0.19 0.65 0.65 1.68 1.87 2.33 0.23 7.60 SW 0.0 0.47 0.75 0.79 1.03 2.38 1.26 0.09 6.77 WSW 0.0 0.23 0.51 0.42 0.93 1.73 0.98 0.0 4.80 W 0.0 0.05 0.37 0.47 1.35 1.31 0.75 0.84 5.14 WNW 0.0 0.19 0.47 0.51 2.43 1.35 1.12 0.37 6.44 NW 0.0 0.14 0.28 0.51 2.10 2.10 1.54 0.19 6.86 NNW 0.0 0.19 0.23 0.33 1.12 2.57 0.84 0.0 5.28 SUBTOTAL 0.10 4.04 9.75 10.40 27.22 27.50 16.99 4.02 100.02 TOTAL HOURS OF VALID WIND OBSERVATIONS 2142 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.2 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.9 MPH

BFN-16 TABLE 2.3-35 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS MAY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.18 0.36 0.68 2.17 1.13 0.23 0.0 4.75 NNE 0.0 0.14 0.36 0.32 2.48 2.80 0.50 0.0 6.60 NE 0.0 0.36 0.23 0.41 1.94 1.31 1.04 0.0 5.29 ENE 0.0 0.23 0.36 0.27 0.95 0.63 0.59 0.0 3.03 E 0.0 0.45 0.81 0.59 1.49 0.68 0.09 0.0 4.11 ESE 0.0 0.36 0.45 0.81 2.98 1.99 0.36 0.05 7.00 SE 0.0 0.36 1.58 2.03 3.48 4.47 0.50 0.05 12.47 SSE 0.0 0.45 1.58 1.45 2.30 1.81 1.08 0.14 8.81 S 0.09 0.59 1.08 1.08 3.03 2.03 1.22 0.18 9.30 SSW 0.0 0.50 0.99 0.95 1.63 1.81 0.32 0.18 6.38 SW 0.05 0.54 0.90 0.54 2.26 2.26 0.45 0.09 7.09 WSW 0.0 0.18 0.77 1.40 1.13 1.04 0.27 0.0 4.79 W 0.05 0.32 0.50 1.13 1.94 0.72 0.50 0.18 5.34 WNW 0.0 0.27 0.45 0.95 2.35 0.86 0.54 0.54 5.96 NW 0.0 0.09 0.77 0.77 1.76 1.08 0.45 0.09 5.01 NNW 0.0 0.41 0.45 0.32 1.26 0.95 0.63 0.09 4.11 SUBTOTAL 0.19 5.43 11.64 13.70 33.15 25.57 8.77 1.59 100.04 TOTAL HOURS OF VALID WIND OBSERVATIONS 2214 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.2 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 10.9 MPH

BFN-16 TABLE 2.3-36 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JUNE (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.05 0.23 0.19 0.65 2.71 2.38 1.03 0.14 7.38 NNE 0.0 0.14 0.19 0.37 2.24 2.06 0.79 0.05 5.84 NE 0.0 0.09 0.42 0.47 1.36 1.31 0.56 0.0 4.21 ENE 0.05 0.05 0.19 0.33 0.98 0.75 0.23 0.0 2.58 E 0.05 0.37 0.47 0.47 0.51 0.09 0.0 0.0 1.96 ESE 0.0 0.37 0.56 0.70 0.93 0.65 0.0 0.0 3.21 SE 0.05 0.47 1.68 1.50 3.32 2.62 0.23 0.0 9.87 SSE 0.0 0.47 1.36 1.50 2.38 1.68 0.09 0.0 7.48 S 0.0 0.70 1.68 1.36 2.71 2.10 0.23 0.0 8.78 SSW 0.0 0.37 1.17 0.98 2.01 1.96 0.33 0.0 6.82 SW 0.0 1.07 1.36 1.40 2.85 1.64 0.28 0.0 8.60 WSW 0.0 0.61 1.17 1.07 2.34 2.20 0.14 0.0 7.53 W 0.0 0.37 1.45 1.54 3.79 1.82 0.0 0.0 8.97 WNW 0.05 0.19 0.70 1.59 3.69 0.89 0.19 0.0 7.30 NW 0.05 0.42 0.79 1.03 2.43 1.03 0.05 0.0 5.80 NNW 0.0 0.28 0.28 0.56 1.68 0.84 0.05 0.0 3.69 SUBTOTAL 0.30 6.20 13.66 15.52 35.93 24.02 4.20 0.19 100.02 TOTAL HOURS OF VALID WIND OBSERVATIONS 2140 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.1 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 9.8 MPH

BFN-16 TABLE 2.3-37 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS JULY (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.18 0.32 0.59 1.68 0.36 0.0 0.0 3.13 NNE 0.0 0.05 0.23 0.45 2.45 2.91 0.27 0.09 6.45 NE 0.0 0.09 0.05 0.59 1.45 2.36 0.14 0.0 4.68 ENE 0.0 0.09 0.36 0.41 1.18 1.09 0.41 0.0 3.54 E 0.0 0.09 0.32 0.50 1.14 0.45 0.23 0.05 2.78 ESE 0.05 0.23 0.64 0.45 1.91 2.27 0.05 0.09 5.69 SE 0.0 0.36 1.95 1.41 2.95 3.00 1.14 0.27 11.08 SSE 0.0 0.45 1.91 1.36 3.23 2.64 0.73 0.09 10.41 S 0.0 0.55 1.32 2.09 3.73 2.04 0.09 0.23 10.05 SSW 0.0 0.59 0.77 1.04 2.54 1.77 0.23 0.09 7.03 SW 0.0 0.45 0.95 0.95 2.82 2.32 0.36 0.05 7.90 WSW 0.0 0.41 1.86 1.95 2.41 0.95 0.18 0.0 7.76 W 0.0 0.18 1.73 1.86 3.95 0.82 0.05 0.0 8.59 WNW 0.0 0.18 0.45 0.91 2.50 0.55 0.09 0.0 4.68 NW 0.0 0.14 0.23 0.73 1.82 0.59 0.14 0.0 3.65 NNW 0.05 0.14 0.23 0.41 1.23 0.50 0.05 0.0 2.61 SUBTOTAL 0.10 4.18 13.32 15.70 36.99 24.62 4.16 0.96 100.03 TOTAL HOURS OF VALID WIND OBSERVATIONS 2201 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 98.6 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 10.2 MPH

BFN-16 TABLE 2.3-38 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS AUGUST (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.09 0.77 0.59 1.67 1.44 0.14 0.0 4.70 NNE 0.0 0.14 0.63 0.59 2.26 2.07 0.23 0.05 5.97 NE 0.0 0.09 0.50 0.36 2.07 1.76 0.54 0.0 5.32 ENE 0.05 0.23 0.36 0.36 0.86 0.77 0.18 0.0 2.81 E 0.0 0.50 0.72 0.41 0.59 0.14 0.05 0.0 2.41 ESE 0.0 0.18 0.54 0.45 1.22 1.22 0.14 0.0 3.75 SE 0.09 0.41 2.17 1.89 5.19 2.80 0.50 0.0 13.05 SSE 0.05 0.81 2.80 3.38 5.37 2.93 0.27 0.0 15.61 S 0.0 0.45 1.85 2.39 4.60 2.21 0.18 0.05 11.73 SSW 0.0 0.41 1.53 1.49 2.71 1.17 0.41 0.05 7.77 SW 0.09 0.81 1.17 1.31 2.93 1.67 0.23 0.0 8.21 WSW 0.09 0.36 1.22 0.54 1.35 0.50 0.0 0.0 4.06 W 0.05 0.41 1.26 1.94 1.31 0.23 0.0 0.0 5.20 WNW 0.05 0.14 0.72 0.68 1.89 0.09 0.09 0.0 3.66 NW 0.0 0.23 0.59 1.08 1.17 0.14 0.14 0.0 3.35 NNW 0.0 0.18 0.45 0.41 0.90 0.50 0.05 0.0 2.49 SUBTOTAL 0.47 5.44 17.28 17.87 36.09 19.64 3.15 0.15 100.09 TOTAL HOURS OF VALID WIND OBSERVATIONS 2217 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.3 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 9.1 MPH

BFN-16 TABLE 2.3-39 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEPT. (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.09 0.19 0.42 1.13 2.92 3.82 0.71 0.14 9.42 NNE 0.0 0.14 0.47 1.04 2.83 3.58 1.41 0.28 9.75 NE 0.0 0.14 0.33 1.08 2.97 4.05 1.08 0.09 9.74 ENE 0.0 0.19 0.24 0.52 1.70 1.79 1.13 0.42 5.99 E 0.0 0.19 0.28 0.42 1.60 1.46 0.38 0.09 4.42 ESE 0.0 0.19 0.52 0.71 3.11 3.25 0.61 0.0 8.39 SE 0.0 0.38 1.55 1.46 6.08 4.62 0.33 0.19 14.61 SSE 0.0 0.57 1.79 1.84 5.32 2.21 0.38 0.42 12.53 S 0.0 0.09 1.08 0.99 2.36 2.21 0.19 0.14 7.06 SSW 0.0 0.47 0.33 0.47 1.08 1.32 0.52 0.0 4.19 SW 0.0 0.09 0.52 0.28 0.24 0.57 0.19 0.0 1.89 WSW 0.0 0.19 0.38 0.33 0.66 0.14 0.05 0.0 1.75 W 0.0 0.09 0.47 0.24 0.85 0.14 0.14 0.05 1.98 WNW 0.0 0.19 0.38 0.28 0.89 0.09 0.05 0.05 1.93 NW 0.0 0.05 0.42 0.52 1.65 0.38 0.33 0.05 3.40 NNW 0.0 0.0 0.33 0.24 1.51 0.71 0.09 0.09 2.97 SUBTOTAL 0.09 3.16 9.51 11.55 35.77 30.34 7.59 2.01 100.02 TOTAL HOURS OF VALID WIND OBSERVATIONS 2123 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 98.3 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 11.5 MPH

BFN-16 TABLE 2.3-40 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS OCTOBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.22 0.36 0.36 2.29 4.85 1.21 0.0 9.29 NNE 0.0 0.13 0.36 0.67 2.07 3.86 1.08 0.0 8.17 NE 0.0 0.04 0.40 0.58 1.93 1.48 0.76 0.04 5.23 ENE 0.0 0.09 0.22 0.31 1.12 0.81 0.13 0.04 2.72 E 0.0 0.18 0.36 0.27 0.72 0.27 0.0 0.0 1.80 ESE 0.0 0.18 0.67 0.94 2.29 1.66 0.90 0.22 6.86 SE 0.04 0.72 1.03 1.57 4.63 3.86 2.02 1.30 15.17 SSE 0.0 0.31 0.49 0.63 2.88 2.25 1.12 0.13 7.81 S 0.0 0.18 0.40 0.40 2.16 2.61 1.03 0.27 7.05 SSW 0.0 0.22 0.31 0.40 1.84 2.47 0.81 0.0 6.05 SW 0.0 0.22 0.36 0.58 1.17 1.84 1.08 0.18 5.43 WSW 0.0 0.13 0.45 0.31 0.54 0.45 0.13 0.04 2.05 W 0.04 0.22 0.31 0.58 1.21 0.94 0.22 0.0 3.52 WNW 0.0 0.27 0.49 1.12 1.89 1.39 0.27 0.0 5.43 NW 0.0 0.22 0.27 0.49 1.93 2.47 0.72 0.13 6.23 NNW 0.0 0.13 0.22 0.36 1.48 3.73 0.99 0.09 7.00 SUBTOTAL 0.08 3.46 6.70 9.57 30.15 34.94 12.47 2.44 99.81 TOTAL HOURS OF VALID WIND OBSERVATIONS 2226 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 99.7 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.7 MPH

BFN-16 TABLE 2.3-41 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS NOVEMBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.09 0.23 0.65 0.79 2.34 3.55 0.47 0.0 8.12 NNE 0.05 0.14 0.33 0.89 3.50 3.22 0.33 0.0 8.46 NE 0.0 0.09 0.42 0.93 2.66 2.90 1.17 0.0 8.17 ENE 0.09 0.28 0.37 0.65 1.17 1.03 0.42 0.0 4.01 E 0.0 0.23 0.51 0.23 1.12 0.65 0.05 0.0 2.79 ESE 0.05 0.19 0.75 0.47 1.45 1.82 1.26 0.14 6.13 SE 0.0 0.37 0.84 0.98 2.15 3.04 2.48 0.75 10.61 SSE 0.05 0.37 0.75 0.75 2.76 3.36 2.62 0.56 11.22 S 0.0 0.47 0.84 1.03 2.71 2.29 1.36 1.03 9.73 SSW 0.0 0.14 0.51 0.75 1.12 0.84 0.0 0.37 3.73 SW 0.0 0.28 0.23 0.51 0.47 0.89 0.14 0.28 2.80 WSW 0.0 0.19 0.42 0.37 1.26 0.89 0.14 0.0 3.27 W 0.09 0.23 0.33 0.37 0.75 0.79 0.19 0.51 3.26 WNW 0.0 0.28 0.42 0.79 1.31 1.45 0.98 0.19 5.42 NW 0.05 0.28 0.61 0.51 1.40 1.64 1.07 0.23 5.79 NNW 0.09 0.33 0.42 0.47 2.10 2.38 0.56 0.05 6.40 SUBTOTAL 0.56 4.10 8.40 10.49 28.27 30.74 13.24 4.11 99.91 TOTAL HOURS OF VALID WIND OBSERVATIONS 2140 TOTAL HOURS OF OBSERVATIONS 2160 RECOVERABILITY PERCENTAGE 99.1 TOTAL HOURS CALM 1 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 12.5 MPH

BFN-16 TABLE 2.3-42 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS DECEMBER (77, 78, 79) WIND WIND SPEED (MPH) DIRECTION 0.6-1.4 1.5-3.4 3.5-5.4 5.5-7.4 7.5-12.4 12.5-18.4 18.5-24.4 >=24.5 TOTAL N 0.0 0.09 0.05 0.19 1.44 3.35 0.79 0.19 6.10 NNE 0.0 0.05 0.19 0.28 1.44 1.91 0.19 0.0 4.06 NE 0.0 0.0 0.37 0.37 1.72 0.65 0.28 0.0 3.39 ENE 0.0 0.05 0.33 0.23 0.84 0.79 0.19 0.0 2.43 E 0.0 0.19 0.47 0.51 0.65 0.14 0.0 0.0 1.96 ESE 0.0 0.33 0.37 0.56 1.35 1.68 0.19 0.09 4.57 SE 0.05 0.23 0.70 0.74 1.63 3.77 4.19 0.65 11.96 SSE 0.0 0.28 0.51 0.65 2.75 5.26 2.51 1.07 13.03 S 0.0 0.47 0.51 0.56 2.14 4.00 2.61 1.40 11.69 SSW 0.0 0.37 0.42 0.47 2.23 2.37 0.93 0.74 7.53 SW 0.05 0.19 0.47 0.65 1.49 1.91 0.61 0.09 5.46 WSW 0.0 0.05 0.28 0.51 1.12 1.35 0.28 0.51 4.10 W 0.0 0.19 0.37 0.33 1.68 1.91 0.51 0.28 5.27 WNW 0.05 0.14 0.33 0.33 1.35 2.70 1.21 0.37 6.48 NW 0.0 0.05 0.33 0.51 0.98 2.42 1.16 0.74 6.19 NNW 0.0 0.09 0.28 0.37 1.21 2.61 1.02 0.23 5.81 SUBTOTAL 0.15 2.77 5.98 7.26 24.02 36.82 16.67 6.36 100.03 TOTAL HOURS OF VALID WIND OBSERVATIONS 2148 TOTAL HOURS OF OBSERVATIONS 2232 RECOVERABILITY PERCENTAGE 96.2 TOTAL HOURS CALM 0 ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY: BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL MEAN WIND SPEED = 14.2 MPH

BFN-16 TABLE 2.3-43 (Sheet 1) WIND DIRECTION PERSISTENCE DATA 10 M Level DISREGARDING STABILITY (JAN 1, 77 - DEC 31, 79) LOST RECORD (%) = 1.79 PERSISTENCE WIND DIRECTION ACC. ACC. (HOURS) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM TOTAL TOTAL FREQUENCY 2 177 164 105 113 93 118 181 202 149 70 35 102 129 96 91 128 1 1954 4968 100.00 3 92 91 51 35 43 48 131 93 90 25 23 53 63 45 47 61 0 991 3014 60.67 4 50 54 31 24 26 35 90 67 65 5 6 24 26 29 37 36 0 605 2023 40.72 5 31 37 20 17 15 18 79 56 26 1 1 17 22 18 27 32 0 417 1418 28.54 6 39 27 11 8 14 17 49 26 36 0 1 6 12 17 13 29 0 305 1001 20.15 7 12 21 12 2 6 10 29 20 23 0 2 9 6 9 9 15 0 185 696 14.01 8 7 11 8 2 7 5 26 10 7 1 1 3 7 8 9 13 0 125 511 10.29 9 6 5 5 3 3 4 24 4 9 0 0 5 4 9 8 10 0 99 386 7.77 10 4 5 6 0 5 3 13 4 5 0 0 1 1 4 4 6 0 61 287 5.78 11 3 9 3 0 2 1 13 3 2 0 0 0 4 2 3 6 0 51 226 4.55 12 2 7 3 0 1 0 11 4 3 0 0 0 1 2 3 7 0 44 175 3.52 13 3 11 0 2 1 3 5 2 3 0 0 0 1 1 1 1 0 34 131 2.64 14 1 0 2 0 0 1 5 2 0 0 0 1 1 0 2 2 0 17 97 1.95 15 0 5 0 0 0 0 6 0 2 0 0 1 2 3 0 1 0 20 80 1.61 16 2 1 0 0 0 0 2 2 1 0 0 0 0 1 0 3 0 12 60 1.21 17 2 3 0 0 0 0 3 0 0 0 0 0 0 1 0 1 0 10 48 0.97 18 0 2 0 1 0 0 1 0 1 0 0 0 1 0 0 0 0 6 38 0.76 19 0 0 0 0 0 0 3 0 3 0 0 0 0 1 0 0 0 7 32 0.64 20 1 1 0 0 0 0 3 0 1 0 0 0 0 0 1 0 0 7 25 0.50 21 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 18 0.36 22 0 2 0 0 0 0 0 0 1 0 0 0 0 0 0 3 0 6 17 0.34 23 0 0 0 0 0 0 1 0 0 0 0 0 1 0 0 0 0 2 11 0.22 24 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 1 9 0.18 25 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 0.16 26 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 8 0.16 27 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7 0.14 28 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 7 0.14 29 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 7 0.14 30 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 6 0.12

BFN-16 TABLE 2.3-43 (Continued) (Sheet 2) WIND DIRECTION PERSISTENCE DATA 10 M Level DISREGARDING STABILITY (JAN 1, 77 - DEC 31, 79) PERSISTENCE WIND DIRECTION ACC. ACC. (HOURS) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM TOTAL TOTAL FREQUENCY 31 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 5 0.10 32 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 1 0 2 4 0.08

         >32      0       0       0      0       0       0      2         0         0       0   0   0   0   0    0       0  0      2   2      0.04 TOTAL            432     458     257    207     216     263    680       495      429      102  69 222 281 246  255    355   1   4968 MAXIMUM PERSISTENCE         20    31      14     18      13      14     36        16       32        8   8  15  23  19   20      32  2 (HOURS) 50.0%       3     3       3      2       3       3      4         3         3       2   2   3   3   3    3       3  2 80.0%       5     6       5      4       5       5      7         5         6       3   3   4   5   6    6       6  2 90.0%       6    10       8      5       7       7     10         7         7       3   4   6   7   8    8       9  2 99.0%      16    20      12     13      11      13     20        13       19        5   8  10  15  16   14      22  2 99.9%      20    31      14     18      13      14     36        16       32        8   8  15  23  19   20      32  2 METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 10.00 METER LEVEL

BFN-16 TABLE 2.3-44 (Sheet 1) WIND DIRECTION PERSISTENCE DATA 93 M LEVEL DISREGARDING STABILITY (JAN 1, 77 - DEC 31, 79) LOST RECORD (%) = 1.89 PERSISTENCE WIND DIRECTION ACC. ACC. (HOURS) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM TOTAL TOTAL FREQUENCY 2 105 125 97 82 83 108 159 193 145 134 109 114 116 112 112 89 0 1883 5070 100.00 3 54 62 52 40 27 59 98 98 100 76 62 45 50 53 72 58 0 1006 3187 62.86 4 46 31 40 33 16 29 64 61 49 42 36 17 36 47 37 35 0 619 2181 43.02 5 27 32 35 15 13 20 50 50 38 27 20 24 25 38 22 20 0 456 1562 30.81 6 16 16 22 14 3 13 39 29 26 13 23 10 14 23 19 23 0 303 1106 21.81 7 13 9 15 6 4 6 27 31 18 12 8 6 9 11 14 10 0 199 803 15.84 8 5 12 11 3 4 3 27 13 10 8 9 2 8 7 11 9 0 142 604 11.91 9 9 12 6 0 0 11 11 12 11 9 6 3 3 9 6 9 0 117 462 9.11 10 4 3 5 2 0 2 11 4 5 2 4 4 4 4 10 5 0 69 345 6.80 11 3 4 5 1 1 7 17 6 11 1 1 1 3 3 2 4 0 70 276 5.44 12 2 6 3 0 0 2 4 6 6 2 0 1 2 1 5 2 0 42 206 4.06 13 7 1 0 0 0 5 6 4 4 1 2 0 2 0 3 2 0 37 164 3.23 14 4 5 2 0 0 2 2 2 2 0 1 1 1 1 3 0 0 26 127 2.50 15 3 1 1 0 0 0 3 3 1 2 0 1 0 1 2 2 0 20 101 1.99 16 3 1 2 0 0 1 2 3 1 1 1 0 0 1 0 1 0 17 81 1.60 17 1 2 1 0 0 1 3 1 2 0 0 0 0 0 1 1 0 13 64 1.26 18 0 1 0 0 0 0 3 1 0 0 1 0 0 0 0 0 0 6 51 1.01 19 1 2 0 0 0 0 0 1 0 0 0 0 0 2 1 0 0 7 45 0.89 20 1 1 0 0 0 0 2 0 0 0 0 0 1 0 0 0 0 5 38 0.75 21 0 1 0 0 0 0 0 1 2 0 0 0 0 1 1 0 0 6 33 0.65 22 3 1 0 0 0 0 0 0 1 0 0 0 1 0 1 1 0 8 27 0.53 23 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 1 19 0.37 24 1 0 0 0 0 0 1 0 0 0 0 0 0 0 1 0 0 3 18 0.36 25 0 1 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 2 15 0.30 26 0 0 0 0 0 1 3 1 0 0 0 0 0 0 0 0 0 5 13 0.26 27 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 8 0.16 28 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 7 0.14 29 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 6 0.12

BFN-16 TABLE 2.3-44 (Continued) (Sheet 2) WIND DIRECTION PERSISTENCE DATA 93 M LEVEL DISREGARDING STABILITY (JAN 1, 77 - DEC 31, 79) PERSISTENCE WIND DIRECTION ACC. ACC. (HOURS) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM TOTAL TOTAL FREQUENCY 30 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 6 0.12 31 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 1 5 0.10 32 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 1 4 0.08

 >32              0      0      0       0      0       0       2        1        0        0         0   0   0    0   0    0      0      3    3   0.06 TOTAL         308     330    297     196    151     270     537     523       433      330      283  229 275  314 323  271      0   5070 MAXIMUM PERSISTENCE      24      28     17      11     11      26      36      38        32       16        18  15  22   21  24   22      1 (HOURS) 50.0%              3      3      3       3      2       3       4        3        3        3         3   3   3    3   3    3      1 80.0%              6      6      6       5      4       6       7        6        6        5         5   5   5    6   6    6      1 90.0%            10       9      8       6      5       9      10        8        9        7         7   6   7    7   9    8      1 99.0%            22      21     16      10      8      16      26      19        17       13       14   12  14   16  19   16      1 99.9%            24      28     17      11     11      26      36      38        32       16       18   15  22   21  24   22      1 METEOROLOGICAL FACILITY: MET FACILITY LOCATED ABOUT 0.5 MI ESE OF BROWNS FERRY NUCLEAR PLANT WIND SPEED AND DIRECTION MEASURED AT THE 93.00 METER LEVEL

BFN-16 Table 2.3-45 TEMPERATURE DATA DECATUR, ALABAMA 1879 - 1958 Avg. Avg. Extreme Extreme Avg. Max. Min. Max. Min. Temp. Temp. Temp. Temp. Temp. Month (°F) (°F) (°F) (°F) (°F) December 43.7 53.0 34.3 78 1 January 42.9 52.3 33.4 79 -5 February 44.6 54.8 34.4 84 -12 Winter 43.7 53.4 34.0 -- -- March 53.1 64.1 42.1 93 4 April 61.8 73.2 50.3 92 26 May 70.4 81.8 59.0 100 34 Spring 61.8 73.0 50.4 -- -- June 78.2 89.3 67.1 108 44 July 80.7 91.2 70.1 107 52 August 79.9 90.6 69.1 107 52 Summer 79.6 90.4 68.8 -- -- September 74.6 85.9 63.3 104 37 October 63.0 75.2 50.8 100 26 November 51.2 62.3 40.1 86 3 Fall 62.9 74.5 51.4 -- -- Annual 62.0 72.8 51.2 -- -- "The Climate of Decatur, Alabama" (1879-1958), Long, Arthur R., U.S. Weather Bureau State Climatologist for Alabama, Weather Bureau Office, Montgomery, Alabama, July 1959.

BFN-16 Table 2.3-46 TEMPERATURE DATA BROWNS FERRY NUCLEAR PLANT January 1, 1977 - December 31, 1979 Avg. Avg. Extreme Extreme Avg. Max. Min. Max. Min. Temp. Temp. Temp. Temp. Temp. Month (°F) (°F) (°F) (°F) (°F) December 42.9 50.9 35.0 73 15 January 30.5 36.5 23.7 60 -1 February 38.0 46.1 30.3 75 13 Winter 37.1 44.5 29.7 -- -- March 52.0 60.2 43.9 81 18 April 62.4 70.9 53.7 83 38 May 68.7 76.3 60.8 88 44 Spring 61.0 69.1 52.8 -- -- June 75.8 83.6 68.0 94 54 July 78.8 86.6 72.0 96 64 August 77.7 86.0 70.4 93 61 Summer 77.4 85.4 70.1 -- -- September 72.3 79.9 65.5 92 54 October 59.3 68.9 50.0 83 36 November 53.1 60.8 45.4 75 22 Fall 61.6 69.9 53.6 -- -- Annual 59.3 67.2 51.6 -- -- Temperature data measurements at about 33 feet. Meteorological facility located about 0.5 miles from Browns Ferry Nuclear Plant.

BFN-16 Table 2.3-47 PRECIPITATION DATA HUNTSVILLE, ALABAMA 1968-1980 1968-1980 1968-1980 1968-1980 Average Number 1941-1970 Extreme Extreme Maximum In of Days With Normals Monthly Max. Monthly Min. 24-Hours Month 0.01 Inch or More (Inches) (Inches) (Inches) (Inches) December 11 5.4 9.9 0.9 5.8 January 12 5.1 10.5 1.7 3.0 February 9 5.2 9.6 0.6 3.9 Winter 32 15.7 March 13 5.8 17.0 3.0 7.7 April 9 4.8 9.1 1.8 3.6 May 11 3.9 9.1 3.1 5.7 Spring 33 14.5 June 9 4.0 7.3 0.8 4.5 July 11 4.9 9.4 1.9 4.5 August 9 3.5 4.0 0.9 2.5 Summer 29 12.4 September 9 3.3 9.8 1.8 4.0 October 6 2.6 12.1 0.8 6.0 November 9 3.9 11.5 1.8 3.3 Fall 24 9.8 Annual 119 52.2 "Local Climatological Data," Annual Summary with Comparative Data, Huntsville, Alabama, NOAA, National Climatic Center, Asheville, NC, 1980.

BFN-16 Table 2.3-48 PRECIPITATION DATA January 1, 1977 - December 31, 1979 Average Number Monthly Extreme Extreme Maximum In of Days With Average Monthly Max. Monthly Min. 24-Hours Month 0.01 Inch or More (Inches) (Inches) (Inches) (Inches) December 9 3.6 6.8 1.9 1.3 January 13 3.8 5.8 2.7 1.4 February 6 2.9 4.9 0.2 2.1 Winter 28 10.3 March 12 5.5 8.5 3.6 3.3 April 9 5.2 7.8 1.2 3.2 May 11 3.7 4.9 1.6 1.9 Spring 32 14.4 June 10 3.3 4.0 2.3 2.1 July 11 3.2 4.5 2.3 2.3 August 10 3.9 7.9 1.5 2.0 Summer 31 10.4 September 10 6.2 8.7 1.7 3.7 October 5 2.6 3.9 0.7 1.4 November 11 6.6 8.6 3.4 3.1 Fall 26 15.4 Annual 117 50.5 Meteorological facility located about 0.5 miles from Browns Ferry Nuclear Plant.

BFN-16 Table 2.3-49 SNOWFALL DATA DECATUR, ALABAMA Jan. Feb. Mar. Apr. May June July Aug. Sept. Oct. Nov. Dec. Annual Average snowfall 0.9 0.8 0.2 T 0 0 0 0 0 T 0.2 0.6 2.7 (inches) Average No. Days 1 1 1

• 0 0 0 0 0 *
• 1 6 (trace or more)

Average No. Days * *

• 0 0 0 0 0 0 0 *
• 1 (0.1 inch or more)

T - trace (not measurable)

• Less than one day "The Climate of Decatur, Alabama" (1879-1958), Long, Arthur R., U.S. Weather Bureau State Climatologist for Alabama, Weather Bureau Office, Montgomery, Alabama, July 1959.

( ( C N

                \/IND SPEED (HPHI 18.6-24.1 l2,S-18,1 S.5*7. 1 J.S*S,-4 1.5*!1.1 AMENDMENT 16 OIINS FERRY NUCLEAlt Pl.

Figure 2.3-1 lt,N l'I 11l,O, 11,81 & 16,SI N TEl'P STAIILIT'I CLAS! A JAN t, 77

• DEC 31,711

( ( ( N

               \illND SPEED CMPHI
                                '2,6-11,'4 7,li-12,<t li . 5-7, I ,S*l.1 AMENDMENT 16
           ~NUCLEAR~

Figure 2.3-2 11 ,N K WllCI, 1*.n I ff,IIJ K TElf' STMlLlt'f CLASS I JAN I , 77

• OEC JI , 7111

( ( (

                 'JIN  SPEED (MPH)
                                  >-Z1,ti 11.s-21.1 12,S*ll,4 7,5*12.1 E

l,!*ll,1 AMENDMENT 16 IMtS fEHY NJCUAlt PL. Figure 2.3-3 11,N l'I WIN>, 11,N & .f6,U H TE!f> SfAlltLtTY CLASS C J-,N I , 77

• DEC J1 , 70

( ( C N WINO SPEED IHPHl te,S-Z1,1 7.5-12.4 li.li-7.4 E J.S-5. 1.s-s.1 AMENDMENT 16 F£RtrY NJCLEAR Pl. Figure 2.3-4 II *

• 11 IIJNJ, II.ti & 4'.!le N TE't" STAJtlllTY Ct.All 0 JAN I, 77
• ore St, 711

C ( ( WIND SPEED [HPHI

• 11 . S*2 ,4 11 s.i.-1 . 1 s.s-s .

1. S-5.i 1.e-1 ...

AMENDMENT 16 0~9 F£Rfll IULW PL Figure 2.3-5 te,N H WIICI, 11,1! ~ S.ff" TEii> STAalLIT~ CLASS E JM,t I , 77 - DEC 11 , 7G

( ( (

                   \JINO SPEED IHPHJ 7.C-12, 11 s.r.-1.

E I.H. 1.5-S,4 AMENDMENT 16 CllilHS f!Plfrr IIJCUAA PL Figure 2, 3-6 II ** 11 lllNJ, 11,IS I -+&,II H TaP ITAIIUTY CUSS F JAN I , 77

• OEC !I , 711

C ( WIND SPEED (MPHI 7.li-12.-t

     ,s
                           ,.,-1 .*

J ,5-S.4 AMENDMENT 16 nMY NUClEAII l't. Figure 2.3-7 II.* " lltlCl, 11,n & 1S,P " TElt'

• tTIIIUL ITY CLASS c; J~ 1 , 77
• DEC SI , 71

( ( ( WlNO SPEED IHPH) 7,1.-12,4 11

                            , .s-1.1 I.S-J ...

1 .e-1.1 AMENDMENT 16 0114S n:Mr NJClEAlt Pl. Figure 2.3-8 11,N 11 VIICl ALL STMllllTY Cl.""° JAN 1 , 77 - OR ,. , 711

( ( C

                 \I l ND SPEED U1PH J 1e,s-2*.*
                                . 12.s-1e.*

7.5-12.-t S,S-7.4

s.s-s.*

1,5-3,-t l.!H,4 AMENDMENT 16 Ill/NS FERRY MJCl.EAR PLAHT Figure 2.3-9 11.N 11 IIIN:> ALL SlAalLll1 CLASSES JAHUARY 177,78,191

( ( ( WIND SPEED INPH) 18.S-2t.t lt.S-18.1 7.S-12.1 S.S-7 .* 3.S-5.4 1.5-S.-t 1.11-1.4 AMENDMENT 16

     \INS FERRY NUCLE.-.R PLAPfT Figure 2.3-10 18.IJII H Wit<<>

ALL ST~BILITY Cl~SSES FEBUARY t77,78,7QI

( ( C WIND SPEED IHPHl 18 .S-2-t,t

                               *12.S-18 . i II S . S-7 .

3 . S-5 . 1.5-3.1 AMENDMENT 16 l.tlS FERRY HUCLEAA PLANT Figure 2. 3-11 Ill.Ill HI/IND

   ~LL 51ABILIT1 Cl~SSE:S
       ~ROI <77,18,7111

C ( (

 *)

WIND SPEED IHPH> 18.5-21.1

                                   . 12.s-11.1 7,S--12.1
           \S S.S-7.

3 . 5-5.1 1.S- 5.1 I.IH.1 AMENDMENT 16 Oll!IS FERRY MJCLEAR PlAlfT Figure 2.3-12 11,fltl H IIIIO ALL iTASILITY C1.~SSES APRIL. en ,78 , n,

( ( ( WINO SPEED (MPHl

                              . 12,S*lt,4 S.S-7,4 J.S*S. 4 1 .S*:S.4 1.11*1 .1 AMENDMENT 16 DlnlS FERRY NUCLEAR PLANT Figure 2.3-13 11,811 H IIIHO ALL STASlltTY CLASSES IIAY 177 ,78 ,791

C WIND SPEED IHPH> l8.S-2t,t

                          . 12.S*IB,4 7.S-12,4 I.Sol;.

1.5*3. AMENDMENT 16

   ~S FERRY "-lCLEAR PLOO Figure 2.3-14 11,11 f1 IHIO AL~ SlASILITY Ct.Assn JUI£ 177 ,78 ,79)

C ( ( WIND SPEED IHPHJ 11.S-2-1 **

                               , 12.s-, . ...

7.S-12,4 S.S-7 . s . s -s . 1.S-5.1 AMENDMENT 16 OWNS FERRY MJCl.EAR Pl.ANT Figure 2,3-15 ll!l,811 H VI~ ALL SJA81LIT1 Cl.ASSES JLU' (77,78 ,7'111

( ( (

                    \IIND SPEED IMPHJ
                                 >*Z-1 ,S 11.S*Z-t.1 12 ,S*l1, 7 . 5-12, S . S-7 .

J.s-s ... 1,5*3 ... AMENDMENT 1 6

      ""'S FERRY NUCLEAA PLANT Figure 2,3-lo 111 . 1111 H 11110 ALL SlASlllTY Cl~SSES
     ~UClJST 177,78,7111

( ( C WIND SPEED IHPHJ 18.5-2i,4

                                  . l2.S-18.

1 . s-12 . s.s-1. 4 3.S*S. 4 AMENDMENT 16 o~s FERRl l'U\.(AA f'\.,\KT Figure 2.3-17 ie.era 11 1111'1l ALL STA81ltlV Cl.~SSES SEPT. 177 ,78 ,TII l

( ( VIND SPEED IMPHl 18 , 5-2 . 12.S-18, 7 .s-,2.1 5.5-7 .1 3,5-S.1 1.5-3,1 1.0-1.1 AMENDMENT 16 ROI/NS FERRY ~EAR PLANT Figure 2.3-18 18,llil 11 WINO ALL STN3JLITY CLASSES OCTOSE:A 177,18,791

( ( ( WIND SPEED <HPH) 18.S-21,'f

                               . 12.s-11.1 7.S-12.

s.s-1. 3.S*S. 1.s-J.1 l.&-1.1 AMENDMENT 16 0""5 FER!t'f MJCLEAR PLANT Figure 2.3-19 II.ell H Ill~ ALL StA!ILITY CLASSES NOl'Ell!f! 177 ,78, 791

( ( C WI ND SPEED (MPH> l8.S-21,1

                              . IJ.S-18,4 7.S-12,4 11 S,S-7.1 I.S*S.-4 AMENDMENT 16 OM FE"RR'f NUCLEI.R PUNT Figure 2.3-20 111,etl 11 IIIHD ALL ST"81LITr ct.~SSES Of:C£"3ER 117,78,79)

C ( C

                 ~IND SPEED [MPH)
                                  > Z , S 11.S-Zf.

7 , 5-IZ,

                                 ,.,-7.4 1.s-s.1 1.s-:s.1 1 .11-1.1 AMENDMENT 16 IICS !'£Riff NJCLEAR Pl Figure 2.3-21 H .N 11 IIJIC>,   -+S,a, I 90,SO N Ttlfl STAIILl1Y Cl.AH A JAN   I , 71

• DC Jl , 79

( ( WIND SPEED IHPH) 11.S--24,1 12.IHl,4 7.S*l2 , 4 r::'\. I.S-7, I.I-I.

 \J                                  1.5-J.1 AMENDMENT 16 CMfS l'VtltY NJCUAA "L Figure 2.3-22 H ... H IIIIO,   ~.le I  IG,SII H TEJP lfAltLITY Cl.ASS I JAH   I, 77

• DEC lit , 79

( ( ( WINO SPEEO IHPHl 12.5-11,

 ~-
 \J S,S-7,1 1,5-S,1 1.0-1,-t AMENDMENT 16 nM'f IU:ltAflt Pl Figure 2.3-23 P,N M WltCI, 41 ...    &  IC.SO K TEW STABILITY CLASS C JAH  1 , 17

• DEC St , 711

( ( ( WlNO SPEED IMPHl ll,S-24,4 12 .s-11. 4 7,S*lt, 4 5 , 5-7.4 I .S*J.1 I.IH.1 AMENDMENT 16

           ,t,m IIJCl.EAII  ~

Fi gure 2.3-24 Ill ** H VJIC), 1' , a a IQ,&;" T£lP ITAIILIT'f c:LASI 0 J'-Jf I , 77 - Ol!C 1' , 79

C ( ( WIND SPEED <MPH> 12.rll,4 I ,S.-:S,4 t.lH.4 AMENDMENT 16 FUlt'f HJCL!AR P\. Figure 2.3-25 ti.* H VII() 1 45.A A 18.H " TE11' IT~ILIT'f CL.All r JAN t, '11

• 0£C SI , 711

( ( (

                ~IND SPEED     CMPHI 11.1-24.1 12.rll,4 ll 5.S-7,1 t .S-J ,1 l.&-1,1 AMENDMENT 16 l'tllff NUCUAR  ~

Figure 2.3-26 ti,. PC VIND, -ta.A & N.IO " TD'P ITAIILlTY CLASI, JAJI I , 11 - OEC JI , 79

( ( (

                   ~IND SPEED         lHPHI 12.i.-11.1 11 I.Ii-Ii.

1 . 5-J. AMENDMENT 16 ft1lll't NJCl.EAR l"I. Figure 2.3-27 01.N N Witt>, +1,11 & 119,lill pt lfPP STAIILlTY C\.ASI ~ JN-1. I , 77

• Ct:C II , 78

( C ( WIND SPEED !MPH> 7,S-IZ, J,S*I,+ I ,S*S,1 l , lt-1,1 AMENDMENT 16 nM'I' MJCLEAA ~ Figure 2, 3-28 tl.N H VIIG AU ITlolllLITY a.ASUS JMI I , 77 - DEC I I

• 70

( ( ( IJlND SPEED lHPHI 11.li-2 ,1

                                . 12,ti-1t, 7.5-12 , 
                                   & .S-7 ,

1,S-!. AMENDMENT 16 nm HUCL£AA PLAlfT Figure 2.3-2 9 11,. "IIIMJ Al.I. ST~II.ITY 0..ASSU JAl<<,IAAY (77.,791

C ( C WIND SPEED IHPHl 12.1.-11.4 7.S-tt,4 S,S-7,4 I ,S-5,4 AMENDMENT 16 DliHS fEMY NUtLEAR PUNT Figure 2.3-30 H.* tt Wl'i>

   -' ALL STASlLITl Cl.ASSU
       ,E8U,t,II., cn.*101

C ( C WINO SPEED (MPH) lt.S-:21.

• 12.s-11.*

7.S-12. c.s-1 ... 1.s-:s.1 1.11-1. AMENDMENT 16

         ,tMY NUQ.QJI PUNT Figure 2.3-31 81.N N IIINl ALL ITAJII\.JTY CUSSES PWICH 111..101

C ( ( WIND SPEED IMPHl 18.5*2-t.1 12.s-11. s.s-1.1 1.IHL1

                                    '.5*:S.

1.e-1

• AMENDMENT 16 DI/HS l'Plfl MJO..EAR Pt.~"1' Figure 2.3-32 H.* P'I VIICI ALL STMIILITY ti.ASSES ArAII. 177-,7a>

( ( ( WIND SPEED <HPHJ

                                 >-Z1,6

11. 5-24,1 12.s-tt.*

7,S*t2,4 S . 5

• 7,1 J.S-i.

I ,S-J.1 1.0-1.1 AMENDMENT 16 ftltftY IUUM PLJ.KT Figure 2.3-33 05 ,11 " IIIICI AL~ 9TAIIILlrY CLASSO 111,Y 177.. 791

( ( C IJINO SPEED IMPHl 18,S*:1!4,t 7.S-12.4 S.5-7,4 s.s-s. 1.11-1.1 AMENDMENT 16 rtftlt1 NUCt.EA~ PLANT Figu-ce 2.3-3~ 11,N" VIN> AI.L STA81t.lTY cv.ssn JlH 171 791

( ( C VIND SPEED CHPHI 12.i*ll,-t 7.5*12,4 J.'ii*S,4 I ,S-3,4 I.II-I ,4 AMENDMENT 16 OWNS n!IRY IU;L~lt PL,-,NT Figure 2.3-35 U.* "VIN> ALL STABILITY Q.AS,e;S JUL)' 177.. 791

( ( ( WIND SPEED IMPHI l8,S-24.1 12.S-18,1 1.s-,.1 I .5*:S. l.&-1.1 AMENDMENT 16 0~9 l"EMY MJCL£Alt PLANT Figure 2.3-36 H,* K 1/ltG ALL STAIIL ITY Cl.ASSO AIJCUST 177,.791

( ( C WINO SPEED IHPHI 18,S-24,4 12.1.-11,4 S.5-7. I .S*J. 1.e--1 .1 AMENDMENT 16 IIMS FtMY NUCl.EA!t PLANT Figure 2.3-37 IIS,N N IIIICI ALL llo\lllLITY Cl.ASSU SEP'T, 177.,791

( ( ( WINO SPEED (MPH) 11.5-21.4 1.s-12.<1 AMENDMENT 16 S nMY NJCLE~R PLOO Figure 2.3-38 Ill,. N Wll'IJ ALL STAIIILlfY cussn OC108Dt <17..791

( ( ( WINO SPEED {HPHI 7,S-12.1 1.S-3,1 1.11-1.1 AMENDMENT 16 11N!1 ,my NUCl.EAR PVJ(T Figure 2.3-39 9S.W H lllhll ALI. !ITA81LITV CLASU9 NOfftt181 177-,70 I .

( ( ( WIND SPEED (HPHl 12.s-11. 7.IH2. I.S-5, 1.5*3. 1.e-1 . AMENDMENT 16 OVliS l'Dlltl IIJCUAR PLIIHT Figure 2.3-40 vs.*" vno All ITAalLITY Cl.~ CitfletR (77..701

BFN-25 2.4 HYDROLOGY, WATER QUALITY, AND AQUATIC BIOLOGY 2.4.1 General The various uses of water in the Browns Ferry area have been investigated. Ground and surface hydrology has been studied to determine the characteristics of both ground and surface water flow in the immediate plant area and the surrounding regional area. Water quality and biological monitoring programs have been developed and implemented to monitor water quality and biological life of Wheeler Reservoir during plant operation. 2.4.2 Hydrology 2.4.2.1 Ground Water Ground water at Browns Ferry is derived from precipitation. Some of the precipitation evaporates, some runs off into streams, and some seeps into the soil. A portion of the water entering the soil is used by vegetation and some of it seeps downward to become ground water. 2.4.2.1.1 Regional Area Studies of subsurface waterflow in the area indicate that ground water flows from the structural highs toward the structural lows. Elevations range from 556 at Wheeler Reservoir to 880 in the east Central part of Limestone County. Rock strata have a regional dip of about 20 ft/mile to the south and southwest, locally altered by minor anticlines and synclines. Both topography and drainage reflect the geologic structure of the area. Rocks exposed in Limestone County are, from oldest to youngest, the Chickamauga limestone, Chattanooga shale, Fort Payne chert, and Tuscumbia limestone. The principal aquifer for Limestone County is the Mississippian Carbonate Regional Aquifer. At this site, the aquifer consists of the Tuscumbia Limestone and the Fort Payne Chert. A mantle of residuum overlies the Fort Payne and Tuscumbia formations. Wells deriving their supply from the residuum are of low capacity. The residuum in the area consists of a mixture of silt, clay, chert, and discontinuous zones of chert gravel. The residuum is capable of storing large amounts of water, which are released at a slow rate to wells, springs, and solution channels in the underlying bedrock. Ground water occurrence is restricted to fractures and solutional cavities in the bedrock. Generally large yields can be anticipated from Tuscumbia and Fort Payne formations. 2.4-1

BFN-28 2.4.2.1.2 Site Area Ground water movement from the regional area into the Browns Ferry site area is controlled by topography and geologic structure. Recharge is also derived from local precipitation that has percolated through the residuum. Natural ground water movement in the area is from the plant site to the Tennessee River. 2.4.2.2 Surface Water Surface water is derived from precipitation remaining after losses due to infiltration and evapotranspiration. It can be generally classified as local surface runoff or streamflow. 2.4.2.2.1 Surface Runoff Surface runoff in the area flows down Poplar Creek, Douglas Branch, and Round Island Creek to the Tennessee River. 2.4.2.2.2 Streamflow Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants, Revision 3, was used as the basis for evaluating the BFN UHS. Per Regulatory Guide 1.27, Sections B, the BFN UHS must be capable of withstanding each of the most severe natural phenomena expected, other site-related events, appropriate combinations of natural phenomena or site-related events, and a single failure of manmade structural features without loss of capability of the UHS to accomplish its safety functions. The most severe phenomena may be considered to occur independently and not simultaneously (e.g., a tornado and an earthquake). In addition, the single failure of manmade structural features need not be considered to occur simultaneously with severe natural phenomena or site-related events unless the severe natural phenomena can cause failure of a manmade structural feature. Per Regulatory Guide 1.27, Sections C.2.a, the BFN UHS must be capable of withstanding, without loss of the UHS safety functions, all of the following events: (1) The most severe natural phenomena expected at the site in accordance with General Design Criteria (GDC) 2, (2) The site-related events (e.g., transportation accident, river diversion) that historically have occurred or that may occur during the plant lifetime, (3) Appropriate combinations of less severe natural phenomena and/or site-related events, 2.4-2

BFN-28 (4) Failure of reservoirs, dams, and other manmade water retaining structures both upstream and downstream of the site including the potential for resultant debris to block water flow; and (5) Potential changes in ocean, river, or lake levels as a result of severe natural events, or possible changes in climatological conditions in the site region resulting from human or natural causes, that may occur during the plant lifetime.

a. Description of UHS The Browns Ferry UHS is the Wheeler Reservoir, which was formed by the damming of the Tennessee River by the Wheeler Dam, located downstream of Browns Ferry, and the Guntersville Dam located upstream of Browns Ferry. This water area has been monitored since 1939. The Wheeler Reservoir provides water flow to the RHRSW system through the plant intake structure.

b. Inflow Since 1939, streamflow records have been maintained at the Guntersville Dam on the Tennessee River. The average daily discharge flow for the period 1939 to 2015 has been 33,500 cfs. The maximum streamflow occurred on March 19, 1973, and was 304,400 cfs. The minimum daily average streamflow, 100 cfs, occurred several times during the period of record, due to regulation of Guntersville Dam. Generally, minimum daily flows are much higher.

Flow duration data for the period 1939 to 2015 were provided by the TVA River Management. This period of record, 76 years, represents a significant period of time (i.e. - much greater than 30 years) and results in highly reliable data. A review of the flow duration data observed during the time period between 1939 and 2015 shows that streamflow equal or exceed the following values for the indicated percentages of the time: 2.4-3

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-28 Percent of Time Streamflow, cfs Equaled of Exceeded 230,298 0.1 129,500 2.3 56,425 15.9 33,500 50.0 17,700 84.1 7,100 97.7 1,319 99.9 It should also be noted that this data only accounts for flow through the Guntersville Dam and does not include additional flow from the various intermediate tributaries l o c a t e d between the Guntersville Dam and the Browns Ferry site. The intermediate tributary flows would add to the total flow available at the site. Channel velocities at the Whitesburg gage, located upstream of Browns Ferry, were measured and average more than 2 ft/sec Under normal winter conditions and a little more than 1 ft/sec under normal summer conditions.** These average winter and summer velocities drop to about 0.7 ft/sec and 0.3 ft/sec, respectively, at Browns Ferry where the reservoir is wider and the slope of the water surface is less. A flood equal to the maximum of record would produce average velocities up to 4 ft/sec in the channel and up to 2 ft/sec in the overbank area. Average velocities produced by a maximum probable flood, regulated, would be about the same magnitude. A location plan and cross-sections of the reservoir at the silt ranges (SR), and adjacent to the plant site are presented in Figures 2.4-1a through 2.4-1e. A longitudinal channel profile throughout Wheeler Reservoir is presented in Figure 2.4-2. The figure also identifies the sources of dependable watershed drainage with direct inflows into Wheeler Reservoir.

• • Velocity data obtained from "Tennessee River Computed Navigation Channel Velocities,"

TVA, July 1963. 2.4-4

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-28 The silt range profiles were taken from a survey made in August 1961; these profiles have been confirmed by 15 additional detailed silt range surveys made in August 1969 and a survey of SR24 and SR25 made in September 1989. The results of six silt range surveys made over a 33-year interval (October 1936, May 1947, May 1953, June 1956, August 1961, and August 1969) confirm that the silting rate is insufficient to require extensive surveys at frequent intervals. This is consistent with findings from TVA's system-wide silt survey studies. Decisions about future updates for the cross sections around Browns Ferry and the frequency of the update are based upon this review. Since silting is the only mechanism for significantly reducing the volume of water available in the pool, this surveillance program will ensure against undetected decreases in pool volume. Silt transport during drawdown following should not be a serious problem because of the large number of deep water pools upstream of the plant site. Similarly, headwater elevation at the Wheeler Dam for the period 1939 to 2015 was provided by TVA River Management. Again, this period of record represents a significant timeframe and thus is considered to be very reliable. A review of the headwater elevation data observed during the time period between 1939 and 2015 shows that the headwater elevation, to the nearest tenth of a foot, equals or exceeds the following values for the indicated percentages of the time: Percent of Time Headwater Elevation, ft Equaled or Exceeded 556.3 0.1 556.0 3.5 (Maximum Normal Level) 555.7 15.9 553.5 50.0 551.1 84.1 550.0 99.2 (Minimum Normal Level) 549.6 99.9 2.4-5

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-28 These minimum flows are combined with the leakage flow from Guntersville Dam to give a minimum flow of 100 cfs in the original river channel. This total flow, in traversing the pool created at the Browns Ferry site by the failure of Wheeler Dam, would have an average residence time of about 11-1/2 days. The Browns Ferry plant intake and discharge arrangements are shown in Figures 12.2-69 through 12.2-75b, sheets 1, 2, and 3. The intake channel to the pumping station is excavated to El. 523 and extends into the reservoir until it connects with the original channel where the aforementioned pool would be trapped. The pumping station floor elevation is at 518 which gives a minimum of 11 ft of water inside the structure. (Eleven feet of water provides adequate submergence for the RHR service water pumps to deliver the shutdown cooling water requirements of 36,000 gpm (80 cfs) to the plant. This is sufficient flow to remove the decay heat from all three reactors plus the heat rejection from eight diesel generator sets operating at full load.) All of the cooling water is discharged from the plant through either the RHRSW diffuser nozzles upstream of the CCW diffusers or the storm sewer into the intake forebay.

c. Safety Evaluation Regulatory Guide 1.27, Section C.2.a(1) - The most severe natural phenomena expected at the site in accordance with GDC-2 The intake structure has been designed for both flooding and seismic events.

As a result, the most severe natural phenomena associated with the Wheeler Reservoir intake is a low level in the reservoir and a low inflow both associated with drought or hot weather conditions. Using the historical lowest flow from Guntersville Dam in combination with the historical lowest Wheeler Reservoir elevation ensures the most severe natural phenomena has been considered. As indicated previously, the minimum inflow from Guntersville Dam was 1,319 cfs and the minimum reservoir level of the Wheeler Reservoir headwater elevation was 549.6. Both of these conditions individually occurred less than 0.13 percent of the time during the period of record. Since the 1,319 cfs influx alone is sufficient to provide a reliable heat sink (i.e. > 80 cfs), and the elevation of the reservoir is much greater than that required for adequate supply and submergence, the reservoir is acceptable for the most severe natural phenomena condition. 2.4-6

BFN-28 Regulatory Guide 1.27, Section C.2.a(2) - The site-related events (e.g., transportation accident, river diversion) that historically have occurred or that may occur during the plant lifetime As specified earlier, the most severe regulation of outflow from Guntersville Dam during the period of record limited the flow to 100 cfs. The actual flow past the Browns Ferry site will be larger than this as a result of inflow from numerous incoming branches between Guntersville Dam and the site. However, the 100 cfs flow rate alone is sufficient to meet the 80 cfs accident and shutdown requirements of the plant. During this time, the water level in the Wheeler Reservoir would be expected to be at its average normal level of 553.5, but would be no lower than its minimum normal level of 550.0, both of which exceed the minimum required level for adequate flow and NPSH to the RHRSW pumps and provide a significant volume of water available to the plant. Additionally, a 30 day low Guntersville Dam discharge flow was determined during a period of several low flow days. The average discharge flow was still approximately 6,000 cfs, which more than supports the flow requirements of RHRSW. Obstruction of the intake pumping station channel resulting from a river transportation accident is discussed in Section 12.2.7.6. At the normal minimum pool level of 550.0, there is sufficient pool depth to ensure that adequate flow and NPSH to the RHRSW pumps is available to the plant. Regulatory Guide 1.27, Section C.2.a(3) - Appropriate combinations of less severe natural phenomena and/or site-related events, From an historical perspective, the inflow from the upstream Guntersville Dam and the Wheeler Reservoir levels combine in a myriad of ways throughout the year based on climatic conditions in the region. However, there are no reasonable combinations of reservoir level and inflow that would invalidate the ability of the reservoir to perform its safety function. The intake structure has been designed for both flooding and seismic events and is therefore capable of performing its safety function for all natural phenomena. The reasonable combination of the minimum normal Guntersville Dam flow of 1,319 cfs and the minimum normal Wheeler Reservoir level of 550.0 feet ensures the ability of the BFN UHS to perform its safety function. Regulatory Guide 1.27, Section C.2.a(4) - Failure of reservoirs, dams, and other manmade water retaining structures both upstream and downstream of the site including the potential for resultant debris to block water flow. 2.4-7

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-28 Therefore, the BFN UHS fulfills its safety requirement after a single failure of a man made structure. Regulatory Guide 1.27, Section C.2.a(5) - Potential changes in ocean, river, or lake levels as a result of severe natural events, or possible changes in climatological conditions in the site region resulting from human or natural causes, that may occur during the plant lifetime. As specified earlier, the BFN UHS has been monitored since 1939 or more than 75 years. During this period the most severe regulation of outflow from Guntersville Dam limited the flow to 100 cfs. The actual flow past the Browns Ferry site will be larger than this because of inflow from numerous incoming branches between Guntersville Dam and the site. However, the 100 cfs flow rate alone is sufficient to meet the 80 cfs accident and shutdown requirements of the plant. During this time, the water level in the Wheeler Reservoir would be expected to be at its average normal level of 553.5, but would be no lower than its minimum normal level of 550.0, both of which exceed the minimum required level for adequate flow and NPSH to the RHRSW pumps and provide a significant volume of water available to the plant. Additionally, a 30 day low Guntersville Dam discharge flow was determined during a period of several low flow days. The average discharge flow was still approximately 6,000 cfs, which more than supports the flow requirements of RHRSW. The Guntersville Dam flow rate and Wheeler Reservoir levels during the monitored period demonstrate that the BFN UHS can perform its safety function following potential changes in river flow or lake levels as a result of severe natural events, or possible changes in climatological conditions. 2.4-8

BFN-28 2.4.2.2.3 Floods The Browns Ferry site is located on the right bank of Wheeler Reservoir at approximately Tennessee River mile (TRM) 294. The lowest natural ground elevation in the site vicinity is about 560 feet above mean sea level and the average ground elevation is about 580. The probable maximum flood (PMF) at Browns Ferry is calculated to reach El. 571.7. However, the site PMF level is being maintained at elevation 572.5. This is the flood which defines the upper limit of potential flooding at the plant. A concise definition of PMF is given in Section 1.2, while the determination of PMF is described in Appendix 2.4A. 2.4.3 Water Quality Information reflecting the water quality, water temperature, and aquatic biota conditions in the vicinity of the Browns Ferry Nuclear Plant (BFN) were incorporated into the Final Environmental Statement, Browns Ferry Nuclear Plant, Units 1, 2, and 3, Volumes 1, 2, and 3, TVA Office of Health and Safety, Chattanooga, Tennessee, September 1, 1972. Results of the preoperational water monitoring program for the period 1968 through 1973 are included in the report "Water Quality and Biological Conditions in Wheeler Reservoir Before Operation of Browns Ferry Nuclear Plant - 1968-1973." Results of the operational monitoring program for the period 1974 through 1980 were included in a series of five semiannual reports followed by five annual reports. The last report including the 1980 monitoring results was "Water Quality and Biological Conditions in Wheeler Reservoir During Operation of Browns Ferry Nuclear Plant January 1, 1980-December 31, 1980," Volumes I and II. These monitoring and reporting requirements under the jurisdiction of NRC were determined to be duplicative of the requirements imposed by the Browns Ferry NPDES permit (AL 0022080) issued on June 30, 1977. In response to TVA's letter dated July 27, 1981, NRC notified TVA of their concurrence with this determination by letter dated December 10, 1981, and accepted TVA's recommendation to delete these monitoring requirements from the Browns Ferry Environmental Technical Specifications. All subsequent water quality, biological, and thermal monitoring and reporting have been and will continue to be in accordance with the requirements of the Browns Ferry NPDES permit (Permit No. AL 0022080) and/or TVA policy. The most recent comprehensive evaluation of the aquatic conditions of Wheeler Reservoir is contained in the report "A Supplemental 316 (a) Demonstration For Alternative Thermal Discharge Limits For Browns Ferry Nuclear Plant, Wheeler Reservoir, Alabama," TVA, February 1983. 2.4-9

BFN-25 2.4.4 Water Use The public and industrial water supplies which withdraw surface water from the Tennessee River in the 61-river mile reach from Decatur, Alabama to Colbert Steam Plant, not including Browns Ferry Nuclear Plant, are listed in Table 2.4-4. 2.4.4.1 Industrial Major industrial water users are located both upstream and downstream of the Browns Ferry project. These users withdraw water from Wheeler Reservoir each day for process and cooling needs. Most of this water is subsequently returned to the reservoir. 2.4.4.2 Public The major public uses of the reservoir are for water supplies, recreation, and waste disposal. Six public water supplies are taken directly from the Tennessee River portion of Wheeler, Wilson, and Pickwick Reservoirs within the reach from Decatur, Alabama, about 12 river miles upstream from the plant, to Colbert Steam Plant, about 49 river miles downstream from the plant. Eleven industrial supplies also withdraw water from the reservoirs in this same reach, and some use a portion of their withdrawal for potable water within the plant. 2.4.4.3 Browns Ferry Nuclear Plant The Browns Ferry Nuclear Plant will use a large volume of water. When all three units are in operation, river water will be pumped through the plant at the rate of about 4,400 cfs. The temperature of this water will be elevated above its natural temperature. Heated condenser cooling water will be diffused into the main channel flow of the Tennessee River by a Diffuser System consisting of three perforated pipes laid side by side on the bottom of the channel near TRM 294. The Diffuser System is detailed in paragraph 12.2.7.5. The containment, treatment, storage (including quantities), and pathways for release of liquid radiological effluents at BFN are detailed in Section 9.2 Liquid Radwaste System. The nearest community surface water supply is at Decatur, Alabama, on Wheeler Reservoir 12 miles upstream from the Browns Ferry site. With normal operation of Guntersville and Wheeler Dams, there would be no flow upstream from Browns Ferry that would reach Decatur. Should a slug release (i.e., a finite volume of contaminant released nearly instantaneously into a receiving waterway) occur at a time when upstream flow to Decatur could conceivably occur, the river control system could be operated to prevent the upstream flow. 2.4-10

BFN-25 The first downstream water intake is the West Morgan-East Lawrence Water Authority intake located at TRM 286.5 on the left bank of Wheeler Reservoir. An analysis was made to determine the minimum dilution to be expected between the diffusers and the intake at West Morgan-East Lawrence for both accidental slug and continuous plane source releases. The following assumptions were used in the analysis.

1. Because the water intake is located on the bank opposite the plant, minimum dilution would occur when the release is fully mixed over the cross section of the reservoir. This is accomplished by configuring the release as a plane source placed vertically across the width of the channel.

2. Mixing calculations are based on steady flow conditions in the reservoir. River flow is assumed to be 33000 ft3/sec. This is the flow which is equaled or exceeded in the reservoir approximately 50 percent of the time.

3. The concentration profile from an instantaneous (i.e. slug) release of contaminant is assumed to be Gaussian in the longitudinal direction.

4. The calculated contaminant concentration is conservative. Material discharged into the river does not degrade through radioactive decay, chemical or biological processes, nor is contaminant removed from the reservoir by adsorption to sediments or by evaporation.

All results are given in units of relative concentration, expressed as C/C0 where C represents the concentration of contaminant at the point of interest, and C0 is the concentration of contaminant at the point where it enters the reservoir. Dilution is the reciprocal of relative concentration. The maximum relative concentration at the West Morgan-East Lawrence Water Authority intake due to a continuous plane source release rate Q (ft3/sec) of contaminant is 3.0 Q x 10-5. The maximum relative concentration at this location due to an instantaneous plant source release of a volume V (ft3) of contaminant is 3.2 V x 10-10. For the instantaneous relative concentration, the following parameter values were used: channel width = 6000 ft, channel depth = 35 ft, longitudinal dispersion parameter = 200, mixing coefficient (manning's n) = 0.03. At the time of initial plant licensing, there were no private ground water wells located within one mile of the reactor building, and there were only eight houses located within one mile of the site perimeter which relied on groundwater as a source of 2.4-11

BFN-25 water supply. Because all local groundwater in the plant site area flows directly to Wheeler Reservoir (see Section 2.4.2.1), it is improbable that any liquid released from the site could contaminate these sources of water supply through contamination of groundwater. Furthermore, with the containment provided for the liquid radwaste system (see Section 9.2), there is little likelihood of the release of liquid radwaste to the groundwater. In the event of any unusual release of radwaste liquid which could contaminate groundwater at the site, special local monitoring will be carried out in accordance with the Radiological Monitoring Plan, Browns Ferry Nuclear Plant, to ensure that the use of these wells will not result in undue hazards to any person, even though there is little likelihood of the wells becoming contaminated. With the very unlikely event that the private wells located within one mile of the site perimeter could become contaminated, the public and industrial groundwater supplies in the site vicinity (all of which are located well beyond one mile from the site) would not be expected to be affected by plant operation. Consequently, the contamination of public and industrial groundwater supplies is not a concern at Browns Ferry requiring the monitoring and/or inventorying of such supplies. However, a periodic inventory of the private wells located within one mile of the site reactor building will be conducted. Table 2.4-6 contains a list of the private wells as inventoried in 1989. Figure 2.4-3 shows the location of the private wells within one mile and two miles of the plant. 2.4.5 Aquatic Biota The historic aquatic biological conditions and their associated routine monitoring and reporting are identified in Section 2.4.3 Water Quality. 2.4.6 Monitoring Programs All Browns Ferry related radiological water quality and aquatic biological monitoring programs are being conducted and reported in accordance with the Browns Ferry Nuclear Plant Radiological Environmental Monitoring Program as described in the Browns Ferry Offsite Dose Calculation Manual. Since 1981, all nonradiological water quality, aquatic biological, and water temperature monitoring programs have been and will continue to be conducted and reported in accordance with the Browns Ferry NPDES permit (Permit No. AL 0022080) and/or TVA policy. (See Section 2.4.3 Water Quality for a discussion of these monitoring programs prior to 1981.) 2.4-12

BFN-25 2.4.7 Conclusions Ground water movement in the area is from the plant site to the Tennessee River. The principal aquifer in the area is overlain by a mantle of residuum that retards the movement of shallow ground water. Migration of radionuclides in the residuum would be quite slow. It is highly unlikely that the private groundwater wells located within one mile of the site perimeter could be contaminated by operation of BFN. Special local groundwater monitoring of these wells would be implemented in the event of a liquid radioactive release to the groundwater at BFN. Consequently, the potential for contamination of the public and industrial groundwater systems in the BFN area is not a concern which requires monitoring or inventorying of these systems. A periodic inventory of private wells within one mile of the site area will be implemented. Surface water runoff from the plant site is to the Tennessee River. Surface water runoff from the plant site is to the Tennessee River. Regulated by the TVA flood control system, the probable maximum flood would result in increasing Wheeler Reservoir level to 572.5 feet above sea level at the site. Safety-related structures are protected against all flood conditions up to El. 578 as discussed in response to Question 2.6 and would not be endangered by the probable maximum flood. All nonradiological water quality, biological, and thermal monitoring and reporting related to BFN has been and will continue to be conducted in accordance with the requirements of the NPDES permit and/or TVA policy. 2.4-13

BFN-19 Table 2.4-1 (Deleted by Amendment 6)

BFN-19 Table 2.4-2 (Deleted by Amendment 6)

BFN-19 Table 2.4-3 (Deleted by Amendment 6)

BFN-19 TABLE 2.4-4 PUBLIC AND INDUSTRIAL SURFACE WATER SUPPLIES WITHDRAWN FROM THE 61 MILE REACH OF THE TENNESSEE RIVER BETWEEN DECATUR ALABAMA AND TVA COLBERT STEAM PLANT Approximate Average Distance Daily From Site Plant Name Use (MGD) Location (River Miles) Type Supply Decatur, (Ala.) 26.92 TRM 306 12 (Upstream) Municipal Solutia Company (Ala.) 120.30 TRM 302 8.0 (Upstream) Industrial 3M Company (Ala.) 12.96 TRM 299.7 5.7 (Upstream) Industrial Amoco Chemicals Corp. (Ala.) 6.63 TRM 299.5 5.5 (Upstream) Industrial Browns Ferry Nuclear Plant (Ala.) TRM 294R 0.0 (Diffuser Location) West Morgan-East Lawrence Water Authority (Ala.) 4 TRM 286.5L 7.5 (Downstream) Municipal Champion International (Ala.) 55.90 TRM 282.6L 11.4 (Downstream) Industrial TVA-Wheeler Dam (Ala)* # TRM 274.9 19.1 (Downstream) Industrial Reynolds Metals Company (Ala.) 3.11 TRM 261L 33.0 (Downstream) Industrial Florence (Ala.) ** 6.70 TRM 259.8R 34.2 (Downstream) Municipal and Cypress Cr 8.4 Muscle Shoals (Ala.) 3.61 TRM 259.6L 34.4 (Downstream) Municipal Fleet Hollow Embayment Mi. 0.4 TVA-Environmental Research Center ERC (Ala.) 4.50 TRM 259.5L 34.5 (Downstream) Industrial Fleet Hollow & Potable Embayment Mi. 0.5 TVA Wilson Dam (Ala.) *** # TRM 259.4 34.6 (Downstream) Industrial Occidental Chemical Co. (Ala.) 14.40 TRM 258.4L 35.6 (Downstream) Industrial Sheffield (Ala.) 2.10 TRM 254.3L 39.7 (Downstream) Municipal TVA Colbert Steam Plant (Ala.) 1250.00 TRM 245 49.0 (Downstream) Industrial Cherokee (Ala.) 12.80 TRM 239.3L 54.7 (Downstream) Municipal Laroche Industries 39.60 TRM 238.7L 55.3 (Downstream) Industrial

• Water used for industrial purposes only. Potable water is purchased from East Lauderdale County Water District.

** Florence has two water treatment plants. The Cypress Creek Plant is not included in this number.

• • • Water used for industrial purposes only. Potable water is from TVA-Property Services and Property Operation (PS&PO).

#  Water usage is not metered.

BFN-19 Table 2.4-5 (Deleted by Amendment 6)

BFN-19 Table 2.4-6 (Sheet 1) Private Water Wells Within Two Miles of BFNP Stack, June 1995 Well Well No. Owner Depth (ft) Use 1A Jesse Crouch (Renter) 38 Not Used 1B Jesse Crouch (Renter) 25.5 Not Used 1C Jesse Crouch (Renter) 32.5 Not Used 2 Unknown -- Unknown 3 Ronnie Crouch 127 Not Used 4 Leonard Hudson 167 Household 5 Thurman C. Burns 185 Not Used 6 Dr. Wm. A. Sims -- Not Used 7 James G. Ratliff -- Unknown 8 J. M. Hall 78 Not Used 9 Leander Faulk 66 Not Used 10 Leland -- Not Used 11 Ted Russ 61 Not Used 12 Nellie Haggermaker -- Not Used 13 Thurman C. Burns 65 Not Used 14 T. C. Bozeman 48 Not Used 15 Vivian Mock 64 Not Used 16 Leech -- Not Used 17 Wm. N. Clingan -- Not Used 18 Name Not Available 2.5 Not Used 19 Lee Townsend 60 Household 20 Wayne Black 65 Not Used 21 John W. Roberts 28 Not Used 22 A. R. Barron 62 Not Used 23 W. D. Boggs -- Unknown 24 Sam Lanza Unknown Renter - James Castle -- 25 Clifford Taylor -- Unknown 26 Unknown -- Unknown 27 Wm. L. Priest 55 Not Used 28 George B. Walker 35 Household 29 Ret. Col. S. H. Wade 60 Not Used 30 Charles McCormack 64 Not Used 31 Unknown -- Unknown 32 W. M. Howard -- Unknown 33 Glenn Ross 30 Not Used 34 Robert R. Leonard -- Not Used

BFN-19 Table 2.4-6 (Cont'd) (Sheet 2) Private Water Wells Within Two Miles of BFNP Stack, June 1995 Well Well No. Owner Depth (ft) Use 35 Mike Morrow 48 Not Used 36 Unknown -- Unknown 37 Joe Cannon 48 Not Used 38 O'Neal Terry 76 Not Used 39 J. D. Walker, Jr. 75 Not Used 40 Loyd Barham 100 Not Used 41 Charles R. Aday 200 Not Used 42 Kenneth King 74 Not Used 43 Mike Basden -- Not Used 44 Russell Mattox 50 Not Used 45 John T. Nelms 45 Not Used 46 Mrs. Lyle (Faye) Cheatham -- Household 47 Neal Holland, Jr. 80 Not Used 48 Percy Terry 40 Not Used 49 Wayne Coggin 90 Not Used 49A Wayne Coggin 50 Irrigation 50 Phillip Geddes 92 Not Used 51 Mrs. M. J. Watkins 100 Not Used 52 Claude Morris 60 Not Used 53 Norman Puckett 52 Not Used 54 Wayne White 62 Household 55 Troy Williams 36 Not Used 56 Dr. Chas. H. Burt 90 Not Used 57 Jim Jordan 60 Not Used 58 Dr. Wayne Brannon 45 Not Used 59 Bobby D. Hamilton 40 Not Used 60 Dr. Sidney Chennault 90 Not Used 61 Dr. Lloyd Nix 38 Not Used 62 Wayne Black 160 Household

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 -ROAO c:::J WATER
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  -    RfACTOR BUILO!NO

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r l FIGURE 2,4-3 BROWNS FERRY NUCLEAR PLANT - PRIVATELY OWNED GRCtlNO'KATER WELLS - 1 ~NO 2 MILE RADIUS FROM THE STACK

C I*. ,; Plant Site Ground Water Supply NOTE: The number associated with the symbol corresponds to the numbering in Table 2.4-!:> Scale of Miles 10 _ .. _o ________ 10___ _ 20 L.H ..1-:f..E .B .r.L _ - - - - .. L .... . .. ___ --~ AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Ground Water Supplies Within 20 mile Radius of Plant Site FIGURE 2.4-4

BFN-25 APPENDIX 2.4A BROWNS FERRY NUCLEAR PLANT PROBABLE MAXIMUM FLOOD (PMF) 2.4A-i

BFN-25 CONTENTS Page Introduction .................................................................................................................................. 2.4A-1 Definition .............................................................................................................................. 2.4A-1 Data Available ...................................................................................................................... 2.4A-1 Summary of Results ..................................................................................................................... 2.4A-2 The Watershed ............................................................................................................................. 2.4A-3 Physiography ....................................................................................................................... 2.4A-3 Climate................................................................................................................................. 2.4A-3 Reservoir System ......................................................................................................................... 2.4A-4 General Description ............................................................................................................. 2.4A-4 Flood Detention Capacity..................................................................................................... 2.4A-4 Hydrograph Determination ........................................................................................................... 2.4A-5 General Description ............................................................................................................. 2.4A-5 Unit Hydrographs ................................................................................................................. 2.4A-5 Reservoir Routing ................................................................................................................ 2.4A-5 Verification ......................................................................................................................... 2.4A-6 Breached Earth Embankments ............................................................................................ 2.4A-6 Probable Maximum Precipitation .................................................................................................. 2.4A-8 Rain-Runoff Relationships ............................................................................................................ 2.4A-9 Conditions Creating Probable Maximum Flood ............................................................................ 2.4A-10 Critical Storm ....................................................................................................................... 2.4A-10 Precipitation Excess............................................................................................................. 2.4A-10 Reservoir Operations ........................................................................................................... 2.4A-10 Dam Failures ....................................................................................................................... 2.4A-10 2.4A-ii

BFN-25 Probable Maximum Flood............................................................................................................. 2.4A-11 Wind Waves......................................................................................................................... 2.4A-11 Maximum Elevation.............................................................................................................. 2.4A-12 Local Drainage ............................................................................................................................. 2.4A-12 Local Stream ........................................................................................................................ 2.4A-12 Main Plant Area ................................................................................................................... 2.4A-14 Cooling Tower System ......................................................................................................... 2.4A-15 2.4A-iii

BFN-25 TABLES

1. Facts about Major TVA Dams and Reservoirs

2. Unit Hydrograph Data

3. Maximum Possible Storm Rainfall and Precipitation Excess

4. Dam Failure Statistics (Deleted)

FIGURES

1. Tennessee Valley Region

2. Seasonal Operating Curve, Cherokee

3. Seasonal Operating Curve, Guntersville

4. Browns Ferry Nuclear Plant Hydrologic Model, Unit Areas

5. Unit Hydrographs (14 Sheets)

6. Hydrologic Model Verification - 1973 Flood, Elevation

7. Hydrologic Model Verification - 1973 Flood, Flow

8. Hydrologic Model Verification - 2004 Flood, Elevation

9. Hydrologic Model Verification - 2004 Flood, Flow

10. Hydrologic Model Verification - Steady-State Profiles

11. Hydrologic Model Verification - Guntersville Tailwater Curve

12. Guntersville Dam, General Plan Elevation and Sections

13. Probable Maximum Precipitation Isohyets (21,400-Square-Mile Downstream)

14. Probable Maximum Precipitation Isohyets (16,170 Square Miles)

15. Browns Ferry Nuclear Plant PMF Discharge

16. Browns Ferry Nuclear Plant PMF Elevations

21. Browns Ferry Nuclear Plant, Watershed, Unnamed Tributary Northwest of Plant

22. Browns Ferry Nuclear Plant, General Plan 22a. Browns Ferry Nuclear Plant, Plant Topography 22b. Probable Maximum Precipitation, Point Rainfall

23. One-Hour Unit Hydrograph for Unnamed Stream Northwest of Plant

24. Maximum Possible Flood, Unnamed Stream Northwest of Plant

25. Channel Relocation West of Cooling Towers, Typical Sections

26. Browns Ferry Nuclear Plant, Channel Relocation West of Cooling Towers, Sheets 1 and 2 2.4A-iv

BFN-25 APPENDIX 2.4A BROWNS FERRY NUCLEAR PLANT PROBABLE MAXIMUM FLOOD (PMF) INTRODUCTION This appendix describes the determination of maximum flood conditions at Browns Ferry Nuclear Plant. The plant could possibly be flooded by the Tennessee River, by a small stream northwest of the plant, and by intense local storms which overtax the site drainage system. Each situation has been examined and is discussed separately in this appendix. The main body of the appendix discusses the determination of maximum flood levels resulting from conditions on the Tennessee River. The Browns Ferry plant is on the right bank of Wheeler Reservoir at Tennessee River mile (TRM) 294, 55 miles downstream from Guntersville Dam and 19 miles upstream from Wheeler Dam. The 27,130-square-mile watershed drains the rugged southern Appalachian Mountains, the second highest annual rainfall-producing area in the conterminous United States, portions of the Cumberland Plateau and Highland Rim, and extends into parts of six states. Definition The term probable maximum flood (PMF), is used by TVA to describe the hypothetical flood which would result from an occurrence of the probable maximum precipitation (PMP) critically centered on the watershed as defined by the National Weather Service. The computational procedures used to translate this PMP rainfall into flood flows result in a flood hydrograph which defines the upper limit of potential flooding at the plant. Such a flood was earlier called the "maximum possible flood" by the Browns Ferry Nuclear Plant. The flood was determined by deterministic procedures, however most hydrologists agree that the probability of occurrence of the flood in a particular year closely approaches zero. Data Available The Tennessee River Basin above Wheeler Dam is a gaged watershed. TVA began its program of hydrologic data collection upon its creation in 1933. The program has continually evolved since that date. There are currently 142 rain gages and 42 recording stream gages which measure rainfall and stream-flows in the basin above Wheeler Dam. The period of above-average hydrologic gaging extends back to 1935 when TVA began its expanded program of hydrologic data collection. 2.4A-1

BFN-25 The nearest location with extensive, formal flood records is 37 miles downstream at Florence, Alabama, where continuous records are available since 1871. Knowledge about significant floods extends back to 1867 based upon newspaper and historical reports. TVA has developed procedures for predicting stream flows for use in the daily operation of the reservoir system. The forecast procedure begins with rainfall and proceeds through all the hydrologic calculations necessary to translate this rainfall into reservoir inflows. The present forecast procedure for Wheeler Reservoir has been in continuous use with modifications dictated by experience since 1952. An important element of these flood forecasts was the analysis of flood events using rainfall developed by the Hydrometeorological Branch, Office of Hydrology, National Weather Service (NWS). The results of NWS rainfall studies are contained in Hydrometeorological Report No. 41, "Probable Maximum and TVA Precipitation Over the Tennessee River Basin Above Chattanooga," published in June 1965, and HMR 47,Meteorlogical Criteria for Extreme Floods for Four Basins in the Tennessee and Cumberland River Watersheds, issued May 1973.

SUMMARY

OF RESULTS Maximum possible discharge from the Tennessee River at Browns Ferry Nuclear Plant is 1,200,000 cubic feet per second (cfs) with corresponding maximum elevation at the plant of 572.5. The PMF elevation includes allowances for failure of portions of an earth saddle dam at Watts Bar dam and earth embankment sections at Chickamauga, Nickajack, and Guntersville Dams located upstream of the plant. The maximum possible discharge from the small stream northwest of the plant is 17,200 cfs. The channel system has been designed to prevent flooding of the plant in the event such a flood should occur. At the plant the maximum concentration of drainage is northwest of the main building complex. Flow restrictions will cause water surface elevations to reach El. 566.6 between the office and service buildings. THE WATERSHED Physiography The Tennessee River at Browns Ferry Nuclear Plant, mile 294, drains a 27,130-square-mile watershed area above the plant. Guntersville Dam, 55 miles upstream, has a drainage area of 24,450 square miles. Wheeler Dam, the next dam downstream, the headwater of which affects flood elevations at the plant, has a drainage area of 29,590 square miles. 2.4A-2

BFN-25 The major tributaries upstream of Chattanooga, drainage area 21,400 square miles, except for the Clinch and Holston Rivers rise to the east in the rugged southern Appalachian Highlands. They flow northwestward through the Appalachian Divide, which is essentially defined by the North Carolina-Tennessee border, to join the Tennessee River which flows southwestward. The drainage pattern is shown on Figure 1. The Tennessee, Clinch, and Holston Rivers flow southwest through the valley and ridge physiographic province which, while not as rugged as the Southern Highlands, features a number of mountains including the Clinch and Powell Mountain chains. Downstream of Chattanooga the only major tributary is the Elk River which joins the Tennessee River above Wheeler Dam and downstream from the plant. About 10 percent of the watershed rises above El. 3000 feet with a maximum elevation of 6684 feet at Mount Mitchell, North Carolina. The watershed is about 60 percent forested with much of the more mountainous area being 100 percent forested. Climate The climate of the watershed is classified as humid temperate. Above Wheeler Dam annual rainfall averages 51 inches and varies from a low of 40 inches at sheltered locations within the mountains to high spots of 90 inches on the southern and eastern divide. Rainfall occurs relatively evenly throughout the year. The lowest monthly average is 3.0 inches in October. The highest monthly average is 5.1 inches in March. Major flood-producing storms are of two general types: the cool season, winter type and the warm season, hurricane type. Most floods at Browns Ferry Nuclear Plant, however, have been produced by winter-type storms in the flood-season months of January through April. Watershed snowfall is relatively light, averaging only about 14 inches annually. Individual snowfalls are normally light with an average of 13 snowfalls per year. Snowfall is not a factor in maximum flood determinations. RESERVOIR SYSTEM General Description The Tennessee River, particularly above Chattanooga, Tennessee, is one of the most regulated rivers in the United States. A prime purpose of the TVA water control system is flood control with particular emphasis on protection for Chattanooga, 170 miles upstream from Browns Ferry Nuclear Plant. 2.4A-3

BFN-25 There are 22 major reservoirs in the TVA system upstream from Browns Ferry Nuclear Plant, 11 of which have substantial reserved flood detention capacity on March 15. Table 1 lists pertinent data for TVA's major dams. In addition, there are six major dams owned by the Aluminum Company of America (ALCOA). The ALCOA reservoirs often contribute to flood reduction, but they do not have dependable reserved flood detention capacity. The locations of these dams are shown on Figure 1. Flood Detention Capacity Flood control above the plant is provided largely by tributary reservoirs. On March 15, near the end of the flood season, these provide a minimum of 4,484,000 acre-feet equivalent to approximately 5.5 inches on the 15,237 square-mile area they control. This is approximately 82 percent of the total available above the plant. The four main river reservoirs-Fort Loudoun, Watts Bar, Chickamauga, and Guntersville-provide 997,400 acre-feet equivalent to 1.6 inches on the 11,893-square-mile area above the plant and lying below the major tributary dams. The flood detention capacity reserved in the TVA system varies seasonally, with the greatest amounts during the January through March flood season. Figures 2 and 3 show typical tributary and main river reservoir seasonal operating guides. Total assured system detention capacity above Wheeler dam varies from 3.8 inches on January 1 to 3.7 inches on March 15 and decreases to 1.1 inch during the summer and fall. Actual detention capacity may exceed these amounts, depending upon inflows and power demands. Wheeler Dam, the headwater elevation of which affects flood elevations at the plant, has a drainage area of 29,590 square miles, 5,140 square miles more than Guntersville Dam. There is one major tributary dam, Tims Ford, in the 5,140-square-mile intervening watershed. On March 15, near the end of the flood season, this project provides a minimum of 167,000 acre-feet equivalent to approximately 5.9 inches on the 529-square-mile controlled area. Wheeler Dam contains 326,500 acre-feet of detention capacity on March 15 equivalent to 1.3 inches on the remaining 4,611 square miles. HYDROGRAPH DETERMINATION General Description The hydrologic model used to determine flood hydrographs at Browns Ferry Nuclear Plant and downstream at Wheeler Dam is one in which the total basin is divided into unit areas, the outflows of which are combined to determine total basin outflows. Unit hydrographs are used to compute flows from unit areas. These flows are combined with appropriate time sequencing to compute inflows into the most upstream reservoirs which in turn are routed through the reservoirs using standard 2.4A-4

BFN-25 techniques. Resulting reservoir outflows are combined with the additional local inflows and continued downstream using appropriate time sequencing or routing procedures. The hydrologic model results ensure that each of the unit areas reflect watershed response to rainfall and the total system will reproduce the largest floods experienced since completion of the basic TVA reservoir system. Unit Hydrographs The total watershed for the Guntersville-Wheeler Reservoir inflow estimating system is divided into 62 unit areas as shown by Figure 4. A unit hydrograph has been developed for each of the unit areas from flood hydrographs either recorded at stream gaging stations or estimated from flood data using reservoir headwater elevation, inflow, and discharge data. Figure 5, which contains 14 sheets, shows the unit hydrographs developed for each unit area. Table 2 contains essential dimensional data of each unit hydrograph. Reservoir Routing Tributary reservoir routings were made using standard reservoir routing procedures and flat pool storage conditions. The tributary reservoirs, Tellico and Melton Hill, and the main river reservoirs were routed using unsteady flow techniques. Unsteady flow routings were computer solved with a mathematical model based on the equations of unsteady flow. This model is described in a paper by Jack M. Garrison, Jean-Pierre Granju, and James T. Price entitled "Unsteady Flow Simulation in Rivers and Reservoirs," Journal of the Hydraulics Division, ASCE, Volume 95, No. HY5, September 1969. Boundary conditions prescribed were inflow hydrographs at the upstream boundary, local inflows, and headwater discharge relationships at the downstream boundary based upon standard operating rules or rating curves where geometry controlled, as appropriate. Verification The total area hydrologic model for estimating flood hydrographs was verified by using it to reproduce the March 1973 and the December 2004 floods which are two of the largest floods of record. A comparison between the observed elevations on Wheeler reservoir and the computed outflow at Wheeler Dam for the 1973 flood are shown in Figures 6 and 7, respectively, and for the 2004 flood in Figures 8 and 9, respectively. The unsteady flow model was used to verify historic flood events in all reservoirs where unsteady flow routing was used. Since the Tellico Dam was not closed until 1979, there is no information available to verify the 1973 flood. Therefore, Federal Emergency Management Agency (FEMA) published 100- and 500-year profiles were used for the verification of this portion of the model. 2.4A-5

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-25 Verifying the reservoir models with actual data approaching the magnitude of the PMF is not possible, because no such events have been observed. Therefore, using flows in the magnitude of the PMF (1,200,000 - 1,400,000 cfs) steady state profiles were computed using the two steady state models. An example of this comparison between the profiles for the two models is shown for Wheeler Reservoir in Figure 10. This approach was applied for each of the unsteady flow reservoir models. Similarly, the tailwater rating curve was compared at each project as shown for Guntersville Dam in Figure 11. In this figure, the initial tailwater curve is compared to results from the steady flow models. Breached Earth Embankments The main river dams upstream from Wheeler include earth embankments which could fail if overtopped. Maximum flood level determinations at Browns Ferry Nuclear Plant are based on the postulated failure of any earth embankment which is overtopped with time of failure based on results of the breach analysis of each embankment. The relationship to compute the rate of erosion in an earth dam failure is that developed by the Bureau of Reclamation. The expression relates the volume of eroded fill material to the volume of water flowing through the breach. This relationship is used only to determine the time of failure of an overtopped earth embankment. At the determined time of failure, the embankment is postulated to fail instantaneously and completely. The solution to determine if an earth embankment would fail begins by solving the erosion equation using a headwater elevation hydrograph assuming no failure. Erosion is assumed to occur across the entire earth section and to start at the downstream edge when headwater elevations reach 0.1 foot above the dam top elevation. When erosion began to lower the dam top elevation, the computations included headwater elevation adjustments for increased reservoir outflow resulting from the breach. During the PMF analysis the earthen embankments at Cherokee, Fort Loudoun, Tellico and Watts Bar dams were shown by the hydraulic model to be susceptible to overtopping. The protection of these embankments was recognized as important to TVA dam safety as well as the protection for the nuclear plants. Therefore, these embankments were protected, temporarily, with HESCO Concertainers and are planned to be replaced by permanent protection measures against overtopping. The additional height of these temporary flood barriers are included in the hydraulic model analysis and show that the embankments at these four dams would not be overtopped. Nickajack Dam and Guntersville Dam embankments were modified as a part of the TVA Dam Safety Program based on analysis at the time and were modified in 1992 and 1995, respectively. 2.4A-6

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-25 PROBABLE MAXIMUM PRECIPITATION Probable maximum precipitation (PMP) for the watersheds above Wheeler and Guntersville Reservoirs has been defined for TVA by the Hydrometeorological Branch of the National Weather Service. Two basic storm situations have the potential to produce a maximum flood at Browns Ferry Nuclear Plant. These are (1) storms producing maximum rainfall on the 21,400-square-mile watershed above Chattanooga with the downstream orographically fixed storm pattern, and (2) storms producing maximum rainfall on the 16,170-square-mile watershed above Wheeler Dam and below the major tributary dams. Previous analyses also considered a third storm situation, the 7,980 square mile storm. However, rainfall depth on the watershed above Wheeler reservoir eliminated the 7,980 square mile storm as a candidate storm. Estimates of PMP for the watershed above Chattanooga are fully defined in HMR No. 41. PMP depths for the 21,400-square-mile watershed above Chattanooga are tabulated below. This storm would occur in March. Two possible isohyetal patterns producing these depths are presented in HMR No. 41. The pattern critical to this study is the "downstream pattern" shown in Figure 13. PMP depths for the 16,170-square-mile watershed above Wheeler Dam but below the major tributaries are contained in HMR No. 47. PMP depths for the 16,170-square-mile watershed are also tabulated below. The isohyetal pattern for the storm is shown in Figure 14. The pattern and depths are for a storm centered within 35 miles of Nickajack Dam. A 72-hour storm 3 days antecedent to the main storm was assumed to occur in all PMP situations with storm depths equivalent to 40 percent of the main storm outlined in Bulletin 41. 2.4A-7

BFN-25 Basin Depth, Inches Storm, 72-Hour Main Storm Sq. Mi. Antecedent Storm 6-Hour 24-Hour 72-Hour 21,400 6.08 4.48 9.43 14.48 16,170 6.22 4.2 9.65 15.55 RAIN-RUNOFF RELATIONSHIPS Precipitation excess resulting from the PMP storm was computed using multivariable relationships developed and used in the day-to-day operation of the TVA system. These relationships, developed from a study of flood records, relate the amount of precipitation excess expected from a given storm rainfall to the week of the year, an antecedent precipitation index (API), and geographical location. The relationships are such that the subtraction from rainfall to compute precipitation excess is greatest at the start of the storm and decreased to no subtraction where precipitation excess is equal to rainfall in the late part of extreme storms. An API determined from an 11-year period of historical rainfall records (1997-2007) was used at the start of the antecedent storm. The precipitation excess computed for the main storm is not sensitive to variations in adopted initial conditions because of the large antecedent storm. CONDITIONS CREATING PROBABLE MAXIMUM FLOOD Critical Storm Enough storm arrangements including different storm centerings, seasonal variability, and consideration of potential dam failures were investigated to ensure selection of the arrangement which would produce the PMF discharge and elevation at Browns Ferry Nuclear Plant. The critical PMP storm was determined to be the 21,400-square-mile downstream centered storm, which would follow an antecedent storm commencing on March 15. The antecedent storm would produce an average precipitation of 6.08 inches on the basins above Wheeler, would be followed by a 3-day dry period, and then by the main storm which would produce an average precipitation of 14.48 inches in 3 days. Figure 13 is an isohyetal map of the maximum 3-day PMP which was used to compute the PMF. Precipitation Excess Median moisture conditions as determined from the 11 year period of historical records (1992-2007) were used to determine the API at the start of the storm sequence. However, the antecedent storm is so large that the precipitation excess computed for the main storm is not sensitive to variations in adopted initial moisture conditions. The precipitation excess from the critical PMP storm was 3.85 inches for 2.4A-8

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-25 the 3-day antecedent storm and 12.74 inches for the 3-day main storm. Table 3 displays the rain and precipitation excess for both the main and antecedent storm for the critical storm for each of the 62 subwatersheds of the hydrologic model. Reservoir Operations Reservoir routings were started with all reservoirs at their respective median mid-March elevations. The reservoir operating guides applied during the unsteady flow model simulations mimic, to the extent possible, operating policies and are within the current reservoir operating flexibility. In addition to spillway discharge, turbine and sluice discharges were used to release water from the tributary reservoirs. Turbine capacities are also used in the main river reservoirs up to the point where the head differentials are too small and/or the powerhouse would flood. All gates were assumed to be operable without failures during the flood. The flood from the antecedent storm occupied about 63 percent of the reserved detection capacity at the beginning of the main storm (day 7 of the event). Reservoir levels are at or above guide levels at the beginning of the main storm in all but Apalachia and Fort Patrick Henry reservoirs, which have no reserved flood detention capacity. Dam Failures PROBABLE MAXIMUM FLOOD The PMF peak discharge at Browns Ferry Nuclear Plant is 1,193,671 cfs. The hydrograph of the calculated discharge is shown in Figure 15. However, the design and licensing basis peak discharge is being maintained as 1,200,000 cfs. Velocities at the site would average 6 feet per second in the channel and up to 4 feet per second in the overbank at the time of peak discharge. The PMF elevation at Browns Ferry Nuclear Plant is 571.7. The hydrograph of the calculated PMF elevation is shown in Figure 16. However, the design and licensing basis PMF elevation is being maintained as 572.5. 2.4A-9

SECURITY-RELATED INFORMATION - WITHHELD UNDER 10 CFR 2.390 BFN-25 Wind Waves Some wind waves are likely when the PMF is cresting at Browns Ferry Nuclear Plant. The flood would be near its crest elevation for a day beginning about 6-1/2 days after cessation of the PMP storm. A reasonably severe windstorm producing 45 mph sustained wind speeds could occur coincidentally with the PMF. A wind from the SE will produce the largest waves at the site. A wind of this magnitude and from this direction can generate 5-foot waves (crest to trough). The analysis of wave heights used the 1% wave of which about 5 per hour will occur. Consequent wave runup above the flood level would be about 5 feet on a vertical wall. Maximum Elevation LOCAL DRAINAGE In addition to flooding from the Tennessee River, Browns Ferry Nuclear Plant could possibly be flooded by intense local storms. Flooding sources include: (1) the small unnamed stream northwest of the plant, a portion of which has been relocated (see Figure 21), (2) the area draining to the switchyard drainage channel, (3) the main plant area, and (4) the area draining to the cooling tower system of channels. These areas, labeled Area 1 through Area 4 respectively, are shown on Figure 22. Direction of flow for runoff is indicated by arrows. Figure 22a shows the plant topography. These areas were evaluated for a local storm producing probable maximum precipitation (PMP). PMP has been defined for TVA by the Hydrometeorological Branch of the National Weather Service and is described in Hydrometeorological Report No. 56 (HMR-56), "Probable Maximum and TVA Precipitation Estimates with Areal Distribution for Tennessee River Drainages less Than 3,000 Mi2 in Area." HMR-56 supersedes HMR-45, which was used to define PMP in the original analysis. A 6-hour storm which would produce a total of 34.4 inches of rainfall with a maximum 1-hour amount of 16.7 inches was determined to be critical and was used to develop probable maximum flood inflows. The mass curve is shown in Figure 22b. Runoff was conservatively assumed equal to rainfall. Ice accumulation would occur only at infrequent intervals because of the temperate climate. Maximum winter precipitation concurrent with ice accumulation would 2.4A-10

BFN-25 impose less severe conditions on the drainage system than would the local PMP, which is associated with severe summer thunderstorm activity. Local Stream An unnamed stream northwest of the plant with drainage area of 1.35 square miles formerly flowed through the plant area. Figure 21 shows this watershed. The stream has now been diverted to flow along the west boundary of the plant as shown in Figure 22. The channel is designed with capacity sufficient to carry the maximum possible (probable maximum) flood without flooding the plant. The peak flood discharge was estimated using a 1-hour unit hydrograph developed synthetically by comparison with gaged watersheds in the region. The adopted unit hydrograph is shown on Figure 23. The maximum possible (probable maximum) peak discharge is 17,200 cfs. This compares with 14,000 cfs determined in the original analysis. The maximum possible (probable maximum) flood hydrograph for this stream and the site PMP hyetograph are shown on Figure 24. The alignment of the relocated channel is shown on Figure 22, typical sections are shown on Figure 25, and plan and profiles, including water surface profiles and minimum grade levels between the channel and plant, are shown on Figures 26, sheets 1 and 2. Maximum water surface elevations were completed using standard step backwater methods. As shown on Figures 26, sheets 1 and 2, the channel will pass 17,200 cfs with maximum water surface elevations below the ground, the dike, and the road which protect the plant and cooling tower areas from flooding. SWITCHYARD DRAINAGE CHANNEL The 100-acre area draining to the switchyard drainage channel is shown on Figure 22 (Area 2). Runoff from this area is diverted through the channel to the Tennessee River southeast of the plant. Two inflow hydrographs were developed: (1) a lateral inflow hydrograph from the 35-acre area adjacent to the channel, distributed uniformly along the length of the channel, and (2) a point inflow hydrograph from the 65-acre area draining to the upstream end of the channel. The lateral inflow hydrograph is equivalent to the PMP hyetograph using 5-minute intervals. The point inflow hydrograph was developed by considering overland flow travel time for a number of discrete points within the 65-acre area. Travel times were estimated for runoff from each respective point to the channel. Peak flood elevations in the channel were computed using unsteady flow routing methods. In the routing it was conservatively assumed that the three culverts at the oil skimmer structure at the downstream end of the channel would be clogged with debris and would provide no discharge capacity during the flood. The channel can pass the maximum possible (probable maximum) flood without flooding safety-related structures. The maximum 2.4A-11

BFN-30 water surface elevation at the holding pond at the downstream end of the channel would be 574.8. The maximum water surface elevation at the north corner of the switchyard would not exceed the switchyard elevation of 578. Main Plant Area Plant buildings could possibly be flooded by an intense local storm which would exceed the capacity of the yard drainage system. The main plant area drains 41 acres and is shown on Figure 22 (Area 3). Plant surface drainage was investigated to determine if a plant maximum possible (probable maximum) storm would exceed plant grade El. 565 and cause flooding of safety-related plant structures. All underground drains were assumed to be clogged. All surface drainage with the exception of that between the office building and service building (patio plaza area) is adequate. Flow from the 3.3-acre patio plaza area northwest of the service building will drain to the employee courtyard west of the service building. Flooding at this location results from runoff collecting in a low point in the area with flood elevations controlled by the narrow opening between the temporary plant engineering building and the radwaste evaporator building. The peak elevation was determined to be 566.6, which is 1.6 feet above plant grade El. 565. The elevation was computed by storage routing the inflow hydrograph equivalent to the PMP hyetograph using 5-minute intervals. The time available between the start of the most intense rainfall and the time flood levels exceed plant grade El. 565 varies from 5 to 21 minutes, depending on the assumed distribution within the critical local PMP 6-hour storm. The Plant Maintenance Building (PMB) was constructed within the main plant drainage area that includes the patio plaza area. Construction of the PMB included demolition of the temporary plant engineering building and grading of the area to accommodate drainage in the area of the PMB. The PMB has a larger footprint than the removed temporary plant engineering building, however, the grading changes for the PMB generally increased storage, more than what was lost due to the footprint of the PMB building. Storage routing was used to evaluate the plant area. The revised storage at each elevation in the vicinity of the PMB was investigated and determined to be greater than the conditions prior to the grading changes and construction of the PMB. The grading changes in the PMB area were analyzed using the revised storage data to determine the effects of the building on local intense precipitation site drainage. Three temporal distributions, identified as early, middle, and late, to reflect the placement of the peak rainfall were analyzed. The resulting water surface elevations did not impact the local intense precipitation site drainage. Site analysis also includes the cooling tower system, the switchyard drainage channel, and an unnamed channel along the west boundary of the plant. The cooling tower system includes runoff from the plant area as an upstream contribution to the system. There is no impact from the PMB on the local intense precipitation site drainage. Therefore, the cooling tower system analysis would also be 2.4A-12

BFN-30 unaffected. The plant area is independent of the switchyard drainage channel and the unnamed channel analyses. Therefore, the PMB does not affect the switchyard drainage channel or the unnamed channel. In the vicinity of the radioactive waste, reactor, and diesel generator buildings, water-surface elevations will not exceed El. 565. Peak water-surface elevations were determined by storage routing the inflow hydrograph using standard weir formulas and flat pool assumptions. The control section was taken to be the perimeter road south of the reactor building with elevation at 564. The total inflow hydrograph was determined by considering overland flow travel time for a number of discrete points within the 41-acre area. Travel times were determined for runoff from each respective point to the perimeter road. The total area was then divided into subareas of equal travel time, with the longest travel times for those areas farthest from the perimeter road. Each subarea contributes to total flow, with respective subarea inflow hydrographs equivalent to the PMP hyetograph using 5-minute time intervals. The total inflow hydrograph was then computed by summing the respective subarea inflow hydrographs, with each lagged by an amount equal to its travel time. Travel times ranged from 0 to 20 minutes, with 0 reflecting instantaneous watershed response. Water-surface elevations at the radioactive waste, reactor, and diesel generator buildings at the time of maximum possible (probable maximum) flood flow could be affected if maximum water levels in the cooling water discharge channel downstream were to exceed El. 564. For a short duration during the possible maximum precipitation (PMP) concurrent with cooling towers operation water surface elevation in cooling water discharge channel reaches 564.93, overflowing the perimeter road and mixing with runoff from the main plant area. Water surface elevation at the radioactive waste, reactor, and diesel generator buildings was evaluated and determined not to exceed El. 565. Cooling Tower System The 179-acre area draining to the cooling tower system of channels is shown on Figure 22 (Area 4). Runoff from this area is diverted to the Tennessee River through the operation of several gate structures. The cooling tower system of channels has sufficient capacity to pass the combined maximum possible (probable maximum) flood runoff and condenser water without flooding the plant for any mode of plant operation. Peak water-surface elevations were determined using storage routing methods. Local inflow hydrographs for each of the basins within the cooling tower system of channels were equivalent to the PMP hyetograph using 5-min ute intervals. These hydrographs were augmented where appropriate for condenser water and main plant area runoff. 2.4A-13

BFN-25 Bellefonte Nuclear Plant, Units 3 & 4 COL Application Table 1 Part 2, FSAR (Sheet 1 of 2) TABLE 2.4.1-202 (Sheet 1 of 2) FACTS ABOUT TVA DAMS AND RESERVOIRS (a) Major Dams Main River Dam Locations Drainage Cost(b) Construction Dam First Unit in Last Unit in Winter Net Number of Location Height Length Type Lock Length of Miles of Reservoir Area of Reservoir Elevation Reservoir Volume (Acre Feet) Jan. 1 Project Number Projects Area (Millions) Began Closure Service Service Dependable Generating of Dam of of Dam of Chamber Reservoir(e) Shoreline(e) Surface Original Controlled of Dams (c) (Feet Above Mean Sea Level) Above (Actual or (Actual or Capacity Units Above Dam(w) (Feet) Dam(d) Size: Width x (Miles) Area(e) River Storage in Dam Scheduled) Scheduled) (Megawatts) Mouth of (Feet) Length x (Acres) Bed (Acre Project River State (Square River Maximum Lift (Acres) Jan. 1 Top of June 1 At Jan. 1 At Top of At June 1 Feet)(f) Miles) (Miles) (Feet) Flood Gates Flood Flood Gates Flood Guide Guide Guide Guide Elevation Elevation Elevation Elevation (g) v) Kentucky Tennessee KY 40,200 128.8 7/1/1938 8/30/1944 9/14/1944 1/16/1948 184 5 22.4 206 8,422 CGE 110x600x75( 184.3 2064.3 160,300 25,200 354.0 375.00 359.0 2,121,000 6,129,000 2,839,000 4,008,000 TN River 9 (i) Pickwick Tennessee TN 32,820 120.9 12/30/1934 2/8/1938 6/29/1938 12/31/1952 229 6 206.7 113 7,715 C G E 110x1000x63 52.7 490.6 42,700 9,580 408.0 418.00 414.0 839,300 1,332,000 1,119,000 492,700 TN River 9 Landing 110x600x63 60x232x48 (h) i) Wilson Tennessee AL 30,750 133.5 4/14/1918 4/14/1924 9/12/1925 4/12/1962 663 21 259.4 137 4,541 CG 110x600x100( 15.5 166.2 15,600 9,108 504.7 507.88 507.7 589,700 640,200 637,200 50,500 TN River 9 60x300x52 60x292x48 Wheeler Tennessee AL 29,590 69.0 11/21/1933 10/3/1936 11/9/1936 12/18/1963 361 11 274.9 72 6,342 CG 60x400x52 74.1 1027.2 67,070 17,600 550.5 556.28 556.0 742,600 1,069,000 1,050,000 326,500 TN River 9 (i) 110x600x52 Guntersville Tennessee AL 24,450 74.2 12/4/1935 1/16/1939 8/1/1939 3/24/1952 124 4 349.0 94 3,979 CGE 60x360x45 75.7 889.1 66,000 12,065 593.0 595.44 595.0 886,600 1,048,700 1,018,000 162,100 TN River 9 (i) 110x600x45 (j) Nickajack Tennessee TN 21,870 56.1 4/1/1964 12/14/1967 2/20/1968 4/30/1968 105 4 424.7 81 3,767 CGE 110x800x41 46.3 178.7 10,200 4,200 632.5- 635.00 632.5- N/A 251,600 N/A N/A TN River 9 110x600x41 634.5 634.5 (v) Chickamauga Tennessee TN 20,790 74.4 1/13/1936 1/15/1940 3/4/1940 3/7/1952 119 4 471.0 129 5,800 CGE 60x360x53 58.9 783.7 36,050 9,500 675.0 685.44 682.5 392,000 737,300 622,500 345,300 TN River 9 (r) Watts Bar Tennessee TN 17,310 66.4 7/1/1939 1/1/1942 2/11/1942 4/24/1944 182 5 529.9 112(x) 2,960 CGE 60x360x70 95.5 721.7 37,500 10,343 735.0 745.00 741.0 796,000 1,175,000 1,010,000 379,000 TN River 9 (s) Fort Loudoun Tennessee TN 9,550 45.3 7/8/1940 8/2/1943 11/9/1943 1/27/1949 162 4 602.3 122(x) 4,190 CGE 60x360x80 60.8 378.2 14,000 4,420 807.0 815.00 813.0 282,000 393,000 363,000 111,000 TN River 9 Pumped Storage Project Raccoon Tennessee TN 1 237.8 7/1/1970 7/11/1978 12/31/1978 8/31/1979 1653 4(k) 230 8,500 ER N/A 528 1530.0- N/A N/A N/A Raccoon Mtn. 1 Mountain 1672.0 Tributary Power Projects Tims Ford Elk TN 529 43.8 3/28/1966 12/1/1970 3/1/1972 3/1/1972 36 1 133.3 175 1,580 ER N/A 34.2 308.7 10,500 565 873.0 895.00 888.0 388,400 608,000 530,000 219,600 Elk River 1 Apalachia Hiwassee NC 1,018 29.4 7/17/1941 2/14/1943 9/22/1943 11/17/1943 82 2 66.0 150 1,308 CG N/A 9.8 31.5 1,100 80 1272.0- 1280.00 1272.0- N/A 57,800 N/A N/A Hiwassee 4 1280.0 1280.0 Hiwassee Hiwassee NC 968 42.5 7/15/1936 2/8/1940 5/21/1940 5/24/1956 141 (l) 75.8 307 1,376 CG N/A 22.2 164.8 5,870 1,000 1485.0 1526.50 1521.0 228,400 434,000 399,000 205,600 Hiwassee 4 2 Chatuge Hiwassee NC 189 9.5 7/17/1941 2/12/1942 12/9/1954 12/9/1954 13 1 121.0 150 2,850 E N/A 13.0 128.0 6,700 107 1918.0 1928.00 1926.0 177,900 240,500 226,600 62,600 Hiwassee 4 Ocoee 1(h)(m) Ocoee TN 595 11.8 8/00/1910 12/15/1911 1/28/1912 0/0/1914 24 5 11.9 135 840 CG N/A 7.5 47.0 1,620 170 820.0 830.76 829.0 64,300 83,300 79,900 19,000 Ocoee 3 (h) Ocoee 2 Ocoee TN 512 28.8 5/00/1912 10/00/1913 10/0/1913 10/00/1913 23 2 24.2 30 450 O N/A N/A N/A N/A N/A N/A 1115.20 N/A N/A N/A N/A N/A Ocoee 3 Ocoee 3 Ocoee TN 492 4.9 7/17/1941 8/15/1942 4/30/1943 4/30/1943 29 1 29.2 110 612.1 CG N/A 7.0 24.0 600 260 1428.0- 1435.00 1428.0- N/A 4,200 N/A N/A Ocoee 3 1435.0 1435.0 (n) Blue Toccoa GA 232 20.4 11/00/1925 12/6/1930 7/0/1931 7/0/1931 13 1 53.0 175 1,000 E N/A 11.0 68.1 3,220 182 1668.0 1691.00 1687.0 127,400 195,900 182,800 68,500 Toccoa/Ocoee 1 Ridge(h)(m) Nottely Nottely GA 214 17.2 7/17/1941 1/24/1942 1/10/1956 1/10/1956 18 1 21.0 197 2,300 RE N/A 20.2 102.1 3,970 170 1762.0 1780.00 1777.0 112,700 174,300 162,000 61,600 Hiwassee 4 Melton Hill Clinch TN 3,343 21.5 9/6/1960 5/1/1963 7/3/1964 11/11/1964 79 2 23.1 103 1,020 CG 75x400x60 44 193.4 5,690 1,645 792.0- 796.00 792.0- N/A 126,000 N/A N/A Clinch 2 795.0 795.0 (u) Norris Clinch TN 2,912 46.1 10/1/1933 3/4/1936 7/28/1936 9/30/1936 110 2 79.6 265 1,860 CGE N/A 129.0 809.2 34,000 2,930 1000.0 1034.00 1,020.0 1,439,000 2,552,000 2,040,000 1,113,000 Clinch 2 (o) (o) (o) (o) (o) Tellico Little TN TN 2,627 117.0 3/7/1967 11/29/1979 0.3 129(x) 3,238 CGE 33.2 357.0 15,600 2,133 807.0 815.00 813.0 304,000 424,000 392,000 120,000 Little TN 2 Fontana Little TN TN 1,571 69.1 1/1/1942 11/7/1944 1/20/1945 2/4/1954 304 3 61.0 480 2,365 CG N/A 29.0 237.8 10,290 1,650 1653.0 1710.00 1703.0 929,000 1,443,000 1,370,000 514,000 Little TN 2 Douglas French TN 4,541 83.0 2/2/1942 2/19/1943 3/21/1943 8/3/1954 111 4 32.3 202 1,705 CG N/A 43.1 512.5 28,070 3,170 954.0 1002.00 994.0 379,000 1,461,000 1,223,500 1,082,000 French Broad 1 Broad Cherokee Holston TN 3,428 29.3 8/1/1940 12/5/1941 4/16/1942 10/7/1953 148 4 52.3 175(x) 6,760 CGE N/A 54.0 394.5 29,560 2,426 1045.0 1075.00 1071.0 791,600 1,541,000 1,422,900 749,400 Holston 4 R Fort Patrick South Fork TN 1,903 18.9 5/14/1951 10/27/1953 12/5/1953 2/22/1954 41 2 8.2 95 737 CG N/A 10.4 31.0 840 339 1258.0- 1263.00 1258.0- N/A 26,900 N/A N/A Holston 4 Henry Holston 1,263.0 1263.0 (t) Boone South Fork TN 1,840 15.5 8/29/1950 12/16/1952 3/16/1953 9/3/1953 89 3 18.6 160 1,532 ECG N/A 32.7 126.6 4,130 719 1364.0 1385.00 1382.0 117,600 193,400 180,500 75,800 Holston 4 Holston (p) South Holston South Fork TN 703 23.1 8/04/1947 11/20/1950 2/13/1951 2/13/1951 44 1 49.8 285 1,600 ER N/A 23.7 181.9 7,600 710 1708.0 1742.00 1729.0 511,300 764,000 658,000 252,800 Holston 4 Holston (p) Watauga Watauga TN 468 22.1 7/22/1946 12/1/1948 8/30/1949 9/29/1949 66 2 36.7 332 900 ER N/A 16.3 104.9 6,440 313 1952.0 1975.00 1959.0 524,200 677,000 568,500 152,800 Watauga 2 (h) Wilbur Watauga TN 471 1.6 00/00/1909 00/00/1912 0/0/1912 7/19/1950 11 4 34.0 76 375.5 CG N/A 1.8 4.8 70 1641.0- 1650.00 1641.0- N/A 714 N/A N/A Watauga 2 1648.0 1648.0 (a)(h) Great Falls Caney Fork TN 1,675 21.4 12/7/1915 12/8/1916 0/0/1916 0/0/1925 36 2 91.1 92 800 CG N/A 22.0 120.0 1,830 1,490 785.0 805.30 800.0 19,700 50,200 40,600 30,500 Caney Fork 1 (q) (q) Nolichucky Nolichucky TN 1,183 0.1 00/00/1913 46.0 94 482 CG 26.0 380 1240.90 Nolichucky 1 (h)( m) (retired)

BFN-25 Table 1 (Sheet 2 of 2) FACTS ABOUT TVA DAMS AND RESERVOIRS a) All in the Tennessee Valley, except for Great Falls which is in the Cumberland Valley. b) Cost of plant including the inception balance of the plant and all additions and retirements from the plant. Transmission assets are not included. c) Winter net dependable capacity as of October 2009. Winter net dependable capacity is the amount of power a plant can produce on an average winter day, minus the electricity used by the plant itself. d) E: Earth; R: Rock fill; G: Gravity; C: Concrete; O: Other (Codes for each dam are listed in order of importance.) e) At June 1 flood guide elevation. f) Volume between the January 1 elevation and top of gates. g) Connected to Barkley Reservoir by 1-1/2 mile canal, which opened July 14, 1966. h) Acquired: Wilson by transfer from the U.S. Army Corps of Engineers in 1933; Ocoee 1, Ocoee 2, Blue Ridge, and Great Falls by purchase from Tennessee Electric Power Company in 1939; Wilbur and Nolichucky (retired) by purchase from East Tennessee Power and Light Company in 1945. Subsequent to acquisition, TVA installed additional units at Wilson and Wilbur. Reconstructed flume at Ocoee 2 was placed in service in November 1983. i) Main locks placed in operation in 1959 at Wilson, 1963 at Wheeler, 1965 at Guntersville, and 1984 at Pickwick Landing. j) Construction of main lock at Nickajack limited to underwater construction. k) Generating units at Raccoon Mountain are reversible Francis type pump-turbine units, each with 428,400 kW generator rating and 612,000 hp pump motor rating. l) Unit 2 at Hiwassee is a reversible Francis type pump-turbine unit with 95,000 kW generator rating and 121,530 hp pump motor rating at 200 ft. net head. m) Ocoee 1 creates Parksville Reservoir, Nolichucky (retired) creates Davy Crockett Reservoir, and Blue Ridge creates Toccoa Reservoir. n) Construction of Blue Ridge discontinued early in 1926; resumed in March 1929. o) Tellico project has no lock or powerhouse. Streamflow through navigable canal to Fort Loudoun Reservoir permits navigation and increases average annual energy output at Fort Loudoun. p) Initial construction of South Holston and Watauga started February 16, 1942; temporarily discontinued to conserve critical materials during WWII. q) Generating units at Nolichucky were removed from system generating capacity in August 1972. The dam was renovated and modified to convert the reservoir for use as a wildlife preserve. r) Includes 72.4 miles up the Tennessee River to Fort Loudoun Dam and 23.1 miles up Clinch River to Melton Hill Dam. s) Includes 6.5 miles up the French Broad River and 4.4 miles up the Holston River. t) Includes 17.4 miles up the South Fork Holston River and 15.3 miles up the Watauga River. u) Includes 73 miles up the Clinch River and 56 miles up the Powell River. v) The U.S. Army Corps of Engineers is increasing the size of lock structures at Kentucky and Chickamauga. w) The structural height of the dam is the vertical distance from the lowest point of the excavated foundation to the top of the dam. Top of dam refers to the highest point of the water barrier on an embankment (or top of parapet wall) and deck elevation (or top of parapet wall) for concrete structures. x) As an interim measure to prevent overtopping, these four dams were raised by HESCO Concertainer floodline units. Watts Bar - 3 feet: embankment at elevation 767 raised to elevation 770. Fort Loudoun - 3.75 feet: embankment at elevation 830 was raised 7 feet to elevation 837 (3.75 feet above top of concrete wall at elevation 833.25). Tellico - 4 feet: embankment at elevation 830 raised to elevation 834. Cherokee - 3 feet: embankment at elevation 1089 raised to elevation 1092.

BFN25 Table 2 (Sheet 1 of 2) UNIT HYDROGRAPH DATA GIS Unit Area Drainage Area Duration Number Name QP CP TP W50 W75 TB (sq.mi.) (hrs.) 1 Asheville 944.4 6 14,000 0.21 12 39 15 168 2 Newport, French Broad 913.1 6 43,114 0.66 12 10 4 48 3 Newport, Pigeon 667.1 6 30,910 0.65 12 8 4 90 4 Embreeville 804.8 4 33,275 0.65 12 10 7 80 5 Nolichucky Local 378.7 6 11,740 0.44 12 14 6 90 6 Douglas Local 835 6 47,207 0.27 6 8 5 60 7 Little Pigeon River 352.1 4 17,000 0.75 12 10 6 66 8 French Broad Local 206.5 6 8,600 0.2 6 13 6 60 9 South Holston 703.2 6 15,958 0.53 18 25 17 96 10 Watauga 468.2 4 37,002 0.74 8 6 3 32 11 Boone Local 667.7 6 22,812 0.16 6 13 7 90 12 Fort Patrick Henry 62.8 6 2,550 0.19 6 12 7 66 13 Gate City 668.9 6 11,363 0.56 24 34 26 108 14&15 Total Cherokee Local 854.6 6 25,387 0.42 12 20 10 54 16 Holston River Local 289.6 6 8,400 0.27 9 18 12 96 17 Little River 378.6 4 11,726 0.68 16 15 7 96 18 Fort Loudoun Local 323.4 6 20,000 0.29 6 10 5 36 19 Needmore 436.5 6 9,130 0.49 18 22 12 126 20 Nantahala 90.9 2 3,130 0.38 8 16 11 54 21 Bryson City 653.8 6 26,000 0.43 10 13 7 60 22 Fontana Local 389.8 4 17,931 0.14 4 14 7 28 23 Little Tennessee Local Fontana to Chilhowee Dam 404.7 6 16,613 0.58 12 10 4 84 24 Little Tennessee Local Chilhowee to Tellico Dam 650.2 6 22,600 0.49 12 15 8 54 25 Watts Bar Local above Clinch River 295.3 6 11,063 0.18 6 10 4 90 26 Norris Dam 2912.8 6 43,773 0.07 6 18 6 102 27 Melton Hill Local 431.9 6 12,530 0.14 6 19 10 90 33 Local above mil 16 37.2 2 4,490 0.94 6 3 2 48 34 Poplar Creek 135.2 2 2,800 0.61 20 26 13 90 35 Emory River 868.8 4 36,090 0.39 8 11 6 84 36 Local Area at Mouth 29.3 2 3,703 0.99 6 3 2 48 37 Watts Bar Local below Clinch River 408.4 6 16,125 0.19 6 10 4 90 38 Chatuge 189.1 1 19,062 0.24 2 3 2 37 39 Nottely 214.3 1 44,477 0.16 1 1 1 12 40 Hiwassee Local 565.1 6 23,349 0.58 12 11 6 96 41 Apalachia 49.8 1 5,563 0.26 2 4 1 23 42 Blue Ridge 231.6 2 11,902 0.4 6 10 7 60 43 Ocoee No. 1 Local 362.6 6 17,517 0.23 6 12 8 36 44A Hiwassee at Charleston 686.6 6 9,600 0.59 30 39 23 108 44B Hiwassee at Mouth 396 6 16,870 1 18 11 6 78 45 Chickamauga Local 792.1 6 32,000 0.38 9 14 7 36 46 South Chickamauga Creek 428.1 6 6,267 0.48 24 39 18 132 47A Nickajack Local 545.7 6 9,059 0.16 9 35 8 144 47B North Chickamauga Creek Local 98.3 4 3,000 0.67 16 15 6 112

BFN25 Table 2 (Sheet 2 of 2) UNIT HYDROGRAPH DATA GIS Unit Area Drainage Area Duration Number Name QP CP TP W50 W75 TB (sq.mi.) (hrs.) 48 Sequatchie 400 4 8,562 0.47 16 16 7 140 49 Guntersville North Local 1027.1 6 22,089 0.4 15 20 11 138 50 Guntersville South Local 1068.9 6 22,963 0.4 15 19 11 132 51 Paint Rock River near Woodvi lle 321.1 6 11,363 1.00 18 10 5 102 52 Paint Rock Local 138.1 6 6,103 1.24 18 10 5 72 53 Flint River near Chase 343.1 6 16,356 0.89 12 12 8 60 54 Flint River Local 224.8 6 7,962 1.33 24 15 7 78 55 Cotaco Creek at Florette 136.2 6 3,174 0.66 18 21 11 96 56 Cotaco Creek Local 101.1 6 2,644 0.98 24 19 9 84 57 Limestone Creek near Athens 121.3 4 10,618 1.64 12 5 3 40 58 Limestone Creek Local 157.4 6 6,407 0.76 12 12 6 54 59 Tims Ford Dam 533.2 6 17,555 0.31 6 16 6 78 60 Elk River Local, Tims Ford to F ayetteville 293.3 6 7,044 0.68 18 24 14 78 61 Elk River Local, Fayetteville to Prospect 490.2 6 8,874 0.85 30 29 16 102 62 Richland Creek at Mouth 488.0 6 11,529 1.11 30 23 15 90 63 Sugar Creek at Mouth 177.0 4 8,512 1.20 16 16 9 92 64 Elk River Local, Mile 16.5 to P rospect Gage 145.1 6 5,913 1.53 24 12 6 72 65 Wheeler Local 1380.0 6 46,747 0.64 12 16 9 78 Definition of Symbols QP = Peak discharge in cfs. CP = Snyder coefficient. TP = Time in hours from beginning of precipitation excess to peak of unit hydrograph. W50 = Width in hours at 50 percent of peak discharge. W75 = Width in hours at 75 percent of peak discharge. TB = Base length in hours of unit hydrograph.

BFN25 Table 3 (Sheet 1 of 2) PROBABLE MAXIMUM FLOOD RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Index Rainfall Runoff Rainfall Runoff No. Sub-basin Name (inches) (inches) (inches) (inches) 1 French Broad River at Asheville 6.18 2.91 18.12 15.44 2 French Broad River, Newport to Asheville 6.18 3.67 18.42 16.43 3 Pigeon River at Newport 6.18 2.91 19.26 16.58 4 Nolichucky River at Embreeville 6.18 3.67 15.30 13.31 5 Nolichucky local, Embreeville to Nolichucky Dam 6.18 3.67 15.42 13.43 6 Douglas Dam local 6.18 4.43 17.16 15.94 7 Little Pigeon River at Sevierville 6.18 3.81 21.12 19.13 8 French Broad River local 6.18 3.81 19.38 17.39 9 South Holston Dam 6.18 4.60 12.12 10.90 10 Watauga Dam 6.18 3.67 12.96 10.97 11 Boone local 6.18 3.81 13.86 11.87 12 Fort Patrick Henry 6.18 4.60 14.34 13.12 13 North Fork Holston River near Gate City 6.18 4.60 12.30 11.08 14-15 Cherokee and Holston River below Fort Pat & Gate City 6.18 4.60 15.42 14.20 16 Holston River local, Cherokee Dam to Knoxville gage 6.18 4.60 16.74 15.52 17 Little River at mouth 6.18 3.81 20.82 18.83 18 Fort Loudoun local 6.18 3.81 17.28 15.29 19 Little Tennessee River at Needmore 6.18 2.73 20.22 17.54 20 Nantahala 6.18 2.73 20.94 18.26 21 Tuckasegee River at Bryson City 6.18 2.91 20.04 17.36 22 Fontana local 6.18 2.91 19.56 16.88 23 Little Tennessee River local, Fontana Dam to Chilhowee Dam 6.18 2.91 22.50 19.82 24 Little Tennessee River local, Chilhowee Dam to Tellico Dam 6.18 2.91 19.26 16.58 25 Watts Bar local above Clinch River 6.18 3.81 15.84 13.85 26 Clinch River at Norris Dam 6.18 4.60 13.56 12.34 27 Melton Hill local 6.18 4.27 15.42 14.01 33 Clinch River local above mile 16 6.18 4.43 15.42 14.01 34 Poplar Creek at mouth 6.18 4.43 14.88 13.47 35 Emory River at mouth 6.18 4.43 12.78 11.37 36 Clinch River local, mouth to mile 16 6.18 4.43 14.94 13.53 37 Watts Bar local below Clinch River 6.18 4.43 14.28 12.87 38 Chatuge Dam 6.18 2.91 21.12 18.44 39 Nottely Dam 6.18 2.91 18.66 15.98 40 Hiwassee River local below Chatuge and Nottely 6.18 2.73 18.18 15.50 41 Apalachia local 6.18 3.81 18.18 16.19 42 Blue Ridge Dam 6.18 2.91 22.14 19.46 43 Ocoee No. 1 local, Ocoee No. 1 to Blue Ridge Dam 6.18 2.91 18.42 15.74 Hiwassee River local, Charleston gage at mile 18.9 to Apalachia 44A and Ocoee No. 1 Dams 6.18 3.81 15.48 13.49 44B Hiwassee River local, mouth to Charleston gage at mile 18.9 6.18 4.27 14.52 13.11 45 Chickamauga local 6.18 4.27 13.56 12.15

BFN25 Table 3 (Sheet 2 of 2) PROBABLE MAXIMUM FLOOD RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Index Rainfall Runoff Rainfall Runoff No. Sub-basin Name (inches) (inches) (inches) (inches) 46 South Chickamauga Creek near Chattanooga 6.18 4.11 12.06 10.65 47A Nickajack local below North Chickamauga Creek @ gage 6.18 4.27 11.46 10.05 47B North Chickamauga Creek @ gage 6.18 4.27 12.30 10.89 48 Sequatchie River at Whitwell 6.18 4.27 12.06 10.65 49 Guntersville North local 6.18 4.27 10.44 9.03 50 Guntersville South local 6.18 4.27 9.90 8.49 51 Paint Rock River near Woodville 5.58 3.57 9.84 8.43 52 Paint Rock Local 5.58 3.57 9.84 8.43 53 Flint River near Chase 5.58 3.72 9.84 8.43 54 Flint River Local 5.58 3.43 9.84 8.43 55 Cotaco Creek at Florette 5.58 3.72 9.84 8.43 56 Cotaco Creek Local 5.58 3.72 9.84 8.43 57 Limestone Creek near Athens 5.58 3.72 9.84 8.43 58 Limestone Creek Local 5.58 3.72 9.84 8.43 59 Tims Ford Dam 5.58 3.28 9.84 8.24 60 Elk River Local, Tims Ford to Fayetteville 5.58 3.28 9.84 8.24 61 Elk River Local, Fayetteville to Prospect 5.58 3.28 9.84 8.24 62 Richland Creek at Mouth 5.58 3.72 9.84 8.43 63 Sugar Creek at Mouth 5.58 3.72 9.84 8.43 64 Elk River Local, Mile 16.5 to Prospect Gage 5.58 3.72 9.84 8.43 65 Wheeler Local 5.58 3.72 9.84 8.43 Basin Averages (inches) 6.08 3.85 14.48 12.74 a Index No. corresponds to Figure 4 numbered areas. b Adopted antecedent precipitation index prior to antecedent storm varies by unit area, ranging from 0.78-1.47 inches. c Computed antecedent precipitatoin index prior to main storm, 3.65 inches.

BFN-25 Table 4 DAM FAILURE STATISTICS Deleted by Amendment 25

034) 2.4A Table 1 - Maximum Possible Flood has been revised and relocated to Item 033) 2.4A Table - Probable Maximum Flood

36° NORTH CAROLJNA

                             /\f)ri2002
      ~    :,:c.ues ll    TVANucloorPo,,,ocPlane,

_;:fl TVAOulofSor,lco~cloorPlants 34°

      "                                              340 1z AMENDMENT 25 FIGURE 1

Seasonal Operating Curve, Charokaa iOBO - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

            - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --TOP OF GATES: EL. 1075.0__

1070 iij' en

1060 *""., "* ~LOODGUIDE Cl

...~ MEDIAN * (I) IE 1050 w i5Ill <( I-

           ......                          --SPILLWAY CREST: El. 1043.0 w 1040 w

u. ~ <( 1030 ~ w 1020 1010 .__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___, JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC AMENDMENT 25 FIGURE 2

Seasonal Operating CUMI, Guntllnvllle 597

                             =NORMAL OPERATING ZONE

?J 596 o, ~~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --TOP OF GATES: EL. 595.44 __ _ ..J i 595

~

c( TOP OF NORMAL

~  594 1--'0""P"'ER""A""T""IN-"G""Z"'O-'N"'E_ __,

IL. MEDIAN ELEVATION BOTTOM OF NORMAL OPERATING ZONE

          *--SPILLWAY CREST: EL. 555.0 592 JAN            FEB         MAR  APR    MAY     JUN      JUL    AUG        SEP OCT   NOV        DEC AMENDMENT 25 FIGURE J

Browns Ferry Nuclear Plant nyum,ogic; 1v10ue1 Uni/Areas KY TN SC GA A N AL W A 25 &> Tennessee River Basins

                                   --*M-=ile=,==    M  Reservoirs AMtNUMtNI  25 FIGURE 4

Unit Hydrographs, Al'NS 1-5 50,000 45,000 40,000 35,000 f/l L1. ~ 30,000 Lil a! 25,000

c 0 20,000 f/l a

15,000 10,000 5,000 0 0 12 24 36 48 60 72 TIME-HOURS

            --AREA 1 - FRENCH BROAD RIVER AT ASHEVILLE: 944.4 SQ. Ml.; 6-HOUR DURATION
            - - - AREA2- FRENCH BROAD RIVER, NEI/\IPORT TO ASHEVILLE; 913.1 SQ. Ml. 6-HOUR DURATION
            ** ** **AREA3- PIGEON RIVER AT NEWPORT; 667 1 SQ. Ml.: 6-HOUR DURATION
            - * -

• AREA4- NOLICHUCKY RIVER AT EMBREEVILLE: 804.B SQ. Ml.; 4-HOUR DURATION

            - * * - AREAS- NOLICHUCKY LOCAL: 378.7 SQ. Ml.; 6-HOUR DURATION AMENDMENT 25 FIGURE 5 SHEET 1 CF 14

Unit Hydrographs, Areas 6-9 50,000 * ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ 45,000 40,000 35,000 "'uIL 30,000 w (.') 0:: 25,000 <t ~ 20,000 c 15,000 10,000 5,000 0 0 12 24 36 48 60 TIME-HOURS

                     --.A.RE,4_ 6, JOUGLAS [,A.~ LOG.A!..., 835 Q SQ. M!, 6-HOlJR DUR.A.T!ON
                     - - - AREA. 7 - LITTLE PIGEON RIVER AT SEVIERVILLE: 352 1 SC. Ml .. 4-HOUR DURATION
                     - * -

• AREA 8- FRENCH BROAD LOCAL: 206 5 SQ. Ml.; 6-HOUR DUP.ATiON

                     * * * * **AREA 9

• SOUTH HOLSTON DAM; 703.2 SQ. Ml.; 6-HOUR DURATION AMENDMENT 25 FI CURE 5

Unit Hydrograph1, Arna 10-13 45.000 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

                                                                                    ----- __::*:: _-.:.::= :::.::..:::

12 24 43 TIME* HOURS

         --AREA 1C** WATAUGA DAM; 468 250, Ml.: 4-HOUR DURATION
         - - - AREA 11 - BOONE lOC,"L 667.7 SQ. M1

• S*HOJR JURATION

                 ../,_.REA 12- roRT PA"'R!CK HENRY ')AJV; 62 8 SC ~-~1: 6-f-'OUR DURATION
         - * -

• AREA 13* NOWH FORK HOLSTO~.J RIVER NEAR GATE CITT: if--8.S SQ ML, 6-HOUR DURATIO~J AMENDvlENT 25 FIGURE 5

Unit Hydrographa, Areas 14-18 30,000 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ 25,000 20,000 15,000 10,000 5,000 12 24 36 48 60 TIME-HOURS

      --AKl:A~ 14 & 1~ - CH<;:KOKbt. LOCAJ..; ;;ei4 6 ::iU Ml. 6-HQUR tJUKAl ION
      - - - AREA 18-HOLSTON R!VER LOCAL CHEROKEE DAM TO KNOXVILLE GAUGE" 319.6S0 Ml &-HOUR DURATION
        *- * *AREA.17 -LITTLE Rr.fER. :mie SQ Ml, 4-HOUR DURATION
      - - -

• AREA. 16- FORT LOUDOUN LOCAL: 323 4 SO. f,/1., 6-HOUR D*JRATION AMENJMENT 25 FIGURE 5

Unit Hydrographs, Areas 19-22 30.000 25,000 fl) LI. 20,000 u w (!) a:: 15,000 <(

i::

u ~ 10,000 C 5.000 48 60 TIME- HOURS

           --AREA E

• LITTLE 7 ENNE.SSE'c R!\IER AT NEEDMGR'c, 43'3.f, SQ. Ml, 13-Cf()ljR DURAT'G*!,

           - - - AREA2v

• HAf,ffAHl,c.ADAfv1, 90.9 SQ. Ml.: 2-HOUR DURATION

           ** ** **AREA21 - TUCK.A.SEGEE RIVER AT BRYSON CITY: 653.8 SQ. Ml. 6-HOUR DURATION
           - * -

• AREA 22. FOMTANA LOCAL: 389 3 SQ Ml.: 4-HOUR DURATION AMENDMENT 25 F !GURE 5 SHEET 5 OF 14

Unit Hydrographs, Areas 23-27 50,000 45,000 -----.,,-;._-----*------------------------------------------------------------------**

                        / \

40,000 '

                      '    'I I

35,000 I (/) I.I.

                                \,

t) 30,000 w ~ 25,000 <t

i:

t) 20,000 (/) 0 15,000 10,000 5,000 12 24 36 48 60 72 84 96

           --ARE.A 23

• LITTLE TENNESSEE RIVER LOCAL, FONTANA TO CHILHOWEE DAM; 404,7 SQ ML: 6-HOUR DURATION

           - - - ARFA ?4. LITTt F TENNESSEE RIVER I.0CAL, CHH.HOWFE TO TELLICO DAM, 650? SQ Ml, &HOUR DURATION
           * * * * .. AREA 25

• WATTS BAR LOCAL ABOVE CLINCH RIVc'R, 295,3 SQ, Ml , 6-HOUR DURATION

           - * -

• AREA 26
• CLINCH RIVER AT NORRIS DAM; 2,912.B SQ, Ml, 6-r.OUR DURATION

           -* * - AREA27

• MELTON HILL LOCAL: 431.SSO, ML: 6-HOUR DURATION AMENDMENT 25 FI CURE 5

Unit Hydrographs, Areas 33, 34, 36 10,000 5,000 0 12 24 36 48 60 TIME* HOURS

        --AREA 33 ~ cu~~CH R!VER :_ocAL ABOVE Ml;__E 16; 37.2 SC: Ml; 2-HOUR DLRAT!ON
        - - - AREA 34. POPLAR CREEK AT MOUTH; 135.2 SO. Ml.; 2-HOUR DURATION
        * * * ***A?-EA 36-CLlt.CH RiVER i.OCAL, MOU",H TO MiLE %, 29.3SQ Ml, 2-HOUR DURATION AMENDMENT 25 FIGURE 5 SHEET 7 OF 14

Unit Hydrographs, Areas 35, 37 45,000 40,000 35,000 30,000 C/) u.. t) w 25,IJOO c:, 5~ 20,000 C/) i:i 15,000 10,000 5,000 12 24 36 48 60 TIME-HOURS

           --AREi, 35 - FMORV R,VER AT M(lci,H_ 868 6 SQ Ml, 4-HOUR DUR,~TION
           ......... AREA. 37 - WATTS BAR LOCAL BELOW CLINCH RNER: 408.4 SQ. f,k.: 6-~0UR DURA.T!CN AMENDMENT 25 FIGURE 5 SHEET 8 OF 14

Unit Hydrographs, Areas 38, 39, 41, 42 50,000 45,000 40,000 35,001) tn u. {,) 30,000 UJ 0 0:: 25,000 <{

i:

~ 20,000 i5 1&,0D0 10,000 S,000 0 0 12 24 36 TIME - HOURS

             --<,REA~$* CHATUGe. DAM_ 189' $0 Ml. 1-HOUR DURAT!ON
             - - - AREA :,9

• NOTTELY DAM. 214.3 SQ Ml.: 1-HOUR DCRA,:ot,

             ******AREA A 1

• APf\LACHIA LOCAL. 49.S SQ. Ml.. 1-HOUR DURATION

             - * -

• AREA 42 - BLUE RIDGE DAM: 231.6 SQ Ml.: 2-HOUR DURATION AMENDMENT 25 FIGURE 5 SHEET 9 a, 14

Unit Hydrographs, Areas 40, 43, <<A, 44B 30,000 25,000 ti) 20,000 LI. u w (!) a:: 15,000 <l:

i:

u ti) i:i 10,000 6,000 72 84 96 TIME-HOURS

          - - A H c;:_,; 40- HIW.A.::':~'..:.:3. PIV[R L(\Cf,:.., ~NS.1 SQ. Ml , 0---1 ~()Uf:: D=~=RP.. fl("lt*J
          - - = ARZA43-C:CG((.f-~O 1 U-:X:::AL:                         VI 0-*1C=L*R;)lJRl.x..T:Ot~
                -ARCA44A H\WA.S:-~[-E-: H1VER fHc)M CrlAHU-:sTOf*J l'(;APAlAGHIANJDOL )f:J. tJO 1 6tb6 SQ_ Ml. &--HOlff{ DURA:!C)t~

• N<:Et-Ji48 - ;~t\~~As:.:.:::::: Rll.'ER FROM ~.1ou*:H :o 0,1,-:-,Ru.::s: O~J 385 ;,J .SU ,.11 ; t- :-!Cl:R 01.;f{,*:. fl:Jf"~

AMENDMENT 25 FIGURE 5 SHEET 10 IF 14

Unit Hyclrographs, Anlas 46, 46, 47.A. 478 40,000 35,000

• 30,000 Cl) ll.. 25,000 0

w Cl a:: 20,000

c 0

C/"J iS,000 ci I I 10,000 I [,/ 5,000 liME-1-iOl!RS

              - - ~       AREA 46

• SOUTH CHICKAMAUGA CREEK NEAR CHATT.~NOOGA; 428.1 SQ Ml : HOUR DURATION

              * * * * *

• ARE4. 47 A~ N!CK.4.JACf( LOCl*L 545.? SQ. ML 6~HOUR DURATION

              - * -

• AREt\ 478 - NORTH CHICKAMAUGA. CREEK: 98 3 SQ Ml ; 4-HOUR DURATION AMENDMENT 25 FIGURE 5 SHEET 11 OF 14

Unit Hydrographs, Areas 48-511 30,000 25,000 20,000 VI u. u w (!) 0:: 15,000 <(

i:

u V, a 10,000 5,000 0 12 24 36 48 60 72 84 96 TIMS: . wn111~i::

              --A.REA 48. SEQUATCHIE RIVER AT \\HTWLL; 400.0 SQ. Ml., 4-HQUR DURATION
              - - - AREA 49- GUNfERSVILLE NORfH LOGAL: 1.0441 SQ M*: 6-HOUR OURAT*ON
              ******A.REA 50

• GUNTERSVILLE SOUTH LC>CAL; 1,154 9 SO. M,.; 6-HOUR DURATION AMENDMENT 25 FIGURE 5 SHEET 12 OF 14

Unit Hydrographs, Areas 51-52, 54-56, 58, 63-65 50,000 - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 45,000 40,000 --- 35,000 ti 30,000 *-- ~ 25,000 u l5 20,000 15,000 --- 10,000 12 24 36 48 60 72 84 96 TIME-HOURS

                         **** *

• AREA 51 - PAINT ROCK CREEK NEAR WOODVILLE; 321 SQ. Ml.; 6-HR DURATION

                         -

• AREA 52 - PAINT ROCK LOCAL; 138 SQ. Ml.; 6-HR DURATION

                              * *AREA 54- FLINT RIVER LOCAL; 225 SQ. Ml.; 6-HR DURATION
                         --- * --- AREA 55 - COTACO CREEK AT FLORETTE; 136 SQ. Ml.; 6-HR DURATION
                         -

• AREA 56- COTACO CREEK LOCAL; 101 SQ. Ml.; 6-HR DURATION

                         -    -    AREA 58- LiiviESTOi\iE CREEK LOCAL; 157 SQ. ivii.; 6-HR DURATiOi\i
                         - - - AREA 63 - SUGAR CREEK AT MOUTH; 177 SQ. Ml.; 4-HR DURATION
                         -        *AREA 64- ELK RIVER LOCAL, MILE 16.5 TO PROSPECT GAGE; 145 SQ Ml.; 6-HR DURATION AMENDMENT 25 FIGURE '.J SHEET 13 OF 14

Unit Hydrographs, Areas 53, 57, 59-62 20,000 18,000 16,000 14,000 1f 12,000 ~ 10,000 ~ !2lC 8,000 6,000 4,000 2,000 12 24 36 48 60 72 84 96 TIME- HOURS

              -

• AREA 53- FLINT RIVER NEAR CHASE; 343 SQ. Ml.; 6-HR DURATION

              --AREA 59-TIMS FORD DAM; 533 SQ. Ml.; 6-HR DURATION
              -

• AREA 60 - ELK RIVER LOCAL, TIMS FORD TO FAYETTEVILLE; 293 SQ. Ml.; 6-HR DURATION

              -  -   AREA 61- ELI< RIVER LOCAL, FAYETTEVILLE TO PROSPECT; 490 SQ Ml.; 6-HR DURATION
              - - - AREA 62- RICHLAND CREEK AT MOUTH; 488 SQ. Ml.; 6-HR DURATION AMENDMENT 25 FIGURE 5 SHEET 14 OF 14

585

                                                                                                                                                                --observed TN River atGuntersvilleTW TRM3490 580  +-----t--+--+---"::!;;-.e::::;=:,,.,_;;::--_+--4--+---+--+-------l-. :~~~~:~~

TRM3490

                                                                                                                                                                --observedTNRrver atWhrtesburg TRM3340 575                                                                                               : i

• I 1

_,___ --- -"- - . '.-1' i t - * ~~~~~~t~~~r atWhrtesburg 1 I  ! *-....: i  ; TRM3340 I * ' I

                                                 -~- ....

570 ......................................

                                                                                                   * ** :** \::********r                      ....... : ....... --observedTNRrver u.

I:*, atDecaturTRM i 3050 .!: i  !*  : C i . I j -*Computadat120 seconds-TNRwer iii 565 I

                                                                           'i I

I atDecaturTRM 3050

                                            .I
                                                                    ~  - -i- -- C
                                                                                          --1                                                                   --observedTNRrver at Browns Ferry I                 i                                                                       TRM29355 560                                                                                                                                                                Comp*fadat12C seconds-TN River fRBM~~\~erry
                                                                                                                                                                --observedTNRrver atWheelerDam 555                                                                                                                                                                TRM2749
                                                                                                                                       ~==:=:::==!- * ~c-0nds--T"!s!tve*

Computedat120 Wheeler Dam TRM 550 +--.....- -......-------...... 3/14/73 3/15/73 3/16/73 3/17/73 3/18/73

                                                                   ----~--+-----------.. . .---.-~. .

3/19/73 3/20/73 3/21/73 3/22/73 3/23/73 3/24/73 3/25/73 3/26/73 3/27/73 2749 Date Hydrologic Model Verification - 1973 Flood AMENDMENT 25 FIGURE 6

450,000 400,000

                                                                                                                       --Observed TN 350,000                                                                                                                   River at WlleelerQs TRM274.9 300,000
                                                                                                                       -

• Computed at 250,000 120 seconds-TN Rivera!

rn Wl1eelerQs LL t) TRM274 9 ,5 200,000

~

ii: 150,000 --Observed TN River at Guntersville TINQsTRM 349.0 100,000 50,000 I I i,, i

                          - ~-,1 1\.
                                  -l-                                                       I-             -!

0 3114173 3115173 3116173 3117173 3118173 3119173 3120173 3121173 3122173 3123173 3124173 3125173 3126173 3127173 Date Hvdralaaic Madel Verification - 1973 Flaad AMENDMENT 25 FIGURE 7

585 580 +-~-+--'-+-++--+--+-+-++--+---+-~+-++4-l-h-t+--+-~-++-- * ~i;;puled at seconds-TN

                      --;-                                                                                                                  River at Guntersville TWTRM a  --
                   - -  ,-  ~

349.0

                                                                                                                                     --Observed TN 575                                                                                                                                      River at
                                        ~~

rv Whitesburg TRM334.0 ~ IL 570 11.~ ~1 :i -

• Computedat 120 seconds-TN lf:

                        'V                                                                                                                  River at

.!: Whitesburg C:  ! :I TRM334.0

>           :__                                                                       i                                              --Observed TN R1verat DecaturTRM iii 565
            -=       //
                  ;J Iii 305.0
                                                                                                                                     --Observed TN River at
                      .,II                         I'--                               !                        **-**-                       Wheeler
             ~ _:,;,-1 i                          I~-                              11                                            _          DamTRM 555 -l-?9s44-3/4-4-..;-,,~+/--::::::sh_,,,eo-f-++++++-F~~~"'-:-++'f""~

-- - I I'-. "\.*. I 274.9 I ----~i::.,:.:....i,. -;.~* ~ *,*- ,~--: . .__, . ~~

                                                                                                                                     -

• Computedat 120

                                                                    ~....*+-r.,..,...-i--,1...................
                                                                                                      ~

seconds-TN

                        ;I                                                                                     -~ 'r-.,,-.., ,.,            River Wheeler 550 +---+--.......~~.....--+-~.........- -.....--+---1--~-+-~......-~1----+----+~--1                                                     DamTRM 274.9 12/4/04 12/6/04 12/8/04 12/10104 12/12/04 12/14/0412/16/04 12/18/04 12/20/04 12/22/04 12/24/04 12/26/04 12/28/04 12/30/04 1/1/LJ=,

Date Hydrologic Model Verification - 2004 Flood AMENDMENT 25 FIGURE 8

4000000

           ,--,-I,I--,--:                 i-,-,-1!---__ :...,..,...II~-,--;_-,--,-I!-,--,.1---,--i-J,-,--f+/-,0e~-,--,--,~----,-P**1..,,........,

_;-,--,--11---,--,---: ,-,--f **.-1

                                                                                                          .           cmtlpul llfFlo.ws
                                                                                                                           '        i 350,000

_, rL.1 I

                     '                 '                  , I
                                       '                                                                                                            --Observed TN R1verat WieelerQs 300,000
                                          ~ti                                                                                                             TRM 274.9 J~' tt 250,000
                              . J :i t            rii

~ (.) ,5 200,000

                        \:
                                'T
                                      ~,  ~

117 i P+--'-+--d--l--ll""IH----1-1-.t--+-'+-+H~':,._--'-,_

                                                          'I Ii       fJ
                                                                       ._     ---+-'-1--'----f----+-++---,--'+---+i--j+--l 1_ _

1 _ j_ - * ~;;puled at seconds-TN River at

~                    '

1 I

                                                                                                .!         i WieelerQs TRM 274.9
                                                                                               ~+

ii: I 150,000

                                                                            '    .;  J,
                                                                        /~f~t.iP,~~)~*~:;t! ,' ! : .I                          I_

h

                                                                        ':i.r ~ ~Tl, 11-l. :j ,I 1

00,000 t':1~---,11~1-tiT-~1:1-_l-lt"::;::#:l::;::j:;J'.'i-1,ttl~tij. i,;11 u-I -f - - i,i! Htitltti.11.lltJMil~-;'lfii~ --Obse,ved TN River al

                                                                                                                                      ~                   Guntersville
                                                                                                   -':'--Tt-:-----lttlffi;""c---t-fflrfttt-lr+--11        ;:_~s   TRM 50,000 ,1-----1--~-+->---+--+-+--+-c---+-+---+---l-h---+---'-+--H--l-'-,_.~....I-Llf---W J;

O ,.__.___,_-*--,---'-l-----'-!----'-1----'-1----1--'----l--'---I---'---!----!- i-___._....-i-.....,' ....---,...'---! I 12/4/04 12/6104 12/8/04 12/10/0412/12/0412/14/0412/16/0412/18/0412/20/0412/22/0412/24/0412/26/0412/28/0412/30/04 1/1/05 Date Uyl'lrnlngir AAnllol 1/orjfirnf'inn - 7nl1A. f:lnnrl AMtNUMtNI 25 FIGURE 9

620

                                                                                      -Model-1 1200K Profile
              ;                                  i 610   --*-                  '
                                                                                      -  -Model-2 I        _,                                                                1200K
       '      c : ' ---                                                                    Profile 600     '                   '

i

                                                                                      -Model-1 1300K Profile
                                                                                      -  -Model-2 1300K Profile 570 560 270    280           290        300       310         320        330       340 350 Tennessee River Mile Hydroiogic Modei Verification - Steady-State Profiies AMENDMENT 25 FIGURE 10

                                                                         +-d . . . . ~+

620 ..--.-,-.....,...,...,..,...,...,.-.,..,.._.,._...,...,_.,......,..-.,...,..,....,..,...,..,....,.......,.........,..,...,.....,..,...,..,-,---,..,...,..........

       ~- --- . . j -- *j-;- -+-----                        .. i---L.                                             i-    l L .. ... - -,              1- +' l-1+ ++ i~ -- -                           +-1-,n1 -- -                    --I-         t \-       h            ,__         ---i-~   -H+/-

I I I +t i I -- / I I I I..L."'T,  ! 610 ttttttttittttl'~:++l_.,..r-~_r-!+-i1:*~~~~1ti1ti1ti~~*~1!.~~~"1~,~'tl1*I I I

I i ,

I

i I~

                                                                                                                                    -             I i I

I I t+-t--t-H++-1h+t--~,1H*+-t---H1H-i:--t-11H-il1'17 +ttiH+H---'-t,-c_...--:.1.,.,,14-1--H+HHl-,tl--:-c-1,--i--i1 -~~~ter 1--+-+-+-1--+-+-+-' 1

                                !        I I             I          _                                      ""!" i       ! '                       I ,           I-      (Mode~1)

I eoo I I I I , I j ! c.. I i !I I ! I I I

                           ' ..J..+                                                                          i l--+-+-r--t--1--t-c-l I

Ii! I I I

                                                            +--r 11 I :
                                                                                                             ' ' ' I
                                                                                                                                             ,       ))II I I I i i I
                           !    I        II                        c,l'io'I"'" 1                             1    i ;      1                      1 1 Ill 590 ++++++++I++++!t+i++I++t+:J,,S,,1,'F-,-,r++'++',++++++++++++H--++-++++H-H-i1                               1             1                           H-' i t--t-li
                           ' I I , I I                      ""' I , I I                                                                           I I I I !             Tallwater c-\- W i            I_ 7 ~1!"" ,...!..~

• J_j_ 1-- i -+ -tt ,_ _J__...µ_ Profile 580

                         +W+~~l~+tR-h1-lt-f -t+H~-Ht+ ~Ltttt+~L+t+n~ ,.
    ~t+-+--+-tlt --t--t--,-t~                                                                                                                                             ,ooe,<J LI llH 11111.M!I _ [J_ J111J. i I1Il_HfEHHE+/-f-H+/-f+/-fH+/-HJY 1                       1     1 1ti Hti:~i.E ~=tt+i-=;. hb.t~~ ::dJtHJ+/-tit+/-r+/-t+/-lU:UH::r.tH~t~1 1

U+~l l i 1111111: I ~W+Wl l l Ill l Ill i I!! 11111111111: 11 ~~1:.~ 510 rttiIDiTfnn~Ir:=n IIETIIT!IIEIJnrr1t~1+HTrE_d~in=1 DRC) I U'l 111111111 : 1I I I i 11 i : I i 111111 i 111 i i I : 111111111 111 i 11 560 i i i n 11111 111 i 11 1 111 1i 1 1 1111111111111 111: 1111 111 111 1 1 1 55ol1lllll]lli l1 it: ,!I :111111111111 1:j))llll1ll 1: :1 1 1 1 1 1 1 1 0 100 200 300 400 500 600 700 600 900 1000 1100 1200 1300 1400 Discharge (1000-cfs) Hydrologlc Model Verification - Guntersvllle Tai/water Curve AMENDMENT 25 FIGURE 11

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 2.4A - Figure 12

Probable Maximum Prec:lpllallon leohyals for 21,-0 lq, Ill, Event, DGwn.traam Pia-nt AMENDMENT 25 FiGURE 13

Probable Maximum Precipitation lsohyets for 16,170 Square Mile Event AMENDMENT 25 FIGURE 14

r------------:---,,n,n1TTT.P'!li~rr-r;:,========:::,;-i 1,200,000 21,400 Sq. Ml. Downstream Centered 1 1

                                                                                                                       ! !;     i"'-         Peak Discharge: 1,193,671 els March Flood Event for BFN                           ' ! ~~~                        i  IJ       ,' ] 1'.

Browns Ferry Nuclear Plant -t-rtt,_,_,-,-_,____,___.,., Note: Design basis PMF discharge TRM 294.0 1 ---,--, i

                                                                                                                  ~l            ! !        , Is maintained as 1,200,000 ds.

I I I -~ ,- - ;-~ / : . ~n~r-r~-i----,-.,....,..i-,----,-,---,-.'1 1---1--H----1-+-++-+-+-+-+-+-H-+--H_,_+---'--H----1-+-+-H~-+-+-+-,--+--Hl-+-+-+-+-+-+-1-.....+-+-+-+-+l--1 L .L _. .,.J_Ll _.. LL_HII/*-+,+*-**+ 1+~ _' , 1,000,000 _i ' ' L-; L _f--,_ ' I I i I J i I J--+-+-+-+-+--'!.-+-+-++-++-+l-+-1--I+-+-+- ' ' ,/ 1 S00,000 -1---1--<.--'--'-'-'~l~~-'-'-~l-+-c...,.l-1-+-"-<f-'--+-'-l-+~~'-+-+-+..;-+-+-1--+--1-+l-t-,H-++'4--i1H-1/4--J---i--"--l-l--+-+-l

                              -Browns Ferry Plant Dlscha'l!*                                  ,_. I     I_                     I                     \ *           :i s              l--1-----'-"=-T-"}..--rt~'

1

                                          -,-+=1=1=+=1=;::.-,..,......,,t- - '       '          j        I            -+-1++1-1-++-l-f-+-+-+--\.+-++-+--+-t+-H-1

• -------* 1t + f T~ -+-

  .. ,--r+** r*-

1

                                                                                                                                                           ,*--H+---

i 600,000 I I I /! I i', 0 *t-*- -

              >-+---~t-+--i-+-+r+-,-f---}-f---1-:          +l+H-
                                                     -~ (l -r
                                                                                        -~+-f--
                                                                          *t-+1,--t-:-,-+-
                                                                                                         ~-,-~---1--1---

Flr T'-rTrr i * *

                                                                                                                          ,-+-t--~--'--+++-:f--t-:--:_7:-r-1/4+--~--H-,-+-+-~
                                                                                                                                                               ~-1\T - -*-
                                                                                                                                                                             -++-l_

40();000 1---1--H--!-.J-J. i+-+-+-+-+-+-+-H-~-+--H-+-+-1-+-+-+-+----+-+-1-+-!-+-H-+--+-+.....,_+-+--'--¥-+-+-+-+-, 1J II I I'\ i!  : ,~ ,, u'I 1,>-+-+-+-+-< 1, ... 200,000 i ~ ~ ~*- - i I- +-. + 1

                                                                                                    -+_*-+-j-t-f--+t-7--o-t-t--+   :+
             -I-I---I--H----l-+.ol!-+-+-+-+-+-+-+--+--H-+--+-+---l-+-l--l-+-++----++-+-+-+,---1-+-+---l-+-+-+-+-++-f->-+-,-+-+"1od I +i-r:-I\~ -
                              ~

Vi i !

                                                  ',.!I-f-l-II I'
                                                                                     ,                   I
                                                                                                                       '---f-1 *,     I'                i -H--H,-

Ii ; i

                                                                                                                                                                          ~

J--+-~1-+-i~!-+ I I i I I I I I I  ! I 3/15 3/17 3/19 3/21 3/23 3/25 3/27 3/29 3/31 4/2 4/4 4/6 4/8 4/10 AMENDMENT 25 FIGURE i5

5so,---------------,-,-,-.....,....,...--.--,--..,..,....,....,...,..............'l""T".,...,.-r-l""l'""T""T-.-,--:-r.. 1 21,400 Sq. Mi. Downstream Centered -**-,--- __ **-<--*,,--**o--,-**~,-+-+--r---t-!-+-i--+- _j._; __ I I _ 0 _ f[ :tF-~ :r= =-I~':~ March Flood Event for BFN i i 'j '

~~:ar

                                                                                       '                         !                   I !

Browns ~':J Plant 1--+--"-t-+-+--,-+-+-*-+-t-+-*-,--,-; i  ! 575 +-+~-+--,+-+  ;-+-<!-+-:+-+.......,->-o--++-+-+-t-+----+-+-<--+-,+-+-+-<t-+-+-J ,,.,J \ . AMENDMENT 24 BROWNS FERRY FINAL SAFETY NUCLEAR PLANT ANALYSIS REPORT

FIGORe NO. 2 2B 35

                                                          ~

0 30

                       !              ~

v--- ------ 25 w

.c:

C z 20

                            /;                                   t
     ..J             /
                       /

\_, ...I 15 LI.

   -<z     10      I a:

5 I 0 I 0 2 4 6 TIME - HOURS PROBABLE MAXIMUM PRECIPtTATION POINT RAINFALL AMENDMENT 16

tlGURE 23 1000--,.--- --1--- ------****** - - - - -* _ ... --* _ soo,-;--;---t------4-----+-----l t2 u 8 g ~ eoo- '-**- -- ~ - - - - *- - . ____ ,__ . ------ ... a: ~ (J (/) Q 4 00 t---t----1- - - + - - **-*------** * - 200i---1,--t-+------l-----l----_j 0) 0 2 4 8 TIME - HOURS BROWNS FERRY NUCLEAR PLANT ONE HOUR UNIT HYDROGRAPH FOR UNNAMED STREAM NORTH-WEST OF PLANT AMENDMENT 16

FIGURE 24 Amendment 16

o.--...,.....--l--~1--.-------.-------.---,---1--~,---

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SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 2.4A Figure 26, Sheet 1

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 2.4A Figure 26, Sheet 2

BFN-26 2.5 GEOLOGY AND SEISMOLOGY 2.5.1 General The regional geologic features in the Browns Ferry site area and the local geologic formations in the immediate plant area have been investigated. The results of extensive drilling, excavation, and testing are presented in this subsection. These results show that the underlying bedrock will provide more than adequate foundation for Browns Ferry plant structures. The seismicity of the site area has also been studied. Data from many earthquakes have been used to compile the seismic history of the area and to evaluate the earthquake hazard at the site. 2.5.2 Geology 2.5.2.1 Introduction The Browns Ferry area was explored first in 1962 as one of several possible sites in the vicinity for a fossil fuel plant. The area is underlain by flat-lying, undeformed limestone of Mississippian age. Exploration indicated and subsequent excavation proved that no significant geologic problems would be encountered in developing satisfactory foundations for the major structures. 2.5.2.2 Geological Investigative Program 2.5.2.2.1 Site Initial geologic investigations were made at the Browns Ferry site from January to May, 1962, and again during February and March, 1963. During these periods 80 holes totaling 5658.6 linear feet were drilled on 200-foot spacing along lettered ranges roughly parallel to the shoreline of Wheeler Lake and numbered sections at right angles to the ranges as shown in Figure 2.5-1. At this time the site was under consideration for a fossil fuel plant. For this reason drilling depth was limited to 10 feet of sound, unweathered rock and, therefore, the majority of the holes penetrated less than 20 feet of rock. The geologic conditions revealed by this drilling are shown on Figures 2.5-2 and 2.5-3. 2.5.2.2.2 Plant Foundations With the decision in 1966 to utilize the site for a nuclear plant, additional exploration was done. As shown on Figure 2.5-4, 29 additional core drill holes and 95 percussion holes were drilled in the main plant area to provide additional geologic information. Graphic logs of the additional core drill holes are shown on 2.5-1

BFN-26 Figure 2.5-5. A summary of geological investigative programs from 1972 to 1980 is presented on Figure 2.5-5a. 2.5.2.2.3 Access Highway Bridge In April of 1972 foundation investigations were conducted for the construction of an access highway bridge over a relocated channel. Ten Nx-wireline core holes were drilled and visually logged. The drill layout is presented on Figure 2.5-5b and the logs are on Figures 2.5-5c through 2.5-5l. 2.5.2.2.4 Deleted 2.5.2.2.5 Low-Level Radwaste Storage Facility During the period from January to April 1980, TVA conducted foundation investigations at the site of the proposed Low Level Radwaste Facility (LLRW) (Figures 2.5-S1 through 2.5-S4). Core borings and various geophysical methods were used to determine the depth to rock, the configuration of the top of rock, and to locate and zone deficiencies in order to assess the ability of the bedrock to support the overlying soils and thus the LLRW facility. Three vertical Nx-wireline core holes were drilled through the near-horizontally bedded limestones of the Mississippian aged Fort Payne Formation into the undifferentiated shales, sandstones, and limestones below (Figure 2.5-S5). The cores from these holes were logged visually, and the holes were logged geophysically. The bedrock was found to lie an average of 50 feet below the ground surface and to have an undulating to slightly undulating top-of-rock surface containing narrow near-vertical solution features. Subsurface geophysical investigations that were carried out in 102 percussion holes drilled 50 feet into the rock, revealed no buried sinkholes or large near-surface cavities which could potentially collapse and cause settlement. 2.5.2.3 Regional Geology 2.5.2.3.1 Geological History The Browns Ferry area lies on the southeastern flank of the Nashville structural dome where it merges into the foreland slope of the Appalachian geosyncline. Throughout most of the Paleozoic Era the region was at or slightly below sea level. During this time more than 5,000 feet of limestone, dolomite, and shale were deposited. Since the end of the Paleozoic Era, some 250,000,000 years ago, the area has been above sea level and has been subjected to numerous cycles of erosion resulting in a general peneplanation. During its history this immediate region has been one of little structural deformation. Major folds and faults are entirely absent. The rock strata are only slightly warped with regional dips of less than 1 degree to the southeast away from the Nashville dome and toward the foreslope of the Appalachian geosyncline. 2.5-2

BFN-26 2.5.2.3.2 Regional Stratigraphy (References 1 and 2) The low plateau on which the Browns Ferry site lies is underlain by near-horizontal limestone strata of Mississippian age having an aggregate thickness of slightly over 1,000 feet. In ascending order the formations and their maximum thicknesses, according to the Alabama Geological Survey, are: Fort Payne, 207 feet; Tuscumbia, 200 feet; Ste. Genevieve, 43 feet; Bethel, 40 feet; Gasper, 160 feet; Cypress, 7 feet; Golconda, 70 feet; Hartselle, 200 feet; and Bangor, 90 feet. Bedrock is mantled by varying thicknesses of cherty clay, silt, sand, and gravel of residual and alluvial origin. The only formations involved directly in the site area are the unconsolidated materials overlying bedrock and the Tuscumbia limestone and the Fort Payne Formation. A brief description of each of these follows. Unconsolidated Deposits - Within the site area bedrock is mantled by an average thickness of 54 feet of red and yellow clay containing some residual chert boulders and lenses of sand and gravel. This material varies in thickness from a known minimum of 41 feet to a known maximum of 69 feet. Tuscumbia Limestone - Only the lower 50 feet of the Tuscumbia formation was encountered at the Browns Ferry site. The Tuscumbia is characterized by medium-to-thick beds of light-gray, medium-to-coarse-crystalline, fossiliferous limestone. Inasmuch as the Tuscumbia Limestone is a relatively pure limestone, it is more affected by solution (than the Fort Payne Formation). Practically all the cavities encountered at the site were developed in this formation. Fort Payne Formation - The maximum known thickness of the Fort Payne formation in northern Alabama is slightly over 200 feet. At the Browns Ferry site the total thickness, penetrated in one drill hole, is 145 feet. The formation consists of medium-bedded, medium to dark gray, silty dolomite and siliceous limestone with a few thin horizons of shale. Near the top of the formation, some of the beds are cherty and some of the cores showed zones which were slightly asphaltic. The most distinguishing lithologic feature is the presence of quartz-and calcite-filled vugs up to 1 inch in diameter. The silty, siliceous nature of the Fort Payne formation inhibits the development of solution cavities and very few were found in cores drilled from this formation. In general, excavation grades for the major structures of the Browns Ferry plant were set in the Fort Payne formation. 2.5.2.3.3 Regional Structure The regional structure in the Browns Ferry area is controlled by the Nashville dome. The area lies on the southeast flank of this dome and the regional dip is a degree or less to the southeast. This regional trend is commonly obscured by the slight local 2.5-3

BFN-26 variations in dip caused by minor warping and folding. In general, axes of these secondary flexures trend northwest-southeast which is compatible with a regional "cross grain" that has developed on the flank of the Nashville dome. In the immediate site area the beds of the Tuscumbia and Fort Payne formations are essentially horizontal. Calculations based on the elevations at which the contact between the Tuscumbia and Fort Payne formations was encountered indicate that the direction of dip varies considerably but has an overall westward major component. As is to be expected in near-horizontal strata, bedrock is cut by a pattern of near-vertical joints. Close to the surface of bedrock, solution channels have developed along these joints especially in the Tuscumbia Limestone. At depth, however, in the less soluble Fort Payne, the joints are tight and most are cemented with calcite. Faulting is not a significant factor in considering the geologic structure in the Browns Ferry area. No active faults showing recent surface displacement are known within a 200-mile radius of the site. The nearest known ancient fault is in Lawrence County, Alabama, 16.5 miles to the west-southwest from the Browns Ferry site and is one of three apparently related near-vertical faults. The vertical displacement varies from 0 to 60 feet and cuts Mississippian bedrock. At the site, the only indications of any rock movement are small shears along bedding planes which represent minor readjustments between beds when the area was uplifted at the end of the Paleozoic Era some 250,000,000 years ago. No accurate measurement of these displacements can be made, but movement was probably on the order of a few hundredths of a foot. 2.5.2.4 Site Geology 2.5.2.4.1 Physiography The area surrounding the Browns Ferry site lies near the southern margin of the Highland Rim section of the Interior Low Plateaus as defined by Fenneman.1 This physiographic subdivision is characterized by a young-to-mature plateau of moderate relief. The general level of the ground rises gradually from 600 feet above sea level at the north shore of Wheeler Lake to around 800 feet above sea level 10 miles north in the vicinity of Athens, Alabama. This surface is modified by the drainage patterns of Poplar, Round Island, and Mud Creeks, which flow across it from northeast to southwest. 1Fenneman, N.M., Physiography of the Eastern United States, pp.415-427. 2.5-4

BFN-26 The plant site is located on an old river terrace surface with an average elevation of 575 feet above sea level. This surface represents an old flood plain of the Tennessee River developed when the river was flowing at a higher level. The most recent flood plain is now inundated by the waters of Wheeler Lake. Plant grade is at El. 565. 2.5.2.4.2 Bedrock The general description of the Tuscumbia and Fort Payne formations has been covered in Section 2.5.2.3.2, "Regional Stratigraphy." This section will deal with the foundation conditions as encountered and treated during the construction period. The first bedrock uncovered was the lower portion of the Tuscumbia Limestone. As had been expected from the results of the exploratory drilling, the Tuscumbia was cut by frequent near-vertical solution channels developed along steeply dipping joints. Before El. 515 had been reached, the near-vertical solution channels had pinched out and the major indications of solution were near-horizontal zones of weathering developed along the bedding. At this stage additional exploratory drilling was done to determine the detailed foundation conditions under the reactor building, turbine building, intake structure, and chimney areas. As shown on Figure 2.5-4 a combination of core drill and percussion holes was drilled. The core holes furnished detailed stratigraphic control, while the primary purpose of the closer spaced percussion holes was to ensure that no large cavities existed below the foundation. Eleven specimens of core selected from holes drilled in the reactor area were tested for unconfined compressive strength in the TVA materials laboratory (Reference 3). The maximum value obtained was 17,200 psi; the minimum was 11,419 psi; and the average was 14,175 psi (Reference 4). Additional core samples from the same area were sent to John A. Blume and Associates for testing of other physical properties. The following data are taken from their report (Reference 5): Elastic Modulus ..................................................................... 8,200,000 psi Shear Modulus ...................................................................... 2,300,000 psi Constrained Modulus.............................................................. 10,000,000 psi Poisson's Ratio ...................................................................... 0.252 Ten selected percussion holes in the Unit 1 and Unit 2 areas were inspected with a borehole television camera to confirm the results of the drilling, and the continuity of horizontal seams was checked in four core drill holes by resistivity and gamma ray logs. In addition, test pits were dug at the centers of Units 1 and 2. 2.5-5

BFN-26 The exploration and initial excavation disclosed a persistent weathered zone at the base of the Tuscumbia Limestone, between El. 509 and 508 at the west side of the Unit 1 portion of the reactor building, which contained one or more partially open seams 0.1 foot to 0.15 foot thick. In some instances this seam was open, and in others it was clay-filled. This weathered zone sloped upward to the east along the contact between the Tuscumbia and Fort Payne formations. Midway between Unit 1 and Unit 2 this contact intersected the rock surface at approximately El. 511. As a result, only the Fort Payne formation is the foundation rock in the Unit 2 and Unit 3 areas, and no cavities or seams were present. Special foundation treatment was required to ensure an adequate foundation for the structures. In the Unit 1 portion of the reactor building this was accomplished by a system of underpinning. Trenches were excavated through the Tuscumbia Limestone to the underlying Fort Payne Formation. These trenches are under the perimeter walls of the reactor building and a doughnut-shaped trench is under the periphery of the central mass under the drywell. These trenches were backfilled with concrete to El. 513, the bottom of the reactor building slab. (Figure 2.5-19) Under the Unit 2 and Unit 3 portions of the reactor building where the Fort Payne formation was below the design elevation of 513, fill concrete was placed to that elevation over the entire area. Foundation for the bearing pile cluster locations in the turbine building area was provided by grouting the seam at the Tuscumbia-Fort Payne contact by conventional consolidation grouting methods. Surficial Deposits Plant Area - A soils investigation program was initiated in the spring of 1966 for the purpose of establishing the allowable bearing value for soil-supported structures and identifying adequate borrow material (Reference 6). The original ground surface occurred at approximately 15 feet above final plant grade in the area of plant structures, and approximately 2 feet above the final transformer and switchyard area grade. The top 15-20 feet is classified as a red-to-reddish-brown, sandy clay and lean-to-medium silty clay with a maximum thickness of 30 feet. This is designated as the preferred borrow material. Underlying these alluvial terrace deposits are approximately 40 feet of residual medium-to-fat clays and plastic silts interbedded with lenses of medium layers of gravelly chert. The ground water table was established at El. 555.1 foot, which corresponds to the level of Wheeler Reservoir. 2.5-6

BFN-26 A total of 13 borings was made. Of these, undisturbed samples were obtained from seven, using 3-inch and 5-inch Shelby Tubes. The remaining six were drilled with a 4-inch power auger, with disturbed samples taken. Undisturbed samples were taken at 3-foot intervals, while disturbed samples were taken at 5-foot intervals. Standard penetration tests were made using a 2-inch split-spoon. Laboratory testing consisted of index tests, soil classification, consolidation tests, and vane shear tests. The laboratory testing and the standard penetration resistance result gave an allowable soil bearing capacity of 1.5 tons per square foot for a mat foundation and 2.0 tons per square foot for spread footings. Intake Channel - To determine the seismic stability of materials in the intake channel, borings and samplings were made in depth at five locations in the side slopes of the channel connecting the intake structure with Wheeler Reservoir. Laboratory tests and analysis then established shear and other design values of this material (see Chapter 12). A vibroseismic survey was also made at the site to identify the shear wave and compression wave velocities of the soils. Using the values obtained, the seismic stability was evaluated and is described in Section 12.2.7. Low Level Radwaste Storage Facility - A soils investigation program was initiated in the Winter of 1980 for the purpose of determining the in-situ and borrow soil properties. Figure 2.5-S1 shows the location of the LLRW facility in relation to main plant and Figure 2.5-S2 shows the location of in-situ soil borings for the LLRW facility. The overburden thickness varies from 37 to 50 feet with existing grade varying from El. 588 to El. 570. Final grade is at approximate El. 580. Figures 2.5-S3 and -S4 are generalized soil profiles in the LLRW area. A predominantly red lean clay layer extends from the ground surface to depths ranging from 2 to 18 feet and averaging 16 feet. This layer is continuous except where previously excavated for borrow. Typically, blow counts (N) from the standard penetration test (ASTM D-1586) fall within the stiff to very stiff consistency range. However, surficial weakness revealed by blow counts less than 10 is not uncommon. These clayey soils represent ancient terrace alluvium. Underlying the lean clay is an intermediate layer consisting of predominantly tan to red, medium to highly plastic residual clay. This layer is not continuous. In places, it has a thickness up to 26 feet and averages about 16 feet. Inclusions of gravelly clay, lean clay, and silt are present. Penetration resistance indicates very stiff consistency with only scattered isolated weakness. 2.5-7

BFN-26 The intermediate layer extends to bedrock and is underlain by a basal cherty clay or clayey chert. The fine fraction of these soils is usually highly plastic, showing a wide variation in color ranging from red to yellow and tan to black. The gravelly clay containing over 10 percent gravel represents a transitional zone between the fat clay and the clayey gravel as shown on the general cross sections, Figures 2.5-S3 and 2.5-S4. The residual clayey gravels sampled with the 2-inch split-spoon usually showed some particle breakage due to the penetration of the samplers. Thus, the in-place material is coarser than the samples grain size tests indicate. This gravel layer is continuous across the site and averages about 18 feet in thickness. Soil consistencies indicate wide areas of relative weakness (N 10) which is most pronounced and persistent immediately above bedrock. Water level readings were taken in borings 1 hour and 24 hours after completion of drilling. Water levels varied from El. 575 in the western area of structures A-1 and A-2 to about El. 569 on the eastern end of A and B structures and the southern end of the C structures. In some cases, water was not initially apparent at the established elevations. On penetrating the clayey gravel (GC) soils, however, water rose and adjusted rapidly, indicating the water is confined. Laboratory testing consisted of index testing of split-spoon samples for classification. Moisture contents were obtained on all samples. Undisturbed samples were tested for moisture content, grain size, Atterberg limits, specific gravity, and unit weight. In addition, representative undisturbed samples were selected for determining engineering properties. Specifically, triaxial compression unconsolidated-undrained (Q-test), triaxial compression consolidated-undrained (R-test), and direct shear consolidated-drained (S-test) strengths were determined. Consolidation and permeability tests were conducted. Three borrow areas were investigated to locate a suitable quantity of an acceptable borrow material. Each area was sampled with auger borings to collect bag samples for laboratory testing. Standard compaction curves (ASTM D-698) were developed for each borrow area, and shear strength, consolidation and permeability tests were conducted on remolded samples of each identified borrow soil class. Each borrow area was evaluated and determined to provide an adequate source for borrow material. An evaluation of the in-situ and borrow soils was made. The results were used to determine the allowable bearing capacity and predicted settlement, which were found to be less than the design criteria limits. In-situ dynamic soil properties were determined using cross hole, uphole, and downhole geophysical tests. The tests were performed at three locations. The tests 2.5-8

BFN-26 were evaluated, and the results indicated a range of shear wave velocities between approximately 500 and 2300 feet per second. 2.5.3 Seismology (References 7 and 8) 2.5.3.1 Introduction The Browns Ferry Nuclear Plant is located in an area far removed from any centers of significant seismic activity. A few major earthquakes centered at distant points, several light-to-moderate shocks at distant points, and several light-to-moderate shocks with nearer centers have affected the area at low-to-moderate intensity (Reference 9). 2.5.3.2 Seismic Investigation Program In order to evaluate the earthquake hazard at the plant site, a study was made of the known seismic history of a large surrounding area. This study was greatly facilitated by research carried on over a period of more than three decades on the seismicity of the southeastern United States in general and the Tennessee Valley region in particular. Voluminous files of earthquake data, collected from a number of sources, were used in the compilation of seismic histories of the several states. By plotting the epicenters of hundreds of these earthquakes, the areas of continuing seismic activity became apparent. The more active areas are as follows:

a. Mississippi Valley, especially the New Madrid region of Missouri, Arkansas, Tennessee and Kentucky. This area has been seismically active since the appearance of the white man and very probably long before. The area has been the center of a few great earthquakes and very numerous lighter shocks which are still occurring at intervals. The New Madrid region is about 170 miles northwest of the plant site.

b. The Lower Wabash Valley of Indiana and Illinois. This area has been the center of several moderately strong earthquakes, some of which were felt as far south as Tennessee. The Lower Wabash Valley is about 225 miles north-northwest from the plant site.

c. Charleston area, South Carolina. One of the country's greatest earthquakes was centered in the Charleston area. Many other light-to-moderate earthquakes have occurred in this area and the activity has continued to the present time. Charleston is about 420 miles east of the plant site.

d. The Southern Appalachian area of western North Carolina and eastern Tennessee. Light-to-moderate earthquakes occur in this area at an average frequency of one or two per year. This area is centered about 200 miles east of Decatur.

2.5-9

BFN-26 In addition to these areas, shocks of light-to-moderate intensity have occurred at many other localities in the southeastern states at various distances from the Browns Ferry plant site. At many of these localities, only a few light-to-moderate shocks from widely scattered centers are known. Seismic history in the vicinity of the plant is discussed in Section 2.5.3.4. Several scales have been devised to evaluate the force of earthquake shocks. Scales which rate earthquakes on the degree of shaking at any given locality are known as intensity scales. In general, the intensity of an earthquake is highest in the epicentral area and diminishes in all directions from the point of maximum intensity. Another factor affecting the intensity of an earthquake is the character of the ground in a given locality. The shaking is much less severe, other things being equal, at a place on bedrock than one on thick alluvium. The most widely used intensity scale is the modified Mercalli scale which has 12 degrees of intensity (see Figure 2.5-7). The degrees of intensity are expressed by Roman numerals from I to XII. The Richter magnitude scale applies to an earthquake as a whole rather than the observations made at some point, or points, within the area affected. Earthquake magnitude is calculated from measurements on seismograms and is expressed by an Arabic numeral and a fraction, such as 6-3/4 or 5.5. 2.5.3.3 Geologic and Tectonic Background As discussed in Section 2.5.2, Browns Ferry Nuclear Plant is founded on a thick succession of essentially horizontal sedimentary rocks. The site is 16.5 miles away from the nearest known inactive fault and approximately 200 miles from the New Madrid region of the Mississippi Valley. Since the site area is very low on the southeastern flank of the Nashville structural dome, it has undergone no tectonic movement except simple uplift. This movement probably ceased at the close of the Paleozoic Era. 2.5.3.4 Seismic History A list of seismic events within a 200-mile radius of the plant site that occurred from 1699 through 1980 is presented in Table 2.5-1. Epicenters of earthquakes within a 200-mile radius of the plant site based on Table 2.5-1 are shown on Figure 2.5-6. 2.5.3.5 Seismicity of the Area Light shocks have been centered near Huntsville, Hazel Green, Anniston, Cullman, Easonville, and Birmingham, but most of them were felt as low intensity shocks or not at all in the Decatur area. The shocks felt most strongly in the area have been major earthquakes centered at distant points, especially in the Mississippi Valley. 2.5-10

BFN-26 There is continuing seismic activity in the Mississippi Valley, and the possibility of another great earthquake in the New Madrid region of Missouri, Arkansas, Tennessee, and Kentucky cannot be discounted. An earthquake of intensity XI or XII at New Madrid might be felt in the Decatur area with an intensity of VII. 2.5.4 Conclusions (References 8 and 10) The site is underlain by massive formations of bedrock, thus providing adequate foundations for all plant structures. The major seismic activity experienced at the site has been caused by distant major earthquakes, especially those at New Madrid and Charleston. For design purposes, a conservative assumption was made that a seismic event at an unstated location could cause an intensity of VII at the plant site. Thus, the design of structures and equipment important to the plant safety features was based on a horizontal ground motion due to a peak acceleration of 0.10g. In addition, the design is such that the plant can be safely shut down during a peak horizontal ground acceleration of 0.20g. Vertical accelerations are two thirds of the horizontal accelerations. Details of the earthquake design of these structures are given in Chapter 12. Design based upon these ground accelerations provides a margin of protection against either minor shocks originating near the site or major shocks originating at New Madrid or Charleston. Since the site is located in an area of extremely low seismicity, it has been principally affected, if at all, by distant, strong earthquakes. The response spectra chosen for the site are shown in Figures 2.5-8 and 2.5-9 for the OBE and DBE, respectively. The time-history method was used in analyzing all structures; and the El Centro, May 1940, N-S component was chosen for this purpose. This record was determined to adequately represent any potential threat to the site. A comparison of the response spectrum produced by this record and the spectrum used for design is shown in Figure 2.5-10. As an alternate basis for design of subsystems, an artificial acceleration time history input ground motion was used. Figure 2.5-11 depicts the acceleration, velocity, and displacement time history of this record. Figures 2.5-12 through 2.5-16 compare the response spectra of this time history to the site design spectra for the various damping levels. The artificial time history meets the enveloping requirements of Section 3.7.1, "Standard Review Plan." 2.5.5 Seismic Instrumentation Program Seismic instrumentation is provided in order to assess the effects on the plant of earthquakes which may occur that exceed the ground acceleration for the Operating Basis Earthquake (OBE = 0.10g ground acceleration). The seismic instrumentation is not safety-related and does not have any effect on safety-related systems or equipment. The seismic instruments were selected to emphasize accuracy and reliability, while at the same time minimizing the maintenance and surveillance 2.5-11

BFN-26 resources required to support the system. The instrumentation that is provided is described in the following sections. 2.5.5.1 Location and Description of Instrumentation The seismic instrumentation locations are shown in Figure 2.5-17. It is solid state digital instrumentation which will enable the processing of data at the plant site within 4 hours of a seismic event. One of the sensors is located at the top-of-rock where the OBE design response spectrum is defined. Therefore, this instrumentation is sufficient to adopt the OBE exceedance criteria described in Reference 2.5.6-11 through 15. The instrumentation consists of the following:

1. A strong-motion triaxial time-history accelerograph at each of the following locations:

a. El. 519.5, Unit 1 Reactor Building, south corner of the Northwest Quad Room,

b. El. 621.75, Unit 1 Reactor Building, upper level in Electrical Board Room 1A, and

c. El. 566.0, Unit 1 and 2 Diesel Generator Building, on the base slab in Room B.

These accelerographs have a full scale range of +2g. The internal recorder is capable of digitally recording a minimum of 25 minutes of data with a minimum of 3 seconds of pre-event memory and 5 seconds of post-event memory. An internal seismic trigger with a bandwidth of 0.1Hz - 12.5 Hz actuates the recording system when a threshold acceleration level is sensed. The unit is equipped with an internal rechargeable battery and an external plug-in type battery charger.

2. The centralized seismic instrumentation panel components are located in the Unit 1 Main Control Room (Elevation 621.0). This panel contains: a) central controller, b) alarm panel, and c) display panel. A description of each item is given below.

a. A central controller consisting of an industrial computer and custom software, which provides a user interface in a multi-tasking operating system that supports simultaneous seismic data acquisition and interrogation. The controller is powered by 120V AC power. The central controller retrieves data files from the internal digital recorders in each remote accelerograph after an event and, for the top-of-rock sensor, 2.5-12

BFN-26 performs automatic analysis of the data. The event-analysis capabilities include computation of the Cumulative Absolute Velocity (CAV), spectral content of the recorded data, and automatic comparison to the site OBE (OBE = 1/2 SSE) response spectrum. Hard copies of the operational data and event analysis will be printed to a Unit 1 Main Control Room printer. The central controllers software capabilities also include automatic event alarm and annunciation (See item 2b). The event analysis functions of the central controller may be performed off-line, if necessary.

b. An alarm panel containing visual alarms to indicate that a seismic event has been recorded, the OBE response spectrum has been exceeded in a damaging frequency range, and system trouble including either loss of AC or DC power. The seismic event alarm is triggered by the accelerographs, while the OBE exceedance and system trouble alarms are triggered by the central controller. Activation of either the event alarm, exceedance alarm, or system trouble alarm also causes corresponding windows on an annunciator panel in the Unit 1 Main Control Room.

c. A display panel to provide a visual display for operation of the centralized system.

3. Annunciator lights in the Unit 1 Main Control Room:

a. Window Legends: START OF STRONG MOTION ACCELEROGRAPH -

Any one of the three strong-motion accelerographs (Item 1) will activate this window if an acceleration greater than or equal to 0.01g is sensed.

b. Window Legends: 1/2 SSE RESPONSE SPECTRUM EXCEEDED - The central controller in conjunction with the alarm panel (Items 2a and 2b) will activate this window if the central controller has determined the OBE response spectrum to be exceeded. This determination is based only on input received from the Reactor Building base slab accelerograph (Item 1a).

c. Window Legends: SEISMIC MONITORING SYSTEM TROUBLE -

The alarm panel will activate this window if the central controller detects a system trouble or if there is loss of AC or DC power. The basis for selecting the Reactor Building for installation of seismic instruments is that it is the rock-supported building most important to safety. The basis for selecting the Diesel Generator Building is that it is the soil-supported building most important to safety. 2.5.5.2 Control Room Operator Notification 2.5-13

BFN-26 The seismic monitoring system provides three independent alarm windows in the Unit 1 Main Control Room. The first annunciator window indicates system trouble which serves to provide warning of equipment operability problems under normal power conditions as well as following a seismic event. The next annunciator window is provided by the accelerographs (Item 1) via the controller, Section 2.5.5.1, which alerts the operator that a seismic event is being recorded. This annunciation indicates that at least one of the accelerographs triggers has sensed seismic motion in excess of 0.01g. The final annunciator window is actuated later and is provided by the central controller (Item 2a), Section 2.5.5.1, and is only received if the event-analysis software indicates that the site OBE design basis response spectrum has been exceeded in a potentially damaging frequency range, as described in Section 2.5.5.3. The basis for establishing the OBE design basis response spectrum for the levels at which control room operator notification is required is that the design of Structures, Systems, and Components (SSCs) for loading combinations, which include OBE, are to design basis allowable stress levels which are well within the elastic limit of the materials. 2.5.5.3 Controlled Shutdown Logic The operator will utilize input from multiple sources to determine the need for a controlled shutdown following the seismic event. The decision for a controlled shutdown will be based primarily on an assessment of the actual damage potential of the event, which will be available within 4 hours, and on the results of short-term inspections, which will be available within 8 hours. The operator may also confirm that ground motion was sensed by plant personnel and/or confirm the occurrence of the seismic event with the National Earthquake Information Center. The purpose of these actions are 1) to perform a preliminary assessment of the effect of the earthquake on the physical condition of SSCs, and 2) to determine if shutdown of the plant is warranted based on observed damage to SSCs, or because the OBE has been exceeded. The walkdowns of plant SSCs in accessible areas of the plant will be performed witin 8 hours following the seismic event. The walkdowns will be performed using the general guidance in Chapter 4 of the Electrical Power Research Institute (EPRI) Report NP-6695 (ref. 2.5.6-12). These walkdowns will include a check of the neutron flux monitoring sensors for changes and an inspection of the containment isolation system to ensure continued containment integrity. The walkdown data will be compared to data previously obtained from baseline and Maintenance Rule inspections in order to obtain a clear understanding of any seismically induced damage. 2.5-14

BFN-26 The assessment of the damage potential of the event will be made within 4 hours following the event using the OBE Exceedance Criteria developed by EPRI and documented in references 2.5.6-11 through -15. As noted above, the indication of damage potential will be provided by event-analysis software installed on the centralized seismic monitoring system described in Section 2.5.5.1. The analysis will be performed for the uncorrected accelerograms recorded from the strong motion triaxial accelerograph located in the Unit 1 Reactor Building on the base slab (item 1a of Section 2.5.5.1). Use of the uncorrected accelerograms is known to be conservative. The basis for use of the seismic motion on the Reactor Building base slab is that the site OBE design response spectrum is defined at top-of-rock, which corresponds to the Reactor Building base slab location. The EPRI OBE Exceedance Criteria uses two indicators of damage potential. The first indicator of damage potential is specified as the Cumulative Absolute Velocity (CAV) of the accelerogram. A meaningful usage of the CAV requires that the recorded data be obtained by the accelerometer mounted in the free-field. As noted above, the OBE design spectrum for Browns Ferry is defined as occurring at top-of-rock (i.e., foundation level of the rock-supported structures); whereas, free-field is defined as top-of-soil at sufficient distance from nearby structures to preclude interference/interaction effects. The Seismic Monitoring System for Browns Ferry does not have a free-field accelerometer. Therefore, the shutdown logic adopted for Browns Ferry will concede CAV exceedance and base the assessment of damage potential solely on the second indicator, as discussed below. In the absence of data from a free-field accelerometer, the second indicator is an evaluation of the frequency range in which the OBE spectrum is exceeded. This criteria is based on research which indicates that exceedances above a frequency of 10 Hz are not damaging to nuclear plant SSCs. The following two measures of damaging potential are used.

• The OBE site design basis response spectrum is exceeded if the 5 percent damping response spectra generated for any one of the three components of the uncorrected accelerograms from the Reactor Building foundation accelerometer is larger than:

1. The corresponding OBE design basis response spectral acceleration in a frequency range between 2-10 Hz, or

2. The corresponding OBE design basis response spectral velocity for frequencies between 1-2 Hz.

Therefore, Browns Ferry will base the assessment of damage potential of the event on either a spectral acceleration exceedance between 2-10 Hz or a spectral velocity exceedance between 1-2 Hz. 2.5-15

BFN-26 Once the results of the walkdown and the assessment of damage potential of the event are available, the operators will determine 1) if a controlled shutdown is required and 2) the condition of the equipment needed to safely achieve shutdown. If the assessment of damage potential indicates that the OBE Exceedance Criteria were not met, and the walkdown results are favorable, the plant will continue to operate. Basing shutdown logic on the actual damage potential of the event and on the results of short-term inspection avoids unnecessary shutdowns while ensuring that the operator has the plant status information needed to make an informed reactor shutdown decision. Post-shutdown actions, including retrieval of data, recalibration of seismic instruments, and comparison of measured and predicted responses will be based on the guidance in Chapters 5 and 6 of EPRI Report NP-6695 (Ref. 2.5.6-12). 2.5.6 References

1. General Geology, Geophysics, and Seismicity of Northwest Alabama, prepared by J. T. Kidd, Geological Survey of Alabama, 1980, NUREG/CR-1519, various pages. (R40990309903)

2. Mississippian Stratigraphy of Alabama, William A. Thomas, Geological Survey of Alabama, Monograph 12, 1972, various pages. (R40990219898)

3. Memorandum - Berlin C. Moneymaker to Ralph O. Lane, Browns Ferry Nuclear Plant Site - Compressive Strength Tests on Cores from Foundation, June 1, 1966. (R40990309901)

4. Memorandum - R. O. Lane to B. C. Moneymaker, Tims Ford and Browns Ferry Sites - Rock Cores - Compressive Strength Test Results, June 10, 1966.

(R40990309902)

5. Static and Dynamic Rock Testing, Browns Ferry Reactor Site, Prepared For:

John Blume and Associates by Woodward - Clyde - Sherard & Associates, November 17, 1966. (R40990219897)

6. Memorandum - G. H. Kimmons to W. F. Emmons, Nuclear Plant X-5, Browns Ferry Site - Soil Investigation, May 4, 1966. (R40990309900)

7. TVA Division of Water Control Planning Geologic Branch - Browns Ferry Nuclear Power Plant Section on Seismology, Berlin C. Moneymaker, May 3, 1966. (R40990219894) 2.5-16

BFN-26

8. TVA Division of Water Control Planning Geologic Branch - Seismology of the Browns Ferry Nuclear Plant, Berlin C. Moneymaker, April 1969.

(R40990219896)

9. Recent Seismic Activity in the Tennessee Valley Area supplied by J. W.

Munsey, Geophysicist with the TVA Resource Group. (R40990309904)

10. Memorandum - W. F. Emmons to W. C. Boop, Browns Ferry Nuclear Plant -

Recommended Earthquake Design Acceleration, June 6, 1966. (R40990219895)

11. EPRI Report NP-5930, A Criterion for Determining Exceedance of the Operating Basis Earthquake, July 1988.

12. EPRI Report NP-6695, Guidelines for Nuclear Plant Response to an Earthquake, December 1989.

13. EPRI Report TR-100082, Standardization of the Cumulative Absolute Velocity, December 1991.

14. Nuclear Regulatory Commission Regulatory Guide 1.166, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions, March 1997

15. Nuclear Regulatory Commission Regulatory Guide 1.12, Nuclear Power Plant Instrumentation for Earthquakes, March 1997 2.5-17

BFN-16 TABLE 2.5-1 Sheet 1 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

1. 1699 Dec 25 IV (Northeast AK) 35.2 90.5

2. 1812 Feb 7 XII New Madrid, MO 36.6 89.5

3. 1816 Jul 25 IV New Madrid, MO 36.6 89.5

4. 1818 Mar V Caruthersville, MO 36.2 89.7

5. 1820 IV New Madrid, MO 36.6 89.5

6. 1824 Aug 22 Jackson, TN 35.6 88.8

7. 1825 Mar 19 Columbia, TN 35.6 87.0

8. 1829 <IV Andrews, NC 35.2 83.8

9. 1829 May IV Jackson, TN 35.6 88.8

10. 1839 Sep 5 V Mayfield, KY 36.7 88.6

11. 1841 Dec 28 V Hickman, KY 36.6 89.2

12. 1842 Nov 4 IV Hickman, KY 36.6 89.2

13. 1842 Mar 28 IV Hickman, KY 36.6 89.2

14. 1843 Jan 5 VIII Marked Tree, AK 35.5 90.4

15. 1843 Jun 13 III Hickman, KY 36.6 89.2

16. 1843 Aug 9 IV Columbia, TN 35.6 87.0

17. 1846 Mar 26 III New Madrid, MO 36.6 89.5

18. 1849 Jan 24 IV Hickman, KY 36.6 89.2

19. 1853 Aug 28 III Hickman, KY 36.6 89.2

20. 1853 Dec 18 V Hickman, KY 36.6 89.2

21. 1855 May 3 IV Cairo, IL 37.0 89.2

22. 1857 Feb IV New Madrid, MO 36.6 89.5

23. 1858 Sep 21 V Hickman, KY 36.6 89.2

24. 1860 Jan 20 NC-SC-GA 0.0 0.0

25. 1865 Aug 17 VII 36.5 89.5

26. 1865 Sep 7 New Madrid, MO 36.6 89.5

27. 1868 Nov 21 Hickman, KY 36.6 89.2

28. 1870 Dec 14 Hickman, KY 36.6 89.2

29. 1871 Jul 25 IV Cairo, IL 37.0 89.2

30. 1872 Feb 8 Cairo, IL 37.0 89.2

31. 1872 Apr 20 Memphis, TN 35.1 90.1

32. 1872 Aug 20 Memphis, TN 35.1 90.1

33. 1873 May 3 IV 36.0 89.6

34. 1873 Aug 22 Memphis, TN 35.1 90.1

35. 1874 Jul 9 IV Cairo, IL 37.0 89.2

36. 1875 Oct 7 IV 36.1 89.6

37. 1875 Oct 28 IV Memphis, TN 35.1 90.1

38. 1877 Jul 15 IV Cairo, IL 37.0 89.2

39. 1877 Nov 19 Cairo, IL 37.0 89.2

40. 1878 Jan 9 Cairo, IL 37.0 89.2

41. 1878 Mar 12 VII Columbus, KY 36.8 89.1

42. 1878 Nov 19 VI 36.7 89.3

43. 1878 Nov 23 <IV Murphy, NC 35.1 84.0

44. 1879 Jul 26 Cairo, IL 37.0 89.2

45. 1879 Sep 26 IV NE AK 35.3 90.3

46. 1800 Jul 14 IV 35.3 90.3

47. 1881 Oct 7 IV Memphis, TN 35.1 90.1

48. 1882 Jul 20 IV Cairo, IL 37.0 89.2

49. 1882 Oct 15 <IV Murphy, NC 35.1 84.0

50. 1883 Jan 1 IV Ashwood, TN 35.6 87.1

51. 1883 Jan 11 V Cairo, IL 37.0 89.2

52. 1883 Apr 12 VI Cairo, IL 37.0 89.2

53. 1883 Jul 6 Cairo, IL 37.0 89.2

54. 1883 Jul 14 V Wickliffe, KY 37.0 89.1

55. 1884 Apr 30 <IV Ogretta, NC 35.2 84.2

56. 1884 Nov 30 IV 35.5 89.7

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 2 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

57. 1886 Feb 5 IV Valley Head, AL 34.6 85.6

58. 1886 Feb 13 VI Demopolis, AL 32.5 87.8

59. 1886 Mar 18 IV Cairo, IL 37.0 89.2

60. 1887 Aug 2 VI Cairo, IL 37.0 89.2

61. 1888 Nov 3 IV Memphis, TN 35.1 90.1

62. 1889 Jan 5 Memphis, TN 35.1 90.1

63. 1889 Jun 6 Memphis, TN 35.1 90.1

64. 1889 Jun 7 IV Benton Co., TN 35.9 88.1

65. 1889 Jul 20 VI Memphis, TN 35.1 90.1

66. 1889 Sep 28 <IV Parksville, TN 35.1 84.6

67. 1891 Jan 14 Memphis, TN 35.1 90.1

68. 1891 Sep 27 IV Cairo, IL 37.0 89.2

69. 1892 Jan 14 IV Memphis, TN 35.1 90.1

70. 1892 Dec 2 V Chattanooga, TN 35.0 85.3

71. 1894 Jul 18 Memphis, TN 35.1 90.1

72. 1895 Jul 27 Savannah, TN 35.2 88.3

73. 1895 Oct 3 Memphis, TN 35.1 90.1

74. 1895 Oct 18 New Madrid, MO 36.3 89.5

75. 1895 Nov 17 Charleston, MO 36.9 89.3

76. 1897 Apr 26 IV Osceola, AK 35.7 90.0

77. 1898 Mar 30 <IV Mt. Hermon, KY 36.8 85.8

78. 1898 Jun 14 IV New Madrid, MO 36.6 89.5

79. 1901 Feb 15 IV 36.0 90.0

80. 1901 Sep 14 Memphis, TN 35.1 90.1

81. 1902 May 29 IV Chattanooga, TN 35.0 85.3

82. 1902 Oct 18 V Chattanooga, TN 35.0 85.3

83. 1903 Nov 4 VII Charleston, MO 36.9 89.3

84. 1903 Nov 24 New Madrid, MO 36.6 89.5

85. 1903 Nov 27 IV New Madrid, MO 36.6 89.5

86. 1904 Mar 5 <IV Maryville, TN 35.8 84.0

87. 1908 Sep 28 IV New Madrid, MO 36.6 89.5

88. 1908 Oct 28 IV Cairo, IL 37.0 89.2

89. 1909 Oct 8 <IV Dalton, GA 34.8 85.0

90. 1913 Mar 13 <IV Calhoun, GA 34.5 85.0

91. 1913 Apr 17 V Madisonville, TN 35.5 84.4

92. 1913 May 2 <IV Madisonville, TN 35.5 84.4

93. 1913 Jun 9 IV Humboldt, TN 35.8 88.9

94. 1914 Jan 24 IV Sweetwater, TN 35.6 84.5

95. 1914 Mar 5 IV Central GA 33.5 84.0

96. 1915 Feb 19 IV Cairo, IL 37.0 89.2

97. 1915 Apr 28 IV Tiptonville, TN 36.4 89.5

98. 1915 Oct 26 V Mayfield, KY 36.7 88.6

99. 1915 Dec 7 V Cairo, IL 37.0 89.2 100. 1916 May 21 IV New Madrid, MO 36.6 89.5 101. 1916 Aug 24 IV New Madrid, MO 36.6 89.5 102. 1916 Oct 18 VII Irondale, AL 33.5 86.7 103. 1916 Oct 19 Mayfield, KY 36.7 88.6 104. 1916 Nov 4 IV Birmingham, AL 33.5 86.8 105. 1916 Dec 19 VI Hickman, KY 36.6 89.2 106. 1917 Jun 9 IV New Madrid, MO 36.6 89.5 107. 1917 Jun 30 IV Rosemary, AL 32.7 87.5 108. 1918 Feb 17 Cairo, IL 37.0 89.2 109. 1918 Jun 22 IV Lenoir City, TN 35.8 84.3 110. 1918 Oct 16 IV 35.2 89.2 111. 1919 May 23 36.6 89.2

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 3 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 112. 1919 May 28 36.6 89.2 113. 1920 Apr 7 II 36.3 88.2 114. 1920 Dec 24 IV Glen Alice, TN 35.8 84.7 115. 1921 Jan 9 IV New Madrid, MO 36.6 89.5 116. 1921 Feb 27 IV Cairo, IL 37.0 89.2 117. 1921 Sep 2 IV Statesville, TN 36.0 86.1 118. 1921 Dec 15 IV Glen Alice, TN 35.0 84.7 119. 1922 Mar 30 <IV Farmington, TN 35.5 86.7 120. 1922 Mar 30 V Caruthersville, MO 36.1 89.7 121. 1923 Mar 27 IV Wyatte, MS 34.6 89.7 122. 1923 May 6 Cairo, IL 37.0 89.2 123. 1923 May 15 Cairo, IL 37.0 89.2 124. 1923 Oct 28 VII Marked Tree, AK 35.5 90.4 125. 1923 Nov 26 IV Marked Tree, AK 35.5 90.4 126. 1923 Nov 28 37.5 87.3 127. 1923 Nov 29 Wickliffe, KY 37.0 89.1 128. 1924 Mar 2 IV 36.9 89.1 129. 1924 Jun 7 IV Tiptonville, TN 36.4 89.5 130. 1925 May 13 IV Mayfield, KY 36.7 88.6 131. 1926 Apr 28 IV Kenton, TN 36.2 89.0 132. 1926 Dec 17 IV Tiptonville, TN 36.4 89.5 133. 1927 Apr 18 Ridgely, TN 36.3 89.5 134. 1927 Jun 16 IV Scottsboro, AL 34.7 86.0 135. 1927 Aug 13 IV Tiptonville, TN 36.4 89.5 136. 1927 Oct 8 IV Chattanooga, TN 35.0 85.3 137. 1928 Mar 7 IV Columbia, TN 35.6 87.0 138. 1928 Apr 23 Hickman, KY 36.6 89.2 139. 1928 May 31 New Madrid, MO 36.6 89.5 140. 1929 May 13 Tiptonville, TN 36.4 89.5 141. 1930 Jan 2 Ripley, TN 35.7 89.5 142. 1930 Feb 18 Marked Tree, AK 35.5 90.4 143. 1930 Feb 25 IV Cairo, IL 37.0 89.2 144. 1930 Mar 27 IV Raleigh, TN 35.2 89.9 145. 1930 Apr 2 IV Caruthersville, MO 36.2 89.7 146. 1930 Aug 13 New Madrid, MO 36.6 89.5 147. 1930 Aug 29 IV Blandville, KY 36.9 88.9 148. 1930 Aug 30 V Kingston, TN 35.9 84.5 149. 1930 Sep 1 V Marston, MO 36.5 89.6 150. 1930 Sep 3 Blandville, KY 36.9 88.9 151. 1931 Apr 1 Hopkinsville, KY 36.9 87.5 152. 1931 Apr 6 IV Berkeley, KY 36.8 89.1 153. 1931 May 5 VI Birmingham, AL 33.5 86.8 154. 1931 Jul 18 IV New Madrid, MO 36.6 89.5 155. 1931 Nov 27 <IV Nashville, TN 36.2 86.8 156. 1931 Dec 10 IV Blytheville, AK 35.9 89.9 157. 1931 Dec 17 VII Charleston, MO 36.9 89.3 158. 1932 Nov 22 IV Blytheville, AK 35.9 89.9 159. 1933 Dec 9 V Manila, AK 35.9 90.2 160. 1934 Jul 3 IV Memphis, TN 35.1 90.1 161. 1934 Aug 20 VII Rodney, MO 37.0 89.2 162. 1935 Jul 24 IV Tiptonville, TN 36.4 89.5 163. 1936 Jan 1 <IV Blue Ridge, CA 34.9 84.3 164. 1936 Feb 17 Hayti, MO 36.2 89.7 165. 1936 Aug 2 IV Tiptonville, TN 36.4 89.5 166. 1936 Oct 20 New Madrid, MO 36.6 89.5

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 4 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 167. 1936 Oct 31 New Madrid, MO 36.6 89.5 168. 1937 Jan 30 V Caruthersville, MO 36.2 89.7 169. 1937 Jun 23 Tiptonville, TN 36.4 89.5 170. 1937 Oct 5 New Madrid, MO 36.6 89.5 171. 1938 Mar 16 New Madrid, MO 36.6 89.5 172. 1938 Mar 31 IV Tapoco, NC 35.5 84.0 173. 1938 Jun 17 IV 35.8 89.8 174. 1938 Sep 18 V 35.5 90.3 175. 1938 Sep 19 Tiptonville, TN 36.4 89.5 176. 1939 Apr 15 IV New Madrid, MO 36.6 89.5 177. 1939 May 5 V Anniston, AL 33.7 85.8 178. 1939 Jun 24 IV Huntsville, AL 34.7 86.6 179. 1940 Feb 14 Blytheville, AK 35.9 89.9 180. 1940 May 31 V Paducah, KY 37.1 88.6 181. 1940 Sep 19 New Madrid, MO 36.6 89.5 182. 1940 Oct 10 36.8 89.2 183. 1940 Oct 19 IV Ryall Springs, TN 35.0 85.1 184. 1941 Sep 8 IV Lookout Mt., TN 35.0 85.4 185. 1941 Oct 8 VI Blytheville, AK 35.9 89.9 186. 1941 Oct 21 IV Cairo, IL 37.0 89.2 187. 1941 Nov 15 IV Memphis, TN 35.1 90.1 188. 1941 Nov 15 Memphis, TN 35.1 90.1 189. 1941 Nov 17 V Covington, TN 35.6 89.6 190. 1942 Aug 31 IV Cairo, IL 37.0 89.2 191. 1944 Dec 23 Caruthersville, MO 36.2 89.7 192. 1945 May 2 IV Marston, MO 36.5 89.6 193. 1945 Jun 14 V Cleveland, TN 35.2 84.9 194. 1945 Aug 6 Caruthersville, MO 36.2 89.7 195. 1945 Sep 23 IV Cairo, IL 37.0 89.2 196. 1945 Oct 27 Point Pleasant, MO 36.5 89.6 197. 1945 Nov 13 IV Cairo, IL 37.0 89.2 198. 1946 Apr 7 IV Cleveland, TN 35.2 84.9 199. 1947 Jan 16 Cairo, IL 37.0 89.2 200. 1947 Mar 26 VI 37.0 88.4 201. 1947 Dec 16 IV Lepanto, AK 35.6 90.3 202. 1947 Dec 28 IV Ryall Springs, TN 35.0 85.1 203. 1949 Jan 14 V Portageville, MO 36.4 89.7 204. 1949 Jan 31 V 36.3 89.7 205. 1949 Aug 13 Caruthersville, MO 36.2 89.7 206. 1950 May 1 Gideon, MO 36.4 89.9 207. 1950 Jun 19 IV Tapoco, NC 35.5 84.0 208. 1950 Sep 17 IV 35.8 90.0 209. 1951 Dec 18 IV New Madrid, MO 36.6 89.5 210. 1952 Feb 2 V Tiptonville, TN 36.4 89.5 211. 1952 Feb 6 V Birmingham, AL 33.5 86.8 212. 1952 Mar 17 IV 36.2 89.6 213. 1952 May 28 Catron, MO 36.6 89.7 214. 1952 Jul 16 V Dyersburg, TN 36.0 89.4 215. 1952 Oct 18 IV Dyersburg, TN 36.0 89.4 216. 1952 Dec 25 IV Blytheville, AK 35.9 89.9 217. 1952 Dec 28 Caruthersville, MO 36.2 89.7 218. 1953 Jan 26 IV Finley, TN 36.0 89.5 219. 1953 Feb 11 IV New Madrid, MO 36.6 89.5 220. 1953 Feb 17 IV Finley, TN 36.0 89.5 221. 1953 May 6 Cairo, IL 37.0 89.2

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 5 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 222. 1953 May 12 Lepanto, AK 35.6 90.3 223. 1953 May 15 IV Cairo, IL 37.0 89.2 224. 1954 Jan 17 IV Dyersburg, TN 36.0 89.4 225. 1954 Jan 23 IV Etowah, TN 35.3 84.5 226. 1954 Apr 27 V Memphis, TN 35.1 90.1 227. 1955 Jan 12 IV Maryville, TN 35.8 84.0 228. 1955 Jan 25 VI Finley, TN 36.0 89.5 229. 1955 Mar 29 V Finley, TN 36.0 89.5 230. 1955 Sep 5 IV Finley, TN 36.0 89.5 231. 1955 Sep 24 IV Tiptonville, TN 36.4 89.5 232. 1955 Dec 13 IV Finley, TN 36.0 89.5 233. 1956 Jan 24 Caruthersville, MO 36.2 89.7 234. 1956 Jan 29 VI Covington, TN 35.6 89.6 235. 1956 Sep 9 IV College Grove, TN 35.8 86.7 236. 1956 Oct 29 IV Caruthersville, MO 36.2 89.7 237. 1957 Mar 26 V Paducah, KY 37.1 88.6 238. 1957 Apr 23 VI Birmingham, AL 33.5 86.8 239. 1957 Jun 23 IV Dixie Lee Junction, TN 35.9 84.2 240. 1957 Aug 17 IV Bogota, TN 36.2 89.4 241. 1957 Nov 7 <IV Powell, TN 36.0 84.0 242. 1958 Jan 26 V Caruthersville, MO 36.2 89.7 243. 1958 Jan 28 V Cairo, IL 37.0 89.2 244. 1958 Apr 8 IV Troy, TN 36.3 89.2 245. 1958 Apr 26 IV Tiptonville, TN 36.4 89.5 246. 1958 May 19 IV Marked Tree, AK 35.5 90.4 247. 1959 Jan 21 V Ridgely, TN 36.3 89.5 248. 1959 Feb 13 IV Bogota, TN 36.2 89.4 249. 1959 Jun 13 IV Tellico Plains, TN 35.4 84.3 250. 1959 Jul 20 Blytheville, AK 35.9 89.9 251. 1959 Aug 12 VI Meridianville, AL 34.8 86.6 252. 1959 Dec 21 V Finley, TN 36.0 89.5 253. 1960 Jan 28 IV Finley, TN 36.0 89.5 254. 1960 Apr 15 IV Maryville, TN 35.8 84.0 255. 1960 Apr 21 IV Tiptonville, TN 36.4 89.5 256. 1962 Feb 2 VI 35.6 89.6 257. 1962 Jun 1 36.0 90.2 258. 1962 Jul 23 VI Dyersburg, TN 36.0 89.4 259. 1963 Mar 31 36.5 89.5 260. 1963 Apr 6 New Madrid, MO 36.6 89.5 261. 1963 May 2 New Madrid, MO 36.6 89.5 262. 1963 Aug 3 V Paduch, KY 37.1 88.6 263. 1963 Nov 14 <IV Nashville, TN 36.2 86.8 264. 1963 Dec 5 <IV Beechmont, KY 37.2 87.0 265. 1963 Dec 15 <IV Beechmont, KY 37.2 87.0 266. 1964 Jan 16 36.8 89.5 267. 1964 Jan 25 36.5 89.5 268. 1964 Feb 18 V Mentone, AL 34.6 85.6 269. 1964 Mar 17 IV Caruthersville, MO 36.2 89.7 270. 1964 May 2 New Madrid, MO 36.6 89.5 271. 1964 May 23 New Madrid, MO 36.6 89.5 272. 1964 Jul 28 <IV Inskip, TN 36.0 84.0 273. 1965 Feb 11 36.4 89.7 274. 1965 Mar 25 IV New Madrid, MO 36.6 89.5 275. 1965 May 25 36.1 89.9 276. 1965 Jun 1 36.5 89.5

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 6 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 277. 1965 Jul 8 36.5 89.5 278. 1965 Nov 8 <IV Canton, GA 34.2 84.5 279. 1965 Dec 19 35.9 89.9 280. 1966 Feb 12 35.9 90.0 281. 1966 Feb 13 IV Covington, TN 35.6 89.6 282. 1966 Mar 13 36.2 90.0 283. 1966 Aug 24 IV Maryville, TN 35.8 84.0 284. 1967 Apr 11 36.1 89.7 285. 1967 Oct 18 36.5 89.5 286. 1968 Jan 23 36.5 89.5 287. 1968 May 30 36.5 89.5 288. 1968 Jul 15 36.5 89.5 289. 1970 Jan 7 IV Raleigh, TN 35.2 89.9 290. 1970 Mar 27 New Madrid, MO 36.6 89.5 291. 1970 Nov 5 36.0 90.0 292. 1970 Nov 17 V Blytheville, AK 35.9 89.9 293. 1970 Nov 30 36.3 89.5 294. 1970 Dec 14 35.7 90.0 295. 1970 Dec 24 IV New Madrid, MO 36.6 89.5 296. 1971 Mar 14 Carrollton, AL 33.3 88.1 297. 1971 Apr 13 35.8 90.1 298. 1971 Jul 13 IV Kingston, TN 35.9 84.5 299. 1971 Oct 18 36.7 89.6 300. 1972 Mar 29 V New Madrid, MO 36.6 89.5 301. 1972 May 7 IV Blytheville, AK 35.9 89.9 302. 1973 Jan 7 Madisonville, KY 37.4 87.5 303. 1973 Oct 3 IV 35.9 90.0 304. 1973 Oct 9 IV New Madrid, MO 36.6 89.5 305. 1973 Nov 30 VI Maryville, TN 35.8 84.0 306. 1973 Dec 20 IV Caruthersville, MO 36.2 89.7 307. 1974 Jan 8 V Bufordville, MO 36.2 89.4 308. 1974 Mar 4 35.7 90.4 309. 1974 Mar 10 36.2 89.5 310. 1974 Mar 12 35.7 89.8 311. 1974 May 13 V East Prairie, MO 36.8 89.4 312. 1974 Dec 25 IV Blytheville, AK 35.9 89.9 313. 1975 Feb 13 V Conran, MO 36.5 89.6 314. 1975 Mar 1 Smithville, MS 34.1 88.4 315. 1975 May 2 Oakdale, TN 36.0 84.6 316. 1975 May 14 Oak Ridge, TN 36.0 84.3 317. 1975 Jun 13 VI Lilbourn, MO 36.6 89.6 318. 1975 Jun 24 IV Fayette, AL 33.7 87.8 319. 1975 Jul 6 Miston, TN 36.2 89.5 320. 1975 Aug 25 36.0 89.8 321. 1975 Aug 29 VI Palmerdale, AL 33.8 86.6 322. 1975 Nov 7 Samantha, AL 33.4 87.6 323. 1975 Dec 3 V New Madrid, MO 36.6 89.5 324. 1976 Feb 4 VI Conasauga, TN 35.0 84.7 325. 1976 Apr 15 V Sacramento, KY 37.4 87.3 326. 1976 May 22 V 36.0 89.8 327. 1976 Oct 23 32.2 88.7 328. 1977 Mar 28 II Marston, MO 36.5 89.6 329. 1977 Jul 27 V Athens, TN 35.4 84.6 330. 1977 Nov 4 V Vardaman, MS 33.6 89.2 331. 1978 Jan 8 near Gainsville, AL 32.8 88.2

BFN-16 TABLE 2.5-1 (Cont'd) Sheet 7 BROWNS FERRY NUCLEAR PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 87.11 W LON 34.71 N LAT FEBRUARY 10, 1982 YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 332. 1978 Jan 18 III Ridgely, TN 36.3 89.5 333. 1978 Mar 1 III near Huntsville, AL 34.4 86.6 334. 1978 Aug 31 V Dyersburg, TN 36.0 89.4 335. 1978 Oct 27 near Jasper, AL 33.8 87.5 336. 1978 Nov 21 II Blytheville, AK 35.9 89.9 337. 1979 Feb 2 III Ridgely, TN 36.3 89.5 338. 1979 Feb 2 II Ridgely, TN 36.3 89.5 339. 1979 Feb 2 III Ridgely, TN 36.3 89.5 340. 1979 Feb 3 Ridgely, TN 36.3 89.5 341. 1979 Feb 5 IV Blytheville, AK 35.9 89.9 342. 1979 Feb 27 Pine Bluff, AK 34.2 90.2 343. 1979 Jun 11 IV near Marston, MO 36.2 89.7 344. 1979 Jun 25 IV Marked Tree, AK 35.5 90.4 345. 1979 Jul 8 IV Wyatt, MO 36.9 89.2 346. 1979 Jul 13 IV near Campbell, MO 36.1 89.8 347. 1979 Aug 13 V near Cleveland, TN 35.2 84.4 348. 1979 Sep 12 V Maryville, TN 35.8 84.0 349. 1980 Apr 21 Maryville, TN 35.8 84.0 350. 1980 Jun 25 IV Maryville, TN 35.8 84.0 351. 1980 Jul 5 IV New Madrid, MO 36.6 89.5 352. 1980 Jul 12 III near Horse Branch, KY 37.3 87.0 353. 1980 Dec 2 V Ridgely, TN 36.3 89.5

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I. TOPOGRAPHY TAKENi FROM KELSH TWO FOOT CONTOUR INTERVAL TOPOGRAPHY OATEO FEBRUARY I, 1977. D 2 ALABAMA "Wc$T" COORDINATES ARc INDICATED IN THE MARGIN

                                       ............... .J            Ir/

I rtHttt--r\---j-------11--j-------lc;f-----t,,' lLlrW A/1£1/ 3. LAYOUT A GRID, WITH STAKES ON 25 FOOT CcNTcRS, IN TH,: LLRW ANO AFR ARcAS, OR/o:NT THc GRID WITH THc BASE L/No:S FOR THc LLRW ARcA. GfflO STAKE~ ARE TO I HAVE COORDINATES, WITH RESPECT TO THE *LLRW AREA II 'I BASE L/No:S, AND GRADE ELEVATION

4. THE GEOLOGIST WILL DETERMINE THE, SEQUENCE OF DRILLING

                                                                ,1                                                                  I                                                                                                                                          FOR THE ROCK INVESTIGATION, DEPTH DF THE CORE BORINGS,

_/*' - __-, AND LOCATION OF THE PERCUSSION BORINGS IN THE LLRW AREA. E \ I - -- )

5. ALL SOIL BORINGS ARE. TO BE MADE uS/NG DRY DRILLING TECHNIQUES.

6. FOR BORING LOCATION PURf'OSf:S EACH LONG STORAGE E

                                                                                                                                              I                                                                                                                              STRUCTURE IN THE LLRW AREA SHOULD BE ASSUMED TO BE 250' LONG.                            .
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T* A STANDARD PENETRATION SPLIT-SPOON BORING IS NEEDED NEAR PLANT GRID COORDINATES O*OON AND 3+00W. THE

                                                                                                                                  ,                                                                                                                                            BOif/NG IS FOR A FUTURE LOW LEVEL RAOWASTE VOLUME
                                                                                                                               ,/                                                                                                                             ; I 11!!         REDUCTION BUILDING. SAMPLING IS TO BE CONT/NOUS TO
                                                                                                                                                                                                                                                 -"-::~'d-/\l-111;18.
                                                                                                                                                                                                                                                                   .*          REF,USAL.A PRECISE LOCATION WILL BE PROVIDED LATER THIS BORING SHOULD BE MADF AFTER BALANCE OF FIELD WORK IS COMPLETED ON OTHER BORINGS.
                                                                                                                                                                                                                                                                          ~

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                                                                                                                                                                                                                                                          'El'."aa.:i=::::1'r,;;;:;;;,:aJ'f'°,..,

11/lff/ff IIIHrHI, BROWNS FERRY NUCLEAR PLANT r;,,,,**** FINAL SAFETY ANALYSIS REPORT Drill Hole Locations and Core Drill Summary FIGURE 2.5-4

uo--------------.,-.,-,.-,---------------r--;;.:;,,:.,,;,,;::;,,,-.,: -:.,-:,,-,.,--~---.-,.-.,,.-,--.~,.-.,-,.-,--------'-----110

                                                                ,.,.._,,11.,                                                           -r*. 1 *                .                                            *r.A.111.r                                   *r:tt11r.o      .,. 11*111,1      -r.,u11,,
      $/I ---.-c,_cc_,cc,.,7,....,.,~..,,,,,..,,...,.--,..-,.=::::..-a---=,;.:"* ,;,:.,,-.----,jF"""'"-~*"""'""~""1-il!----l!l---'5----:.,.:Mc.-:,.-:,.-T-1                                                                      *. 1-.-,.-,.--i!t----M---*---                                      616 4

ca:- -u. C<. l,fl. ~:;~-4 i Or *l,II, tJ* *l.1'. tJ* *l. .t: ,:- *l,II, C* *I.If, l* *I.It

      .,. --&---~,.---.-,_- ,...,~li-.ca*-=*--li-----111'----ii----11-'.,_-<,....,.-f-i!l-----/ il---*---l!---3/4---ii----!;i---------*16 C*      *I.If,
      ***--*----1---------a----f!!l----11t----;1---1--:-1....---m--------sa-----111'---iil----l!-------*.,
      ... --s----1s------11--------11---*-----1a-----.--tl---- ---------------------------***

0 , - - - - - - - - - - - - - - " - - - l ! i - - - - - - - - ; . -'- - - - , - - - - - - - - -- - - - - - - - - - - - - ,.. 00 l*ll.llt """ , .., , " '

      ... ------------------~- - ----------------------------$10
                                                                             ,.,,.IN O*#I
      ... -----------------------------------------------IIF"C.C",s*'<:""-H*
          --.,~.,=,.~,-~,,..=,~-,-=-~,,.,,,.a-,~.,=--,..,..,.cc,.c,--:::.,:::,.:--:,:----::,."',..,*-=---:,,:,,_:c,:----,,c,,,c..,::a,
                                                                                                                 ~~*            O*HI            l"-HN
                                                                                                                                                          .--;-,.-,,cs,,:,-11-:-.--,.,,,..,.c:1 --,:,.,.,,cc,--"',.""",--.:,...,.,.,--~,.~,.,,,=--~,~.,.=,--~,~,,~.,=-- oo I,...,,, ,,.11.,N l*ION                                 ,.,,.,11 -!'"

c,.-

                                                                                                                                                                                                                                                   ,.,.,,,          0*40,,,
                                                                                                                                                                                                                                                                              ,i.*'*   (#>IOI
                                                                                                                                                                                                                                                                                        ,.n,.,*,.""*
                                                                                                                                                                                                                                                                            *I.ti,* IH,f1wi,,.,n,. S** .tlf#,
                                                                                                                                                                                                                                                                            *C:£.*f"* /u1~**M*r.r11,, ..~,,1,,11,
                                                                                                                                                                                                                                          -1.11.
                                                                                                                                                                                                                                                                            . '******'- ,,-IN.
                                                                                                                                                                                                                                                                            - ,,,, '91~ ,.~*Hu.

I

            ----------------------,-------..,.--~------11 ---...e----;11--~u*
                                                                                                                                                                                                                                                                            ,_H,,tt,-,tfl,_,_IWf,lwlll,.,1
                                                                                                                                                                                                                                                                            ..,.,,,,_,,.,.,.,.,.,,,w,,,.~
            ------------------------------'---------i!---3/4---llf----Sl6                                                                                                                                                                                                     """,.,..,
      ... ---------------------------------------e- ---;a----111---610
      ,,. ---~-,.-;_~-,,-,--\'.:-,..-,.~-,.-,---c-'\'.-,,-,,:-*-- ,;::-,;-_:-**-*-=='("",c.,:;;,.',:.,--,;,:-,-:!~'.,-.,--\:-,-:.i-,~,.-,--'(-;-::~...,.,-_-,-==-".:"':.;*!.;;,._-;-,,~~:",'-_,.._.,____________ ,,,                                                         AMENDMENT 16
      ,,. --*---il----i!l---lil----ll----il----ill-----!!--i--ll-- '-'-,!!--------------11*
      ,., ---111----ls-- - - - - - - i i - - - - l l - - - - A - - - - i ~ - - - - ! i l - - i - - S - - - - " ' - - - - - - - - - - - - - - * * *
      *** ---;         11; - - - ,1,:.11 1 ,;:;4.,                , - - - :1:
                                      ;;;1                ,c:,.;;;

1- ' - - ; 1, - - - - ;1 6 ,::;,_;;:;  ;,c.,.,.,:----,

                                                                                                 /  1111 1---;

1::,~:,_,;---;:,.:;.,.,,- 10;1. 6 ,:;;  :--*i-; I. 1.;;;,,;;,c---; 1*1111 1:;.,;,_,;---,:::.:;;,_;;;,--1;-:,:;;,.:::,c---;:,_,a,_;-:,:---~IOO l IOM ,..,,.,, BROWNS FERRY NUCLEAR PLANT

                , ..1111          **40#                OflON                 OflOI                               ** ,,,,,        ,~,ON           l*UN                                                   ** ,, .. , ,   ,.,,:,.,                                         FINAL SAFETY ANALYSIS REPORT

'\ Graphic Logs of Core Drill Holes FIGURE 2.5-5

\ AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT MISCELLANEOUS FOUNDATION INVESTIGATIONS 1972 TO 1980 FIGURE 2.5-5a

                       --z.----                      ~
                                      .......,       II)
                                                     ...iii Cl) g              l?

J! ii 1/J (II a: BFI To H U.S. ~ b1 Sia l_________

                   ,tbr9Abull
                                              .l-52+ 33.!il  'BFZ                                                                                 ABUT 2 PIER B ABUT I PLAN Not to Scolt brg abut or l pier

t. nlwy 5 l/2°cune

• AMENDMENT 16

                                                            ~       _,__L                      BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT TanQenl to trdwy at l brg                                            ACCESS HIGHWAY BRIDGE or l pier _ _ __.-

DRILL LAYOUT FIGURE 2.5-5b TYP LAYOUT

TVA 4215A (WCP,8-71) Page 1 of l GEOLOGIC RECORD OF DRILL HOLE PROJECT BROWN~ EERBY BRIDCE Hole Number Location Geologic Formation

  ~i;--T                                                                          Ft. Pavne Elevation of Surface                                 Exploration              Elevation of Water Loss r:: 70 &.

Abutment tll None Ete*,at1on Top ot SedrO<;k Thickness of Overburden Elevation of Water Gained c;. 1. 1 n Elevation Bottom of Hole Ml F. Size of Core None Driller I Collins c;?c; ? Recommended Foundation Grade NY-*.ri ri> l i ,,.,. Bottom of Weathering Encountered c; i:n 7 Date Started 4/25/72 I Date Completed Elevation Depth Thickness of From ot Material Stratum Surface Stratum Dip Description 0 t7:ERBURDE ~ Red clav 579.6 0.0 ].0 I Grav clav 576.6 3,0 6.0 I Yellow clav 570.6 9,0 12.0 I I Yellow clay and I chert 558.6 - 21.0 27.6 I l1bCK DRU LING Light gray, tine grain, oroiceq ana Limestone 531.0 48.6 5,8 ground to 530.7; calcite vug at Sll:S. 7, ::>ltl.], )l),l I 525.2 54.4  ! BOTTOM OF HOLE I i I I BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT 'NJt!!r teH : ACCESS HIGHWAY BRIDGE Eleva!ron G.P.M. P.S.J. HOLE BF-1 FICURE 2.5-5c Fot' location of boring, see fi ure 2.5-5b. AMENDMENT 1 6 Logged By A. D. Soderberg

TVA 4215A (WCP-8-71) Page 1 of 1 GEOLOGIC RECORD OF DRILL HOLE PROJECT BRO,WNS_EJIB.RL.8R!!)GF. ____ 1-lole Number Location Geologic Formation Ft. Pavne I!

 'B'F-2 t:.lcva 11011 of Surface                               Exploration             Elevation of Water Loss                  !

57~ 8 Elevation Top of Bedrock Abutm"'"~ l,Z Thickness of Overburden Elevation of Water Gilined I I

 '1'H 7                 .

Elevation Bottom of Hole 47.l Size of Core Driller

                                                                                                                         )

I

 <;?.L4 Recommended Foundation Grade Nx-wireline Bottom of Weathering Encountered Collins Date Started           I4/28/72 Date Completed I
 ---                                     S3L7                                   4/28/72 i

Elevation Depth Thickness of From of Material Stratum Surface Stratum Olp Oescriotion CJ IJERBURDE N Red and yellow clav 578,8 o.o 3.0 Clay and chert 575.8 3.0 44.1 l BOCK DRU LING I Light gray, fine grain, calcite I Limestone 531.7 47.1 7.3 vug at .528.8, 527.8~ 5274., 525.3; I 0.1' clay seam at 527.2 I II BOTTOM OF ROLE 524.4 54.4 r

                                                                                                                       \

i I I I i I I . . BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT ACCESS HIGHWAY BRIDCE W:iter !est: HOLE BF-2 Elev.1t10n- G.P.M. P.S.I. FIGURE 2.5-Sd For location of boring, see tigure 2.5-5b. AMENDMENT 16 Logged By A. o. Soderber~

TVA 4,215A (WCP*8*71) P.ige 1 of 1 GEOLOGIC RECORD OF DRILL HOLE PROJECT BROWNS FERRY BRIDGE Hofe Number Loc11tion Geologic Formation i BF-3 Ft. Pavne I E1evation of Surface Exploration Elevatlon of Water Los~ I 579.5 Pier A N'nne I Elevation Top of Bedrock Thickness of Ovl!rburden Elevation of Water Gained 531.8 . Elevation Bottom of Hofe 47.7 Size of Core None Driller I Nx-wireline Anr-lerson j 515.2 Recommended Foundation Grade Bottom of Weathering Encountered Date St~rted I 4/25/72 Date Completed j

 ----                                    531. 8                                  al?r:,/7?                               i i

Elev at Ion Depth Thickness of From of Material Stratum Surface Stratum Dip Description 0 7ERBURDE ~ Red elav 579.5 o.o 3.0 - j Brown clav 576.5 3.0 2.0 Grav clav 574.5 s.o 5.0 Brown clay and I boulders 569.5 10.0 37.7 ' I ROCK DRII .. ING Light co medium gray, tine grain, Limestone 531.8 47.7 16.6 calcite vug at528.S, 528.1, 527.2, 522.8, 51S.n; CJ.l. shaJ.e ac ,;u. u; near vertical joint 522.5 to 521.9 BOTTOM OF HOLE 515.2 64.J I I I I I I BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT ACCESS HlGHW'AY BR[DGE W.:iter test: Ele11atron G,P.M. P,S.I. HOLE BF-3 FIGURE 2.5-5e For location of boring, see figure

                 .,~.5-Sb .            AMENDMENT 16                         Logged By      A. D. Soderber~

TVA 4215A (WCP-8*71) Page 1 of l GEOLOGIC RECORD OF DRILL HOLE PROJECT _ _ _B_R_O_WN_S_FE_*_R_R_Y_B_R_I_D_G_E_ __ Hole Number .Location Geologic Formation  : U'C'_/, Ft Pavne ' Elevation of Surface Exploration Elevation of Water Loss Pier A i

 <;7() 6.                                                                           Mn"'<>                                     '

Elevation Top of BeClrock Thickness of O11er0urden Elevation of Water Gained

                                        /.~    Q                                    N'nni:>                                    i Elevation Bottom of Hole               Si~e Qf Core                                Driller r,.,,,-{nc:,                              !

Recommended Foundation Grade 1'.Tv-t,1-{..-<>1 -t.-.o Bottom of We,llhering Encountered

                                        .,,., 1 Date Started
                                                                                    "- /'>7 /72 IOate Completed 4/?.7 /72 I

II t Elevation Depth Thickness j Material of Stratum From Surface of Stratum Dip I Description I I I Cl!/ERBURDI N  !

 '(ellow and red clav                          579.4          o.o                2.0                                                         II Yellow clay       &  chert    577.4           2.0            44.9                                                            t 1
                                                        ~CK DRII LING                                                        I Light gray, Ei ne grain, weatherea              'i Limestone                     532.S        46.9               lS.6          to 532.3; calcite vug at 531.8,                 !

I 529.IJ, SL.1:1.0; o. l clay seam at I 529.4; weathered parting 519.8; I r cherty 518.5 to 518.l i j I BOTTOM OF HOLE 516.9 62. 5 I i i' I  ! I I I I I I I I i BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT ACCESS HIGHWAY BRIDGE W.iter test: Etev.it,on G.P.M. P.S.I. HOLE BF-4 FIGURE 2.5-5f For location of boring, see figure 2.5-Sb. AMENDMENT 16 Logged Sy A. D. Soderberg

TVA 4215A {WCP*B-71) Page 1 of 1 GEOLOGIC RECORD OF DRILL HOLE PROJECT BROWNS FERRY BRIDGE Hole Number Location Geologic Formation  ! BF-5 Ft. Pavne I Elevation of Surface Exploration Elevation of Water' Loss I 578.7 Pi?r A None i Elevation Top of Bedrock ThIct<ness of Overburden Elevation of Water-Gained  ! 53:;:,2 46. 5 None  ! Elcv.:it1on Bottom of Hole Size of Core ormer i 516.6 ~x-wirel-lne Cnllini-: I Recommended Foundation Grade Bottom of Weathering Encountered Date Started J Date Completed I I 532 _2 1,./27 /72 4/27172 l i Elev.it ion Depth Thickness  ! Material of Stratum Fro~ Surface of Stralurn Dip Desc:rlotion  ! CivERBURDI N Clay and chert 578.7 0.0 46.5 tll"lrv n'DTT T T1'rr - II Light gray, fine grain, calcite vugi T l ~ -- C: 'l., ., I . I! cc , C: ,t! _ .. C:'11 ., C::7'1 c:. ",, ,,.., -**. ~8-- i I 527.7 i j I

 ;i.,T--,P: n'll' t.lr"IT i::'

0::1 -- ~ .:.? I

                                                                                                                                 ,                    I I

i r*- ' j II I Ii i BROWNS FERRY NUCLEAR PLANT r FINAL SAFETY i l ~EMA fl KS : ANALYSIS REPORT ACCESS HfGHWAY BRIDGE Water test: HOLE BF-5 Elevation G.P.M. P.S , I, FIGURE 2,5-Sg For location of boring, i ---- see figure 2.,-Sb. AMENDMENT 16

                          ~

Logged By A. D. Soderberg

TVA 4215A (WCP-8-71} Page 1 of GEOLOGIC RECORD OF DRILL HOLE PROJECT _ _ _ _ _ _ _ _ _ _ _ _ _ __ HojoNumber .Location Geologic Formation BF-6 Ft. u ......, .. Elevatron ot Surface EXploration Elevation ol Water Loss 579.6 p-r.,._r B v~-.,. Elevation Top of Bedrock Thickness of Overburden Elevation ot Water Gained II 540.8 48.8 Nnne I Elevation Bottom of Hole Si.ze of Core Driller I i 515.2 Nx-wireline Recommended Foundation Grade 540.8 Bottom of Weatnering Encountered 526.1 r,.,11*n~ Date Started 4/24/72 IDate Completed 4/?/, /7? I I I Ele11.1tion Depth Thickness Material Stratum of From Surface of Stratum Dip Description II r Ul:D'RT!'Dl\t :.,

  'D .. rl ,.. , ....              C.70     ~              n n           ' n r:r::i.v rTnu                   i;n ,;                   c: n         ~    n I

,.,.,,.11, .... ,.., .... r:~A ~ ,, n ., n lh.rl *"- nal 1-** ,..1,,., C:t:,T /!. 1R n 17 n Brown clay and

  .. --..1                        <;/, I,  &;          -:i:c: n       1 'l   R I                                                                1:nrv      m,n     r T'lll~                                                               I LimestonP                       i::.t.n  R           4R -~          1 i:; ,:;

Medium gray, fine grain, weathered

                                                                                            "'~,.f<'i ".,. c;; 1n c;;   .... .,.,., ..  <;?~ ~            I weathered parting 526.1 1 near                                I 1,,:,,-t-ir,.,1 ini"t- *;?c; j<; l"'n 'i?'i a 524.7 to 524.3, cherty 525.8 to                               !

525.3 I 1 BOTTOM OF l:IOLE 515.2 fi4. 4 I I I I;

                                                                                                           .                                           ...l BROWNS FERRY NUCLEAR PLANT                                 l REMARKS:

FINAL SAFETY ANALYSIS REPORT J Water test: ACCESS HIGHWAY BRIDGE Elevation G.P.M. P.S.I . HOLE BF-6 FIGURE 2.5-Sh For location of boring, see figure 2,5-Sb. AMENDMENT 16 Logg&d By A n, sggerherg

TVA 4215A (WCP-8-71) Page l of 1 GEOLOGIC RECORD OF ORJLL HOLE PROJECT __B_R_O_WN_S_F_F._.R_R_Y_B_R_ID~r_;E_*_ __ HoJe Number Location Geologic Formation* BF-7 .,.. t- 1';,vnR Elevation of*Surface Exploration Elevation of Water Loss 579.8 Pier B Nnm* Elevation Top of Bedrock Thickness of Overburden Elevation of Water Gained I 532.4 47.4 l'Jnn,a I Elevation Bottom of Hole 516.0 Recommended Foundation Grade Size of Core Nx-wireline Bottom of weathering Encountered Driller Anderson Date Started I I 531.l t../?7 /72 Date Completed l.l'-7 /72 I I Elevation Depth Thickness of From of Material Stratum Surface Stratum Dip Descrlotion (~ER.BURDE ~ Red clav 579.8 o.o 2.0 Brown clav 577.8 2.0 3.0 Yel1rn.r clav 574.8 5.0 2.0 Brown clay and boulders 572.8 7. 0 FOCK DRI1 LING Light gray, fine grain; weach~red Limestone 532.4 47.4 6.1 to 531. l; calcite vug at 527.9, 527.7; near verticai Joint ~l~.4; clay seam 528.1 Medium gray, nne grain witn Limestone 526.3 53.5 10.3 scattered shale BOTTOM Of HOLE 516.0 63.8 BROWNS FERRY NUCLEAR PLANT i'INAL SAFETY FIEMAAKS: ANALYSIS REPORT Water test: Elevation G.P,M. P.S.I. ACCESS HIGHWAY BRIDGE HOLE BF-7 FIGURE 2.5-Si. For locaeion of boring, see figure 2.5-Sb. AMENDMENT 16 Logged By A, D, Soderber~

TVA 4215A (WCP-8-71) Page l of 1 GEOLOGIC RECORD OF DRILL HOL.E PROJECT BROWNS FERRY BRIDGE Hole Numtier Location Geologic Formation BF-8 Ft, Payne Elevation of Surface Exploration Elevation or Water Loss 579.l Pier B None Elevation Top ol Bedrock Thickness of Overburden Elevation of Water Gained 1 48.0 None II 531. l Elevation Bottom of Hole Size of Core Driller 516.0 Nx-wireline Bottom of Weathering Encountered Anderson Date Started I4/27/72 I Recommended Foundation G,ade 530.8 4/27/72 Date Completea II I Elevation Depth Thickne55 I of From of Material Stratum Surface Stratum Oio Oescrlotlon

                                                                .._,.,.*u.._   ...,

t)-..J _, -- ~7() 1 n n  ? n

 'Rrnun ,.1 .....                  l:i77 1            ?  n                'l   t.

r.~.,... ,.. , ..... ,;_'7-t ,;_ .. i:. 'l n Red clay and  !

 .. .. ,.r1 ......
   ~                                ,;.-,n ,;_        ~  f:.           '10     li I

t "'lr'Tr nT.1TT .nrr. Light gray, fine grain, weath~red T.f ...,.., .. ,.,".,. ,:; 1,, , &Rn , ,:; - 1 t-n c;-:i.n R! ,...,,l,..i l"e VU"' at 529.6. 528.6; calcite on near vertical

                                                                                           ;oint 525.9~ clav seam at 526.4.                   !

526.0; calcite on weathered near vertical ioint 517.1 to 516.9 j I BOTT0!-1 OF HOLE 516.0 I 63.1 I iI II II I I I BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMA'AKS: ANALYSIS REPORT W.iter test: ACCESS HIGHWAY BRIDGE Elevation G.P.M. P.S.I, HOLE BF-8 FIGURE 2. S-5j For loc3tion of boring, see figure 2.5-Sb. AMENDMENT 16 Logged By A D Scderhe'"i

TVA 4215A (WCP*8*71) Page 1 of 1 GEOLOGIC RECORD OF DRILL HOLE PROJECT BROWNS FERRY RRIDGE Hole Number LOC3tior) Geologic Formation BF-9 Ft. Payne Elevation of Surface Exploration Elevation of Water Loss 579.0 Abutment 112 None I Eiavatlon Top of Bedrock Thickness of Overburden Elevation ot Water Gained 531.5 Elevation Bottom of Hole 47.5 Size of Core None . Driller I 516.5 Recommended Foundation Grade Nx-wireline Bottom of Weathering Encountered Anderson Date Started I4/25/72 Date Completed 531.5 519.0 4/24/72 Elevation Depth Thickness of From of Material Stratum Surface Stratum Oip Descrlotlon

                                                         .       u
!) - .J    1 ---

• o:;70 n n n ., n A

Brown ,.., ....  ? n 1 fl

                               " n r.,-..,o   ,-1 ~**             c;7,:, 0      1 (I            'l  n Brown clay and hni,1Aol"~                     ,;71   n      r:. n         b.1   c; l1 lf'l( nllTT  .nlr.

Medium gray, fine grain, near l' ~-..o,i::f"n"o ._.H c; h.7 c; 1 c; n **~-.. .;,.. .. , ;n{..,t- t:;":11 t:; t-n c;-:i, n, I calcite filled vuggy at 530.5; I c-l;i.v .,,.,,.m i27.7 I" 'i27.l* rha.rt 524.1 to 523.3; weathered parting

                                                                              ,;10 n BOTTOM OF HOLE                 516-5       62.5 I

BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT W.:iter test: ACCESS HIGHWAY BRIDGE Elevation G,P.M, P,5.1. HOLE BF-9 FIGURE 2.5-5k For location of boring, see figure 2.5-Sb. AMENDMENT 16 Logged By A D Soderheri;:

TVA 4215A (WCP,8-71) Page _1_ of_1__ GEOLOGIC RECORD OF DRILL HOLE PROJECT BROWNS FERRY BRIDGE Hole Number Location Geologic: Formation BF-10 tlevatlon of Surface Exploration 1,'~

                                                                                                     'P---

ElevaUori of Water Leiss 579,75 A h01 ---- .- 11 'J 1'T--- Elevation Top of Bedrock Thickness of Overburden Elevation of Water Gained 527.8 'i? .0 :Mn.. ,.. Elevation Bottom of Hole Size of Core Driller

                                                 ..,. 'T 'f " "

522.1 Recommended Foundation Grade r-lv-*-* Bottom of Weathering Encountered

                                       <;?7 SI A .. A ... - - - -

Date Started I, /')g /7'1 IDate Completed I Elevation Oeptti Thickness of Frorn of Material Stratum Surface Stratum Dip Descrlotlon 07ERBURDEIJ Red clay 579.8 o.o 3.0 Gray clay 576.8 3.0 6.0 Brown clay 570.8 9.0 41.0 Gray clay and boulders 529.8 so.a 2.0 ROCK DRIL :..ING Limestone 527.8 52.0 5.7 Liitht erav. fine strain BOTTOM OF HOLE 522.l 57.7 1 i I I BROWNS FERRY NUCLEAR PLANT FINAL SAFETY REMARKS: ANALYSIS REPORT Water test: Elevation G.P.M. P.S.I. ACCESS HIGHWAY BRIDGE HOLE BF-10 For location of boring, FIGURE 2. 5- S1 see fi g ure 2.5-Sb. AMENDMENT 16 Logged By A. D, Soderberg

BFN-21 Figures 2.5-5m through 2.5-5aj (Deleted by Amendment 13)

B~OWNS *ERRY PLANT

                 ~IST0RICqL E~RTHOUR~E ~AP CEPICE~TEQS THROLlGH 19801 Z00 ~CLE RRDIUS ~~OUNO 97 11 "~ON                       J4 . ll    ~ LRT l-MO Cl KY         I CJ Cl ISi 0               0 BROWNS C

FERRY SITE_.,

                                                                   ;;r   g I:!

MS AL GA Cl 0 CI 0 31°30' 910 la'g0°' 0'89° 0'aa 0 0'8? 0 0'86° 0'85° 0'84° 0' STFITllTt 111L£.S

                                                                        =I 9           *II        90     li!B SCl=!LE::     8 HI                 I LEGE:ND :

CltNTtNStiY <* lV D tNTE:NS trY .. I/ CJ PHE~S !TV .. v l AMENDMENT 16 OtNTE:NSITY. '/Ii: 0INT£~SITY

• VIII BROWNS FERRY NUCLEAR PLANT OrNTE~srrv ,. tx FINAL SAFETY ANAL YSJS REPORT

 ~   "1UL TIPLE: E:VE~TS NOTE:

• HISTORICAL EAJlTHQUAll MAP FIGURE 2.5-6

THE MODIFIED MERCALI INTENSITY SCALE OF 1931 considerable in pnorlv built or badly designed (ABRIDGED)* structures; ,ome chimneys broken. Noticed by persons driving motorcars. IV Ill Rossi-Forel Scale.) I. Not felt except by a vary few under 9pecially VI 11. Damage slight in specially designad structures; favorable circumstances. (I Rossi-Forti Scale.I considerable in ordinilfV subatantial buildings with II. Felt only by a few parsons at rest, aspecially on partial collapse; great in poorly built structures. upper floors of buildings. Oelicatalv suspend<<! Panel walls thrown out of frame structures. Fall of objects may swing. II to 11 Rossi*Foral Scale. I chimneys. factory stacks. columns, monuments, Ill. F11lt qulto noticeably ind(lorg, especially nn upptJr walls. H1111vy l11rni1111r. ovftl'turned. Sand and mud flt,or, of buildings, but 111eny people do not fljuct11d m 1inall 111m111111~. Ctum1195 in wall watlll' . raCOQnize it as an eartllquake. Standing mo.torcnrs Ptirsom. driving nKJlorcars diliturb!'ll. (VIII+ to IX*-* mav rock slightly. Vibration like pas.sing of truck . Rossi-Forel Scale.I Duration estimatod.1111 Rossi-Forel ~le .) IX. Damage considerable in specially designed frame IV. During the day felt indoors by 1Y1any, outdoors by structures; well-doslgnod frame structuras thrown few. At night some awakened. Dishes. windows, out of plumb: great In substantial buildlttgs, with doors disturbed; walls make Cfllaking sound. partial collapse. Buildings shifted off foundations. Sensation like heavy truck miking building. Ground cracked conspicuoutly. (IX+ Roai-Forel Standing motorcars rocked noticeably. (IV to V Scale.) Rossi-Forel Scala.) X. Some well-built wooden structures destroyed: most V. Falt by nearly avervone, many awakenl!d. Some ma$0nry and 1rame structures destroyed with dlshn, wind~. etc,. broken; a few instancas of foundations; ground badly cracked. Rails bent. cracked plaster; unltlble objects overturned. Landslides considerable from riverbanks and steep Disturbances of trees, poles, and other tall objects 11op&9. Shihed sand and mud. Watar splt111hed 10metlr"1H noticed. Pendulum clod<1 m*v stop. IV (slopped) nver banks. (X Rossi*Forel Scale.) to VI Roul*Forel Scale.t XI. Few, if any. (moa.omy) structures remain standing. VI. Falt hy 11II, ,mr,y frightened and ru11 outdoon . 8 r idnes d eatrovlld. Rrolld flssur11s in grour11'1. Soma hflllVV furnlturn movllll; a faw in-1ant:a11 of \J111lnr11ro1111il tli11nlinn1 cnmplntnly out of M1rvicn. f11U11n µlastar or demeged c:himney1. Darnaga slight. l:.llllh ~lump~ ,md hmtl *hJl!i m r.ofl ground . Rail~ 4v1 to VII Rossi*For11I Scala.) bent greatly. VII. Evervbody runs outdoors. Damage negligible In XU. Damage total. Waves seen on ground surfaces. Lines buildings of good design and construction; $light to of ,ight and level distorted. Objects thrown upward moderate In wall*built ordinary structures; into air. AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT MODIFIED HERCALLI INTENSITY SCALE OF 1931 (ABRIDGED) FtGURF. 2.5-7

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SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 2.5-17

BFN-16 Figure 2.5-18 (Deleted)

SECURITY-RELATED INFORMATION, WITHHELD UNDER 10CFR2.390 Figure 2.5-19

BFN-17 2.6 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM 2.6.1 General The preoperational environmental radiological monitoring program established a baseline of data on the distribution of natural and manmade radioactivity in the environment near the plant site. By comparing data from the operational program with this background information and with data from control monitoring stations, accumulation or buildup of radioactivity in the environment can be identified. Sample collection and analysis was initiated in April 1968 and will continue indefinitely. TVA participates in an Interlaboratory Comparison Program. Samples supplied by the comparison laboratory are analyzed by TVA and the results submitted to the laboratory for comparison. Reports describing the results of the environmental radiological monitoring activities are submitted routinely to the Nuclear Regulatory Commission as required by the plant Technical Specifications. 2.6.2 Monitoring Program A general discussion of the environmental radiological monitoring program follows. More specific information can be found in the BFN Offsite Dose Calculation Manual. 2.6.2.1 Atmospheric Monitoring The atmospheric monitoring network is divided into three subgroups. Local air monitors are located on or adjacent to the plant site. Perimeter monitors are located in areas of high population density out to approximately 10 miles from the plant. Remote air monitors are located at distances greater than 15 miles. Each monitor is capable of continuously sampling air through a particulate filter. In series with, but downstream of the particulate filter, is a charcoal filter used to collect iodine. 2.6.2.2 Terrestrial Monitoring External gamma radiation levels are monitored at selected locations near the site boundary in different sectors around the plant. 2.6-1

BFN-17 Samples of fresh milk are collected routinely from selected dairy farms in the vicinity of the plant. In addition, vegetation grown near the plant is sampled during the growing season. Municipal water systems downstream from the plant as specified in the BFN Offsite Dose Calculation Manual are also sampled on a routine basis. Six groundwater monitoring wells near the Low-Level Radwaste Storage Area were sampled quarterly for one year (9/82 to 6/83) to obtain baseline radiological data. The wells are all hydrologically downgradient from the storage area. Parameters analyzed were: gamma scan, gross beta count, strontium 89 and 90, and tritium. Results show very low background levels of radioactivity. 2.6.2.3 Reservoir Monitoring Reservoir water samples are collected from sampling locations upstream and downstream from the plant. In addition, samples of sediment and biological media are also taken from the reservoir in the vicinity of the plant. Due to the expected low concentrations of radioisotopes in the plant effluent and the dilution provided by the streamflow past the plant site, it is expected that the radioactivity levels in the reservoir will be well below the limits established in Plant Technical Specifications. Radioactivity levels in the river at the nearest downstream water supply intake resulting from accidental slug release are expected to be well below the effluent concentration limits in 10 CFR 20, particularly since public water supplies are located at considerable distances from the site. Adequate environmental monitoring shall be provided to ensure the potable water supplies will not exceed regulatory limits for radioactivity due to operations at Browns Ferry Nuclear Plant. Both public and private potable water supplies taking water from the Tennessee River downstream from the plant will be monitored periodically. One upstream water supply will also be monitored and used as a control station. 2.6.2.4 Other Monitoring Samples of terrestrial biological specimens may be collected and analyzed at periodic intervals as required to aid in the evaluation of overall radiological control programs. Types and frequencies of such samples shall be determined by TVA. 2.6-2

BFN-29 REACTOR TABLE OF CONTENTS 3.0 REACTOR ................................................................................................................................. 3.1-1 3.1 Summary Description ................................................................................................................ 3.1-1 3.2 Fuel Mechanical Design ............................................................................................................ 3.2-1 3.2.1 Power Generation Objective ...................................................................................... 3.2-1 3.2.2 Power Generation Design Basis ................................................................................ 3.2-2 3.2.3 Safety Design Basis ................................................................................................... 3.2-2 3.2.4 Description ................................................................................................................. 3.2-3 3.2.5 Safety Evaluation ....................................................................................................... 3.2-5 3.2.6 Inspection and Testing ............................................................................................... 3.2-7 3.2.7 References................................................................................................................. 3.2-8 3.3 Reactor Vessel Internals Mechanical Design ............................................................................ 3.3-1 3.3.1 Power Generation Objective ...................................................................................... 3.3-1 3.3.2 Power Generation Design Basis ................................................................................ 3.3-1 3.3.3 Safety Design Basis ................................................................................................... 3.3-1 3.3.4 Description ................................................................................................................. 3.3-2 3.3.5 Safety Evaluation ....................................................................................................... 3.3-8 3.3.6 Inspection and Testing ............................................................................................... 3.3-17 3.4 Reactivity Control Mechanical Design ....................................................................................... 3.4-1 3.4.1 Safety Objective ......................................................................................................... 3.4-1 3.4.2 Power Generation Objective ...................................................................................... 3.4-1 3.4.3 Safety Design Basis ................................................................................................... 3.4-1 3.4.4 Power Generation Design Basis ................................................................................ 3.4-2 3.4.5 Description ................................................................................................................. 3.4-2 3.4.6 Safety Evaluation ....................................................................................................... 3.4-22 3.4.7 Inspection and Testing ............................................................................................... 3.4-31 3.5 Control Rod Drive Housing Supports ........................................................................................ 3.5-1 3.5.1 Safety Objective ......................................................................................................... 3.5-1 3.5.2 Safety Design Basis ................................................................................................... 3.5.1 3.5.3 Description ................................................................................................................. 3.5-1 3.5.4 Safety Evaluation ....................................................................................................... 3.5-3 3.5.5 Inspection and Testing ............................................................................................... 3.5-3 3.6 Nuclear Design .......................................................................................................................... 3.6-1 3.6.1 Power Generation Objective ...................................................................................... 3.6-1 3.6.2 Power Generation Design Basis ................................................................................ 3.6-1 3.6.3 Safety Design Basis ................................................................................................... 3.6-1 3.6.4 Description ................................................................................................................. 3.6-2 3.6.5 Analytical Methods ..................................................................................................... 3.6-7 3.0-i

BFN-29 REACTOR TABLE OF CONTENTS (Cont'd) 3.6.6 Reactivity of Fuel in Storage ...................................................................................... 3.6-7 3.6.7 References................................................................................................................. 3.6-8 3.7 Thermal and Hydraulic Design .................................................................................................. 3.7-1 3.7.1 Power Generation Objective ...................................................................................... 3.7-1 3.7.2 Power Generation Design Basis ................................................................................ 3.7-1 3.7.3 Safety Design Basis ................................................................................................... 3.7-1 3.7.4 Thermal and Hydraulic Limits..................................................................................... 3.7-1 3.7.5 Description of Thermal-Hydraulic Design of the Reactor Core................................... 3.7-2 3.7.6 Description of the Thermal-Hydraulic Design of the Reactor Coolant System ........... 3.7-8 3.7.7 Evaluation .................................................................................................................. 3.7-12 3.7.8 References................................................................................................................. 3.7-25 3.8 Standby Liquid Control System ................................................................................................. 3.8-1 3.8.1 Safety Objective ......................................................................................................... 3.8-1 3.8.2 Safety Design Basis ................................................................................................... 3.8-1 3.8.3 Description (Figures 3.8-1, 3.8-2, 3.8-3, 3.8-5, 3.8-6, and 3.8-8 ................................ 3.8-2 3.8.4 Safety Evaluation ....................................................................................................... 3.8-5 3.8.5 Inspection and Testing ............................................................................................... 3.8-9 3.0-ii

BFN-29 REACTOR LIST OF TABLES Table Title 3.2-1 (Deleted) 3.2-2 (Deleted) 3.2-3 (Deleted 3.2-4 (Deleted) 3.2-5 (Deleted) 3.3-1 Reactor Vessel Internals, Design Data 3.6-1 Definition of Fuel Design Limits 3.0-iii

BFN-29 REACTOR LIST OF FIGURES Figure Title 3.2-1 (Deleted) 3.3-1 Reactor Vessel Internals 3.3-2 Reactor Vessel Internal Flow-Schematic 3.3-3 Steam Separator 3.3-4 Fuel Support Pieces 3.3-5 Jet Pump 3.3-6 Steam Dryer 3.3-7 Reactor Internal Pressure Difference (RIPD) Locations 3.3-8 (Deleted) 3.3-9 Thermal Shock Transient Analysis Zones 3.3-10 Material Behavior Graph-Cycles vs. Stress for Stainless Steel 3.3-11 (Deleted) 3.4-1 Control Rod-Isometric 3.4-2 Control Rod-Isometric 3.4-3 Control Rod to Control Rod Drive Coupling-Isometric 3.4-4 Control Rod Velocity Limiter-Isometric 3.4-5 Control Rod Drive, Simplified Component Illustration 3.4-6 Control Rod Drive, Schematic Diagram 3.4-7 Control Rod Drive Hydraulic Control System, Simplified Component Illustration 3.4-8a sht 1 CRD Hydraulic System - Mechanical Control Diagram 3.4-8a sht 2 Control Rod Drive Hydraulic System - Mechanical Control Diagram 3.4-8a sht 3 Control Rod Drive Hydraulic System - Mechanical Control Diagram 3.4-8a sht 4 Control Rod Drive Hydraulic System - Mechanical Control Diagram 3.4-8a sht 5 Control Rod Drive Hydraulic System - Mechanical Control Diagram 3.4-8b Control Rod Hydraulic System - Flow Diagram 3.4-8c Control Rod Hydraulic System - Flow Diagram 3.4-8d CRD Hydraulic System - Mechanical Control Diagram 3.4-8e Control Rod Drive Hydraulic System - Flow Diagram 3.4-8f Control Rod Drive Hydraulic System - Mechanical Control Diagram 3.4-8g CRD Hydraulic System - Mechanical Control Diagram 3.4-8h CRD Hydraulic System - Mechanical Control Diagram 3.4-9 Control Rod Unit Drive-Cutaway Illustration 3.4-9a Modified Control Rod Drive (BWR/6) 3.4-10 (Deleted) 3.4-11 (Deleted) 3.4-12 Marathon Control Rod - Isometric 3.4-13 Westinghouse CR-82M-1 Control Rod 3.5-1 Control Rod Drive Housing Support-Isometric 3.6-1 (Deleted) 3.6-2 (Deleted) 3.6-3 (Deleted) 3.6-4 (Deleted) 3.0-iv

BFN-29 REACTOR LIST OF FIGURES (Cont'd) Figure Title 3.6-5 (Deleted) 3.6-6 (Deleted) 3.6-7 (Deleted) 3.6-8 (Deleted) 3.6-9 (Deleted) 3.6-10 (Deleted) 3.6-11 (Deleted) 3.6-12 (Deleted) 3.6-13 (Deleted) 3.7-1 Operating Map, Unit 1 3.7-2 Operating Map, Unit 2 3.7-3 Operating Map, Unit 3 3.8-1 Standby Liquid Control System Flow Diagram 3.8-2 Standby Liquid Control System Mechanical Control Diagram 3.8-3 Standby Liquid Control System - Flow Diagram 3.8-4 (Deleted) 3.8-5 Standby Liquid Control System - Flow Diagram 3.8-6 Standby Liquid Control System - Mechanical Control Diagram 3.8-7 (Deleted) 3.8-8 Standby Liquid Control System - Mechanical Control Diagram 3.0-v

BFN-16 3.0 REACTOR 3.1

SUMMARY

DESCRIPTION The subsections included in the "Reactor" section describe and evaluate those systems most pertinent to the fuel barrier and the control of core reactivity. The "Fuel Mechanical Design" subsection describes the mechanical aspects of the fuel material (uranium dioxide), the fuel cladding, the fuel rods, and the arrangement of fuel rods in bundles. Of particular interest is the ability of the fuel to serve as the initial barrier to the release of radioactive material. The mechanical design of the fuel is sufficient to prevent the escape of significant amounts of radioactive material during normal modes of reactor operation. The "Reactor Vessel Internals Mechanical Design" subsection describes both the arrangements of the supporting structure for the core and the reactor vessel internal components which are provided to properly distribute the coolant delivered to the reactor vessel. In addition to their main function of coolant distribution, the reactor vessel internals separate the moisture from the steam leaving the vessel and provide a floodable inner volume inside the reactor vessel that allows sufficient submergence of the core under accident conditions to prevent additional damage to the fuel and the gross release of fission products from the fuel. The reactor vessel internals are designed to allow the control rods and Core Standby Cooling Systems to perform their safety functions during abnormal operational transients and accidents. The "Reactivity Control Mechanical Design" subsection describes the mechanical aspects of the moveable control rods. They are provided to control core reactivity. The Control Rod Drive Hydraulic System is designed so that sufficient energy is available to force the control rods into the core under conditions associated with abnormal operational transients and accidents. Control rod insertion speed is sufficient to prevent fuel damage as a result of any abnormal operational transient. Control Rod Housing Supports are located underneath the reactor vessel near the control rod housings. These supports limit the travel of a control rod in the event that a control rod housing is ruptured. They prevent a significant nuclear excursion as a result of the housing failure, thus protecting the fuel barrier and the primary system. The "Nuclear Design" subsection describes the nuclear aspects of the reactor. The design of the boiling water reactor core and fuel is based on a proper combination of design variables, such as moderator-to-fuel volume ratio, core power density, thermal-hydraulic characteristics, fuel exposure level, nuclear characteristics of the core and fuel, heat transfer, flow distribution, void content, heat flux, and operating pressure. All of these conditions are dynamic functions of operating conditions. 3.1-1

BFN-17 However, design analyses and calculations, verified by comparison with data from operating plants, are performed for specific steady state, transient, and accident conditions. Included in the Nuclear Design subsection are discussions of operating and shutdown reactivity control requirements. Also included are discussions of the reactivity coefficients and xenon characteristics of the core. Transient and accident analyses are discussed in Chapter 14, Plant Safety Analysis. Results of steady state, transient, and accident analyses for current reload core designs are contained in Appendix N. The "Thermal and Hydraulic Design" subsection describes the thermal and hydraulic characteristics of the core. The low coolant saturation temperature, high heat transfer coefficient, and neutral water chemistry of the boiling water reactor are significant advantages in minimizing Zircaloy temperatures and associated temperature-dependent hydride pickup. This results in improved cladding performance at long exposures. The relatively uniform fuel cladding temperatures throughout the boiling water reactor core minimize migration of the hydrides to cold cladding zones and reduce the thermal stresses. A discussion of fuel failure mechanisms and the parameters associated with fuel damage is included in the subsection. The "Standby Liquid Control System" provides a redundant, independent, and different way from the control rods to make the reactor subcritical, even in the cold condition. The Standby Liquid Control System is never expected to be needed because of the large number of control rods available to shut down the reactor. However, in the unlikely event that control rod insertion were to be impaired, the Standby Liquid Control System has the capability of bringing the reactor from rated power to cold shutdown (MODE 4) with the control rods remaining withdrawn in the rated power pattern. 3.1-2

BFN-26 3.2 FUEL MECHANICAL DESIGN The fuel assembly is comprised of the fuel bundle, channel, and channel fastener. The fuel bundle is comprised of fuel rods, water rods (water channels), spacers, upper and lower tie plates, springs, and fittings. Fuel licensing acceptance criteria for General Electric (GE) fuel designs are specified in GESTAR II (General Electric Standard Application For Reactor Fuel) (Reference 1). Amendment 22 of GESTAR II established an approved set of licensing acceptance criteria for which fuel design compliance constitutes USNRC acceptance and approval without specific USNRC review. Current GE fuel designs that have received specific USNRC review and approval, or that have been shown to meet the approved fuel licensing acceptance criteria are documented in Reference 2 (General Electric Fuel Bundle Designs). GE designs documented in Reference 2 and approved for use in Browns Ferry reload cores include the GE13 and GE14 fuel product lines. The initial core 7x7 and 8x8 designs, the reload core unpressurized 8x8R design, and the pressurized P8x8R, BP8x8R, GE9B, GE11 designs that are currently in spent fuel storage will not be reinserted in any future reload cores. Fuel licensing acceptance criteria for AREVA fuel designs are specified in ANF-89-98(P)(A) Revision 1 and Supplement 1. Generic Mechanical Design Criteria for BWR Designs (Reference 4). This document contains an approved set of licensing acceptance criteria for which fuel design compliance constitutes USNRC acceptance and approval without specific USNRC review. For GE fuel, the generic information contained in GESTAR II is supplemented by plant cycle-unique information and analytical results. This cycle-unique information includes a list of the fuel to be loaded in the core and safety analysis results. This information is documented in a separate plant-unique cycle-dependent submittal for each reload. The format for this Supplemental Reload Licensing Report (SRLR) and a description of the transient and accident methods used are given in the country-specific supplement to GESTAR II (Reference 3). For AREVA fuel, a reload-specific fuel mechanical design report is prepared to document compliance with Reference 4 Generic Design Criteria. The Generic Design Criteria lists the approved methodology documents. The fuel mechanical design report is referenced in the applicable cycle-specific Safety Analysis Report. 3.2.1 Power Generation Objective The objective of the nuclear fuel is to provide a high integrity assembly of fissionable material which can be arranged in a critical array. The assembly must be capable of efficiently transferring the generated fission heat to the circulating coolant water while maintaining structural integrity and containing the fission products. 3.2-1

BFN-26 3.2.2 Power Generation Design Basis The nuclear fuel shall be designed to assure (in conjunction with the core nuclear characteristics, the core thermal and hydraulic characteristics, the plant equipment characteristics, and the capability of the nuclear instrumentation and reactor protection system) that fuel damage limits will not be exceeded during either planned operation or abnormal operational transients caused by any single equipment malfunction or single operator error. Limits are established to assure fuel operation remains within design bases. Three types of thermal limits are used. The first is the Minimum Critical Power Ratio Operating Limit. This limit is generated to protect against the phenomena of dryout, where the liquid film on the fuel cladding surface is boiled/stripped away, thereby creating conditions for a rapid rise in cladding temperature due to the elimination of effective boiling heat transfer. The second limit is the Linear heat Generation Rate. This limit protects the fuel rod thermal/mechanical integrity during normal operation, as well as during anticipated operational occurrences. The third limit is the Maximum Average Planar Linear Heat Generation Rate. This limit protects the fuel from exceeding fuel performance requirements identified by Title 10 Code of Federal Regulations Part 50.46. Power and Flow dependent multipliers are applied to all three types of thermal limits to protect operation at off-rated conditions. The basic limit types, along with the power/flow multipliers are generated for all fuel types, regardless of fuel vendor or unit. Power and flow dependent limits (or multipliers) are applied to both the MCPR and thermal-mechanical limits to account for off-rated conditions. The thermal-mechanical limits are protected with off-rated corrections applied directly to the LHGR limits. 3.2.3 Safety Design Bases In meeting the power generation objectives, the nuclear fuel cladding shall be utilized as the initial barrier to the release of fission products. The fission product retention capability of the nuclear fuel shall be substantial during normal modes of reactor operation so that significant amounts of radioactivity are not released from the reactor fuel barrier. For GE fuel, the detailed fuel thermal-mechanical design bases and limits are provided in GESTAR II. For AREVA fuel, the thermal-mechanical design bases and limits are provided in the AREVA Generic Design Criteria document (Reference 4). The design bases address each of the fuel system damage, failure, and coolability criteria identified in the Standard Review Plan (NUREG-0800). 3.2-2

BFN-26 3.2.4 Description A core cell is defined as a control rod and the four fuel assemblies which immediately surround it. Each core cell is associated with a 4-lobed fuel support piece. Around the outer edge of the core, certain fuel assemblies are not immediately adjacent to a control rod and are supported by individual peripheral fuel support pieces. A description of the fuel assembly and various fuel assembly components is provided in the following sub-sections. 3.2.4.1 Fuel Assembly The fuel assembly consists of a fuel bundle and a channel which surrounds it. The fuel bundle contains fuel rods and water rods (or water channel) which are spaced and supported in a square array by upper and lower tieplates, as well as fuel rod spacers. The lower tieplate has a nosepiece which has the function of supporting the fuel assembly in the reactor. The upper tieplate has a handle for transferring the fuel bundle from one location to another. The identifying fuel assembly serial number is engraved on the top of the handle. No two assemblies bear the same serial number. A boss projects from one side of the handle to aid in ensuring proper orientation of the assembly in the core. Finger springs located between the lower tieplate and channel are utilized to control the bypass flow through that flow path. 3.2.4.1.1 Fuel Rods Three types of fuel rods are used in a GE fuel bundle; tie rods, standard rods, and (in some designs) part length rods. The tie rods in each fuel bundle have lower end plugs which thread into the lower tieplate and threaded upper end plugs which extend through the upper tieplate. A nut and locking tab are installed on the upper end plug to hold the fuel bundle together. The tie rods support the weight of the assembly during fuel handling operations. All of the standard rods are full length rods. The part length rods are approximately 2/3 of the length of standard fuel rods. AREVA fuel assemblies contain two basic rod types: standard rods and part length fuel rods (PLFRs). Tie rods are not necessary because the structural tie between the lower and upper tieplate is provided by the water channel. The PLFRs lengths vary by fuel design, relative to standard fuel rods. During operation, the GE assembly is supported by the lower tieplate. The end plugs of the standard rods have shanks which fit into bosses in the tieplates. An expansion spring is located over the upper end plug shank of each rod in the bundle to keep the rods seated in the lower tieplate. 3.2-3

BFN-26 For the AREVA design, the lower ends of the fuel rods rest on top of the lower tieplate grid. The lower ends of the fuel rods are laterally restrained by an additional spacer grid located just above the lower tieplate. No expansion springs are necessary on each fuel rod because a single, large reaction spring is used on the water channel to hold the upper tieplate in the latched position. Each fuel rod contains high density ceramic uranium dioxide fuel pellets stacked within Zircaloy cladding. The fuel rod is evacuated, backfilled with helium, and sealed with end plugs welded into each end. U-235 enrichments may vary from fuel rod to fuel rod within a bundle to reduce local peak-to-average fuel rod power ratios. Selected fuel rods within each bundle may include small amounts of Gadolinia as a burnable poison. Adequate free volume is provided within each fuel rod in the form of a pellet-to-cladding gap and a plenum region. A plenum spring, or retainer, is provided in the plenum space to minimize the movement of the column of fuel pellets inside the fuel rod during shipping and handling. For GE fuel product lines through GE13, a hydrogen getter has historically been provided in the plenum space as assurance against chemical attack from inadvertent admission of moisture or hydrogenous impurities into the fuel rod during manufacture. With enhanced hydrogen controls in place, the optional feature of a reactive getter has been removed for the current GE13 and GE 14 product lines. Likewise, hydrogen and moisture controls eliminate the need for a getter in the AREVA fuel rod design. 3.2.4.1.2 Water Rods or Water Channel For the GE fuel designs, water rods are hollow Zircaloy tubes with several holes around the circumference near each end to allow coolant to flow through. One water rod in each bundle axially positions the spacers. The AREVA design instead uses one larger internal water channel that has a square cross-section in contrast to round water rods. The water channel displaces 3x3 array of fuel rods in the fuel assembly interior. 3.2.4.1.3 Fuel Spacer The primary function of the spacer is to provide lateral support and spacing of the fuel rods, with consideration of thermal-hydraulic performance, fretting wear, strength, neutron economy, and producibility. 3.2.4.1.4 Finger Springs Finger Springs are employed to control the bypass flow through the channel-to-lower tieplate flow path. 3.2-4

BFN-26 3.2.4.1.5 Debris Filter Lower Tie Plate Fuel assemblies may include a debris filter as part of the lower tie plate products designed to reduce the probability of foreign material entering the fuel during normal operation. Both GE and AREVA debris filter designs utilize non-line-of-sight coolant flow paths to maximize the ability to stop and retain foreign material. The debris filter does not play a part in the structural performance of the fuel assembly. Designs are explicitly tested to determine hydraulic performance impacts. 3.2.4.1.6 Channels The BWR Zircaloy fuel channel performs the following functions: (1) Forms the fuel bundle flow path outer periphery for bundle coolant flow. (2) Provides surfaces for control rod guidance in the reactor core. (3) Provides structural stiffness to the fuel bundle during lateral loadings applied from fuel rods through the fuel spacers. (4) Minimizes, in conjunction with the finger springs and bundle lower tieplate, coolant bypass flow at the channel/lower tieplate interface. (5) Transmits fuel assembly seismic loadings to the top guide and fuel support of the core internal structures. (6) Provides a heat sink during loss-of-coolant accident (LOCA). (7) Provides a stagnation envelope for incore fuel sipping. The channel is open at the bottom and makes a sliding seal fit on the lower tieplate surface. The upper end of the fuel assemblies in a four-bundle cell are positioned in the corners of the cell against the top guide beams by the channel fastener springs. At the top of the channel, two diagonally opposite corners have welded tabs supporting the weight of the channel on the threaded raised posts of the upper tieplate. One of these raised posts has a threaded hole. The channel is attached to the fuel bundle using the threaded channel fastener assembly, which also includes the fuel assembly positioning spring. Channel-to-channel spacing is assured by the fuel bundle spacer buttons located on the upper portion of the channel adjacent to the control rod passage area. 3.2.5 Safety Evaluation The GE thermal-mechanical evaluations performed for the fuel are described in GESTAR II. AREVA thermal-mechanical evaluations are described in the Generic Design Criteria (Reference 4). Areas evaluated include: (1) Fuel System Damage -- stress/strain, fatigue, fretting wear, oxidation, hydriding, corrosion, dimensional changes, internal gas pressure, and hydraulic loads. 3.2-5

BFN-26 (2) Fuel Rod Failure -- hydriding, cladding collapse, fretting wear, overheating of cladding, overheating of pellets, excessive fuel enthalpy, pellet-cladding interaction, bursting, and mechanical fracturing. (3) Fuel Coolability -- cladding embrittlement, violent expulsion of fuel, generalized cladding melting, fuel rod ballooning, and structural deformation. 3.2.5.1 Evaluation Methods The GE methods used in performing thermal-mechanical evaluations for the fuel are described in GESTAR II. These evaluations are performed primarily using the NRC-approved GESTR-MECHANICAL fuel rod thermal-mechanical performance model. The GESTR-MECHANICAL fuel rod performance model performs best estimate coupled thermal and mechanical analyses of a fuel rod experiencing a variable operating history. The model explicitly addresses the effects of:

• Fuel and cladding thermal expansion
• Fuel and cladding creep and plasticity
• Cladding irradiation growth
• Cladding irradiation hardening and thermal annealing of that irradiation hardening
• Fuel irradiation swelling
• Fuel irradiation-induced densification
• Fuel cracking and relocation
• Fuel hot pressing
• Fission gas generation and exposure-enhanced fission gas release including fission product helium release
• Differential axial expansion of the fuel and cladding reflecting axial slip or lockup of the fuel pellets with the cladding
• Fuel phase change volumetric expansion upon melting The GESTR-MECHANICAL material properties and component models represent the latest experimental information available.

The fuel rod cladding stress analyses are performed using a Monte Carlo statistical method in conjunction with distortion energy theory. Fuel cladding plasticity analyses are also performed when required by the loading conditions. AREVA methods used in performing the thermal-mechanical evaluations for the fuel are described in the Generic Design Criteria (Reference 4) and by the approved topical reports referenced thereof. These analyses are performed primarily using the NRC-approved RODEX2A fuel rod thermal-mechanical performance code. This code combines best estimate and conservative models, coupled with a specific input methodology to produce conservative results. The code contains a number of different models to address the following phenomena: 3.2-6

BFN-26

• Fuel densification
• Fuel gaseous and solid swelling
• A physically-based fission gas release model coupled to the swelling model
• Columnar grain growth
• Instantaneous plastic deformation of the fuel
• Fuel cracking, crack volume closure and creep deformation
• Fuel pore migration
• Cladding anisotropic creep deformation, irradiation growth
• Cladding corrosion and hydriding
• Thermal-hydraulic conditions of the fuel rod sub-channel The code performs interactive calculations on a time incremental basis with conditions updated at each calculated increment. Cladding strain, fuel and cladding temperatures, fission gas release, rod internal pressure are calculated as well as time and burnup-dependent fuel and cladding properties. In addition, the code is used to establish initial conditions for ramping and accident analyses.

Starting with the ATRIUM-10X fuel design, the NRC has approved use of the newer RODEX4 method at BFN. RODEX4 supports thermal mechanical analysis and development of the LHGR limit. RODEX4 is considered a best estimate method, utilizing a Monte-Carlo analysis process. The older ROCEX2A methodology is still retained to provide conservative input to transient and LOCA analyses. 3.2.5.2 Evaluation Results The thermal-mechanical evaluations described above have been completed by GE for all fuel designs included in Reference 2 (General Electric Fuel Bundle Designs). The evaluations demonstrate that these fuel designs meet the required thermal-mechanical licensing criteria documented in GESTAR II. Similarly, AREVA thermal-mechanical evaluations are described in the applicable fuel mechanical design report prepared for each cycle. These evaluations document compliance of the fuel design to the NRC approved Generic Design Criteria topical report. 3.2.6 Inspection and Testing The GE fuel quality assurance program is described in GESTAR II. The AREVA fuel quality assurance program is described in FQM U.S. Version, Framatome ANP Fuel Sector Quality Manual. The program covers the quality control areas associated with the manufacture and inspection of new fuel for the areas of: (1) Material and component procurement. 3.2-7

BFN-26 (2) Fabrication and assembly of components and systems. (3) Inspection and testing. (4) Cleaning, packaging, and shipping. GE and AREVA also have active programs of interim and post-irradiation surveillance of both lead use assemblies and developmental BWR fuel. The GE program and the inspection techniques used are described in GESTAR II. The AREVA program is described in the referenced reports contained in the applicable fuel mechanical design report. 3.2.7 References

1. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, (See Appendix N for applicable revision).

2. General Electric Fuel Bundle Designs, NEDE-31152P, Rev. 7, June 2000.

3. General Electric Standard Application for Reactor Fuel (Supplement for United States), NEDE-24011-P-A-US, (See Appendix N for applicable revision).

4. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A)

Revision 1 and Supplement 1, Advanced Nuclear Fuels Corp., May 1995. 3.2-8

BFN-16 Table 3.2-1 (Deleted by Amendment No. 16)

BFN-16 Table 3.2-2 (Deleted by Amendment No. 16)

BFN-16 Table 3.2-3 (Deleted by Amendment No. 16)

BFN-16 Table 3.2-4 (Deleted by Amendment No. 16)

BFN-16 Table 3.2-5 (Deleted by Amendment No. 16)

BFN-21 Figure 3.2-1 (Deleted by Amendment 21)

BFN-26 3.3 REACTOR VESSEL INTERNALS MECHANICAL DESIGN 3.3.1 Power Generation Objective Reactor vessel internals (exclusive of fuel, control rods and incore flux monitors) are provided to achieve the following objectives:

a. Maintain partitions between regions within the reactor vessel to provide proper coolant distribution, thereby allowing power operation without fuel damage due to inadequate cooling.

b. Provide positioning and support for the fuel assemblies, control rods, incore flux monitors, and other vessel internals to assure that control rod movement is not impaired.

c. Provide a source of neutrons to assure meaningful nuclear measurements at reactor low power levels.

3.3.2 Power Generation Design Basis

1. The reactor vessel internals shall be designed to provide proper coolant distribution during all anticipated normal operating conditions to allow power operation of the core without fuel damage.

2. The reactor vessel internals shall be arranged to facilitate refueling operations.

3. The reactor vessel internals shall include devices that permit assessment of the core reactivity condition during periods of low power and subcritical operations.

4. Adequate working space and access shall be provided to permit adequate inspection of reactor vessel internals.

3.3.3 Safety Design Basis

1. The reactor vessel internals shall be arranged to provide a floodable volume in which the core can be adequately cooled in the event of a breach in the nuclear system process barrier external to the reactor vessel.

2. Deflections and deformation of reactor vessel internals shall be limited to assure that the control rods and the Core Standby Cooling Systems can perform their safety functions during abnormal operational transients and accidents.

3.3-1

BFN-26

3. The reactor vessel internals mechanical design shall assure that safety design bases 1 and 2 are satisfied in accordance with the loading criteria of Appendix C, so that the safe shutdown of the plant and removal of decay heat are not impaired.

3.3.4 Description The reactor vessel internals are installed inside the reactor vessel to properly distribute the flow of coolant delivered to the vessel, to locate and support the fuel assemblies, and to provide an inner volume containing the core that can be flooded following a break in the nuclear system process barrier external to the reactor vessel. The reactor vessel internals are the following components: Core shroud Shroud head and steam separator assembly Core support (core plate) Top guide Fuel support pieces Control rod guide tubes Jet pump assemblies Steam dryers Feedwater spargers Core spray lines and spargers Vessel head cooling spray nozzle Differential pressure and liquid control line Incore flux monitor guide tubes Startup neutron sources Surveillance sample holders The overall arrangement of the internals within the reactor vessel is shown in Figure 3.3-1. Table 3.3-1 gives detailed design data for the various reactor vessel internals. Although not mandatory, the design of the reactor vessel internals is in accordance with the intent of Section III of the ASME Boiler and Pressure Vessel Code. The material used for most of the fabrication of the reactor vessel internals is solution heat-treated, unstabilized type 304 austenitic stainless steel conforming to ASTM specifications. Weld procedures and welders are qualified in accordance with the intent of Section IX of the ASME Boiler and Pressure Vessel Code. The floodable inner volume of the reactor vessel is shown on Figure 4.3-4. It is the volume inside the core shroud up to the level of the jet pump nozzles. The boundary of the floodable inner volume consists of the following (see Figure 3.3-1):

a. The jet pumps from the jet pump nozzles down to the shroud support.

3.3-2

BFN-26

b. The shroud support, which forms a barrier between the outside of the shroud and the inside of the reactor vessel.

c. The reactor vessel wall below the shroud support.

d. The core shroud up to the level of the jet pump nozzles.

3.3.4.1 Core Structure The core structure surrounds the active core of the reactor and consists of the core shroud, shroud head and steam separator assembly, core support, and top guide. This structure is used to form partitions within the reactor vessel to sustain pressure differentials across the partitions, to direct the flow of the coolant water, and to laterally locate and support the fuel assemblies, control rod guide tubes, and steam separators. Figure 3.3-2 shows the reactor vessel internal flow paths. The core structure is designed in accordance with the structural loading criteria of Appendix C. 3.3.4.1.1 Core Shroud The core shroud is a stainless steel cylindrical assembly which provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow. This partition separates the core region from the downcomer annulus thus providing a floodable region following a recirculation line break. The volume enclosed by the core shroud is characterized by three regions each with a different shroud diameter. The upper shroud has the largest diameter and surrounds the core discharge plenum which is bounded by the shroud head on top and the top guide below. The central portion of the shroud surrounds the active fuel and forms the longest section of the shroud. This section has the intermediate diameter and is bounded at the bottom by the core support assembly. The lower shroud, surrounding part of the lower plenum, has the smallest diameter and at the bottom is welded to the reactor vessel shroud support (see Subsection 4.2 "Reactor Vessel and Appurtenances Mechanical Design"). 3.3.4.1.2 Shroud Head and Steam Separator Assembly The shroud head and steam separator assembly is bolted to the top of the upper shroud to form the top of the core discharge plenum. This plenum provides a mixing chamber for the steam-water mixture before it enters the steam separators. Long holddown bolts are used for easy access during removal. The individual stainless steel axial flow steam separators shown in Figure 3.3-3 are attached to the top of standpipes which are welded into the shroud head. 3.3-3

BFN-26 The centrifugal type steam separators have no moving parts. In each separator, the steam-water mixture rising through the standpipe passes vanes which impart a spin to establish a vortex separating the water from the steam. The steam exits from the top of the separator and rises up to the dryers. The separated water exits from under the separator cap and flows out between the standpipes, draining into the recirculation flow downcomer annulus. 3.3.4.1.3 Core Support (Core Plate) The core support consists of a circular stainless steel plate stiffened with a rim and beam structure. Perforations in the plate provide lateral support and guidance for the control rod guide tubes, peripheral fuel support pieces, incore flux monitor guide tubes, and startup neutron sources. The entire assembly is bolted to a support ledge between the central and lower portions of the core shroud after proper positioning has been assured by alignment pins which fit into slots in the ledge. 3.3.4.1.4 Top Guide The top guide is formed by a series of stainless steel beams joined at right angles to form square openings. Each opening provides lateral support and guidance for four fuel assemblies. Detent sockets are provided beneath the top guide to anchor dry tubes, power range monitor incore detectors, and neutron sources. The top guide is positioned by alignment pins which fit into radial slots in the top of the shroud. 3.3.4.2 Fuel Support Pieces The fuel support pieces, shown in Figure 3.3-4, are of two basic types--peripheral and four-lobed. The peripheral fuel support pieces, which are welded to the core support, are located at the outer edge of the active core and are not adjacent to control rods. Each peripheral fuel support piece will support one fuel assembly and contains a replaceable orifice assembly designed to assure proper coolant flow to the fuel assembly. The four-lobed fuel support pieces will each support four fuel assemblies and are provided with orifice plates to assure proper coolant flow distribution to each fuel assembly. The four-lobed fuel support pieces rest in the top of the control rod guide tubes and are supported laterally by the core support. The control rod blades pass through slots in the center of the four-lobed fuel support pieces. A control rod and the four fuel assemblies which immediately surround it represent a core cell (see Subsection 3.2, "Fuel Mechanical Design"). 3.3-4

BFN-26 3.3.4.3 Control Rod Guide Tubes The control rod guide tubes, located inside the vessel, (see Figure 3.3-1) extend from the top of the control rod drive housings up through holes in the core support. Each tube is designed as the lateral guide for a control rod and as the vertical support for a four-lobed fuel support piece and the four fuel assemblies surrounding the control rod. The bottom of the guide tube is supported by the control rod drive housing (see Subsection 4.2, "Reactor Vessel and Appurtenances Mechanical Design") which in turn transmits the weight of the guide tube, fuel support piece, and fuel assemblies to the reactor vessel bottom head. A thermal sleeve is inserted into the control rod drive housing from below and is rotated to lock the control rod drive tube in place. A key is inserted into a locking slot in the bottom of the control rod drive housing to hold the thermal sleeve in position. 3.3.4.4 Jet Pump Assemblies The jet pump assemblies are located in two semicircular groups in the downcomer annulus between the core shroud and the reactor vessel wall. Each stainless steel jet pump consists of a driving nozzle, suction inlet, throat or mixing section, and diffuser (Figure 3.3-5). The driving nozzle, suction inlet, and throat are joined together as a removable unit and the diffuser is permanently installed. High pressure water from the recirculation pumps (see Subsection 4.3, "Reactor Recirculation System") is supplied to each pair of jet pumps through a riser pipe welded to the recirculation inlet nozzle thermal sleeve. A riser brace is welded to cantilever beams extending from pads on the reactor vessel wall. The jet pump diffuser is a gradual conical section changing to a straight cylindrical section at the lower end. The diffuser is supported vertically by the shroud support, a flat ring which is welded to the reactor vessel wall and to which is welded the shroud support cylinder. The joint between the throat and the diffuser is a slip fit. A metal-to-metal spherical to conical seal joint is used between the nozzle entry section and riser with firm contact maintained by a clamp arrangement which fits under posts on the riser and utilizes a bolt to provide a downward force on a pad on top of the nozzle entry section. The throat section is supported laterally by a bracket attached to the riser. The jet pump diffuser section is welded to the shroud support and provides a positive seal. This permits reflooding the core to the top of the jet pump inlet following a design basis loss-of-coolant accident.1 1 "Design and Performance of GE BWR Jet Pumps," General Electric Co., Atomic Power Equipment Department, July 1968. (APED-5460). 3.3-5

BFN-28 3.3.4.5 Steam Dryers The steam dryer removes moisture from the wet steam which exits from the steam separator. The wet steam leaving the steam separator flows across the dryer vanes and the moisture flows down through collecting troughs and tubes to the water above the downcomer annulus (see Figure 3.3-6). A skirt extends down into the water to form a seal between the wet steam plenum and the dry steam flowing out the top of the dryer to the steam outlet nozzles. Vertical guide rods facilitate positioning the dryer and shroud head in the vessel. Replacement steam dryers (RSD) have been designed to support Extended Power Uprate (EPU) operation. The RSDs hare curved hood six-bank dryers constructed mainly of type 304L stainless steel. The RSD design incorporates design features that were developed to accommodate flow induced vibration (FIV) acoustic loads that lead to steam dryer failures at a BWR-3 plant. The dryer rests on steam dryer support brackets attached to the reactor vessel wall. The original steam dryers are restricted from lifting by steam dryer holddown brackets which are attached to the reactor vessel closure head over the top of the steam dryer lifting lugs when the head is in place. RSDs are restricted from vertical lifting by latch assemblies which hook underneath two of the steam dryer support brackets. 3.3.4.6 Feedwater Spargers As a result of cracks discovered in the feedwater nozzle blend radius, nozzle bore regions, and around the sparger flow holes, the General Electric Company (GE) developed an improved nozzle/sparger design which would reduce this cracking. This new design has been installed in all three units. A separate sparger is fitted to each feedwater nozzle (6) with a double piston ring thermal sleeve. Each sparger is shaped to conform to the curve of the vessel wall and is attached to the wall with two end brackets. These end brackets are bolted to the vessel wall brackets which support the weight of the spargers and position the sparger away from the vessel wall. Flow nozzles are welded to the inner radii of the sparger. Feedwater flow enters the center of the sparger and is discharged radially inward and downward through the nozzles to mix the cooler feedwater with the downcomer flow from the steam separators before it contacts the vessel wall. The feedwater also serves to collapse the steam voids and to subcool the water flowing to the jet pumps and recirculation pumps. This improved nozzle/sparger design, in conjunction with an ultrasonic testing inspection program, will preclude the possibility of any crack growing to a depth which would endanger the pressure vessel integrity. 3.3-6

BFN-26 3.3.4.7 Core Spray Lines The two 100-percent capacity core spray lines separately enter the reactor vessel through the two core spray nozzles as shown in Figures 4.2-1 and 4.2-3 (see Subsection 4.2, "Reactor Vessel and Appurtenances Mechanical Design"). The lines divide immediately inside the reactor vessel. The two halves are routed to opposite sides of the reactor vessel and are supported by clamps attached to the vessel wall. The header halves are then routed downward into the downcomer annulus and pass through the upper shroud immediately below the flange. The flow divides again as it enters the center of the semicircular sparger ring which is routed halfway around the inside of the upper shroud. The ends of the two sparger rings for each line are supported by slip-fit brackets designed to accommodate thermal expansion of the rings. The header routings and supports are designed to accommodate differential movement between the shroud and the vessel. The lower portion of Core Spray Downcomer C has been replaced with a sectional replacement (Unit 3 only). The other core spray line is identical except that the header enters the opposite side of the vessel and the sparger rings are at a slightly different elevation in the shroud. The proper spray distribution pattern is provided by a combination of distribution nozzles pointed radially inward and downward from the sparger rings (see Section 6, "Core Standby Cooling Systems"). 3.3.4.8 Vessel Head Cooling Spray Nozzle The vessel head cooling spray nozzle is mounted to a short length of pipe and a flange, which is bolted to a mating flange above a reactor vessel head nozzle. The vessel head cooling spray nozzles are still installed but are not functional since blind flanges are permanently installed at the mating flange. 3.3.4.9 Differential Pressure and Liquid Control Line The differential pressure and liquid control line serves a dual function within the reactor vessel to inject liquid control solution into the coolant stream (discussed in Subsection 3.8, "Standby Liquid Control System") and to sense the differential pressure across the core support assembly (see Subsection 4.2, "Reactor Vessel and Appurtenances Mechanical Design"). The line enters the reactor vessel at a point below the core shroud as two concentric pipes. In the lower plenum, the two pipes separate. The inner pipe terminates near the lower shroud with a perforated length below the core support assembly. It is used to sense the pressure below the core support during normal operation and to inject liquid control solution when required. This location assures that good mixing and dispersion are facilitated. The use of the inner pipe also reduces the thermal shock to the vessel nozzle should the Standby Liquid Control System ever be used. The outer pipe terminates immediately above the core support assembly and senses the pressure in the region outside the fuel assembly channels. 3.3-7

BFN-26 3.3.4.10 Incore Flux Monitor Guide Tubes The incore flux monitor guide tubes are welded to the top of the incore flux monitor housings (see Subsection 4.2, "Reactor Vessel and Appurtenances Mechanical Design") in the lower plenum and extend up to the top of the core support. The power range detectors for the power range monitoring units and the dry tubes for the source range monitoring/intermediate range monitoring (SRM/IRM) detectors are inserted through the guide tubes and are held in place below the top guide by spring tension. A lattice work of clamps, tie bars, and spacers is bolted around the guide tubes at the approximate level of the reactor vessel shroud support to give lateral support and rigidity to the guide tubes. The bolts and clamps are welded after assembly to prevent loosening during reactor operation. 3.3.4.11 Startup Neutron Sources Startup neutron sources are used to provide a sufficient neutron population to assure that the core neutron flux is detectable by installed neutron monitors and to assure that significant changes in core reactivity are readily detectable by installed neutron flux instrumentation. Antimony-beryllium neutron sources were used for cycle 1. Antimony-beryllium or californium neutron sources may be used in later cycles if spent fuel alone cannot provide the required neutron population. (See Subsection 7.5, "Neutron Monitoring System"). 3.3.4.12 Surveillance Sample Holders The surveillance sample holders are welded baskets containing impact and tensile specimens capsules (see Subsection 4.2, "Reactor Vessel and Appurtenances Mechanical Design"). The baskets hang from brackets on the inside diameter of the reactor vessel at the mid height of the active core and at radial positions of 30°, 120°, and 300°. These locations are chosen to expose the specimens to the same environment and maximum neutron fluxes experienced by the reactor vessel itself while at the same time avoiding jet pump removal interference or damage. 3.3.5 Safety Evaluation 3.3.5.1 Evaluation Methods To determine that the safety design basis is satisfied, the responses of the reactor vessel internals to loads imposed during normal operation, abnormal operational transients, and accidents were examined. Determination of these effects on the ability to insert control rods, cool the core, and flood the inner volume of the reactor vessel was made. The various structural loading combinations assumed to be 3.3-8

BFN-26 imposed on the reactor vessel internals are as described in Appendix C for Class I equipment. These loading combinations include upset loads, emergency loads, and faulted loads. The ASME Boiler and Pressure Code, Section III for Class A vessels, is used as a guide to determine limiting stress intensities and cyclic loadings for the reactor vessel internals. For those components, for which buckling is not a possible failure mode and stresses are within those stated in the ASME Code, it was concluded that the safety design basis is satisfied. For those components, for which either buckling is a possible failure mode or stresses exceed those presented in the ASME Code, then either the elastic stability of the structure or the resulting deformation was examined to determine if the safety design basis was satisfied. 3.3.5.1.1 Specific Events to be Evaluated Examination of the spectrum of conditions for which the safety design basis must be satisfied reveals four significant events as follows:

a. Loss-of-coolant accident. This accident is a break in a recirculation line. The accident results in flow induced loads and acoustic shock loads on some of the reactor vessel internals.

b. Steamline break accident. This accident is a break in one main steamline between the reactor vessel and the flow restrictor. The accident results in significant pressure differentials across the reactor vessel internals.

c. Thermal shock. The most severe thermal shocks to the reactor vessel internals occur when low pressure coolant injection or high pressure coolant injection (LPCI or HPCI) operations reflood the reactor vessel inner volume following either a recirculation line break or a main steamline break (see Section 6, "Core Standby Cooling System").

d. Earthquake. This event subjects the reactor vessel internals to significant forces as a result of ground motion.

Analysis of other conditions existing during normal operation, abnormal operational transients, and accidents showed that the loads affecting the reactor vessel internals are less severe than the four postulated events. 3.3-9

BFN-30 3.3.5.1.2 Reactor Internal Pressure Differentials (RIPD) The core flow, feedwater temperature, reactor pressure, and power level that result in the maximum loads are used as initial conditions. For GE analyses, the normal condition is analyzed using a digital computer code2 for the steady-state thermal-hydraulic analysis of a BWR core (ISCOR). The reactor internal pressure differences (RIPDs) are calculated based on the ISCOR results. Appropriate adders or multipliers, which have been conservatively established based on the GE transient analysis methods on a generic basis for GE BWRs, are applied to the normal condition values to determine the upset condition values. For AREVA reload analyses, core and fuel channel RIPDs are evaluated based on XCOBRA analyses. The locations at which RIPDs are applicable and calculated are shown in Figure 3.3-7. The bounding RIPD values which result from the range of possible reactor power and flow conditions are then determined. The RIPD values acting on the major components at normal increased core flow conditions are documented in the following reference: GEH 002N3430, Revision 0, Task T0304: Reactor Internal Pressure Differences and Fuel Lift Evaluation, March 2015, Table 3.3.1.1. NOTE: The fuel type utilized in the reference is GE13, which is a bounding design for RIPD analysis. AREVA fuel designs in use at BFN are bounded by GE13 as documented in the following reference: 0000-0160-0876-R0, T304 RIPD Assessment for Browns Ferry Control Blade-Channel Interface SC 11-05 Seismic Analysis, GE Hitachi Nuclear Energy, April 2013. RIPDs were subsequently evaluated for operation in the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain and found to be bounded by the reference evaluation. 3.3.5.2 Recirculation Line Break This accident is the same design basis loss-of-coolant accident as described in Section 6, "Core Standby Cooling Systems," and Section 14, "Plant Safety Analysis." It is assumed that an instantaneous, circumferential break occurs in one recirculation 2 EPIC/ISCOR, GE Propriety Code 3.3-10

BFN-30 loop. The reactor is assumed to be operating at design power with rated recirculation flow at the time of the break. The recirculation line break LOCA results in short term transient loads which affect those components in the vicinity of the recirculation outlet nozzle. The resulting flow-induced and acoustic loads from the recirculation line break LOCA have been determined for a reference BWR plant using a recent three-dimensional thermal-hydraulic transient analysis computer code (TRACG, GE Propriety Code). Browns Ferry specific loads are then determined by scaling the reference plant loads by the effects of the geometric and thermal-hydraulic parameter differences between the reference plant and Browns Ferry. The flow-induced and acoustic loads resulting from the introduction of the TRACG method of calculation are greater than the results determined for past Browns Ferry evaluations. In order to determine the impact of off-rated power/flow conditions, load multipliers are calculated normalized to a multiplier of 1.0 for the normal power and rated core flow condition. Thus, the effect of off-rated power and flow conditions or increased core flow is illustrated by the normalized multipliers. The geometric scaling is based on the dimensional differences between the Browns Ferry reactor configuration and the reference BWR plant in the vicinity of the recirculation outlet and shroud annulus region. The thermal-hydraulic scaling is determined by using ISCOR program. The load multipliers for the selected points on the power flow map are also determined by using the ISCOR program. The acoustic loads are short term transient impulse loads and are conservatively evaluated by determining a static equivalent load. The flow induced loads are more slowly applied transient loads as compared to acoustic loads; therefore, the loads are not added. The higher load governs structural evaluations. Detailed evaluation of these loads indicated that the acoustic loads resulted in higher loads and are the loads used in structural evaluations. Flow-induced and acoustic loads for a recirculation line break LOCA from operation at minimum recirculation pump speed typically results in the highest acoustic load regardless of the rated power or core flow; however, the probability of a LOCA occurring at this in-frequent operational condition is extremely remote. Therefore, evaluations are based on the more probable LOCA during maximum extended load line limit (MELLL) and MELLLA+ operation. An analysis has been performed to evaluate the potential leakage from within the floodable inner volume of the reactor vessel during the recirculation line break and subsequent LPCI reflooding. The two possible sources of leakage are:

a. Jet pump throat to diffuser joint.

b. Jet pump nozzle to riser joint.

3.3-11

BFN-30 The jet pump to shroud support joint is welded and therefore is not a possible source of leakage. The throat to diffuser joints for all jet pumps by analysis are shown to leak no more than a total of 225 gpm. The jet pump nozzle to riser joint by analysis is shown to leak no more than 582 gpm for the pumps through which the vessel is being flooded. The summary of maximum leakage is then: 225 gpm Total leakage through all throat-to-diffuser joints 582 gpm Total leakage through all nozzle-to-riser joints 807 gpm Total maximum rate (Unit 3) 157 gpm Additional leakage through bolted design access hole covers (Unit 1 only) 964 gpm Total maximum rate (Unit 1) 209 gpm Additional leakage through bolted design access hole covers (Unit 2 only) 1016 gpm Total maximum rate (Unit 2) LPCI capacity is sized to accommodate 3000 gpm leakage at these locations. It is concluded that the reactor vessel internals retain sufficient integrity during the recirculation line break accident to allow reflooding the inner volume of the reactor vessel. 3.3.5.3 Steamline Break Accident The analysis of this accident assumes an instantaneous circumferential break of one main steamline between the reactor vessel and the main steamline flow restrictor. This is not the same accident as that described in Chapter 14, "Plant Safety Analysis," (which postulates a break downstream of the flow restrictors) because greater differential pressures across the reactor vessel internals result from this accident. It is noteworthy that this accident results in greater loading of the reactor vessel internals and a higher depressurization rate than does the recirculation line break. This is because the depressurization rate is proportional to the mass flow rate and the excess of fluid escape enthalpy above saturated water enthalpy. However, mass flow rate is inversely proportional to escape enthalpy, he; therefore, depressurization rate is proportional to 1-hf/he. Consequently, depressurization rate decreases as h decreases, that is, depressurization is less for mixture flow than for steam flow. The main steam line (MSL) break inside containment is postulated for calculating the RIPDs, except for the steam dryer RIPD for which the steamline break outside containment is postulated, as specified in Section 3.6 of the UFSAR. These faulted loading conditions are analyzed using the LAMB thermal-hydraulic analysis code. The bounding RIPD values for MSL break from the low reactor power/high core flow (point I shown on FSAR Figure 3.7-1) or full power/high core flow (point F shown on FSAR Figure 3.7-1) operation is determined and used for reactor internals structural evaluations. The maximum RIPD values for this faulted condition are identified in the following reference: 3.3-12

BFN-28 GEH 002N3430, Revision 0, Task T0304: Reactor Internal Pressure Differences and Fuel Lift Evaluation, March 2015, Table 3.3.1.1. Note: The fuel type utilized in the reference is GE13, which is a bounding design for RIPD analysis. AREVA fuel designs in use at BFN are bounded by GE13 as documented in the following reference: 0000-0160-0876-R0, T304 RIPD Assessment for Browns Ferry Control Blade Channel Interface Sc 11-05 Seismic Analysis, GE Hitachi Nuclear Energy, April 2013. These maximum differential pressures are used, in combination with other assumed structural loads as described in Appendix C, to determine the total loading on the various reactor vessel internals. The various internals are then examined to assess the extent of deformation and collapse, if any. Of particular interest are the responses of the fuel bundle, the core support, the guide tubes, and the metal channels around the fuel bundles. 3.3.5.3.1 Core Support The two considerations important to the core support evaluation are sliding of the core support and buckling of the supporting beams. Evaluations have determined that the core support will not slide under the postulated accident conditions with preload on the holddown bolts. Additional resistance to sliding is provided by aligners which further stabilize the core support. The core plate buckling pressure is evaluated by a computer program3 that calculates core plate stiffener beam-buckling capability. It uses the Rayleigh-Ritz energy method to determine the applied moment to begin yielding, and then to buckle a given tee beam. The tee beam models a segment of a BWR core plate with a stiffener beam. The pressure differential across the plate that would have created this moment is calculated for the longest core plate beam. The MSL break pressure of 28.5 psid (3952 MWt) is within the permissible load allowed by the Buckling Stability Criteria of Table C.2-3(b) for the safety factors in accordance with Appendix C: Sections C.2 and C.2.6 required emergency and faulted load combinations. 3.3.5.3.2 Guide Tubes Because the guide tube experiences external-to-internal pressure differentials, the guide tubes were examined for buckling under these conditions. For a guide tube with minimum wall thickness and maximum allowed ovality, the pressure which causes buckling is 105 psi compared to the main steam line break design pressure of 28.5 psid (3952 MWt). This pressure is well within the permissible load allowed 3 PIPST01, GE Proprietary Code 3.3-13

BFN-28 by the Buckling Stability Criteria of Table C.2-3(b) for the safety factors in accordance with Appendix C: Sections C.2 and C.2.6 required emergency and faulted load combinations. It is concluded that the guide tube will not fail under the assumed accident conditions. 3.3.5.3.3 Fuel Channels The description of the testing and analysis performed for initial core channels is provided in this section. The NRC approval of the channels used with the current AREVA reload is provided in EMF-93-177(P)(A) Revision 1 with the plant and cycle specific analyses reported in the reload specific fuel mechanical design report. The fuel channel load due to an internally applied pressure was examined utilizing a fixed-fixed beam analytical model under a uniform load. Tests were conducted to verify the applicability of the analytical model. The results indicate that the analytical model is conservative. The fuel channels may deform sufficiently outward to cause some interference with movement of the control rod blade. There are about 15 factors such as fuel channel deformation, core support hole tolerance, top guide beam location, etc., that determine the clearance between the control rod blade and fuel channel. If each of these tolerance factors are assumed to be at the worst extreme of the tolerance range, then a slight interference would develop under an 18 psi pressure difference across the channel wall. At the top of the control rod there is a roller or spacer pad to guide the blade as it is inserted. The clearance between channels is 70 mils less than the diameter of the roller or spacer pad, causing it to slide or skid instead of roll. As the rod is inserted about halfway there is a tendency for the control rod sheath to push inward on the channel. This is a control rod surface to channel surface contact. A "worst case" study indicates a possibility of a 50-mil interference. The possibility of a worst case developing is extremely remote. A statistical analysis utilizing a normal distribution for each of the 15 variables indicates that no interference occurs within 3 sigma limits, where sigma is the standard deviation in a point distribution of events. Three sigma lies in the 0.995 percentile of probability of nonoccurrence. However, even if interference occurs, the result is negligible. About 1 pound of lateral force is required to deflect the channel inboard 1 mil. The friction force developed is an extremely small percentage of the total force available to the control rod drives. The above discussion presupposes the control rod has not moved when the fuel channel experiences the largest magnitude of pressure drop. Analysis indicates that the rod is about 70 percent to 90 percent inserted. If the rod is beyond 70 percent inserted, then no interference is likely to develop because all the channel deformation is in the lower portion of the fuel channel, whereas the roller or spacer 3.3-14

BFN-26 pad is at the top of the rod. It is concluded that the main steamline break accident can pose no significant interference to the movement of control rods. Also, the calculated maximum pressure differential across the core is approximately 12 psi below the 42 psi required to lift a fuel bundle. The AREVA methodology for evaluating the fuel channel deformation due to the internal pressure and irradiation growth makes use of a Monte Carlo analysis to determine the probability of having a stuck control blade condition. A 95/95 statistical statement is made, taking into account variations in core tolerances, fuel channel tolerances, bulge and bow, to demonstrate acceptable interface with the surfaces of the control blades. Fuel lift is evaluated for AREVA fuel assuming maximum differential pressure conditions. A substantial margin is calculated to exist prior to lift off. 3.3.5.4 Thermal Shock The most severe thermal shock effects for the reactor vessel internals result from the reflooding of the reactor vessel inner volume. For some vessel internals, the limiting thermal shock occurs from LPCI operation and for others HPCI operation is controlling, dependent upon the location of the component. These effects occur as a result of any large loss-of-coolant accident, such as the recirculation line break and the steamline break accidents previously described. Three specific locations are of particular interest, as shown in Figure 3.3-9. The locations are as follows:

a. Shroud support plate,

b. Shroud-to-shroud support plate discontinuity, and

c. Shroud inner surface at highest irradiation zone.

The peak strain resulting in the shroud support plate is about 6.5 percent. This strain is higher than the 5.0 percent strain permitted by the ASME Code, Section III, for ten cycles, but for one cycle, peak strain corresponds to about six allowable cycles of an extended ASME Code as applied to less than ten cycles. Figure 3.3-10 illustrates both the ASME Code curve and the basic material curves from which it was established (with the safety factor of 2 on strain or 20 on cycles, whichever is more conservative). The extension of the ASME Code curve represents a similar criteria to that used in the ASME Code, Section III, but applied to fewer than ten cycles of loading. For this type 304 stainless steel material, a 10 percent peak strain corresponds to one allowable cycle of loading. Even a 10 percent strain for a single-cycle loading represents a very conservative suggested 3.3-15

BFN-28 limit because this has a large safety margin below the point at which even minor cracking is expected to begin. Because the conditions which lead to the calculated peak strain of 6.5 percent are not expected to occur even once during the entire reactor lifetime, the peak strain is considered tolerable. The results of the analysis of the shroud-to-shroud support plate discontinuity region are as follows: Amplitude of Alternating Stress ......................................................... 180,000 psi Peak Strain ....................................................................................... 1.34 percent The ASME Code, Section III, allows 220 cycles of this loading, thus no significant deformations result. The most irradiated point on the inner surface of the shroud is subjected to a total integrated neutron flux of 2.7 x 1020 nvt (>1 MeV) by the end of plant life. The peak thermal shock stress is 155,700 psi, corresponding to a peak strain of 0.57 percent. The shroud material is type 304 stainless steel, which is not significantly affected by irradiation. The material does experience some hardening and an apparent loss in uniform elongation, but it does not experience a loss in reduction of area. Because reduction of area is the property which determines tolerable local strain, irradiation effects can be neglected. The peak strain resulting from thermal shock at the inside of the shroud represents no loss of integrity of the reactor vessel inner volume. 3.3.5.5 Earthquake The seismic loads on the RPV and RPV internals are determined from dynamic earthquake analysis described in Section 12.2 using the mathematical models of the RPV and internals shown in Figures 12.2-27B and 12.2-27C. The design of the RPV and internals are described in Section C.4 of Appendix C and Appendices J, K, and L. RPV Internals and Structural Integrity Evaluations were performed with GE fuel per the following

Reference:

002N4782, Revision 0, Task T0303: RPV Internals Structural Integrity Evaluation, GE Hitachi Nuclear Energy, April 2015. For AREVA fuels, the evaluations were performed per the following

Reference:

0000-0166-4147, Revision 0, RPV Internals Structural Integrity Evaluations for AREVA ATRIUM-10 and ATRIUM-10XM Fuels, GE Hitachi Nuclear Energy, January 2014. 3.3-16

BFN-26 3.3.5.6 Conclusions The analyses of the responses of the reactor vessel internals to situations imposing various loading combinations on the internals show that deformations are sufficiently limited to allow both adequate control rod insertion and proper operation of the Core Standby Cooling Systems. Sufficient integrity of the internals is retained in such situations to allow successful reflooding of the reactor vessel inner volume. The analyses considered various loading combinations, including loads imposed by external forces. Thus, safety design bases 1, 2, and 3 are satisfied. 3.3.6 Inspection and Testing Quality control methods were used during the fabrication and assembly of reactor vessel internals to assure that the design specifications were met. The reactor coolant system, which includes the reactor vessel internals, was thoroughly cleaned and flushed before fuel was loaded initially. During the preoperational test program, operational readiness tests are performed on various systems. In the course of these tests such reactor vessel internals as the feedwater spargers, the core spray lines, the vessel head cooling spray nozzle, and the Standby Liquid Control System line are functionally tested. Steam separator-dryer performance tests were run to determine carryunder and carryover characteristics on the first 1098 MWe boiling water reactor plant to go into operation. Samples were taken from the inlet and outlet of the steam dryers and from the inlet to the main steamlines at various reactor power levels, water levels, and recirculation flow rates. Moisture carryover was determined from sodium-24 activity in these samples and in reactor water samples. Carryunder was determined from measured flows and temperatures determined by heat balances. Vibration analysis of reactor vessel internals is included in the design to reduce failures due to vibration. When necessary, vibration measurements are made during startup tests to determine the vibration characteristics of the reactor vessel internals and the recirculation loops under forced recirculation flow. Vibratory responses are recorded at various recirculation flow rates using strain gauges on fuel channels and control rod guide tubes, accelerometers on the shroud support plate and recirculation loops, and linear differential transducers on the upper shroud and shroud head-steam separator assembly. The vibration analyses and tests are designed to determine any potential, hydraulically induced equipment vibrations and to check that the structures will not fail due to fatigue. The structures were analyzed for natural frequencies, mode shapes, and vibrational magnitudes that could lead to fatigue at these frequencies. With this analysis as a guide, the reactor internals were instrumented and tested to ascertain that there were no gross instabilities. The 3.3-17

BFN-28 cyclic loadings are evaluated using as a guide the cyclic stress criteria of the ASME Code, Section III. These field tests were performed only on reactor vessel internals that represented a significant departure from design configurations previously tested and found to be acceptable. Field test data were correlated with the analyses to ensure validity of the analytical techniques on a continuing basis4. The reactor vessel and internals are designed to assure adequate working space and access for inspection of selected components and locations5. The criteria for selecting the components and locations to be inspected are based on the probability of a defect occurring or enlarging at a given location and includes areas of known stress concentrations and locations where cyclic strain or thermal stress might occur. The reactor vessel internals inspection program is detailed in Subsection 4.12, "Inservice Inspection and Testing." After installation, the RSDs are evaluated during initial power ascension to assess their structural performance. The RSD for the lead unit is assessed for structural integrity through measured on-dryer strains and acceleration. The RSDs for the follow on units are assessed through the analysis of dryer loads projected from main steam line strain gauge measurements. Reactor parameters that have been indicative of past dryer structural failures are also monitored during initial power ascension. 4 Quad-Cities Station Units 1 and 2 Docket No.s 50-254 and 265, Amendment 19 5 Brandt, F. A., Design Provision for In-Service Inspection, General Electric Company. Atomic Power Equipment Department, April 1967 (APED-5450) 3.3-18

BFN-28 Table 3.3-1 REACTOR VESSEL INTERNALS, DESIGN DATA Core Shroud Jet Pumps Upper Portion,o.d.,in. 220 Number 20 Central Portion,o.d.,in. 207 Throat Diameter, in. 8.18 Lower Portion,o.d.,in. 201 Weight, lb. 22,700 Central Portion, thickness, in. 2 Original Steam Dryers Weight, lb. 116,900 Weight, lb 90,000 Shroud Head-Steam Separator Assembly Replacement Steam Dryers Head Thickness, in. 2.0 Weight, lb. 119,000 Number of Separators 211 Feedwater Sparger Separator o, d., in. 12.75 Dimensions, in. 6 Sched. 40 Number 6 Cross Section Area, ft2 0.2006 Standpipe i. d., in. 6.065 Number 6 Standpipe o, d., in. 6.625 Core Spray Sparger Rings Weight, lb. 139,600 Diameter, in. 4 Sched. 40 S Core Support Cross Section Area, ft2 0.088 Weight, lb. 20,500 Number of Spray Outlets 260 Top Guide Weight, lb. 4,317 Weight, lb. 15,200 Vessel Head Cooling Spray Nozzle Fuel Support Pieces Pipe Size, in. 4 Number of Preipheral 24 Scheduled 40 Four Lobe Differential Pressure & Liquid Control Line Number without Plugs 185 Inner Pipe (Liquid Control), in 1 Sched. 40 Number with Plugs 0 Outer Pipe, in. 2 Sched. 40 Weight, lb. 11,300 Incore Flux Monitor Guide Tubes Control Rod Guide Tubes Number 55 Number 185 Weight, lb. 46,250 Surveillance Sample Holders 3 Total Weight of Reactor Vessel 477,000 Internals, pounds (excluding fuel)

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BFN-16 Figure 3.3-11 Deleted by Amendment 10.

BFN-27 3.4 REACTIVITY CONTROL MECHANICAL DESIGN 3.4.1 Safety Objective The safety objective of the reactivity control mechanical design is to provide a means to quickly terminate the nuclear fission process in the core so that damage to the fuel barrier is limited. The objective is met by inserting reactivity control devices into the reactor core. 3.4.2 Power Generation Objective The power generation objective of the reactivity control mechanical design is to provide a means to control power generation in the fuel. This objective is met by positioning reactivity control devices in the reactor core. 3.4.3 Safety Design Basis

1. The reactivity control mechanical design shall include control rods.

a. The control rods shall be so designed and have sufficient mechanical strength to prevent the displacement of their reactivity control material.

b. The control rods shall have sufficient strength and be of such design as to prevent deformation that could inhibit their motion.

c. Each control rod shall include a device to limit its free fall velocity to such a rate that the nuclear system process barrier is not damaged due to a pressure increase caused by the rapid reactivity increase resulting from the free fall of a control rod from its fully inserted position.

2. The reactivity control mechanical design shall provide for a sufficiently rapid insertion of control rods so that no fuel damage results from any abnormal operating transient.

3. The reactivity control mechanical design shall include positioning devices each of which individually support and position a control rod.

4. Each positioning device shall:

a. Prevent its control rod from withdrawing as a result of a single malfunction.

b. Avoid conditions which could prevent its control rod from being inserted.

c. Be individually operated such that a failure in one positioning device does not affect the operation of any other positioning device.

3.4-1

BFN-27

d. Be individually energized when rapid control rod insertion (scram) is signaled so that failure of a power source external to the positioning device does not prevent other control rods from being inserted.

e. Be locked to its control rod to prevent undesirable separation.

f. Provide positive indication of control rod position to the operator.

3.4.4 Power Generation Design Basis

1. The reactivity control mechanical design shall include reactivity control devices (control rods) which shall contain and hold the reactivity control material necessary to control the excess reactivity in the core.

2. The reactivity control mechanical design shall include provisions for adjustment of the control rods to permit control of power generation in the core.

3. The reactivity control mechanical design shall provide indication of the CRDM temperature to the operator.

3.4.5 Description The reactivity control mechanical design consists of control rods which can be positioned in the core, during power operation, by individual control rod drive (CRD) mechanisms. The CRD mechanisms are part of the CRD System. The CRD System hydraulically operates the CRD mechanisms using water from the condensate storage system as a hydraulic fluid. The CRD mechanisms are used to manually position individual control rods during normal operation but act automatically to rapidly insert all control rods during abnormal (scram) conditions. The control rods, CRD mechanisms, and that part of the CRD Hydraulic System necessary for scram operation are designed as Class I equipment in accordance with Appendix C, "Structural Qualification of Subsystems and Components." 3.4.5.1 Reactivity Control Devices 3.4.5.1.1 Control Rods The control rods perform the dual function of power shaping and reactivity control. Power distribution in the core is controlled during operation of the reactor by manipulation of selected patterns of control rods. The control rods are positioned in 3.4-2

BFN-27 a manner which counterbalances steam void effects at the top of the core and results in significant power flattening. Five General Electric control rod designs and one Westinghouse design are currently approved for use in BFN reactors: (1) Original Equipment, (2) Modified BWR/6, (3) Hybrid I, (4) Marathon, (5) Ultra, and (6) Westinghouse CR-82M-1. All of these control rods are "Matched Worth" designs. The reactivity worth of the replacement control rods is nearly identical to the Original Equipment control rod design so that all the designs can be used interchangeably without affecting lattice physics and core reload analyses or core monitoring software. These control rods are also designed to be interchangeable considering system performance and mechanical fit. A brief description of each design follows. Original Equipment Control Rod The Original Equipment control rod consists of a sheathed cruciform array of neutron absorber rods consisting of stainless steel tubes filled with boron-carbide powder. The control rods are 9.75 inches in total span and are located uniformly through the core on a 12-inch pitch. Each control rod is surrounded by four fuel assemblies. The main structural member of a control rod is made of type 304 stainless steel and consists of a top casting which incorporates a handle, a bottom casting which incorporates a velocity limiter and control rod drive coupling, a vertical cruciform center post, and four U-shaped absorber tube sheaths. The two end castings and the center post are welded into a single skeletal structure. The U-shaped sheaths are resistance welded to the center post and castings to form a rigid housing to contain the neutron absorber rods. Rollers or spacer pads at the top and the rollers at the bottom of the control rod provide guidance for the control rod as it is inserted and withdrawn from the core. The control rods are cooled by the fuel assembly bypass flow. The U-shaped sheaths are perforated to allow the coolant to freely circulate about the absorber tubes. Operating experience has shown that control rods constructed as described above are not susceptible to dimensional distortions, as required by safety design basis 1.b. The boron-carbide (B4C) powder in the stainless steel absorber tubes is compacted to about 70 percent of its theoretical density; the boron-carbide contains a minimum of 76.5 percent by weight natural boron. The Boron-10 (B-10) content of the boron is 18.0 percent by weight minimum. The absorber tubes are made of type 304 or a high purity type 348 stainless steel. An absorber tube is 0.188 inch in outside diameter and has a 0.025 inch wall thickness (Figure 3.4-2). An absorber tube is sealed by a plug welded into each end. The boron-carbide is separated longitudinally into individual compartments by stainless steel balls at approximately 16-inch intervals. The steel balls are held in place by a slight crimp of the tube. Should the boron-carbide tend to compact further in service, the steel balls will distribute the resulting voids over the length of the absorber tube. 3.4-3

BFN-27 The end of control blade life occurs when any quarter segment of the control blade reaches a 10-percent reduction in relative reactivity worth. The reduction in blade worth is due to a combination of Boron-10 depletion and boron carbide loss resulting from cracking of the absorber rod tubes. The mechanism identified as causing the tube failures is B4C swelling resulting in stress corrosion cracking. Given sufficient exposure, the B4C swelling may initiate small stress corrosion cracks on the tube surface. Examinations of high exposure blades have shown that these surface cracks may exist at average segment Boron-10 depletions greater than 20 percent. Ultimately, the cracks will propagate through the tube wall allowing reactor coolant to enter the tube. In this condition, B4C can be leached out slowly by the reactor coolant, resulting in a loss of control blade worth. Examinations have shown that the combination of Boron-10 depletion and loss of B4C result in a 10-percent reduction in relative control blade worth at approximately 34 percent average Boron-10 depletion. This is the defined end of useful blade life for the standard all B4C control blades.1 Modified BWR/6 Control Rod The Modified BWR/6 control rods differ from the Original Equipment blades by having increased wing thickness, increased neutron absorber, and a double bail handle (Figure 3.4-1). The absorber tubes have a 0.220 inch outside diameter and a 0.027 inch wall thickness. The BWR/6 control rods have been modified with replacement pins and rollers made from low-cobalt materials and sized for BWR/4 application. Cobalt reduction is desirable to reduce activation products within the reactor system and to reduce radiation levels of spent control rods. The design blade life for the BWR/6 control rods is 34 percent average Boron-10 depletion, which is the same as for the Original Equipment blades. Hybrid I Control Rod Hybrid I Control Rod (HICR) assemblies contain improved B4C absorber rod material to eliminate cracking during assembly lifetime.2,3 Also in the HICR design, the three outermost absorber tubes in each wing are replaced with solid hafnium rods which increase blade lifetime compared to standard all-B4C control blades. End-of-life boron equivalent depletion for the HICRs is 56 percent for any quarter segment of the control blade. 2 Safety Evaluation of General Electric Hybride I Control Rod Assembly, NEDE-22290-A, September 1983. 3 General Electric BWR Control Rod Lifetime, NEDE-30931-P, March 1985. 3.4-4

BFN-27 Marathon Control Rod The Marathon control rod (Figure 3.4-12) differs from the preceding designs by replacement of the absorber tube and sheath arrangement with an array of square tubes, which results in reduced weight and increased absorber volume4. The square tubes are fabricated from a high purity stabilized Type-304 stainless steel that provides high resistance to irradiation-assisted stress corrosion cracking. The absorber tubes are welded lengthwise to form the four wings of the control rod. For the BFN BWR/4 D-lattice design, each wing is comprised of 14 absorber tubes. The absorber tubes each act as an individual pressure chamber for the retention of helium which is produced during neutron absorption reactions. The four wings are welded to a central tie rod to form the cruciform-shaped member of the control rod. The square tubes are circular inside and are loaded with either B4C or hafnium. The absorber tubes have an inside diameter of 0.250 inches and a nominal wall thickness of 0.024 inches. The B4C is contained in separate capsules to prevent its migration. The capsules are placed inside the absorber tubes and are smaller than the absorber tube inside diameter, allowing the B4C to swell before it makes contact with the absorber tubes thereby providing improved resistance to stress corrosion. The B4C capsules are fabricated from stainless steel tubing and have stainless steel caps attached by rolling the tubing into grooves in the caps. The capsules are loaded into the individual absorber tubes, which are then sealed at each end by welded end caps. The capsules securely contain the B4C while allowing the helium to migrate through the absorber tube. The Marathon design offers increased blade lifetimes due to the increased absorber loading and absorber tube design improvements. The Marathon blades have a quarter segment, end-of-life Boron-10 equivalent depletion limit of 68 percent. Ultra Control Rod The GEH manufactured Ultra MD and Ultra HD control rods (formerly known as Marathon-5S and Marathon-Ultra, respectively) are similar to the original Marathon design shown in Figure 3.4-12. The NRC acceptance of the Marathon-5S and Marathon-Ultra Control Rods is documented by Licensing Topical Reports (LTRs). 9,10 .The Ultra MD Control Rod consist of simplified absorber tubes, edge welded together to form the control rod wings, and welded to a full-length tie rod to form the cruciform assembly shape. The absorber tubes are filled with a combination of boron carbide (B4C) capsules, and empty capsules. The Ultra HD 9 NEDE-33284P-A Rev. 2, Marathon-5S Control Rod Assembly, October 2009. 10 NEDE-33284 Supplement 1P-A Rev. 1, Marathon-Ultra Control Rod Assembly, March 2012. 3.4-5

BFN-27 Control Rod is a derivative version of the Ultra MD Control Rod in that it uses an identical outage structure. The only differences for the Ultra HD Control Rod is the inclusion of full length hafnium rods in high-depletion absorber tubes, and the use of a then-wall boron carbine capsule, similar in geometry to the previous Marathon control rod design. 4 The original Marathon design was intended to reduce stress/strain associated with B4C swelling. The Ultra MD and Ultra HD designs were developed to eliminate stress/strain associated with B4C swelling to prevent end of life cracking issues. The Ultra MD and Ultra HD blades have quarter segment, end-of-life Boron-10 equivalent depletion limit of less than 57 percent to 67 percent, respectively. Westinghouse CR-82M-1 Control Rod The Westinghouse CR-82M-1 control rod (Figure 3.4-13) consists of four stainless steel sheets welded together to form the cruciform shaped rod.5,6 Each sheet has horizontally drilled holes to contain the absorber materials (boron carbide powder and hafnium). The hafnium tip of the CR-82M-1 design protects the control rod from absorber material swelling when operated in the fully withdrawn position. The blade material used in the CR-82M-1 design is AISI 316L stainless steel. AISI 316L stainless steel as structural material is an irradiation resistant steel not readily sensitized to irradiation assisted stress corrosion cracking, and also has an extremely low cobalt content (< 0.02 percent). The design with horizontally drilled absorber holes limits the washout of boron carbide in case of a defect in a wing, thus maintaining full reactivity worth. The absorber hole geometry for the Westinghouse CR-82M-1 control rod is optimized to provide a matched worth (within 5 percent) to the initial worth of an original equipment control rod. The nuclear end-of-life (10 percent worth decrease from initial original equipment value) depletion limits for the CR-82M-1 design are: 85 percent equivalent Boron-10 depletion for the top quarter segment, and 89 percent equivalent Boron-10 depletion for the other quarter segments. [Note: These are equivalent depletion limits developed to allow the utility to monitor the CR-82M-1 control rod as if it were an original equipment control rod. Conclusion Thus, the control rods and absorber tubes meet the requirements of safety design basis 1.a. 3.4-6

BFN-27 3.4.5.1.2 Control Rod Velocity Limiter (Figures 3.4-3 and 3.4-4) The control rod velocity limiter is an integral part of the bottom assembly of each control rod. This static engineered safeguard protects against a high reactivity insertion rate by limiting the control rod velocity in the event of a control rod drop accident. It is a one way device, in that the control rod scram velocity is not significantly affected but the control rod dropout velocity is reduced to a permissible limit. The velocity limiter is in the form of two nearly mated conical elements that act as a large clearance piston and baffle inside the control rod guide tube over the length of the control rod stroke. It is fabricated of type 304 stainless steel. It has a nominal diameter of approximately 9.2 inches, and is fitted inside the control rod guide tube which has an inside diameter of 10.4 inches. This configuration results in an annulus between the limiter and the guide tube of approximately 0.6 inch. The limiter always remains in the guide tube except when the control rod is removed. Four adjacent fuel assemblies and the fuel assembly support casting must be removed before the control rod can be removed because of the shape of the velocity limiter. The hydraulic drag forces on a control rod are approximately proportional to the square of the rod velocity and are negligible during normal rod withdrawal or rod insertion. However, during the scram stroke the rod reaches high velocity and the drag forces could become appreciable. In order to limit control rod velocity during dropout but not during scram, the velocity limiter is provided with a streamlined profile in the scram (upward) direction. Thus, when the control rod is scrammed, the velocity limiter assembly offers little resistance to the flow of water over the smooth surface of the upper conical element into the annulus between the guide tube and the limiter. In the dropout direction, however, water is trapped by the lower conical element and discharged through the annulus between the two conical sections. Because this water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod assembly to less than 5 ft/sec at 70 degrees F.7 3.4.5.2 Control Rod Drive Mechanisms (Figures 3.4-3, 3.4-5, 3.4-6, and 3.4-9) The CRD mechanism (drive), used for positioning the control rod in the reactor core, is a double-acting, mechanically latched, hydraulic cylinder using water from the condensate storage system as its operating fluid. The individual drives are mounted on the bottom head of the reactor pressure vessel. Each drive is an integral unit 3.4-7

BFN-27 contained in a housing extending below the reactor vessel. The lower end of each drive housing terminates in a flange to which the drive is bolted. The drives do not interfere with refueling and are operative even when the head is removed from the reactor vessel. The bottom location makes maximum utilization of the water in the reactor as a neutron shield giving the least possible neutron exposure to the drive components. The use of reactor water from the condensate storage system as the operating fluid eliminates the need for special hydraulic fluid. Drives utilize simple piston seals whose leakage does not contaminate the reactor vessel and helps cool the drive mechanisms. The drives are capable of inserting or withdrawing a control rod at a slow controlled rate for reactor power level adjustment, as well as providing rapid insertion when required. A locking mechanism on the drive allows the control rod to be locked at every six inches of stroke over the twelve foot length of the core. A coupling at the top end of the drive index tube (piston rod) engages and locks into a mating socket at the base of the control rod. The weight of the control rod is sufficient to engage and lock this coupling. Once locked, the drive and rod form an integral unit which must be manually unlocked by specific procedures before a drive and its rod can be separated; this prevents accidental separation of a control rod from its drive. Each drive positions its control rod in 6-inch increments of stroke and holds it in these distinct latch positions until actuated by the hydraulic system for movement to a new position. Indication is provided for each rod that shows when the insert travel limit or withdraw travel limit is reached. An alarm annunciates when withdraw overtravel limit on the drive is reached. Normally, the control rod seating at the lower end of its stroke prevents the drive withdraw overtravel limit from being reached. If the drive can reach the withdrawal overtravel limit, it indicates that the control rod is uncoupled from its drive. The overtravel limit alarm allows the coupling to be checked. Individual rod position indicators are grouped together on the control panel in one display and correspond to the relative rod location in the core. Each rod indicator gives continuous rod position indication in digital form. A separate, smaller display is located just below the large display on the vertical part of the bench board. This display presents the positions of the control rod selected for movement and the other rods in the rod group. For display purposes the control rods are considered in groups of four adjacent rods centered around a common core volume monitored by four LPRM strings. Rod groups at the periphery of the core may have less than four rods. The small rod display shows the positions in digital form of the rods in the group to which the selected rod belongs. A white light indicates which of the four rods is the one selected for movement. 3.4-8

BFN-27 3.4.5.2.1 Components Figure 3.4-5 illustrates the principles of operation of a drive. Figures 3.4-6, 3.4-9 and 3.4-9a illustrate the drive in more detail. Currently BWR/6 drives (Figure 3.4-9a) obtained from Hartsville Nuclear Plant and modified by General Electric are acceptable replacements for BWR/4 drives (Figures 3.4-6 and 3.4-9). BWR/6 drives are currently in use at Browns Ferry Nuclear Plant. Throughout the following sections, details for the BWR/4 drives are being maintained for historical purposes. Following is a description of the main components of the drive and their functions. Drive Piston and Index Tube The drive piston is mounted at the lower end of the index tube which functions as a piston rod. The drive piston and index tube make up the main moving assembly in the drive. The drive piston operates between positive end stops, with a hydraulic cushion provided at the upper end only. The piston has both inside and outside seal rings and operates in an annular space between an inner cylinder (fixed piston tube) and an outer cylinder (drive cylinder). The effective piston area for down-travel or withdraw is about 1.2 square inches versus 4.0 square inches for up-travel or insertion. This difference in driving area tends to balance out the control rod weight and makes it possible to always have a higher insertion force than withdrawal force. The index tube is a long hollow shaft made of nitrided type 304 stainless steel (XM-19 for BWR/6 drives). Any index tubes which are found to need replacing during normal CRD maintenance are replaced by index tubes of identical design made of Grade XM-19 stainless steel. This tube has circumferential locking grooves spaced every 6 inches along the outer surface. These grooves transmit the weight of the control rod to the collet assembly. Collet Assembly The collet assembly serves as the index tube locking mechanism. It is located in the upper part of the drive unit. The collet assembly prevents the index tube from accidentally moving downward. The collet assembly consists of the collet fingers, a return spring, a guide cap, a collet housing (part of the cylinder, tube, and flange assembly), and the collet piston seals. Locking is accomplished by six fingers mounted on the collet piston at the top of the drive cylinder. In the locked or latched position, the fingers engage a locking groove in the index tube. 3.4-9

BFN-27 The collet piston is normally held in the latched position by a return spring force of approximately 150 pounds. Metal piston rings are used to seal the collet piston from reactor vessel pressure. The collet assembly will not unlatch until the collet fingers are unloaded by a short, automatically sequenced, drive in signal. A pressure of approximately 180 psi above reactor vessel pressure acting on the collet piston is required to overcome spring force, slide the collet up against the conical surface in the guide cap, and spread the fingers out so that they do not engage a locking groove. The collet piston is nitrided to minimize wear due to rubbing against the surrounding cylinder surfaces. Fixed in the upper end of the drive assembly is a guide cap. This member provides the unlocking cam surface for the collet fingers. It also serves as the upper bushing for the index tube and is nitrided to provide a compatible bearing surface for the index tube. As reactor water is used to supplement accumulator pressure during a scram, it is drawn through a filter on the guide cap. Piston Tube and Stop Piston Extending upward inside the drive piston and index tube is an inner cylinder or column called the piston tube. The piston tube is fixed to the bottom flange of the drive and remains stationary. Water is brought to the upper side of the drive piston through this tube. A series of orifices at the top of the tube combined with drive seals and bushings provides progressive water shutoff to cushion the drive piston at the end of its scram stroke. A stationary piston, called the stop piston, is mounted on the upper end of the piston tube. This piston provides the seal between reactor vessel pressure and the space above the drive piston. It also functions as a positive end stop at the upper limit of control rod travel. A stack of spring washers just below the stop piston helps absorb the final mechanical shock at the end of control rod travel on the BWR/4 design. In the BWR/6 design a buffer piston is included between the drive piston and the stop piston. This isolates the higher pressures from the drive piston seals during the deceleration phase of the scram stroke. The piston rings are similar to the outer drive piston rings. A bleed-off passage to the center of the piston tube is located between the two pairs of rings. This arrangement allows seal leakage from the reactor vessel (during a scram) to be bled directly to the discharge line, rather than to the space above the drive piston. The lower pair of seals is used only during the cushioning of the drive piston at the upper end of the stroke. Position Indicator The center tube of the drive mechanism forms a well to contain the position indicator probe. The position indicator probe is an aluminum extrusion attached to a cast 3.4-10

BFN-30 aluminum housing. Mounted on the extrusion are a series of hermetically sealed, magnetically operated, position indicator switches. Each switch is sheathed in a braided glass sleeve, and the entire probe assembly is protected by a thin-walled stainless steel tube. The switches are actuated by a ring magnet of Alnico 5 carried at the bottom of the drive piston. The drive piston, piston tube, and indicator tube are all of nonmagnetic stainless steel, allowing the individual switches to be operated by the magnet as the piston passes. One switch is located at each position corresponding to an index tube groove, thus allowing indication at each latching point. An additional switch is located at each midpoint between latching points, allowing indication of the intermediate positions during drive motion. Thus, indication is provided for each 3 inches of travel. Duplicate switches are provided for the full-in and full-out positions. One additional switch (an overtravel switch) is located at a position below the normal full-out position. Because the limit of down-travel is normally provided by the control rod itself as it reaches the backseat position, the index tube can pass this position and actuate the overtravel switch only if it is uncoupled from its control rod. A convenient means is thus provided to verify that the drive and control rod are coupled after installation of a drive or at any time during plant operation. A thermocouple is located in each position indicator to indicate drive mechanism temperature in the control room. This satisfies safety design basis 4.f. Cylinder, Tube, and Flange Assembly The cylinder, tube, and flange assembly consists of an inner cylinder, outer tube, and a flange. Both the cylinder tube and outer tube are welded to the drive flange. The tops of these tubes have a sliding fit to allow for differential expansion. A sealing surface on the upper face of this flange is used in making the seal to the drive housing flange. Teflon-coated, stainless steel "O" rings are used for these seals. In addition to the reactor vessel seal, the two hydraulic control lines to the drive are sealed at this face. A drive can thus be replaced without removing the control lines, which are permanently welded into the housing flange. The drive flange contains the integral ball or two-way check (shuttle) valve. This valve is so situated as to direct reactor vessel pressure or driving pressure, whichever is higher, to the underside of the drive piston. Reactor vessel pressure is admitted to this valve from the annular space between the drive and drive housing through passages in the flange. A screen is provided to intercept foreign material at this point. A cooling water orifice, adjacent to the insert port of the flange, permits cooling water flow from the CRD through the annulus formed by the CRD outer tube and the thermal sleeve in the CRD housing. Water used to operate the collet piston passes between the outer tube and the cylinder tube. The inside of the cylinder tube is honed to provide the surface required for the drive piston seals. 3.4-11

BFN-27 Coupling Spud, Plug, and Unlocking Tube The upper end of the index tube is threaded to receive a coupling spud. The coupling (Figure 3.4-3) is designed to accommodate a small amount of angular misalignment between the drive and the control rod. Six spring fingers allow the coupling spud to enter the mating socket on the control rod. The control rod weight (approximately 250 pounds) is sufficient to force the spud fingers to enter the socket and push the lock plug up, allowing the spud to enter the socket completely and the plug to snap back into place. However, with the lock plug in place, a force in excess of 50,000 pounds is required to pull the coupling apart. Two means of uncoupling are provided. With the reactor vessel head removed, the lock plug may be raised against the spring force of approximately 50 pounds by a rod extending up through the center of the control rod to an unlocking handle located above the control rod velocity limiter. The control rod, with the lock plug raised, can then be separated from the drive. The lock plug may also be pushed up from below to uncouple a drive without removing the reactor pressure vessel head for access to change a CRD drive or perform maintenance or inspections. In this case, the central portion of the drive mechanism is pushed up against the uncoupling rod assembly which raises the lock plug and allows the coupling spud to disengage the socket as the drive piston and index tube are driven down. The coupling spud and locking tube meet the requirements of safety design basis 4.e. 3.4.5.2.2 Materials of Construction Factors determining the choice of materials are listed below:

a. The index tube must withstand the locking and unlocking action of the collet fingers. A compatible bearing combination must be provided which is able to withstand moderate misalignment forces. The reactor environment limits the choice of materials suitable for corrosion resistance. The column and tensile loads can be satisfied by an annealed 300 series stainless steel. The wear and bearing requirements are provided by Malcomizing the completed tube.

To obtain suitable corrosion resistance, a carefully controlled process of surface preparation is employed. Index tubes for BWR/6 drives and replacement index tubes are made of XM-19.

b. The coupling spud is made of Inconel 750 which is aged to produce maximum physical strength and also provide the required corrosion resistance. As misalignment tends to produce a chafing in the semispherical contact area, the entire part is protected by a thin vapor-deposited chromium plating (Electrolyzing). This plating also serves to prevent galling of the threads attaching the coupling spud to the index tube.

3.4-12

BFN-27

c. Inconel 750 is used for the collet fingers, which must function as leaf springs when cammed open to the unlocked position. Colmonoy 6 hard facing is applied to the area contacting the index tube and unlocking cam surface of the guide cap to provide a long-wearing surface adequate for design life.

d. For BWR/4 CRDs, Graphitar 14 is selected for seals and bushings on the drive piston and stop piston. The material is inert and has a low friction coefficient when water lubricated.

For BWR/6 CRDs, a composite material of nickel, chrome, graphite and resin is selected for seals and bushings on the drive piston and stop piston. The material reduces the leakage due to excessive wear or premature breakage. The drive is supplied with cooling water to normally hold temperatures below 250 degree F. The CRD high temperature alarm set point is 350 degrees F based on the BWR/6 seals and bushings being more temperature resistant than the BWR/4 seals. The Graphitar is relatively soft, which is advantageous when an occasional particle of foreign matter reaches a seal. The resulting scratches in the seal reduce sealing efficiency until worn smooth, but the drive design can tolerate considerable water leakage past the seals into the reactor vessel. These seals determine the service life of the CRDM. All drive components exposed to reactor vessel water are made of AISI 300 series stainless steel except the following:

a. Seals and bushings on the drive piston and stop piston are Graphitar 14. For later model BWR/6 CRDs, seals and bushings on the drive piston and stop piston are a composite material of nickel, chrome, graphite and resin.

b. All springs and members requiring spring-action (collet fingers, coupling spud, and spring washers) are made of Inconel 750.

c. The ball check valve is a Haynes Stellite cobalt-base alloy.

d. Elastometric 0-ring seals are ethylene propylene.

e. Collet piston rings are Haynes 25 alloy.

f. Certain wear surfaces are hard faced with Colmonoy 6.

g. Nitriding by a proprietary New Malcomizing process, Electrolyzing (a vapor deposition of chromium), and chromium plating are used in certain areas where resistance to abrasion is necessary.

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BFN-27

h. The drive piston head is made of Armco 17-4Ph.

i. Replacement index tubes and piston tubes are made of grade XM-19 stainless steel.

j. BWR/6 drives also use XM-19 for the index tubes and piston tubes.

k. The buffer assembly for BWR/6 CRDs consists of the stop piston, buffer piston, seal ring, nut, locking cup and the buffer shaft which is secured to the top of the piston tube assembly. The materials used to fabricate these components are Inconel X-750, Inconel-600, Armco 17-4PH and Haynes 25.

Pressure containing portions of the drives are designed and built in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. 3.4.5.3 Control Rod Drive Hydraulic System (Figures 3.4-8a Sheets 1, 2, 3, 4 and 5, 3.4-8b, 3.4-8c, 3.4-8d, 3.4-8e, 3.4-8f, 3.4-8g, and 3.4-8h) The Control Rod Drive Hydraulic System supplies and controls the pressure and flow requirements to the drives. There is one supply subsystem which supplies water at the proper pressures and sufficient flow to the hydraulic control units (HCU's). Each HCU controls the flow to and from a drive. The water discharged from the drives during a scram flows through the HCU's to the scram discharge volume. The water discharged from a drive, during a normal control rod positioning operation, flows through its HCU and into the exhaust header. The discharged water then backflows through the other 184 CRD exhaust valves into the reactor vessel via the cooling water header. 3.4.5.3.1 CRD Hydraulic Supply and Discharge Subsystems (Figures 3.4-7, 3.4-8a Sheets 1, 2, 3, 4 and 5, 3.4-8c, 3.4-8d, 3.4-8e, 3.4-8f, 3.4-8g and 3.4-8h) The CRD hydraulic supply and discharge subsystems control the pressure and flows required for the operation of the control rod drive mechanisms. These hydraulic requirements identified by the function they perform are as follows:

a. An accumulator charging pressure of approximately 1400 to 1500 psig is required. Flow is required only during scram reset or during system startup.

b. Drive pressure of about 260 psi above reactor vessel pressure is required at a flow rate of approximately 4 gpm to insert a control rod and 2 gpm to withdraw a control rod during normal operation.

3.4-14

BFN-27

c. Cooling water to the drives is normally supplied at pressures greater than reactor pressure and at adequate flow rate to prevent drive component degradation due to elevated temperatures.

d. The exhaust water header is maintained at a pressure about 20 psig above vessel pressure to direct the flow of the water displaced during normal control operation of the drives back into the reactor vessel by backflowing through the other 184 CRD exhaust valves.

e. A scram discharge volume of approximately 3.3 gallons per drive to receive the water displaced from the drives during a scram is required. The scram discharge volume is vented and drained except during scram when it is isolated and filled with scram water until the scram signal is cleared and the scram reset. The scram discharge volume will reach reactor pressure following a scram.

The CRD hydraulic supply and discharge subsystems provide the required functions with the pumps, filters, valves, instrumentation, and piping shown in Figures 3.4-8a sheets 1, 2, 3, and 4, 3.4-8b, 3.4-8c, 3.4-8d, 3.4-8e, and 3.4-8f and described in the following paragraphs. Duplicate components are included, where necessary, to assure continuous system operation if an in-service component requires maintenance. Pumps One 100 percent capacity supply pump is provided for each unit to pressurize the system with water from the condensate storage system. One common 100 percent capacity spare pump is provided for Units 1 and 2. It can supply water to either control rod drive hydraulic systems. Unit 3's system is separate and has one spare 100 percent capacity pump. Change over (or selection) of the pumps is performed manually, either locally or from the main control room. Each pump is installed with a suction strainer and a discharge check valve to prevent bypassing flow backwards through the nonoperating pump. A minimum flow bypass connection between the discharge of the pumps and the condensate storage tank prevents overheating of the pumps in the event that the pump discharge is inadvertently closed. In addition, a portion of the CRD flow is directed to the recirculation and the reactor water cleanup pump bearing seals for pump seal cooling and Reactor Vessel Level Instrumentation System (RVLIS) reference leg back filler. Pump discharge pressure is indicated locally at the inlet to the drive water filters by a pressure indicator. 3.4-15

BFN-27 Filters Two parallel filters remove foreign material larger than 50 microns absolute (25 microns nominal) from the hydraulic supply subsystem water. The isolated filter can be drained, cleaned, and vented for reuse while the other is in service. A differential pressure indicator monitors the filter element as it collects foreign material. A strainer in the filter discharge line guards the hydraulic system in the event of filter element failure. Accumulator Charging Pressure The accumulator charging pressure is maintained automatically by a flow-sensing element, controller, and an air-operated flow control valve. During normal operation, the accumulator charging pressure is established upstream from the flow control valve by the restriction of the flow control valve. During scram, the flow-sensing system upstream of the accumulator charging header detects high flow in the charging header and closes the flow control valve. The flow control valve is closed so that the proper flow to recharge the accumulators is diverted from the hydraulic supply header to the accumulator charging header. The parallel spare valve is provided with isolation valves to permit maintenance of the noncontrolling valve. The pressure in the charging header is monitored in the control room with a pressure indicator and high pressure alarm. During normal operation, the constant flow established through the flow control valve is the sum of the maximum water required to cool all the drives and that amount of water needed to provide a stable hydraulic system for insertion and withdrawal of the mechanism. Drive Water Pressure The drive water pressure control valve, which is manually adjusted from the control room, maintains the required pressure in the drive water header. A flow rate of approximately 6 gpm (the sum of the flow rates required to insert and to withdraw a control rod) normally passes from the drive water pressure header through two solenoid-operated stabilizing valves (arranged in parallel) and then goes into the line downstream from the cooling pressure control valve. The two solenoid-operated stabilizing valves also have identical, backup function, solenoid-operated stabilizing valves which are selectable from the main control room in the event that the normal path valves become inoperative or require online maintenance. One stabilizing valve passes flow equal to the nominal drive insert flow; the other passes flow equal to the drive withdrawal flow. The appropriate stabilizing valve is closed 3.4-16

BFN-27 when operating a drive to divert the required flow to the drive. Thus, the flow through the drive pressure control valve is always constant. Flow indicators are provided in the drive water header and in the line downstream from the stabilizing valves, so that flow rate through the stabilizing valves can be adjusted. Differential pressure between the reactor vessel and the drive water pressure header is indicated in the control room. Cooling Water Pressure The cooling water header passes the flow from the drive water pressure control valve through the control rod drives and into the vessel. At normal flow rates, the cooling water header pressure will be approximately 10 to 15 psi above reactor vessel pressure. A differential pressure indicator in the control room indicates the difference between reactor vessel pressure and the drive cooling water pressure. Although the drives can function without cooling, the life of their seals is shortened by exposure to reactor temperatures. Exhaust Water Header The exhaust water header takes water discharged during a normal control rod positioning operation and directs it through the other CRD exhaust valves into the reactor vessel. If necessary, the exhaust water may be directed into the reactor vessel via the RWCU system by opening a normally closed valve. Scram Discharge System The scram discharge system is used to contain the reactor vessel water from all the drives during a scram. This system is provided in the two scram discharge volumes (SDVs) which each drain to a scram discharge instrument volume (SDIV). Water level monitors on the SDIVs provide an alarm if water is retained in the system. During normal plant operation, the volumes are empty with their drain and vent valves open. Upon receipt of a scram signal, the drain and vent valves close. Position indicator switches on the drain and vent valves indicate valve position by lights in the control room. During a scram, the scram discharge volume partially fills with water which is discharged from above the drive pistons. While scrammed, the control rod drive seal leakage continues to flow to the discharge volume until the discharge volume pressure equals reactor vessel pressure. There is a check valve in each HCU which prevents reverse flow from the scram discharge volume to the drive. When the 3.4-17

BFN-27 initial scram signal is cleared from the reactor protection system, the scram discharge volume scram signal may be overridden with the override switch and the scram discharge system drained. A control system interlock will not allow the drives to be withdrawn until the discharge system is emptied to a safe level. A test pilot valve allows the discharge volume valves to be tested without disturbing the reactor protection system. Closing the discharge volume valves allows the outlet scram valve seats to be leak tested by timing the accumulation of leakage inside the scram discharge volume. As an alternative to the test pilot valve in Units 1, 2 and 3, a key-lock test switch is provided to de-energize each of the SDV Drain and Vent Pilot Solenoid valves. This will allow stroke time testing of SDV Drain and Vent Valves to be performed. Six level switches on the scram discharge instrument volume, set at three different water levels, guard against operation of the reactor without sufficient free volume present in the scram discharge volume to receive the scram discharge water in the event of a scram. At the first (lowest) level, one level switch initiates an alarm for operator action. At the second level, one level switch initiates a rod withdrawal block to prevent further withdrawal of any control rod. At the third (highest) level, the four level switches (two for each reactor protection system trip system) initiate a scram to shut down the reactor while sufficient free volume is still present to receive the scram discharge. After a scram, these same level switches must be cleared by draining the scram discharge volume before reactor operation can be resumed. The piping and equipment pressure parts in the CRD hydraulic supply and discharge subsystems are designed in accordance with USAS B 31.1.0. 3.4.5.3.2 Hydraulic Control Units (Figures 3.4-7, 3.4-8a Sheets 2 and 4, 3.4-8c, 3.4-8e, and 3.4-11) Each hydraulic control unit controls a single CRD. The basic components in each hydraulic control unit are manual, pneumatic and electrically operated valves, an accumulator, filters, related piping, and electrical connections. Each hydraulic control unit furnishes pressurized water, upon signal, to a control rod drive. The drive then positions its control rod as required. Operation of the electrical system which supplies scram and normal control rod positioning signals to the hydraulic control unit is described in Subsection 7.7, "Reactor Manual Control System." 3.4-18

BFN-27 The basic components contained in each hydraulic control unit and their functions are as follows: Insert Drive Valve The insert drive valve is a solenoid-operated valve which opens on an insert or withdrawal signal to supply drive water to the bottom side of the main drive piston. Insert Exhaust Valve The insert exhaust valve is a solenoid-operated valve which opens on an insert or withdrawal signal to discharge water from above the drive piston to the exhaust header. Withdrawal Drive Valve The withdrawal drive valve is a solenoid-operated valve which opens on a withdrawal signal to supply drive water to the top side of the drive piston. Withdrawal Exhaust Valve The withdrawal exhaust valve is a solenoid-operated valve which opens on a withdrawal signal to discharge water from below the main drive piston to the exhaust header. Speed Control Valves The speed control valves, which regulate the control rod insertion and withdrawal rates during normal operation, are manually adjustable flow control valves used to regulate the water flow to and from the volume beneath the main drive piston. Once a speed control valve is properly adjusted, it is not necessary to readjust the valve except to compensate for changes in piston seal leakage. Scram Pilot Valves The scram pilot valves are operated from the Reactor Protection System Trip System. Two scram pilot valves control both the scram inlet valve and the scram exhaust valve. The scram pilot valves are identical, 3-way, solenoid-operated, normally energized valves. On loss of electrical signal to the pilot valves, the pressure ports are closed and the exhaust ports are opened on both pilot valves. The pilot valves are arranged as shown in Figures 3.4-7, 3.4-8a sheets 2 and 4, 3.4-8c, and 3.4-8e so that the trip system signal must be removed from both valves before air pressure is discharged from the scram valve operators. 3.4-19

BFN-27 Scram Inlet Valve The scram inlet valve is opened to supply scram water pressure to the bottom of the drive piston. The scram inlet valve is a globe valve which is opened by the force of an internal spring and system pressure and closed by air pressure applied to the top of its diaphragm operator. The opening force of the spring is approximately 700 pounds. The valve opening time is approximately 0.1 second from start to full open. Scram Exhaust Valve The scram exhaust valve opens slightly before the scram inlet valve, exhausting water from above the drive piston during a scram. Quicker opening times are achieved because of a greater spring force in the valve operator. Otherwise this valve is similar to the scram inlet valve. The scram inlet and scram exhaust valves have a position indicator switch which energizes a light in the control room as soon as both valves open. Scram Accumulator The scram accumulator stores sufficient energy to insert a control rod to the fully inserted position during a scram independent of any other source of energy. The accumulator consists of a water volume pressurized by a volume of nitrogen. The accumulator has a piston separating the water on top from the nitrogen below. A check valve in the charging line to each accumulator retains the water in the accumulator in the event supply pressure is lost. During normal plant operation, the accumulator piston has a differential pressure across it which is equal to the difference in the charging water pressure and the nitrogen cylinder pressure. Loss of nitrogen causes a decrease in the nitrogen pressure which actuates the pressure switch and sounds an alarm in the control room. Also, to ensure that the accumulator is always capable of producing a scram it is continuously monitored for water leakage. A float-type level switch actuates an alarm if water leaks past the barrier and collects in the accumulator instrumentation block. The accumulator instrumentation block is located below the accumulator (nitrogen side) in such a way that it will receive any water which leaks past the accumulator piston. The scram accumulator meets the requirements of safety design basis 4.d. 3.4-20

BFN-27 3.4.5.4 Control Rod Drive System Operation The control rod drive system performs three operational functions: rod insertion, row withdrawal, and scram. The functions are described below. Rod Insertion Rod insertion is initiated by a signal from the operator to the insert valve solenoids which opens both insert valves. The insert drive valve applies reactor pressure plus approximately 90 psig to the bottom of the drive piston. The insert exhaust valve allows water from above the drive piston to discharge to the exhaust header. As is illustrated in Figure 3.4-6, the locking mechanism is a ratchet-type device and does not interfere with rod insertion. The speed at which the drive moves is determined by the pressure drop through the insert speed control valve which is approximately 4 gpm for a shim speed (nonscram operation) of 3 inches per second. During normal insertion, the pressure on the downstream side of the speed control valve is 90 to 100 psi above reactor vessel pressure. However, if the drive slows down for any reason, the flow through and pressure drop across the insert speed control valve will decrease and the full drive water differential pressure will be available to cause continued insertion. With 250 psi differential pressure acting on the drive piston, the piston exerts an upward force of 1000 pounds. Rod Withdrawal Drive withdrawal is, by design, more involved. First the collet fingers (latch) must be raised to reach the unlocked position as in Figure 3.4-5. The notches in the index tube and the collet fingers are shaped so that the downward force on the index tube holds the collet fingers in place. The index tube must be lifted before the collet fingers can be released. This is done by opening the drive insert valves (in the manner described in the preceding paragraph) for approximately 1 second using an automatic sequence timer. The withdraw valves are then opened (by the sequence timer mechanism), applying driving pressure above the drive piston and opening the area below the piston to the exhaust header. Pressure is simultaneously applied to the collet piston. As the collet piston raises, the collet fingers are cammed outward, away from the index tube, by the guide cap. The pressure required to release the latch is set and maintained high enough to overcome the force of the latch return spring plus the force of reactor pressure opposing movement of the collet piston. When this occurs, the index tube is unlatched and free to move in the withdrawal direction. Water displaced by the drive piston flows out through the withdrawal speed control valve which is set to give the control rod a shim withdrawal of approximately 3 inches per second. The entire 3.4-21

BFN-27 valving sequence is automatically controlled and is initiated by a single operation of the rod withdraw switch. Rod Scram During a scram the scram pilot valves and scram valves are operated as previously described. With the scram valves open, accumulator pressure is admitted under the drive piston and the area over the drive piston is vented to the scram discharge volume. The large differential pressure (initially about 1400 psi and always several hundred psi depending on reactor vessel pressure) produces a large upward force on the index tube and control rod, giving the rod a high initial acceleration and providing a large margin of force to overcome any possible friction. The characteristics of the hydraulic system are such that, after the initial acceleration is achieved (approximately 30 milliseconds after start of motion), the drive continues at a fairly constant velocity of approximately 5 feet per second. This characteristic provides a high initial rod insertion rate. As the drive piston nears the top of its stroke, the piston seals close off the large passage in the stop piston tube and the drive slows down. In the BWR/6 design a buffer piston is included between the drive piston and the stop piston. This isolates the higher pressures from the drive piston seals during the deceleration phase of the scram stroke. Each drive requires about 2.5 gallons of water during the scram stroke. There is adequate water capacity in each drive's accumulator to complete a scram in the required time at low reactor vessel pressure. At higher reactor vessel pressures, the accumulator is assisted on the upper end of the stroke by reactor vessel pressure acting on the drive via the ball check (shuttle) valve. As water is forced from the accumulator, the accumulator discharge pressure falls below reactor vessel pressure. This causes the check valve to shift its position to admit reactor pressure under the drive piston. Thus, reactor vessel pressure furnishes the force needed to complete the scram stroke at higher reactor vessel pressures. When the reactor vessel is up to full operating pressure, the accumulator is actually not needed to meet scram time requirements. With the reactor at 1000 psig and the scram discharge volume at atmospheric pressure, the scram force without an accumulator is over 1000 pounds. 3.4-22

BFN-27 SCRAM TIMES (a)(b) (seconds) NOTCH POSITION REACTOR STEAM DOME PRESSURE

                                                             > 800 psig 46                                              0.45 36                                              1.08 26                                              1.84 06                                              3.36 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b) Scram times as a function of reactor steam dome pressure, when <800 psig are within established limits. 3.4.6 Safety Evaluation 3.4.6.1 Evaluation of Control Rods As discussed above, it has been determined that the control rods meet the design basis requirements. The description also indicates how the control rod-to-drive coupling unit meets design basis requirements. 3.4.6.2 Evaluation of Control Rod Velocity Limiter The control rod velocity limiter limits the free fall velocity of the control rod to a value which cannot result in nuclear system process barrier damage, as required by safety design basis 1.c. This velocity is evaluated by the rod drop accident analysis in Section 14, "Plant Safety Analysis." The following sequence of events is necessary to postulate an accident in which the control rod velocity limiter is required:

1. The rod-to-drive coupling fails.

2. The control rod sticks near the top of the core.

3. The drive is withdrawn and the control rod does not follow.

3.4-23

BFN-27

4. The operator fails to notice the lack of plant response as the control rod drive is withdrawn.

5. The control rod later becomes loose and falls freely to the fully withdrawn position.

3.4.6.3 Evaluation of Scram Time The rod scram function of the Control Rod Drive System provides the negative reactivity insertion which is required by safety design basis 2. The scram time shown in the description is adequate as shown by the transient analyses of Section 14, "Plant Safety Analysis." 3.4.6.4 Analysis of Malfunctions Relating to Rod Withdrawal There are no known single malfunctions which could cause even a single rod to withdraw. The following malfunctions have been postulated and the results analyzed:

a. Drive Housing Fails at Attachment Weld The bottom head of the reactor vessel has a penetration with an internal nozzle for each control rod drive location. A drive housing is raised into position inside each penetration and fastened to the top of the internal nozzle with a J-weld. The drive is raised into the drive housing and bolted to a flange at the bottom of the housing. The basic failure considered is a complete circumferential crack through the housing wall at an elevation just below the J-weld. The housing material is seamless type 304 stainless steel pipe with a minimum tensile strength of 75,000 psi.

Static loads on the housing wall include the weight of the drive and control rod, the weight of the housing below the attachment weld to the vessel nozzle, and reactor pressure acting on the 6-inch diameter cross-sectional area of the housing and the drive. Dynamic loading is due to the reaction force during drive operation. If the housing were to fail, as described above, the following sequence of events is foreseen. The housing would separate from the vessel and the control rod, the drive and the housing would be blown downward against the support structure by reactor pressure acting on the cross-sectional area of the housing and the drive. The amount of downward motion of the drive and associated parts would be determined by the gap between the bottom of the drive and the support structure and by the amount the support structure deflects under load. In the current design, maximum deflection is approximately 3 inches. If the collet were to remain latched, no further control rod ejection would occur.8 The housing would not drop far enough to clear the vessel 3.4-24

BFN-27 penetration. Reactor water would leak through the 0.06-inch diametral clearance between the housing o.d. and the vessel penetration i.d. at a rate of approximately 440 gpm. If the basic housing failure were to occur at the same time the control rod is being withdrawn (this is a small fraction of the total drive operating time), and if the collet were to stay unlatched, the housing would separate from the vessel, the drive and housing would be blown downward against the control rod drive housing support and calculations indicate that the steady state rod withdrawal velocity would be 0.3 ft/sec. During withdraw, pressure under the collet piston would be approximately 250 psi greater than the pressure over it. Therefore, the collet would be held in the unlatched position until driving pressure is removed from the pressure-over port.

b. Rupture of Either or Both Hydraulic Lines to a Drive Housing Flange (1) Pressure-Under Line Breaks In this case, a partial or complete circumferential opening is postulated at or near the point where the line enters the housing flange. Failure is more likely to occur after another basic failure wherein the drive housing, or housing flange, separates from the reactor vessel. Failure of the housing, however, does not necessarily lead directly to failure of the hydraulic lines.

If the pressure-under line were to fail, and if the collet were latched, no control rod withdrawal would occur. There would be no pressure differential across the collet piston in this case, and therefore no tendency to unlatch the collet. Consequently, it would not be possible to either insert or withdraw the control rod involved. If reactor pressure were to shift the drive ball check valve against its upper seat, the broken pressure-under line would be sealed off. If the ball check valve were to be prevented from seating, reactor water would leak to the atmosphere. Cooling water could not be supplied to the drive involved because of the broken line. Loss of cooling water would cause no immediate damage to the drive. However, prolonged drive exposure to temperatures at or near reactor temperature could lead to deterioration of material in the seals. High temperature would be indicated to the operator by the thermocouple in the position indicator probe. If the basic line failure were to occur at the same time the control rod is being withdrawn, and if the collet were to remain open, calculations indicate that the steady state control rod withdrawal velocity would be 2 ft/sec. In this case, however, there would not be sufficient hydraulic force to hold the collet open and spring force would normally cause the collet to latch, stopping rod withdrawal. 3.4-25

BFN-27 (2) Pressure-Over Line Breaks The failure considered is complete breakage of the pressure-over line at or near the point where the line enters the housing flange. If the line were to break, pressure over the drive piston would drop from reactor pressure to atmospheric pressure. If there were any significant reactor pressure (approximately 500 psig or greater) it would act on the bottom of the drive piston, and the drive would insert to the fully inserted position. Drive insertion would occur regardless of the operational mode at the time of the failure. After full insertion, reactor water would leak past the stop piston seals, the contracting seals on the drive piston and the collet piston seals. This leakage would exhaust to atmosphere through the broken pressure-over line. In an experiment to simulate this failure, a leakage rate of 80 gpm has been measured with reactor pressure at 1000 psi. If the reactor were hot, drive temperature would increase. The reactor operator would be apprised of the situation by indication of the fully inserted drive, by high drive temperature indicated and printed out on a recorder in the control room, and by operation of the drywell sump pump. (3) Coincident Breakage of Both Pressure-Over and Pressure-Under Lines This failure would require simultaneous occurrence of the failures described above. Pressures above and below the drive piston would drop to zero and the ball check valve would shift to close off the broken pressure-under line. Reactor water would flow from the annulus outside of the drive through the vessel ports to the space below the drive piston. As in the pressure-over line break case, the drive would then insert at a speed dependent on reactor pressure. Full insertion would occur regardless of the operational mode at the time of failure. Reactor water would leak past the drive seals and out of the broken pressure-over line to the atmosphere as described above. Drive temperature would increase. The reactor operator would be apprised of the situation by indication of the fully inserted drive, high drive temperature printed out and alarmed by a recorder in the control room, and by operation of the drywell sump pump.

c. All Drive Flange Bolts Fail in Tension Each control rod drive is bolted by eight cap screws to a flange at the bottom of a drive housing which is welded to the reactor vessel. Bolts are made of AISI-4140 steel.

Replacement cap screws are made of AISI-4340 or SA-540 B23 steel which is more resistant to stress corrosion cracking. In the event that progressive or simultaneous failure of all the bolts were to occur, the drive would separate from the housing and the control rod and the drive would be blown 3.4-26

BFN-27 downward against the support structure due to reactor pressure acting on the cross-sectional area of the drive. Impact velocity and support structure loading would be slightly less than in drive housing failure, since reactor pressure would act on the drive cross sectional area only and the housing would remain attached to the reactor vessel. The drive would be isolated from the cooling water supply. Reactor water would flow downward past the velocity limiter piston and through the large drive filter into the annular space between the thermal sleeve and the drive. For worst case leakage calculations, it is assumed that the large filter would be deformed or swept out of the way so that it would offer no significant flow restriction. At a point near the top of the annulus, where pressure has dropped to 350 psi, the water would flash to steam and choke-flow conditions would exist. Steam would flow down the annulus and out the space between the housing and the drive flanges to the atmosphere. Steam formation would limit the leakage rate to approximately 840 gpm. If the collet were latched, control rod ejection would be limited to the distance the drive can drop before coming to rest on the support structure. Since pressure below the collet piston would drop to zero, there would be no tendency for the collet to unlatch. Pressure forces, in fact, exert 1435 pounds to hold the collet in the latched position. If the bolt failure were to occur while the control rod is being withdrawn, pressure below the collet piston would drop to zero and the collet, with 1650 pounds return force, would latch, stopping rod withdrawal.

d. Weld Joining Flange to Housing Fails in Tension The failure considered is a crack in or near the weld joining the flange to the housing that extends through the wall and completely around the circumference of the housing so that the flange can separate from the housing. The flange material is a forged type 304 stainless steel and the housing material is seamless type 304 stainless steel pipe.

A conventional full penetration weld of type 308 stainless steel is used to join the flange to the housing. Minimum tensile strength is approximately the same as the parent metal. The design pressure is 1250 psig and the design temperature is 575F. A combination of reactor pressure acting downward on the cross-sectional area of the drive; the weight of the control rod, drive and flange; and the dynamic reaction force during drive operation result in a maximum tensile stress at the weld of approximately 6,000 psi. In the event that the basic failure described above were to occur, the flange and the attached drive would be blown downward against the support structure. The support structure loading would be slightly less severe than in drive housing failure, since reactor pressure would act only on the drive cross-sectional area. Since there would be no differential pressure across the collet piston, the collet would remain latched and control rod motion would be limited to approximately 3 inches. Downward drive movement would be small; therefore, most of the drive would remain inside the housing. The pressure-under and pressure-over lines are flexible enough to withstand the small 3.4-27

BFN-27 downward displacement and remain attached to the flange. Reactor water would follow the same leakage path described in c, above, except that the exit to the atmosphere would be through the gap between the lower end of the housing and the top of the flange. Water would flash to steam in the annulus surrounding the drive. The leakage rate would be approximately 840 gpm. If the basic flange-to-housing joint failure were to occur at the same time the control rod is being withdrawn (a small fraction of the total operating time), and if the collet were held unlatched, the flange would separate from the housing, the drive and flange would be blown downward against the support structure, and the calculated steady state rod withdrawal velocity would be 0.13 ft/sec. Since the pressure-under and pressure-over lines remain intact, driving water pressure would continue to be supplied to the drive and the normal exhaust line restriction would exist. The pressure below the velocity limiter piston would decrease below normal due to leakage out of the gap between the housing and the flange to the atmosphere. This differential pressure across the velocity limiter piston would result in a net downward force of approximately 70 pounds. However, leakage out of the housing would greatly reduce the pressure in the annulus surrounding the drive so that the net downward force on the drive piston would be less than normal. The overall effect would be a reduction of rod withdrawal speed to a value approximately one-half of normal speed. The collet would remain unlatched with a 560-psi differential across the collet piston, but should relatch as soon as the drive signal is removed.

e. Housing Wall Ruptures The failure considered in this case is a vertical split in the drive housing wall just below the bottom head of the reactor vessel. The hole was considered to have a flow area equivalent to the annular area between the drive and the thermal sleeve so that flow through this annular area, rather than flow through the hole in the housing, would govern leakage flow. The housing is made from type 304 stainless steel seamless pipe.

If the housing wall rupture described above were to occur, reactor water would flash to steam and leak to the atmosphere at approximately 1030 gpm through the hole in the housing. Choke flow conditions described in c, above would exist. In this case, however, the leakage flow would be greater because the flow resistance is less; that is, the leaking water and steam would not have to flow down the length of the housing to reach the atmosphere. Critical pressure at which the water would flash to steam is 350 psi. There would be no pressure differential across the collet piston tending to cause collet unlatching, but the drive would insert due to loss of pressure in the drive housing and, therefore, in the space above the drive piston. 3.4-28

BFN-27 If the basic housing wall failure were to occur at the same time the control rod is being withdrawn (a small fraction of the total operating time), the drive would stop withdrawing, but the collet would remain unlatched. The drive stoppage would be caused by a reduction in the net downward force acting on the drive line. This would occur when the leakage flow of 1030 gpm reduces the pressure in the annulus outside the drive approximately 540 psig and therefore reduces the pressure acting on the top of the drive piston to this value. There would be a pressure differential of approximately 710 psi across the collet piston, holding the collet unlatched as long as the operator held the withdraw signal.

f. Flange Plug Blows Out A 3/4-inch-diameter hole is drilled in the drive flange to connect the vessel ports with the bottom of the ball check valve. The outer end of this hole is sealed with an 0.812-inch-diameter plug 0.250 inch thick. The plug is held in place with a full-penetration weld of type 308 stainless steel. The failure considered is a full circumferential crack in this weld and subsequent blow-out of the plug.

If the weld were to fail and the plug were to blow out, there would be no control rod motion provided the collet were latched. There would be no pressure differential across the collet piston tending to cause collet unlatching. Reactor water would leak past the velocity limiter piston, down the annulus between the drive and the thermal sleeve through the vessel ports and drilled passage and out the open plug hole to the atmosphere at approximately 320 gpm. This leakage calculation is based on liquid only exhausting from the flange as a worst case. Actually, hot reactor water would flash to steam, and choke-flow conditions would exist, so that the expected leakage rate would be lower than the calculated value. Drive temperature would rise, and the alarm would signal the operator. If the basic plug weld failure were to occur at the same time the control rod is being withdrawn (a small percentage of the total operating time), and if the collet were to stay unlatched, calculations indicate that control rod withdrawal speed would be approximately 0.24 ft/sec. Leakage out of the open plug hole in the flange would cause reactor water to flow downward past the velocity limiter piston. The small differential pressure across the piston would result in an insignificant driving force of approximately 10 pounds tending to increase withdraw velocity. The collet would be held unlatched by a 295 psi pressure differential across the collet piston as long as the driving signal was maintained. The exhaust path from the drive would have normal flow resistance since the ball check valve would be seated at the lower end of its travel by pressure under the drive piston. 3.4-29

BFN-27

g. Pressure Regulator and Bypass Valves Fail Closed (Reactor Pressure O psig)

Pressure in the drive water header supplying all drives is controlled by regulating the amount of water from the supply pump that is bypassed back to the reactor. This is accomplished primarily with the drive water control valves, and secondarily with the pressure stabilizing valves. There are two drive water control valves arranged in parallel. One is a motor-operated valve that can be adjusted from the control room. This valve is normally in service and is partially open to maintain a pressure of reactor pressure plus 260 psig in the header just upstream from the valve. The other is a hand-operated valve that is normally closed but that can be valved in and operated locally whenever the motor-operated valve is out of service. The pressure stabilizing valves are solenoid-operated and have built-in needle valves for adjusting flow. The two valves are arranged in parallel between the drive water header and the cooling water header. The two solenoid-operated stabilizing valves also have identical, backup function, solenoid-operated stabilizing valves which are selectable from the main control room in the event that the normal path valves become inoperative or require online maintenance. One valve is set to bypass 2 gpm, and closes when any drive is given a withdraw signal, so that flow is diverted to the drive being operated rather than back to the reactor. Relatively constant header pressure is thus maintained. Similarly, the other valve is set to bypass 4 gpm, and closes when any drive is given an insert signal. The failure considered is when all of these valves are closed so that maximum supply pump head of 1700 psi builds up in the drive water header. The major portion of the bypass flow normally passes through the motor-operated valve; therefore, closure of this valve is most critical. Since lowest exhaust line pressure exists when reactor pressure is zero, this reactor condition is also assumed. If the valve closure failure described above were to occur at the same time the control rod is being withdrawn, calculations indicate that steady-state withdrawal speed would be approximately 0.5 ft/sec or twice normal velocity. The collet would be held unlatched by a 1670-psi pressure differential across the collet piston. Flow would be upward past the velocity limiter piston, but retarding force would be negligible.

h. Ball Check Valve Fails to Close Off Passage to Vessel Ports The failure considered in this case depends upon the following sequence of events. If the ball check valve were to seal off the passage to the vessel ports during the "up"-signal portion of the job withdraw cycle, the collet would be unlatched. This is the normal withdrawal sequence. Then if the ball were to move up and become jammed in the ball cage by foreign material or prevented from reseating at the bottom by foreign material that settles out on the seat surface, water from below the drive piston would return to the reactor through the vessel ports and the annulus between the drive and the 3.4-30

BFN-27 housing. Since this return path would have lower than normal flow resistance, the calculated withdrawal speed would be 2 ft/sec. During withdrawal, there would be differential pressure across the collet piston of approximately 40 psi. Therefore, the collet would tend to latch and would have to stick open before continuous withdrawal at 2 ft/sec could occur. Water would flow upward past the velocity limiter piston and a small retarding force would be generated (approximately 120 pounds).

i. Hydraulic Control Unit Valve Failures Various failures of the valves in the HCU can be postulated, but none are capable of producing differential pressures which approach those described in the preceding paragraphs and none are capable alone of producing a high velocity withdrawal.

Leakage through either or both of the scram valves produces a pressure which tends to insert the control rod rather than withdraw it. If the pressure in the scram discharge volume should exceed reactor pressure following a scram, a check valve in the line to the scram discharge header prevents this pressure from operating the drive mechanisms.

j. Failure of the Collet Fingers to Latch The drive continues to withdraw (after removal of the signal) at a fraction of its normal withdrawal speed. There is no known means for the collet fingers to become unlocked without some initiating signal. Failure of the withdrawal drive valve to close following a rod withdrawal has the same effect as failure of the collet fingers to latch in the index tube and is immediately apparent to the operator. Accidental opening of the withdrawal drive valve normally does not unlock the collet fingers because of the characteristic of the collet fingers to remain locked until unloaded.

k. Withdrawal Speed Control Valve Failure Normal withdrawal speed is determined by differential pressures at the drive and set for a nominal value at 3 in./sec. The characteristics of the pressure regulating system are such that withdrawal speed is maintained independent of reactor vessel pressure.

Tests have determined that accidental opening of the speed control valve to the full open position produces a velocity of approximately 6 in./sec. The Control Rod Drive System prevents rod withdrawal as required by safety design basis 4.a. It is shown above that only multiple failures in a drive unit and its control unit could cause an unplanned rod withdrawal. 3.4-31

BFN-27 3.4.6.5 Scram Reliability High scram reliability is the result of a number of features of the CRD System, such as the following:

a. There are two sources of scram energy to insert each control rod when the reactor is operating: accumulator pressure and reactor vessel pressure.

b. Each drive mechanism has its own scram and pilot valve so that only one drive can be affected by failure of a scram valve to open. Two pilot valves are provided for each drive. Both pilot valves must be vented to initiate a scram.

c. The Reactor Protection System and HCU's are designed so that the scram signal and mode of operation override all others.

d. The collet assembly and index tube are designed so that they will not restrain or prevent control rod insertion during scram.

e. The scram discharge volume is monitored for accumulated water and will scram the reactor before the volume is filled to a point that could interfere with a scram.

The scram reliability meets the requirements of safety design basis 4.b and 4.c. 3.4.6.6 Control Rod Support and Operation As shown in the description, each control rod is independently supported and controlled as required by safety design basis 3. 3.4.7 Inspection and Testing 3.4.7.1 Development Tests The development drive (one prototype) testing included over 5000 scrams and approximately 100,000 latching cycles during 5000 hours of exposure to simulated operating conditions. These tests have demonstrated the following:

a. That the drive withstands the forces, pressures, and temperatures imposed without difficulty.

b. That wear, abrasion, and corrosion of the nitrided type 304 stainless parts are negligible. That mechanical performance of the nitrided surface is superior to materials used in earlier operating reactors.

3.4-32

BFN-27

c. That the basic scram speed of the drive has a satisfactory margin above minimum plant requirements at any reactor vessel pressure.

d. That usable seal lifetimes greater than 1000 scram cycles may be expected.

3.4.7.2 Factory Quality Control Tests Quality control of welding, heat treatment, dimensional tolerances, material verification, etc., was maintained throughout the manufacturing process to assure reliable performance of the mechanical reactivity control components. Some of the quality control tests on the control rods, control rod drive mechanisms, and hydraulic control units were as follows: Control Rod Absorber Tube Tests

a. The tubing and end plug material integrity was verified by ultrasonic inspection.

b. Boron content of the Boron-10 fraction of each lot of boron-carbide was verified.

c. The weld integrity of the finished absorber tubes was verified by helium leak testing.

CRD Mechanism Tests

a. Hydrostatic testing of the drives to check pressure welds was in accordance with ASME codes.

b. Electrical components were checked for electrical continuity and resistance to ground.

c. All drive parts which could not be visually inspected for dirt were flushed with filtered water at high velocity. No significant foreign material was permissible in effluent water.

d. Seal leakage tests were performed to demonstrate proper seal operation.

e. Each drive was tested for shim motion, latching, and control rod position indicating.

f. Each drive was subjected to cold scram tests at various reactor pressures to verify proper scram performance.

3.4-33

BFN-27 Hydraulic Control Unit Tests Each HCU received the following tests:

a. All hydraulic systems were hydrostatically tested in accordance with USAS B31.1.0.

b. All electrical components and systems were tested for electrical continuity and resistance to ground.

c. The correct operation of the accumulator pressure and level switches was verified.

d. The unit's ability to perform its part of a scram was demonstrated.

e. Proper operation and adjustment of the insert and withdrawal valves was demonstrated.

3.4.7.3 Operational Tests After installation, all rods, hydraulic control units, and drive mechanisms were tested through their full range for operability. Details of the preoperational test are given in Subsection 13.4. During normal operation, each time a control rod is withdrawn a notch, the operator can observe the in-core monitor indications to verify that the control rod is following the drive mechanism. All control rods that are partially withdrawn from the core can be tested for rod following by inserting or withdrawing the rod one notch and returning it to its original position, while the operator observes the in-core monitor indications. To make a positive test of control rod to control rod drive coupling integrity, the operator can withdraw a control rod to the end of its travel and then attempt to withdraw the drive to the overtravel position. Failure of the drive to overtravel demonstrates rod-to-drive coupling integrity. Hydraulic supply subsystem pressures can be observed from instrumentation in the control room. Scram accumulator pressures can be observed locally on the nitrogen pressure gages. 3.4-34

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9.75" U3" COUPLIIIG 1/IElEASE MMDL£ AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL S,V£TY AHA.LYSIS REPORT COUPLING SOCKET Control Rod - Isometric FIGURE 3.4-2

CONTROL ROO ASSEMBLY w/VELOClTY LIMITER

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INSERT ftUIClff unu GIIIOITUII WITHDRAW AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT CONTROL ROD VOLOCITY LIMITOR-ISOMETRIC FIGURE 3.4-4

80TTOII or HACTOR VESSEL CRIVE HOUSING DRIVE WITHilRAW DRJVt INSERT LIN£ - - - - - - - . LINE ARROIIS SNOl 'fATEI FLOW Wtt£ft THE OAIVE 1$ Ill THE l'ITHDR.UAL IIQOE OF.OPtRllTION. HLl CKECM P~ESSURES SHOllt ARE IIAXIUUII. VAL¥£ PR*REACTOR PRE!SURt BROWNS FERRY NUCLEAR PLANT FlHAL SAFETY ANALYSIS REPORT Control Rod Drive, Simplified Component Illustration AMENDMENT 16 FIGURE 3.4-6 I

GUIOE TUBE NOTE: A CIRCLEO NUMBER INDICATES A PRESSURE POINT. n 0 TTER PIN  ;! ET-FINGERSr~ LET-SPRING EHiSJDH RINGS PISTON _ _ __ DRIVE TUBE Pl5TON ANO~H,LS PRESSURE .HOUSI~ TO F'LANGE IIELD OVER PORT FLANGE HUB DRIVE PRESSURE WITHDRAW LINE UNDER PORT HOUSING DRIVE INSERT fl.ANGE LINE BALL CHECK VALVE VESSEL PORTS DRIVE MAIN FLANGE WELDtD PLUG POSITION INDICATOR HOUSING AMENDMENT 16 BROWNS FERRY NUCLEAR PLAHT FIN.AL SAFETY ANALYSIS REPORT Control Rod Drhre, Schematic Diagram FIGURE 3.4-6

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                                                                                                                                                          !ie~                                                                                                                                      CRD HYDRAULIC SYSTEM MECHANICAL CONTROL DIAGRAM FIGURE 3 . 4-Sh 10

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• POSlTIOH INDICATOR SW!TCHES 80 - BUFFER SHAFT PORTS BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT AMENDMENT 16 MODIFIED CONTROL. ROD DRIVE (BWR/61 L__ _ _ _ _ _ _ _ _ _ _ ___:___ _ _ _ _ _ _ _ _ ___i..______;...:..;;;.;;.;.~;.;;._---1 FIGURE 3.4-9a ,

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BFN-16 Figure 3.4-10 Deleted by Amendment 7.

BFN-22 Figure 3.4-11 (Deleted by Amendment 22)

HANDLE 0 () BLADE NEUTRON ABSORBER RODS COUPLING RELEASE HANDLE VELOCITY LIMITER COUPLING SOCKET AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Marathon Control Rod - Isometric FIGURE 3.4-12

ADMENDMENT 20 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANAL YS!S REPORT WESTINGHOUSE CR-82M-l CONTROL ROD FIGURE 3.4-13

BFN-16 3.5 CONTROL ROD DRIVE HOUSING SUPPORTS 3.5.1 Safety Objective The control rod drive housing supports protect against additional damage to the nuclear system process barrier or damage to the fuel barrier by preventing any significant nuclear transient in the event a drive housing breaks or separates from the bottom of the reactor vessel. 3.5.2 Safety Design Basis

1. Control rod downward motion shall be limited, following a postulated control rod drive (CRD) housing failure, so that any resulting nuclear transient could not be sufficient to cause fuel damage or additional damage to the process barrier.

2. Clearance shall be provided between the housings and the supports to prevent vertical contact stresses due to their respective thermal expansion during plant operation.

3.5.3 Description The control rod drive housing supports are illustrated in Figure 3.5-1. Horizontal beams are installed immediately below the bottom head of the reactor vessel between the rows of control rod housings, and are bolted to brackets welded to the steel form liner of the reactor support pedestal. Hanger rods, about 10 feet long by 1-3/4 inches in diameter, are supported from the beams on stacks of disc springs which compress about 2 inches under the design load. The support bars are bolted between the bottom ends of the hanger rods. The spring pivots at the top and the beveled loose-fitting ends on the support bars prevent substantial bending moment in the hanger rods if the support bars are ever loaded. Individual grids rest on the support bars between adjacent beams. Because a single-piece grid would be difficult to handle in the limited work space and because it is necessary that control rod drive, position indicators, and in-core instrumentation components be accessible for inspection and maintenance, each grid is designed to be assembled or disassembled in place. Each grid assembly is made from two grid plates, a clamp and a bolt. The top part of the clamp acts as a guide to assure that each grid is correctly positioned directly below the respective CRD housing which it would support in the postulated accident. 3.5-1

BFN-17 When the support bars and grids are installed, a 1 +/- .25 gap is provided between the grid and the bottom contact surface of the control rod drive flange. During system heatup this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, a minimum gap of 7/16 is maintained. In the postulated CRD housing failure, the CRD housing supports are loaded when the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hanger rods, their disc springs, and two adjacent beams. The American Institute of Steel Construction (AISC), "Specification for the Design, Fabrication, and Erection of Structural Steel for Building," was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90 percent of yield, and the shear stress as 60 percent of yield. These are 1.5 times the corresponding AISC allowable stresses of 60 percent and 40 percent of yield. This stress criterion is considered desirable for this application and adequate for the "once in a lifetime" loading condition. For mechanical design purposes, the postulated failure resulting in the highest forces is an instantaneous circumferential separation of the CRD housing from the reactor vessel, with an internal pressure of 1250 psig (reactor vessel design pressure) acting on the area of the separated housing. The weight of the separated housing, control rod drive, and blade, plus the force of 1250 psig pressure acting on the area of the separated housing, gives a force of approximately 35,000 pounds. This force is multiplied by a factor of 3 for impact, conservatively assuming the housing travels through a 1-inch gap before contacting the supports. The total force (10.5 x 104 pounds) is then treated as a static load in design formulas. The control rod drive housing supports are designed as Class I equipment in accordance with Appendix C, "Structural Qualification of Subsystems and Components". All control rod drive housing support subassemblies are fabricated of ASTM-A-36 structural steel, except for the following: Grid ASTM-A-441 Disc springs Schnorr Type BS-125-71-8 Hex bolts and nuts ASTM-A-307 3.5-2

BFN-17 3.5.4 Safety Evaluation Downward travel of a CRD housing and its control rod following the postulated housing failure is the sum of the compression of the disc springs under dynamic loading and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the approximate 2-inch spring compression plus the installed 1 +/- .25 gap. If the reactor were hot and pressurized, the gap would be greater than a 7/16 and the spring compression slightly less than in the cold condition. In either case, the control rod movement following a housing failure is limited substantially below one drive "notch" movement (6 inches). The nuclear transient from sudden withdrawal of any control rod through a distance of one drive notch at any position in the core does not result in a transient sufficient to cause damage to any radioactive material barrier. This meets the fuel damage limitation of safety design basis 1. The control rod drive housing supports are in place any time the reactor is to be operated. At plant operating temperature, a minimum gap of 7/16 is maintained between the CRD housing and the supports, at lower temperatures the gap is greater. Because the supports do not come in contact with any of the CRD housings, except during the postulated accident condition, vertical contact stresses are prevented as required by safety design basis 2. 3.5.5 Inspection and Testing The control rod drive housing supports were inspected after initial installation. When the reactor is in the shutdown mode (MODE 4 or MODE 5), the control rod drive housing supports may be removed to permit inspection and maintenance of the control rod drives. When the support structure is reinstalled, it is inspected for proper assembly, particular attention being given to assure that the correct gap between the CRD flange lower contact surface and the grid is maintained. 3.5-3

STEEL FORIILINElt

    ..... .-*              CRD HOUSIIG 1--tt--+--ltMGEI IIOD GRID CUMP AMENDMENT 16 BROWNS FERRY NUCLIAR PLAtlT PINAL SAl'ln .lMAL TSIS REPORT C-1"f Rod Drift HOUli119 &IPSIO"
                                       *-*c FIGURE 3.5-1

BFN-28 3.6 NUCLEAR DESIGN This section describes the nuclear core design basis and the models used to analyze the fuel discussed in Subsection 3.2, Fuel Mechanical Design. The nuclear design criteria of AREVA fuel is described in Reference 15 with the detailed design provided in a reload specific fuel nuclear design report. 3.6.1 Power Generation Objective The objectives of the fuel nuclear design are as follows:

a. To attain rated power generation from the nuclear fuel for a given period of time,

b. To attain reactor nuclear stability throughout core life,

c. To allow normal power operation of the nuclear fuel without sustaining fuel damage.

3.6.2 Power Generation Design Basis

1. Fuel nuclear design shall provide sufficient excess reactivity during power operation to achieve the core design burnup.

2. Fuel nuclear design shall provide sufficient negative reactivity feedback to facilitate normal maneuvering and control.

3. Fuel nuclear design shall, in combination with reactivity control systems, allow continuous, stable regulation of core excess reactivity.

3.6.3 Safety Design Basis The design bases are those that are required for the plant to operate, meeting all safety requirements. Safety design bases fall into two categories: (1) the reactivity basis, which prevents an uncontrolled positive reactivity excursion, and (2) the overpower bases, which prevent the core from operating beyond the fuel integrity limits. 3.6.3.1 Reactivity Basis The nuclear design shall meet the following basis: The core shall be capable of being made subcritical at any time or at any core condition with the highest worth control rod fully withdrawn. 3.6-1

BFN-28 3.6.3.2 Overpower Bases The Technical Specification limits on Minimum Critical Power Ratio (MCPR), the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR), and the Maximum Linear Heat Generation Rate (MLHGR) are determined such that the fuel will not exceed required licensing limits during abnormal operational occurrences or accidents. 3.6.4 Description The BWR core design consists of a light-water moderated reactor, fueled with slightly enriched uranium-dioxide. The use of water as a moderator produces a neutron energy spectrum in which fissions are caused principally by thermal neutrons. At normal operating conditions, the moderator boils, producing a spatially variable distribution of steam voids in the core. The BWR design provides a system for which reactivity is reduced by an increase in the steam void content in the moderator. Void feedback effects are one inherent safety feature of the BWR. Any system input which increases reactor power, either in a local or gross sense, produces additional steam voids which reduce reactivity and thereby reduce the power. 3.6.4.1 Nuclear Design Description The reference loading pattern for each cycle is documented by the vendor specific reload licensing analysis report. The current licensing report for each BFN unit is included in Appendix N of the FSAR. The reference loading pattern is the basis for all fuel licensing. It is designed with the intent that it will represent, as closely as possible, the actual core loading pattern; however, there will be occurrences where the number and/or types of bundles in the reference design and the actual core loading do not agree exactly. Any differences between the reference loading pattern and the actual loading pattern are evaluated as described in Reference 12 for AREVA analyzed reload cores. To assure that licensing calculations performed on the reference core are applicable to the as-loaded core, certain key parameters, which affect the licensing calculations, are examined to assure that there is no adverse impact. If the final loading plan does not meet the necessary criteria, a re-examination of the parameters that determine the operating limits is performed. Only when this examination has been completed and it has been established that the as-loaded core satisfies the licensing basis will the core be operated. 3.6-2

BFN-28 3.6.4.2 Power Distribution The core power distribution is a function of fuel bundle design, core loading, control rod pattern, core exposure distribution, and core coolant flow rate. The thermal performance parameters MAPLHGR, MLHGR, and MCPR (defined in Table 3.6-1) limit unacceptable core power distributions. 3.6.4.2.1 Power Distribution Measurements The techniques for measurement of the power distribution within the reactor core, together with instrumentation correlations and operation limits, are discussed in Reference 11 for AREVA analyzed cores. 3.6.4.2.2 Power Distribution Accuracy The accuracy of the calculated power distribution is discussed in References 11 and 13 for AREVA analyzed reload cores. 3.6.4.2.3 Power Distribution Anomalies Stringent inspection procedures are utilized to ensure the correct arrangement of the core following fuel loading. A fuel loading error (a mislocated or a misoriented fuel bundle in the core) would be a very improbable event, but calculations have been performed to determine the effects of such events on CPR and LHGR. The fuel loading error is discussed further References 12 and 17. The inherent design characteristics of the BWR are well suited to limit gross power tilting. The stabilizing nature of the large moderator void coefficient effectively reduces the effect of perturbations on the power distribution. In addition, the in-core instrumentation system, together with the on-line computer, provides the operator with prompt information on the power distribution so that he can readily use control rods or other means to limit the undesirable effects of power tilting. Because of these design characteristics, it is not necessary to allocate a specific margin in the peaking factor to account for power tilt. If, for some reason, the power distribution could not be maintained within normal limits using control rods and flow, then the total core power would have to be reduced. 3.6.4.3 Reactivity Coefficients Reactivity coefficients, the differential changes in reactivity produced by differential changes in core conditions, are useful in calculating stability and evaluating the response of the core to external disturbances. The base initial condition of the system and the postulated initiating event determine which of the several defined 3.6-3

BFN-28 coefficients are significant in evaluating the response of the reactor. The coefficients of interest, relative to BWR systems, are discussed here individually. There are two primary reactivity coefficients that characterize the dynamic behavior of boiling water reactors; these are the Doppler reactivity coefficient and the moderator void reactivity coefficient. Also associated with the BWR is a power reactivity coefficient and a temperature coefficient. The power coefficient is a combination of the Doppler and void reactivity coefficients in the power operating range, and the temperature coefficient is merely a combination of the Doppler and moderator temperature coefficients. Power and temperature coefficients are not specifically calculated for reload cores. 3.6.4.3.1 Doppler Reactivity Coefficient The Doppler coefficient is of prime importance in reactor safety. The Doppler coefficient is a measure of the reactivity change associated with an increase in the absorption of resonance-energy neutrons caused by a change in the temperature of the material in question. The Doppler reactivity coefficient provides instantaneous negative reactivity feedback to any rise in fuel temperature, on either a gross or local basis. The magnitude of the Doppler coefficient is inherent in the fuel design and does not vary significantly among BWR designs. For most structural and moderator materials, resonance absorption is not significant, but in U-238 and Pu-240 an increase in temperature produces a comparatively large increase in the effective absorption cross-section. The resulting parasitic absorption of neutrons causes a significant loss in reactivity. In BWR fuel, in which approximately 97% of the uranium in UO2 is U-238, the Doppler coefficient provides an immediate negative reactivity response that opposes increased fuel fission rate changes. Although the reactivity change caused by the Doppler effect is small compared to other power-related reactivity changes during normal operation, it becomes very important during postulated rapid power excursions in which large fuel temperature changes occur. The most severe power excursions are those associated with rod drop accidents. A local Doppler feedback associated with a 3000°F to 5000°F temperature rise is available for terminating the initial excursion. The application of the Doppler coefficient to the AREVA analysis of the rod drop accident is discussed in Reference 12. 3.6.4.3.2 Moderator Void Coefficient The moderator void coefficient should be large enough to prevent power oscillation due to spatial xenon changes yet small enough that pressurization transients do not unduly limit plant operation. In addition, the void coefficient in a BWR has the ability to flatten the radial power distribution and to provide ease of reactor control due to 3.6-4

BFN-28 the void feedback mechanism. The overall void coefficient is always negative over the complete operating range since the BWR design is under moderated. A detailed discussion of the methods used to calculate void reactivity coefficients, their accuracy, and their application to plant transient analyses is presented in References 11 and 12 for AREVA analyzed cores. 3.6.4.4 Control Requirements The BWR control rod system is designed to provide adequate control of the maximum excess reactivity anticipated during the plant operation. The shutdown capability is evaluated assuming a cold, xenon-free core. 3.6.4.4.1 Shutdown Reactivity The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn and all other rods fully inserted. The shutdown margin is determined by using a BWR simulator code (see Subsection 3.6.5, Analytical Methods) to calculate the core multiplication at selected exposure points with the strongest rod fully withdrawn. The shutdown margin is calculated based on the carryover of the minimum expected exposure at the end of the previous cycle. The core is assumed to be in the cold, xenon-free condition in order to ensure that the calculated values are conservative. Further discussion of the uncertainty of these calculations is given in References 11 and 12 for AREVA analyses. As exposure accumulates and burnable poison depletes in the lower exposure fuel bundles, an increase in core reactivity may occur. The nature of this increase depends on specifics of fuel loading and control state. The cold keff is calculated with the strongest control rod out at various exposures through the cycle. A value R is defined as the difference between the strongest rod out keff at BOC and the maximum calculated strongest rod out keff at any exposure point. The strongest rod out keff at any exposure point in the cycle is equal to or less than: keff = keff(Strongest rod withdrawn)BOC + R

where, R is always greater than or equal to 0. The value of R includes equilibrium Sm.

3.6-5

BFN-26 3.6.4.4.2 Reactivity Variations The excess reactivity designed into the core is controlled by the control rod system supplemented by gadolinia-urania fuel rods. Control rods are used during the cycle partly to compensate for burnup and partly to control the power distribution. 3.6.4.4.3 Standby Liquid Control System The Standby Liquid Control System (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from a full power and minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon-free state. The requirements of this system are dependent primarily on the reactor power level and on the reactivity effects of voids and temperature between full-power and cold, xenon-free conditions. The shutdown capability of the SLCS is reported in the AREVA Reload Licensing Analysis Report. 3.6.4.5 Criticality of Reactor During Refueling The core is subcritical at all times during refueling. This is ensured by a combination of refueling interlocks and analytical verification of shutdown margin. Shutdown margin is determined by using a BWR simulator code (see Subsection 3.6.5, Analytical Methods) to calculate the core multiplication with the strongest rod fully withdrawn for the final reload core configuration and for limiting interim core configurations in the case of an incore shuffle. 3.6.4.6 Stability 3.6.4.6.1 Xenon Transients Boiling water reactors do not have instability problems due to xenon. This has been demonstrated by: (1) Never having observed xenon instabilities in operating BWRs (2) Special tests which have been conducted on operating BWRs in an attempt to force the reactor into xenon instability (3) Calculations All of these indicators have proven that xenon transients are highly damped in a BWR due to the large negative power coefficient. Analyses and experiments conducted in this area are reported in Reference 8. 3.6-6

BFN-28 3.6.4.6.2 Thermal Hydraulic Stability The compliance of GE fuel designs to the criteria set forth in General Design Criterion 12 is demonstrated provided that the following stability compliance criteria are satisfied using approved methods: (1) Neutron flux limit cycles, which oscillate up to 120% APRM high neutron flux scram setpoint or up to the LPRM upscale alarm trip (without initiating scram) prior to operator mitigating action shall not result in exceeding specified acceptable fuel design limits. (2) The individual channels shall be designed and operated to be hydrodynamically stable or more stable than the reactor core for all expected operating conditions (analytically demonstrated). The AREVA for demonstrating the above has been reviewed and approved by the NRC in Reference 14. See Subsection 3.7.6.2, Thermal Hydraulic Stability Performance, for additional information regarding core thermal-hydraulic stability. 3.6.5 Analytical Methods AREVA nuclear evaluations are performed using the analytical tools and methods described in References 11 and 12. The lattice analyses are performed during the bundle design process. The results of these single bundle calculations are reduced to libraries of lattice reactivities, relative rod powers, and few group cross-sections as functions of instantaneous void, exposure, exposure-void history, exposure-control history, control state, and fuel and moderator temperature, for use in the core analysis. These analyses are dependent upon fuel lattice parameters only and are, therefore, valid for all plants and cycles to which they are applied. The core analysis is unique for each cycle. It is performed in the months preceding the cycle loading to demonstrate that the core meets all applicable safety limits. The principal tool used in the core analysis is a three-dimensional BWR simulator code, which computes power distributions, exposure, and reactor thermal-hydraulic characteristics, with spatially varying voids, control rods, burnable poisons, and other variables. 3.6-7

BFN-30 3.6.6 Reactivity of Fuel in Storage The NRC amended its regulations in December 1998 to give licensees the option of either meeting the criticality accident requirements of 10 CFR 70.24 paragraphs (a) through (c) in handling and storage areas for Special Nuclear Material (SNM), or electing to comply with certain requirements in a new section, 50.68, in 10 CFR part 50. Browns Ferry has chosen to comply with the new requirements of 10 CFR 50.68(b). To meet these new requirements the quantity of SNM, other than nuclear fuel stored onsite, shall be less than the quantity necessary for a critical mass. The quantity of SNM specified to be enough for a critical mass in Section 1.1 of Regulatory Guide 10.3, Guide for the Preparation of Applications for Special Nuclear Material Licenses of Less than Critical Mass Quantities is 350 gram of U-235, 200 grams of U-233, and 200 grams of Pu-239. The combined total of non-fuel SNM maintained at BFN is far less than this quantity. The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five percent by weight. Requirements for new fuel storage are described in Section 10.2 of the FSAR. Requirements for spent fuel storage are described in FSAR, Section 10.3.5. Requirements for the reactor building ventilation radiation monitors are described in FSAR, Section 7.12.5. The existing radiation monitors on the refuel zone are considered to meet 10 CFR 50.68(b) requirements. The remaining 50.68 (b) requirements for fuel handling outside of approved storage areas are contained in plant procedures. The basic criterion associated with the storage of irradiated (spent) and new fuel is that the effective multiplication factor of fuel stored under normal conditions will be 0.95 for high density racks. For storage of new fuel in the new fuel storage vaults, the effective multiplication factor will be 0.90 for dry conditions and 0.95 for flooded conditions. [Note: Placement of fuel in the new fuel storage vaults is currently prohibited at Browns Ferry. This restriction is administratively controlled by BFN Site Procedures.] The current and legacy fuel products have been assessed and shown to meet the criteria per References 10 and 26. 3.6.7 References

1. (Deleted)

2. (Deleted) 3.6-8

BFN-30

3. (Deleted)

4. (Deleted)

5. (Deleted)

6. (Deleted)

7. (Deleted)

8. R. L. Crowther, Xenon Considerations in Design of Boiling Water Reactors, APED-5640, June 1968.

9. (Deleted)

10. ANP-3160(P) Revision 1, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM' 10XM, December 2015.

11. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.

12. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, Exxon Nuclear Company, March 1983.

13. EMF-2493(P) Revision 0, MICROBURN-B2 Based Impact of Failed/Bypassed LPRMs and TIPs, Extended LPRM Calibration Interval, and Single Loop Operation on Measured Radial Bundle Power Uncertainty, Siemens Power Corporation, December 2000.

14. EMF-CC-074(P)(A) Volume 4 Revision 0, BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.

15. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.

3.6-9

BFN-30

16. ANP-10307PA Revision 0 AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

17. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, April 1986.

18. (Deleted)

19. (Deleted)

20. (Deleted)

21. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.

22. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-Factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.

23. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.

24. (Blank)

25. (Blank)

26. ANP-3910P Revision 3, Browns Ferry Nuclear Plant Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Analysis for ATRIUM 11 Fuel, Framatome Inc., April 2022.

3.6-10

BFN-28 Table 3.6-1 DEFINITION OF FUEL DESIGN LIMITS Maximum Linear Heat Generation Rate (MLHGR) The MLHGR is the maximum linear heat generation rate expressed in kW/ft for the fuel rod with the highest surface heat flux at a given nodal plane in the bundle. The MLHGR operating limit is bundle type dependent. The MLHGR is monitored to assure that all mechanical design requirements will be met. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) The MAPLHGR is the maximum average linear heat generation rate (expressed in kW/ft) in any plane of a fuel bundle allowed by the plant Technical Specifications for that fuel type. This parameter is obtained by averaging the linear heat generation rate over each fuel rod in the plane, and its limiting value is selected such that: (a) The peak clad temperature during the design basis loss-of-coolant accident will not exceed 2200F in the plane of interest, and (b) All fuel design limits specified in Subsection 3.2, Fuel Mechanical Design, will be met if the MLHGR is not monitored for that purpose. Minimum Critical Power Ratio (MCPR) The critical power ratio is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. The critical power is determined at the same mass flux, inlet temperature, and pressure which exists at the specified reactor condition. Thermal margin is stated in terms of the minimum value of the critical power ratio, MCPR, which corresponds to the most limiting fuel assembly in the core. Operating Limit MCPR The MCPR operating limit is the minimum CPR specified in the Core Operating Limits Report for a given bundle type. The minimum CPR is a function of several parameters, the most important of which are bundle power, bundle flow, and empirically derived correlation coefficients (R-factors in GE terminology; feff or K-factor for AREVA). The empirical correlation coefficients are dependent upon the local power distribution and details of the bundle mechanical design including channel bow considerations. The limiting value of CPR is selected for each bundle type such that, during the most limiting event of moderate frequency, the calculated CPR in that bundle is not less than the safety limit CPR. The MCPR safety limit is attained when parameters (e.g., bundle power, flow, pressure, subcooling, etc.) are applied to approved licensing correlations, supporting the Technical Specification value. [*Note: The K-factor is used for AREVA reload analyses and monitoring with the ACE critical power correlation, per References 21 and 22. Fuel designs monitored using the AREVA SPCB correlation, a similar term, Feff, is developed to account for local power distribution and details of the bundle mechanical design, per Reference 23. Channel bow considerations are accounted during MCPR Safety Limit analyses, per (Reference 16)]

BFN-16 Figures 3.6-1 through 3.6-13 (Deleted by Amendment No. 16)

BFN-26 3.7 Thermal and Hydraulic Design 3.7.1 Power Generation Objective The objective of the thermal and hydraulic design of the core is to achieve power operation of the fuel over the life of the core without sustaining fuel damage. 3.7.2 Power Generation Design Basis The thermal hydraulic design of the core shall provide the following characteristics:

a. The ability to achieve rated core power output throughout the design lifetime of the fuel without sustaining fuel damage.

b. The flexibility to adjust core power output over the range of plant load and load maneuvering requirements without sustaining fuel damage.

3.7.3 Safety Design Basis

1. The thermal hydraulic design of the core shall establish limits for use in setting devices of the nuclear safety systems so that no fuel damage occurs as a result of abnormal operational transients (see Chapter 14, Plant Safety Analysis).

2. The thermal hydraulic design of the core shall establish a thermal hydraulic safety limit for use in evaluating the safety margin relating the consequences of fuel barrier failure to public safety.

3.7.4 Thermal and Hydraulic Limits 3.7.4.1 Requirements for Steady-State Conditions For purposes of maintaining adequate fuel performance margin during normal steady-state operation, the Minimum Critical Power Ratio (MCPR) must not be less than the required MCPR operating limit, the Average Planar Linear Heat Generation Rate (APLHGR) must be maintained below the required Maximum APLHGR limit (MAPLHGR) and the Linear Heat Generation Rate (LHGR) must be maintained below the required Maximum LHGR limit (MLHGR). The steady-state MCPR, MAPLHGR, and MLHGR limits are determined by analysis of the most severe moderate frequency Abnormal Operational Transients (AOTs) to accommodate uncertainties and provide reasonable assurance that no fuel damage results during moderate frequency AOTs at any time in life. 3.7-1

BFN-28 3.7.4.2 Requirements for Abnormal Operational Transients (AOTs) The MCPR, MAPLHGR, and MLHGR limits are established such that no safety limit is expected to be exceeded during the most severe moderate frequency AOT event as defined in Chapter 14, Plant Safety Analysis. 3.7.4.3 Summary of Design Bases In summary, the steady-state operating limits have been established to assure that the design bases are satisfied for the most severe moderate frequency AOT. Demonstration that the steady-state MCPR, MAPLHGR, and MLHGR limits are not exceeded is sufficient to conclude that the design bases are satisfied. 3.7.5 Description of Thermal - Hydraulic Design of the Reactor Core 3.7.5.1 Critical Power Ratio A description of the critical power ratio is provided in Subsection 3.7.7.1, Critical Power. Criteria used to calculate the critical power ratio safety limit are given in References 32 and 42 for AREVA reload analyses. 3.7.5.2 Average Planar Linear Heat Generation Rate (APLHGR) Models used to calculate the APLHGR limit are given in Subsection 3.2.5.1, Evaluation Methods, as pertaining to the fuel mechanical design limits, and in Subsection 6.5.2.1, Analysis Model, as pertaining to 10 CFR 50, Appendix K limits. 3.7.5.3 Core Coolant Flow Distribution and Orificing Pattern The flow distribution to the fuel assemblies and bypass flow paths is calculated on the assumption that the pressure drop across all fuel assemblies and bypass flow paths is the same. This assumption has been confirmed by measuring the flow distribution in boiling water reactors (References 2, 3, and 4). The components of bundle pressure drop considered are friction, local, elevation, and acceleration (Subsections 3.7.5.4.1 through 3.7.5.4.4, respectively). Pressure drop measurements made in operating reactors confirm that the total measured core pressure drop and calculated core pressure drop are in good agreement. There is reasonable assurance, therefore, that the calculated flow distribution throughout the core is in close agreement with the actual flow distribution of an operating reactor. An iteration is performed on flow through each flow path (fuel assemblies and bypass flow paths), which equates the total differential pressure (plenum to plenum) across each path and matches the sum of the flows through each path to the total core flow. The total core flow less the control rod cooling flow enters the lower 3.7-2

BFN-26 plenum. A fraction of this passes through various bypass flow paths. The remainder passes through the orifice in the fuel support plate (experiencing a pressure loss) where some of the flow exits through the fit-up between the fuel support and the lower tieplate and through the lower tieplate holes into the bypass flow region. All reload core fuel bundles have lower tieplate holes. The majority of the flow continues through the lower tieplate (experiencing a pressure loss) where some flow exits through the flow path defined by the fuel channel and lower tieplate into the bypass region. This bypass flow is lower for those fuel assemblies with finger springs. The bypass flow paths considered in the analysis and typical values of the fraction of bypass flow through each flow path are given in Reference 5. Within the fuel assembly, heat balances on the active coolant are performed nodally. Fluid properties are expressed as the bundle average at the particular node of interest. In evaluating fluid properties a constant pressure model is used. For core design and monitoring, assembly-specific relative radial and axial power distributions are used with the bundle flow to determine the axial coolant property distribution, which gives sufficient information to calculate the pressure drop components within each fuel assembly type. When the equal pressure drop criterion described above is satisfied, the flow distributions are established. 3.7.5.4 Core Pressure Drop and Hydraulic Loads The components of bundle pressure drop considered are friction, local, elevation, and acceleration pressure drops. Pressure drop measurements made in operating reactors confirm that the total measured core pressure drop and calculated core pressure drop are in good agreement. 3.7.5.4.1 Friction Pressure Drop Friction pressure drop is calculated with a basic model as follows: w2 fL Pf 2 TPF 2 2 gc DH Ach where 3.7-3

BFN-28 P f = friction pressure drop, psi w = mass flow rate g c = conversion factor

     = average nodal liquid density DH = channel hydraulic diameter Ach = channel flow area L = incremental length f = friction factor TPF = two-phase friction multiplier AREVA pressure drop methodology is described in References 33, 34, and 35.

3.7.5.4.2 Local Pressure Drop The local pressure drop is defined as the irreversible pressure loss associated with an area change, such as the orifice, lower tieplate, and spacers of a fuel assembly. The general local pressure drop model is similar to the friction pressure drop and is w2 K 2 PL 2 gc A 2 TPL where PL = local pressure drop, psi K = local pressure drop loss coefficient A = reference area for local loss coefficient TPL = two-phase local multiplier 3.7-4

BFN-28 and w, g, and are defined above. For AREVA analyses the Reference 33, 34, and 35 methodologies are used. For advanced spacer designs a quality modifier has been incorporated in the two-phase multiplier to better fit the data. Empirical constants were added to fit the results to data taken for the specific designs of the BWR fuel assembly. These data were obtained from tests performed in single-phase water to calibrate the orifice, the lower tieplate, and the holes in the lower tieplate, and in both single- and two-phase flow, to derive the best fit design values for spacer and upper tieplate pressure drop. The range of test variables was specified to include the range of interest for boiling water reactors. New test data are obtained whenever there is a significant design change to ensure the most applicable methods are used. 3.7.5.4.3 Elevation Pressure Drop The elevation pressure drop is based on the relation: g PE L  ; gc f ( 1 ) g where PE = elevation pressure drop, psi L = incremental length

     = average mixture density g = acceleration of gravity
    = nodal average void fraction f , g = saturated water and vapor density, respectively AREVA void fraction models are described in References 33, 34, and 35.

3.7-5

BFN-26 3.7.5.4.4 Acceleration Pressure Drop A reversible pressure change occurs when an area change is encountered, and an irreversible loss occurs when the fluid is accelerated through the boiling process. The basic formulation for the reversible pressure change resulting from a flow area change in the case of single-phase flow is given by: w2 PACC (1 A2 ) 2 gc f A22 A2 A A1 where PACC = acceleration pressure drop A2 = final flow area A1 = initial flow area In the case of two-phase flow, the liquid density is replaced by a density ratio so that the reversible pressure change is given by: w 2 H PACC (1 )2 A 2 gcKE2 A22 where 1 x (1 x )

                        , homogeneous density H     g      f 1        x3       (1  x )3 2 2 2                  , kinetic energy density KE 2

g f (1 )2

     = void fraction at A2 x = steam quality at A2 and other terms are as previously defined. The basic formulation for the acceleration pressure change due to density change is:

3.7-6

BFN-28 w2 1 1 PACC 2 gcAch OUT IN where is either the homogeneous density, H , or the momentum density, M 1 x2 (1 x )2 M g f (1 ) and is evaluated at the inlet and outlet of each axial node. Other terms are as previously defined. The total acceleration pressure drop in boiling water reactors is on the order of a few percent of the total pressure drop. 3.7.5.5 Correlation and Physical Data AREVA has obtained substantial amounts of physical data in support of the pressure drop and thermal-hydraulic loads discussed in Subsection 3.7.5.4, Core Pressure Drop and Hydraulic Loads. Correlations have been developed to fit these data to the formulations discussed. 3.7.5.5.1 Pressure Drop Correlations AREVA has taken significant amounts of friction pressure drop data in multi-rod geometries representative of BWR plant fuel bundles and correlated both the friction factor and two-phase multipliers on a best fit basis using the pressure drop formulations reported in Subsections 3.7.5.4.1 and 3.7.5.4.2. Tests are performed in single-phase water to calibrate the orifice and the lower tie-plate, and in both single-phase and two-phase flow to arrive at best fit design values for spacer and upper tie-plate pressure drop. The range of test variables is specified to include the range of interest to boiling water reactors. New data are taken whenever there is a significant design change to ensure the most applicable methods are in use at all times. Applicability of the single-phase and two-phase hydraulic models discussed in Subsections 3.7.5.4.1 and 3.7.5.4.2 have been confirmed by full scale prototype flow testing. 3.7.5.5.2 Void Fraction Correlation The void fraction correlation includes effects of pressure, flow direction, mass velocity, quality, and subcooled boiling. 3.7-7

BFN-28 3.7.5.5.3 Heat Transfer Correlation The fuel heat transfer correlations for AREVA reload analyses are described in Reference 37. 3.7.5.6 Thermal Effects of Abnormal Operational Transients The evaluation of the cores capability to withstand the thermal effects resulting from abnormal operational transients is covered in Chapter 14, Plant Safety Analysis. 3.7.5.7 Uncertainties in Estimates Uncertainties in thermal-hydraulic parameters are considered in the statistical analysis which is performed to establish the fuel cladding integrity safety limit documented in Subsection 3.7.7.1.1, Fuel Cladding Integrity Safety Limit. 3.7.5.8 Flux Tilt Considerations For flux tilt considerations, refer to Subsection 3.6.4.2, Power Distribution. 3.7.6 Description of the Thermal-Hydraulic Design of the Reactor Coolant System 3.7.6.1 Power/Flow Operating Map 3.7.6.1.1 Performance Range for Normal Operations A boiling water reactor must operate within certain restrictions due to pump net positive suction head (NPSH) requirements, overall plant control characteristics, core thermal power limits, etc. A typical operating power-flow map for BFN is shown in Figure 3.7-1. The boundaries on the maps are as follows: Natural Circulation Line (Line A in Figure 3.7-1) The operating state of the reactor moves along this line for the normal control rod withdrawal sequence in the absence of recirculation pump operation. 3.7-8

BFN-29 20 Percent Pump Speed Line (Line B in Figure 3.7-1) The operating state for the reactor follows this line for the normal control rod withdrawal sequence with recirculation pumps operating at approximately 20 percent speed. 100 Percent Rod Line (Line which runs through point E in Figure 3.7-1) The 100% rod line passes through 100 percent power at 100 percent flow. The operating state for the reactor follows this line (or one roughly parallel to it) for recirculation flow changes with a fixed control rod pattern. The line is based on constant xenon concentration. APRM Rod Block Line (Shown in Figure 3.7-1) The line shown on the graph limits control rod withdrawal to within the constraint of the control rod block line. Pump Constant Speed Line The lines passing through Points H and I and Points G and J show the change in flow associated with power reduction by control rod insertion from 3458 MWt, 105% core flow and 3458 MWt, 100% core flow, respectively, while maintaining constant recirculation pump speed. Minimum Expected Flow Control Line (Shown in Figure 3.7-1) This line approximates the expected flow control line that the plant would follow if core flow were rapidly reduced from 107.7% of rated core flow at 35.4% power by reducing recirculation pump speed to 20% with a constant control rod pattern. Minimum Power Line (Shown in Figure 3.7-1) Operation above the line passing through Points L and K prevents cavitation of the jet pumps and reactor recirculation pumps by ensuring adequate subcooling and therefore net positive suction head for all normal modes of jet pump and reactor recirculation pump operation. Increased Core Flow (ICF) Region (The region bounded by Points E, G, J, I, H, and F in Figure 3.7-1) The plant is licensed for Increased Core Flow (ICF) operation up to a maximum of 105% of rated core flow at 100% rated power. At core thermal powers less than rated but greater than or equal to 3458 MWt, the maximum allowable core flow is 3.7-9

BFN-29 limited to 105% of rated core flow. At thermal powers less than 3458 MWt, the maximum core flow is set by the constant recirculation pump speed line that passes through Point H, up to a maximum core flow of 112.6% of rated flow at 35.4% rated power on the Power/Flow operating map. ICF can be used to extend full power operation beyond the point where all rods are out at rated power and flow conditions (End of Full Power Life - EOFPL). ICF may be used prior to reaching EOFPL for operating flexibility. Maximum Extended Load Line Limit Analysis (MELLLA) Region (shown in Figure 3.7-1) The plant is licensed for Maximum Extended Load Line Limit Analysis (MELLLA) which allows operation at 3458 MWt down to 81% rated flow conditions. As shown in Figure 3.7-1, with power uprate to 3952 MWt, MELLLA allows operation at full power down to 99% rated flow conditions. The plant is also licensed for operation in the MELLLA+ region, which restores much of the operating flexibility at rated power that previously existed in the MELLLA domain prior to implementation of power uprate to the licensed core thermal power level of 3952 MWt. This is accomplished by expanding the licensed operating domain to allow operation above the MELLLA upper boundary at core flows greater than or equal to 55% of the rated core flow with reactor power up to 100% of rated thermal power. The MELLLA+ region upper boundary passes through 100% power at 85% core flow (point N). The operating state of the reactor follows this line (or one similar) for recirculation flow changes with a fixed control rod pattern at a constant xenon condition. The MELLLA+ upper boundary line terminates at 55% core flow (point O). 3.7.6.1.2 Flow Control The following simplified description of BWR operation summarizes the principle modes of normal power range operation. Prior to unit startup the recirculation pumps are started one at a time and typically held at a pump speed of 28 percent or less of rated speed. The first part of the startup sequence is achieved by withdrawing control rods with the recirculation pumps at a pump speed of 28 percent or less of rated speed. Core power, steam flow, and feedwater flow increase as control rods are withdrawn by the operator, until feedwater flow increases to a point above the feedwater flow interlock. The low feedwater flow interlock (approximately 19% feedwater flow) prevents low power-high recirculation flow combinations which may create recirculation system NPSH problems. The natural circulation characteristics of the BWR are still very influential in this part of the power flow region. 3.7-10

BFN-29 Once the feedwater interlock has been cleared the recirculation flow in each loop can be increased to increase power. The operator then can achieve full power by a combination of control rod withdrawals and pump speed increases, depending on operating and core management strategies. A typical strategy for plant startup is to increase core flow to a mid-range value (typically greater than 55% core flow). Then control rods are withdrawn to a point just below the upper boundary of the licensed operating domain. Core flow can then be increased until the desired high power condition is reached. The normal power range operation is bounded by the licensed operating domain maximum rod line and 100% power. The large negative operating coefficients, which are inherent in the BWR, provide the following important advantages:

1. Stable load change response following with well damped behavior and little undershoot or overshoot in the heat transfer response.

2. Load changes with recirculation flow control.

3. Strong damping of spatial power disturbances.

To increase reactor power, it is necessary only to increase the recirculation flow rate which reduces core average void content, causing an increase in core reactivity. As the reactor power increases, more steam is formed and the reactor stabilizes at a new power level with the transient excess reactivity balanced by the new void formation. No control rods are moved to accomplish this power level change. Conversely, when a power reduction is required, it is necessary only to reduce the recirculation flow rate. When this is done, more voids are formed in the moderator, and the reactor power output automatically decreases to a new power level commensurate with the new recirculation flow rate. No control rods are moved to accomplish the power reduction. Varying the power level by varying the recirculation flow rate (flow control) is more advantageous than using control rod positioning. Flow variations perturb the reactor uniformly in the horizontal planes and thus, allow operation with flatter power distribution and reduced transient allowances. As the flow is varied, the power and void distributions remain approximately constant at the steady-state end points for a wide range of flow variations. These constant distributions provide the important advantage that the operator can adjust the power distribution at a reduced power and flow by movement of control rods and then bring the reactor to rated conditions by increasing flow, with the assurance that the power distribution will remain approximately constant. Subsection 7.9, "Recirculation Flow Control System," describes the means by which recirculation flow is varied. 3.7-11

BFN-29 3.7.6.2 Thermal-Hydraulic Stability Performance The AREVA analytical methodology for demonstrating stability compliance for AREVA fuel designs is described in Subsection 3.6.4.6, Stability. To provide additional assurance that regional instabilities will not occur, Browns Ferry has implemented the long-term stability solution designated as Option III in NEDO-31960, Supplement 1, BWR Owners Group Long-Term Stability Solution Licensing Methodology, for Units 2 and 3 and DSS-CD in NEDC-33075P-A, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density, for Unit 1. As part of the implementation of the MELLLA+ operating domain expansion, the Option III stability solution is replaced as the licensing basis stability solution by the Detect and Suppress Solution - Confirmation Density (DSS-CD) long term stability solution (NEDC-33075P-A, Revision 8, GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density). The existing Option III algorithms are retained, with generic setpoints, to provide defense-in-depth protection for unanticipated stability events. For DSS-CD long-term stability solution, the Oscillation Power Range Monitor (OPRM) Upscale Trip function of the Power Range Neutron Monitoring (PRNM) system is enabled. [Note: See Section 7.5.7.3.5 for a detailed description of the OPRM system.] The OPRM Upscale Trip function provides protection from exceeding the fuel MCPR Safety Limit in the event of thermal-hydraulic power oscillations. The OPRM receives input signals from the Local Power Range Monitors (LPRMs) within the reactor core. An Upscale Trip is issued if oscillatory changes in the neutron flux are detected. The OPRM Upscale Trip function is required to be operable when the plant is in a region of power-flow operation where actual thermal-hydraulic oscillations might occur (Tech Spec enabled region - greater than 23% rated thermal power and less than 75% recirculation drive flow. The DSS-CD setpoints that provide protection of the Safety Limit MCPR during instability events are established on a cycle independent basis and are confirmed to be applicable for each reload cycle via an NRC approved process using established applicability checklists. If necessary, cycle specific setpoint adjustments can be implemented to provide adequate Safety Limit MCPR protection. For Unit 1, a cycle specific DSS-CD stability analysis is performed for each reload cycle to conform the plant-applicability checklist in NEDC-33075P-A remains satisfied and define the cycle specific BSP regions in the COLR. The confirmation of the checklist ensures that the DSS-CD setpoints are applicable for that cycle and provide protection of the fuel limits. If the OPRM trip function should become inoperable, alternate methods of stability protection are implemented in accordance with the Technical Specifications. 3.7-12

BFN-28 3.7.7 Evaluation The thermal-hydraulic design of the reactor core and reactor coolant system is based upon an objective of no fuel damage during normal operation or during abnormal operational transients. This design objective is demonstrated by analysis as described in the following sections. 3.7.7.1 Critical Power The objective for normal operation and AOTs is to maintain nucleate boiling and thus avoid a transition to film boiling. Operating limits are specified to maintain adequate margin to the onset of the boiling transition. The figure of merit utilized for plant operation is the critical power ratio (CPR). This is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. The critical power is determined at the same mass flux, inlet temperature, and pressure which exists at the specified reactor condition. Thermal margin is stated in terms of the minimum value of the critical power ratio (MCPR), which corresponds to the most limiting fuel assembly in the core. To ensure that adequate margin is maintained, a design requirement based on a statistical analysis was selected as follows: Moderate frequency AOTs caused by a single operator error or equipment malfunction shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition (Reference 32). Both the transient (safety) and normal operating thermal limits in terms of MCPR are derived from this basis. 3.7.7.1.1 Fuel Cladding Integrity Safety Limit The generation of the Minimum Critical Power Ratio (MCPR) limit requires a statistical analysis of each reload core near the limiting MCPR condition. The MCPR Fuel Cladding Integrity Safety Limit applies not only for core wide AOTs, but is also applied to the localized rod withdrawal error AOT. The cycle-specific Safety Limit MCPR is derived based on methodology documented in References 32 and 42 for AREVA reload analyses. For AREVA reload analyses, the Reference 43, 44, and 46-47 approved critical power correlations are used as appropriate to specific fuel types. The resulting safety limit MCPR for each cycle is given in the AREVA Reload Analysis Report for each BFN unit (included in Appendix N of the FSAR). 3.7-13

BFN-28 Statistical Model The statistical analysis utilizes a model of the BWR core which simulates the process computer function. This code produces a critical power ratio (CPR) map of the core based on inputs of power distribution, flow, and heat balance information. Details of the procedure are documented in References 32 and 42 for AREVA reload analyses. Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical representation of the core and the resulting bundle critical power ratios are calculated. Uncertainties typical of AREVA cycle-specific analyses are provided in Reference 39. The minimum allowable critical power ratio is set to correspond to the criterion that 99.9% of the rods are expected to avoid boiling transition by interpolation among the means of the distributions formed by all the trials. BWR Statistical Analysis Statistical analyses are performed for each operating cycle that provide the fuel cladding integrity safety limit MCPR. This safety limit MCPR is derived based on methodology documented in Reference 42. 3.7.7.1.2 MCPR Operating Limit Calculational Procedure A plant-unique MCPR operating limit is established to provide adequate assurance that the cycle-specific fuel cladding integrity safety limit for the plant is not exceeded for any moderate frequency AOT. This operating requirement is obtained by addition of the maximum CPR value for the most limiting AOT (including any imposed adjustment factors) from conditions postulated to occur at the plant to the cycle-specific fuel cladding integrity safety limit. Calculational Procedure for AOT Pressurization Events Core-wide rapid pressurization events (turbine trip w/o bypass, load rejection w/o bypass, feedwater controller failure) are analyzed using the system model COTRANSA2 (Reference 40) for AREVA reload analyses. The Time Varying Axial Power Shape (TVAPS) calculation is performed by COTRANSA2 (Reference 40) for AREVA reload analyses. The TVAPS is a short time period phenomena that occurs during the control rod scram that terminates an AOT. The analytical procedures used to evaluate the AOTs account for TVAPS either in a bounding manner or explicitly, depending on the AOT and the fuel design. 3.7-14

BFN-28 Calculational Procedure for AOT Slow Events For AREVA reload analyses, the MICROBURN-B2 (Reference 35) 3-D simulator code is used for quasi-steady-state loss of feedwater heating (LFWH) transients; for transient events that cannot be handled in a quasi-steady-state manner, COTRANSA2 (Reference 40) is used. Inadvertent HPCI startup is not analyzed due to the fact that the core enthalpy change for the event is similar to the loss of feedwater heating event and the severity of the inadvertent HPCI startup is generally bounded by the load reject no bypass and feedwater controller failure events. The loss of feedwater heating event is analyzed each cycle to demonstrate that the loss of feedwater heating event and inadvertent HPCI startup events remain non-limiting. When necessary, it is analyzed using COTRANSA2 for AREVA analyses. Rod Withdrawal Error Calculational Procedure The reactor core behavior during the rod withdrawal error transient is calculated by doing a series of steady-state three-dimensional coupled nuclear-thermal-hydraulic calculations using the 3-D BWR Simulator (Reference 35). Event Descriptions For AREVA analyses, the AOT descriptions in Chapter 14 of this UFSAR are utilized with appropriate cycle-specific input parameter updates. MCPR Operating Limit Calculation For AREVA reload analyses, the CPR for rapid AOTs is calculated using XCOBRA (Reference 41) for the initial steady-state analysis and XCOBRA-T (Reference 37) for the transient thermal margin analysis of the limiting fuel assembly. MCPR Uncertainty Considerations for AREVA Reload Analyses For fast transient events, the one-dimensional kinetic thermal-hydraulic COTRANSA2 code is used for the reactor system analysis, with the XCOBRA/XCOBRA-T codes evaluating the initial and transient hot channel hydraulics and CPR. The NRC approved application methodology of References 40, 41, and 42 provides adequate conservatism by accounting for uncertainties in the computed CPR results. Therefore, for a given transient event, the required operating limit MCPR (OLMCPR) is simply the XCOBRA-T calculated CPR added to the safety limit MCPR (SLMCPR). The results of the system pressurization transients are sensitive to the control rod scram speed used in the calculations. To take advantage of average scram speeds 3.7-15

BFN-28 faster than those associated with the technical specifications surveillance times, scram speed-dependent MCPR limits are provided. If the control rod scram time performance is equal or better than the nominal scram speed (NSS) insertion times specified in the core operating limits report (COLR), the NSS-based MCPR operating limits apply. If the control rod scram time performance is equal or better than the optimal scram speed (OSS) insertion times specified in the COLR, the OSS-based MCPR operating limits apply. Otherwise, MCPR operating limits are applied that are based on the technical specification scram speed (TSSS) control rod insertion times. The plant technical specifications allow for operation with a certain number and arrangement of slow control rods as well as one stuck control rod. Conservative adjustments to the OSS, NSS, and TSSS scram speeds are input to the reload transient analyses to account for these slow and stuck rod effects on scram reactivity. TSSS, NSS, and OSS-based power-dependent MCPR operating limits are reported in the reload licensing analysis report. Low Flow and Low Power Effects on MCPR The operating limit MCPR must be increased for low flow and, for plants with ARTS, low power conditions. This is because, in the BWR, power increases as core flow increases which results in a corresponding lower MCPR. If the MCPR at a reduced flow condition were at the 100% power and flow MCPR operating limit, a sufficiently large inadvertent flow increase could cause the MCPR to decrease below the Fuel Cladding Integrity Safety Limit MCPR. The Average Power Range Monitor, Rod Block Monitor, and Technical Specification (ARTS) Improvement Program imposes both power- and flow-dependent limits imposed on the operating limit MCPR (OLMCPR). The flow-dependent required OLMCPR, MCPRf, is defined as a function of the core flow rate and maximum rated power core flow capability. For powers between 100% of rated and the bypass point for the turbine stop valve/turbine control valve fast closure scram signal (about 26% of rated), the power-dependent OLMCPR, MCPRp, is directly supplied. For powers between 23% rated and the bypass point, the MCPRp limits are absolute values and are defined separately for high core flows (>50% of rated flow) and for low core flows (50% of rated flow) conditions. There is no thermal limits monitoring required below 23% of rated power. The OLMCPR to be used at powers less than 100% becomes the most limiting value of either MCPRf or MCPRp. End-of-Cycle Coastdown Considerations AOT analyses are performed at the full power, EOC, all-rods-out condition. Once an individual plant reaches this condition, it may shutdown for refueling or it may be placed in a coastdown mode of operation. In the end-of-cycle coastdown type of operation, if coastdown is initiated from a reactor power level 3.7-16

BFN-28 greater than 3458 MWt, recirculation pump speed must be decreased as reactor power coasts down in order to maintain core flow less than or equal to 105% of rated. Once reactor power is less than or equal to 3458 MWt, the plant is allowed to coastdown to a lower percent of rated power while maintaining recirculation pump speed less than or equal to the constant pump speed line corresponding to 105% of rated core flow at 3458 MWt. For GE methods, in Reference 29, evaluations were made at 90%, 80%, and 70% power level points on the linear curve. The results show that the pressure and MCPR from the limiting pressurization AOT exhibit a larger margin for each of these points than the EOC full power, full flow case. MLHGR limits for the full power, rated flow case are conservative for the coastdown period, since the power will be decreasing and rated core flow will be maintained. Therefore, it can be concluded that the coastdown operation beyond full power operation is conservatively bounded by the analysis at the EOC conditions. In Reference 30, this conclusion is confirmedfor coastdown operation down to 40% power and is shown to hold for analyses performed with ODYN. For AREVA methods, the nominal start of coastdown cycle exposure is conservatively extended. Coastdown limits are then determined at the final cycle exposure bounding of anticipated operation, forming the licensing basis maximum core average exposure (CAVEX). 3.7.7.2 Analysis Options 3.7.7.2.1 MCPR Margin Improvement Options Several MCPR margin improvement options have been developed for operating BWRs. The following options are utilized at Browns Ferry: (1) Recirculation Pump Trip (2) Thermal Power Monitor (3) Exposure-Dependent Limits (4) Improved Scram Times The exposure-dependent limits option is used on an as-needed basis. AREVA Reload Analysis Report, for each unit indicates which options are currently analyzed. 3.7-17

BFN-28 Recirculation Pump Trip Due to operator concerns associated with the difficulty of restarting the recirculation pumps following a reactor scram, the recirculation pump trip (RPT) feature is maintained in a bypass condition. For many of the plant operating cycles, the limiting AOTs are the turbine trip, generator load rejection, or other AOTs which result in a turbine trip. A significant improvement in thermal margin can be realized if the severity of these transients is reduced. The RPT feature accomplishes this by cutting off power to the recirculation pump motors anytime that the turbine control valve or turbine stop valve fast closure occurs. This rapid reduction in recirculation flow increases the core void content during the AOT, thereby reducing the peak AOT power and heat flux. Basically, the RPT consists of switches installed in both the turbine control valves and the turbine stop valves. When these valves close, breakers are tripped which releases the recirculation pumps to coast down under their own inertia. Thermal Power Monitor The APRM simulated thermal power trip (APRM thermal power monitor) is a minor modification to the APRM system. The modified APRM system generates two upscale trips. On one, the APRM signal (which is proportional to the thermal neutron flux) is compared with a reference which is not dependent on flow rate. During normal reactor operations, neutron flux spikes may occur due to conditions such as transients in the recirculation system, transients during large flow control load maneuvers, or transients during turbine stop valve tests. The neutron flux leads the heat flux during transients because of the fuel time constant. And the neutron flux for these transients trips upscale before the heat flux increases significantly. (High heat flux is the precursor of fuel damage.) Thus, increased availability can be achieved without fuel jeopardy by adding a trip dependent on heat flux (thermal power). For this trip, the APRM signal is passed through a low pass RC filter. It is compared with a recirculation flow rate dependent reference which decreases approximately parallel to the flow control lines. In addition to increased availability, another benefit is that with the minor operational spikes filtered out, the heat flux trip setpoint is lower than the neutron flux trip setpoint. For long, slow AOTs such as the loss-of-feedwater heater, the heat flux and neutron flux are almost in equilibrium. For these AOTs, the lower trip setpoint 3.7-18

BFN-28 results in an earlier scram with a smaller increase in heat flux and a corresponding reduction in the consequences. The APRM Simulated Thermal Power Trip at Browns Ferry is non-safety grade and is not taken credit for in any of the licensing transient analyses. Exposure-Dependent Limits The severity of any plant AOT pressurization event is worst at the end of the cycle primarily because the EOC all-rods-out scram curve gives the worst possible scram response. It follows that some limits relief may be obtained by analyzing the AOTs at other interim points in the cycle and administering the resulting limits on an exposure dependent basis. This technique is straightforward and consists merely of repeating certain elements of the AOT analyses for selected midcycle exposures. Because the scram reactivity function monotonically deteriorates with exposure (after the reactivity peak), the limit determined for an exposure Ei is administered for all exposures in the interval Ei-1 E Ei where Ei-1 is the next lower exposure point for which a limit was determined. This results in a table of MCPR limits to be applied through different exposure intervals of the cycle. Improved Scram Times As described in Section 3.7.7.1.2, subsection titled MCPR Uncertainty Considerations for AREVA Reload Analyses, for AREVA reload analyses power-dependent MCPR limits are provided for OSS, NSS, and TSSS bases. OSS and NSS bases limits may be used depending on scram speed measurements. Otherwise, the TSSS MCPR limits are applied. The TSSS and NSS-based power-dependent MCPR operating limits are reported in the reload licensing analysis report. 3.7.7.2.2 Operating Flexibility Options A number of operating flexibility options have been developed for BWRs. The following options are utilized at Browns Ferry: (1) Maximum Extended Load Line Limit (2) Increased Core Flow (3) Feedwater Temperature Reduction (4) Turbine Bypass Out of Service (5) ARTS Program (6) Recirculation Coolant Pump Out of Service (Single-Loop Operation) (7) End-of-Cycle Recirculation Pump Trip Out of Service 3.7-19

BFN-28 (8) Power Load Unbalance Out of Service (9) Maximum Extended Load Line Limit Plus The AREVA Reload Licensing Report for each unit indicates which options are currently analyzed (included in Appendix N of the FSAR). Maximum Extended Load Line Limit The Maximum Extended Load Line Limit Analysis (MELLLA) expands the operating domain to allow operation at 3458 MWt down to 81% rated flow conditions and at 3952 MWt down to 99% rated flow conditions. Addition of the MELLLA region provides improved power ascension capability to full power and additional flow range at rated power. Evaluations performed for MELLLA conditions include normal and AOTs, LOCA analysis, containment responses, and stability. The reload fuel dependent results of these analyses are re-evaluated each cycle. Increased Core Flow Operation Analyses are performed in order to justify operation at core flow rates in excess of the 100% rated flow condition. The limiting AOTs that are analyzed at rated flow as part of a standard supplemental reload licensing report are reanalyzed for increased core flow operation. In addition, the loss-of-coolant accident (LOCA), fuel loading error, rod drop accident, and rod withdrawal error are also re-evaluated for increased flow operation. The effects of the increased pressure differences on the reactor internal components, fuel channels, and fuel bundles as a result of the increased flow are analyzed in order to ensure that the design limits will not be exceeded. The thermal-hydraulic stability is re-evaluated for increased core flow operation, and the effects of flow-induced vibration are also evaluated to assure that the vibration criteria will not be exceeded. Feedwater Temperature Reduction Analyses are performed in order to justify operation at a reduced feedwater temperature at rated thermal power within the MELLLA domain. Usually, the analyses are performed for end-of-cycle operation with the last-stage feedwater heaters valved out of service. However, throughout cycle operation, an additional feedwater temperature reduction can be justified by analyses at the appropriate operating conditions. The limiting AOTs are reanalyzed for operation at a reduced feedwater temperature. In addition, the loss-of-coolant accident (LOCA), fuel loading error, rod drop 3.7-20

BFN-28 accident, and rod withdrawal error are also re-evaluated for operation at a reduced feedwater temperature. Turbine Bypass Out of Service Operation of the turbine bypass system is assumed in the analysis of the Feedwater Controller Failure (FWCF)-maximum demand event. If this event is limiting or near limiting, the operating limit MCPR basis may be invalid if the bypass system cannot be demonstrated as fully functional. Reload specific evaluations may incorporate a FWCF without credit for bypass operation calculation as a provision when temporary factors render the system unavailable. Additionally, for extended operation with degraded bypass system operation, evaluations in support of this condition are augmented with the appropriate limiting events, such as the FWCF, for the applicable cycle. ARTS Program The ARTS program is a comprehensive project involving the Average Power Range Monitor (APRM), the Rod Block Monitor (RBM), and Technical Specification improvements. Implementing the ARTS program provides for the following improvements which enhance the flexibility of the BWR during power level monitoring. (1) The Average Power Range Monitor (APRM) trip setdown requirement is replaced by power-dependent and flow-dependent MCPR operating limits to reduce the need for manual setpoint adjustments. In addition, another set of power- and flow-dependent limits (LHGR and/or MAPLHGR) are also specified for more rigorous fuel thermal protection during postulated transients at off-rated conditions. These power- and flow-dependent limits are verified for plant-specific application during the initial ARTS licensing implementation. For AREVA reload analyses, the power-and flow-dependent limits are reviewed and updated as needed each cycle. (2) The RBM system is modified from flow-biased to power-dependent trips to allow the use of a new generic non-limiting analysis for the Rod Withdrawal Error (RWE) and to improve response predictability to reduce the frequency of nonessential alarms. For AREVA reload analyses, the RWE analysis with ARTS is re-evaluated each cycle. The resulting improvements in the flexibility of the BWR provided by ARTS are designed to significantly minimize the time to achieve full power from startup conditions. 3.7-21

BFN-29 Recirculation Coolant Pump Out of Service The plant is licensed to allow extended Single-Loop Operation (SLO) in the MELLLA domain. The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability/outage planning standpoint, in the event maintenance of a recirculation pump or other components renders one loop inoperative. SLO analyses evaluate the plant for continuous operation at a maximum expected power output. To justify SLO, safety analyses have to be reviewed for one-pump operation. The MCPR fuel cladding integrity safety limit, AOT analyses, operating limit MCPR, and non-LOCA accidents are evaluated. Increased uncertainties in the total core flow and traversing incore probe (TIP) readings result in a small increase in the fuel cladding integrity safety limit MCPR. SLO can also result in changes to plant response during a LOCA. These changes are accommodated by the application of reduction factors to the two-loop operation LHGR and/or MAPLHGR limits, if required. Reduction factors are evaluated on a plant and fuel type dependent basis. In each subsequent reload, reduction factors are checked for validity and, if new fuel types are added, new reduction factors may be needed in order to maintain the validity of the SLO analysis. End-of-Cycle Recirculation Pump Trip Out of Service The EOC-RPT-OOS contingency mode of operation eliminates the automatic recirculation pump trip signal when turbine trip or load rejection occurs. As such, the core flow decreases at a slower rate following the recirculation pump trip due to the anticipated transient without scram (ATWS) high pressure recirculation system trip, thus, increasing the severity of the transient responses. Power Load Unbalance Out of Service The load rejection event scenario depends on whether the initial power level is sufficient for the PLU feature to operate. The PLU causes fast closure of the TCV. If the PLU does not operate as the result of a load rejection, the TCV closes at the maximum demand rate for speed control (servo mode). For the PLUOOS analysis, the PLU is assumed inoperable at any power level and does not initiate fast TCV closure. Maximum Extended Load Line Limit Plus (MELLLA+) MELLLA+ expands the licensed operating domain to allow operation above the MELLLA upper boundary at core flows greater than or equal to 55% of rated core flow with reactor power up to 100% of rated thermal power. The MELLLA+ region upper boundary passes through 100% power (3952 MWt) at 85% core flow. The 3.7-22

BFN-29 MELLLA+ upper boundary line terminates at 55% core flow. Final Feedwater Temperature Reduction (greater than 10F from design feedwater temperature) and Single Loop Operation are not allowed in the MELLLA+ domain. Addition of the MELLLA+ region provides improved power ascension capability to full power and additional flow range at rated power. Evaluations performed for MELLLA+ conditions include normal and AOTs, LOCA analysis, containment responses, Anticipate Transient without SCRAM (ATWS) analysis, reactor and reactor internals structural analyses, fluence analyses, and stability. The fuel dependent results of these analyses are re-evaluated each cycle. 3.7.7.3 Core Hydraulics Core hydraulics models and correlations are discussed in Subsection 3.7.5, Description of Thermal-Hydraulic Design of the Reactor Core. 3.7.7.4 Influence of Power Distributions The influence of power distributions on the thermal-hydraulic design is discussed in Reference 41. 3.7.7.5 Core Thermal Response The thermal response of the core for accidents and expected AOT conditions is given in Chapter 14, Plant Safety Analysis. 3.7.7.6 Analytical Methods The analytical methods, thermodynamic data, and hydrodynamic data used in determining the thermal and hydraulic characteristics of the core are documented in Subsection 3.7.7.1.2, MCPR Operating Limit Calculational Procedure. 3.7.8 References

1. (Deleted)

2. Core Flow Distribution in a Modern Boiling Water Reactor as Measured in Monticello, NEDO-10299A, October 1976.

3. H. T. Kim and H. S. Smith, Core Flow Distribution in a General Electric Boiling Water Reactor as Measured in Quad Cities Unit 1, NEDO-10722A, August 1976.

3.7-23

BFN-28

4. Brunswick Steam Electric Plant Unit 1 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations, NEDO-21215, March 1976.

5. Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration, NEDE-21156 (Proprietary), February 1976.

6. (Deleted)

7. (Deleted)

8. (Deleted)

9. (Deleted)

10. (Deleted)

11. (Deleted)

12. (Deleted)

13. (Deleted)

14. (Deleted)

15. (Deleted) 16.(Deleted) 17.(Deleted) 18.(Deleted) 3.7-24

BFN-28 19.(Deleted) 20.(Deleted) 21.(Deleted) 22 (Deleted) 23.(Deleted) 24.(Deleted) 25.(Deleted) 26.(Deleted) 27.(Deleted) 28.(Deleted)

29. Letter from R. A. Bolger (Commonwealth Edison Co.) to B. C. Rusche (USNRC),

QC-2 Proposed Amendment to Facility License No. DPR-30, Docket No. 50-265.

30. Letter from R. E. Engel (GE) to T. A. Ippolito (NRC), End of Cycle Coastdown Analyzed With ODYN/TASC, September 1, 1981.

31. (Deleted)

32. ANF Critical Power Methodology for Boiling Water Reactors, ANF-524(P)(A)

Revision 2 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, November 1990.

33. Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies, XN-NF-79-59(P)(A), Exxon Nuclear Company, November 1983.

34. Siemens Power Corporation Methodology for Boiling Water Reactors:

Evaluation and Validation of CASMO-4/MICROBURN-B2, EMF-2158(P)(A) Revision 0, Siemens Power Corporation, October 1999.

35. MICROBURN-B2: Theory Manual, FS1-0009248 Revision 1, AREVA NP Inc.,

January 2013. 3.7-25

BFN-28

36. (Deleted)

37. XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, Exxon Nuclear Company, February 1987.

38. (Deleted)

39. Letter, T. A. Galioto (FANP) to J. F. Lemons (TVA), Browns Ferry Unit 3 Cycle 12 MCPR Safety Limit Analysis Revision 1 Core Design, TAG:04:018, January 30, 2004

40. COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.

41. Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Company, January 1987.

42. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.

43. SPCB Critical Power Correlation, EMF-2209(P)(A) Revision 3, AREVA NP, September 2009.

44. Application of Siemens Power Corporations Critical Power Correlations to Co-Resident Fuel, EMF-2245(P)(A) Revision 0, Siemens Power Corporation, August 2000.

45. (Deleted)

46. ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.

47. ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.

3.7-26

BFN-29 Figure 3.7-1 OPERATING MAP BFN Core Flow (1\lllb/hr) 0 10 20 30 40 50 60 70 80 90 100 11 0 120 120 100~ Ra~ed The m.al P we r 39 52 . <t 100-% Cor e F_ w 102 .5 Ml b / h r 4500 110 A: Na~ural Circulacion B: 20% Dual P\mp S~ ed C: 5 4 . 3%. Power / 37 .3 % Cl ow N DE F 100 ~0% Power / 99 . 0% , l ow 4000 100 E: 100 . 0% Power / 100 . 0% 2' l ow 39521\IWt LLA+ U

             ?:   l 0 . % Power /1 05 . 0% fl ow G:    81 . 5% Power / 1 00 . 0% Flow                                                    Boun 90   H:    87 . 5% Powe r / 10 5. 0% 2' l ow                                                                                          H I:    35. 4% Powe r / 12 . 61; ?l ow                                                                                                3458 i\IWt       3500 J:    35 . 4% Powe r / 10 7 . 7% 2'low K:        . 3% Power /1 00 . 0% fl ow                                                                                                                     ,.-._

~ 0 80 L: 1 . 3% Power / 37 . 7% fl ow .... '--' N: 100 . 0% Powe r / 85 . 0% l l ow 3000 ~ I. 0: 7 . 6% Power / 55. 0% Flow

             ?: 6 .4% Power / 55. 0% ~l ow                                                                                                                             ~
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                                                                                              ~ u m Power Line I       K                         500 0

0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow(%)

BFN-28 Figure 3.7-2 OPERATING MAP BFN UNIT 2 DELETED BY AMENDMENT 28

            

BFN-28 Figure 3.7-3 OPERATING MAP BFN UNIT 3 DELETED BY AMENDMENT 28 

              

BFN-27 3.8 STANDBY LIQUID CONTROL SYSTEM 3.8.1 Safety Objective The safety objective of the Standby Liquid Control System is to provide a backup method, which is independent of the control rods, to make the reactor subcritical over its full range of operating conditions and provide sufficient buffering agent to maintain the suppression pool pH at or above 7.0 following a DBA LOCA involving fuel damage (see Section 14.6.3.5). Making the reactor subcritical is essential to permit the nuclear system to cool to the point where corrective actions can be carried out. Maintaining the suppression pool pH at or above 7.0 following a LOCA involving fuel damage supports the LOCA radiological dose analyses that do not consider the re-evolution of iodine to the containment atmosphere. The Standby Liquid Control System is classified as a special safety system. 3.8.2 Safety Design Basis

1. Backup capability for reactivity control shall be provided, independent of normal reactivity control provisions in the nuclear reactor, to shut down the reactor if the normal control is impaired so that cold shutdown (MODE 4) cannot be obtained with control rods alone.

2. The backup system shall have the capacity for controlling the reactivity difference between the steady-state rated operating condition of the reactor and the cold shutdown condition (MODE 4), including shutdown margin, to assure complete shutdown from the most reactive condition at any time in the core life.

3. The time required for actuation and effectiveness of the backup reactivity control shall be consistent with the nuclear reactivity rate of change predicted between rated operating and cold shutdown conditions (MODE 4). A scram of the reactor or operational control of fast reactivity transients is not specified to be accomplished by this system.

4. Means shall be provided by which the functional performance capability of the system components can be verified periodically under conditions approaching actual use requirements. Demineralized water, rather than the actual neutron absorber solution, is injected into the reactor to test the operation of all components of the redundant control system.

5. The neutron absorber shall be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage, dilution, or imperfect mixing.

3.8-1

BFN-27

6. The system shall be reliable to a degree consistent with its role as a special safety system.

7. The possibility of unintentional or accidental shutdown of the reactor by this system shall be minimized.

8. The system shall be capable of supplying buffering agent to the suppression pool in the event of a large recirculation break. Sufficient buffering agent shall be provided to ensure that the pH of the suppression pool for DBA post-LOCA events involving fuel damage remains at or above 7.0 for 30 days.

3.8.3 Description (Figures 3.8-1, 3.8-2, 3.8-3, 3.8-5, and 3.8-6) The Standby Liquid Control System is manually initiated from the Main Control Room to pump a boron neutron absorber solution into the reactor if:

1. The operator determines the reactor cannot be shut down or kept shut down with the control rods; or

2. Fuel damage occurs post-LOCA.

The Standby Liquid Control System is required to shut down the reactor at a steady rate within the capacity of the shutdown cooling systems and to keep the reactor from going critical again as it cools. The Standby Liquid Control System is needed in the improbable event that not enough control rods can be inserted in the reactor core to accomplish subcriticality in the normal manner. The Standby Liquid Control System is also required to supply sodium pentaborate solution for post-LOCA events that involve fuel damage to maintain the suppression pool pH at or above 7.0. The radiological dose analyses for the DBA LOCA assumes concentrations of iodine species consistent with a suppression pool pH at or above 7.0 (i.e., re-evolution of iodine to the containment atmosphere is not considered). The sodium pentaborate solution is credited as a buffering agent to offset the post-LOCA production of acids (e.g., radiolysis products). The system consists of a boron solution tank, a test water tank, two positive-displacement pumps, two explosive-actuated valves, and associated local valves and controls. They are mounted in the Reactor Building outside the primary containment. The liquid is piped into the reactor vessel via the differential pressure and liquid control line and discharged near the bottom of the core lower support plate through a standpipe so it mixes with the cooling water rising through the core (see Sections 4.2, "Reactor Vessel and Appurtenances Mechanical Design," and 3.3, "Reactor Vessel Internals Mechanical Design"). 3.8-2

BFN-28 The Boron-10 isotope absorbs thermal neutrons and thereby terminates the nuclear fission chain reaction in the uranium fuel. The specified neutron absorber solution is enriched sodium pentaborate (Na2B10016-10H20). It consists of a mixture of borax, enriched boric acid, and demineralized water prepared in accordance with approved plant procedures to ensure the proper volume and enriched sodium pentaborate concentration is present in the standby liquid control tank. A sparger is provided in the tank for mixing, using air. To prevent system plugging, the tank outlet is raised above the bottom of the tank and is fitted with a strainer. At all times when it is possible to make the reactor critical, the configuration of the Standby Liquid Control System shall satisfy the following equation: C Q E 1.0 8.7 WT% 50 GPM 94 ATOM% C = sodium pentaborate solution weight percent concentration Q = SLCS pump flow rate in gpm E = Boron-10 atom percent enrichment in the sodium pentaborate solution The SLC system is used to control the Suppression Pool pH in the event of a DBA LOCA by injecting sodium pentaborate into the reactor vessel. The solution is then transported to the suppression pool by mixing with the ECCS flow circulating through the reactor and flowing out of the recirculation break into the suppression chamber. The amount of sodium pentaborate solution that must be available for injection following a DBA LOCA is determined as part of the DBA LOCA radiological dose analysis. This quantity is maintained in the storage tank as specified in the Technical Specifications. The solution concentration is normally limited to a maximum of 9.2 weight percent to preclude unwanted precipitation of the sodium pentaborate. The saturation temperature of the 9.2 percent solution is 40F which provides a 10F thermal margin below the lowest temperature predicted for the SLCS equipment area. Tank heating components provide backup assurance that the sodium pentaborate solution temperature will never fall below 50. The sodium pentaborate solution concentration is allowed to be 9.2 weight percent provided the concentration and temperature of the solution are within the limits permitted by the technical 3.8-3

BFN-27 specifications. High or low temperature, high or low liquid level, or a shorted heater causes an alarm in the control room. Tank level indication is also provided in the control room. Each positive displacement pump was originally sized to inject sodium pentaborate into the reactor in 50 to 125 minutes (approximately 50 gpm), depending on the amount of solution in the tank, at the reactor vessel maximum operating pressure. The minimum quantity of enriched sodium pentaborate is injected when required in less than 2 hours. The pump and system design pressure is 1500 psig. The two relief valves are set at approximately 1425 psig to exceed the reactor operating pressure by a sufficient margin to avoid valve leakage. To prevent bypass flow from one pump, in case of relief valve failure in the line from the other pump, a check valve is installed downstream of each relief valve line in the pump discharge pipe. A bladder-type pneumatic-hydraulic accumulator is installed on the discharge piping near each relief valve to dampen pulsations from the pumps to protect the system. Unit 1 is equipped with a maintenance-free suction accumulator at the SLC pump-inlet flange to provide suction stabilization and protect the system. Unit 2 is equipped with a maintenance-free suction accumulator at the SLC pump-inlet flange to provide suction stabilization and protect the system. Unit 3 is equipped with a maintenance-free suction accumulator at the SLC pump-inlet flange to provide suction stabilization and protect the system. The two explosive-actuated injection valves provide high assurance of opening when needed and ensure that the boron solution will not leak into the reactor even when the pumps are being tested. The valves have a demonstrated firing reliability in excess of 99.99 percent. Each explosive valve is closed by a plug in the inlet chamber. The plug is circumscribed with a deep groove so the end will readily shear off when pushed with the valve plunger. This opens the inlet hole through the plug. The sheared end is pushed out of the way in the chamber and is shaped so it will not block the ports after release. The shearing plunger is actuated by an explosive charge with dual ignition primers inserted in the side chamber of the valve. Ignition circuit continuity is monitored by a trickle current, and an alarm occurs in the control room if either circuit opens. Indicator lights show which primer circuit is opened. To service a valve after firing, a 6-inch length of pipe (spool piece) must be removed immediately upstream of the valve to gain access to the shear plug. The Standby Liquid Control System is actuated by a five-position spring return to "normal" keylock switch located on the control room console. The keylock feature 3.8-4

BFN-27 ensures that switching from the "stop" position is a deliberate act (safety design basis 7). Momentarily placing the switch to either "start A" or "start B" position starts the respective injection pump, opens both explosive valves, and closes the Reactor Water Cleanup System isolation valves to prevent loss or dilution of the boron solution. A green light in the control room indicates that power is available to the pump motor contactor, but that the contactor is open (pump not running). A red light indicates the contactor is closed (pump running). A white light indicates that the motor has tripped or the local handswitch is in the test position. A red light beside the switch turns on when liquid is flowing through an elbow style flow meter and associated flow indicating switch downstream of the explosive valves. If the flow light or pump lights indicate that the liquid may not be flowing, the operator can immediately turn the switch to the other side, which actuates the alternate pump. Crosspiping and check valves assure a flow path through either pump and either explosive valve. The chosen pump will start even though its local switch at the pump is in the "stop" position for test or maintenance. Pump discharge pressure indication is also provided in the control room. Equipment drains and tank overflow are piped not to the waste system but to separate containers (such as 55-gallon drums) that can be removed and disposed of independently to prevent any trace of the boron solution from inadvertently reaching the reactor. Instrumentation is provided locally at the standby liquid control tank consisting of solution temperature indication and control, tank level, and heater status. 3.8.4 Safety Evaluation 3.8.4.1 Reactivity Control The Standby Liquid Control System is a special safety system not required for normal plant operation, and is never expected to be needed for reactor shutdown because of the large number of control rods available to shut down the reactor. The system is designed to make the reactor subcritical from rated power to a cold shutdown (MODE 4) at any time in core life. The reactivity compensation provided will reduce reactor power from rated to the after-heat level and allow cooling the nuclear system to normal temperature with the control rods remaining withdrawn in the rated power pattern. It includes the reactivity gains due to complete decay of the rated power xenon inventory. It also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced Doppler effect in uranium, reduction of neutron leakage from the boiling to cold condition, 3.8-5

BFN-28 and decreasing control rod worth as the moderator cools. A licensing analysis is performed each cycle to verify adequate SLCS shutdown capacity. The analysis assumes the specified minimum final concentration of boron in the reactor core and allows for calculational uncertainties. The SLCS shutdown capacity is reported in Appendix N. The specified minimum average concentration of natural boron in the reactor to provide the specified shutdown margin, after operation of the Standby Liquid Control System, is 720 ppm (parts per million). The minimum quantity of sodium pentaborate to be injected into the reactor is calculated based on the required 720 ppm average concentration in the reactor coolant, Boron-10 enrichment, the quantity of reactor coolant in the reactor vessel, recirculation loops, and the entire RHR system in the shutdown cooling mode, at 70F and reactor normal water level. The result is increased by 25 percent to allow for imperfect mixing, leakage, and volume in other piping connected to the reactor. This minimum concentration is achieved by preparing the solution as defined in paragraph 3.8.3 and maintaining it above saturation temperature. This satisfies safety design basis 5. Cooldown of the nuclear system will take several hours, at a minimum, to remove the thermal energy stored in the reactor, cooling water, and associated equipment, and to remove most of the radioactive decay heat. The controlled limit for the reactor coolant temperature cooldown is 100F per hour. Normal operating temperature is about 550F. Usually, shutting down the plant with the main condenser and various shutdown cooling systems will take 10 to 24 hours before the reactor vessel is opened, and much longer to reach room temperature (70F). The addition of RHR shutdown cooling volume results in the dilution of the dissolved boron. Therefore, the pressure at which RHR shutdown cooling is activated represents the point of maximum reactivity, when the control rods are still withdrawn, and is the point which requires the maximum boron. Analyses are performed to bound the saturation temperature at this pressure using the equivalent of 720 ppm at 70 F. Analyses demonstrating that adequate shutdown capability exists under these conditions ensure that safety design basis 2 is met. The specified boron injection rate is limited to the range of 7 to 40 ppm per minute change of boron concentration in the reactor pressure vessel and recirculation loop piping water volumes. The lower rate ensures that the boron is injected into the reactor in less than 2 hours, which is considerably faster than the cooldown rate. The upper limit injection rate insures that there is sufficient mixing such that the boron does not recirculate through the core in uneven concentrations which could possibly cause asymmetric power oscillations in the core. This satisfies safety design basis 3. 3.8-6

BFN-27 3.8.4.2 Suppression Pool pH Control The Standby Liquid Control System is required to supply sodium pentaborate solution for post-LOCA events that involve fuel damage to maintain the suppression pool pH at or above 7.0. The radiological dose analysis for the DBA LOCA assumes concentrations of iodine species consistent with a suppression pool pH at or above 7.0 (i.e., re-evolution of iodine to the containment atmosphere is not considered). The quantity of sodium pentaborate necessary to offset the post-LOCA production of acid and maintain the suppression pool pH at or above 7.0 has been documented as part of the LOCA radiological dose analysis. This quantity is maintained in the storage tank as specified in the technical specifications. Maintaining the suppression pool pH at or above 7.0 is a concern following a DBA LOCA involving fuel damage. With a LOCA involving a recirculation pipe break, there will be sufficient flow from the ECCS systems through the reactor vessel and out of the break to transport the buffering agent to the suppression pool. The calculation methodology for suppression pool pH control was based on the approach outlined in NUREG-1465 and NUREG/CR-5950. The design inputs were conservatively established to maximize the post-LOCA production of acids and to minimize the post-LOCA production and/or addition of bases. Other design input values such as initial suppression pool volume and pH were selected to minimize the calculated pH. It is expected that the initial effects on post-accident suppression pool pH will come from rapid fission product transport and the formation of cesium compounds, which would result in increasing the suppression pool pH. However, cesium compounds are not credited in the long-term pH analyses and the determination of the final (30 day) pH value. As radiolytic production of nitric acid and hydrochloric acid proceeds, and these acids are transported to the pool over the first days of the event, the pH would become more acidic. The buffering effect of sodium pentaborate solution injection within several hours is sufficient to offset the affects of these acids that are transported to the suppression pool. In these events, the addition of a buffering agent to the suppression pool offsets the radiolysis production of acids. This satisfies safety design basis 8. 3.8.4.3 System Safety Evaluation The Standby Liquid Control System is classified as a special safety system. To assure the availability of the Standby Liquid Control System, two sets of the components required to actuate the pumps and explosive valves are provided in parallel for redundancy (safety design basis 6). 3.8-7

BFN-28 The SLC components required for the performance of the safety-related suppression pool pH control function are qualified for the post-LOCA environmental conditions they will be subjected to during the performance of this function. The Standby Liquid Control System is designed as a Class I system for withstanding the specified earthquake loadings (see Appendix C). Nonprocess equipment such as the test tank is designed as Class II. The system piping and equipment are designed, installed, and tested in accordance with USAS B31.1.0, Section I. For the reactivity control function, the Standby Liquid Control System is not required to be designed to meet the single failure criterion because it serves as a backup to the control rods. System reliability is enhanced by providing redundancy of pumps and valves. Hence, redundancy is not required for the tank heater or the heating cable. For the suppression pool pH control function, the SLC system does not completely meet the single failure criteria with regard to the containment isolation check valves and the main control room selector switch. Although a single failure to open one of the two check valves could prevent SLC injection, the potential for failure is very low based on the quality as established by its procurement as an ASME, Section III, Class 2 safety-related valve, periodic testing and inspection, and historical performance of the component. Also, although a failure of the selector switch in the main control room could prevent either train or both trains of injection from functioning, the switch is a highly reliable component at an accessible location. The switch could be easily replaced or bypassed to start one of the SLC trains if it were to fail. The Standby Liquid Control System is required to be operable in the event of a station power failure so the pumps, valves, and controls are powered from the standby AC power supply in the absence of normal power. The pumps and valves are powered and controlled from separate buses and circuits so that a single failure affecting power supply will not prevent system operation. The essential instruments and lights are powered from the 120-V AC instrument power supply. The Standby Liquid Control System and pumps have sufficient pressure margin, up to the system relief valve nominal setting of 1425 psig, to assure solution injection into the reactor at a pressure of at least three percent above the lowest setpoint of the main steam relief valves (1174 psig). The nuclear system is protected from overpressurization during operation of the Standby Liquid Control System positive displacements pumps by the nuclear system main steam relief valves. For the Standby Liquid Control System suppression pool pH function, this operating condition is consistent with other two train systems. Although the Standby Liquid Control System is not strictly a two train system, active components most 3.8-8

BFN-27 susceptible to failure (e.g., pumps and squib valves) are redundant which provides additional assurance that most single failures will not impede the ability of the system to perform its function. Only one of the two standby liquid control pumps is needed for proper system operation. If one pump is inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. The system pumps are powered by a diesel backed source and are not load shed. The period during which one redundant component upstream of the explosive valves may be out of operation will be consistent with the very small probability of failure of both the control rod shutdown capability and the alternate component in the Standby Liquid Control System, together with the fact that nuclear system cooldown takes 10 or more hours while liquid control solution injection takes less than 2 hours. This indicates the considerable time available for testing and restoring the Standby Liquid Control System to operable condition after testing while reactor operation continues. Assurance that the system will still fulfill its function during repairs is obtained by demonstrating operation of the operable pump. It can be seen that the Standby Liquid Control System satisfies safety design basis 1. 3.8.4.4 Quality Assurance The equipment that performs the special safety functions of the Standby Liquid Control System (provide a backup method to make the reactor subcritical and provide sufficient buffering agent to maintain the suppression pool pH at or above 7.0 following a DBA LOCA involving fuel damage) are classified as quality related. As delineated by condition of the Units 1, 2, and 3 BFN Operating Licenses, the Augmented Quality Program for the Standby Liquid Control System provides the quality control elements to ensure component reliability for the required alternative source term function as governed by the BFN Quality Assurance Program. 3.8.5 Inspection and Testing Operational testing of the Standby Liquid Control System is performed in at least two parts to avoid injecting boron into the reactor inadvertently. By opening two closed valves to the solution tank, the boron solution may be recirculated by turning on either pump with its local switch. With the valves to and from the solution tank closed and the three valves opened to and from the test tank, the demineralized water in the test tank can be recirculated by turning on either pump locally. After pumping boron solution, demineralized water is pumped to flush out the pumps and pipes. Functional testing of the injection portion of the system is accomplished by closing the open valve from the solution tank, opening the closed valve from the test tank, and actuating the switch in the control room to either the A or B circuit. This starts one pump and ignites one of the explosive actuated injection valves to open. 3.8-9

BFN-27 The lights and alarms in the control room indicate that the system is functioning. This satisfies safety design basis 4. After the functional test, the affected injection valve and explosive charge must be replaced and all the valves returned to their normal positions as indicated in Figures 3.8-1, 3.8-2, 3.8-3, 3.8-5, and 3.8-6. By closing a local normally open valve to the reactor in the containment, leakage through the injection valves can be detected at a test connection in the line between the containment isolation check valves. (A position indicator light in the control room indicates when the local valve is full open and ready for operation.) Leakage from the reactor through the first check valve can be detected by opening the same test connection whenever the reactor is pressurized. The test tank contains sufficient demineralized water for testing pump operation. Demineralized water from the makeup or condensate storage system is available at 30 gpm for refilling or flushing the system. Should the boron solution ever be injected into the reactor, either intentionally or inadvertently, then after making certain that the normal reactivity controls will keep the reactor subcritical, the boron is removed from the reactor coolant system by flushing for gross dilution followed by operation of the reactor cleanup system. There is practically no effect on reactor operations when the boron concentration has been reduced below about 50 ppm. The sodium pentaborate solution weight percent in the SLCS storage tank is periodically determined by titration or equivalent chemical analysis. The Boron-10 isotopic atom percent concentration of the solution is also determined periodically, utilizing mass spectrometry or equivalent technology. The gas pressure in the discharge accumulators is measured periodically to detect leakage. A pressure gauge and portable nitrogen supply are required to test and recharge the accumulators. 3.8-10

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1. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 2 ANO SYSTEM 63 UNLESS OTHERWISE 1-1/2" J J SPOOL FOR SERVICING l---'-1*_ _, NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNI ARE NOT DEPICTED.

                                                                                                         ~     ~       EXPLOSIVE VALVE                                                                                                                                                                                                                                                                                                                 FOR ADDITIONAL GUIDANCE, REFER TO NEDP-4.

CKV-525 ISV-524 ( TYP 2),c=:-,_ 1 rev,

                                                                                                                                                                                                                      .                  RFV-512    1                                           .                 RFV-513 1                                                                                                                           EXAMPLES: MEL UNID                                                                   DRAWING UNID N                                                                         N m

N S"lJ~B ' ' '=aaaF""N-""'o'"'-"s"Hv~-~0~1~8~-0~5~0=2 BFN-1-HS-031-0160A SHV-18-502 HS-31-160A

                                                                                                                                                                                                                                     -SET TO RELIEVE o                                                                                                                                                                                                                BFN-2-PT-00>-0204C                                                   PT-3-204C cc                                                  1-1/2"      ,i,J_ ..       ..                               -                                                                                                                      SET TO RELIEVE o 1425 PSIG                                                                                                                                                                                               2. ALL VALVES ARE SA'-£ SIZE AS PIPE UNLESS OTHERWISE NOTED.
                                                                   ~

N cc PIPE-12B-1"-'

                                                                                                                                                            **~~

N 0 LY-1I

• N

                                                                                                                                                                                                             -'         STANDBY LIQUID catTROL PUMP 50 GPM o 1425 PSIG

3. OPERATIONAL VALVES ARE SHOWN IN THEIR NORMAL OPERATING POSITION.

4. [I].[]) . ETC, DENOTE DESIGN PRESSURE ANO TEMPERATURE AS GIVEN IN DATA TABLE ON THIS DRAWING.

                                                           "                                                                                                 ,:::                                      >/4"             1500 PSIG                                              >/4"

5. DELETED.

6. DELETED.

                                                                                                                                                               "'          I         1:11 CHECK VALVE TEST CONN
                                                                                                                                               +--*~Q PMP-6A                  lrn VTV-573 EL 639.0 EL 621.25 VTV-574 EL 639.0 EL 621.25

7. VENT, DRAIN, AND TEST CONNECTIONS 1-1/2* AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399.

                                            "" t.tPEN-
                                            "t  100-42
                                                                                                                                                                                            .      ACC-578                                                                   ACC-579 r-EL 639.0                                                                                                                                                                                                                                             D T

X EL 639.0 EL 639.0

,-47E854-1,G1 1"

DRV-575 I 17 ~ m

                                                                                                                                                                                                               ;:;'                        .   '              -  1*
                                                                                                                                                                                                                                                                                 -21 EL 621.25                                              EL 621.25 REFERENCE DRAWINGS:

l2-47E854-1,G1 )>------1N 110111 17 21 MEL ................... TABULATION VALVE MARKER TAGS CKV-508 SHv-5091 "' cKV-s10 sHv-s11 *N MEL .....*............. TABULATION OF INSTRLt.tENTS 0-47W462-1, -3 ........ PIPING DRAWINGS 2-471462-2 ............ PIPING DRAWINGS 3/4" mlm m--..r;i PUMP BASE & / - 0-47W600-56 ........... MECHANICAL - INSTRUMENTS AND CONTROLS 2-47E610-3-SERIES ..... CONTROL DIAGRAM - REACTOR FEEDWATER SYSTEM 2-47E610-63-1 ......... CONTROL DIAGRAM - STANDBY LIQUID CONTROL SYSTEM 2-47E610-68-SERIES .... CONTROL DIAGRAM - REACTOR WATER RECIRCULATION SYSTEM SHV-538 SHV-539 FLOOR DRAIN~ EL 639.0 - - 2-47E610-75-SERIES .... CONTROL DIAGRAM - CORE SPRAY SYSTEM 2-47E610-85-SERIES .... CONTROL DIAGRAM - CONTROL ROD DRIVE SYSTEM EL 621.25 47E800-1 .............. FLOW DIAGRAM - GENERAL PLANT SYSTEMS

                        ~

0-47E800-2 ............ MECHANICAL SYMBOLS ANO FLOI DIAGRAM DRAWING INDEX

                        ~

J ISV 63-1

                                                    ~                                                                                                                                           EL  639 .0                                                              EL 639 .0 2-47E817-1 ............ FLOW DIAGRAM - REACTOR WATER RECIRCULATION SYSTEM 0-47E845-2 ............ FLOI DIAGRAM - COMPRESSED AIR-STATION SERVICE 2-47E856-2 ............ FLOW DIAGRAM - DEMINERALIZED WATER

(=

                        ~

GE DRAWINGS:

                      "0u "'
                   *    ~
                        ~
                           ~
                                                                                                                                                                                 ,1!41------------------..--------

EL 621 .25 2-1/2" EL 621 .25 919D926 ............... FUNCTIONAL CONTROL SYSTEM - SLC SYSTEM 22A1315 ............... DESIGN SPECIFICATION - SLC SYSTEM 22A1315AB ............. DESIGN SPECIFICATION - SLC SYSTEM m '~ N

                                                                                                                                                                                                                                                                                                                                                                                           -'                                                                                                                                                                           C m
                           ~
                           ~

N SYMBOL:

                                                                                                                                                                                                                                                                                                                                                                                                                                                >-I-~0"--!v~---'>I HEATING TAPE
                      "'"' '                                                                                                                                                           N
                        ~  N
       * *            "                                                                                                                                                                                                                                                                                                                                                                                                                               FAIL AS JS N

0 O

• 0 0

                              ~

REACTOR PRESSURE VESSEL

                             .~

J

                              ~

15 "'.~ 0 J

                                   -m
                                    ~
                              ~

m

                                    ~
                                    ~
                                                                                                                                                                                                                                                     ' 1----',DRV 63-1'
                                                                                                                                                                                                                                                                ~

B

                              ~

N PRESS. ABOVE CORE PLATE .

                                                                                                                                                                                                                                                                                                            .                                                                                                        ~

LIQUID CONTROL N 2* -' "

• AMENDMENT 28 POWERHOUSE UNIT 2 N~~ '

                                                                                              ;_ >-~PRESS. BELOW                                                                                                                                                                                                                                                                                             DRAIN CART TNK-584 BROWNS FERRY NUCLEAR PLANT CORE PLATE DETAIL 2A1 SYSTEM PRESS.- TEMP DATA PRESS.         DESIGN         DESIGN                                                                                                         DRAIN VALVE FINAL SAFETY ANALYSIS REPORT TEMP PRESSURE       (PSJG) TEMP.(°F)                                                                                                         DRV-542---1.

INDEX 1500 150 EL 621 .25~ "--... 2 150 150 ' J 3.14"1

• A 3

4 1150 100 560 1"0 STANDBY LIQUID CONTROL SYSTEM 5 6 67 1250 150 575 COMPANION DRAIINGS: 1-47E854-1 FLOW DIAGRAM 7 110 100 3-47E854-1 FIGURE 3.8-1 8 7 6 5 4 3

lt0ML-t9-O1 9 ] Lt- Z l'I £9 H G d

                            ' __ ~.,. ,.,------- --------------@        '

D L Y,.,.,.,,.,. '

                                  ~r C

B AMENDM EN T 27 POIERHOUSE UNIT2 BROWNS FERRY NUCL EAR PLANT F I NAL SAFE TY ANALYSI S REPORT STANDBY LIQUID CONTROL SYST EM A MECHA NICAL CONTROL DIAGRAM F I GURE 3. 8 - 2

l-tS83Lt- l Vl L9 SERVICE AIR 1* SHV-536

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  .  -        1/2' 0-47E845-2, HS   I CKV-535 3/4" VENT
                                                                                                                                          ,                                                                                                                                                                                                                                                                                                                                                       -m
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                   -                                       H DEMINERALIZED WATER ij                                                                                                                                                                                                                                                                                                                                                1*
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  .-        IF     K1-47E856-2, GB HATCH FOR ADDING CHEMICALS & TAKING SAIO'LES\

SHV-534 CKV-533 g ~ 4~ VENT 1 -* SHV-520 1-1/2" 2"

                                                                                                                                                     -                                                                                                                                                          -                                                                                  DEMINERALIZED WATER
                                                                                                                                                                                                                                                                                                                          .                                                                                                                              TNK 583
                                                                                                                                                                                                                                                                                                                          "'                                                                     I   1-47E856-2, GB
                                                                                                                                                                   ,------------------------------7-------------------                                                                                                                                                                                                                                                                n OEMINERALIZEO WATER STANDBY LIOU ID ITT        -'57 I

I 0 1:-_-:4::7::EB:,5:,6:-_-:-,-,-:G::8--.,.._ CONTROL TANK (GROSS VOL- 4901 BJ-2 G

                                                                                                                                                  -           -I                                                                                                                                                                                                                                                             IF                    GAL)

I I I I I I 4 I 1-1/2'

                                                                                                                                                                                                                                                                                                                                                    +--'"---1~~--!.'........---,0-F SHV-519                                                                                                                                                                                                  SHV-532         CKV-531 TEST TANK                                                                                                   HTR-2 2

(VOL* 210 GAL} TNK 582 ELE~ 2-1/2" CKV-SOJ SHV-14 ----,--_,_- DURING MIXING)_/ ~ (CLOSE VALVE /SHV-SOO EXPANSION LOOP 1* \_ AIR SPARGER PIPE-13A SPOOL FOR SERVICING EXPLOSIVE VALVE RTV-7 256-1/8- DIA HOLES IN SPARGER F 0 1/col2~*..,,,.__VEHT

                                                                                                                                                                                                                                                                                                                                               ~
                                                                                                                                                                                                                                                                                                                                               ~
                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ~

0 0

                                                                                                                               "        *                                                                                                                                                                                                  "0                                                                                                                  ~
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                                                                                                                                                                                                                                                                                                                                            ~                                                                                                                  ~

I ~ w

                                                                                                                                                                                                                                    "'                                     I                                               '
                                                                                                                                                                                                                                                                                                                           "'              "'w                                                                                                               ",_

I ~

                                                                                                                                                                                                                                                                                                                                                                                                                                                              "0
                                                , ' ~VENT OR                                                                                                                                                         PI I

I PI

                                                ~ r         TEST CONN                                                                                                                                                                                                      I
                                                ~     :;:

080 I I 581 i ACCUMULATOR,

                                                                                                                                                                                                           "    <I-SHV-580 I  ACCUMULATOR                 <I-
                                                                                                                                                                                                                                                                                                             '  SHV-581                                                                                                                                        -'

13 -1 I 1-1/2" I

                                                                                                                                                                                                               ' \ ..._       ACC-580 I

I I ACC-581 5 NOTES,

1. ALL VALVE NUMBERS BEAR A SYSTEU DESIGNATION PREFIX (BJ).

E PIPE-128 I 1 1 2 r,;,_-,;~1;;,_".1.,"T_ _ t.,,...----f>,c:J,-----,i CKV-525 SHV-524 e" FE' ,.__-I SPOOL FOR SERVICING EXPLOSIVE VALVE - 1* RFV-512 ITJ I I I

                                                                                                                                                                                                                                                                                                 ..              1*

RFV-513 1 2. J. 4. ALL VALVES ARE SAME SIZE AS PIPE UNLESS OTHERWISE NOTED. DELETED OPERATIONAL VALVES ARE SHOWN IN THEIR NORMAL OPERATING POSITION. I 5, [I] , (!] , ETC., DENOTE DESIGN PRESSURE AND TE*PERATURE

                                                                                                                                                                                                                 -'                   SET TO RELIEVE o 1425 PSIG I

I

                                                                                                                                                                                                                                                                                                           -'                                                                                                                                                  AS GIVEN IN DATA TABLE ON THIS DRAWING.

B. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS "11..J ' 121-- I SET TO RELIEVE e 1425 PSIG SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 1 UNLESS OTHERWISE NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED.

                                                                                                                                                                                                !:!                                                                                                                                                                                                                                                            FOR ADDITIONAL GUIDANCE, REFER TO NEOP-4.

EXAMPLE: MEL UNID DRAWING UNID PUMP 1A ll 1B BFN-O-SHV-018-0502 SHV-18-502

                                 '                                                                                                                                                                                            50 GPM o                                                                                                                                                                                                                                         BFN-1-HS-OJ1-0160A     HS-J1-160A 8                                                                                        1500 PSIG                                                                                                                                                                                                                                        BFN-2-PT-003-0204C     PT-J-204C LlJ Fllll, A           1-rn                           "<w .,
                                                                                                                                                                                                                    ~
                                                                                                                                                                                                                        -                               '          Fl]
                                                                                                                                                                                                                                                       /1!,=L.l-!=1;;::B::!J          1--m

7. VENT. DRAIN. AND TEST CONNECTIONS 1-1/2- AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399.

                                                                                                                                                                                                                                                                                ~~-

DRYWELL MTR-BA MTR-6B PENETRATION C

                                                                                                                                                                           \_IPMP~6A                                w    N N                                                                                                                                                                                                                                   REFERENCE DRAIINGS'.
                                                                                                                                                                                                                    ~    .,'                                                             PMP~BB VTV-574                                                                                                                                                             MEL ................... TABULATION VALVE MARKER TAGS AND INSTRUMENTS VTV-573
                               *                                                                                                                            ~                                                     "zw ......w                                                    -                                                                                                                                                                      0-471462-1, -2, -3 .... PIPING DRAWINGS                                            D
                               "'                                                                                    ~                                                                                            -                                                                                                           -EL 639.0                                                                                                                 1-47E610-63-SERIES,., .CONTROL DIAGRAM v                                   "w' -'

I i..,.---ACC-579 0-47E800-1 ............ FLOW DIAGRAM-GENERAL PLANT SYSTEMS DEMINERALIZED WATER I @]

                                                                                                                                                                                         *cc-578                   C

• 0-47E800-2 ............ MECHANICAL SYMBOLS AND FLOW DIAGRAM DRAWING INDEX 1[~1-~4~7~E~85~6~-!2~,~G~s::::::>---( IF )--f'"--Jf---ol-,::----

1*

                                                                                                                                                                                                             -                                                   r.......,6_                                                 ,

J CKV-508 SHV-509 DRV-575

                                                                                                                                                                                                      '° l(0 '    CKY-510   SHV-511 1
                                                                                                                                                                                                                                                              ;RV-576
                                                                                                                                                                                                                                                                              '67-       ....fz]
                                                                   -                                                                 61                              fl;i1J6;J-<+---l?lU2LI
                                                                                                                                                                     -                 -            7~
                                                                                                                                                                                                                                                                              -            -                         1" SHV-538           SHV-539

,; 12 61 en PUMP BASE A FD _I/

Tl 3-4 C

..                                                                                                                                                                                                        2-1/2" i-;

Q D PRESSURE BELOI CORE PLATE -" z

                                                                                                                                                                                                                                                 -g 1  1-47EB17-1, DJ                                                                                                                                                                                  z 0
                                                                                                                                                                                                                                                  ~

REACTOR VESSEL "'* B AMENDMENT 28 FRCN STANDBY LIQUID CONTROL 1" PANEL 25-19 l'-'o----*1__*--*--00'----s--*_~,,.>------, POWERHOUSE I U UNIT 1 DRAIN CART BROWNS FERRY NUCLEAR PLANT TNK-584 SYSTEM PRESS. PRESS. DESIGN ll TEMP DATA DESIGN FINAL SAFETY ANALYSIS REPORT TEIO' DRAIH--h INDEX PRESSURE (PSIG) TEMP(DF) 2 3 5 6 1500 150 1250 100 67 110 150' 150' 575* 150' 1S0' 100* CCMPANION DRAWINGS: EL 621.25-.

                                                                                                                                                                                                                                                                                                                                                                                                                                 ,~,1*1,                               STANDBY LIQUID CONTROL SYSTEM FLOW DIAGRAM A

2-47E854-1 1-47E854-1-ISI 3-47E854-1 FIGURE 3.8-3 8 7 6 5 4 3

BFN-22 Figure 3.8-4 (Deleted by Amendment 22)

F - *- ** m* H I---1 i

                   ~--------""'----~---------'--~ ~
                                                                                                       --- ~  T G

i '""'""""'"'""****** ~

                    .,,,~~i                           i
                    "' '" ' ~: i              ,------,----7 I-                                                                                                                D T

C B AMENDMENT 27

     ~~~.,,_,_                                                     ,,,._;,,,_::;
                                                                   ~

POWERHOUSE UNITJ BROWNS FERRY NUCL EAR PLANT

    """""'                                                                                  FINAL SAFETY ANALYSI S REPORT

!!:Ol"-

A STANDBY LFt~w1Dof?tRTlJ>L SYS TE~ tmr.::: FIGURE 3 . 8-5

ooo~ r- U - Ol 9l Lt - t l'I L9 H G D tt!liii'.°'.'* **** ****f,li!,'i1Ji1!,;',i/{.,;:.:o1/4lll'" maa.. --. *** Brtii;.wm?11t,;"""""'

                                ***mm,, **.B:oill1*1Bi:!1!i: ,111,.:,,,.,,.,              C B

AM END MENT 27 POWERHOUSE UNITJ BROWNS FERRY NUCL EAR PLANT FI NAL SAFE TY ANAL YS I S REPORT STA NDBY LI QUID CONTROL SYS TE~ A

                                      ~ECHA NICAL CONTROL DI AGR AM F I GURE 3. 8 - 6

BFN-25 Figure 3.8-7 (Deleted by Amendment 25)

1100* f-U-0 19lH- t l'I £9 H G

                          *-~"-"r-~-1'
                          *t'~L..~r                                           '

' l' ,. - , ,,... ,, ,

I

                                                               ~--- --------                 !

L--~ ~~------------------------@-: ____________ j C B AME ND MEN T 27 POIERHOUSE UNIT! BROWNS FERRY NUCL EAR PLANT FI NAL SAFE TY ANALY S I S REPORT ST ANOBY LI OU IO CONTROL SYS A MECHA NICAL CONTROL DIAGRAM F I GURE 3. 8 - 8

BFN-29 REACTOR COOLANT SYSTEM TABLE OF CONTENTS 4.0 Reactor Coolant System............................................................................................................ 4.1-1 4.1 Summary Description ................................................................................................................ 4.1-1 4.2 Reactor Vessel And Appurtenances Mechanical Design........................................................... 4.2-1 4.2.1 Power Generation Objective ...................................................................................... 4.2-1 4.2.2 Power Generation Design Basis ................................................................................ 4.2-1 4.2.3 Safety Design Basis ................................................................................................... 4.2-1 4.2.4 Description ................................................................................................................. 4.2-2 4.2.5 Safety Evaluation ....................................................................................................... 4.2-10 4.2.6 Inspection and Testing ............................................................................................... 4.2-15 4.3 Reactor Recirculation System ................................................................................................... 4.3-1 4.3.1 Power Generation Objective ...................................................................................... 4.3-1 4.3.2 Power Generation Design Basis ................................................................................ 4.3-1 4.3.3 Safety Design Basis ................................................................................................... 4.3-1 4.3.4 Description ................................................................................................................. 4.3-1 4.3.5 Safety Evaluation ....................................................................................................... 4.3-7 4.3.6 Inspection and Testing ............................................................................................... 4.3-8 4.4 Nuclear System Pressure Relief System ................................................................................... 4.4-1 4.4.1 Safety Objective ......................................................................................................... 4.4-1 4.4.2 Power Generation Objective ...................................................................................... 4.4-1 4.4.3 Safety Design Basis ................................................................................................... 4.4-1 4.4.4 Power Generation Design Basis ................................................................................ 4.4-2 4.4.5 Description ................................................................................................................. 4.4-2 4.4.6 Safety Evaluation ....................................................................................................... 4.4-9 4.4.7 Inspection and Testing ............................................................................................... 4.4-10 4.5 Main Steam Line Flow Restrictor ............................................................................................... 4.5-1 4.5.1 Safety Objective ......................................................................................................... 4.5-1 4.5.2 Safety Design Basis ................................................................................................... 4.5-1 4.5.3 Description ................................................................................................................. 4.5-1 4.5.4 Safety Evaluation ....................................................................................................... 4.5-2 4.5.5 Inspection and Testing ............................................................................................... 4.5-3 4.6 Main Steam Isolation Valves ..................................................................................................... 4.6-1 4.6.1 Safety Objectives ....................................................................................................... 4.6-1 4.6.2 Safety Design Basis ................................................................................................... 4.6-1 4.6.3 Description ................................................................................................................. 4.6-2 4.6.4 Safety Evaluation ....................................................................................................... 4.6-5 4.6.5 Inspection and Testing ............................................................................................... 4.6-8 4.0-i

BFN-29 REACTOR COOLANT SYSTEM TABLE OF CONTENTS (Cont'd) 4.7 Reactor Core Isolation Cooling System ..................................................................................... 4.7-1 4.7.1 Power Generation Objective ...................................................................................... 4.7-1 4.7.2 [Deleted]..................................................................................................................... 4.7-1 4.7.3 Power Generation Design Basis ................................................................................ 4.7-1 4.7.4 Safety Design Basis ................................................................................................... 4.7-1 4.7.5 Description ................................................................................................................. 4.7-1 4.7.6 Safety Evaluation ....................................................................................................... 4.7-4 4.7.7 Inspection and Testing ............................................................................................... 4.7-4 4.8 Residual Heat Removal System (RHRS) .................................................................................. 4.8-1 4.8.1 Safety Objective ......................................................................................................... 4.8-1 4.8.2 Power Generation Objective ...................................................................................... 4.8-1 4.8.3 Safety Design Basis ................................................................................................... 4.8-1 4.8.4 Power Generation Design Basis ................................................................................ 4.8-2 4.8.5 Summary Description ................................................................................................. 4.8-2 4.8.6 Description ................................................................................................................. 4.8-4 4.8.7 Safety Evaluation ....................................................................................................... 4.8-8 4.8.8 Inspection and Testing ............................................................................................... 4.8-8 4.9 Reactor Water Cleanup System ................................................................................................ 4.9-1 4.9.1 Power Generation Objective ...................................................................................... 4.9-1 4.9.2 Power Generation Design Basis ................................................................................ 4.9-1 4.9.3 Description (Figures 4.9-1, 4.9-2, 4.9-3, 4.9-5, 4.9-6, 4.9-7, 4.9-8, 4.9-9, and 4.9-10) ................................................................................................ 4.9-1 4.9.4 Inspection and Testing ............................................................................................... 4.9-3 4.10 Nuclear System Leakage Rate Limits ....................................................................................... 4.10-1 4.10.1 Safety Objective ......................................................................................................... 4.10-1 4.10.2 Safety Design Basis ................................................................................................... 4.10-1 4.10.3 Description ................................................................................................................. 4.10-1 4.10.4 Safety Evaluation ....................................................................................................... 4.10-7 4.10.5 Inspection and Testing ............................................................................................... 4.10-7 4.11 Main Steam Lines, Feedwater Piping, and Drains ..................................................................... 4.11-1 4.11.1 Power Generation Objective ...................................................................................... 4.11-1 4.11.2 Safety Design Basis ................................................................................................... 4.11-1 4.11.3 Power Generation Design Bases ............................................................................... 4.11-1 4.11.4 Description ................................................................................................................. 4.11-1 4.11.5 Safety Evaluation ....................................................................................................... 4.11-3 4.11.6 Inspection and Testing ............................................................................................... 4.11-3 4.0-ii

BFN-29 REACTOR COOLANT SYSTEM TABLE OF CONTENTS (Cont'd) 4.12 Inservice Inspection And Testing ............................................................................................... 4.12-1 4.12.1 Introduction ................................................................................................................ 4.12-1 4.12.2 Scope ......................................................................................................................... 4.12-2 4.12.3 Responsibility ............................................................................................................. 4.12-2 4.12.4 Area and Extent of Examination................................................................................. 4.12-2 4.0-iii

BFN-29 REACTOR COOLANT SYSTEM LIST OF TABLES Table Title 4.2-1 Reactor Pressure Vessel Materials 4.2-2 Reactor Vessel Data 4.2-3 Reactor Vessel Attachments 4.3-1 Reactor Recirculation System Design Characteristics (3952 MWt) 4.3-1b (Deleted) 4.4-1 (Deleted) 4.4-1A Nuclear System Main Steam Relief Valves (Units 2 and 3) 4.7-1 Reactor Core Isolation Cooling System Turbine - Pump Design Data 4.8-1 Residual Heat Removal System Equipment Design Data 4.9-1 Reactor Water Cleanup System Equipment Design Data 4.0-iv

BFN-29 REACTOR COOLANT SYSTEM LIST OF ILLUSTRATIONS Figure Title 4.2-1 Reactor Vessel 4.2-2 Reactor Vessel Nozzles and Penetrations 4.2-3 Reactor Vessel 4.2-4 Reactor Vessel 4.3-1 Recirculation System--Elevation, Isometric 4.3-2a sht 1 Nuclear Boiler Flow Diagram 4.3-2a sht 2 Nuclear Boiler Flow Diagram 4.3-2a sht 3 Nuclear Boiler Flow Diagram 4.3-2b (Deleted) 4.3-3 Jet Pump--Operating Principle 4.3-4 Recirculation System--Core Flooding Capability 4.4-1 2-Stage Safety/Relief Valve Schematic (Closed Position) 4.4-2 2-Stage Safety/Relief Valve Schematic (Open Position) 4.4-3 Safety Valve Sizing Analysis 4.4-4 Deleted by Amendment 13 4.4-5 Deleted by Amendment 13 4.4-6 T-Quencher for Safety/Relief Discharge 4.4-7 Mechanical Main Steam Relief Valve Vent Piping 4.4-8 Mechanical Main Steam Relief Valve Vent Piping 4.5-1 Primary Steam Piping 4.5-2 Primary Steam Piping 4.5-3 Primary Steam Piping 4.6-1 Main Steam Isolation Valve 4.7-1a Reactor Core Isolation Cooling System Flow Diagram 4.7-1b Reactor Core Isolation Cooling System, Mechanical Control Diagram 4.7-1c Reactor Core Isolation Cooling System - Flow Diagram 4.7-1d Reactor Core Isolation Cooling System - Mechanical Control Diagram 4.7-1e Reactor Core Isolation Cooling System - Mechanical Control Diagram 4.7-1f Reactor Core Isolation Cooling System - Flow Diagram 4.7-2a (Deleted) 4.7-2b (Deleted) 4.7-2c (Deleted) 4.7-2d (Deleted) 4.7-2e (Deleted) 4.7-2f (Deleted) 4.7-2g (Deleted) 4.7-2h (Deleted) 4.8-1 Residual Heat Removal System, Unit Cross Connections and Standby Coolant Supply 4.9-1 Reactor Water Cleanup System Flow Diagram 4.9-2 Reactor Water Cleanup Demineralizer Flow Diagram 4.9-3 Reactor Water Cleanup System, Mechanical Control Diagram 4.9-4 (Deleted) 4.0-v

BFN-29 REACTOR COOLANT SYSTEM LIST OF ILLUSTRATIONS (contd) Figure Title 4.9-4a (Deleted) 4.9-4b (Deleted) 4.9-4c (Deleted) 4.9-4d (Deleted) 4.9-5 Reactor Water Cleanup System - Flow Diagram 4.9-6 Reactor Water Cleanup Demineralizer - Flow Diagram 4.9-7 Reactor Water Cleanup System - Mechanical Control Diagram 4.9-8 Reactor Water Cleanup System - Flow Diagram 4.9-9 Reactor Water Cleanup Demineralizer - Flow Diagram 4.9-10 Reactor Water Cleanup System - Mechanical Control Diagram 4.10-1 Drywell Leak Detection System Diagram 4.10-2 Deleted 4.10-3 Axial Through-Wall Crack Data Correlation 4.11-1 Feedwater Piping Arrangement 4.0-vi

BFN-16 4.0 REACTOR COOLANT SYSTEM 4.1

SUMMARY

DESCRIPTION The subsections in the "Reactor Coolant System" section describe those systems and components that form the major portions of the nuclear system process barrier. These systems and components contain or transport the fluids coming from or going to the reactor core. The "Reactor Vessel and Appurtenances Mechanical Design" subsection describes the reactor vessel and the various fittings with which other systems are connected to the vessel. The major safety considerations for the reactor vessel are concerned with the ability of the vessel to function as a radioactive material barrier. Various combinations of structural loading are considered in the vessel design. The vessel meets the requirements of various applicable codes and criteria. The possibility of brittle fracture is considered, and suitable limits are established that avoid conditions where brittle fracture is possible. Periodic, cumulative-fatigue usage evaluations are performed for each reactor vessel to verify that the vessel does not approach usage limits. The Reactor Recirculation System pumps coolant through the core. Adjustment of the core coolant flow rate changes reactor power output, thus providing a means of following plant load demand or manually changing reactor output without adjusting control rods. The recirculation system is designed with sufficient fluid and pump inertia that fuel thermal limits will not be exceeded as a result of recirculation system malfunctions. The arrangement of the recirculation system is designed so that a piping failure cannot compromise the integrity of the floodable inner volume of the reactor vessel. The Nuclear System Pressure Relief System is designed to protect the nuclear system process barrier from damage due to overpressure. To accomplish overpressure protection a number of main steam relief valves are provided that can discharge steam from the nuclear system to the primary containment. The Nuclear System Pressure Relief System also acts to automatically depressurize the nuclear system in the event of a loss-of-coolant accidents in which the High Pressure Coolant Injection System (HPCIS) fails to deliver sufficient flow. The depressurization of the nuclear system allows low pressure Emergency Core Cooling Systems to supply enough cooling water to adequately cool the fuel. Six of the main steam relief valves used to provide overpressure protection are arranged to effect automatic depressurization. The main steam line flow restrictors are venturi-type flow devices. One restrictor is installed in each main steam line close to the reactor vessel but downstream of the main steam relief valves. The restrictors are designed to limit the loss of coolant resulting from a main steam line break outside the primary containment. The 4.1-1

BFN-16 coolant loss is limited so that reactor vessel water level remains above the top of the core during the time required for the main steam isolation valves to close. This action protects the fuel barrier. Two main steam isolation valves are installed on each main steam line. One valve in each line is located inside the primary containment, the other outside. These valves act automatically to close off the nuclear system process barrier in the event a pipe break occurs downstream of the valves. This action limits the loss of coolant and the release of radioactive materials from the nuclear system. In the event that a main steam line break occurs inside the primary containment, closure of the isolation valve outside the containment acts to seal the primary containment itself. The Reactor Core Isolation Cooling System (RCICS) includes a turbine-pump driven by reactor vessel steam. Under certain conditions the system automatically starts in time to prevent the core from becoming uncovered without the use of the Core Standby Cooling Systems. The system provides the ability to cool the core during a reactor shutdown in which feedwater flow is not available. The Residual Heat Removal System (RHRS) includes a number of pumps and heat exchangers that can be used to cool the nuclear system under a variety of situations. During normal shutdown and reactor servicing, the RHRS removes residual and decay heat. One operational mode of the RHRS is low pressure coolant injection (LPCI). LPCI operation is an engineered safeguard for use during a loss-of-coolant accident; this operation is described in Section 6.0, "Emergency Core Cooling Systems." Another mode of RHRS operation allows the removal of heat from the primary containment following a loss-of-coolant accident. The Reactor Water Cleanup System functions to maintain the required purity of reactor coolant by circulating coolant through a system of filter/demineralizers. The "Nuclear System Leakage Rate Limits" subsection establishes the limits on nuclear system leakage inside the primary containment so that appropriate action can be taken before the nuclear system process barrier is threatened by a crack large enough to propagate rapidly. The main steam lines, feedwater piping, and their associated drains are attached to the reactor vessel and provide core coolant flow paths external to it. These lines penetrate the primary containment and specified portions of these lines must provide adequate nuclear system process barrier for normal and accident conditions. Four steam lines are utilized between the reactor and the turbine which permit turbine stop valve and main steam isolation valve tests during plant operation with a minimum amount of load reduction. In addition, differential pressures on reactor internals under assumed accident conditions of a broken steam line are limited. Feedwater lines provide water to the reactor vessel entering near the top of the 4.1-2

BFN-16 vessel downcomer annulus. Drains are provided at the low point of each main steam line, at the reactor vessel bottom head, and on each side of the recirculation pumps. The program for preoperational examination and periodic inservice examination of Reactor Coolant System is also defined. 4.1-3

BFN-27 4.2 REACTOR VESSEL AND APPURTENANCES MECHANICAL DESIGN 4.2.1 Power Generation Objective The reactor vessel power generation design objective is to provide a volume in which the core can be submerged in coolant, thereby allowing power operation of the fuel. The reactor vessel and appurtenances design provides the means for the attachment of pipelines to the reactor vessel and the means for the proper installation of vessel internal components. 4.2.2 Power Generation Design Basis

1. The location and design of the external and internal supports provided as an integral part of the reactor vessel shall be such that stresses in the reactor vessel and supports due to reactions at these supports are within ASME Boiler and Pressure Vessel Code limits.

2. The reactor vessel design lifetime shall be 40 years. Time Limited Aging Analyses (TLAAs) have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Section O.3.1 and O.3.2.

3. The design of the reactor vessel and appurtenances shall allow for the accomplishment of a suitable program of periodic inspection and surveillance.

4.2.3 Safety Design Basis

1. The reactor vessel and appurtenances shall be designed to withstand adverse combinations of loadings and forces resulting from operation under abnormal and accident conditions.

2. The reactor vessel shall be designed and fabricated to a high standard of quality to provide assurance of an extremely low probability of failure.

3. To minimize the possibility of brittle fracture failure of the nuclear system process barrier, the following shall be required: (1) the initial ductile-brittle transition temperature of materials used in the reactor vessel shall be known by reference or established empirically; (2) expected shifts in transition temperature during design service life due to environmental conditions, such as neutron flux, shall be determined and employed in the reactor vessel design; and (3) operation margins to be observed with regard to the transition temperature shall be designated for each mode of operation.

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4. The design shall provide for material surveillance specimens which may be used to verify predicted radiation exposure and to measure the effect of radiation on the vessel material.

4.2.4 Description 4.2.4.1 Reactor Vessel The reactor vessel is a vertical, cylindrical pressure vessel with hemispherical heads of welded construction. The reactor vessel is designed and fabricated for a useful life of 40 years based upon the specified design and operating conditions. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. The vessel for each unit is designed, fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section III, 1965 edition, Summer 1965 addenda (Unit 3 vessel - Summer 1966 addenda), code cases 1332-1, 1332-2, 1332-3, 1334, 1335-2 (paragraph 4), 1336, 1339, applicable requirements for Class A vessels as defined therein, and additional GE requirements. The reactor vessel and its supports are designed as Class I equipment in accordance with the loading criteria of Appendix C. The materials used in the design and fabrication of the reactor pressure vessel are shown in Table 4.2-1. The Browns Ferry Unit 1 vessel was fabricated by B&W. The Browns Ferry Units 2 and 3 vessels were fabricated by Ishikawajima-Harima Heavy Industries Co. (IHI) in Japan, under a contract between B&W and IHI. IHI had previously manufactured the Fukushima I and II vessels. These vessels are built to the ASME Boiler and Pressure Vessel Code and GE specifications. Reactor vessel data is presented in Table 4.2-2. The cylindrical shell and bottom hemispherical head of the reactor vessel are fabricated of low alloy steel plate which is clad on the interior with weld overlay. The cylindrical shell is clad with stainless steel, and the bottom hemispherical head is clad with Inconel. The plates and forgings are ultrasonically tested and magnetic-particle-tested over 100 percent of their surfaces after forming and heat treatment. Full-penetration welds are used at all joints, including nozzles, throughout the vessel, except for nozzles of less than 3-inch nominal size and the CRD housing-to-stub tube welds. Nozzles of less than 3-inch nominal size are partial-penetration-welded as permitted by ASME Boiler and Pressure Vessel Code, Section III. The electroslag weld process was used on the Browns Ferry pressure vessels. Electroslag welding process variables, quality control procedures and technical details were presented in Appendix F, Dresden 2/3 FSAR, Docket Nos. 50-237 and 50-249. 4.2-2

BFN-27 Although little corrosion of plain carbon or low-alloy steels occurs at temperatures of 500F to 600F, higher corrosion rates occur at temperatures around 140F. The 0.125-inch minimum-thickness cladding provides the necessary corrosion resistance during reactor shutdown and also helps maintain water clarity during refueling operations. Since the vessel head is exposed to a saturated steam environment throughout its operating lifetime, stainless steel cladding is not required over its interior surfaces. Exterior, exposed ferritic surfaces of pressure-containing parts have a minimum corrosion allowance of 1/16 inch. The interior surfaces of the top head and all carbon and low-alloy steel nozzles exposed to the reactor coolant have a corrosion allowance of 1/16 inch. The vessel shape is designed to limit coolant retention pockets and crevices. The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle, rather than ductile, manner. The NDT temperature increases as a function of neutron exposure at integrated neutron exposures greater than about 1 x 1017 nvt with neutrons of energies in excess of 1 MeV. Since the material NDT temperature dictates the minimum operating temperature at which the reactor vessel can be pressurized, it is desirable to keep the NDT temperature as low as possible. One way that this is accomplished is by selecting fine-grained steels and by using advanced fabrication techniques to minimize radiation effects. The as-fabricated initial NDT temperature for all carbon and low-alloy steel used in the main closure flanges, closure bolting material, and the shell and head materials connecting to these flanges, including the connecting circumferential weld material, is limited to a maximum of 10F as determined by ASTM E208. For each main closure flange forging, a minimum of 1 tensile, 3 Charpy V-notch, and 2 drop weight test specimens have been tested from each of two locations about 180 apart on the flange. For all other carbon and low-alloy steel pressure-containing materials, including weld materials and the vessel support skirt material, the initial NDT temperature is no higher than 56F for Unit 1, and 40F for Units 2 and 3. A grain size of 5 or finer, as determined by the method in ASTM E112, is maintained. Another way of minimizing any changes (elevating) to the NDT temperature is by reducing the integrated neutron exposure at the inner surface of the reactor vessel. The coolant annulus between the vessel and core shroud and the core location in the vessel limit the integrated neutron exposure of reactor vessel material to less than 1 x 1019 nvt from neutrons with energy levels greater than 1 MeV, within the 40-year design lifetime of the vessel. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. This is not the expected exposure, nor is it the absolute limit of safe exposure; it is an exposure value that can be demonstrated to be safe and practical to maintain. The maximum calculated exposure for neutrons of 1 MeV or greater is 1.58 X 1018 nvt for Unit 1, per GEH Report No. 0000-0166-0632-R0. The maximum calculated exposure for 4.2-3

BFN-29 neutrons of 1 MeV or greater is 1.93 X 1019 nvt for Unit 2, per GEH Report No. 000N2175-R1. The maximum calculated exposure for neutrons of 1 MeV or greater is 2.23 X 1018 nvt for Unit 3, per GEH Report No. 000N2183-R0. These maximum calculated exposures encompass the Browns Ferry unit power history since initial operation as well as a conservative prediction of future Browns Ferry operation up to 3952 MWt within the MELLL domain. Operation in the MELLLA+ domain results in enhanced spectral shift during the operating cycle which results in more top-peaked axial power shape/flux shape. In addition, the MELLLA+ operating domain expansion results in a slightly higher operating neutron flux in the upper portion of the reactor core due to decreased water density. The net effect of spectral shift and water density reduction is a small increase in peak flux above the active fuel. Refinements in the calculations to support MELLLA+, documented in GEH Report No. NEDC-33877P, shows that the previous calculated exposures for Units 1, 2, and 3 from neutrons of 1 MeV or greater remain bounding. The vessel top head is secured to the reactor vessel by studs and nuts which are designed to be tightened with a stud tensioner. The vessel flanges are sealed by two concentric metallic seal-rings designed for no detectable leakage through the inner or outer seal at any operating condition, including: (a) cold hydrostatic pressure test at the hydro-pressure specified in the ASME code, and (b) heating to operating pressure and temperature at a maximum rate of 100F/hr. To detect lack of seal integrity, a 1-inch vent tap is provided in the area between the two seal-rings, and a monitor line is attached to the tap to provide an indication of leakage from the inner seal-ring seal (see Subsection 7.8). A 1-inch tap is also provided in the area outside the outer seal-ring for use in monitoring leakage. This tap is used only if the inner seal fails and is piped to an accessible place in the drywell and capped. The head and vessel flanges are low-alloy steel forgings. The sealing surfaces of the reactor vessel head and shell flanges are weld-overlay clad with Inconel 82 (ERNiCr material. The clad thickness is 0.25 inches on both the head flange and shell flange sealing surfaces. All sensitized austenitic stainless steel has been replaced on the Browns Ferry pressure vessels, except the jet pump riser brace pads on all units. These components have been clad with nonfurnace-sensitized stainless steel weld overlay. Austenitic stainless steel used in other component parts of the reactor coolant pressure boundary, including relief and safety valves, is fully annealed to preclude sensitization. Welding processes were limited to 110,000 joules per inch and the interpass temperature limited to 350F to avoid local sensitization of stainless steel. 4.2-4

BFN-27 Stainless steel with deliberate additions of nitrogen for enhancing the material strength has not been used. The vessel nozzles (Figure 4.2-2) are low-alloy steel forgings made in accordance with ASTM A508 CL2 as modified by ASME code case 1332-2, paragraph 5. Nozzles of 3-inch nominal size or larger are full-penetration welded to the vessel. Nozzles of less than 3-inch nominal size may be partial-penetration-welded as permitted by ASME Boiler and Pressure Vessel Code, Section III. Nozzles which are partial-penetration welded are nickel-chromium-iron forgings made in accordance with ASME SB166 as modified by code case 1336. The vessel top head nozzles are provided with flanges with small groove facing. The drain nozzle is of the full-penetration weld design and extends 16 inches below the bottom outside surface of the vessel. The recirculation inlet nozzles are located as shown in Figures 4.2-1, 4.2-3, and 4.2-4; feedwater inlet nozzles, core spray inlet nozzles, and the control rod drive hydraulic system return nozzle have thermal sleeves similar to those shown in the detail of Figure 4.2-2. As a result of cracks discovered in the feedwater nozzle blend and nozzle bore regions of several operating reactors, General Electric and the NRC performed an extensive study of the problem. The program, the solutions, and NRC acceptance of the modifications are fully described in NEDE 21821-A, "Boiling Water Reactor Feedwater Nozzle/Sparger - Final Report," February 1980 (proprietary version). The modifications to the BFNP feedwater nozzles included: (1) removal of the stainless-steel-clad and heat affected zone of the feedwater nozzle bore and nozzle bend radius, and (2) machining the safe end and nozzle bore and inner bend radius to accept the improved double piston ring seal, interference fit spargers with forged tee design, and orificed elbow discharges. Implementing these modifications increased the assurance of maintaining vessel integrity by minimizing the potential for crack initiation due to thermal cycling. The nozzle for the core differential pressure and standby liquid control pipe is designed with a transition so that the stainless steel outer pipe of the differential pressure and standby liquid control line (see Subsection 3.3, "Reactor Vessel Internals Mechanical Design") can be socket-welded to the inner end of the nozzle and so that the inner pipe passes through the nozzle. This design provides an annular region between the nozzle and the inner standby liquid control line to minimize thermal shock effects on the reactor vessel in the event that use of the Standby Liquid Control System is required. The jet pump instrumentation penetration seal is welded directly to the outer end of the jet pump instrumentation nozzle. The stainless steel recirculation loop piping (see Subsection 4.3, "Reactor Recirculation System") is welded to the outer end of the recirculation outlet nozzle. The main steam line piping is welded to the outer 4.2-5

BFN-27 end of the steam outlet nozzle. The piping attached to the vessel nozzle is designed, installed, and tested in accordance with the requirements of USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. Thermocouple pads are located on the exterior of the vessel (see Table 4.2-3). At each thermocouple location, two 3/4-inch-diameter pads are provided: an end pad to hold the end of a 3/16-inch-diameter thermocouple and a clamp pad equipped with a set screw to secure the thermocouple. The reactor vessel is laterally and vertically supported and braced to make it as rigid as possible without impairing the movements required for thermal expansion. Where thermal requirements prohibit the use of rigid supports, spring anchors or hydraulic snubbers are employed to resist earthquake forces, while allowing sufficient flexibility for thermal expansion. 4.2.4.2 Shroud Support The reactor vessel shroud support is a radial, cylindrical shell that surrounds the reactor core assembly and is designed so that stresses due to reactions at the shroud support are within ASME code, Section III, requirements for normal, upset, emergency, and faulted loading conditions. The design of the shroud support also takes into account the restraining effect of the components attached to the support, their weight, and earthquake loadings. The vessel shroud support and other internal attachments (jet pump riser support pads, diffuser brackets, guide rod brackets, steam dryer support brackets, dryer holddown brackets, feedwater sparger brackets, and core spray brackets) are as shown in Figures 4.2-1, 4.2-3, and 4.2-4. 4.2.4.3 Reactor Vessel Support Assembly The reactor vessel support assembly consists of a ring girder, sole plates, and the various bolts, shims, and set screws necessary to position and secure the assembly between the reactor vessel support skirt and the support pedestal. The concrete and steel support pedestal is constructed integrally with the building foundation. Steel anchor bolts are set in the concrete with the threads extending above the surface. The sole plates are set flat and level on the concrete, and the lower flange of the ring girder is set on top of the sole plates and shimmed as necessary to level the ring girder. The anchor bolts extend through both the sole plates and the ring girder bottom flange. High strength bolts are used to bolt the flange of the reactor vessel support skirt to the top flange of the ring girder. The ring girder and sole plates are fabricated of ASTM A36 structural steel according to AISC specifications. 4.2-6

BFN-27 4.2.4.4 Vessel Stabilizers The vessel stabilizers are connected between the reactor vessel and the top of the shield wall surrounding the vessel to provide lateral stability for the upper part of the vessel. Eight stabilizer brackets are attached by full-penetration welds to the reactor vessel at evenly spaced locations around the vessel below the flange. Each vessel stabilizer consists of a stabilizer rod, threaded at the ends, springs, washers, nut, a plate, and a bumper bracket with tapered shims. The stabilizers are attached to each bracket and apply tension in opposite directions. The stabilizers are evenly preloaded with tensioners to the values of the residual loads. The stabilizers are designed to permit radial and axial vessel expansion, to limit horizontal vibration, and to resist seismic and jet reaction forces. 4.2.4.5 Refueling Bellows The refueling bellows form a seal between the reactor vessel and the surrounding primary containment drywell to permit flooding of the space (reactor well) above the vessel during refueling operations. The refueling bellows assembly (see Figures 4.2-1, 4.2-3, and 4.2-4) consists of a Type 304 stainless steel bellows, a backing plate, a spring seal, and a removable guard ring. The backing plate surrounds the outer circumference of the bellows to protect it and is equipped with a tap for testing and for monitoring leakage. The self-energizing spring seal is located in the area between the bellows and the backing plate and is designed to limit water loss in the event of a bellows rupture by yielding to make a tight fit to the backing plate when subjected to full hydrostatic pressure. The guard ring attaches to the assembly and protects the inner circumference of the bellows. The guard ring can be removed from above to inspect the bellows. The assembly is welded to the reactor bellows support skirt and the reactor well seal bulkhead plate. The reactor bellows support skirt is welded to the reactor vessel shell flange, and the reactor well seal bulkhead plate bridges the distance to the primary containment drywell wall. Six watertight, hinged covers are bolted in place for normal refueling operation. For normal operation, these covers are opened and removable air supply ducts and air return ducts permit circulation of ventilation air in the region above the reactor well seal. 4.2.4.6 Control Rod Drive Housings The control rod drive housings are inserted through the control rod drive penetrations in the reactor vessel bottom head and are welded to the stub tubes extending into the reactor vessel1 (Figure 4.2-2). 1 Kobsa, I. R., and Wetzerl, V. R., "Design and Analysis of Control Rod Drive Reactor Vessel Penetrations," General Electric Co., Atomic Power Equipment Department, November 1968 (APED-5703). 4.2-7

BFN-27 Each housing transmits a number of loads to the bottom head of the reactor. These loads include the weight of a control rod and control rod drive, which are bolted to the housing from below (see Subsection 3.4, "Reactivity Control Mechanical Design"), the weight of a control rod guide tube, one four-lobed fuel support piece, and the four fuel assemblies which rest on the top of the fuel support piece (see Subsection 3.3, "Reactor Vessel Internal Mechanical Design"). The housings are fabricated of Type 304 austenitic stainless steel. 4.2.4.7 Control Rod Drive Housing Supports The control rod drive housing support is designed to prevent a nuclear transient in the unlikely event that there is a control rod drive housing failure. This device consists of a grid structure located below the reactor vessel from which housing supports are suspended. The supports allow only slight movement of the control rod drive or housing in the event of failure. The control rod drive housing support is described in detail in Subsection 3.5, "Control Rod Drive Housing Supports." 4.2.4.8 In-Core Neutron Flux Monitor Housing The in-core neutron flux monitor housings are inserted up through the in-core penetrations in the bottom head of the reactor vessel and are welded to the inner surface of the bottom head (Figure 4.2-2). An in-core flux monitor guide tube is welded to the top of each housing (see Subsection 3.3, "Reactor Vessel Internals Mechanical Design"), and either a source range monitor/intermediate range monitor (SRM/IRM) drive unit or a local power range monitor (LPRM) is bolted to the seal-ring flange at the bottom of the housing (see Subsection 7.5, "Neutron Monitoring System"). 4.2.4.9 Reactor Vessel Insulation The reactor vessel insulation is an all-metal, reflective insulation having an average maximum heat transfer rate of approximately 80 Btu/hr-ft2 at the operating conditions of 550F for the vessel and 135F for the outside air. The maximum insulation thickness ranges from 4 inches for the upper head to 3-1/2 inches for the cylindrical shell and nozzles and 3 inches for the bottom head. The insulation is designed to permit complete submersion in water without loss of insulating material, contamination from the water, or adverse effect on the insulation efficiency of the insulation assembly after draining and drying. The lower head and cylindrical shell insulation is permanently installed for the 60 year operating life of the vessel. The insulation panels for the cylindrical shell of the vessel are held in place by vessel insulation supports located at two elevations on the vessel. The support brackets for each support are full-penetration-welded to the vessel at 12 evenly spaced locations 4.2-8

BFN-27 around the circumference. Provisions are made for removing insulation during inservice inspection. 4.2.4.10 Other Reactor Coolant Pressure Boundary Ferritic Components The fracture or notch-toughness properties and the operating temperature of ferritic materials are controlled to ensure adequate toughness when the system is pressurized to more than 20 percent of the design pressure. Such assurance is provided by maintaining the lowest service metal temperature, when the system pressure exceeds 20 percent of design pressure, at least 60F above the nil-ductility transition temperature (NDTT). The lowest service-metal temperature is the lowest temperature which the metal will experience in service while the plant is in operation. It is established by appropriate calculations considering atmosphere ambient temperatures, the insulation or enclosure provided, and the minimum temperature maintained. Further interpretations and requirements are as follows: A. Charpy V-notch (American Society for Testing and Material Standard A370 Type A) or drop weight (per ASTM E208) tests have been performed to demonstrate that all materials and weld metal meet brittle fracture requirements at test temperature. Test specimens, for the surveillance capsule pulled in 1994, were prepared and tested with minimum impact energy requirements in accordance with Table N-421 and the general provisions of N-313, N-331, N-332, and N-511 of Section III of the ASME Boiler and Pressure Vessel Code. For the surveillance capsule pulled in 2011, per BWRVIP-271/NP, the Charpy impact tests were conducted in accordance with ASTM Standards E185-82 and E23-02. Prior to the Summer 1972 Addenda of the 1971 ASME Section III Boiler and Pressure Vessel Code, impact testing was not required on materials with a nominal section thickness of 1/2 inch or less. However, this 1/2 inch thickness exclusion was increased to 5/8 inch by the ASME Boiler and Pressure Vessel Code, Section III, 1971 Edition, Summer 1972 Addenda. Therefore, after issuance of the Summer 1972 Addenda, impact testing is not required on materials with a nominal section thickness of 5/8 inch or less. The welding procedures used were qualified by impact testing of weld metal and heat affected zone to the same requirements as the base metal in accordance with N-541. B. Impact tests were not required for the following:

1. Bolting, including nuts, 1-inch nominal diameter or less,

2. Bars with a nominal cross-sectional area not exceeding 1 square inch,

3. Materials with a nominal (section) wall thickness of less than 1/2 inch or 5/8 inch (refer to paragraph 4.2.4.10.A),

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4. Components including pumps, valves, piping, and fittings with a nominal inlet pipe size of 6-inch-diameter and less, regardless of thickness, and

5. Consumable insert material, austenitic stainless steel, and nonferrous materials.

C. Impact testing was not required on components or equipment pressure parts having a minimum service temperature of 250F or more when pressured over 20 percent of the design pressure. Example: Steam line is excluded from brittle fracture test requirement since the steam temperature will be over 250F when the steam line pressure is at the 20 percent design pressure. D. Impact testing was not required on components or equipment pressure parts whose rupture could not result in a loss of coolant exceeding the capability of normal makeup systems to maintain adequate core cooling for the duration of a reactor shutdown and orderly cooldown. E. These criteria apply to components and equipment pressure parts, including flange bolts of the reactor coolant pressure boundary, and do not apply to related components such as anchors, anchor bolts, hangers, suppressors, and restraints. All components for the Browns Ferry plant were designed and fabricated giving consideration to brittle-fracture control requirements as stated above. However, these specific conditions were not a part of the initial Browns Ferry Units 1 and 2 plant requirements, and due to the status of fabrication on two items, the requirements could not be imposed without scrapping all materials. On Browns Ferry Units 1 and 2 these two items are: (1) feedwater piping through the second containment isolation valve, and (2) the 14-inch HPCI testable check valve (HPCI pump return into feedwater pipe outside the containment). Charpy V-notch impact tests were performed on these items where possible, and results indicate they generally would not meet the conditions under A, above, if they had been imposed. 4.2.5 Safety Evaluation The reactor vessel design pressure of 1250 psig is determined by an analysis of margins required to provide a reasonable range for maneuvering during operation, with additional allowances to accommodate transients above the operating pressure without causing operation of the safety valves. The design temperature for the reactor vessel (575F) is based on the saturation temperature of water corresponding to the design pressure. 4.2-10

BFN-27 To withstand external and internal loadings while maintaining a high degree of corrosion resistance, a high-strength, carbon-alloy steel is used as the base metal with an internal cladding applied by weld overlay to the cylindrical shell and bottom head. Use of the ASME Boiler and Pressure Vessel Code, Section III, Class A, pressure vessel code design criteria provides assurance that a vessel designed, built, and operated within its design limits has an extremely low probability of failure due to any known failure mechanism. The reactor vessel is designed for a 40-year life and will not be exposed to more than 1 X 1019 nvt of neutrons with energies exceeding 1 MeV. The reactor vessel is also designed for the transients which could occur during the 40-year life as indicated below. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. 4.2-11

BFN-27 No. of Type of Cycle Cycles Boltup 123 Design hydrostatic test at 1250 psig 130 Startup (100F/hr heatup rate) 120 Daily reduction to 75 percent power 10,000 Weekly reduction to 50 percent power 2,000 Control rod worth test 400 Loss of feedwater heaters (80 cycles total) Turbine trip at 25 percent power 10 Feedwater heater bypass 70 Scram (200 cycles total) Loss of feedwater pumps, isolation valves close 10 Turbine trip, feedwater on, isolation valves stay open 40 Reactor overpressure with delayed scram, feedwater stays on, isolation valves stay open 1 Single safety relief valve blowdown 2 All other scrams 147 Improper start of cold recirculation loop 5 Sudden start of pump in cold recirculation loop 5 Shutdown (100F/hr cooldown rate) 118 Hydrostatic test at 1563 psig 3 Unbolt 123 Stress analysis and load combinations for the reactor vessel are evaluated for the cycles listed above, with the conclusion that ASME code limits are satisfied. The details of assumed loading combinations are described in Appendix C for Class 1 equipment. It is possible that the specified number of cycles for some of the events listed above may be exceeded over the life of the plant. A plant procedure has been implemented at Browns Ferry to maintain surveillance on the number of cycles which have occurred and the resulting fatigue usage factors. When the fatigue usage factor reaches a value of 0.7, the procedure requires a reevaluation to be completed in a timely manner to assure that the allowable fatigue usage factor of 1.0 is not exceeded. Operating limits on pressure and temperature during inservice hydrostatic testing were established using as a guide Appendix G to the ASME Boiler and Pressure Vessel Code, Section III, 1971, which was first added to the code in the summer 1972 addenda. The intent of Appendix G is to set criteria based on fracture toughness to provide a margin of safety against a nonductile failure. The resulting operating limits ensure that a large postulated surface flaw, having a depth of one-quarter of the material thickness and a length of one and one-half of the material thickness, can be safely accommodated in regions of the reactor vessel shell remote from discontinuities. Operating limits on temperature and pressure 4.2-12

BFN-28 when the core is critical were established by using 10 CFR 50, Appendix G, "Fracture Toughness Requirements," paragraph IV.A.2.C. The 1998 Edition of the ASME Section XI Boiler and Pressure Vessel Code including 2000 Addenda was used in the development of the Unit 1 P-T curves. The P-T curve methodology includes the following: 1) the use of K1C from Figure 4200-1 of Appendix A to Section XI and 2) the use of the Mm calculation in the ASME Code paragraph G.2214 of Appendix G to Section XI for a postulated defect normal to the direction of maximum stress. An exemption from specific requirements of 10 CFR Part 50, Appendix G is taken by use of ASME Code Case N-640 for Unit 2 and Unit 3. ASME Code Case N-640 permits the use of an alternative reference fracture curve K1c for RPV materials for use in determining the PT limits. The PT limit curves based on the K1c fracture toughness curve enhance overall plant safety by minimizing challenges to operators since requirements for maintaining a high vessel temperature during pressure testing are lessened. ASME Code Case N-588 methodology was also used as a basis for the PT curves. This code case permits the use of an alternative procedure for calculating applied stress intensity factors during normal operation and pressure test conditions due to pressure and thermal gradients for axial flaws. This methodology is incorporated into the ASME, Section XI Code, 1995 Edition, 1996 Addenda, which is the current code of record for the Unit 2 inservice inspection program. Since Unit 3 uses an earlier code of record for the inservice inspection program, Unit 3 implements the requirements of only the 1995 Edition, 1996 Addenda of ASME Section XI, Appendix G to allow the use of the ASME Code Case N-588 methodology for PT curves. The operating limits are provided in the technical specifications for Browns Ferry. For the purpose of setting these operating limits, the initial RTNDT (nil-ductility reference temperature) was determined from the impact test data taken in accordance with the requirements of the code to which the reactor vessels were designed and manufactured. The maximum NDT temperature allowed by the vessel specifications was 40F. Although test data on beltline base material show lower NDT temperatures, an assumed RTNDT of 40F was used in the vessel beltline area, as well as the areas remote from the beltline because the generally accepted NDT temperature for electroslag welds used in the beltline longitudinal seams is 40F. The current operating limits on the pressure/temperature (P/T) curves in the technical specifications are based on the following (RTNDT) values. Unit 1 has used 23.1F for the (RTNDT) value, Unit 2 has used 23.1F for the (RTNDT) value, and Unit 3 has used 23.1F for the (RTNDT) value. For the current P/T curves, fluences were conservatively calculated for licensed operating periods of 38 EFPY for Unit 1, 48 EFPY for Unit 2, and 54 for Unit 3. These periods reflect 60-year reactor pressure vessel operating life and a conservative period of plant operation at 3952 MWt power level. The higher fluence was used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G in accordance with Regulatory Guide 1.99, Revision 2. The end-of-life shelf 4.2-13

BFN-27 energy was evaluated by an equivalent margin analysis (EMA). The results of these evaluations indicated that: (a) The results of the upper shelf energy EMA for limiting welds and plates for the three vessels remain less than the acceptance criterion in all cases. (b) The effective full power year (EFPY) shift is slightly increased and, consequently requires a change in the adjusted reference temperature (ART), which is the initial RTNDT plus the shift. The beltline material ART will remain within the 200 degree screening criterion. In addition to the minimum requirements of the ASME Boiler and Pressure Vessel Code, the following precautions were taken and tests made either to ensure that the initial NDT temperature of the reactor vessel material is low or to reduce the sensitivity of the material to irradiation effects.

a. The material was selected and fabrication procedures were controlled to produce as fine a grain size as practical. It is an objective in fabrication to maintain a grain size of 5 or finer.

b. Drop weight impact tests were performed on each heat and heat treatment charge of all low-alloy steel-plate material in its "as-fabricated" condition.

c. Drop weight impact tests were made on the weld metal, the heat-affected zone of the base metal, and the base metal of the weld test plates simulating seams. If different welding procedures were used for nozzle welds, drop weight tests of similarly prepared coupons were made. The NDT temperature test criteria for the weld and heat-affected zone of the base material are the same as for the unaffected base metal.

d. The actual NDT temperature of the plates opposite the center of the reactor core was determined. In other areas it was sufficient to demonstrate that the two drop weight test specimens did not break at 10F above the design NDT temperatures. The area of the vessel located opposite the core was fabricated entirely of plate and was not penetrated by nozzles, nor were there any other structural discontinuities in this area which would act as stress risers.

The reactor assembly is designed such that the average annular distance from the outermost fuel assemblies to the inner surface of the reactor vessel is approximately 80 centimeters. This annular volume, which contains the core shroud, the jet pump assemblies, and reactor coolant, serves to attenuate the fast neutron flux incident upon the reactor vessel wall. Using assumptions of plant operation at 3440 Mw(t), 100 percent plant availability, and 40-year plant life, the neutron fluence at the inner 4.2-14

BFN-27 urface of the vessel was calculated to be 3.8 X 1017 nvt for neutrons having energies greater than 1 MeV. The results of the analyses of the vessel wall neutron dosimeters which were removed from the Browns Ferry reactor vessels at the end of the first core cycle indicated that the neutron fluence at the inner surfaces of the vessels at the end of 40-year plant life would be 1.56 x 1018, 1.34 x 1018, and 1.31 x 1018 nvt for Units 1, 2, and 3, respectively. These results ranged from 3-1/2 to 4 times the calculated fluence of 3.8 x 1017 nvt. Thus, additional analyses were required to predict the shifts in RTNDT based on fluence obtained from the analyses of the vessel wall neutron dosimeters. The procedures in Regulatory Guide 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Revision 1, April 1977 were used to predict the RTNDT shifts. Response to Generic Letter 92-01 provides updated fluence data. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. Quality control methods were used during the fabrication and assembly of the reactor vessel and appurtenances to ensure that the design specifications were met. The fabrication test program was carried out by the reactor vessel vendors on material representative of the formed, heat-treated, and fully fabricated vessel. Tests of base metal and welded joint were performed, and the results were reported during the early stages of vessel construction. Tensile specimens (of 0.505 inch in diameter) from the shell plate material were prepared for various thickness levels of the plate material. These specimens were tested at various temperatures per ASTM Specifications E8 and E21 to determine tensile strength, yield strength, elongation, and reduction of area. Tensile specimens whose gauge diameter is at least 80 percent of the reactor vessel wall thickness were prepared from base metal and weld material. These specimens were tested at room temperature per ASTM Specification E8 to provide stress-strain curves, tensile strength, yield strength, elongation, reduction of area, and macrophotographs of the breaks. Charpy V-notch impact specimens were prepared from base metal and tested per ASTM Specification E23, Type A, to establish curves for determining the transition temperature at which 30 ft-lb of absorbed energy result in ductile fracture for various thickness levels of the plate material. The Reactor Coolant System was cleaned and flushed before fuel was loaded initially. During the preoperational test program, the reactor vessel and Reactor Coolant System were given a hydrostatic test in accordance with code requirements at 125 percent of design pressure. The vessel temperature is maintained at a minimum of 60F above the NDT temperature prior to pressurizing the vessel for a hydrostatic test. A hydrostatic test at a pressure not to exceed system operating pressure is made following each removal and replacement of the reactor vessel head. Other preoperational tests included calibrating and testing the reactor vessel 4.2-15

BFN-27 flange seal-ring leakage detection instrumentation, adjusting reactor vessel stabilizers, checking all vessel thermocouples, and checking the operation of the vessel flange stud tensioner. During the startup test program, the reactor vessel temperatures were monitored during vessel heatup and cooldown to assure that thermal stress on the reactor vessel was not excessive during startup and/or shutdown. The average rate of reactor coolant temperature change during normal heatup and cooldown is limited to 100F in any 1-hour period. Only during some postulated events, or in local areas, would this rate of fluid temperature change be exceeded as a result of rapid blowdown, valve operation, or rupture accident. 4.2.6 Inspection and Testing The inservice inspection and testing program for the reactor vessel and appurtenances is outlined and detailed in Subsection 4.12. Extent and areas of examination, inspection methods, and frequency of examination are established therein. The surveillance test program provides for the preparation of a series of Charpy V-notch impact specimens and tensile specimens from the base metal of the reactor vessel, weld heat-affected zone metal, and weld metal from a reactor steel joint which simulates a welded joint in the reactor vessel. The reactor vessel material surveillance program is described in report NEDO-10115, Mechanical Property Surveillance of General Electric BWR Vessels, by J. P. Higgins and F. A. Brant. It describes the specimens, specimen inventory, capsule design, associated equipment, material selection and instructions for handling the specimens. All the requirements of paragraphs 3.1 through 3.3 of ASTM E-185-66 are met. All the requirements of paragraphs 4, 5, 6, 7, and 8 of ASTM E-185-66 are met, except that thermal control specimens discussed in paragraph 4.3 are not used. NEDO-10115, paragraph 5.7 states, "Because the BWR is a constant-temperature device, no special temperature monitoring devices are required." It is felt paragraph 4.3 of E-185-66 is a recommendation rather than a requirement. The vessel surveillance samples were prepared in accordance with GE purchase specification 21A1111, Rev. No. 9, Attachment B. 4.2-16

BFN-27 The NDT temperatures for the three core region plates were as follows. Heat No. Plate No. NDTT (F) C2884-2 6-139-19 0 C2868-2 6-139-20 0 C2753-1 6-139-21 -20 The two test plates furnished by Babcock & Wilcox under the requirements of paragraph 3.1.1 of attachment B to specification 21A1111 were fabricated from Heat No. C2884-2 and C2868-2. The two plates were electroslag-welded (B&W Weld Procedure WR-12-4) and heat-treated the same as the core region plates. Tensile and Charpy impact specimen samples were removed as indicated in Figures 3, 4, 5, 6, and 7 of attachment B to 21A1111. (See FSAR Appendices J, K, and L.) The surveillance test plate 610-0127 was 139 in. long and 60 in. wide, and all excess material is under TVA control in the event that additional material is needed. It is estimated that enough extra material is available for several hundred additional Charpy specimens. No weak direction specimens were included in the reactor vessel material surveillance program. All Charpy V-notch specimens were taken parallel to the direction of rolling. The majority of developmental work on radiation effects has been with longitudinal specimens. This is considered the best specimen to be used for determination of changes in transition temperature. At the low neutron fluence levels of BWR plants, no change in transverse shelf level is expected and transition temperature changes are minimal. The specimens and neutron monitor wires were placed near core midheight adjacent to the reactor vessel wall where the neutron exposure is similar to that of the vessel wall (see Subsection 3.3). The specimens were installed at startup or just prior to full-power operation. For Units 1, 2, and 3, Integrated Surveillance Program (ISP) implementation and surveillance specimen schedule withdrawal and testing is governed and controlled by BWRVIP-86 Revision 1-A, the BWRVIP responses to NRC RAIs dated May 30, 2001, December 22, 2001, and January 11, 2005, and the NRCs Safety Evaluation dated February 1, 2002. (NOTE: WRVIP-86, Revision 1-A, was approved by NRC and issues in October 2012, superseding both BWRVIP-86-A and BWRVIP-116.) Surveillance and chemistry data for all representative materials in the BWRVIP ISP have been consolidated into BWRVIP-135 {Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations.} A test specimen surveillance capsule (the second set of Unit 2 test specimens located at Azimuth 120º) was withdrawn in accordance with the ISP in 2011 during Unit 2 Refueling Outage 16 (U2R16) at approximately 23 EFPY of operation. An additional test specimen surveillance capsule is scheduled for withdrawal during the license 4.2-17

BFN-28 renewal period, this being the third set of Unit 2 test specimens located at Azimuth 300º, which are currently scheduled for removal in the refueling outage closest to without exceeding 40 EFPY of operation. At the present time, this would correspond to Unit 2 Refueling Outage 24 (U2R24) in 2027. Presently, there are no plans to withdraw any capsules from either Unit 1 or Unit 3, as per BWRVIP-135, the BFN Unit 2 capsules provide the best representative plate material for all three units and the best representative weld material for Units 2 and 3. Supplemental Surveillance Program (SSP) Capsules A, B, D, G, E, and I provide the best representative weld material for Unit 1. Test results will provide the necessary data to monitor embrittlement for Units 1, 2, and 3. Since the predicted transition temperature shift of the reactor vessel beltline steel is less than 100F at end-of-life, the use of the capsules per the ISP meets the requirements of 10 CFR 50, Appendix H, and ASTM E185-82. Revisions to fluence calculations using data obtained from the surveillance capsule specimens will use an NRC approved methodology that meets Regulatory Guide 1.190. {By letter dated August 14, 2008 (EDMS Number L44 080828 014), NRC issued License Amendment 273 for BFN Unit 1, and by letter dated January 28, 2003 (EDMS Number L44 030204 001), NRC issued License Amendment Numbers 279 and 238, for BFN Units 2 and 3 respectively, authorizing adoption of the BWRVIP Integrated Surveillance Program to address the requirements of Appendix H to 10 CFR Part 50.] 4.2-18

BFN-26 TABLE 4.2-1 REACTOR PRESSURE VESSEL MATERIALS Component Form Material *Spec. (ASTM/ASME) Heads, Shell ...................... rolled plate low-allow steel SA-302 B cc 1339 Closure Flange .................. forged rings low-alloy steel A-508 CL 2 cc 1332-2 Cladding ............................ weld overlay austenitic SA-371 type ER309-type stainless steel ER308 (and carbon content

                                                         -inconel            <0.08 w/o)-Inconel 82 and 182 Nozzles.............................. forged shapes       low-alloy steel     A-508-CL2 cc 1332-2 Control Rod Drive .............. forged tubes             Inconel             SB-166 cc 1336 Stub Tubes Control Rod Drive .............. pipe                     austenitic          --

Housings stainless steel In-Core Housings .............. pipe austenitic -- stainless steel Vessel Supports- ............... rolled plate low-alloy steel SA-302 Gr.B External Shroud Support- ................ forging Inconel SB-168 Annealed cc 1336 Internal Nozzle Safe Thermal ......... pipe austenitic SA-312 TP.304 Sleeves stainless steel Nozzle Safe Ends .............. forging austenitic SA-336-F8/F8M stainless steel and some low-carbon SA-105-2 cc 1332-1 steel Nozzle for Instrument ........ forging Inconel SB-166 cc 1336 para. 1 Penetrations

• cc - Code Case that modifies/augments the material specification.

BFN-26 Table 4.2-2 Table 4.2-3 REACTOR VESSEL DATA REACTOR VESSEL ATTACHMENTS Reactor Vessel Qty. Inside Diameter, in. (min.) ..............................251 3/8 in. Internal Attachments Inside Length ............................................ 73 ft 11-1/2 in. Guide Rod Bracket ......................................... 2 Design Pressure and Temperature, Steam Dryer Support Bracket ......................... 4 psig @ °F ................................................... 1250 @ 575 Dryer Holddown Bracket ................................. 4 Feedwater Sparger Bracket .......................... 12 Vessel Nozzles (number and size) Jet Pump Riser Support Pads....................... 20 Recirculation Outlet ................................2-36 in. to 28 in. Jet Pump Diffuser Bracket ............................ 20 Steam Outlet ....................................................... 4-26 in. Core Spray Bracket ........................................ 4 Recirculation Inlet ............................................. 10-12 in. Feedwater Inlet................................................... 6-12 in. External Attachments Core Spray Inlet ................................................. 2-10 in. Stabilizer Bracket ............................................ 8 Instrument (one of these is Head Spray)** ......... 2- 6 in. Top Head Lifting Lug....................................... 4 Control Rod Drive ............................................. 185- 6 in. Insulation Supports ......................................... 2 Jet Pump Instrumentation..................................... 2- 4 in. Insulation Support Brackets ...... 12 ea; 2 places Vent ...................................................................... 1- 4 in. Thermocouple Pad ....................................... 36 Instrumentation ..................................................... 6- 2 in. Control Rod Drive Hydraulic System Return * ...... 1- 4 in. Core Differential Pressure and Liquid Control ...... 1- 2 in. Drain ..................................................................... 1- 2 in. In-Core Flux Instrumentation .............................. 55- 2 in. Head Seal Leak Detection .................................... 2- 1 in. Approximate Weights (in pounds) Bottom Head ...................................................... 207,500 Vessel Shell........................................................ 842,000 Vessel Flange..................................................... 106,000 Support Skirt......................................................... 28,000 Other Vessel Components ................................... 65,500 Total Vessel without Top Head........................ 1,249,000 Top Head1 .......................................................... 252,000 Total Vessel..................................................... 1,501,000 1 This weight includes 60,095 lbs. which is the weight of the reactor pressure vessel studs, nuts and washers.

• CRD Return Line nozzle is capped.
  • Unit 1, 2, and 3 head spray nozzle line is capped.

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