Text

Enclosure 1: Sequoyah Nuclear Plant, Units 1 and 2, Redacted Updated Final Safety Analysis Report - Amendment 31 (Public Use Only)
	ML23349A014
Person / Time
Site:	Sequoyah
Issue date:	10/12/2023
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
	Wentzel M
Shared Package
ML23285A087	List: ML23285A088; ML23349A014;
References
	Download: ML23349A014 (1)
	v • d • e

{{#Wiki_filter:Sequoyah Nuclear Plant Updated Final Safety Analysis Report Amendment 31 TENNESSEE VALLEY AUTHORITY Redacted Version

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 1 26 2 24 3 27 4 18 CHAPTER 1 1-1 19 1-2 17 1-3 17 1.1-1 31 1.1-2 31 1.2-1 13 1.2-2 17 1.2-3 24 1.2-4 31 1.2-5 25 1.2-6 25 1.2-7 21 1.2-8 13 1.2-9 16 1.2-10 13 1.2-11 19 Figure 1.2.3-1 13 Figure 1.2.3-2 18 Figure 1.2.3-3 13 Figure 1.2.3-4 13 Figure 1.2.3-5 13 Figure 1.2.3-6 13 Figure 1.2.3-7 13 Figure 1.2.3-8 13 Figure 1.2.3-9 13 Figure 1.2.3-10 13 Figure 1.2.3-11 18 Figure 1.2.3-12 24 Figure 1.2.3-13 18 Figure 1.2.3-14 13 Figure 1.2.3-15 13 Figure 1.2.3-16 13 Figure 1.2.3-17 13 Figure 1.2.3-18 13 Figure 1.2.3-19 18 1.3-1 27 Table 1.3.1-1 (Sheet 1) 13 Table 1.3.1-1 (Sheet 2) 13 Table 1.3.1-1 (Sheet 3) 25 Table 1.3.1-1 (Sheet 4) 13 Table 1.3.1-1 (Sheet 5) 13 Note: The Effective Amendment designation indicates the affected page was either changed as a result of the Amendment or the page number was altered as a result of the Amendment. EPL 1

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 1.3.1-2 (Sheet 1) 24 Table 1.3.1-2 (Sheet 2) 24 Table 1.3.1-2 (Sheet 3) 13 Table 1.3.2-1 (Sheet 1) 24 Table 1.3.2-1 (Sheet 2) 13 Table 1.3.2-1 (Sheet 3) 13 Table 1.3.2-1 (Sheet 4) 17 Table 1.3.2-1 (Sheet 5) 13 Table 1.3.2-1 (Sheet 6) 13 Table 1.3.2-1 (Sheet 7) 13 Table 1.3.2-1 (Sheet 8) 13 Table 1.3.2-1 (Sheet 9) 18 1.4-1 24 1.5-1 13 1.5-2 13 1.5-3 13 1.5-4 13 1.5-5 31 1.5-6 13 1.5-7 13 1.5-8 13 1.5-9 13 1.5-10 13 1.5-11 13 1.5-12 13 1.5-13 13 1.5-14 13 1.6-1 17 Table 1.6.1-1 (Sheet 1) 13 Table 1.6.1-1 (Sheet 2) 13 Table 1.6.1-1 (Sheet 3) 13 Table 1.6.1-1 (Sheet 4) 31 Table 1.6.1-1 (Sheet 5) 31 Table 1.6.1-1 (Sheet 6) 31 Table 1.6.1-1 (Sheet 7) 13 Table 1.6.1-1 (Sheet 8) 13 Table 1.6.1-1 (Sheet 9) 13 Table 1.6.1-1 (Sheet 10) 13 Table 1.6.1-1 (Sheet 11) 31 Table 1.6.1-1 (Sheet 12) 24 Table 1.6.1-1 (Sheet 13) 24 1.7-1 17 1.7-2 13 1.7-3 13 1.7-4 28 1.7-5 26 1.7-6 27 1.7-7 24 1.7-8 24 1.7-9 24 1.7-10 13 EPL 2

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 1.7-11 13 1.8-1 13 CHAPTER 2 2-1 13 2-2 13 2-3 24 2-4 24 2-5 13 2-6 13 2-7 13 2-8 13 2-9 21 2-10 13 2-11 13 2-12 17 2-13 13 2-14 13 2-15 13 2-16 17 2-17 13 2-18 17 2-19 17 2-20 13 2-21 13 2.1-1 20 2.1-2 13 2.1-3 13 2.1-4 13 2.1-5 13 2.1-6 19 Table 2.1.1-1 1 Table 2.1.3-1 13 Table 2.1.3-2 13 Table 2.1.3-3 13 Table 2.1.3-4 13 Table 2.1.3-5 13 Table 2.1.3-6 13 Table 2.1.3-6a 13 Table 2.1.3-7 13 Table 2.1.3-8 13 Table 2.1.3-9 13 Table 2.1.3-10 13 Table 2.1.3-11 13 Table 2.1.3-12 13 Table 2.1.3-12a 13 Table 2.1.3-13 13 Table 2.1.3-14 13 Table 2.1.3-15 13 Table 2.1.3-16 13 Table 2.1.3-17 13 Table 2.1.3-18 13 EPL 3

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 2.1.3-19 13 Table 2.1.3-20 21 Table 2.1.4-1 13 Table 2.1.4-2 13 Figure 2.1.1-1 13 Figure 2.1.1-2 13 Figure 2.1.1-3 Original Figure 2.1.2-1 26 Figure 2.1.2-2 20 Figure 2.1.3-1 Original Figure 2.1.4-1 Original Figure 2.1.4-2 Original 2.2-1 13 2.2-2 13 2.2-3 13 2.2-4 16 2.2-5 13 2.2-6 13 2.2-7 16 2.2-8 13 2.2-9 26 Table 2.2.2-1 13 Table 2.2.2-1a 13 Table 2.2.2-1b 13 Table 2.2.3-1 (Sheet 1) 13 Table 2.2.3-1 (Sheet 2) 13 Table 2.2.3-2 (Sheet 1) 13 Table 2.2.3-2 (Sheet 2) 13 Table 2.2.3-3 13 2.3-1 13 2.3-2 18 2.3-3 21 2.3-4 18 2.3-5 18 2.3-6 18 2.3-7 15 2.3-8 24 2.3-9 23 2.3-10 24 2.3-11 23 2.3-12 28 2.3-13 20 2.3-14 13 2.3-15 13 2.3-16 21 2.3-17 21 2.3-18 23 2.3-19 23 Table 2.3.2-1 13 Table 2.3.2-2 13 Table 2.3.2-3 13 EPL 4

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 2.3.2-4 13 Table 2.3.2-5 13 Table 2.3.2-6 13 Table 2.3.2-7 13 Table 2.3.2-8 13 Table 2.3.2-9 13 Table 2.3.2-10 13 Table 2.3.2-11 13 Table 2.3.2-12 13 Table 2.3.2-13 13 Table 2.3.2-14 (Sheet 1) 13 Table 2.3.2-14 (Sheet 2) 13 Table 2.3.2-15 13 Table 2.3.2-16 18 Table 2.3.2-17 13 Table 2.3.2-18 13 Table 2.3.2-19 13 Table 2.3.2-20 18 Table 2.3.2-21 13 Table 2.3.2-22 13 Table 2.3.2-23 13 Table 2.3.2-24 13 Table 2.3.2-25 13 Table 2.3.2-26 13 Table 2.3.2-27 13 Table 2.3.2-28 13 Table 2.3.2-29 13 Table 2.3.2-30 13 Table 2.3.2-31 13 Table 2.3.2-32 13 Table 2.3.2-33 13 Table 2.3.2-34 13 Table 2.3.2-35 13 Table 2.3.2-36 13 Table 2.3.2-37 13 Table 2.3.2-38 13 Table 2.3.2-39 13 Table 2.3.2-40 13 Table 2.3.2-41 13 Table 2.3.2-42 13 Table 2.3.2-43 13 Table 2.3.2-44 13 Table 2.3.2-45 13 Table 2.3.2-46 13 Table 2.3.2-47 13 Table 2.3.2-48 13 Table 2.3.2-49 13 Table 2.3.2-50 13 Table 2.3.2-51 13 Table 2.3.2-52 13 Table 2.3.2-53 13 EPL 5

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 2.3.2-54 13 Table 2.3.2-55 13 Table 2.3.2-56 13 Table 2.3.2-57 13 Table 2.3.2-58 13 Table 2.3.4-1 13 Table 2.3.4-2 13 Table 2.3.4-3 13 Table 2.3.4-4 13 Table 2.3.4-5 13 Table 2.3.4-6 13 Table 2.3.4-7 13 Table 2.3.4-8 13 Table 2.3.4-9 13 Table 2.3.4-10 13 Table 2.3.4-11 13 Table 2.3.4-12 13 Table 2.3.4-13 13 Table 2.3.4-14 13 Figure 2.3.1-1 Original Figure 2.3.1-2 Original Figure 2.3.2-1 Original Figure 2.3.2-2 Original Figure 2.3.2-3 1 Figure 2.3.2-4 1 Figure 2.3.2-5 1 Figure 2.3.2-6 1 Figure 2.3.2-7 1 Figure 2.3.2-8 1 Figure 2.3.2-9 1 Figure 2.3.2-10 1 Figure 2.3.2-11 1 Figure 2.3.2-12 1 Figure 2.3.2-13 1 Figure 2.3.2-14 1 Figure 2.3.2-15 1 Figure 2.3.2-16 1 Figure 2.3.2-17 1 Figure 2.3.2-18 1 Figure 2.3.2-19 1 Figure 2.3.2-20 1 Figure 2.3.2-21 1 Figure 2.3.2-22 1 Figure 2.3.2-23 (Sheet 1) Original Figure 2.3.2-23 (Sheet 2) Original Figure 2.3.2-23 (Sheet 3) Original Figure 2.3.2-23 (Sheet 4) Original Figure 2.3.2-23 (Sheet 5) Original Figure 2.3.2-23 (Sheet 6) Original Figure 2.3.2-23 (Sheet 7) Original Figure 2.3.2-23 (Sheet 8) Original EPL 6

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 2.3.2-23 (Sheet 9) Original 2.4-1 17 2.4-2 17 2.4-3 17 2.4-4 17 2.4-5 17 2.4-6 27 2.4-7 17 2.4-8 26 2.4-9 17 2.4-10 17 2.4-11 17 2.4-12 26 2.4-13 26 2.4-14 17 2.4-15 17 2.4-16 17 2.4-17 17 2.4-18 17 2.4-19 17 2.4-20 17 2.4-21 17 2.4-22 17 2.4-23 17 2.4-24 17 2.4-25 17 2.4-26 17 2.4-27 17 2.4-28 17 2.4-29 17 2.4-30 17 2.4-31 17 2.4-32 21 2.4-33 27 2.4-34 24 2.4-35 21 2.4-36 21 2.4-37 21 2.4-38 21 2.4-39 21 2.4-40 25 2.4-41 23 2.4-42 17 2.4-43 24 Table 2.4.1-1 17 Table 2.4.1-2 17 Table 2.4.1-3 17 Table 2.4.1-4 17 Table 2.4.1-5 (Sheet 1) 17 Table 2.4.1-5 (Sheet 2) 17 Table 2.4.3-1 (Sheet 1) 13 EPL 7

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 2.4.3-1 (Sheet 2) 13 Table 2.4.3-2 13 Table 2.4.4-1 17 Table 2.4.13-1 (Sheet 1) 17 Table 2.4.13-1 (Sheet 2) 17 Table 2.4.13-1 (Sheet 3) 17 Table 2.4.13-2 (Sheet 1) 17 Table 2.4.13-2 (Sheet 2) 17 Table 2.4.13-2 (Sheet 3) 17 Table 2.4.13-2 (Sheet 4) 17 Figure 2.4.1-1 Original Figure 2.4.1-2 17 Figure 2.4.1-3 (Sheet 1) 17 Figure 2.4.1-3 (Sheet 2) 17 Figure 2.4.1-3 (Sheet 3) 17 Figure 2.4.1-3 (Sheet 4) 17 Figure 2.4.1-3 (Sheet 5) 17 Figure 2.4.1-3 (Sheet 6) 17 Figure 2.4.1-3 (Sheet 7) 17 Figure 2.4.1-3 (Sheet 8) 17 Figure 2.4.1-3 (Sheet 9) 17 Figure 2.4.1-3 (Sheet 10) 17 Figure 2.4.1-3 (Sheet 11) 17 Figure 2.4.1-3 (Sheet 12) 17 Figure 2.4.1-3 (Sheet 13) 17 Figure 2.4.1-3 (Sheet 14) 17 Figure 2.4.2-1 17 Figure 2.4.3-1 17 Figure 2.4.3-2 17 Figure 2.4.3-3 17 Figure 2.4.3-4 17 Figure 2.4.3-5 17 Figure 2.4.3-6 (Sheet 1) 17 Figure 2.4.3-6 (Sheet 2) 17 Figure 2.4.3-6 (Sheet 3) 17 Figure 2.4.3-6 (Sheet 4) 17 Figure 2.4.3-6 (Sheet 5) 17 Figure 2.4.3-6 (Sheet 6) 17 Figure 2.4.3-6 (Sheet 7) 17 Figure 2.4.3-6 (Sheet 8) 17 Figure 2.4.3-6 (Sheet 9) 17 Figure 2.4.3-6 (Sheet 10) 17 Figure 2.4.3-6 (Sheet 11) 17 Figure 2.4.3-7 17 Figure 2.4.3-8 17 Figure 2.4.3-9 17 Figure 2.4.3-10 17 Figure 2.4.3-11 17 Figure 2.4.3-12 17 Figure 2.4.3-13a 26 Figure 2.4.3-14 Original EPL 8

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 2.4.3-15 Original Figure 2.4.3-16 Original Figure 2.4.3-17 Original Figure 2.4.4-1 6 Figure 2.4.4-2 17 Figure 2.4.4-3 17 Figure 2.4.4-4 Original Figure 2.4.4-5 17 Figure 2.4.4-6 6 Figure 2.4.4-7 1 Figure 2.4.4-8 Original Figure 2.4.4-9 Original Figure 2.4.4-10 Original Figure 2.4.4-11 17 Figure 2.4.4-12 Original Figure 2.4.4-13 17 Figure 2.4.4-14 17 Figure 2.4.4-15 Original Figure 2.4.4-16 Original Figure 2.4.4-17 Original Figure 2.4.4-18 Original Figure 2.4.4-21 17 Figure 2.4.4-24 Original Figure 2.4.4-25 17 Figure 2.4.4-26 Original Figure 2.4.4-27 Original Figure 2.4.4-28 Original Figure 2.4.4-29 Original Figure 2.4.4-30 17 Figure 2.4.4-31 Original Figure 2.4.4-37 17 Figure 2.4.4-38 17 Figure 2.4.4-39 17 Figure 2.4.8-1 17 Figure 2.4.13-1 6 Figure 2.4.13-2 6 2.4A-i 13 2.4A-ii 13 2.4A-iii 17 2.4A-1 17 2.4A-2 27 2.4A-3 17 2.4A-4 21 2.4A-5 23 2.4A-6 17 2.4A-7 27 2.4A-8 13 2.4A-9 13 2.4A-10 17 2.4A-11 17 2.4A-12 17 EPL 9

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 2.4A-13 17 2.4A-14 17 2.4A-15 17 2.4A-16 24 Table 2.4A-2 17 Figure 2.4A-2 28 Figure 2.4A-3 28 Figure 2.4A-4 17 2.5-1 13 2.5-2 13 2.5-3 13 2.5-4 13 2.5-5 13 2.5-6 13 2.5-7 13 2.5-8 13 2.5-9 13 2.5-10 13 2.5-11 13 2.5-12 13 2.5-13 13 2.5-14 13 2.5-15 13 2.5-16 13 2.5-17 13 2.5-18 13 2.5-19 13 2.5-20 13 2.5-21 13 2.5-22 13 2.5-23 13 2.5-24 13 2.5-25 13 2.5-26 13 2.5-27 13 2.5-28 13 2.5-29 13 2.5-30 13 2.5-31 13 2.5-32 13 Table 2.5.1-1 13 Table 2.5.1-2 (Sheet 1) 13 Table 2.5.1-2 (Sheet 2) 13 Table 2.5.1-3 13 Table 2.5.1-4 13 Table 2.5.1-5 13 Table 2.5.1-6 13 Table 2.5.1-7 13 Table 2.5.1-8 13 Table 2.5.1-9 13 Table 2.5.1-10 13 EPL 10

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 2.5.1-11 (Sheet 1) 13 Table 2.5.1-11 (Sheet 2) 13 Table 2.5.1-12 13 Table 2.5.1-13 (Sheet 1) 13 Table 2.5.1-13 (Sheet 2) 13 Table 2.5.1-14 13 Table 2.5.2-1 (Sheet 1) 13 Table 2.5.2-1 (Sheet 2) 13 Table 2.5.2-1 (Sheet 3) 13 Table 2.5.2-1 (Sheet 4) 13 Table 2.5.2-1 (Sheet 5) 13 Table 2.5.2-1 (Sheet 6) 13 Figure 2.5.1-1 Original Figure 2.5.1-2 Original Figure 2.5.1-3 Original Figure 2.5.1-4 Original Figure 2.5.1-5 Original Figure 2.5.1-6 Original Figure 2.5.1-7 Original Figure 2.5.1-8 Original Figure 2.5.1-9 Original Figure 2.5.1-10 Original Figure 2.5.1-11 Original Figure 2.5.1-12 23 Figure 2.5.1-12a 13 Figure 2.5.1-12b 13 Figure 2.5.1-13 13 Figure 2.5.1-13a 13 Figure 2.5.1-14 Original Figure 2.5.1-15 Original Figure 2.5.2-1 Original Figure 2.5.2-2 Original Figure 2.5.2-3 Original Figure 2.5.2-4 Original Figure 2.5.2-5 Original Figure 2.5.2-6 Original Figure 2.5.2-7 Original Figure 2.5.2-8 Original Figure 2.5.2-9 Original Figure 2.5.2-10 Original Figure 2.5.2-11 Original Figure 2.5.2-12 Original Figure 2.5.2-13 Original Figure 2.5.2-14 Original Figure 2.5.5-1 Original Figure 2.5.6-1 6 Figure 2.5.6-2 Original 2.6-1 13 Table 2.6-1 13 EPL 11

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment CHAPTER 3 3-1 13 3-2 26 3-3 13 3-4 16 3-5 13 3-6 13 3-7 31 3-8 26 3-9 18 3-10 13 3-11 24 3-12 24 3-13 25 3-14 25 3-15 31 3-16 13 3-17 13 3-18 13 3-19 18 3-20 18 3-21 13 3-22 13 3.1-1 13 3.1-2 13 3.1-3 13 3.1-4 13 3.1-5 13 3.1-6 13 3.1-7 13 3.1-8 13 3.1-9 29 3.1-10 13 3.1-11 13 3.1-12 13 3.1-13 13 3.1-14 13 3.1-15 13 3.1-16 13 3.1-17 13 3.1-18 18 3.1-19 13 3.1-20 13 3.1-21 13 3.1-22 13 3.1-23 13 3.1-24 20 3.1-25 20 3.1-26 13 3.1-27 13 3.1-28 21 EPL 12

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.1-29 13 3.1-30 13 3.1-31 13 3.1-32 13 3.1-33 18 3.2-1 13 3.2-2 15 3.2-3 13 3.2-4 30 3.2-5 30 Table 3.2.1-1 26 Table 3.2.1-2 (Sheet 1) 26 Table 3.2.1-2 (Sheet 2) 13 Table 3.2.1-2 (Sheet 3) 13 Table 3.2.1-2 (Sheet 4) 23 Table 3.2.1-2 (Sheet 5) 26 Table 3.2.1-2 (Sheet 6) 31 Table 3.2.1-2 (Sheet 7) 13 Table 3.2.1-2 (Sheet 8) 13 Table 3.2.1-2 Notes (Sheet 9) 18 Table 3.2.1-2 Notes (Sheet 10) 19 Table 3.2.1-2 Notes (Sheet 11) 13 Table 3.2.1-3 (Sheet 1) 18 Table 3.2.1-3 (Sheet 2) 13 Table 3.2.1-3 (Sheet 3) 13 Table 3.2.1-3 (Sheet 4) 30 Table 3.2.1-3 (Sheet 5) 30 Table 3.2.2-1 13 Table 3.2.2-2 (Sheet 1) 13 Table 3.2.2-2 (Sheet 2) 13 Table 3.2.2-3 (Sheet 1) 13 Table 3.2.2-3 (Sheet 2) 13 Table 3.2.2-3 (Sheet 3) 19 3.3-1 21 3.3-2 26 3.3-3 25 3.3-4 13 Figure 3.3.2-1 Original 3.4-1 13 3.5-1 13 3.5-2 13 3.5-3 13 3.5-4 13 3.5-5 13 3.5-6 13 3.5-7 13 3.5-8 13 3.5-9 13 3.5-10 13 3.5-11 26 3.5-12 13 EPL 13

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.5-13 13 3.5-14 13 3.5-15 13 3.5-16 21 3.5-17 13 3.5-18 13 3.5-19 13 3.5-20 13 3.5-21 19 3.5-22 13 3.5-23 23 3.5-24 13 Table 3.5.1-1 (Sheet 1) 13 Table 3.5.1-1 (Sheet 2) 26 Table 3.5.2-1 25 Table 3.5.5-1 13 Table 3.5.5-2 26 Table 3.5.5-3 (Sheet 1) 13 Table 3.5.5-3 (Sheet 2) 13 Table 3.5.5-3 (Sheet 3) 13 Table 3.5.5-4 13 Table 3.5.5-5 13 Figure 3.5.2-1 6 Figure 3.5.2-2 Original Figure 3.5.2-3 Original Figure 3.5.4-1 Original Figure 3.5.4-2 Original Figure 3.5.4-3 Original Figure 3.5.4-4 Original 3.6-1 29 3.6-2 13 3.6-3 13 3.6-4 13 3.6-5 13 3.6-6 13 3.6-7 13 3.6-8 13 3.6-9 18 3.6-10 13 3.6-11 13 3.6-12 13 3.6-13 13 3.6-14 13 3.6-15 13 3.6-16 13 3.6-17 13 3.6-18 18 3.6-19 24 3.6-20 13 3.6-21 13 3.6-22 13 EPL 14

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.6-23 30 3.6-24 13 3.6-25 24 Table 3.6.1-1 (Sheet 1) 21 Table 3.6.1-1 (Sheet 2) 21 Table 3.6.1-1 (Sheet 3) 21 Table 3.6.2-1 18 Table 3.6.7-1 18 Table 3.6.7-2 13 Table 3.6.7-3 13 Figure 3.6.1-1 13 Figure 3.6.2-1 18 Figure 3.6.4-2 Original Figure 3.6.7-1 18 Figure 3.6.7-2 18 3.7-1 13 3.7-2 17 3.7-3 31 3.7-4 31 3.7-5 13 3.7-6 13 3.7-7 25 3.7-8 13 3.7-9 13 3.7-10 24 3.7-10a 18 3.7-11 13 3.7-12 13 3.7-13 13 3.7-14 13 3.7-15 13 3.7-16 13 3.7-17 13 3.7-18 13 3.7-19 13 3.7-20 13 3.7-21 13 3.7-22 13 3.7-23 26 3.7-24 18 3.7-25 13 3.7-26 18 3.7-27 17 3.7-28 13 3.7-29 25 3.7-29a 18 3.7-30 13 3.7-31 13 3.7-32 13 3.7-33 13 3.7-34 13 EPL 15

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.7-35 19 3.7-35a 19 3.7-36 16 3.7-37 16 3.7-38 16 3.7-39 16 3.7-40 22 3.7-41 13 3.7-42 26 Table 3.7.1-1 13 Table 3.7.1-2 13 Table 3.7.1-3 (Sheet 1) 13 Table 3.7.1-3 (Sheet 2) 13 Table 3.7.1-3A (Sheet 1) 13 Table 3.7.1-3A (Sheet 2) 13 Table 3.7.1-4 13 Table 3.7.1-5 13 Table 3.7.2-1 13 Table 3.7.2-2 (Sheet 1) 16 Table 3.7.2-2 (Sheet 2) 13 Table 3.7.2-6 13 Table 3.7.2-7 13 Table 3.7.2-10 13 Table 3.7.2-11 13 Table 3.7.2-12 13 Table 3.7.2-13 13 Table 3.7.2-14 25 Table 3.7.2-15 13 Table 3.7.2-16 13 Table 3.7.2-17 13 Table 3.7.2-18 13 Table 3.7.2-19 13 Table 3.7.2-20 13 Table 3.7.2-21 13 Table 3.7.2-22 13 Table 3.7.2-23 13 Table 3.7.2-24 13 Table 3.7.2-25 25 Table.3.7.2-26 13 Table 3.7.2-27 13 Table 3.7.2-28 13 Table 3.7.2-29 13 Table 3.7.2-30 13 Table 3.7.2-31 25 Table 3.7.2-32 13 Table 3.7.2-33 13 Table 3.7.2-34 25 Table 3.7.2-35 13 Table 3.7.2-36 13 Table 3.7.2-37 13 Table 3.7.2-38 13 EPL 16

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 3.7.2-39 13 Table 3.7.2-40 13 Table 3.7.2-41 13 Table 3.7.2-42 13 Table 3.7.2-43 13 Table 3.7.2-44 13 Table 3.7.2-45 13 Figure 3.7.1-1 13 Figure 3.7.1-2 13 Figure 3.7.2-3 Original Figure 3.7.2-4 Original Figure 3.7.2-5 Original Figure 3.7.2-6 Original Figure 3.7.2-7 Original Figure 3.7.2-8 Original Figure 3.7.2-9 Original Figure 3.7.2-10 Original Figure 3.7.2-11 Original Figure 3.7.2-12 Original Figure 3.7.2-13 Original Figure 3.7.2-14 Original Figure 3.7.2-15 Original Figure 3.7.2-16 Original Figure 3.7.2-17 Original Figure 3.7.2-18 Original Figure 3.7.2-19 Original Figure 3.7.2-20 Original Figure 3.7.2-21 Original Figure 3.7.2-22 Original Figure 3.7.2-23 Original Figure 3.7.2-24 Original Figure 3.7.2-25 Original Figure 3.7.2-26 Original Figure 3.7.2-27 Original Figure 3.7.2-28 Original Figure 3.7.2-29 Original Figure 3.7.2-30 Original Figure 3.7.2-31 Original Figure 3.7.2-32 Original Figure 3.7.2-33 Original Figure 3.7.2-34 Original Figure 3.7.2-35 Original Figure 3.7.2-36 Original Figure 3.7.2-37 Original Figure 3.7.2-38 Original Figure 3.7.2-39 Original Figure 3.7.2-40 Original Figure 3.7.2-41 Original Figure 3.7.2-42 Original Figure 3.7.2-43 Original Figure 3.7.2-44 Original EPL 17

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 3.7.2-45 Original Figure 3.7.2-46 Original Figure 3.7.2-47 Original Figure 3.7.2-48 6 Figure 3.7.2-49 Original Figure 3.7.2-50 Original Figure 3.7.2-51 Original Figure 3.7.2-52 Original Figure 3.7.2-53 Original Figure 3.7.2-54 6 Figure 3.7.2-55 6 Figure 3.7.2-56 6 Figure 3.7.2-57 6 Figure 3.7.2-58 6 Figure 3.7.2-59 6 Figure 3.7.2-60 6 Figure 3.7.2-61 6 Figure 3.7.2-62 6 Figure 3.7.2-63 6 Figure 3.7.2-64 6 Figure 3.7.2-65 6 Figure 3.7.2-66 6 Figure 3.7.2-67 6 Figure 3.7.2-68 6 Figure 3.7.2-69 6 Figure 3.7.2-70 6 Figure 3.7.2-71 6 Figure 3.7.2-72 6 Figure 3.7.2-73 6 Figure 3.7.2-74 6 Figure 3.7.2-75 6 Figure 3.7.2-76 6 Figure 3.7.2-77 Original Figure 3.7.2-78 Original Figure 3.7.2-79 18 3.8-1 20 3.8-2 24 3.8-3 23 3.8-4 13 3.8-5 17 3.8-6 13 3.8-7 13 3.8-8 13 3.8-9 13 3.8-10 13 3.8-11 13 3.8-12 24 3.8-13 24 3.8-14 13 3.8-15 24 3.8-16 13 EPL 18

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.8-17 21 3.8-18 13 3.8-19 13 3.8-20 13 3.8-21 13 3.8-22 24 3.8-23 13 3.8-24 20 3.8-25 13 3.8-26 13 3.8-27 13 3.8-28 13 3.8-29 13 3.8-30 13 3.8-31 21 3.8-32 13 3.8-33 13 3.8-34 13 3.8-35 13 3.8-36 13 3.8-37 13 3.8-38 13 3.8-39 21 3.8-40 13 3.8-41 13 3.8-42 13 3.8-43 26 3.8-44 26 3.8-45 24 3.8-46 13 3.8-47 24 3.8-48 13 3.8-49 17 3.8-50 18 3.8-51 17 3.8-52 13 3.8-53 13 3.8-54 13 3.8-55 13 3.8-56 13 3.8-57 13 3.8-58 24 3.8-58a 18 3.8-59 13 3.8-60 13 3.8-61 13 3.8-62 13 3.8-63 13 3.8-64 13 3.8-65 16 3.8-66 13 EPL 19

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.8-67 13 3.8-68 13 3.8-69 13 3.8-70 13 3.8-71 13 3.8-72 18 3.8-73 13 3.8-74 13 3.8-75 19 3.8-76 13 3.8-77 23 3.8-78 13 3.8-79 23 3.8-80 26 3.8-81 13 3.8-82 13 3.8-83 13 3.8-84 13 3.8-85 13 3.8-86 13 3.8-87 13 3.8-88 19 3.8-89 13 3.8-90 13 3.8-91 13 3.8-92 13 3.8-93 13 3.8-94 13 3.8-95 13 3.8-96 13 3.8-97 13 3.8-98 13 3.8-99 13 3.8-100 13 3.8-101 13 3.8-102 13 3.8-103 13 3.8-104 13 3.8-105 13 3.8-106 13 3.8-107 23 3.8-108 28 3.8-109 13 3.8-110 13 3.8-111 28 3.8-112 19 3.8-113 19 3.8-114 13 Table 3.8.1-1 13 Table 3.8.1-2 13 Table 3.8.2-1 (Sheet 1) 13 EPL 20

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 3.8.2-1 (Sheet 2) 13 Table 3.8.2-2 (Sheet 1) 13 Table 3.8.2-2 (Sheet 2) 13 Table 3.8.2-2 (Sheet 3) 13 Table 3.8.3-1 13 Table 3.8.3-2 13 Table 3.8.3-2A 24 Table 3.8.3-3 13 Table 3.8.3-4 (Sheet 1) 13 Table 3.8.3-4 (Sheet 2) 13 Table 3.8.3-4 (Sheet 3) 13 Table 3.8.3-5 13 Table 3.8.3-6 13 Table 3.8.3-7 13 Table 3.8.3-8 21 Table 3.8.3-9 13 Table 3.8.3-10 (Sheet 1) 13 Table 3.8.3-10 (Sheet 2) 13 Table 3.8.3-11 13 Table 3.8.4.1 (Sheet 1) 13 Table 3.8.4-1 (Sheet 2) 21 Table 3.8.4-1 (Sheet 3) 21 Table 3.8.4-1 (Sheet 4) 13 Table 3.8.4-1 (Sheet 5) 13 Table 3.8.4-2 (Sheet 1) 13 Table 3.8.4-2 (Sheet 2) 13 Table 3.8.4-2 (Sheet 3) 13 Table 3.8.4-3 (Sheet 1) 13 Table 3.8.4-3 (Sheet 2) 13 Table 3.8.4-4 13 Table 3.8.4-5 13 Table 3.8.4-6 13 Table 3.8.4-7 13 Table 3.8.4-8 (Sheet 1) 13 Table 3.8.4-8 (Sheet 2) 13 Table 3.8.4-9 (Sheet 1) 13 Table 3.8.4-9 (Sheet 2) 13 Table 3.8.4-10 13 Table 3.8.4-11 13 Table 3.8.4-12 13 Table 3.8.4-13 13 Table 3.8.4-14 (Sheet 1) 13 Table 3.8.4-14 (Sheet 2) 13 Table 3.8.4-15 (Sheet 1) 13 Table 3.8.4-15 (Sheet 2) 13 Table 3.8.4-16 13 Table 3.8.4-17 (Sheet 1) 13 Table 3.8.4-17 (Sheet 2) 13 Table 3.8.4-18 13 Table 3.8.6-1 (Sheet 1) 13 Table 3.8.6-1 (Sheet 2) 13 EPL 21

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 3.8.1-1 13 Figure 3.8.2-1 13 Figure 3.8.2-2 13 Figure 3.8.2-3 13 Figure 3.8.2-4 13 Figure 3.8.2-5 13 Figure 3.8.2-6 13 Figure 3.8.2-7 27 Figure 3.8.2-8 13 Figure 3.8.2-9 13 Figure 3.8.2-10 13 Figure 3.8.2-11 13 Figure 3.8.3-1 13 Figure 3.8.3-2 13 Figure 3.8.4-1 23 Figure 3.8.4-2 13 Figure 3.8.4-3 13 Figure 3.8.4-4 13 Figure 3.8.4-5 13 Figure 3.8.4-6 13 Figure 3.8.4-7 19 Figure 3.8.4-8 13 Figure 3.8.4-9 19 Figure 3.8.4-10 13 Figure 3.8.4-11 13 Figure 3.8.6-1 13 Figure 3.8.6-2 13 Figure 3.8.6-3 28 Figure 3.8.6-4 13 Figure 3.8.6-5 28 Figure 3.8.6-6 18 Figure 3.8.6-7 18 Figure 3.8.6-8 18 Figure 3.8.6-9 13 Figure 3.8.6-10 13 Figure 3.8.6-11 13 Figure 3.8.6-12 Original 3.8A-1 13 3.8A-2 13 Figure 3.8A-1 6 Figure 3.8A-2 6 3.8B-1 13 3.8B-2 13 3.8B-3 13 3.8B-4 13 3.8B-5 13 3.8B-6 13 3.8C-1 13 3.8C-2 13 Figure 3.8C-1 Original Figure 3.8C-2 6 EPL 22

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 3.8C-3 6 Figure 3.8C-4 6 3.8D-1 13 3.8D-2 13 3.8D-3 13 Table 3.8D-1 13 Table 3.8D-2 13 Table 3.8D-3 13 Table 3.8D-4 13 Table 3.8D-5 13 Table 3.8D-6 13 Table 3.8D-7 13 Table 3.8D-8 13 Figure 3.8D-1 Original Figure 3.8D-2 Original 3.8E-1 13 3.8E-2 13 3.8E-3 13 3.8E-4 13 3.8E-5 13 Figure 3.8E-1 Original 3.9-1 13 3.9-2 13 3.9-3 13 3.9-4 13 3.9-5 13 3.9-6 31 3.9-7 13 3.9-8 13 3.9-9 13 3.9-10 13 3.9-11 13 3.9-12 13 3.9-13 13 3.9-14 13 3.9-15 18 3.9-16 18 3.9-17 13 3.9-18 26 3.9-19 13 3.9-20 13 3.9-21 13 3.9-22 13 3.9-23 13 3.9-24 13 3.9-25 13 3.9-26 13 3.9-27 13 3.9-28 13 3.9-29 13 3.9-30 13 EPL 23

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.9-31 13 3.9-32 31 3.9-33 31 3.9-34 21 3.9-35 31 3.9-36 31 3.9-37 31 3.9-38 13 3.9-39 31 3.9-40 31 3.9-41 31 3.9-42 31 Table 3.9.2-1 26 Table 3.9.2-2 13 Table 3.9.2-3 22 Table 3.9.2-3a 22 Table 3.9.2-4 (Sheet 1) 13 Table 3.9.2-4 (Sheet 2) 13 Table 3.9.2-5 (Sheet 1) 14 Table 3.9.2-5 (Sheet 2) 14 Table 3.9.3-1 13 Figure 3.9.1-1 Original Figure 3.9.1-2 Original Figure 3.9.1-3 Original Figure 3.9.1-4 Original Figure 3.9.1-5 (Sheet 1) Original Figure 3.9.1-5 (Sheet 2) Original Figure 3.9.1-6 6 Figure 3.9.1-7 Original Figure 3.9.1-8 18 Figure 3.9.2-1 10 3.10-1 13 3.10-2 13 3.10-3 13 3.10-4 13 3.10-5 13 3.10-6 13 3.10-7 13 3.10-8 13 3.10-9 13 Table 3.10.2-1 13 3.11-1 13 3.11-2 13 Table 3.11.1-1 (Sheet 1) 20 Table 3.11.1-1 (Sheet 2) 13 3.12-1 21 3.12-2 21 3.12-3 21 3.12-4 21 3.12-5 21 3.12-6 21 EPL 24

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 3.12-7 21 3.13-1 26 CHAPTER 4 4-1 24 4-2 31 4-3 24 4-4 17 4-5 31 4-6 31 4-7 31 4-8 13 4-9 16 4-10 24 4.1-1 31 4.1-2 31 4.1-3 31 4.1-4 31 Table 4.1-1 (Sheet 1) 31 Table 4.1-1 (Sheet 2) 31 Table 4.1-1 (Sheet 3) 31 Table 4.1-1 (Sheet 4) 31 Table 4.1-2 (Sheet 1) 31 Table 4.1-2 (Sheet 2) 31 Table 4.1-2 (Sheet 3) 31 Table 4.1-3 13 4.2-1 13 4.2-2 31 4.2-3 31 4.2-4 31 4.2-5 31 4.2-6 31 4.2-7 31 4.2-8 31 4.2-9 31 4.2-10 31 4.2-11 31 4.2-12 31 4.2-13 31 4.2-14 31 4.2-15 31 4.2-16 31 4.2-17 31 4.2-18 31 4.2-19 31 4.2-20 31 4.2-21 31 4.2-22 31 4.2-23 31 4.2-24 31 4.2-25 31 4.2-26 31 EPL 25

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 4.2-27 31 4.2-28 31 4.2-29 31 4.2-30 31 4.2-31 31 4.2-32 31 4.2-33 31 4.2-34 31 4.2-35 31 4.2-36 31 4.2-37 31 4.2-38 31 4.2-39 31 4.2-40 31 4.2-41 31 4.2-42 31 4.2-43 31 4.2-44 31 4.2-45 31 4.2-46 31 4.2-47 31 4.2-48 31 4.2-49 31 4.2-50 31 4.2-51 31 4.2-52 31 4.2-53 31 4.2-54 31 4.2.55 31 4.2.56 31 Table 4.2.2-1 13 Figure 4.2.1-1 31 Figure 4.2.1-1A 31 Figure 4.2.1-1B 31 Figure 4.2.1-2 Original Figure 4.2.1-2A 8 Figure 4.2.1-2B 31 Figure 4.2.1-3 31 Figure 4.2.1-3A 8 Figure 4.2.1-3A Notes 31 Figure 4.2.1-3B 31 Figure 4.2.1-4 Original Figure 4.2.1-5 31 Figure 4.2.1-6a 31 Figure 4.2.1-6b 31 Figure 4.2.1-7B 8 Figure 4.2.1-7c Original Figure 4.2.1-7d 31 Figure 4.2.1-8 31 Figure 4.2.1-8b 31 Figure 4.2.1-9 31 EPL 26

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 4.2.1-10 31 Figure 4.2.1-11 31 Figure 4.2.1-12 31 Figure 4.2.2-1 Original Figure 4.2.2-2 Original Figure 4.2.2-3 Original Figure 4.2.2-4 Original Figure 4.2.2-5 Original Figure 4.2.3-1 17 Figure 4.2.3-2 Original Figure 4.2.3-2A Original Figure 4.2.3-3 Original Figure 4.2.3-3A Original Figure 4.2.3-3B Original Figure 4.2.3-3C Original Figure 4.2.3-4 Original Figure 4.2.3-5a 10 Figure 4.2.3-5b 10 Figure 4.2.3-6 1 Figure 4.2.3-7 Original Figure 4.2.3-8 Original Figure 4.2.3-9 Original Figure 4.2.3-10 Original 4.3-1 13 4.3-2 31 4.3-3 31 4.3-4 26 4.3-5 31 4.3-6 31 4.3-7 13 4.3-8 13 4.3-9 31 4.3-10 13 4.3-11 31 4.3-12 31 4.3-13 13 4.3-14 13 4.3-15 31 4.3-16 31 4.3-17 13 4.3-18 13 4.3-19 15 4.3-20 13 4.3-21 13 4.3-22 13 4.3-23 13 4.3-24 13 4.3-25 17 4.3-26 26 4.3-27 26 4.3-28 16 EPL 27

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 4.3-29 16 4.3-30 16 4.3-31 13 4.3-32 13 4.3-33 13 4.3-34 13 4.3-35 13 4.3-36 31 4.3-37 31 4.3-38 31 4.3-39 31 4.3-40 31 Table 4.3.2-1 (Sheet 1) 25 Table 4.3.2-1 (Sheet 2) 13 Table 4.3.2-2 13 Table 4.3.2-3 30 Table 4.3.2-4 13 Table 4.3.2-5 13 Table 4.3.2-6 13 Table 4.3.2-7 13 Table 4.3.2-8 13 Table 4.3.2-9 13 Table 4.3.2-10 13 Table 4.3.2-11 6 Figure 4.3.2-1 Original Figure 4.3.2-2 Original Figure 4.3.2-3 Original Figure 4.3.2-4 Original Figure 4.3.2-5 Original Figure 4.3.2-6 10 Figure 4.3.2-7 10 Figure 4.3.2-8 10 Figure 4.3.2-9 6 Figure 4.3.2-10 10 Figure 4.3.2-11 10 Figure 4.3.2-12 Original Figure 4.3.2-13 Original Figure 4.3.2-14 Original Figure 4.3.2-15 Original Figure 4.3.2-16 Original Figure 4.3.2-17 Original Figure 4.3.2-21 13 Figure 4.3.2-22 13 Figure 4.3.2-23 13 Figure 4.3.2-24 10 Figure 4.3.2-25 Original Figure 4.3.2-26 Original Figure 4.3.2-27 Original Figure 4.3.2-28 Original Figure 4.3.2-29 Original Figure 4.3.2-30 Original EPL 28

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 4.3.2-31 Original Figure 4.3.2-32 Original Figure 4.3.2-33 Original Figure 4.3.2-34 Original Figure 4.3.2-35 Original Figure 4.3.2-36 30 Figure 4.3.2-37 Original Figure 4.3.2-38 Original Figure 4.3.2-39 Original Figure 4.3.2-40 8 Figure 4.3.2-41 Original Figure 4.3.2-42 Original Figure 4.3.2-43 Original Figure 4.3.2-44 Original Figure 4.3.2-45 Original Figure 4.3.2-46 Original 4.4-1 31 4.4-2 31 4.4-3 31 4.4-4 31 4.4-5 31 4.4-6 31 4.4-7 31 4.4-8 31 4.4-9 31 4.4-10 31 4.4-11 31 4.4-12 31 4.4-13 31 4.4-14 31 4.4-15 31 4.4-16 31 4.4-17 31 4.4-18 31 4.4-19 31 4.4-20 31 4.4-21 31 4.4-22 31 4.4-23 31 4.4-24 31 4.4-25 31 4.4-26 31 4.4-27 31 4.4-28 31 4.4-29 31 4.4-30 31 4.4-31 31 4.4-32 31 4.4-33 31 4.4-34 31 Table 4.4.2-1 31 EPL 29

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 4.4.2-2 13 Table 4.4.2-3 13 Figure 4.4.2-3 Original Figure 4.4.2-4 13 Figure 4.4.2-5 13 Figure 4.4.2-6 Original Figure 4.4.2-7 10 Figure 4.4.2-8 10 Figure 4.4.2-9 10 Figure 4.4.2-10 Original Figure 4.4.5-1 30 4.5-1 31 4.5-2 31 4.5-3 24 4.5-4 24 4.5-5 24 4.5-6 24 4.5-7 24 4.5-8 24 4.5-9 24 4.5-10 24 4.5-11 24 4.5-12 24 4.5-13 24 4.5-14 24 4.5-15 24 4.5-16 24 4.5-17 24 4.5-18 24 4.5-19 24 4.5-20 24 4.5-21 24 4.5-22 24 4.5-23 24 4.5-24 24 4.5-25 24 4.5-26 24 4.5-27 24 4.5-28 24 4.5-29 24 4.5-30 24 4.5-31 24 4.5-32 24 4.5-33 24 4.5-34 27 4.5-35 26 4.5-36 24 4.5-37 24 4.5-38 24 4.5-39 24 4.5-40 24 EPL 30

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 4.5-41 24 4.5-42 24 4.5-43 24 4.5-44 24 4.5-45 24 4.5-46 29 4.5-47 24 4.5-48 24 4.5-49 24 4.5-50 31 4.5-51 24 4.5-52 24 4.5-53 24 4.5-54 24 4.5-55 24 4.5-56 31 4.5-57 29 4.5-58 29 4.5-59 24 4.5-60 24 4.5-61 24 4.5-62 24 4.5-63 31 4.5-64 24 4.5-65 24 Table 4.5.1-1 24 Table 4.5.1-2 24 Table 4.5.1-3 13 Table 4.5.2-1 13 Table 4.5.2-2 25 Table 4.5.2-3 25 Table 4.5.3-1 24 Table 4.5.4.2-1 (Sheet 1) 29 Table 4.5.4.2-1 (Sheet 2) 29 Figure 4.5.2-1 16 Figure 4.5.2-2 13 Figure 4.5.2-3 13 Figure 4.5.2-4 13 Figure 4.5.2-5A 16 Figure 4.5.2-5B 16 Figure 4.5.2-6 16 Figure 4.5.2-7A 16 Figure 4.5.2-7B 16 Figure 4.5.2-8A 16 Figure 4.5.2-8B 16 Figure 4.5.2-9 13 Figure 4.5.2-10 13 Figure 4.5.2-11 13 Figure 4.5.2-12 13 Figure 4.5.2-13 13 Figure 4.5.2-14 13 EPL 31

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 4.5.2-15 16 Figure 4.5.2-16a 24 Figure 4.5.2-16b 24 Figure 4.5.2-17 24 Figure 4.5.2-18 24 Figure 4.5.2-19 24 Figure 4.5.2-20 24 Figure 4.5.2-21 24 Figure 4.5.2-22 24 Figure 4.5.2-23 24 Figure 4.5.2-24 24 Figure 4.5.2-25 24 Figure 4.5.2-26 24 Figure 4.5.2-27 24 Figure 4.5.4.2-7 24 Figure 4.5.4.2-8 24 Figure 4.5.4.2-9 24 CHAPTER 5 5-1 18 5-2 18 5-3 25 5-4 18 5-5 26 5-6 13 5-7 25 5-8 18 5.1-1 18 5.1-2 24 5.1-3 13 5.1-4 20 5.1-4a 20 5.1-5 13 5.1-6 28 5.1-7 28 5.1-8 20 Table 5.1-1 (Sheet 1) 31 Table 5.1-1 (Sheet 2) 31 Figure 5.1-1 27 Figure 5.1-2 13 Figure 5.1-3 13 5.2-1 18 5.2-2 25 5.2-3 18 5.2-4 18 5.2-5 18 5.2-6 18 5.2-7 18 5.2-8 18 5.2-9 24 5.2-10 28 5.2-11 18 EPL 32

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 5.2-12 24 5.2-13 18 5.2-14 18 5.2-15 24 5.2-16 24 5.2-17 18 5.2-18 18 5.2-19 25 5.2-20 18 5.2-21 25 5.2-22 18 5.2-23 25 5.2-24 25 5.2-25 23 5.2-26 18 5.2-27 24 5.2-28 24 5.2-29 18 5.2-30 18 5.2-31 31 5.2-32 19 5.2-33 31 5.2-34 18 5.2-35 18 5.2-36 18 5.2-37 26 5.2-38 19 5.2-39 24 5.2-40 28 5.2-41 28 5.2-42 18 5.2-43 18 5.2-44 18 5.2-45 18 5.2-46 18 5.2-47 18 5.2-48 20 5.2-49 18 5.2-50 18 5.2-51 18 5.2-52 18 5.2-53 18 5.2-54 25 5.2-55 25 5.2-56 28 5.2-57 18 5.2-58 18 5.2-59 22 5.2-60 22 5.2-61 25 Table 5.2-1 13 EPL 33

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 5.2.1-1 (Sheet 1) 31 Table 5.2.1-1 (Sheet 2) 31 Table 5.2.1-22 13 Table 5.2.1-23 13 Table 5.2.1-24 13 Table 5.2-2 13 Table 5.2.2-1 13 Table 5.2.2-2 13 Table 5.2-3 (Sheet 1) 13 Table 5.2-3 (Sheet 2) 13 Table 5.2.3-1 (Sheet 1) 24 Table 5.2.3-1 (Sheet 2) 24 Table 5.2.3-1 (Sheet 3) 24 Table 5.2.3-2 (Sheet 1) 30 Table 5.2.3-2 (Sheet 2) 13 Table 5.2.3-3 25 Table 5.2-4 25 Table 5.2.4-1 25 Table 5.2.4-2 25 Table 5.2.4-3 25 Table 5.2.4-4 13 Table 5.2.4-5 13 Table 5.2.4-6 13 Table 5.2.4-7 13 Table 5.2.4-8 (Sheet 1) 28 Table 5.2.4-8 (Sheet 2) 28 Table 5.2-5 18 Table 5.2.5-1 13 Figure 5.2.1-4 24 Figure 5.2.1-5 Original Figure 5.2.1-6 24 Figure 5.2.1-8 Original Figure 5.2.1-9 Original Figure 5.2.1-10 Original Figure 5.2.1-11 Original Figure 5.2.1-12 Original Figure 5.2.1-13 Original Figure 5.2.1-14 18 Figure 5.2.1-15 Original Figure 5.2.1-16 Original Figure 5.2.6-1 Original Figure 5.2.7-1 13 5.3-1 31 Figure 5.3.4-1 13 5.4-1 13 5.4-2 13 5.4-3 17 5.4-4 17 5.4-5 24 5.4-6 17 5.4-7 29 EPL 34

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 5.4-8 29 5.4-9 25 5.4-10 25 5.4-11 25 5.4-12 25 5.4-13 25 5.4-14 25 5.4-15 25 5.4-16 28 Table 5.4.2-1 13 Table 5.4.4-1 13 Figure 5.4.2-1 Original Figure 5.4.3-1 Original Figure 5.4.3-2 Original Figure 5.4.3-3 29 5.5-1 18 5.5-2 18 5.5-3 25 5.5-4 18 5.5-5 24 5.5-6 18 5.5-7 25 5.5-8 18 5.5-9 24 5.5-10 24 5.5-11 24 5.5-12 25 5.5-13 25 5.5-14 25 5.5-15 25 5.5-16 18 5.5-17 18 5.5-18 18 5.5-19 18 5.5-20 18 5.5-21 18 5.5-22 27 5.5-23 18 5.5-24 18 5.5-25 31 5.5-25a 19 5.5-26 18 5.5-27 18 5.5-28 18 5.5-29 18 5.5-30 18 5.5-31 18 5.5-32 18 5.5-33 27 5.5-34 30 5.5-35 18 EPL 35

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 5.5-36 24 5.5-37 31 5.5-38 18 5.5-39 18 Table 5.5.1-1 16 Table 5.5.1-2 13 Table 5.5.2-1 24 Table 5.5.2-2 13 Table 5.5.3-1 13 Table 5.5.7-1 17 Table 5.5.7-2 16 Table 5.5.10-1 21 Table 5.5.10-2 13 Table 5.5.11-1 13 Table 5.5.12-1 18 Table 5.5.13-1 13 Figure 5.5.1-1 Original Figure 5.5.1-2 Original Figure 5.5.1-3 Original Figure 5.5.2-1a 24 Figure 5.5.2-1b 24 Figure 5.5.7-1 13 Figure 5.5.7-2 22 Figure 5.5.10-1 Original Figure 5.5.14-1 Original Figure 5.5.14-2 24 Figure 5.5.14-3 Original Figure 5.5.14-4 Original Figure 5.5.14-5 24 5.6-1 28 5.6-2 13 5.6-3 28 5.6-4 28 5.6-5 28 5.6-6 28 CHAPTER 6 6-1 29 6-2 20 6-3 13 6-4 15 6-5 13 6-6 18 6-7 21 6-8 20 6-9 13 6-10 13 6-11 13 6-12 14 6-13 14 6-14 14 6-15 20 EPL 36

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6-16 13 6-17 13 6.1-1 30 6.1-2 29 6.2-1 17 6.2-2 13 6.2-3 13 6.2-4 13 6.2-5 30 6.2-6 14 6.2-7 31 6.2-8 13 6.2-9 13 6.2-10 13 6.2-11 13 6.2-12 13 6.2-13 25 6.2-14 13 6.2-15 13 6.2-16 13 6.2-17 25 6.2-18 18 6.2-19 18 6.2-20 18 6.2-21 24 6.2-22 24 6.2-22a 24 6.2-23 13 6.2-24 13 6.2-25 13 6.2-26 13 6.2-27 13 6.2-28 14 6.2-29 14 6.2-30 14 6.2-31 17 6.2-32 14 6.2-33 25 6.2-34 14 6.2-35 25 6.2-36 24 6.2-37 14 6.2-38 24 6.2-39 24 6.2-40 18 6.2-41 25 6.2-42 18 6.2-43 18 6.2-44 24 6.2-45 18 6.2-46 24 EPL 37

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6.2-47 24 6.2-48 13 6.2-49 22 6.2-50 22 6.2-51 15 6.2-52 30 6.2-53 20 6.2-53a 21 6.2-54 13 6.2-55 13 6.2-56 30 6.2-57 21 6.2-58 23 6.2-59 21 6.2-60 13 6.2-61 16 6.2-62 13 6.2-63 13 6.2-64 19 6.2-65 31 6.2-66 13 6.2-67 13 6.2-68 13 6.2-69 13 6.2-70 26 6.2-71 13 6.2-72 13 6.2-73 13 6.2-74 13 6.2-75 13 6.2-76 23 6.2-77 13 6.2-78 18 6.2-79 13 6.2-80 13 6.2-81 13 6.2-82 24 6.2-83 13 6.2-84 28 6.2-85 28 6.2-86 28 6.2-87 30 6.2-88 16 6.2-89 20 6.2-90 29 6.2-91 29 6.2-92 20 6.2-93 21 6.2-94 20 6.2-95 26 6.2-96 26 EPL 38

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6.2-97 20 6.2-98 20 6.2-99 20 6.2-100 20 6.2-101 20 6.2-102 30 Table 6.2.1-1 (Sheet 1) 18 Table 6.2.1-1 (Sheet 2) 13 Table 6.2.1-1 (Sheet 3) 13 Table 6.2.1-1 (Sheet 4) 13 Table 6.2.1-1 (Sheet 5) 16 Table 6.2.1-1 (Sheet 6) 14 Table 6.2.1-1 (Sheet 7) 26 Table 6.2.1-1 (Sheet 8) 15 Table 6.2.1-2 13 Table 6.2.1-3 13 Table 6.2.1-4 13 Table 6.2.1-5 (Sheet 1) 13 Table 6.2.1-5 (Sheet 2) 13 Table 6.2.1-6 14 Table 6.2.1-7 14 Table 6.2.1-8 18 Table 6.2.1-9 18 Table 6.2.1-10 18 Table 6.2.1-12 13 Table 6.2.1-13 14 Table 6.2.1-15 14 Table 6.2.1-16 14 Table 6.2.1-18 14 Table 6.2.1-21 18 Table 6.2.1-22 14 Table 6.2.1-23 14 Table 6.2.1-24 24 Table 6.2.1-25 13 Table 6.2.1-26 (Sheet 1) 13 Table 6.2.1-26 (Sheet 2) 13 Table 6.2.1-27 24 Table 6.2.1-28 (Sheet 1) 13 Table 6.2.1-28 (Sheet 2) 13 Table 6.2.1-28 (Sheet 3) 13 Table 6.2.1-29 (Sheet 1) 13 Table 6.2.1-29 (Sheet 2) 13 Table 6.2.1-29 (Sheet 3) 13 Table 6.2.1-29 (Sheet 4) 13 Table 6.2.1-29 (Sheet 5) 13 Table 6.2-1-30 (Sheet 1) 13 Table 6.2.1-30 (Sheet 2) 13 Table 6.2.1-31 (Sheet 1) 17 Table 6.2.1-31 (Sheet 2) 17 Table 6.2.1-32 (Sheet 1) 18 Table 6.2.1-32 (Sheet 2) 18 EPL 39

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 6.2.1-32 (Sheet 3) 21 Table 6.2.1-33 24 Table 6.2.1-34 13 Table 6.2.1-35 13 Table 6.2.1-36 17 Table 6.2.1-37 13 Table 6.2.1-38 13 Table 6.2.1-39 (Sheet 1) 13 Table 6.2.1-39 (Sheet 2) 13 Table 6.2.1-40 25 Table 6.2.1-41 (Sheet 1) 13 Table 6.2.1-41 (Sheet 2) 13 Table 6.2.1-42 (Sheet 1) 13 Table 6.2.1-42 (Sheet 2) 13 Table 6.2.1-43 13 Table 6.2.1-44 13 Table 6.2.2-1 15 Table 6.2.2-2 21 Table 6.2.2-3 13 Table 6.2.3-1 (Sheet 1) 29 Table 6.2.3-1 (Sheet 2) 29 Table 6.2.3-2 (Sheet 1) 29 Table 6.2.3-2 (Sheet 2) 29 Table 6.2.4-1 (Sheet 1) 29 Table 6.2.4-1 (Sheet 2) 26 Table 6.2.4-1 (Sheet 3) 23 Table 6.2.4-1 (Sheet 4) 23 Table 6.2.4-1 (Sheet 5) 24 Table 6.2.4-1 (Sheet 6) 24 Table 6.2.4-1 (Sheet 7) 26 Table 6.2.4-1 (Sheet 8) 24 Table 6.2.4-1 (Sheet 9) 24 Table 6.2.4-1 (Sheet 10) 24 Table 6.2.4-1 (Sheet 11) 24 Table 6.2.4-1 (Sheet 12) 24 Table 6.2.4-1 (Sheet 13) 30 Table 6.2.4-1 (Sheet 14) 24 Table 6.2.4-1 (Sheet 15) 31 Table 6.2.4-1 (Sheet 16) 31 Table 6.2.6-1 26 Figure 6.2.1-1 13 Figure 6.2.1-2 Original Figure 6.2.1-3 13 Figure 6.2.1-4 13 Figure 6.2.1-5 13 Figure 6.2.1-6 13 Figure 6.2.1-7 13 Figure 6.2.1-8 Original Figure 6.2.1-9 Original Figure 6.2.1-10 Original Figure 6.2.1-11 Original EPL 40

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 6.2.1-12 Original Figure 6.2.1-13 Original Figure 6.2.1-14 Original Figure 6.2.1-15 18 Figure 6.2.1-16 18 Figure 6.2.1-17 18 Figure 6.2.1-18 18 Figure 6.2.1-19 18 Figure 6.2.1-22 13 Figure 6.2.1-23 13 Figure 6.2.1-24 Original Figure 6.2.1-25 Original Figure 6.2.1-26 Original Figure 6.2.1-27 Original Figure 6.2.1-28 Original Figure 6.2.1-29 Original Figure 6.2.1-30 Original Figure 6.2.1-31 Original Figure 6.2.1-32 Original Figure 6.2.1-33 Original Figure 6.2.1-34 Original Figure 6.2.1-41 Original Figure 6.2.1-42 Original Figure 6.2.1-43 18 Figure 6.2.1-44 18 Figure 6.2.1-45 18 Figure 6.2.1-46 18 Figure 6.2.1-47 18 Figure 6.2.1-48 18 Figure 6.2.1-49 18 Figure 6.2.1-50 18 Figure 6.2.1-51 18 Figure 6.2.1-52 18 Figure 6.2.1-53 18 Figure 6.2.1-54 18 Figure 6.2.1-55 18 Figure 6.2.1-56 18 Figure 6.2.1-57 18 Figure 6.2.1-58 18 Figure 6.2.1-59 18 Figure 6.2.1-60 18 Figure 6.2.1-61 Original Figure 6.2.1-62 Original Figure 6.2.1-63 Original Figure 6.2.1-63A 13 Figure 6.2.1-63B 13 Figure 6.2.1-63C 13 Figure 6.2.1-63D 13 Figure 6.2.1-63E 13 Figure 6.2.1-63F 13 Figure 6.2.1-64 8 EPL 41

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 6.2.1-65 8 Figure 6.2.1-66 8 Figure 6.2.1-67 8 Figure 6.2.1-68 8 Figure 6.2.1-69 8 Figure 6.2.1-70 Original Figure 6.2.1-71 Original Figure 6.2.1-72 Original Figure 6.2.1-73 Original Figure 6.2.1-74 Original Figure 6.2.1-75 Original Figure 6.2.1-76 Original Figure 6.2.1-77 Original Figure 6.2.1-78 Original Figure 6.2.1-79 Original Figure 6.2.1-80 Original Figure 6.2.1-81 Original Figure 6.2.1-82 Original Figure 6.2.1-83 14 Figure 6.2.2-1 15 Figure 6.2.2-2 22 Figure 6.2.2-3 15 Figure 6.2.3-1 13 Figure 6.2.4-1 31 Figure 6.2.5A-l 1 Figure 6.2.5A-2 1 Figure 6.2.5A-3 1 Figure 6.2.5A-4 1 Figure 6.2.5A-5 1 Figure 6.2.5B-1 20 Figure 6.2.6-1 26 Figure 6.2.6-2 26 Figure 6.2.6-3 26 6.3-1 13 6.3-2 13 6.3-3 13 6.3-4 27 6.3-5 27 6.3-6 29 6.3-7 26 6.3-8 26 6.3-9 17 6.3-10 21 6.3-11 13 6.3-12 13 6.3-13 23 6.3-14 13 6.3-15 13 6.3-16 29 6.3-17 13 6.3-18 23 EPL 42

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6.3-19 23 6.3-20 23 6.3-21 23 6.3-22 24 6.3-23 24 6.3-24 24 6.3-25 23 6.3-26 27 Table 6.3.2-1 (Sheet 1) 17 Table 6.3.2-1 (Sheet 2) 26 Table 6.3.2-1 (Sheet 3) 29 Table 6.3.2-2 16 Table 6.3.2-3 26 Table 6.3.2-4 (Sheet 1) 21 Table 6.3.2-4 (Sheet 2) 21 Table 6.3.2-4 (Sheet 3) 13 Table 6.3.2-5 27 Table 6.3.2-6 13 Table 6.3.2-7 (Sheet 1) 26 Table 6.3.2-7 (Sheet 2) 13 Table 6.3.2-7 (Sheet 3) 13 Table 6.3.2-8 13 Table 6.3.3-1 13 Figure 6.3.2-1 29 Figure 6.3.2-2 22 Figure 6.3.2-3 13 Figure 6.3.2-4 21 Figure 6.3.2-5 21 Figure 6.3.2-6 21 Figure 6.3.2-7 21 6.4-1 13 6.4-2 19 6.4-3 13 6.4-4 13 6.4-5 16 6.4-6 16 6.5-1 15 6.5-2 13 6.5-3 15 6.5-4 18 6.5-5 24 6.5-6 15 6.5-7 13 6.5-8 26 6.5-9 13 6.5-10 13 6.5-11 13 6.5-12 13 6.5-13 13 6.5-14 13 6.5-15 26 EPL 43

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6.5-16 13 6.5-17 15 6.5-18 13 6.5-19 13 6.5-20 15 6.5-21 26 6.5-22 13 6.5-23 14 6.5-24 31 6.5-25 31 6.5-26 30 6.5-27 13 6.5-28 25 6.5-29 31 6.5-30 13 6.5-31 30 6.5-32 13 6.5-33 13 6.5-34 31 6.5-35 13 6.5-36 13 6.5-37 22 6.5-38 22 6.5-39 13 6.5-40 13 6.5-41 26 6.5-42 13 6.5-43 13 6.5-44 13 6.5-45 13 6.5-46 13 6.5-47 13 6.5-48 13 6.5-49 13 6.5-50 13 6.5-51 13 6.5-52 13 6.5-53 13 6.5-54 13 6.5-55 13 6.5-56 17 6.5-57 13 6.5-58 13 6.5-59 13 6.5-60 13 6.5-61 26 6.5-62 13 6.5-63 21 6.5-64 26 6.5-65 13 6.5-66 21 EPL 44

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 6.5-67 21 6.5-68 25 6.5-69 21 6.5-70 13 6.5-71 15 6.5-72 13 6.5-73 17 Table 6.5.2-1 (Sheet 1) 13 Table 6.5.2-1 (Sheet 2) 13 Table 6.5.2-2 13 Table 6.5.2-3 13 Table 6.5.3-1 13 Table 6.5.3-2 13 Table 6.5.3-3 13 Table 6.5.3-4 13 Table 6.5.3-5 13 Table 6.5.3-6 13 Table 6.5.3-7 13 Table 6.5.4-1 13 Table 6.5.4-2 13 Table 6.5.4-3 25 Table 6.5.4-4 15 Table 6.5.4-5 13 Table 6.5.4-6 13 Table 6.5.4-7 13 Table 6.5.4-8 13 Table 6.5.4-9 13 Table 6.5.4-10 13 Table 6.5.4-11 13 Table 6.5.4-12 13 Table 6.5.4-13 15 Table 6.5.4-14 15 Table 6.5.4-15 15 Table 6.5.4-16 15 Table 6.5.5-1 13 Table 6.5.6-1 (Sheet 1) 31 Table 6.5.6-1 (Sheet 2) 17 Table 6.5.9-1 13 Table 6.5.10-1 (Sheet 1) 13 Table 6.5.10-1 (Sheet 2) 13 Table 6.5.10-2 (Sheet 1) 13 Table 6.5.10-2 (Sheet 2) 13 Table 6.5.10-2 (Sheet 3) 13 Table 6.5.10-2 (Sheet 4) 13 Table 6.5.10-2 (Sheet 5) 13 Table 6.5.10-2 (Sheet 6) 13 Table 6.5.10-2 (Sheet 7) 13 Table 6.5.10-2 (Sheet 8) 13 Table 6.5.10-2 (Sheet 9) 13 Table 6.5.10-2 (Sheet 10) 13 Table 6.5.10-2 (Sheet 11) 13 EPL 45

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 6.5.10-2 (Sheet 12) 13 Table 6.5.10-2 (Sheet 13) 13 Table 6.5.10-2 (Sheet 14) 13 Table 6.5.11-1 13 Table 6.5.11-2 13 Table 6.5.12-1 13 Table 6.5.15-1 Original Figure 6.5.1-1 Original Figure 6.5.1-2A 13 Figure 6.5.1-2B 13 Figure 6.5.1-3 Original Figure 6.5.2-1 Original Figure 6.5.2-2 Original Figure 6.5.2-3 Original Figure 6.5.3-1 6 Figure 6.5.3-2 Original Figure 6.5.3-3 Original Figure 6.5.3-4 Original Figure 6.5.3-5 Original Figure 6.5.4-1 Original Figure 6.5.4-2 Original Figure 6.5.6-1 13 Figure 6.5.7-1 Original Figure 6.5.9-1 Original Figure 6.5.9-2 13 Figure 6.5.9-3 Original Figure 6.5.9-4 Original Figure 6.5.9-5 Original Figure 6.5.9-6 Original Figure 6.5.9-7 6 Figure 6.5.10-1 Original Figure 6.5.10-2 Original Figure 6.5.10-3 Original Figure 6.5.10-4 Original Figure 6.5.10-5 Original Figure 6.5.10-6 Original Figure 6.5.10-7 Original Figure 6.5.10-8 Original Figure 6.5.10-9 Original Figure 6.5.10-10 Original Figure 6.5.10-11 Original Figure 6.5.10-12 Original Figure 6.5.10-13 Original Figure 6.5.10-14 Original Figure 6.5.10-15 Original Figure 6.5.10-16 Original Figure 6.5.11-1 6 Figure 6.5.12-1 Original Figure 6.5.12-2 6 Figure 6.5.13-1 Original Figure 6.5.15-1 15 EPL 46

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 6.5.15-2 Original Figure 6.5.15-3 Original Figure 6.5.15-4 Original 6.6-1 13 6.6-2 13 6.6-3 13 6.7-1 13 6.8-1 20 6.9-1 23 6.9-2 16 6A-1 13 6A-2 13 6A-3 13 6A-4 21 Figure 6A-1 13 CHAPTER 7 7-1 30 7-2 28 7-3 28 7-4 28 7-5 13 7-6 13 7.1-1 28 7.1-2 13 7.1-3 13 7.1-4 13 7.1-5 28 7.1-6 13 7.1-7 30 7.1-8 30 7.1-9 26 7.1-10 26 7.1.10a 13 7.1-11 13 7.1-12 30 7.1-13 13 7.1-14 30 7.1-15 30 7.1-16 30 7.1-17 28 7.1-18 28 7.1-19 28 7.1-20 28 7.1-21 30 7.1-22 30 Figure 7.1.3-1 Original Figure 7.1.3-2 Original Figure 7.1.3-3 Original Figure 7.1.3-4 Original Figure 7.1.3-5 Original Figure 7.1.4-1 28 EPL 47

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 7.2-1 13 7.2-2 13 7.2-3 31 7.2-4 13 7.2-5 13 7.2-6 13 7.2-7 13 7.2-8 13 7.2-9 30 7.2-10 13 7.2-11 13 7.2-12 13 7.2-13 16 7.2-14 30 7.2-15 31 7.2-16 24 7.2-17 13 7.2-18 13 7.2-19 28 7.2-20 28 7.2-21 28 7.2-22 28 7.2-23 28 7.2-24 29 7.2-25 28 7.2-26 28 7.2-27 28 7.2-28 28 7.2-29 28 7.2-30 28 7.2-31 28 7.2-32 28 7.2-33 28 Table 7.2.1-1 (Sheet 1) 31 Table 7.2.1-1 (Sheet 2) 30 Table 7.2.1-2 13 Table 7.2.1-3 (Sheet 1) 31 Table 7.2.1-3 (Sheet 2) 13 Table 7.2.1-3 (Sheet 3) 13 Table 7.2.1-4 31 Table 7.2.1-5 (Sheet 1) 31 Table 7.2.1-5 (Sheet 2) 13 Figure 7.2.1-1 (Sheet 1) 13 Figure 7.2.1-1 (Sheet 2) 13 Figure 7.2.1-1 (Sheet 3) 31 Figure 7.2.1-1 (Sheet 4) 13 Figure 7.2.1-1 (Sheet 5) 13 Figure 7.2.1-1 (Sheet 6) 13 Figure 7.2.1-1 (Sheet 7) 13 Figure 7.2.1-1 (Sheet 8) 13 Figure 7.2.1-1 (Sheet 9) 29 EPL 48

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 7.2.1-1 (Sheet 10) 28 Figure 7.2.1-1 (Sheet 11) 28 Figure 7.2.1-1 (Sheet 12) 28 Figure 7.2.1-1 (Sheet 13) 23 Figure 7.2.1-1 (Sheet 13A) 22 Figure 7.2.1-1 (Sheet 14) 23 Figure 7.2.1-1 (Sheet 14A) 22 Figure 7.2.1-1 (Sheet 15) 13 Figure 7.2.1-1 (Sheet 16) 23 Figure 7.2.1-1 (Sheet 17) 13 Figure 7.2.1-1 (Sheet 18) 13 Figure 7.2.1-1 (Sheet 19) 13 Figure 7.2.1-1 (Sheet 20) 13 Figure 7.2.1-2 Original Figure 7.2.2-1 17 Figure 7.2.2-2 Original 7.3-1 13 7.3-2 13 7.3-3 13 7.3-4 13 7.3-5 13 7.3-6 13 7.3-7 13 7.3-8 13 7.3-9 13 7.3-10 13 7.3-11 13 7.3-12 13 7.3-13 13 7.3-14 13 7.3-15 28 7.3-16 13 Table 7.3.1-1 13 Table 7.3.1-2 13 Table 7.3.1-3 16 Table 7.3.1-4 (Sheet 1) 24 Table 7.3.1-4 (Sheet 2) 24 Table 7.3.1-4 (Sheet 3) 15 Table 7.3.1-4 (Sheet 4) 26 Table 7.3.1-4 (Sheet 5) 26 Table 7.3.2-1 13 7.4-1 13 7.4-2 20 7.4-3 13 7.4-4 13 7.4-5 13 7.5-1 13 7.5-2 13 7.5-3 13 7.5-4 15 7.5-5 15 EPL 49

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 7.5-6 15 7.5-7 30 7.5-8 13 Table 7.5-1 15 Table 7.5-2 (Sheet 1) 13 Table 7.5-2 (Sheet 2) 18 Table 7.5-2 (Sheet 3) 31 Table 7.5-2 (Sheet 4) 13 Table 7.5-2 (Sheet 5) 21 Table 7.5-2 (Sheet 6) 13 Table 7.5-2 (Sheet 7) 15 Table 7.5-2 (Sheet 8) 13 Table 7.5-2 (Sheet 9) 13 Table 7.5-2 (Sheet 10) 13 Table 7.5-2 (Sheet 11) 13 7.6-1 17 7.6-2 13 7.6-3 13 7.6-4 13 7.6-5 26 7.6-6 28 7.6-7 28 7.6-8 13 7.6-9 26 Figure 7.6.6-1 24 Figure 7.6.7-1 28 Figure 7.6.9-1 (Sheet 1) 6 Figure 7.6.9-1 (Sheet 2) Original Figure 7.6.9-1 (Sheet 3) 6 7.7-1 28 7.7-2 26 7.7-3 28 7.7-4 28 7.7-5 13 7.7-6 28 7.7-7 28 7.7-8 28 7.7-9 28 7.7-10 28 7.7-11 28 7.7-12 28 7.7-13 28 7.7-14 28 7.7-15 28 7.7-16 28 7.7-17 28 7.7-18 30 7.7-19 28 7.7-20 28 7.7-21 28 7.7-22 28 EPL 50

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 7.7.1-1 30 Figure 7.7.1-1 28 Figure 7.7.1-2 28 Figure 7.7.1-3 Original Figure 7.7.1-4 28 Figure 7.7.1-5 13 7A-1 13 7A-2 28 7A-3 13 7A-4 13 7A-5 13 Figure 7A-1 Original Figure 7A-2 13 Figure 7A-3 13 Figure 7A-4 13 Figure 7A-5 13 Figure 7A-6 13 CHAPTER 8 8-1 24 8-2 20 8-3 20 8-4 13 8-5 13 8.1-1 25 8.1-2 13 8.1-3 26 8.1-4 30 8.1-5 13 8.1-6 30 Table 8.1.2-1 (Sheet 1) 13 Table 8.1.2-1 (Sheet 2) 13 Figure 8.1.1-1 17 Figure 8.1.2-1 31 Figure 8.1.2-2 26 8.2-1 31 8.2-2 31 8.2-3 25 8.2-4 25 8.2-5 25 8.2-6 25 8.2-7 25 8.2-8 25 8.2-9 25 8.2-10 25 8.2-11 25 8.2-12 24 8.2-13 24 8.2-14 27 8.2-15 31 8.2-16 31 Table 8.2.1-1 (Sheet 1) 25 EPL 51

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 8.2.1-1 (Sheet 2) 29 Table 8.2.1-1 (Sheet 3) 13 Table 8.2.1-1 (Sheet 4) 13 Table 8.2.1-1 (Sheet 5) 13 Figure 8.2.1-1 31 Figure 8.2.1-2 31 Figure 8.2.1-3 26 Figure 8.2.1-4 26 Figure 8.2.1-5 31 Figure 8.2.1-6 13 8.3-1 13 8.3-2 13 8.3-3 21 8.3-4 29 8.3-5 29 8.3-6 29 8.3-7 29 8.3-8 29 8.3-9 29 8.3-10 29 8.3-11 30 8.3-12 29 8.3-13 29 8.3-14 29 8.3-15 29 8.3-16 29 8.3-17 30 8.3-18 29 8.3-19 29 8.3-20 31 8.3-21 31 8.3-22 29 8.3-23 29 8.3-24 29 8.3-25 29 8.3-26 29 8.3-27 29 8.3-28 30 8.3-29 29 8.3-30 29 8.3-31 29 8.3-32 29 8.3-33 29 8.3-34 29 8.3-35 29 8.3-36 29 8.3-37 29 8.3-38 29 8.3-39 29 8.3-40 29 8.3-41 29 EPL 52

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 8.3-42 29 8.3-43 29 8.3-44 29 8.3-45 29 8.3-46 29 8.3-47 29 8.3-48 29 8.3-49 29 8.3-50 29 Table 8.3.1-1 21 Table 8.3.1-2 16 Table 8.3.1-3 13 Table 8.3.1-4 25 Table 8.3.1-5 13 Table 8.3.1-6 13 Table 8.3.1-7 13 Table 8.3.1-8 13 Table 8.3.1-9 13 Table 8.3.1-11 13 Table 8.3.2-1 13 Figure 8.3.1-1 29 Figure 8.3.1-2 28 Figure 8.3.1-3 31 Figure 8.3.1-4 13 Figure 8.3.1-5 31 Figure 8.3.1-6 29 Figure 8.3.1-7 17 Figure 8.3.1-8 19 Figure 8.3.1-9 30 Figure 8.3.1-10 17 Figure 8.3.1-11 30 Figure 8.3.1-12 30 Figure 8.3.1-13 30 Figure 8.3.1-14 26 Figure 8.3.1-15 19 Figure 8.3.1-16 30 Figure 8.3.1-17 23 Figure 8.3.1-18 13 Figure 8.3.1-19 13 Figure 8.3.1-20 13 Figure 8.3.1-21 18 Figure 8.3.1-22 13 Figure 8.3.2-1 27 CHAPTER 9 9-1 19 9-2 19 9-3 19 9-4 19 9-5 19 9-6 24 9-7 25 EPL 53

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 9-8 16 9-9 16 9-10 21 9.1-1 28 9.1-2 24 9.1-3 18 9.1-4 18 9.1-5 18 9.1-6 18 9.1-7 18 9.1-8 20 9.1-9 18 9.1-10 18 9.1-11 18 9.1-12 18 9.1-13 18 9.1-14 31 9.1-15 19 9.1-16 19 9.1-17 31 9.1-18 18 9.1-19 31 9.1-20 18 9.1-21 30 9.1-22 31 9.1-23 27 9.1-24 27 9.1-25 28 9.1-26 27 Table 9.1.3-1 (Sheet 1) 13 Table 9.1.3-1 (Sheet 2) 13 Table 9.1.3-2 (Sheet 1) 20 Table 9.1.3-2 (Sheet 2) 13 Table 9.1.3-4 18 Table 9.1.5-1 19 Table 9.1.5-2 27 Figure 9.1.1-1 13 Figure 9.1.1-2 21 Figure 9.1.2-1 Original Figure 9.1.2-2 13 Figure 9.1.3-1 13 Figure 9.1.3-2 13 Figure 9.1.4-1 15 Figure 9.1.4-3 Original Figure 9.1.4-4 Original Figure 9.1.4-5 13 Figure 9.1.4-6 13 Figure 9.1.4-7 18 Figure 9.1.4-8 Original 9.2-1 21 9.2-2 25 EPL 54

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 9.2-3 25 9.2-4 19 9.2-5 13 9.2-6 13 9.2-7 25 9.2-8 25 9.2-9 27 9.2-10 27 9.2-11 27 9.2-12 27 9.2-13 27 9.2-14 29 9.2-15 27 9.2-16 27 9.2-17 27 9.2-18 27 9.2-19 27 9.2-20 27 9.2-21 27 9.2-22 27 9.2-23 27 9.2-24 27 9.2-25 27 9.2-26 27 9.2-27 27 9.2-28 27 9.2-29 29 Table 9.2.1-1 (Sheet 1) 13 Table 9.2.1-1 (Sheet 2) 13 Table 9.2.1-2 (Sheet 1) 13 Table 9.2.1-2 (Sheet 2) 13 Table 9.2.2-1 13 Table 9.2.2-2 13 Table 9.2.7-1 13 Figure 9.2.1-1 17 Figure 9.2.1-2 25 Figure 9.2.1-3 25 Figure 9.2.1-4 22 Figure 9.2.1-5 13 Figure 9.2.1-6 13 Figure 9.2.1-7 13 Figure 9.2.2-1 24 Figure 9.2.2-2 24 Figure 9.2.2-3 18 Figure 9.2.2-3a 16 Figure 9.2.2-4 30 Figure 9.2.2-4a 30 Figure 9.2.2-5 24 Figure 9.2.2-6 13 Figure 9.2.3-1 13 Figure 9.2.3-2 16 EPL 55

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 9.2.7-1 31 Figure 9.2.7-2 31 Figure 9.2.7-3 31 Figure 9.2.7-4 27 9.3-1 16 9.3-2 19 9.3-2a 19 9.3-3 26 9.3-4 16 9.3-5 25 9.3-6 16 9.3-7 19 9.3-8 13 9.3-9 13 9.3-10 19 9.3-11 17 9.3-12 13 9.3-13 31 9.3-14 19 9.3-15 13 9.3-16 29 9.3-17 29 9.3-18 29 9.3-19 13 9.3-20 29 9.3-21 29 9.3-22 17 9.3-23 13 9.3-24 13 9.3-25 29 9.3-26 29 9.3-27 13 9.3-28 27 9.3-29 17 9.3-30 13 9.3-31 28 9.3-32 13 9.3-33 13 9.3-34 13 Table 9.3.1-1 27 Table 9.3.2-1 (Sheet 1) 17 Table 9.3.2-1 (Sheet 2) 16 Table 9.3.2-1 (Sheet 3) 16 Table 9.3.2-1 (Sheet 4) 13 Table 9.3.4-1 19 Table 9.3.4-2 (Sheet 1) 13 Table 9.3.4-2 (Sheet 2) 13 Table 9.3.4-2 (Sheet 3) 13 Table 9.3.4-2 (Sheet 4) 19 Table 9.3.4-2 (Sheet 5) 19 Table 9.3.4-2 (Sheet 6) 13 EPL 56

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 9.3.1-1 17 Figure 9.3.1-2 13 Figure 9.3.1-3 17 Figure 9.3.3-1 13 Figure 9.3.3-2 13 Figure 9.3.4-1 31 Figure 9.3.4-2 31 Figure 9.3.4-3 17 Figure 9.3.4-4 15 Figure 9.3.4-5 26 Figure 9.3.4-6 13 Figure 9.3.6-1 13 9.4-1 16 9.4-2 13 9.4-3 16 9.4-4 13 9.4-5 13 9.4-6 16 9.4-7 13 9.4-8 13 9.4-9 16 9.4-10 30 9.4-11 13 9.4-12 14 9.4-12a 14 9.4-13 27 9.4-14 27 9.4-15 13 9.4-16 13 9.4-17 13 9.4-18 19 9.4-19 26 9.4-20 21 9.4-21 15 9.4-22 13 9.4-23 19 9.4-24 22 9.4-25 18 9.4-26 21 9.4-27 18 9.4-28 28 9.4-29 18 9.4-30 25 Table 9.4.1-1 (Sheet 1) 16 Table 9.4.1-1 (Sheet 2) 16 Table 9.4.7-1 13 Figure 9.4.1-1 25 Figure 9.4.2-1 13 Figure 9.4.2-2 13 Figure 9.4.2-2a 13 Figure 9.4.2-3 17 EPL 57

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 9.4.2-4 16 Figure 9.4.2-5 16 Figure 9.4.5-1 30 Figure 9.4.7-1 18 Figure 9.4.10-1 13 9.5-1 26 9.5-2 13 9.5-3 26 9.5-4 26 9.5-5 26 9.5-6 26 9.5-7 13 9.5-8 26 9.5-9 13 9.5-10 13 9.5-11 24 9.5-12 13 9.5-13 13 9.5-14 17 9.5-15 24 Table 9.5.2-1 26 Table 9.5.10-1 15 Figure 9.5.4-1 26 Figure 9.5.5-1 13 Figure 9.5.6-1 29 CHAPTER 10 10-1 13 10-2 13 10-3 13 10-4 13 10-5 17 10.1-1 17 Table 10.1-1 (Sheet 1) 28 Table 10.1-1 (Sheet 2) 13 Table 10.1-1 (Sheet 3) 25 Table 10.1-1 (Sheet 4) 13 Table 10.1-1 (Sheet 5) 26 Figure 10.1-1 13 Figure 10.1-2 18 Figure 10.1-3 24 10.2-1 24 10.2-2 30 10.2-3 13 10.2-4 19 10.2-5 19 10.2-6 18 10.2-7 13 10.2-8 13 10.2-9 25 10.2-10 13 10.2-11 13 EPL 58

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 10.2-12 13 10.2-13 13 10.2-14 13 10.2-15 13 10.2-16 18 Table 10.2.3-1 13 Table 10.2.3-2 13 Table 10.2.3-3 (Sheet 1) 13 Table 10.2.3-3 (Sheet 2) 13 Table 10.2.3-4 13 Table 10.2.3-5 13 Figure 10.2.3-1 Original Figure 10.2.3-2 Original Figure 10.2.3-3 Original Figure 10.2.3-4 6 10.3-1 26 10.3-2 18 10.3-3 24 10.3-4 18 10.3-5 24 10.3-6 18 10.3-7 29 Table 10.3.2-1 14 Table 10.3.3-1 14 Figure 10.3.2-1 28 10.4-1 13 10.4-2 29 10.4-3 13 10.4-4 13 10.4-5 13 10.4-6 13 10.4-7 28 10.4-7a 28 10.4-8 13 10.4-9 21 10.4-10 27 10.4-11 13 10.4-12 13 10.4-13 16 10.4-14 15 10.4-15 17 10.4-16 13 10.4-17 23 10.4-18 22 10.4-19 22 10.4-20 22 10.4-21 13 10.4-22 19 10.4-23 14 10.4-24 13 10.4-25 13 EPL 59

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 10.4-26 26 10.4-27 24 10.4-28 14 10.4-29 25 10.4-30 31 10.4-31 31 10.4-32 31 10.4-33 13 10.4-34 17 10.4-35 25 10.4-36 25 10.4-37 26 10.4-38 25 10.4-39 24 10.4-40 13 10.4-41 20 10.4-42 20 10.4-43 13 10.4-44 13 10.4-45 25 10.4-46 31 10.4-47 28 10.4-48 19 10.4-49 19 Table 10.4.1-1 13 Table 10.4.7-1 26 Table 10.4.7-2 14 Table 10.4.7-3 13 Table 10.4.7-4 (Sheet 1) 13 Table 10.4.7-4 (Sheet 2) 25 Table 10.4.7-5 13 Table 10.4.7-6 (Sheet 1) 14 Table 10.4.7-6 (Sheet 2) 14 Table 10.4.7-6 (Sheet 3) 14 Table 10.4.7-6 (Sheet 4) 14 Table 10.4.8-1 (Sheet 1) 14 Table 10.4.8-1 (Sheet 2) 14 Figure 10.4.2-1 17 Figure 10.4.5-1 24 Figure 10.4.5-2 23 Figure 10.4.7-1 29 Figure 10.4.7-2 28 Figure 10.4.7-3 31 Figure 10.4.7-4 19 Figure 10.4.7-5 28 Figure 10.4.7-6 13 Figure 10.4.7-7 13 Figure 10.4.8-1 15 Figure 10.4.8-2 13 Figure 10.4.8-3 13 Figure 10.4.9-1 28 EPL 60

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 10.4.9-2 28 CHAPTER 11 11-1 13 11-2 20 11-3 13 11-4 15 11-5 13 11.1-1 25 11.1-2 25 11.1-3 13 11.1-4 13 Table 11.1.1-1 (Sheet 1) 13 Table 11.1.1-1 (Sheet 2) 13 Table 11.1.1-2 (Sheet 1) 13 Table 11.1.1-2 (Sheet 2) 13 Table 11.1.1-3 13 Table 11.1.1-4 13 Table 11.1.1-5 13 Table 11.1.2-1 13 Table 11.1.2-2 (Sheet 1) 13 Table 11.1.2-2 (Sheet 2) 13 Table 11.1.2-2 (Sheet 3) 13 Table 11.1.2-3 (Sheet 1) 13 Table 11.1.2-3 (Sheet 2) 13 Table 11.1.2-4 (Sheet 1) 13 Table 11.1.2-4 (Sheet 2) 13 Table 11.1.2-4 (Sheet 3) 13 11.2-1 17 11.2-2 22 11.2-3 17 11.2-4 31 11.2-5 17 11.2-6 17 11.2-7 19 11.2-8 17 11.2-9 22 11.2-10 31 11.2-11 19 11.2-12 13 11.2-13 17 11.2-14 27 11.2-15 27 11.2-16 17 11.2-17 16 11.2-18 16 11.2-19 13 11.2-20 13 Table 11.2.2-1 (Sheet 1) 17 Table 11.2.2-1 (Sheet 2) 13 Table 11.2.2-2 13 Table 11.2.3-1 (Sheet 1) 13 EPL 61

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 11.2.3-1 (Sheet 2) 13 Table 11.2.3-1 (Sheet 3) 17 Table 11.2.3-1 (Sheet 4) 28 Table 11.2.3-1 (Sheet 5) 13 Table 11.2.4-1 13 Table 11.2.9-1 13 Table 11.2.9-2 13 Figure 11.2.2-1 28 Figure 11.2.2-2 27 11.3-1 13 11.3-2 14 11.3-3 14 11.3-4 13 11.3-5 13 11.3-6 19 11.3-7 17 11.3-8 13 11.3-9 13 11.3-10 15 Table 11.3.2-1 (Sheet 1) 13 Table 11.3.2-1 (Sheet 2) 14 Table 11.3.2-1 (Sheet 3) 13 Table 11.3.2-1 (Sheet 4) 14 Table 11.3.2-2 15 Table 11.3.3-1 13 Table 11.3.6-1 15 Table 11.3.6-2 (Sheet 1) 15 Table 11.3.6-2 (Sheet 2) 15 Table 11.3.9-1 13 Table 11.3.9-2 13 Table 11.3.9-3 13 Table 11.3.9-4 15 Table 11.3.9-5 15 Figure 11.3.2-1 16 Figure 11.3.2-2 21 11.4-1 13 11.4-2 23 11.4-3 19 11.4-4 19 11.4-5 23 11.4-6 17 11.4-7 20 11.4-8 20 11.4-9 16 11.4-10 28 Table 11.4.2-1 (Sheet 1) 23 Table 11.4.2-1 (Sheet 2) 19 Table 11.4.2-2 (Sheet 1) 31 Table 11.4.2-2 (Sheet 2) 16 Table 11.4.2-2 (Sheet 3) 16 Table 11.4.2-3 (Sheet 1) 17 EPL 62

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 11.4.2-3 (Sheet 2) 16 Table 11.4.2-3 (Sheet 3) 15 Table 11.4.2-3 (Sheet 4) 16 Table 11.4.2-3 (Sheet 5) 19 Table 11.4.2-3 (Sheet 6) 19 11.5-1 17 11.5-2 17 11.5-3 13 11.5-4 25 11.5-5 16 Table 11.5.2-1 13 11.6-1 13 11.6-2 13 11.6-3 13 11.6-4 13 11.6-5 17 Table 11.6.4-1 13 11A-1 13 11A-2 17 11A-3 17 11A-4 17 Table 11A-1 21 Table 11A-2 21 CHAPTER 12 12-1 13 12-2 13 12-3 16 12.1-1 29 12.1-2 18 12.1-3 13 12.1-4 18 12.1-5 13 12.1-6 13 12.1-7 13 12.1-8 27 12.1-9 27 12.1-10 23 12.1-11 23 12.1-12 23 12.1-13 23 12.1-14 23 12.1-15 24 Table 12.1.2-1 26 Table 12.1.4-1 20 Table 12.1.6-1 13 12.2-1 13 12.2-2 13 12.2-3 13 12.2-4 16 12.2-5 26 12.2-6 13 EPL 63

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 12.2.4-1 25 Table 12.2.4-2 25 Table 12.2.4-3 26 Table 12.2.6-1 13 12.3-1 26 12.3-2 24 12.4-1 13 Table 12.4.2-1 13 CHAPTER 13 13-1 13 13-2 27 13.1-1 13 13.1-2 28 13.1-3 26 13.2-1 31 13.3-1 13 13.4-1 15 13.5-1 28 13.5-2 16 13.5-3 13 13.5-4 13 13.5-5 13 Figure 13.5.1-1 13 13.6-1 19 13.6-2 19 13.7-1 29 13.8-1 15 13.9-1 26 13.9-2 29 13.9-3 29 13.9-4 27 13.9-5 27 13.9-6 27 13.9-7 27 13.9-8 29 13.9-9 29 13.9-10 29 13.9-11 29 13.9-12 29 13.9-13 31 13.9-14 29 13.9-15 28 13.9-16 29 13.9-17 29 13.9-18 29 13.9-19 28 13.9-20 29 13.9-21 29 13.9-22 31 13.9-23 29 13.9-24 29 EPL 64

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 13.9-25 28 13.9-26 29 13.9-27 29 13.9-28 29 13.9-29 28 13.9-30 28 13.9-31 30 13.9-32 29 13.9-33 29 13.9-34 28 13.9-35 29 13.9-36 29 13.9-37 28 13.9-38 28 13.9-39 29 13.9-40 31 13.9-41 31 13.9-42 31 13.9-43 31 13.9-44 29 13.9-45 29 13.9-46 29 13.9-47 29 13.9-48 29 13.9-49 28 13.9-50 28 13.9-51 28 13.9-52 28 13.9-53 28 13.9-54 29 13.9-55 29 13.9-56 29 13.9-57 29 13.9-58 29 13.9-59 31 13.9-60 31 13.9-61 31 13.9-62 30 13.9-63 29 13.9-64 30 13.9-65 29 13.9-66 29 13.9-67 29 13.9-68 29 13.9-69 29 13.9-70 29 13.9-71 29 13.9-72 29 13.9-73 29 13.9-74 29 13.9-75 29 EPL 65

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 13.9-76 31 13.9-77 29 13.9-78 29 13.9-79 29 13.9-80 31 13.9-81 29 13.9-82 29 13.9-83 29 13.9-84 29 13.9-85 29 13.9-86 29 13.9-87 29 13.9-88 29 13.9-89 29 13.9-90 29 13.9-91 29 13.9-92 29 13.9-93 29 13.9-94 29 13.9-95 29 13.9-96 29 13.9-97 29 13.9-98 29 13.9-99 29 13.9-100 29 13.9-101 31 13.9-102 29 13.9-103 29 13.9-104 29 13.9-105 29 CHAPTER 14 14-1 13 14.1-1 13 14.1-2 13 14.1-3 13 14.1-4 13 14.1-5 13 14.1-6 13 14.1-7 13 14.1-8 13 14.1-9 13 14.1-10 13 14.1-11 13 14.1-12 13 14.1-13 13 Table 14.1-1 (Sheet 1) 13 Table 14.1-1 (Sheet 2) 13 Table 14.1-1 (Sheet 3) 13 Table 14.1-1 (Sheet 4) 13 Table 14.1-1 (Sheet 5) 13 Table 14.1-1 (Sheet 6) 13 EPL 66

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 14.1-1 (Sheet 7) 13 Table 14.1-1 (Sheet 8) 13 Table 14.1-1 (Sheet 9) 13 Table 14.1-1 (Sheet 10) 13 Table 14.1-1 (Sheet 11) 13 Table 14.1-1 (Sheet 12) 13 Table 14.1-1 (Sheet 13) 13 Table 14.1-1 (Sheet 14) 13 Table 14.1-1 (Sheet 15) 13 Table 14.1-1 (Sheet 16) 13 Table 14.1-1 (Sheet 17) 13 Table 14.1-1 (Sheet 18) 13 Table 14.1-1 (Sheet 19) 13 Table 14.1-1 (Sheet 20) 13 Table 14.1-1 (Sheet 21) 13 Table 14.1-1 (Sheet 22) 13 Table 14.1-1 (Sheet 23) 13 Table 14.1-1 (Sheet 24) 13 Table 14.1-1 (Sheet 25) 13 Table 14.1-1 (Sheet 26) 13 Table 14.1-1 (Sheet 27) 13 Table 14.1-1 (Sheet 28) 13 Table 14.1-1 (Sheet 29) 13 Table 14.1-1 (Sheet 30) 13 Table 14.1-1 (Sheet 31) 13 Table 14.1-1 (Sheet 32) 13 Table 14.1-1 (Sheet 33) 13 Table 14.1-1 (Sheet 34) 13 Table 14.1-1 (Sheet 35) 13 Table 14.1-1 (Sheet 36) 13 Table 14.1-1 (Sheet 37) 13 Table 14.1-1 (Sheet 38) 13 Table 14.1-1 (Sheet 39) 13 Table 14.1-1 (Sheet 40) 13 Table 14.1-1 (Sheet 41) 13 Table 14.1-1 (Sheet 42) 13 Table 14.1-1 (Sheet 43) 13 Table 14.1-1 (Sheet 44) 13 Table 14.1-1 (Sheet 45) 13 Table 14.1-1 (Sheet 46) 13 Table 14.1-1 (Sheet 47) 13 Table 14.1-1 (Sheet 48) 13 Table 14.1-1 (Sheet 49) 13 Table 14.1-1 (Sheet 50) 13 Table 14.1-1 (Sheet 51) 13 Table 14.1-1 (Sheet 52) 13 Table 14.1-1 (Sheet 53) 13 Table 14.1-1 (Sheet 54) 13 Table 14.1-1 (Sheet 55) 13 Table 14.1-1 (Sheet 56) 13 Table 14.1-1 (Sheet 57) 13 EPL 67

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 14.1-1 (Sheet 58) 13 Table 14.1-1 (Sheet 59) 13 Table 14.1-1 (Sheet 60) 13 Table 14.1-1 (Sheet 61) 13 Table 14.1-1 (Sheet 62) 13 Table 14.1-1 (Sheet 63) 13 Table 14.1-1 (Sheet 64) 13 Table 14.1-1 (Sheet 65) 13 Table 14.1-1 (Sheet 66) 13 Table 14.1-1 (Sheet 67) 13 Table 14.1-1 (Sheet 68) 13 Table 14.1-1 (Sheet 69) 13 Table 14.1-1 (Sheet 70) 13 Table 14.1-1 (Sheet 71) 13 Table 14.1-1 (Sheet 72) 13 Table 14.1-1 (Sheet 73) 13 Table 14.1-1 (Sheet 74) 13 Table 14.1-1 (Sheet 75) 13 Table 14.1-1 (Sheet 76) 13 Table 14.1-1 (Sheet 77) 13 Table 14.1-1 (Sheet 78) 13 Table 14.1-1 (Sheet 79) 13 Table 14.1-1 (Sheet 80) 13 Table 14.1-1 (Sheet 81) 13 Table 14.1-1 (Sheet 82) 13 Table 14.1-1 (Sheet 83) 13 Table 14.1-1 (Sheet 84) 13 Table 14.1-1 (Sheet 85) 13 Table 14.1-1 (Sheet 86) 13 Table 14.1-1 (Sheet 87) 13 Table 14.1-1 (Sheet 88) 13 Table 14.1-1 (Sheet 89) 13 Table 14.1-1 (Sheet 90) 13 Table 14.1-1 (Sheet 91) 13 Table 14.1-1 (Sheet 92) 13 Table 14.1-2 (Sheet 1) 13 Table 14.1-2 (Sheet 2) 13 Table 14.1-2 (Sheet 3) 13 Table 14.1-2 (Sheet 4) 13 Table 14.1-2 (Sheet 5) 13 Table 14.1-2 (Sheet 6) 13 Table 14.1-2 (Sheet 7) 13 Table 14.1-2 (Sheet 8) 13 Table 14.1-2 (Sheet 9) 13 Table 14.1-2 (Sheet 10) 13 Table 14.1-2 (Sheet 11) 13 Table 14.1-2 (Sheet 12) 13 Table 14.1-2 (Sheet 13) 13 Table 14.1-2 (Sheet 14) 13 Table 14.1-2 (Sheet 15) 13 Table 14.1-3 (Sheet 1) 13 EPL 68

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 14.1-3 (Sheet 2) 13 Table 14.1-3 (Sheet 3) 13 Table 14.1-3 (Sheet 4) 13 Table 14.1-3 (Sheet 5) 13 Table 14.1-3 (Sheet 6) 13 Table 14.1-3 (Sheet 7) 13 Table 14.1-3 (Sheet 8) 13 Table 14.1-3 (Sheet 9) 13 Table 14.1-3 (Sheet 10) 13 Table 14.1-3 (Sheet 11) 13 Table 14.1-3 (Sheet 12) 13 Table 14.1-3 (Sheet 13) 13 Table 14.1-3 (Sheet 14) 13 Figure 14.1-1 Original Figure 14.1-2 Original Figure 14.1-3 Original Figure 14.1-4 Original 14.2-1 13 14.2-2 13 14.2-3 13 Figure 14.2.2-1 13 Figure 14.2.2-2 13 CHAPTER 15 15-1 31 15-2 31 15-3 31 15-4 31 15-5 31 15-6 31 15-7 31 15-8 31 15-9 31 15-10 31 15-11 31 15-12 31 15-13 31 15-14 31 15-15 31 15-16 31 15-17 31 15-18 31 15-19 31 15-20 31 15-21 31 15-22 31 15-23 31 15-24 31 15.1-1 31 15.1-2 17 15.1-3 31 15.1-4 31 EPL 69

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 15.1-4a 31 15.1-5 31 15.1-6 13 15.1-7 13 15.1-8 13 15.1-9 13 15.1-10 31 15.1-11 31 15.1-12 31 15.1-13 31 15.1-14 31 15.1-15 31 15.1-16 31 Table 15.1.2-1 17 Table 15.1.2-2 (Sheet 1) 31 Table 15.1.2-2 (Sheet 2) 31 Table 15.1.2-2 (Sheet 3) 31 Table 15.1.3-1 (Sheet 1) 31 Table 15.1.3-1 (Sheet 2) 31 Table 15.1.7-1 13 Table 15.1.7-2 13 Figure 15.1.3-1 31 Figure 15.1.5-1 31 Figure 15.1.5-2 31 Figure 15.1.5-3 31 Figure 15.1.6-1 31 Figure 15.1.8-1 1 Figure 15.1.9-1 13 15.2-1 18 15.2-2 18 15.2-3 31 15.2-4 31 15.2-5 31 15.2-6 31 15.2-7 31 15.2-8 31 15.2-9 31 15.2-10 31 15.2-11 31 15.2-12 26 15.2-13 31 15.2-14 31 15.2-15 31 15.2-16 31 15.2-17 31 15.2-18 31 15.2-19 31 15.2-20 31 15.2-21 31 15.2-22 31 15.2-23 31 EPL 70

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 15.2-24 31 15.2-25 31 15.2-26 31 15.2-27 31 15.2-28 31 15.2-29 31 15.2-30 31 15.2-31 31 15.2-32 31 15.2-33 31 15.2-34 31 15.2-35 31 15.2-36 31 Table 15.2-1 (Sheet 1) 31 Table 15.2-1 (Sheet 2) 31 Table 15.2-1 (Sheet 3) 31 Table 15.2-1 (Sheet 4) 31 Table 15.2-1 (Sheet 5) 31 Table 15.2-1 (Sheet 6) 31 Table 15.2-1 (Sheet 7) 31 Table 15.2-1 (Sheet 8) 31 Figure 15.2.1-1 31 Figure 15.2.1-2 31 Figure 15.2.1-3 31 Figure 15.2.2-1 31 Figure 15.2.2-2 31 Figure 15.2.2-3 31 Figure 15.2.2-4 31 Figure 15.2.2-5 31 Figure 15.2.2-6 31 Figure 15.2.2-7 31 Figure 15.2.3-1 Original Figure 15.2.3-2 Original Figure 15.2.5-1 31 Figure 15.2.5-2 31 Figure 15.2.5-3 31 Figure 15.2.5-4 31 Figure 15.2.5-5 31 Figure 15.2.7-1 31 Figure 15.2.7-2 31 Figure 15.2.7-3 31 Figure 15.2.7-4 31 Figure 15.2.7-5 31 Figure 15.2.7-6 31 Figure 15.2.7-7 31 Figure 15.2.7-8 31 Figure 15.2.7-9 31 Figure 15.2.7-10 31 Figure 15.2.7-11 31 Figure 15.2.7-12 31 Figure 15.2.7-13 31 EPL 71

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 15.2.7-14 31 Figure 15.2.7-15 31 Figure 15.2.7-16 31 Figure 15.2.7-17 31 Figure 15.2.7-18 31 Figure 15.2.7-19 31 Figure 15.2.7-20 31 Figure 15.2.7-21 31 Figure 15.2.7-22 31 Figure 15.2.8-1a & Figure 15.2.8-1b 31 Figure 15.2.8-2a & Figure 15.2.8-2b 31 Figure 15.2.8-3a & Figure 15.2.8-3b 31 Figure 15.2.8-4a & Figure 15.2.8-4b 31 Figure 15.2.9-1a & Figure 15.2.9-1b 31 Figure 15.2.9-2a & Figure 15.2.9-2b 31 Figure 15.2.9-3a & Figure 15.2.9-3b 31 Figure 15.2.9-4a & Figure 15.2.9-4b 31 Figure 15.2.10-1 31 Figure 15.2.10-2 31 Figure 15.2.10-3 31 Figure 15.2.10-4 31 Figure 15.2.10-5 31 Figure 15.2.10-6 31 Figure 15.2.10-7 31 Figure 15.2.11-1 31 Figure 15.2.11-2 31 Figure 15.2.11-3 31 Figure 15.2.11-4 31 Figure 15.2.11-5 31 Figure 15.2.11-6 31 Figure 15.2.11-7 31 Figure 15.2.11-8 31 Figure 15.2.12-1 31 Figure 15.2.12-2 31 Figure 15.2.12-3 31 Figure 15.2.12-4 31 Figure 15.2.13-1 31 Figure 15.2.13-2 31 Figure 15.2.14-1 31 Figure 15.2.14-2 31 15.3-1 31 15.3-2 24 15.3-3 30 15.3-4 31 15.3-5 31 15.3-6 31 15.3-7 28 15.3-8 31 15.3-9 31 15.3-10 31 15.3-11 31 EPL 72

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 15.3-12 31 15.3-13 31 Table 15.3.1-1 24 Table 15.3.1-2 24 Table 15.3.4-1 31 Table 15.3.7-1 31 Figure 15.3.1-1 24 Figure 15.3.1-2 24 Figure 15.3.1-3 24 Figure 15.3.1-4 24 Figure 15.3.1-5 24 Figure 15.3.1-6 24 Figure 15.3.1-7 24 Figure 15.3.3-1 Original Figure 15.3.3-2 Original Figure 15.3.3-3 Original Figure 15.3.3-4 Original Figure 15.3.3-5 Original Figure 15.3.4-1 31 Figure 15.3.4-2 31 Figure 15.3.4-3 31 Figure 15.3.4-4 31 Figure 15.3.7-1a 31 Figure 15.3.7-1b 31 Figure 15.3.7-2a 31 Figure 15.3.7-2b 31 15.4-1 31 15.4-2 24 15.4-3 21 15.4-4 24 15.4-5 24 15.4-6 24 15.4-7 31 15.4-8 31 15.4-9 31 15.4-10 31 15.4-11 31 15.4-12 31 15.4-13 31 15.4-14 31 15.4-15 31 15.4-16 31 15.4-17 31 15.4-18 31 15.4-19 31 15.4-20 31 15.4-21 31 15.4-22 31 15.4-23 31 15.4-24 31 15.4-25 31 EPL 73

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 15.4-26 31 15.4-27 31 15.4-28 31 15.4-29 31 15.4-30 31 15.4-31 31 15.4-32 31 15.4-33 31 15.4-34 31 15.4-35 31 15.4-36 31 15.4-37 31 15.4-38 31 15.4-39 31 15.4-40 31 15.4-41 31 15.4-42 31 Table 15.4.1-1 (Sheet 1) 21 Table 15.4.1-1 (Sheet 2) 24 Table 15.4.1-2 24 Table 15.4.1-2a 21 Table 15.4.1-3a 24 Table 15.4.1-3b 24 Table 15.4.1-4 24 Table 15.4.1-5 21 Table 15.4.1-9 (Sheet 1) 13 Table 15.4.1-9 (Sheet 2) 15 Table 15.4.1-9 (Sheet 3) 31 Table 15.4.1-12 31 Table 15.4.1.a-13 (Sheet 1) 31 Table 15.4.1.a-13 (Sheet 2) 31 Table 15.4.1.a-14 (Sheet 1) 31 Table 15.4.1.a-14 (Sheet 2) 31 Table 15.4.1.a-15 31 Table 15.4.1.a-16 31 Table 15.4.1.a-17 31 Table 15.4.1.a-18 31 Table 15.4.1.a-19 31 Table 15.4.1.a-20 31 Table 15.4.1.a-21 31 Table 15.4.2-1 31 Table 15.4.4-1 31 Table 15.4.6-1 31 Figure 15.4.1-1 24 Figure 15.4.1-2 24 Figure 15.4.1-3 24 Figure 15.4.1-4 24 Figure 15.4.1-6 24 Figure 15.4.1-7 24 Figure 15.4.1-9 24 Figure 15.4.1-10 24 EPL 74

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 15.4.1-11 24 Figure 15.4.1-14 24 Figure 15.4.1-15 24 Figure 15.4.1-16 24 Figure 15.4.1-17 24 Figure 15.4.1-18 24 Figure 15.4.1-19 24 Figure 15.4.1-20 24 Figure 15.4.1-21 24 Figure 15.4.1-22 24 Figure 15.4.1.a-23 31 Figure 15.4.1.a-24 31 Figure 15.4.1.a-25 31 Figure 15.4.1.a-26 31 Figure 15.4.1.a-27 31 Figure 15.4.1.a-28 31 Figure 15.4.1.a-29 31 Figure 15.4.1.a-30 31 Figure 15.4.1.a-31 31 Figure 15.4.1.a-32 31 Figure 15.4.1.a-33 31 Figure 15.4.1.a-34 31 Figure 15.4.1.a-35 31 Figure 15.4.1.a-36a 31 Figure 15.4.1.a-36b 31 Figure 15.4.1.a-37 31 Figure 15.4.1.a-38 31 Figure 15.4.1.a-39 31 Figure 15.4.1.a-40 31 Figure 15.4.1.a-41 31 Figure 15.4.1.a-42 31 Figure 15.4.1.a-43 31 Figure 15.4.1.a-44 31 Figure 15.4.1.a-45 31 Figure 15.4.1.a-46 31 Figure 15.4.1.a-47 31 Figure 15.4.2-1 31 Figure 15.4.2-2 31 Figure 15.4.2-3 31 Figure 15.4.2-4 31 Figure 15.4.2-5 31 Figure 15.4.2-6 31 Figure 15.4.2-7 31 Figure 15.4.2-8 31 Figure 15.4.2-9 31 Figure 15.4.2-10 31 Figure 15.4.2-11 31 Figure 15.4.2-12 31 Figure 15.4.2-13 31 Figure 15.4.2-14 31 Figure 15.4.2-15 31 EPL 75

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Figure 15.4.2-16 31 Figure 15.4.2-39a & 15.4.2-39b 31 Figure 15.4.2-39c & 15.4.2-39d 31 Figure 15.4.2-40a & 15.4.2-40b 31 Figure 15.4.2-40c & 15.4.2-40d 31 Figure 15.4.2-40e & 15.4.2-40f 31 Figure 15.4.2-41a & 15.4.2-41b 31 Figure 15.4.2-41c & 15.4.2-41d 31 Figure 15.4.2-42a & 15.4.2-42b 31 Figure 15.4.2-42c & 15.4.2-42d 31 Figure 15.4.2-42e & 15.4.2-42f 31 Figure 15.4.4-1 31 Figure 15.4.4-2 31 Figure 15.4.4-3 31 Figure 15.4.4-4 31 Figure 15.4.4-5 31 Figure 15.4.4-6 31 Figure 15.4.6-1 31 Figure 15.4.6-2 31 Figure 15.4.6-3 31 Figure 15.4.6-4 31 15.5-1 18 15.5-2 24 15.5-3 24 15.5-4 31 15.5-5 18 15.5-6 24 15.5-7 30 15.5-8 30 15.5-9 18 15.5-10 21 15.5-11 21 15.5-12 18 15.5-13 27 15.5-14 21 15.5-15 21 15.5-16 21 15.5-17 18 15.5-18 18 15.5-19 18 15.5-20 24 15.5-21 25 15.5-22 25 15.5-23 18 15.5-24 26 15.5-25 26 15.5-26 26 15.5-27 25 15.5-28 25 Table 15.5.1-1 18 Table 15.5.2-1 18 EPL 76

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment Table 15.5.2-2 13 Table 15.5.3-1 (Sheet 1) 13 Table 15.5.3-1 (Sheet 2) 30 Table 15.5.3-2 13 Table 15.5.3-3 27 Table 15.5.3-3A 30 Table 15.5.3-4 31 Table 15.5.3-5 13 Table 15.5.3-6 18 Table 15.5.3-7 31 Table 15.5.3-8 (Sheet 1) 13 Table 15.5.3-8 (Sheet 2) 13 Table 15.5.4-1 18 Table 15.5.5-1 (Sheet 1) 20 Table 15.5.5-1 (Sheet 2) 18 Table 15.5.6-1 (Sheet 1) 26 Table 15.5.6-1 (Sheet 2) 26 Table 15.5.6-1 (Sheet 3) 26 Figure 15.5.1-1 10 Figure 15.5.1-2 10 Figure 15.5.1-3 10 Figure 15.5.3-1 Original 15A-1 18 15A-2 18 15A-3 18 15A-4 18 Table 15A-1 18 Table 15A-2 18 15B-1 13 15B-2 13 15B-3 13 15B-4 13 15B-5 13 15B-6 13 15B-7 13 Table 15B-1 13 Table 15B-2 13 Table 15B-3 13 Figure 15B-1 10 Figure 15B-2 Original Figure 15B-3 10 Figure 15B-4 Original Figure 15B-5 10 Figure 15B-6 10 Figure 15B-7 10 Figure 15B-8 10 Figure 15B-9 10 Figure 15B-10 10 Figure 15B-11 10 15C-1 13 15C-2 13 EPL 77

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING Effective Page Amendment 15C-3 13 15C-4 13 15C-5 13 15C-6 13 15C-7 13 Figure 15C-1 11 15D-1 18 15D-2 26 15D-3 26 Table 15D-1 26 Table 15D-2 18 Table 15D-3 18 CHAPTER 17 17-1 13 17.1-1 18 17.2-1 19 EPL 78

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE EFFECTIVE PAGE LISTING SQN SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT UPDATE LIST OF AMENDMENTS Amendment Number Date Submitted to NRC Updated FSAR Submitted April 14, 1983 Amendment 1 April 16, 1984 Amendment 2 April 11, 1985 Amendment 3 April 11, 1986 Amendment 4 April 20, 1987 Amendment 5 April 20, 1988 Amendment 6 April 14, 1989 Amendment 7 April 27, 1990 Amendment 8 April 15, 1991 Amendment 9 April 15, 1992 Amendment 10 April 14, 1994 Amendment 11 May 12, 1995 Amendment 12 December 6, 1996 Amendment 13 March 25, 1998 Amendment 14 May 4, 1998 Amendment 15 November 9, 1999 Amendment 16 May 10, 2001 Amendment 17 November 8, 2002 Amendment 18 May 28, 2004 Amendment 19 October 13, 2005 Amendment 20 June 12, 2007 Amendment 21 December 5, 2008 Amendment 22 May 24, 2010 Amendment 23 December 14, 2011 Amendment 24 July 3, 2013 Amendment 25 October 6, 2014 Amendment 26 June 10, 2016 Amendment 27 November 28, 2017 Amendment 28 June 5, 2019 Amendment 29 October 28, 2020 Amendment 30 April 28, 2022 Amendment 31 October 12, 2023 EPL 79

SQN-26 TABLE OF CONTENTS SECTION TITLE PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.3 COMPARISON TABLES 1.3-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 1.7 LIST OF ABBREVIATIONS 1.7-1 1.8 TECHNICAL QUALIFICATIONS OF APPLICANT (HISTORICAL) 1.8-1 2.0 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1-1 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND 2.2-1 MILITARY FACILITIES 2.3 METEOROLOGY 2.3-1 2.4 HYDROLOGIC ENGINEERING 2.4-1 APPENDIX 2.4A FLOOD PROTECTION PLAN 2.4A-1 2.5 GEOLOGY AND SEISMOLOGY 2.5-1

2.6 CONCLUSION

S 2.6-1 3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1-1 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, 3.2-1 AND COMPONENTS 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.5 MISSILE PROTECTION 3.5-1 3.6 PROTECTION AGAINST EFFECTS ASSOCIATED 3.6-1 WITH THE POSTULATED RUPTURE OF PIPING 3.7 SEISMIC DESIGN 3.7-1 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 APPENDIX 3.8A SHELL TEMPERATURE TRANSIENTS 3.8A-1 APPENDIX 3.8B CONTAINMENT VESSEL PENETRATIONS 3.8B-1 APPENDIX 3.8C CONTAINMENT ANCHORAGE 3.8C-1 APPENDIX 3.8D COMPUTER PROGRAMS USED IN 3.8D-1 STRUCTURAL ANALYSIS APPENDIX 3.8E DESIGN PROCEDURE FOR REINFORCED 3.8E-1 CONCRETE BLOCK WALLS 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3.10-1 AND ELECTRICAL EQUIPMENT 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.12 CONTROL OF HEAVY LOADS 3.12-1 3.13 FLEX RESPONSE SYSTEM 3.13-1 1

SQN-24 TABLE OF CONTENTS SECTION TITLE PAGE 4.0 REACTOR 4.1

SUMMARY

DESCRIPTION 4.1-1 4.2 MECHANICAL DESIGN 4.2-1 4.3 NUCLEAR DESIGN 4.3-1 4.4 THERMAL AND HYDRAULIC DESIGN 4.4-1 4.5 REACTOR (HTP AND MARK-BW FUEL) 4.5-1 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTION 5.1-1 5.2 INTEGRITY OF THE REACTOR COOLANT SYSTEM 5.2-1 BOUNDARY 5.3 THERMAL HYDRAULIC SYSTEM DESIGN 5.3-1 5.4 REACTOR VESSEL AND APPURTENANCES 5.4-1 5.5 COMPONENT AND SUBSYSTEM DESIGN 5.5-1 5.6 INSTRUMENTATION APPLICATION 5.6-1 6.0 ENGINEERED SAFETY FEATURES 6.1 GENERAL 6.1-1 6.2 CONTAINMENT SYSTEMS 6.2-1 6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1 6.4 HABITABILITY SYSTEMS 6.4-1 6.5 ICE CONDENSER SYSTEM 6.5-1 APPENDIX 6A IODINE REMOVAL IN THE ICE CONDENSER 6A-1 SYSTEM 6.6 AIR RETURN FANS 6.6-1 6.7 AUXILIARY FEEDWATER SYSTEM 6.7-1 6.8 PUMP AND VALVE INSERVICE TESTING PROGRAM 6.8-1 6.9 MOTOR-OPERATED VALVE PROGRAM - GENERIC 6.9-1 LETTER 89-10 7.0 INSTRUMENTATION AND CONTROLS

7.1 INTRODUCTION

7.1-1 7.2 REACTOR TRIP SYSTEM 7.2-1 7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 7.3-1 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.4-1 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION FOR POST 7.5-1 ACCIDENT MONITORING (PAM) 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY 7.6-1 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 7.7-1 APPENDIX 7A INSTRUMENTATION IDENTIFICATIONS AND 7A-1 SYMBOLS 2

SQN-27 TABLE OF CONTENTS SECTION TITLE PAGE 8.0 ELECTRIC POWER

8.1 INTRODUCTION

8.1-1 8.2 OFFSITE POWER SYSTEM 8.2-1 8.3 ONSITE POWER SYSTEM 8.3-1 9.0 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.2 WATER SYSTEMS 9.2-1 9.3 PROCESS AUXILIARIES 9.3-1 9.4 HEATING, VENTILATING, AND AIR CONDITIONING 9.4-1 9.5 OTHER AUXILIARY SYSTEMS 9.5-1 10.0 STEAM AND POWER CONVERSION SYSTEM 10.1

SUMMARY

DESCRIPTION 10.1-1 10.2 TURBINE GENERATOR 10.2-1 10.3 MAIN STEAM SUPPLY SYSTEM 10.3-1 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION 10.4-1 SYSTEM 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS 11.1-1 11.2 LIQUID WASTE SYSTEMS 11.2-1 11.3 GASEOUS WASTE SYSTEMS 11.3-1 11.4 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING 11.4-1 SYSTEMS 11.5 SOLID WASTE MANAGEMENT SYSTEM 11.5-1 11.6 OFFSITE RADIOLOGICAL MOITORING PROGRAM 11.6-1 APPENDIX 11A TRITIUM CONTROL 11A-1 12.0 RADIATION PROTECTION 12.1 SHIELDING 12.1-1 12.2 VENTILATION 12.2-1 12.3 RADIOLOGICAL CONTROL PROGRAM 12.3-1 12.4 LEAKAGE REDUCTION PROGRAM 12.4-1 13.0 CONDUCT OF OPERATIONS 13.1 ORGANIZATIONAL STRUCTURE 13.1-1 13.2 TRAINING PROGRAMS 13.2-1 13.3 EMERGENCY PLANNING 13.3-1 13.4 REVIEW AND AUDIT 13.4-1 13.5 SITE INSTRUCTIONS AND SECTION INSTRUCTION LETTERS 13.5-1 13.6 PLANT RECORDS 13.6-1 13.7 NUCLEAR SECURITY 13.7-1 13.8 TECHNICAL REQUIREMENTS MANUAL 13.8-1 13.9 LICENSE RENEWAL PROGRAM 13.9-1 3

SQN-18 TABLE OF CONTENTS SECTION TITLE PAGE 14.0 INITIAL TESTS AND OPERATIONS 14.1 TEST PROGRAM 14.1-1 14.2 AUGMENTATION OF PLANT STAFF FOR INITIAL 14.2-1 TESTS AND OPERATIONS (HISTORICAL) 15.0 ACCIDENT ANALYSES 15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL 15.1-1 TRANSIENTS 15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY 15.2-1 15.3 CONDITION III - INFREQUENT FAULTS 15.3-1 15.4 CONDITION IV - LIMITING FAULTS 15.4-1 15.5 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-1 APPENDIX 15A DOSE MODELS USED TO EVALUATE THE 15A-1 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS APPENDIX 15B SENSITIVITY ANALYSIS OF LOCA DOSE 15B-1 CALCULATIONS (HISTORICAL INFORMATION) APPENDIX 15C FUEL ROD MODEL DISCUSSION (HISTORICAL INFORMATION) 15C-1 APPENDIX 15D PRIMARY COOLANT ACTIVITY MODEL USED TO EVALUATE 15D-1 THE RADIOLOGICAL CONSEQUENCES OF ACCIDENTS 16.0 TECHNICAL SPECIFICATION (SEE SEPARATE DOCUMENT) 17.0 QUALITY ASSURANCE 17.1 A) QUALITY ASSURANCE DURING DESIGN AND 17.1-1 CONSTRUCTION B) WESTINGHOUSE QUALITY ASSURANCE PROGRAM 17.2 QUALITY ASSURANCE PROGRAM FOR STATION 17.2-1 OPERATION 4

SQN-19 TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

1.1-1 1.1.1 LICENSING BASIS DOCUMENTS 1.1-1 1.1.2 PROGRAMMATIC COMMITMENTS 1.1-2 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 SITE CHARACTERISTICS 1.2-1 1.2.1.1 Location 1.2-1 1.2.1.2 Demography 1.2-1 1.2.1.3 Meteorology 1.2-1 1.2.1.4 Hydrology 1.2-1 1.2.1.5 Geology 1.2-1 1.2.1.6 Seismology 1.2-1 1.2.2 FACILITY DESCRIPTION 1.2-2 1.2.2.1 Design Criteria 1.2-2 1.2.2.2 Nuclear Steam Supply System 1.2-2 1.2.2.3 Control and Instrumentation 1.2-3 1.2.2.4 Fuel Handling System 1.2-4 1.2.2.5 Waste Processing System 1.2-4 1.2.2.6 Steam and Power Conversion System 1.2-5 1.2.2.7 Plant Electrical System 1.2-5 1.2.2.8 Cooling Water 1.2-6 1.2.2.9 Component Cooling System and Essential Raw Cooling Water 1.2-7 1.2.2.10 Chemical and Volume Control System 1.2-7 1.2.2.11 Sampling and Water Quality System 1.2-8 1.2.2.12 Ventilation 1.2-8 1.2.2.13 Fire Protection System 1.2-9 1.2.2.14 Compressed Air System 1.2-9 1.2.2.15 Engineered Safety Features 1.2-9 1.2.2.16 Shared Facilities and Equipment-Safety Related 1.2-10 1.2.3 GENERAL ARRANGEMENT OF MAJOR STRUCTURES 1.2-11 AND EQUIPMENT 1.3 COMPARISON TABLES 1.3-1 1.3.1 COMPARISONS WITH SIMILAR FACILITY 1.3-1 DESIGNS 1.3.2 COMPARISONS OF FINAL AND PRELIMINARY 1.3-1 DESIGNS 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 S1-0toc.doc 1-1

SQN-17 1.5.1 PROGRAMS REQUIRED FOR PLANT OPERATION 1.5-1 1.5.2 PROGRAMS NOT REQUIRED FOR PLANT OPERATION 1.5-3 1.5.2.1 Burnable Poison Program 1.5-3 1.5.2.2 Fuel Development Program for 1.5-4 Operation at High Power Densities 1.5.2.3 FLECHT (Full Length Emergency Core Cooling 1.5-4 Heat Transfer Test) 1.5.2.4 Loss of Coolant Analysis Program 1.5-4 1.5.2.5 Reactor Vessel Thermal Shock 1.5-4 1.5.2.6 Blowdown Forces Program 1.5-5 1.5.3 17 x 17 FUEL ASSEMBLY VERIFICATION TEST 1.5-5 1.5.3.1 Rod Cluster Control Spider Tests 1.5-6 1.5.3.2 Grid Tests 1.5-6 1.5.3.3 Fuel Assembly Structural Tests 1.5-7 1.5.3.4 Guide Tube Tests 1.5-7 1.5.3.5 Prototype Assembly Tests 1.5-7 1.5.3.6 Departure from Nucleate Boiling 1.5-8 1.5.3.7 Incore Flow Mixing 1.5-8 1.5.4 INPILE FUEL DENSIFICATION 1.5-8 1.5.5 LOCA HEAT TRANSFER TESTS (17 x 17) 1.5-9 1.5.5.1 Blowdown Heat Transfer Testing 1.5-9 1.5.5.2 Results 1.5-11 1.5.5.3 Single Rod Burst Test (SRBT) 1.5-11 1.

5.6 REFERENCES

1.5-12 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 1.6.1 TOPICAL REPORTS 1.6-1 1.7 LIST OF ABBREVIATIONS 1.7-1 1.7.1 ABBREVIATIONS OF ORGANIZATIONS 1.7-1 1.7.2 ABBREVIATIONS AND SYMBOLS 1.7-2 1.8 TECHNICAL QUALIFICATIONS OF APPLICANT (HISTORICAL) 1.8-1 S1-0toc.doc 1-2

SQN-17 LIST OF TABLES Number Title 1.3.1-1 Design Comparison (Excluding Secondary Cycle) 1.3.1-2 Design Comparison - Secondary Cycle 1.3.2-1 Major Design Changes Since Submittal of the PSAR 1.6.1-1 Westinghouse and Framatome Topical Reports Incorporated by Reference LIST OF FIGURES Number Title 1.2.3-1 Equipment Plans - Roof Plans 1.2.3-2 Equipment Plan - EL 749.0 and Above 1.2.3-3 Equipment Plan - EL 734.0 and EL 732.0 1.2.3-4 Equipment Plan - EL 714.0 and EL 706.0 1.2.3-5 Equipment Plan - EL 690.0 and EL 685.0 1.2.3-6 Equipment Plan - EL 669.0 and EL 662.5 1.2.3-7 Equipment Plan - EL 727.5, EL 706.0, EL 653.0, and EL 643.0 1.2.3-8 Equipment - Plant Section 1.2.3-9 Equipment - Auxiliary Building Section 1.2.3-10 Equipment - Turbine Building Section 1.2.3-11 Equipment - Reactor Building 1.2.3-12 Equipment - Reactor Building 1.2.3-13 Equipment - Reactor Building 1.2.3-14 Equipment - ERCW Pumping Station 1.2.3-15 Equipment - ERCW Pumping Station 1.2.3-16 Equipment - ERCW Pumping Station 1.2.3-17 Equipment - Diesel Generator Building 1.2.3-18 Equipment - CCW Pumping Station 1.2.3-19 Equipment - CCW Pumping Station S1-0toc.doc 1-3

SQN-31

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This Final Safety Analysis Report is in support of the Tennessee Valley Authority (TVA) facility operating licenses for a two-unit nuclear power plant located approximately 7.5 miles northeast of Chattanooga at the Sequoyah site in Hamilton County, Tennessee. This facility has been designated the Sequoyah Nuclear Plant (SNP). The plant has been designed, built, and is operated by TVA. Each of the two identical units employs a Pressurized Water Reactor Nuclear Steam Supply System with four coolant loops furnished by Westinghouse Electric Corporation. These units are similar to those of the Watts Bar Nuclear Plant, and other plants reviewed by the U.S. Nuclear Regulatory Commission. Each of the two reactor cores is rated at 3,455 MWt and, at this core power, each NSSS will operate at 3,467 MWt. The additional 12 MWt is due to the contribution of heat of the Primary Coolant System from nonreactor sources, primarily reactor coolant pump heat. Each of the reactor cores has an Engineered Safeguards design rating of approximately 3565 MWt and each NSSS 3577 MWt. The total generator output is approximately 1,199 MWe for the rated core power. The containment for each of the reactors consists of a freestanding steel vessel with an ice condenser and separate reinforced concrete shield building. The ice condenser was designed by the Westinghouse Electric Corporation. The freestanding containment vessel was designed by Chicago Bridge & Iron (CBI). Unit 1 began commercial operation in July 1, 1981. Unit 2 began commercial operation on June 1, 1982. 1.1.1 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 operating license. The following list provides references on the review and approval requirements for the listed documents. INCORPORATED REGULATORY OR BY REFERENCE DOCUMENT REQUIREMENT IN FSAR Updated Final Safety Analysis Report 10 CFR 50.59 N/A 10 CFR 50.71(e) Technical Requirements Manual Technical Requirement 6.0 10 CFR 50.59 Yes 10 CFR 50.36(c)(2)(ii) Technical Specification Bases 10 CFR 50.59 No Organizational Topical Report 10 CFR 50.71(e) Yes Quality Assurance Plan 10 CFR 50.54(a)(3) Yes S1-1.doc 1.1-1

SQN-31 INCORPORATED REGULATORY OR BY REFERENCE DOCUMENT REQUIREMENT IN FSAR Fire Protection Report Unit 1 License Condition 2.C.16 Yes Unit 2 License Condition 2.C.13 Yes Offsite Dose Calculation Manual Technical Specification 5.5.1 No Physical Security Plan 10 CFR 50.54(p) No Radiological Emergency Plan 10 CFR 50.54(q) No Core Operating Limits Report Technical Specification 5.6.3 No Pressure Temperature Limits Report Technical Specification 5.6.4 No Licensing Requirements for the 10 CFR 72 No Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste General License Issued 10 CFR 72.210 No Conditions of General License 10 CFR 72.212 No Issued Under 10 CFR 72.210 1.1.2 PROGRAMMATIC COMMITMENTS The following programmatic commitments are incorporated to ensure control under the licensing basis process. Technical Specification Change (TSC 04-08) TVA is using an industry database (e.g., the industry's Consolidated Data Entry [CDE] program, currently being developed and maintained by the Institute of Nuclear Power Operations) to send to the NRC the operating data (for each calendar month) that is described in Generic Letter 97-02, "Revised Contents of the Monthly Operating Report," by the last day of the month following the end of each calendar quarter. This regulatory commitment will be implemented to prevent any gaps in the monthly operating statistics and shutdown experience provided to the NRC (i.e., data for all months will be provided using one or both systems monthly operating reports and CDE). S1-1.doc 1.1-2

SQN 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site Characteristics 1.2.1.1 Location The plant site, consisting of approximately 525 acres, is located in southeastern Tennessee on the west shore of Chickamauga Lake approximately 7.5 miles northeast of Chattanooga. 1.2.1.2 Demography The population density of the area surrounding the site is relatively low. The site consists of an owner controlled exclusion area. A low population zone surrounds the plant site. 1.2.1.3 Meteorology Meteorological data has been collected since April 1971 at the site. Selected data has been used for the description of the local weather and for the calculation of the dispersion factors. In addition, data from stations within 75 miles of the site was used to calculate the regional climatology. The probability of a tornado occurrence at the site is estimated to be about once in 6,000 years. Despite this low probability, the design of plant Category I structures included consideration of the effects of tornadic winds. 1.2.1.4 Hydrology The Design Basis Flood could exceed plant grade at the plant site. The plant grade has been established at approximately elevation 705 feet. The flood elevation includes wave runup on vertical surfaces resulting from an over water wind. The plant design considered the effects of this flood and the plant can be placed in a safe shutdown condition before the flood exceeds plant grade. The potential for floods resulting from seismically induced dam failure and/or dam failure permutations has been investigated. The results indicate that floods of this type will exceed plant grade, but to an elevation lower than the Design Basis Flood. 1.2.1.5 Geology The controlling feature of the geologic structure at the site is the Kingston thrust fault which developed some 250 million years ago. The fault has been inactive for many millions of years and recurrence of movement is not expected. The fault crosses to the northwest of the site area; however, it was not involved directly in the foundation for any of the major plant structures. 1.2.1.6 Seismology The seismic history of the southeastern United States indicates that there has been no significant seismic activity originating in the site area. The Safe Shutdown Earthquake (SSE) for the plant has been established as having a maximum horizontal acceleration of 0.18g and a simultaneous maximum vertical acceleration of 0.12g. S1-2.doc 1.2-1

SQN-17 1.2.2 Facility Description 1.2.2.1 Design Criteria The design criteria for the Sequoyah Nuclear Plant are discussed in Section 3.1. 1.2.2.2 Nuclear Steam Supply System The Nuclear Steam Supply System consists of a reactor and four closed reactor coolant loops connected in parallel to the reactor vessel. Each loop contains a reactor coolant pump, a steam generator, loop piping, and instrumentation. The Nuclear Steam Supply System also contains an electrically heated pressurizer and certain auxiliary systems. High pressure water circulates through the reactor core to remove the heat generated by the nuclear chain reaction. The heated water exits the reactor vessel and passes via the coolant loop piping to the steam generators. Here it gives up its heat to the feedwater to generate steam for the turbine generator. The cycle is completed when the water is pumped back to the reactor vessel. The inherent design of the pressurized water, closed-cycle reactor minimizes the quantities of fission products released to the atmosphere. Three barriers exist between the fission product accumulation and the environment. These are the fuel cladding, the reactor vessel and coolant loops, and the reactor containment. The consequences of a breach of the fuel cladding are greatly reduced by the ability of the uranium dioxide lattice to retain fission products. Escape of fission products through fuel cladding defect would be contained within the pressure vessel, loops and auxiliary systems. Breach of these systems or equipment would release the fission products to the reactor containment where they would be retained. The reactor containment is designed to adequately retain these fission products under the most severe accident conditions, as analyzed in Chapter 15. The reactor core, with its related Control and Protection System, is designed to function throughout its design lifetime without exceeding the acceptable fuel damage limits defined in Section 4.2. The core design, together with process and residual heat removal systems, provides for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations, including, as examples, the effects of the loss of reactor coolant flow, turbine trips due to steam and power conversion system malfunctions, and loss of external electrical load. The reactor core is a multi-region cycled core. The fuel rods are zirconium alloy tubes containing slightly enriched uranium dioxide fuel. The fuel assembly is a canless type with the basic assembly consisting of the Rod Cluster Control (RCC) guide thimbles welded to the top nozzle and mechanically fastened to the grids and bottom nozzle. The fuel rods are held by the spring clip grids in this assembly. The internals, consisting of the upper and lower core support structure, are designed to support, align, and guide the core components, direct the coolant flow to and from the core components, and to support and guide the in-core instrumentation. Dissolved boric acid is used as reactivity control device to minimize the use of RCC assemblies and assist in the control of power peaking. S1-2.doc 1.2-2

SQN-24 Full length RCC assemblies and burnable poison rods are inserted into the guide thimbles of the fuel assemblies. The control rod drive mechanisms for the full length RCC assemblies are of the magnetic latch type. The latches are controlled by three magnetic coils. They are so designed that upon a loss of power to the coils, the RCC assembly is released and falls into the core by gravity to shut down the reactor. Pressure in the system is controlled by the pressurizer, where system pressure is maintained through the use of electrical heaters and sprays. Steam can either be formed by the heaters, or condensed by a pressurizer spray to minimize pressure variations due to contraction and expansion of the coolant. Instrumentation used in the Reactor Coolant system is described in Chapter 7. Spring-loaded safety valves and power-operated relief valves for overpressure protection are connected to the pressurizer and discharge to the pressurizer relief tank, where the discharge steam is condensed and cooled by mixing with water. The reactor coolant pumps are Westinghouse vertical, single-stage, mixed flow pumps of the shaft-seal type. The power supply system to the pumps is designed so that adequate coolant flow is maintained to cool the reactor core under all credible circumstances. The original steam generators (OSG) were Westinghouse vertical U-tube units which contain Inconel tubes. Integral moisture separation equipment reduces the moisture content of the steam. The replacement steam generators (RSG) for Unit 1and Unit 2 are similar in design and are supplied by Westinghouse Electric Company, Combustion Engineering Nuclear Power LLC (CENP). The RSGs on Unit 1 were installed during the U1C12 RFO (March - May 2003). The RSGs on Unit 2 were installed during the U2C18 RFO (October - December 2012). The reactor coolant piping and the pressure-containing and heat transfer surfaces in contact with reactor water are stainless steel clad except the steam generator tubes and fuel tubes, which are Inconel and zirconium alloy, respectively. Reactor core internals, including control rod drive shafts, are stainless steel. Auxiliary system components are provided to charge the Reactor Coolant System and add makeup water, purify reactor coolant water, provide chemicals for corrosion inhibition and reactor control, cool system components, remove decay heat when the reactor is shut down, and provide for emergency coolant injection. 1.2.2.3 Control and Instrumentation The reactor is controlled by temperature coefficients of reactivity, control rod clusters, and a soluble neutron absorber, boron, in the form of boric acid. Instrumentation and controls are provided to monitor and maintain essential reactor facility operating variables such as neutron flux, primary coolant pressure, temperature, and control rod positions within prescribed operating ranges. The non-neutronic process and containment instrumentation measures temperatures, pressure, flows, and levels in the Reactor Coolant system, steam systems, containment, and auxiliary systems. Process variables which are required on a continuous basis for the startup, power operation, and shutdown of the plant are monitored in a controlled access area. The quantity and types of process instrumentation provided are adequate for safe and orderly operation of all systems and processes over the full operating range of the plant. S1-2.doc 1.2-3

SQN-31 Reactor protection is achieved by defining a region of power and coolant temperature conditions allowed by the principal tripping functions: the overpower delta temperature trip, the overtemperature delta temperature trip, and the nuclear overpower trip. The allowable operating region within these trip settings is designed to prevent any combination of power, temperatures, and pressure which would result in exceeding departure from nucleate boiling ratio limits. Additional tripping functions such as a high-pressurizer pressure trip, low-pressurizer pressure trip, high-pressurizer water-level trip, loss of coolant flow trip, steam generator low-low water-level trip, turbine trip, safety injection trip, nuclear source and intermediate range trips, positive neutron flux rate trip, and manual trip are provided to support the principal tripping functions for specific accident conditions and mechanical failures. Independent and redundant channels are combined in logic circuits which improve tripping reliability and minimize trips from spurious causes. Protection interlocks, initiation signals to the Safety Injection System, containment isolation signals, and turbine runback signals further assist in plant protection during operation. 1.2.2.4 Fuel Handling System New fuel assemblies are removed one at a time from the shipping cask and stored dry in the fuel storage racks located in the fuel storage area or wet in the spent fuel pool. New fuel is delivered to the reactor vessel by placing a fuel assembly into the new fuel elevator, lowering it into the transfer canal, storing it in the spent fuel pit or taking it through the fuel transfer system and placing it in the core by the use of the manipulator crane. Spent fuel is removed from the reactor vessel by the manipulator crane and placed in the Fuel Transfer System. In the spent fuel pool, the fuel is removed from the Transfer System and placed in the storage racks. After a suitable decay period, the fuel may be removed from storage and loaded in a shipping cask for removal from the site or the spent fuel assemblies may be placed in interim storage at SQN Independent Spent Fuel Storage Installation (ISFSI) (Section 9.1.5). Spent fuel is handled entirely under water from the time it leaves the reactor vessel until it is placed in a cask for shipment from the site or the spent fuel assemblies may be placed in interim storage at SQN Independent Spent Fuel Storage Installation (ISFSI) (Section 9.1.5). Underwater transfer of spent fuel provides an effective, economic and transparent shield, as well as a reliable cooling medium for removal of decay heat. 1.2.2.5 Waste Processing Systems The Waste Processing System provides equipment necessary for controlled treatment, and preparation for retention or disposal of liquid, gaseous, and solid wastes produced as a result of reactor operation. The Liquid Waste System collects and processes reactor grade water, removes or concentrates radioactive constituents and processes them until suitable for release or shipment offsite. The Gaseous Waste Processing System functions to remove fission product gases from the reactor coolant. The system also collects the gases from S1-2.doc 1.2-4

SQN-25 various tanks and processes. The waste processing systems, including both liquid and gas, are designed to ensure that the quantities of radioactive releases from the total plant to the surrounding environment will not exceed the 10 CFR 20 limits and are as low as practicable. The solid waste management system functions to prepare slurries and solid radwaste for shipment or for temporary onsite storage in compliance with the requirements in 10 CFR 61, 10 CFR 71, and 49 CFR 170 through 178. Waste inputs are divided into two categories: dry active waste (DAW) and wet active waste (WAW). DAW is further divided into compactible and non-compactible wastes. WAW is primarily composed of two types of waste: evaporator concentrates and spent resins. 1.2.2.6 Steam and Power Conversion System The Steam and Power Conversion System consists of a turbine-generator, main condenser, vacuum pumps, Turbine Seal System, Turbine Bypass System, hot well pumps, condensate booster pumps, main feed pumps, main feed pump turbines (MFPT), condenser-feedwater heater, feedwater heaters, heater drain pumps, and Condensate Storage System. The system is designed to convert the heat produced in the reactor to electrical energy through conversion of a portion of the energy contained in the steam supplied from the steam generators, to condense the turbine exhaust steam into water, and to return the water to the steam generator as feedwater. Each turbine generator unit consists of a tandem arrangement of one double-flow high-pressure turbine and three double-flow low-pressure turbines driving a direct-coupled generator at 1800 RPM. The generator has a nameplate rating of 1,356,200 KVA at 0.9 PF with 75 psig hydrogen pressure. Each unit employs a horizontal, single pressure, triple shell, single pass surface condenser. Return to the steam generator is through three stages of feedwater pumping and seven stages of feedwater heating. Safety relief valves and power operated relief valves, as well as a turbine bypass to the condenser are provided in the steam lines. 1.2.2.7 Plant Electrical System For Unit 1 and 2, the Plant Electric Power System consists of the main generator, the generator circuit breaker, the unit station service transformers, the common station service transformers, the main bank transformers, the diesel generators, the batteries, and the electric distribution system. The main generator supplies electrical power through isolated-phase buses to the main bank transformers and the unit station service transformers located adjacent to the Turbine Building. The primaries of the unit station service transformers are connected to the isolated-phase bus at a point between the generator circuit breaker terminals and the low-voltage connection of the main bank transformers. During normal operations the auxiliary power is typically supplied by unit power through the unit station service transformers. During startup and shutdown the auxiliary power is typically supplied by the 500-kV system through the main bank and unit station service transformers for Unit 1 and the 161-kV system through the main bank and unit station service transformers for Unit 2. During startup, shutdown, and normal operations auxiliary power may be supplied by the 161-kV system through the common station service transformers. The standby onsite power is supplied by four diesel generators. The power to the 6.9-kV common boards is supplied by the 161-kV system through the common station service transformers. S1-2.doc 1.2-5

SQN-25 The Plant Distribution System can receive AC power from either the Unit 1 or 2 nuclear power unit, the two independent preferred (offsite) power circuits, or the four diesel-generator standby (onsite) power sources and distribute it to safety-related and nonsafety-related loads as required. The two preferred circuits have access to the TVA transmission network which in turn has multiple interties with other transmission networks. The safety-related loads for the plant are divided into two redundant groups. Each load group has access to each of the two preferred offsite sources. One load group with its two associated diesel generators can provide all safety functions in each unit. The electrical systems are described in Section 8.2 and 8.3. The vital AC and DC Control and Instrument Power system consists of four 125V batteries, four battery chargers and eight 120V AC inverters with their respective safety-related loads. A spare 125V battery and spare chargers are available as needed. Each channel has a spare inverter which can be manually aligned to replace the Unit 1 or Unit 2 inverter. The 125V DC Distribution System is a safety-related system which receives power from four independent battery chargers and four 125V DC batteries and distributes it to safety-related loads of both units. The 120V AC Distribution System receives power from eight independent inverters and distributes it to the safety-related loads of both units. These systems are described in Section 8.3. 1.2.2.8 Cooling Water The Condenser Circulating Water System provides cooling water to the main turbogenerator condensers and auxiliary cooling equipment. Water from this system may also be used to dilute and disperse low level radioactive liquid waste. For each unit, three pumps are provided in the intake pumping station located at the land end of the intake channel. Water flows into the intake channel under a skimmer wall from the river. Each pump has a separate suction well with individual traveling screens and discharges through a motor-operated butterfly valve into a common single square conduit tunnel which carries the cooling water to the condensers. Unit 1 was started up before completion of the separate, permanent ERCW pumping station. Unit 1 startup utilized the ERCW pumps in the Intake Pumping Station which houses the permanent condenser circulating water (CCW) pumps. The Main Cooling Towers and the permanent ERCW pumping station were completed prior to startup of Unit 2. The new ERCW station is located offshore in the lake, at the skimmer wall, and is capable of taking suction from the river channel on loss of the downstream dam. Unit 1 operation before startup of unit 2 utilized once-through CCW and ERCW cooling, the discharge water passing through an embayment and a diffuser discharge system in the lake. The transition to the Main Cooling Towers and the permanent ERCW pumping station was made prior to startup of Unit 2. With the Main Cooling Towers operable, the CCW System may be operated in any of three modes as follows:

1. Once-through as described above, with the discharge stream passing through the Cooling Tower supply pumping station.

S1-2.doc 1.2-6

SQN-21

2. Helper mode in which the main condenser discharge stream is pumped by the Cooling Tower supply pumps into one or both natural draft Cooling Towers where the heat load is dumped to the atmosphere. Seven pumps are provided. The cooled stream then passes through the holding pond and diffuser pipes as in the once-through mode.

3. Closed cycle, in which the discharge stream is returned from the Cooling Towers to the CCW intake pumping station forebay, from which it is recycled.

1.2.2.9 Component Cooling System & Essential Raw Cooling Water The Component Cooling system (CCS) is the intermediate, closed-loop cooling water between various components handling Reactor Coolant system fluids, and the Essential Raw Cooling Water (ERCW). Two basic purposes of the CCS are:

1. To remove heat from the components and heat exchangers that are handling radioactive fluids.

2. To serve as a buffer against leakage from the nuclear systems to the ERCW and thus to the environment.

The CCS system is vital to plant operation. The system is designed as a safety system with components necessary for heat removal from other safety systems. The ERCW system is the cooling water supply and discharge to the ultimate heat sink, the Tennessee River. 1.2.2.10 Chemical and Volume Control System The Chemical and Volume Control System (CVCS) performs the following functions:

1. Fills the Reactor Coolant system (RCS).

2. Provides a source of high pressure water for pressurizing the RCS when cold.

3. Maintains the water level in the pressurizer when the RCS is hot.

4. Reduces the concentration of corrosion and fission products in the reactor coolant.

5. Adjusts the boric acid concentration of the reactor coolant for chemical shim control.

6. Provides high pressure seal water for the reactor coolant pump seals.

7. Provides a means of reactor coolant water chemistry control.

During power operation, a continuous feed-and-bleed stream is normally maintained to and from the RCS. Letdown water leaves the RCS and flows through the shell side of the regenerative heat exchanger where it gives up its heat to makeup water being returned to the RCS. The letdown water then flows through the orifices where its pressure is reduced, then through the letdown heat exchanger, followed by a second pressure reduction by a S1-2.doc 1.2-7

SQN low-pressure letdown valve. The letdown normally flows through a mixed bed demineralizer, where ionic impurities are removed, then flows either through the cation demineralizers or directly through the reactor coolant filter, and into the volume control tank via a spray nozzle. The vapor space in the volume control tank normally contains hydrogen which dissolves in the coolant. Fission gases can be removed from the system by venting of the volume control tank. The charging pumps take the coolant from the volume control tank and send it along two parallel paths: (1) to the RCS through the tube side of the regenerative heat exchanger and (2) to the seals of the reactor coolant pumps. Some RCS seal water flows into the RCS and the remainder leaves the pumps as seal leakage. From the pumps, the leakage water goes to the seal water heat exchanger and then returns for another circuit. If the normal letdown and charging path through the regenerative heat exchanger is not operable, water injected into the RCS through the reactor coolant pump seals is returned via the excess letdown heat exchanger. Surges from the RCS accumulate in the volume control tank unless a high water level in the tank causes flow to be diverted to the Hold Up Tanks. Makeup to the CVCS comes from the following sources:

1. Demineralized and deaerated water supply, when the concentration of the dissolved neutron absorber is to be reduced.

2. Boric acid tank, when the concentration of dissolved neutron absorber is to be increased.

3. A blend of demineralized and deaerated water and concentrated boric acid to match or regulate the reactor coolant boron concentration for normal plant makeup.

4. Refueling water storage tank for emergency makeup of borated water.

5. Chemical mixing tank for small quantities of hydrazine for oxygen scavenging or lithium hydroxide for pH control.

1.2.2.11 Sampling and Water Quality System The Sampling and Water Quality System provides the equipment necessary to provide required process samples for laboratory analysis. These analyses provide the essential chemical and radiochemical data required for the operation of the various process systems in each of the two units. 1.2.2.12 Ventilation The internal environments of the various buildings of the plant are controlled within acceptable limits for safety, comfort, and equipment protection by several heating, cooling, and ventilating systems. Filtration is provided in exhaust systems as required to reduce contaminants. S1-2.doc 1.2-8

SQN-16 Heating systems involve both electric and hot water systems while cooling utilizes fan coil units supplied with direct expansion, chilled water, or raw water coils. Ventilation is, for the most part, by both supply and exhaust with central intakes and exhausts for proper treatment of the air. Redundant equipment is provided for safety-related equipment. 1.2.2.13 Fire Protection System The Fire Protection system will provide a reliable water and CO2 system to extinguish fires both inside and outside the buildings. The systems are designed to provide early detection and extinguishing of fires with an overall objective of minimizing fire hazards and limiting the consequences in the event of a fire. The Fire Protection System is discussed in the Fire Protection Report (see 9.5.1). 1.2.2.14 Compressed Air System The Compressed Air System is common to both units and is divided into two subsystems: the station control and service air system, and the auxiliary control air system. The station control and service air system supplies compressed air for general plant service, instrumentation, testing, and control. The auxiliary air systems provide, as a minimum, sufficient air for an orderly plant shutdown, including Safe Shutdown Earthquake and Maximum Possible Flood. Only the auxiliary air systems are considered to be Engineered Safety Features. For detailed description see Subsection 9.3.1. 1.2.2.15 Engineered Safety Features Several Engineered Safety Features have been incorporated into the plant design to reduce the consequences of a loss-of-coolant accident. One of these safety features is an Emergency Core Cooling System (ECCS) which automatically delivers borated water via the cold legs to the reactor core for continued cooling and for negative reactivity insertion following an accidental steam release. Another safety feature which has been included is the Ice Condenser Containment system. Basically, this system provides for very rapid absorption of the energy released from the Reactor Coolant System in the improbable event of a loss-of-coolant accident. The energy is absorbed by condensing steam in a low temperature heat sink, consisting of a suitable quantity of ice permanently stored inside the containment. The ice containment system markedly reduces the peak containment pressure that would otherwise result in the event of a loss-of-coolant accident. The peak pressure is reduced to an even lower value within a few minutes. The system also removes radioactive iodine from the containment atmosphere by the action of sodium tetraborate impregnated ice. There are several other systems which help mitigate the consequences of a LOCA by aiding the systems mentioned above or by the performance of other specific functions. The first of these is the Containment Spray System which sprays cool water into the containment atmosphere to insure that the containment pressure limit is not exceeded. The air return S1-2.doc 1.2-9

SQN fans also aid in the operation of the Containment Spray System and the ice condenser by returning to the lower compartment air which is displaced through the ice condenser into the upper compartment. This system also limits hydrogen concentration by ensuring a flow of air in potentially stagnated regions. The containment isolation systems maintain containment integrity by isolating systems that pass through the containment as required. The radioactivity that may be released in the containment will be confined there by this system. To help reduce radioactive nuclide releases to the atmosphere, this plant is provided with gas treatment systems. The Emergency Gas Treatment System and the Auxiliary Building Gas Treatment System establish and maintain the air pressure below atmospheric in the Shield Building annulus and the Auxiliary Building Secondary Containment Enclosure (ABSCE), respectively. These systems reduce the concentration of radioactive nuclides in the air released from the annulus and the ABSCE. 1.2.2.16 Shared Facilities and Equipment-Safety Related Separate and similar systems and equipment are provided for each unit of the two unit Sequoyah Nuclear Plant when required. In certain instances, systems or some components of a system are shared by both units. A common control room and Auxiliary Building is provided with shared HVAC and air cleanup systems. Other principal components/systems which are shared are identified below. System Components Shared Quantity Provided Chemical and Volume Boric Acid Tanks 3 Control System Boric Acid Transfer Pumps 4 Component Cooling Pump Total of 5, up to 3 System shared Heat Exchangers Total of 6- two are shared. Spent Fuel Pit Cooling Spent Fuel Pit 1 System Spent Fuel Pit Pumps 3 Spent Fuel Pit Filter 1 Spent Fuel Pit Heat Exchanger 2 Refueling Water Purification Pumps 2 Refueling Water Filters 2 Waste Disposal System A common Waste Disposal System is used for the two units. Each containment structure has its own reactor coolant drain tank and containment sump and each is serviced by two reactor coolant drain tank pumps. All other waste disposal equipment is sized to or contracted to adequately serve two units and common Auxiliary and Service Building. S1-2.doc 1.2-10

SQN-19 System Components Shared Quantity Provided Emergency Gas Portions of the Air Cleanup Subsystem of the Treatment Systems and Gas Treatment Systems shared components Air Cleanup Systems include ducting, air purification filter and absorbers, fans and flow control dampers. Essential Raw Cooling The water supply and distribution system is Water System essentially common to both units. Standby AC power The Standby AC Power System supplies System power to both units. Vital 125V DC Control Four 125V Vital Batteries and Boards, each Power System supply two static inverters of the Vital 120V AC Control Power system on each unit. Each channel has a spare inverter which can be manually aligned to replace the Unit 1 or Unit 2 Inverter. A spare vital battery is also provided as needed. Offsite Power System The offsite power grid serves as the preferred (Preferred Power power supply for both units. Supply) 1.2.3 General Arrangement of Major Structures and Equipment The major structures are two Reactor Buildings, a Turbine Building, Auxiliary Building, a Control Building, a Service and Office Building, a Diesel Generator Building, an Intake Pumping Station, ERCW Pumping Station, two natural draft Cooling Towers, and an Independent Spent Fuel Storage Installation (ISFSI) (Section 9.1.5). The arrangement of these structures is shown in Figure 2.1.2-1. Plant arrangement plans and cross sections are presented in Figures 1.2.3-1 through 1.2.3-19. S1-2.doc 1.2-11

Security-Related Information - Figure 1.2.3-1 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-2 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-3 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-4 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-5 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-6 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-7 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-8 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-9 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-10 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-11 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-12 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-13 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-14 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-15 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-16 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-17 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-18 Withhold Under 10 CFR 2.390 Security-Related Information - Figure 1.2.3-19 Withhold Under 10 CFR 2.390 SQN-27 1.3 COMPARISON TABLES When originally submitted the following information was valid. 1.3.1 Comparisons With Similar Facility Designs Table 1.3.1-1 presents a design comparison of the Sequoyah Nuclear Steam Supply System design with that of Donald C. Cook Units 1 and 2 and Trojan. Table 1.3.1-2 presents a detailed design comparison of the Sequoyah Nuclear Plant Secondary Cycle with that of Diablo Canyon, D. C. Cook and Zion. 1.3.2 Comparison of Final and Preliminary Designs Table 1.3.2-1 lists the significant design changes that have been made since the submittal of the Preliminary Safety Analysis Report. S1-3.doc 1.3-1

SQN TABLE 1.3.1-1 (Sheet 1) DESIGN COMPARISON (EXCLUDING SECONDARY CYCLE) Sequoyah Nuclear Plant Units 1 and 2 - Comparison with Donald C. Cook Units 1 and 2 and Trojan CHAPTER CHAPTER TITLE REFERENCES SIGNIFICANT SIGNIFICANT NUMBER SYSTEM/COMPONENT (FSAR) SIMILARITIES DIFFERENCES 3.0 Steel Containment Section 3.8.2 D. C. Cook Units 1 & 2 The use of freestanding steel System primary containment vessel. 4.0 Reactor Fuel Section 4.2.1 Trojan None. Reactor Vessel Section 4.2.2 D. C. Cook Units 1 & 2, D. C. Cook Units 1 and 2 and Sequoyah Internals Trojan Units 1 and 2 have thermal shields. Trojans has neutron pads. Sequoyah upper internals have been modified to incorporate UHI. Reactivity Control Section 4.2.3 D. C. Cook Units 1 & 2, None. System Trojan Nuclear Design Section 4.3 D. C. Cook Units 1 & 2, None. Trojan Thermal-Hydraulic Section 4.4 D. C. Cook Units 1 & 2, The total primary heat output and coolant Design Trojan temperatures are higher for Sequoyah and Trojan than for the D. C. Cook Plant. T131-1.doc

SQN TABLE 1.3.1-1 (Sheet 2) DESIGN COMPARISON (EXCLUDING SECONDARY CYCLE) Sequoyah Nuclear Plant Units 1 and 2 - Comparison with Donald C. Cook Units 1 and 2 and Trojan CHAPTER CHAPTER TITLE REFERENCES SIGNIFICANT SIGNIFICANT NUMBER SYSTEM/COMPONENT (FSAR) SIMILARITIES DIFFERENCES 5.0 Reactor Coolant Sections 5.1, 5.2 D. C. Cook Units 1 & 2, The following have been added System and Connected Trojan or changed: Systems, Integrity of the Reactor Coolant - New requirements for fracture System Boundary toughness testing,

                                                                                              - New means of determining heat-up and cool-down rates.

Reactor Vessel Section 5.4 D. C. Cook Units 1 & 2, None. and Appurtenances* Trojan Reactor Coolant Section 5.5.1 D. C. Cook Units 1 & 2, None. Pumps* Trojan Steam Generators* Section 5.5.2 D. C. Cook Units 1 & 2, None. Trojan Reactor Coolant Section 5.5.3 D. C. Cook Units 1 & 2, None. Piping* Trojan Residual Heat Section 5.5.7 D. C. Cook Units 1 & 2, None. Removal System Trojan Pressurizer* Section 5.5.10 D. C. Cook Units 1 & 2, None. Trojan

All components designed and manufactured to Code edition in effect at date of purchase order.

T131-1.doc

SQN-25 TABLE 1.3.1-1 (Sheet 3) DESIGN COMPARISON (EXCLUDING SECONDARY CYCLE) Sequoyah Nuclear Plant Units 1 and 2 - Comparison with Donald C. Cook Units 1 and 2 and Trojan CHAPTER CHAPTER TITLE REFERENCES SIGNIFICANT SIGNIFICANT NUMBER SYSTEM/COMPONENT (FSAR) SIMILARITIES DIFFERENCES 6.0 Engineered Safety Features Emergency Core Section 6.3 D. C. Cook Units 1 & 2, None Cooling System Trojan Ice Condenser Section 6.5 D. C. Cook Units 1 & 2 Trojan does not use an ice System condenser. 7.0 Instrumentation & Controls Reactor Trip System Section 7.2 System functions are None similar to D. C. Cook Units 1 and 2, Trojan Engineered Safety Section 7.3 Systems functions are None. Features Actuation similar to D. C. Cook System Units 1 and 2, Trojan Systems Required Section 7.4 System functions are None. for Safe Shutdown similar to D. C. Cook Units 1 and 2, Trojan Safety-Related Section 7.5 Parametric display is Actual physical configuration Display Instrumen- similar to that of may differ due to customer tation D. C. Cook Units 1 & 2, design philosophy. Trojan T131-1.doc

SQN TABLE 1.3.1-1 (Sheet 4) DESIGN COMPARISON (EXCLUDING SECONDARY CYCLE) Sequoyah Nuclear Plant Units 1 and 2 - Comparison with Donald C. Cook Units 1 and 2 and Trojan CHAPTER CHAPTER TITLE REFERENCES SIGNIFICANT SIGNIFICANT NUMBER SYSTEM/COMPONENT (FSAR) SIMILARITIES DIFFERENCES All Other Systems Section 7.6 Operational functions are None. Required for Safety similar to D. C. Cook Units 1 & 2, Trojan Control Systems Section 7.7 Operational functions are The Sequoyah Nuclear Plant not required similar to D. C. Cook Units has approximately 50-percent load rejection for Safety 1 & 2, Trojan capability while that of the D. C. Cook Plant is 100 percent. The rod position indication for the Sequoyah Nuclear Plant and the D. C. Cook Plant is an analog system; Trojan's RPI is a digital system. 9.0 Auxiliary Systems Chemical and Volume Section 9.3.4 D. C. Cook Units 1 & 2, The Sequoyah Nuclear Plant Control System Trojan does not have deboration demineralizers. T131-1.doc

SQN TABLE 1.3.1-1 (Sheet 5) DESIGN COMPARISON (EXCLUDING SECONDARY CYCLE) Sequoyah Nuclear Plant Units 1 and 2 - Comparison with Donald C. Cook Units 1 and 2 and Trojan CHAPTER CHAPTER TITLE REFERENCES SIGNIFICANT SIGNIFICANT NUMBER SYSTEM/COMPONENT (FSAR) SIMILARITIES DIFFERENCES 11.0 Radioactive Waste Management Source Terms Section 11.1 D. C. Cook Units 1 & 2, Differences are based upon Trojan plant operational influences. Liquid Waste Section 11.2 Performance characteris- The Sequoyah Nuclear Plant Systems tics similar to D. C. Cook has a dissimilar segregated Units 1 & 2, Trojan liquid drain system. Gaseous Waste Section 11.3 D. C. Cook Units 1 and 2, None. Systems Trojan Process and Effluent Section 11.4 Functionally similar to None. Radiological D. C. Cook Units 1 and 2, Monitoring Systems Trojan 15.0 Accident Analysis Chapter 15 Similar to D. C. Cook Units The Accident Analysis sections 1 and 2, Trojan have been updated. New sections have been added, e.g., single RCA withdrawal, accidental despressurization of the RCS, computer code descriptions, etc. T131-1.doc

SQN-24 TABLE 1.3.1-2 (Sheet 1) DESIGN COMPARISON - SECONDARY CYCLE Referenced Sequoyah Diablo Feature FSAR Section Nuclear Plant Canyon D. C. Cook Zion Turbine Generator Net Generator Output (kW) 10.1, 10.2 1,183,192 *1,026,000; 1,100,000 1,050,000

                                                                      **1,122,000 Turbine Cycle Heat Rate (Btu/kW-Hr)              10.1                9,871                 *10,075;       *10,208;      ***

10,033 **10,232 Type/LSB Length 10.2 TC6F/44 TC6F/44 *TC6F/43; TC6F/44

                                                                                     **TC6F/52 Cylinders (No.)           10.2                1 H.P.-3 L.P.        1 H.P.-3 L.P. 1 H.P.-3 L.P. 1 H.P.-3 L.P.

Steam Conditions at Throttle Valve Flow (lb/hr) 10.2 14,420,210 *13,934,600; 14,120,000 13,989,300

                                                                      **14,239,300 Pressure (psia)           10.2                861.9                725           728           690 Temperature (°F)          10.2                526.9                507           507.5         501.5 Moisture Content (%)      10.1, 10.2          0.21                 *.65; **.53   NA            .25 Turbine Cycle Arrangement Steam Reheat Stages (No.) 10.1                2                    2             1             1 Feedwater Heating         10.1, 10.4.7,       7                    6             6             6 Stages (No.)             10.4.9 T131-2.doc

SQN-24 TABLE 1.3.1-2 (Sheet 2) DESIGN COMPARISON - SECONDARY CYCLE Referenced Sequoyah Diablo Feature FSAR Section Nuclear Plant Canyon D. C. Cook Zion Strings of Feedwater 10.1, 10.4.7, 3 3 3 Lowest 3 Heaters (No.) 10.4.9 Pressure; 2 All Others Heaters in Condenser 10.4.1 3 0 1 Neck (No.) Heater Drain System 10.4.9 All Drains High Pressure High Pressure High Pressure (Type) Pumped Pumped Pumped Pumped Forward Forward; Low Forward: Low Forward: Low Pressure Pressure Pressure Cascaded Cascaded Cascaded Hotwell Pumps (No.) 10.1, 10.4.7 3 3 3 4 Condensate Booster 10.1, 10.4.7 3 3 3 4 Pumps (No.) Heater Drain Pumps (No.) 10.1, 10.4.9 3 H.P.-2 L.P. 3 3 3 Main Feed Pumps 10.1, 10.4.7 2 - Turbine 2 - Turbine 2 - Turbine 2 - Turbine (No. and Type) Driven Driven Driven Driven Main Steam Bypass Capacity (%) 10.4.4 40% 40% 85% 40% Final Feedwater Temperature 10.1 435.7 *432.1; *434.8; NA

                                                                       **432.9       **430.5 T131-2.doc

SQN-13 TABLE 1.3.1-2 (Sheet 3) DESIGN COMPARISON - SECONDARY CYCLE Referenced Sequoyah Diablo Feature FSAR Section Nuclear Plant Canyon D. C. Cook Zion Condenser Type 10.1, 10.4.1 Single Single Single Single Pressure Pressure Pressure Pressure Number of Shells 10.1, 10.4.1 3 2 3 3 Design Back Pressure 10.1, 10.4.1 2 1.5 *1.71; **1.41 1.5 (In. Hg Abs) Total Condenser Duty 10.1, 10.4.1 7.829 x 109 7.6 x 109 2.5 x 109 7.18 x 109 (Btu/Hr) (Approx) (Approx) (Approx)

      *Unit 1
    **Unit 2
   ***Commonwealth Edison will not release these heat rates.

T131-2.doc

SQN-24 TABLE 1.3.2-1 (Sheet 1) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Containment 3.8.2 The steel containment was modified as a result of analyses to include stiffeners in both the vertical and horizonal directions. Fuel 4.2.1 The reactors will be fueled with 17 x 17 fuel assemblies in lieu of 15 x 15 fuel assemblies. Reactor Internals 4.2.2 The reactor internals have been modified to accept 17 x 17 fuel assemblies. Reactor 5.0 The Unit 1 and Unit 2 Steam Generators have been replaced. 5.5.15 A Reactor Vessel Head Vent System has been provided. Containment 6.2 Design of the following has been modified: Ice Condenser (1) Ice Baskets (2) Lower inlet door and hinges (3) Lower support structure (4) Lattice Frames (5) Lattice frame support columns (6) Wall panels (7) Intermediate deck doors (8) Top deck doors (9) Air handling unit supports (10) Top deck beams (11) Ice Condenser crane, crane rail, and supports (12) Stud material and diameter in containment, end walls, and crane wall (13) Number of air handling units (14) Number of refrigeration packages and associated hardware The following have been deleted: (1) Floor air-cooling duct (2) Access platform to lower inlet doors T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 2) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Containment (Cont.) The following have been added: (1) Ice basket tie-down (2) Lattice frame tangential-tie-member (3) Closer spacing of lattice frames (4) Lower inlet door arrester (5) Turning vanes on lower support structure and floor (6) Jet impingement plate (7) Foam concrete in floor (8) Glycol cooling of floor (9) Defrosting capability of wall panels and floor (10) Floor support columns (11) Wall panel cradle (12) Rounded entrance to lower doors Containment Carbon absorbers added to containment purge exhaust. The pressure vessels of the containment spray heat exchangers will conform to ASME Boiler and Pressure Vessel Code, Section VIII. 6.2.5 Electric recombiners for post LOCA hydrogen control have been added. 6.2.4 A Reactor Vessel Level Indicating System has been provided. 6.2.6 A containment vacuum relief system has been added to limit pressure differential across the steel containment vessel. T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 3) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Emergency core cooling 6.3 Safety injection pumps will normally inject into the four System cold legs of the reactor coolant system but provision for injection into the hot legs has been retained. An UHI Accumulator System was added and the Reactor head and internals modified. The UHI system was later deleted. Cables 7.1.2, Tray loading has been modified to 30 percent of the cross-section area for low-voltage power 8.3.1.2 trays (except when a single layer of cables are used) and to 60 percent for control and instrument cables trays to reflect current design practices (unless evaluated as a design criteria exception). Instrument & Controls 7.1 The process protection system has been replaced by a Westinghouse Eagle 21 System. The Main Control Room has undergone a Control Room Design Review and layout modifications. Reactor trip system 7.2 Protection system logic design has been changed from relay to solid state. Engineered safety 7.3 Increased online testability has been provided for the Engineered features Safety Features. Onsite AC power 8.3.1 A fourth diesel generator has been added to the plant. The capacity of each diesel generator has been increased. Onsite DC power 8.3.2 The two 250-volt battery systems have been replaced by four 125-volt battery systems shared between the two nuclear generating units to achieve greater diversity of the onsite dc power supplies. A fifth vital battery has been added as a spare. T132-1.doc

SQN-17 TABLE 1.3.2-1 (Sheet 4) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Component cooling 9.2.1 During normal full power operation with maximum spent fuel pit cooling available, two CCS pumps and one heat exchanger pair may be required for the unit assigned the spent fuel pit load and one CCS pump and heat exchanger pair for the other unit. It was indicated in the PSAR that two pumps and one heat exchanger were capable of serving all operating components in both units, but that two pumps and heat exchangers would normally be operated. Two CCS pumps and one heat exchanger pair and not three CCS pumps and two heat exchanger pairs may be required to remove the residual and sensible heat load plus the aligned component loads for minimum cooldown rate. However, unit cooldown operations design allows assignment of a pump and heat exchanger pair to each train of the safeguards systems thereby increasing cooldown capability. Normal alignment of the CCS has been changed to assure two independent trains of cooling. Automatic actuation has been added to start any standby pumps on the normally operating headers to help assure a continuous supply of cooling water to all loads. The following equipment is no longer served by the CCS: (1) Reactor coolant pump bearing coolers (2) Reactor vessel supports (3) Safety injection pump oil coolers (4) Charging pump oil coolers T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 5) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Component Cooling (cont) 9.2.1 The following equipment is served by the CCS: (cont) (1) Reactor coolant pump thermal barriers and motor oil coolers (2) Residual heat removal pump seal water heat exchangers (3) Safety injection pump mechanical seal coolers (4) Charging pumps mechanical seal coolers (5) Waste gas compressors The water temperature detectors have been repositioned to the outlet of each heat exchanger pair or heat exchanger group and to the main return headers to the CCS pumps. Radiation monitors have been provided at each CCS heat exchanger pair outlet. Four booster pumps (two per unit) have been included to provide the additional head necessary to overcome the high head loss through the RCS thermal barriers. A seal collection station has been provided to collect seal leakage from the CCS pumps. The three shell and tube CCS HTXs have been replaced by six plate heat exchangers. T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 6) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Spent fuel storage pool 9.1.2 The pool in the Auxiliary Building has been modified by the addition of a concrete wall separating the cask set down area from the fuel area to protect the spent fuel from an accidental drop of the cask. The storage capacity of the spent fuel pool has increased. 9.1.3 The volume of the pool has been modified. Essential raw 9.2.2 A new ERCW pump station has been provided. The AERCW system has been deleted. The auxiliary charging pumps and auxiliary letdown heat exchangers are no longer served by the ERCW system. The reactor coolant pump motor coolers and the control rod drive motor coolers are additional equipment served by the ERCW system. Sodium hypochlorite can be injected into the ERCW system in the pumping station ERCW pump compartment to control Asiatic clams. Demineralized 9.2.3 A new makeup water treatment was provided and a Contractor supplied source of makeup water can be provided. Auxiliary control air 9.3.1 Credit is now taken for auxiliary air system as a safety feature. T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 7) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Heating, ventilation, 9.4.1 Two redundant emergency cleanup air supply fans have been and air conditioning provided to recirculate a portion of the main control room air through the HEPA-charcoal filter trains during control room isolation. A capability to isolate major sections of the Auxiliary Building during emergencies and keep it at a slight negative pressure is provided. 9.4.7 An annulus vacuum control subsystem was included in the emergency gas treatment system to continuously maintain the Shield Building annulus space at a negative pressure during plant operation. Fire Protection 9.5.1 C02 storage has been moved outside the Control Building. Post Accident Sampling 9.5.10 A post accident sampling facility has been provided. Facility Main steam supply 10.3 32 OD piping has been used instead of the 33 ID indicated in the PSAR. Main condenser 10.4.2 An optional HEPA filter-charcoal adsorber system was provided to evacuation restrict radioactive effluents to a level as low as practicable. Condenser circulating 10.4.5 Cooling towers and a cooling tower supply pumping station water have been provided. T132-1.doc

SQN TABLE 1.3.2-1 (Sheet 8) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Condensate-feedwater 10.4.7 Bypass feedwater regulator valves have been included to provide additional feedwater stability during startup conditions. A motor-operated feedwater isolation valve in the piping to each steam generator was included to provide redundant valve closure in feedwater isolation signal. This system has been modified to provide the capability of restoring 85 percent of the feedwater flow to the steam generators within 20 seconds after the loss of a main feed pump by: (1) Increasing rated speed of drive turbine, and (2) Starting all auxiliary feedwater pumps. The main feed pump turbine condenser cooling is performed by the condensate instead of the raw water as indicated in the PSAR. Secondary side heat exchanger components have undergone a copper reduction program. Auxiliary Feedwater 10.4.8 A steam driven turbine auxiliary feedwater pump system has Systems been added to each unit. Redundant and independent isolation valves have been added to guard against a loss of auxiliary feedwater during a major accident. T132-1.doc

SQN-18 TABLE 1.3.2-1 (Sheet 9) MAJOR DESIGN CHANGES SINCE SUBMITTAL OF THE PSAR Reference System Section Changes Heater, drains, 10.4.9 The Unit Main Turbine Generator will receive a signal to run and vents the unit back to approximately 78% (Unit 2) and 76.6% (Unit 1) if: load (a) either No. 3 Heater Drain Tank bypass valve is open, (b) the main turbine generator is loaded to greater than 83%, (Unit 2) and 81.6% (Unit 1) and (c) after receiving a delayed indication of low flow from the discharge header of the No. 3 Heater Drain Tank Pumps. Additional logic has been provided to close level control valve at No. 3 heater drain pump discharge on loss of one drain pump to protect the remaining operating pumps when required. Waste disposal 11.2 The drains have been segregated into tritiated and non-tritiated systems. 11.3 Holdup time for the gaseous waste system has been increased to 60 days. 11.2 Provisions have been made to supply nitrogen to the steam generators when they are drained to inert them. A mobile waste system is provided as needed to process radwaste. T132-1.doc

SQN-24 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS The Westinghouse Electric Corporation contracted the design and fabrication of NSSS components at the Sequoyah Nuclear Plant including the two reactors. In addition, they supplied the initial fuel loading. TVA has also contracted with Framatome for replacement core fuel starting with the Cycle 9 reloads for both units. TVA contracted with Westinghouse Electric Company, Combustion Engineering Nuclear Power LLC for the design and fabrication of replacement steam generators (RSG) on Unit 1and Unit 2. Removal and installation construction work associated with the Unit 1 RSGs was performed by Bechtel. Removal and installation construction work associated with the Unit 2 RSGs was performed by SGT LLC. TVA's Nuclear Engineering Group (formerly the Division of Engineering Design (EN DES) in the Office of Engineering Design and Construction) had the responsibility for the design of the remainder of the plant plus additional design changes as they became necessary. TVA's Nuclear Construction Group (formerly the Division of Construction in the Office of Engineering Design and Construction) had the responsibility for construction of the plant. TVA Nuclear has the responsibility for operating the plant. TVA utilizes consultants, as necessary, to perform selected design work and to obtain specialized services. Weston Geophysical Engineering, Inc., was contracted to assist in soil foundation dynamic analyses. Engineering Data Systems, Inc., of San Francisco, assisted in seismic analysis of piping. Chicago Bridge and Iron Company, Chicago, Illinois, was contracted to design and construct the free standing steel containments for both units. Certification of material used for containment flexible seals to withstand extreme radiation and temperature conditions was done by the Presray Corporation, Pawling, New York. S1-4.doc 1.4-1

SQN 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Sequoyah Nuclear Plant is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. The term "research and development" as used in this section is the same as that used by the Commission in Section 50.2 of 10 CFR Part 50 as follows: (n) "Research and development" means (1) theoretical analysis, exploration or experimentation; or (2) the extension of investigative findings and theories of a scientific nature into practical application for experimental and demonstration purposes including the experimental production and testing of models, devices, equipment, materials and processes." The research and development discussed in the FSAR is to confirm the engineering and design values normally used to complete equipment and system designs. It does not involve the creation of new concepts or ideas. The technical information generated by these research and development programs are used either to demonstrate the safety of the design and more sharply define margins of conservatism, or to lead to design improvements. Each research and development program is briefly summarized for identification and its relationship to the Sequoyah Nuclear Plant is discussed. Detailed discussions of each program are available in a more expanded summary form in the references incorporated throughout this section. Information regarding the Mark-BW fuel assembly is provided in the referenced Topical Reports in Section 1.6 and the text in Chapter 4.5. 1.5.1 Programs Required for Plant Operation In the PSAR, the following programs were identified as required for plant design and operation:

1. Core Stability Evaluation (Item 1 in Reference 1)

The purpose of this program was to establish means for the detection and control of potential xenon oscillations and for the shaping of the axial power distribution for improved core performance. The research and development portions of this program have been completed, as discussed below. The development program for power distribution control is divided into four general areas, namely:

a. Confirmation of the capability of the out-of-core detector system to indicate axial and diametrical gross core power distributions sufficiently to permit control of xenon oscillations within specified operating limits.

S1-5.doc 1.5-1

SQN

b. Development of a control system utilizing the out-of-core detector system for axial power shaping (such a system is used in the Robert Emmett Ginna, Indian Point Unit 2, and all subsequent Westinghouse reactors).

c. Verification, during startup tests of other Westinghouse reactors, that the control system specified in Item b can control the core power distribution.

d. Verification that adequate margin exist to operate the Sequoyah Nuclear Plant at the licensed power rating by measurements taken during prior operation of other Westinghouse reactors.

Items a and b of this program have been completed satisfactorily. Items c and d were evaluated on Westinghouse reactors going into operation prior to the Sequoyah Nuclear Plant. These include Donald C. Cook Units 1 and 2 (Docket Nos. 50-315 and 50-316), Zion Units No. 1 and 2 (Docket Nos. 50-295 and 50-304) and Trojan (Docket No. 50-344). Safe operation at the design power level experimentally demonstrated, at the time of Sequoyah's initial startup, that the actual power shapes at full power are no worse than those used in the calculation of core integrity. The analytical model used to predict these power shapes has been justified by these and earlier measurements.

2. Fuel Rod Burst Program (Item 2 in Reference 1)

The original rod burst program, a study of the performance of Zircaloy cladding under simulated loss-of-coolant accident (LOCA) conditions, has been completed. It has supplied empirical data from which the effect of geometry distortion on the ability of the emergency core cooling system (ECCS) to meet the LOCA design criteria has been determined using present analytical design techniques. The program included burst and quench tests on single rods and burst tests on rod bundles. As a result of single rod tests, specific design limits have been established on peak clad temperature and allowable maximum metal-water reaction to assure effective core cooling. The multi-rod burst tests demonstrated that even when rod-to-rod contact does occur after burst, the remaining flow area is always sufficient to ensure adequate core cooling. The single rod burst test program for the 17 x 17 fuel pin array is discussed in Section 1.5.5.3.

3. Ice Condenser Containment Program (Item 4 in Reference 1)

In order to confirm the functional adequacy and the structural integrity of the designed ice condenser components, an extensive test program was performed. This program confirmed the prototype design, and was validated by additional confirmatory tests on selected production components. A summary of the completed test program is presented below, the results of which are reported in the indicated References. S1-5.doc 1.5-2

SQN Title Reference ICE BASKET TESTS Static Load Test of Ice Basket 2, 8, 10 Failure Load Test of Ice Basket 2 Dynamic Load Test of Ice Basket 2 Ice Fallout from Seismic Loading 9 Stress Analysis Report 8 WALL PANEL TESTS Wall Panel Leak Test 2 Wall Panel Radial Load Test 2 Wall Panel Shear Test 2 LATTICE FRAME TESTS Lattice Frame Load-Deflection Test 2 LOWER INLET DOOR TESTS Lower Inlet Door Dynamic Load Test 3 Lower Inlet Door Heat Transfer and 4 Leak Rate Test Lower Inlet Door Shock Absorber 5, 6 Dynamic Tests TOP DECK DOOR TESTS Dynamic Load Tests 4, 6 INTERMEDIATE DOOR TESTS Intermediate Door Dynamic Test 5 PERFORMANCE TESTS Full Scale Section Test 7, 11, 12, 14 ICE TECHNOLOGY Iodine Removal Effectiveness 13 1.5.2 Programs Not Required for Plant Operation Other areas of research and development, as outlined below, are those which give added confirmation that the designs are conservative. 1.5.2.1 Burnable Poison Program (Item 7 in Reference 1) Burnable poison rod program is complete. The burnable poison rods are borosilicate glass encased in stainless steel tubes. The fixed rods are used in the first core to reduce the concentration of boric acid poison in the moderator, thereby ensuring that the moderator coefficient of reactivity is always negative at operating temperature. S1-5.doc 1.5-3

SQN 1.5.2.2 Fuel Development Program for Operation at High Power Densities (Item 8 in Reference 1) To demonstrate satisfactory operation of fuel at high burnup and power densities, and to define design margins, a program was designed to test fuel in both the Saxton and Zorita reactors. The Saxton loose-lattice irradiation program was designed to demonstrate fuel performance at conditions significantly in excess of PWR design limits, and would establish power burnup limits for the fuel. The Zorita reactor is the first PWR with a Zircaloy core to operate at similar core conditions as the current design units. Because of the timely manner in which fuel can be irradiated in Zorita, four fuel assemblies are being tested there to demonstrate satisfactory operation of the fuel in a commercial PWR environment. Sustained successful operation of special Zorita fuel rods at peak design power levels, in excess of those planned for the Sequoyah Nuclear Plant, will increase assurance that the fuel has adequate performance margins to accommodate transient overpower operation. The Saxton Loose Lattice Irradiation and Saxton Parametric Irradiation subprograms have been completed. It is concluded that the loose lattice program has satisfactorily completed the test objective. The work of the loose lattice assemblies was partly performed under USAEC Contract AT (11-1)-3044 and has been reported on a quarterly basis (Reference 15); a fuel materials performance report has been published (Reference 16). 1.5.2.3 FLECHT (Full Length Emergency Core Cooling Heat Transfer Test) (Item 12 in Reference 1) The objective of the FLECHT program was to obtain experimental reflooding heat transfer data under simulated loss-of-coolant accident conditions for use in evaluating the heat transfer capabilities of pressurized water reactor emergency core cooling systems. The test results verified the ability of a bottom flooding ECCS design to terminate the temperature increase during a LOCA. The LOCA evaluation presented in this application utilizes the results of the FLECHT Program for the analysis of the reflooding phase of the accident. 1.5.2.4 Loss of Coolant Analysis Program (Item 14 in Reference 1) This program has been completed with the results of the Flashing Heat Transfer Program (Item 13 in Reference 1) being incorporated in the core thermal design codes used in the LOCA analysis presented in this application. The loss of coolant analysis program was established to integrate, as appropriate, the more realistic heat transfer models obtained from experimental and analytical development programs into the core thermal design codes used to evaluate the loss-of-coolant accident. 1.5.2.5 Reactor Vessel Thermal Shock (Item 16 in Reference 1) The effects of safety injection water on the integrity of the reactor vessel following a postulated loss-of-coolant accident, has been analyzed using data on fracture toughness of heavy section steel both at beginning of plant life and after irradiation corresponding to approximately 40 years of equivalent plant life. The results show that under the postulated accident conditions, the integrity of the reactor vessel is maintained. S1-5.doc 1.5-4

SQN 31 Fracture toughness data are obtained from a Westinghouse experimental program which is associated with the Heavy Section Steel Technology (HSST) Program at ORNL and from Euratom programs. Since results of the analyses are dependent on the fracture toughness of irradiated steel, efforts are continuing to obtain additional confirmatory data. Data on two-inch thick specimens became available in 1970 from the HSST Program. Their data indicated a strong temperature dependence with a rapid increase in toughness at approximately NDT. For results obtained in the HSST Program, the HSST Semiannual Progress Report, issued by the Oak Ridge National Laboratory (quarterly, beginning in 1974), should be consulted. 1.5.2.6 Blowdown Forces Program (Item 15 in Reference 1) The objective of the Blowdown Forces Program was to develop a digital computer program for the calculation of pressure, velocity, and force transients in the Reactor Coolant System during a loss-of-coolant accident, and to utilize this code in the calculation of blowdown forces on the fuel assemblies and reactor internals to ensure that the stress and deflection criteria used in the design of these components are met. Westinghouse has completed the development of BLODWN-2, an improved digital computer program for the calculation of local fluid pressure, flow and density transients in the Primary Coolant System during a loss of coolant accident. Extensive comparisons have been made between BLODWN-2 and test data. Agreement between code predictions and data has been good. The validity of the BLODWN-2 Code has been demonstrated, therefore the program is considered to be complete. MULTIFLEX is an extension of the BLODWN-2 code and includes mechanical structure models and their interactions with the thermal-hydraulic system. MULTIFLEX 3.0 includes additional enhancements such as the use of a two dimensional flow network to represent the vessel downcomer region, the allowance of nonlinear boundary conditions at the vessel and downcomer interface at the radial keys and the upper core barrel flange and the allowance for vessel motion in lieu of rigid vessel assumptions. Analyses using the MULTIFLEX and MULTIFLEX 3.0 Programs to evaluate the effects of Blowdown Forces are presented in Sections 3.9 and 5.2 of the Sequoyah FSAR. It was concluded from these analyses that the design of this reactor meets the established design criteria. 1.5.3 17 x 17 Fuel Assembly Verification Tests (Item 23 in Reference 1) A comprehensive test program for the 17 x 17 assembly has been successfully completed by Westinghouse. Reference 1 contains a summary discussion of the program. Some of the verification work described herein was conducted using 17 x 17 assemblies of seven grid design whereas the selected 17 x 17 assembly design has eight grids. Tabulated below are those 17 x 17 tests which utilized a seven grid geometry and the effect of adding an eighth grid. Test Parameter Effect Fuel Assembly Axial Stiffness Negligible effect at blowdown Structural Structural Test Test impact forces (Reference 17) Lateral Impact Additional grid shares impact load (Reference 17) S1-5.doc 1.5-5

SQN Test Parameter Effect Prototype Pressure Drop The margin between 7 grid design P Assembly Test and D loop results (Reference 18) is adequate to cover the additional P resulting from the additional grid (< 5% increase in P). Lift Force The margin between 7 grid design lift force and D loop results (Reference 18) is adequate to cover the additional lift force resulting from the additional grid. Rod Vibration Decreased span length results in improved vibration characteristics and reduced rod wear. Departure from DNB Correlation Addition of a grid increases mixing which Nucleate Boiling increases DNB margin Incore Flow TDC TDC increases as grid Mixing spacing decreases (Reference 19) The above tabulation shows that (1) additional design changes are not required (e.g. no new fuel assembly holddown spring) due to the addition of a grid and (2) seven grid test information can be used to assess the adequacy of the eight grid design. Additional testing to specifically investigate the eight grid assembly is not required. 1.5.3.1 Rod Cluster Control Spider Tests The 17 x 17 rod cluster control (RCC) spider is conceptually similar to, but geometrically different from the 15 x 15 spider. The 17 x 17 spider supports 24 rodlets (the 15 x 15 design supports 20) with no vane supporting more than two rodlets (same as the 15 x 15 design). The RCC spider tests verified the structural adequacy of the design. The RCC Spider tests have been completed. A vertical static load test approximately seven times the design dynamic load did not result in spider vane to hub joint failure. A spider was tested to 2.8 x 106 steps without failure. The spider loading was 110% of the design value for 1.8 x 106 cycles and 220% of the design loading for 1 x 106 cycles. Design load is 3600 pounds compression and 1800 pounds tension. The spring test resulted in negligible preload loss. 1.5.3.2 Grid Tests The 17 x 17 grid is conceptually similar but geometrically different from the 15 x 15 "R" grid. The purpose of the grid tests is to verify the structural adequacy of the grid design. S1-5.doc 1.5-6

SQN The grid tests have been completed. Test results are in agreement with pretest design values. The test results, along with fuel assembly structural test results, were factored into the seismic analysis (Reference 17). 1.5.3.3 Fuel Assembly Structural Tests The 17 x 17 fuel assembly tests were performed to determine mechanical strength and properties. The fuel assembly parameters obtained were as follows: lateral and axial stiffness, impact and internal structural damping coefficients, vibrational characteristics and the lateral and axial impact response for postulated accident loads. The parameters obtained from the lateral dynamic tests are used for seismic analysis, while those obtained from the axial tests are incorporated in the loss-of-coolant (blowdown) accident analysis. There is a general axial test buckling criterion which does not allow local buckling of components which could preclude control rod insertion during an accident. The fuel assembly overall buckling and component local buckling is checked during the axial static and dynamic tests. The lateral displacement associated with the fuel assembly overall (beam type) buckling is constrained by the reactor internals and therefore does not reduce the fuel assembly ultimate strength. Local component buckling was not experienced during either the static or dynamic tests for loads well in excess of the design values. The general acceptance was not violated. These tests were completed at the Westinghouse Engineering Mechanics Laboratory. A general description of the test procedure, including a description of use of the parameters as related to seismic and blowdown is presented in Reference 17. 1.5.3.4 Guide Tube Tests To verify the structural adequacy of the guide tubes, an extensive series of tests were conducted to determine guide tube deflection with simulated blowdown forces comparable to those expected during a loss-of-coolant accident and to determine the maximum acceptable deflection which assures insertion of a control rod by free fall. Additional tests were conducted to determine fatigue strength, displacement as a function of strain and the natural frequencies of the guide tubes for use in dynamic analyses. Refer to References 19 and 20 for a discussion of these tests. 1.5.3.5 Prototype Assembly Tests The purpose of these tests was to demonstrate that the 17 x 17 fuel assembly and control rod hardware designs will perform as predicted. Two prototype assemblies were sequentially tested in order to obtain the required experimental data. A single set of control rod hardware, including driveline, was used in the tests. The fuel assemblies were subjected to flow and system conditions covering those most likely to occur in a plant during normal operation as well as during a pump overspeed transient. Seismic testing is not included in the test sequence. These tests were used to verify the integrated fuel assembly and RCC performance in several areas. Data obtained included pressures and pressure drops throughout the system, hydraulic loadings on the fuel assembly and drive line, control rod drop time and stall velocity, fuel rod vibration and control rod, drive line, guide tube, and guide thimble wear during a lifetime of operation. S1-5.doc 1.5-7

SQN The D-Loop testing has been completed. The results of the testing are given in Reference 18. 1.5.3.6 Departure from Nucleate Boiling (DNB) The effect of the 17 x 17 fuel assembly geometry on the DNB heat flux has been determined experimentally and has been incorporated in a modified spacer factor for use with the W-3 correlation. The effect of cold-wall thimble cells in the 17 x 17 geometry has also been quantified. A similar program was conducted to quantify the DNB performance of the R-type mixing vane grid as developed for the 15 x 15 fuel assembly design (References 21 and 22). The results of that program were used to develop a modified spacer factor which quantifies the power capability associated with the use of the R mixing vane grid as well as the change in power capability due to the axial spacing of the grids. The modified spacer factor, along with the W-3 correlation with the cold-wall factor, was shown to be applicable to cold-wall thimble cells in the 15 x 15 geometry (Reference 22). The program has been completed and the results are reported in Reference 23. 1.5.3.7 Incore Flow Mixing In the thermal-hydraulic design of a reactor core, the effect of mixing or turbulent energy transfer within the hot assembly is evaluated using the THINC code. The rate of turbulent energy transfer is formulated in the THINC analysis in terms of a thermal diffusion coefficient (TDC). A program (Reference 19) to determine the proper value of TDC for the R grid vane, as used in the 15 x 15 fuel assembly design, has been completed and showed that a design value of 0.038 (for 26 inch spacing) can be used for TDC. These results also showed that TDC was independent of Reynold's number, mass velocity, pressure, and quality over the ranges tested. A similar TDC experimental program employed a geometry typical of the 17 x 17 fuel assembly to determine the effects of the geometry on mixing and to determine an appropriate value for TDC. A uniform axial heat flux was used. There is no analytical reason to expect that the mixing coefficient would be affected by a non-uniform axial heat flux. The THINC computer code considers the mixing in each increment along the heated length and within that increment the heat flux is considered uniform. The tests reported by Cadek (Reference 24) indicate that there was no difference, within experimental accuracy, between a test section with a uniform flux (Pitt) and one half of a cosine flux (Columbia). The heat flux will vary between the simulated fuel rods in the test section to create a thermal gradient in the radial direction. Using different flow rates and inlet temperatures, the TDC for the 17 x 17 geometry will be determined. The TDC tests are completed and the results are reported in Reference 25. 1.5.4 Inpile Fuel Densification (Item 22 in Reference 1) Operating experience with uranium dioxide fuel has indicated that the fuel may densify under irradiation, to a greater density than that to which it was manufactured. This densification can lead to shorter active fuel length stacks, increased initial rod-to-clad radial gaps, and pellet-to-S1-5.doc 1.5-8

SQN pellet axial gaps. The shorter fuel stack length gives rise to a small increase in overall, average linear power density (kW/ft). Increased radial gap dimensions result in reduced gap conductance and lead to higher pellet temperatures. Axial gaps give rise to local power peaking due to decreased neutron absorption. Westinghouse fuel densification research was directed toward producing fuel with a structure which minimizes inpile densification (hereafter called stable fuel). The objective of the program was to define material characteristics and manufacturing processes which lead to stable fuel. Stable fuel is defined as fuel whose densification is small. Residual effects of densification were evaluated on a model developed by this program. A more detailed description of the program and results is presented in Reference 26. 1.5.5 LOCA Heat Transfer Tests (17 x 17) Extensive experimental programs have been completed to determine the thermal hydraulic characteristics of 15 x 15 fuel assemblies, and to obtain experimental heat transfer data under simulated loss-of-coolant accident conditions. Complementary experimental programs were completed with a simulated 17 x 17 assembly to determine its behavior under similar loss-of-coolant accident (LOCA) conditions. Results from the 17 x 17 programs were compared with data from the 15 x 15 assembly test programs and were used to confirm predictions made by correlations and codes based on the 15 x 15 test results. Refer to Reference 27 for a more detailed discussion of these results. 1.5.5.1 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling) The NRC Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactor was issued in Section 50.4. of 10 CFR 50 on December 28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of Emergency Core Cooling Systems (ECCS). Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions. To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A DNB correlation will be developed by Westinghouse from these test results for use in the ECCS analyses. S1-5.doc 1.5-9

SQN Objective The objective of the Blowdown Heat Transfer Test was to determine the time that DNB occurs under LOCA conditions. This information will be used to confirm the existing, or develop a new Westinghouse transient DNB correlation. The steady state DNB data obtained from 15 x 15 and 17 x 17 test programs can be used to assure that the geometrical differences between the two fuel arrays can be correctly treated in the transient correlations. Program The program was divided into two phases. The Phase I tests started from steady state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown. Phase I provided separate-effects data for heat transfer correlation development. Typical parameters used for Phase I testing are shown below. Parameters Nominal Value Initial Steady State Conditions Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5 x 106 lb/hr-ft2 Core inlet temperature 550 to 600°F Maximum heat flux 306,000 to 531,000 Btu/hr-ft2 Transient Ramp Conditions Pressure decrease 0 to 350 psi/sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100 percent/sec Inlet enthalpy Constant Phase II simulates PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady state operating conditions. The fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures reach a predetermined limit. Typical parameters used for Phase II testing are shown below. S1-5.doc 1.5-10

SQN Parameter Nominal Value Initial Steady State Conditions Pressure 2250 psia Test section mass velocity 2.5 x 106 lb/hr-ft2 Inlet coolant temperature 545°F Maximum heat flux 531,000 Btu/hr-ft2 Transient Conditions Simulated break Double-ended cold leg guillotine breaks Test Description The experimental program was conducted in the J-Loop at the Westinghouse Forest Hills Facility with a full length 5 x 5 rod bundle simulating a section of a 15 x 15 assembly to determine DNB occurrence under LOCA conditions. The heater rod bundles used in this program were internally-heated rods, capable of a maximum power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples. 1.5.5.2 Results The experiments in the DDNB Facility resulted in cladding temperature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds. Facility modifications and installation of the initial test bundle were completed. A series of shakedown tests in the J-Loop were performed. These tests provided data for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975. Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November and plans were made for a new test bundle and further testing during 1978-1979. These tests were completed in December of 1979. A DNB correlation will be developed from these test results for use in Westinghouse ECCS analyses. 1.5.5.3 Single Rod Burst Test (SRBT) The single rod burst test results were used to quantify the maximum assembly flow blockage which is assumed in LOCA analyses. S1-5.doc 1.5-11

SQN The single rod burst test program for the 17 x 17 fuel assembly rods consisted of testing specimens, at the two internal pressures and the three heating rates listed below in a steam atmosphere. Heating Rate Internal Pressure (725°F to 1940°F) psi 5°F/sec 1200, 1800 25°F/sec 1200, 1800 100°F/sec 1200, 1800 All specimens were then heated 5°F/sec from 1940°F to about 2300°F, held for a short time and then cooled 5°F/sec to 1200°F. Metallography was done on specimens to determine the degree of wall thinning and the extent of oxygen embrittlement. In addition, tests were run on 15 x 15 fuel assembly rods to insure reproducibility of the 1972 single rod burst test results. The single rod burst tests are complete. The tests showed that the LOCA behavior of 17 x 17 clad in comparison to that of 15 x 15 clad exhibited no significant differences in failure ductility. Because of the results and the geometric scaling, the flow blockage as determined by 15 x 15 MRBT simulation can be used for 17 x 17 fuel geometry. 1.5.6 References

1. "Topical Report - Safety Related Research and Development for Westinghouse Pressurized Water Reactors - Program Summaries - Winter 1977 - Summer 1978" WCAP-8768, Rev. 2.

2. "Test Plans and Results for the Ice Condenser System," WCAP-8110, April 16, 1973.

3. "Test Plans and Results for the Ice Condenser System," WCAP-8110, Supplement 1, April 30, 1973.

4. "Test Plans and Results for the Ice Condenser System," WCAP-8110, Supplement 2, June 19, 1973.

5. "Test Plans and Results for the Ice Condenser System," WCAP-8110, Supplement 3, July 18, 1973.

6. "Test Plans and Results for the Ice Condenser System," WCAP-8110, Supplement 4, October 1, 1973.

7. "Test Plans and Results for the Ice Condenser System," WCAP-8110, Supplement 7, May 1974.

S1-5.doc 1.5-12

SQN

8. "Test Plans and Results for the Ice Condenser System-Stress and Structural Analysis and Testing of Ice Baskets," WCAP-8110, Supplement 8, May 1974.

9. "Test Plans and Results for the Ice Condenser System-Ice Fallout from Seismic Testing of Fused Ice Baskets," WCAP-8110, Supplement 9, May 1974.

10. "Test Plans and Results for the Ice Condenser System-Static Testing of Production Ice Baskets," WCAP-8110, Supplement 10, September 1974.

11. "Test Plans and Results for the Ice Condenser System-Ice Condenser Full-Scale Section Test at the Walz Mill FAcility," WCAP-8110, Supplement 6, May 1974.

12. "R. Salvatori (Approved), "Ice Condenser Containment Pressure Transient Analysis Method,"

WCAP-8078, March, 1973.

13. D. D. Malinowsky, "Iodine Removal in the Ice Condenser System," WCAP-7426, March 1970.

14. Final Report Ice Condenser Full-Scale Section Tests at the Waltz Mill Facility WCAP-8282 Prop. including Addenda.

15. The WCAP-3385 Series (specifically, 3385-18, -20, and -22 through -37) Reports Data from the Saxton Reactor.

16. W. R. Smalley, "Evaluation of Saxton Core II Fuel Materials Performance," WCAP-3385-57, July 1974.

17. L. Gesinski, D. Chiang, and S. Nakazato, "Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8288, December 1973.

18. E. E. De Mario and S. Nakazato, "Hydraulic Flow Test of the 17x17 Fuel Assembly," WCAP-8279, February 1974.

19. F. F. Cadek, F. E. Motley, and D. P. Dominicis, "Effect of Axial Spacing on Interchannel Thermal Mixing with R Mixing Vane Grid," WCAP-7941-L, June 1972 (Westinghouse Proprietary); and WCAP-7959, October 1972.

20. Cooper, F. W., Jr., "17 x 17 Driveline Component Tests - Phase IB, II, III, D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974.

21. F. E. Motley and F. F. Cadek, "DNB Results for New Mixing Vane Grid (R), "WCAP-7695-L, July 1972 (Westinghouse Proprietary) and WCAP-7958, October 1972.

S1-5.doc 1.5-13

SQN

22. F. E. Motley and F. F. Cadek, "DNB Test Results for R Grids with Thimble Cold Wall Cells,"

WCAP-7695-L, Addendum 1, October 1972 (Westinghouse Proprietary) and WCAP-7958, Addendum 1, October 1972.

23. K. W. Hill, F. E. Motley, F. F. Cadek, and A. H. Wenzel, "Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8297, March 1974.

24. F. F. Cadek, "Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L, May 1971 (Westinghouse Proprietary) and WCAP-7775, September 1971.

25. F. E. Motley, A. H. Wenzel, and F. F. Cadek, "The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8299, March 1974.

26. "Safety-Related Research and Development for Westinghouse Pressurizer Water Reactors, Program Summaries, Fall 1974," WCAP-8485, March 1975.

27. "Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary), November 1975.

S1-5.doc 1.5-14

SQN-17 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6.1 Topical Reports Table 1.6.1-1 lists those Westinghouse topical reports (WCAPs) and Framatome Topical Reports BAWs) referenced throughout the Sequoyah FSAR. These WCAPs and BAWs provide information additional to that provided in the FSAR and have been filed separately with the NRC in support of this and other applications. S1-6.doc 1.6-1

SQN Table 1.6.1-1 (Sheet 1) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8768, Rev.2 "Topical Report - Safety Related Research and Winter 1977 1.5 Development for Westinghouse - Program Summaries" Summer 1978 8110 "Test Plans and Results for the Ice Condenser System" April 6, 1973 1.5, 6.5 A 8110, Supplement 1 "Test Plans and Results for the Ice Condenser System" April 30, 1973 1.5, 6.5 A 8110, Supplement 2 "Test Plans and Results for the Ice Condenser System" June 19, 1973 1.5, 6.5 A 8110, Supplement 3 "Test Plans and Results for the Ice Condenser System" July 18, 1973 1.5, 6.5 A 8110, Supplement 4 "Test Plans and Results for the Ice Condenser System" October 1, 1973 1.5, 6.5 A 8110, Supplement 7 "Test Plans and Results for the Ice Condenser System" May 1974 1.5, 6.5 A 8110, Supplement 8 "Test Plans and Results for the Ice Condenser System - Stress and Structural Analysis and Testing of Ice Baskets" May 1974 1.5, 6.5 A 8110, Supplement 9 "Test Plans and Results for the Ice Condenser System - Ice Fallout from Seismic Testing of Fused Ice Baskets" May 1974 1.5, 6.5 A 8110, Supplement 10 "Test Plans and Results for the Ice Condenser System - Static Testing of Production Ice Baskets" September 1974 1.5, 6.5 A T161-1.doc

SQN Table 1.6.1-1 (Sheet 2) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8110, Supplement 6 "Test Plans and Results for the Ice Condenser System - Ice Condenser" May 1974 1.5 A 8078 "Ice Condenser Containment Pressure Transient Analysis Method" March 1973 1.5, 6.2 A 7426 "Iodine Removal in the Ice Condenser System" March 1970 1.5, 15.5, APX6A A 8282 (Prop. incl. "Final Report Ice Condenser Full-Scale Addenda) (Prop.) Section Tests at Waltz Mill Facility" May 1974 1.5, 6.5 A 3385 Series (specif. "Reports Data from Saxton Reactor" -- 1.5 0 3385-18, 20 and 22 through 37) 3385-57 "Evaluation of Saxton Core II Fuel Materials Performance" July 1974 1.5 0 8288 "Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident" December 1973 1.5 A 8279 "Hydraulic Flow Test of the 17x17 Fuel Assembly" February 1974 1.5 A 7941-L (Prop.) "Effect of Axial Spacing on Interchannel June 1972 7959 (Non-Prop.) Thermal Mixing with R Mixing Vane Grid" October 1972 1.5 A 8446 (Prop.) "17x17 Driveline Component Tests - Phase IB, 8449 (Non-Prop.) II, III, D-Loop Drop and Deflection" December 1974 1.5 A T161-1.doc

SQN Table 1.6.1-1 (Sheet 3) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 7695-L (Prop.) "DNB Results for New Mixing Vane Grid (R)" October 1972 1.5 A 7958 (Non-Prop.) 7695-L, Addendum 1 "DNB Test Results for R Grids with Thimble (Prop.) Cold Wall Cells" October 1972 1.5 A 7958, Addendum 1 (Non-Prop.) 8297 "Effect of 17x17 Fuel Assembly Geometry on DNB" March 1974 1.5 A 7667-L (Prop.) "Interchannel Thermal Mixing with Mixing May 1971 7775 (Non-Prop.) Vane Grids" September 1971 1.5 A 8299 "The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing" March 1974 1.5 A 8485 "Safety-Related Research and Development for Westinghouse Pressurizer Water Reactors, Program Summaries, Fall 1974" March 1974 1.5 8622 (Prop.) "Westinghouse ECCS Evaluation Model - 8623 (Non-Prop.) October Version" November 1975 1.5 A 7822 "Indian Point Unit No. 2 Internals Mechanical Analysis for Blowdown Excitation" December 1971 3.9 7920 "Indian Point Unit No. 2 Primary Loop Vibration Test Program" -- 3.9 T161-1.doc

SQN-31 Table 1.6.1-1 (Sheet 4) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8373 "Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974" -- 3.9 7558 "Seismic Vibration Testing with Sine Beats" October 1971 3.9 8516 (Prop.) "UHI Plant Internals Vibration Measurement 8517 (Non-Prop.) Program and Pre- and Post- Hot Functional A Examinations" March 1975 3.9 A 9645 (Prop.) "Verification of Upper Head Injection Reactor 9646 (Non-Prop.) Vessel Internals for Pre-Operational Tests on Sequoyah 1 Power Plant" March 1981 3.9 7422 "Westinghouse PWR Core Behavior Following a Loss-of-Coolant Accident" September 1971 3.9 7950 "Fuel Assembly Safety Analysis for Combined Seismic and Loss-of-Coolant Accident" July 1972 3.9 A 7422 "Westinghouse PWR Core Behavior Following a Loss-of-Coolant Accident" September 1971 3.9 7918, Rev. 1 "Description of the BLODWN-2 Computer Code" October 1970 3.9 8708 (Prop.) MULTIFLEX - A Fortran-IV Computer Program for February 1976 3.9 A analyzing Thermal-Hydraulic-Structure System Dynamics 8709 (Non-Prop) MULTIFLEX - A Fortran-IV Computer Program for September 1977 3.9 A analyzing Thermal-Hydraulic-Structure System Dynamics 9735 Rev. 2 (Prop) MULTIFLEX 3.0 A FORTRAN-IV Computer Program for February 1998 3.9 Analyzing Thermal-Hydraulic-Structural System Dynamics (III) Advanced Beam Model T161-1.doc

SQN-31 Table 1.6.1-1 (Sheet 5) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 9736 Rev. 2 (Non-Prop) MULTIFLEX 3.0 A FORTRAN-IV Computer Program for February 1998 3.9 Analyzing Thermal-Hydraulic-Structural System Dynamics (III) Advanced Beam Model 7401 "Loss-of-Coolant Accident Analysis: Comparison Between BLODWN-2 Code Results and Test Data" November 1969 3.9 7817 "Seismic Test of Electrical and Control Equipment" December 1971 3.10 7817, Supplement 1 "Seismic Testing of Electrical and Control Equipment" December 1971 3.10 7817, Supplement 2 "Seismic Testing of Electrical and Control Equipment" January 1971 3.10 7817, Supplement 3 "Seismic Testing of Electrical and Control Equipment" January 1971 3.10 7774, Volume 1 "Enrironmental Testing of Engineered Safety Features Related Equipment (NSSS-Standard Scope)" August 1971 6.3 8301 (Prop.) "LOCTA-IV Program: Loss-of-Coolant Transient 8305 (Non-Prop.) Analysis" June 1974 15.3, 15.4 7422-L (Prop.) "Westinghouse PWR Core Behavior Following January 1970 7422 (Non-Prop.) a Loss-of-Coolant Accident" August 1971 15.4 8219 "Fuel Densification Experimental Results and Model for Reactor Application" October 1973 15.4 7750 "A Comprehensive Space-Time Dependent Analysis of Loss-of Coolant (Satan 4 Digital Code)" August 1971 15.4, 5.2 7665 "PWR FLECHT (Full Length Emergency Core Heat Transfer), Final Report" April 1971 15.4 T161-1.doc

SQN-31 Table 1.6.1-1 (Sheet 6) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 7437-L (Prop.) "LOCTA-R2 Program" Loss-of-Coolant Transient January 1970 7835 (Non-Prop.) Analysis" January 1972 15.4 7909 "MARVEL - A Digital Computer Code for Transient Analysis of A Multiloop PWR System" June 1972 15.4, 15.1 7969 "Calculation of Flow Coastdown after Loss of Reactor Coolant Pump (PHOENIX Code)" September 1972 15.4 7907 "LOFTRAN Code Description" June 1972 15.4, 15.1 7306 "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors" April 1969 15.4,7.1,7.2 7588, Rev. 1 "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Method" December 1971 15.4, 15.5 A 7979 "TWINKLE - A Multi-Dimensional neutron Kinetics Computer Code" November 1972 15.4, 15.1 A 8339 (Prop.) "Westinghouse ECCS Evaluation Model - Summary" June 1974 15.4 8339 (Non-Prop.) July 1974 15.4 A 8302 (Prop.) "SATAN-VI Program: Comprehensive Space-Time June 1974 15.4 A 8306 (Non-Prop.) Dependent Analysis of Loss-of-Coolant" June 1974 15.4 A 8170 (Prop.) "Calculational Model for Core Reflooding After June 1974 15.4 A 8171 (Non-Prop.) a Loss-of-Coolant Accident (WREFLOOD CASE)" June 1974 15.4 A T161-1.doc

SQN Table 1.6.1-1 (Sheet 7) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8354, Supplement 1 "Long Term Ice Condenser Containment (Prop.) LOTIC Code Supplement 1" July 1974 15.4 A 8355, (Non-Prop.) May 1975 15.4 A 9220 (Prop.) "Westinghouse ECCS Evaluation Model February 1978 15.4 9221 (Non-Prop.) February 1978 Version" February 1978 15.4 7372 "Control of the Hydrogen Concentration Following a Loss-of-Coolant Accident by Containment Venting for the H. B. Robinson Plant" November 1969 15.4 A 8370, Rev. 7A "Quality Assurance Plan Westinghouse Nuclear Energy Systems Divisions" February 1975 17.1B A 8370, Rev. 8A "Westinghouse Water Reactor Divisions Quality Assurance Plan" September 1977 17.1B A 8370, Rev. 8A "Westinghouse Water Reactor Divisions Quality Assurance Plan" October 1979 17.1B 7800, Rev. 5 "Nuclear Fuel Division Quality Assurance Program Plan" December 1977 17.1B 7800, Rev. 5 "Nuclear Fuel Division Quality Assurance Program Plan" December 1977 17.1B 8336 (Prop.) "Ice Condenser System Lower Inlet Door May 1974 A 8110, Supplement 5 Shock Absorber Test Plans and Results" May 1974 6.5 A (Non-Prop.) T161-1.doc

SQN Table 1.6.1-1 (Sheet 8) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8304 (Prop.) "Stress and Structural Analysis and Testing of Ice Baskets" May 1974 6.5 A 8110, Supplement 9-A "Ice Fallout From Seismic Testing of Fused Ice Basket" May 1974 6.5 9725 "Westinghouse Technical Support Complex" June 1980 7.8 8200, Rev. 2 (Prop.) "WFLASH-4 FORTRAN-IV Computer Program for A 8261, Rev. 1 Simulation of Transients in a Multi-Loop PWR" (Non-Prop) August 1974 15.3 A 8219 "Fuel Densification Experimental Results and Model for Reactor Application" October 1973 15.3 A 7835 "LOCTRA-R2 Program Loss-of-Coolant Transient Analysis" January 1972 15.3 7213 (Prop.) "The TURTLE 24.0 Diffusion Depletion Code" June 1968 15.1 A 7758 (Non-Prop.) September 1971 15.3 A 3296-26 "LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094" September 1963 15.3, 15.4, 15.1 7969 "Calculation of Flow Coastdown after Loss of Reactor Coolant Pump (PHOENIX Code)" September 1972 15.3, 15.1 7907 "LOFTRAN Code Description" June 1972 15.3 7908 "FACTRAN, A Fortran-IV Code for Thermal Transients in UO Fuel Rods" June 1972 15.3, 15.1 2 T161-1.doc

SQN Table 1.6.1-1 (Sheet 9) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8479, Rev. 2 (Prop.) "Westinghouse Emergency Core Cooling System 8480, Rev. 2 Evaluation Model Application to Plants (Non-Prop.) Equipped with Upper Head Injection" January 1975 15.3, 15.4 7894 "Long Term Transient Analysis Program for PWRs (BLKOUT Code)" June 1972 15.1 7980 "WIT-6 Reactor Transient Analysis Computer Program Description" November 1972 15.1 A 7756 "Power Distribution in the R. E. Ginna PWR" October 1971 7.7 A 7571 "Rod Position Monitoring" April 1971 7.7 A 7778 "Solid State Rod Control System, Full Length" December 1971 7.7 7769, Rev. 1 "Overpressure Protection for Westinghouse Pressurized Water Reactors" June 1972 5.2 7706 "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients" September 1971 7.1,7.2,7.3 7862 "Isolation Tests - Process Instrumentation Isolation Amplifier - Westinghouse Computer and Instrumentation Division" September 1972 7.2 A 7705 "Engineered Safeguards Final Device or Activator Testing" February 1973 7.3 7924 "Basis for Heatup and Cooldown Limit Curves" August 1972 5.2 A T161-1.doc

SQN Table 1.6.1-1 (Sheet 10) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 7488-L (Prop.) "Solid State Logic Protection System Description" March 1971 7.2, 7.3 A 7672 "Solid State Logic Protection System Description" May 1971 7.1,7.2,7.3 A 7380-L (Prop.) "Nuclear Instrumentation System" January 1971 7.2, 7.7 7506-L (Prop.) "Nuclear Instrumentation System Isolation Amplifier" October 1970 7.2, 7.7 A 7819 "Nuclear Instrumentation System Isolation Amplifier" January 1972 7.2 A 7744 (Vol. I & II) "Environmental Testing of Engineered Safety Sept. 1971 (I) 3.11, 6.3 Features Related Equipment" Jan. 1972 (II) 7.3 7607 "In-Core Instrumentation (Flux-Mapping System and Thermocouples)" July 1971 7.7 7921 "Damping Valves of Nuclear Power Plant Components" November 1972 3.7 A 7671 "Process Instrumentation for Westinghouse Nuclear Steam Supply Systems" May 1971 5.2,7.2,7.3 8004 "Topical Report - Safety Related Research and Development for Westinghouse Pressurized Water Reactor Program Summaries" Fall 1972 1.5 8077 (Prop.) "Ice Condenser Containment Pressure Transient Analysis Method" March 1973 6.2 A T161-1.doc

SQN-31 Table 1.6.1-1 (Sheet 11) Westinghouse Topical Reports Incorporated by Reference WCAP Section(s) NRC Review Number Title Date Referenced Status 8185 (Vol. 1 & 2) "Reference Core Report 17x17" December 1973 4.0, 15.1, 15.2, 15.3 15.4 7861 "Methods of Determining the Probability of a Turbine Missile Hitting a Particular Plant Region" February 1972 10.2 7623 "Heavy Section Steel Technology Program Technical Report No. 13 - Dynamic Fracture Toughness Properties of Heavy Section Steel" December 1970 5.2

7503, Rev. 1 "Determination of Design Pipe Breaks for the Westinghouse Reactor Coolant System February 1972 5.2 5890 "Ultimate Strength Criteria to Ensure No Loss-of-Function of Piping and Vessels Under Earthquake Loading" 1969 5.2 7820, Supplement 2 "Electric Hydrogen Recombiner for PWR Containments, Equipment Qualification Report" November 1973 3.11 16996, Rev. 1 Realistic LOCA Evaluation Methodology Applied November 2016 15.4 A (Vol. 1, 2, & 3) to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)

T161-1.doc

SQN-24 Table 1.6.1-1 (Sheet 12) (Continued) AREVA Topical Reports Incorporated by Reference BAW Section(s) NRC Review Number Title Date Referenced Status BAW-1419 PEEL - A Transport Code for Special Depletion May 1978 4.5 Approved BAW-10054P Fuel Densification Report May 1973 4.5 Approved Rev. 2 BAW-10084P-A Program to Determine In-Reactor Rev. 3 Performance of B&W Fuels - Cladding Creep Collapse July 1995 4.5 Approved BAW-10096A, B&W NPGD Quality Assurance Program Rev. 4 for the Nuclear Steam System and Nuclear Steam Core Product Lines March 1982 4.5 Approved BAW-10115A NULIF - Neutron Spectrum Generator, Few-Group Constant Calculator and Fuel Depletion Code February 1972 4.5 Approved BAW-10133P Mark C Fuel Assembly LOCA-Seismic Rev. 1 Analysis May 1979 4.5 Approved BAW-10147P-A Fuel Rod Bowing in Babcock & Rev. 1 Wilcox Fuel Designs May 1983 4.5 Approved BAW-10156-A -LYNXT- Core Transient Rev. 1 Thermal-Hydraulic Program August 1993 4.5 Approved BAW-10159P-A BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies July 1990 4.5 Approved BAW-10162P-A TAC03 - Fuel Pin Thermal Analysis Computer Code October 1989 4.5 Approved BAW-10163P-A Core Operating Limits Methodology for Westinghouse Designed PWRs June 1989 4.5 Approved T161-1.doc

SQN-24 Table 1.6.1-1 (Sheet 13) (Continued) AREVA Topical Reports Incorporated by Reference BAW Section(s) NRC Review Number Title Date Referenced Status BAW-10168A B&W LOCA Evaluation Model for Rev. 3 Recirculating Steam Generator Plants November 1993 4.5 Approved BAW-10170P-A Statistical Core Design For Mixing Vane Cores December 1988 4.5 Approved BAW-10172P-A Mark-BW Mechanical Design Report December 1989 4.5 Approved Rev. 1 BAW-10180-A NEMO - Nodal Expansion Method Rev. 1 Optimized March 1993 4.5 Approved BAW-10183P-A Fuel Rod Gas Pressure Criterion (FRGPC) July 1995 4.5 Approved BAW-10184P-A GDTACO - Urania-Gadolinia Thermal Analysis Code February 1995 4.5 Approved BAW-10186P-A Extended Burnup Evaluation June 2003 4.5 Approved BAW-10189P CHF Testing and Analysis of the Mark-BW Fuel Assembly Design August 1993 4.5 Approved BAW-10199P The BWU Critical Heat Flux Correlations December 1994 4.5 Approved BAW-10220P Mark-BW Fuel Assembly Application for Sequoyah Nuclear Units 1 and 2 March 1996 4.5 Submitted BAW-10227P-A Evaluation of Advanced Cladding and Rev. 1 Structural Material (M5) in PWR Reactor Fuel June 2003 4.5, 11A Approved BAW-10231P-A COPERNIC Fuel Rod Design Rev. 1 Computer Code January 2004 4.5 Approved T161-1.doc

SQN-17 1.7 LIST OF ABBREVIATIONS 1.7.1 Abbreviations of Organizations AACC American Association for Contamination Control ACI American Concrete Institute AEC Atomic Energy Commission AISC American Institute of Steel Construction AMRA Air Moving and Conditioning Association ANS American Nuclear Society ANSI American National Standards Institute ARC Alliance Research Center ARI Air Conditioning and Refrigeration Institute ASCE American Society of Civil Engineers ASHRAE American Society of Heating, Refrigeration, and Air Conditioning Engineers ASME American Society of Mechanical Engineers ASTM American Society for Testing Materials AWS American Welding Society AWWA American Water Works Association BAW Framatome Cogema Fuels Topical Reports BWFC Babcock and Wilcox Fuels Company BWNT Babcock and Wilcox Nuclear Technologies CE Civil Engineering of NE CTI Cooling Tower Institute DOT Department of Transportation EEB Electrical Engineering Branch of NE EDS Engineering Data Systems FRA Framatome HEI Heat Exchange Institute ICRP International Commission on Radiological Protection IEEE Institute of Electrical and Electronics Engineers INEL Idaho National Engineering Laboratories IPCEA Insulated Power Cable Engineers Association MIT Massachusetts Institute of Technology MTB Mechanical Technology Branch of NE NBS National Bureau of Standards NE Nuclear Engineering of NP NED Nuclear Equipment Division of Westinghouse NEMA National Electric Manufacturers' Association NES Nuclear Energy Systems of Westinghouse NFD Nuclear Fuel Division of Westinghouse NFI Nuclear Fuel Industries NFPA National Fire Protection Association NT Nuclear Technology of NE NOAA National Oceanic and Atmospheric Administration NP Nuclear Power S1-7.doc 1.7-1

SQN NRC Nuclear Regulatory Commission NSD Nuclear Service Division of Westinghouse NSF National Science Foundation NSRB Nuclear Safety Review Board ORNL Oak Ridge National Laboratory PD Pensacola Division of Westinghouse PORC Plant Operations Review Committee PWR-SD Pressurized Water Reactor Systems Division of Westinghouse RGE Rochester Gas and Electric Company SAMA Scientific Apparatus Makers Association SMACNA Sheet Metal and Air Conditioning Contractors National Association, Inc. SMD Specialty Metals Division of Westinghouse SNEC Saxon Nuclear Experimental Corporation SQN Sequoyah Nuclear Plant TD Tampa Division of Westinghouse TEMA Tubular Exchange Manufactures Association TVA Tennessee Valley Authority UEM Union Electricia Madrilina USACE United States Army Corps of Engineers USAS United States of American Standard USGS United States Geological Survey USWB United States Weather Bureau VAA Volunteer Army Ammunition WRC Welding Research Council W Westinghouse Electric Corporation 1.7.2 Abbreviations and Symbols A-Auto Accident-Automatic AUX BLDG Auxiliary Building ABGTS Auxiliary Building Gas Treatment System ABI Auxiliary Building Isolation ABN Abnormal ac Alternating Current A/C Air Conditioning ACC Accumulator ACR Auxiliary Control Room ACS Auxiliary Charging System ADS Automatic Dispatch System AERCW Auxiliary Essential Raw Cooling Water AFD Axial Flux Difference AFW Auxiliary Feedwater AHU Air Handling Unit ALM Alarm ALT Alternate/Alteration AMB Ambient A Ampere AMT Auxiliary Make-up Tank ANAL Analysis ANALZ Analyzer AO Axial Offset AP-Auto Accident, Process-Automatic APDMS Axial Power Distribution Monitoring System API Atecedent Precipitation Index AT Accumulator Tank S1-7.doc 1.7-2

SQN ATM Atmosphere AUO Assistant Unit Operator AUTO Automatic AUX Auxiliary AVG Average AWG American Wire Gage AZ Azimuth Beff Effective Delayed Neutron Fraction BAL Balance BAT Boric Acid Tank BTRY Battery BLDG Building BLWDN Blow Down BLK Block BO Blackout BOL Beginning of Life BRG Bearing BKR Breaker BPRA Burnable Poison Rod Assembly BTD Bearing Thrust Trip Device BTU British Thermal Unit BTUH British Thermal Unit per Hour BWG Birmingham Wire Gage BWR Boiling Water Reactor C Centigrade CAL Caloric CAV Cavity CB Control Board CC Cubic Centimeters CCHX Component Cooling Heat Exchanger CCP Centrifugal Charging Pump CCS Component Cooling System CCSDT Component Cooling Pump Seal Drain Tank CCST Component Cooling Surge Tank CCW Condenser Circulating Water CDT Chemical Drain Tank CECC Central Emergency Control Center CFM Cubic Feet per Minute CLFM Centerline Fuel Melt CFS Cubic Feet per Second CHEM Chemical CHF Critical Heat Flux CIRC Circular CKV Check Valve CMPNT Component CNDS Condensate CNFP Commercial Nuclear Fuel Plant CNTM Containment COL Column COLR Core Operating Limits Report CONT Control/Controller S1-7.doc 1.7-3

SQN-28 COMM Communication CONC Concentration COND Condenser CONN Connect/Connection CPM Count per Minute CPU Central Processing Unit CRDL Control Rod Driveline CRDM Control Rod Drive Mechanism CS Containment Sump CSP Containment Spray Pump CSSTR Common Station Service Transformer CSTG Casting CT Control Transformer CV Control Valve CVCS Chemical & Volume Control System CVN Charpy V-Notch CWA Cask Work Area CWS Chilled Water Supply CYL Cylinder DB Dry Bulb DBA Design Basis Accident DBF Design Basis Flood dc Direct Current DCB Diesel Control Board DCS Distributed Control System DNB Departure from Nucleate Boiling DECON Decontamination DEG Degree DEMIN Demineralizer DEPT Department DES Design DET Detector DF Decontamination Factor DISCH Discharge DISTR Distribution DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio DOP Dioctye Phthalate Test dp Differential Pressure DR Drain DSL Diesel DWG Drawing ECC Emergency Core Cooling ECCS Emergency Core Cooling System EEP Environs Emergency Plan EFL Effluent EGTS Emergency Gas-Treatment System E-H Electro Hydraulic Control System EHC Electrohydraulic Control System E/I Voltage to Current EJCTR Ejector EL Elevation ELEC Electric S1-7.doc 1.7-4

SQN-26 ELEM Elementary/Element EMD Electromechanical Device EMERG Emergency EMF Electro-mechanical Force EOC End of Cycle EOL End of Life E/P Voltage to Pneumatic EQUIP Equipment ERCW Essential Raw Cooling Water ERCWS Essential Raw Cooling Water System ESF Engineered Safety Features EST Estimation EVAP Evaporator EXCH Exchange EXH Exhaust EXT External EXT STW Extraction Steam F Farenheit FCV Flow Control Valve FD Feed FDCT Floor Drain Collector Tank FESB FLEX Equipment Storage Building FL Floor FLD Field FLTR Filter FLX Flexible FPM Feet per Minute FPS Feet per Second FS Flow Switch FSAR Final Safety Analysis Report/Updated Final Safety Analysis Report FT Feet FW Feedwater GA Gauge GAL Gallon GCB Generator Circuit Breaker GDC General Design Criteria GDT Gas Decay Tank GEF General Exhaust Fan GEN Generator GEN General GND Ground GNN Generator End GOV Governor GPD Gallons per Day GPM Gallons per Minute GSF General Supply Fan GTCC Greater Than Class C GVN Governor End GWPS Gaseous Waste Processing System H2 Hydrogen HCF Hot Channel Factor HD Head HDR Header HEPA High Efficiency Particulate Air S1-7.doc 1.7-5

SQN-27 HGR Hanger HI-STORM 100 Holtec International Storage and Transfer Operation Reinforced Module HI-STORM FW Holtec International Storage Module Flood and Wind HI-TRAC Holtec International Transfer Cask HOR Horizontal hp Horsepower HP High Pressure HPFP High Pressure Fire Protection System HR Hour HRZ Horizontal HS Hand Switch HSDT Hot Shower Drain Tank HSG Housing HTR Heater HVAC Heating, Ventilating and Air Conditioning HYDR Hydraulic HYDRO Hydrostatic HZ Hertz ICC Inspection Control Card I/E Current to Voltage I/I Current to Current IMP Impeller IN Inch INDR Indicator INFO Information INJ Injection INOP Inoperative INSP Inspection INST Instructions I/O Input/Output I/P Current to Pneumatic ISFSI Independent Spent Fuel Storage Installation S1-7.doc 1.7-6

SQN-24 ISOL Isolation JB Junction Box JCT Junction K Kip KIP 1000 Pounds kJ Kilojoules kV Kilovolt kVA Kilovolt Ampere kW Kilowatt kWH Kilowatt Hours LAB Laboratory LB Pounds LCO Limiting Conditions for Operation LCV Level Control Valves LHR Linear Heat Rate LLC Limited Liability Company LOCA Loss of Coolant Accident LP Low Pressure LPT Low Profile Transporter LPZ Low Population Zone LS Limit Switch LSS Lower Support Structure LTDN Letdown LWPS Liquid Waste Processing System MAN Manual MAP Maximum Allowable Peak Mark-BW Mark-BW fuel MCC Motor Control Center MCR Main Control Room MECH Mechanical MFPT Main Feedwater Pump Turbine MFRR Manufacturer MISC Miscellaneous MK NO Mark Number MOV Motor Operated Valve MPC Multi-Purpose Canister mR Millirem MSR Moisture Separator Reheater MKUP Makeup MULT Multiple MV Millivolt MVA Megavoltamperes MW Megawatt MWH Megawatt-Hour MWT Megawatt Thermal N2 Nitrogen NDT Nondestructive Testing NDTT Nil Ductility Transition Temperature NIM Nuclear Instrumentation Module NIS Nuclear Instrumentation System NOM Nominal S1-7.doc 1.7-7

SQN-24 NOR Normal NQAM Nuclear Quality Assurance Manual NPSH Net Positive Suction Head NSSS Nuclear Steam Supply System NUC Nuclear NVT Fast Neutron Exposure (No. x Velocity x Time) O2 Oxygen OD Outside Diameter OPER Operator ORF Orifice OSC Oscillograph OSGSF Old Steam Generator Storage Facility OSG Original Steam Generators P-AUTO Process-Automatic PAX Private Automatic Exchange PCB Power Circuit Breaker PCI Pellet Cladding Interaction PD Positive Displacement PDIS Pressure Differential Indicating Switch PDS Pressure Differential Switch PF Power Factor pH Measure of Acidity and Basicity PIE Post Irradiation Exam PLT Plant PMF Probable Maximum Flood PMP Pump PMWS Primary Makeup Water System PNEU Pneumatic PNL Panel POSN Position PPM Parts Per Million PRESS Pressure PRI Primary PROC Procedure PROP Proportional PROT Protection PRT Pressurizer Relief Tank PZR Pressurizer PS Pressure Switch PSAR Preliminary Safety Analysis Report PSCC Power System Control Center PSIA Pounds Per Square Inch, Absolute PSIG Pounds Per Square Inch, Gauge P Signal High Containment Pressure Signal PT Point PW Primary Water PWR Pressurized Water Reactor Px Power Supply PWR Sply Power Supply S1-7.doc 1.7-8

SQN-24 QA Quality Assurance QC Quality Control QTY Quantity QUAL Quality RAD Radiation RAD DET Radiation Detector RADWASTE Radioactive Waste RC Reactor Coolant RCC Rod Cluster Control RCCA Rod Cluster Control Assembly RCDT Reactor Coolant Drain Tank RCL Reactor Coolant Loop RCP Reactor Coolant Pumps RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RCW Raw Cooling Water REAC Reactor RECIP Reciprocating RECIRC Recirculation REF Reference REG Regular REGEN Regenerative REP Radiological Emergency Plan RSVR Reservoir REV Revision RHR Residual Heat Removal RHRP Residual Heat Removal Pump R/HR Rem Per Hour RM Radiation Monitor RMS Radiation Monitoring System RO Reactor Operator RPS Reactor Protection System RSG Replacement Steam Generators RTNDT Reference Temperature Ni1 Ductility Frans RTD Resistance Temperature Detector RW Raw Water RWMS Reactor Water Makeup System RWST Refueling Water Storage Tank RV Reactor Vent SAC Service Air Compressor SAF Safety SCD Statistical Core Design SCFM Standard Cubic Feet Per Minute SCL Scale SFP Spent Fuel Pit SFPCS Spent Fuel Pit Cooling System SG Steam Generator SGT Steam Generating Team SD Shutdown S1-7.doc 1.7-9

SQN SDL Statistical Design Limit SI Safety Injection SIP Safety Injection Pump SIS Safety Injection System SKIM Skimmer SLV Sleeve SM Shift Manager SMPL Sampling SQN Sequoyah Nuclear Plant SOL Solenoid SP Set Point SP GR Specific Gravity SRO Senior Reactor Operator SRST Spent Resin Storage Tank SS Stainless Steel SSE Safe Shutdown Earthquake S Signal Safety Injection System Signal SSPS Solid State Protection System STBY Standby STD Standard STM Steam STM GEN Steam Generator STP Standard Temperature and Pressure SUCT Suction SW Switch SWG Switch Gear SWP Screen Wash Pump SYS System TC Thermocouple TD (removing existing TD, not used) Theoretical Density TDC Thermal Diffusion Coefficient TDCT Tritiated Drain Collector Tank TEMP Temperature TFTR Transportable Flow Test Rig THERM Thermal THERMO Thermostat TIG Tungsten Inert Gas TK Tank TR Transmitter-Receiver TRANS Transfer/Transformer TRM Tennessee River Mile TURB Turbine TWR Tower UHI Upper Head Injection UO Unit Operator UPSTR Upstream US Unit Supervisor S1-7.doc 1.7-10

SQN USST Unit Station Service Transformer UV Undervoltage V Volts Vac Volts - ac Vdc Volts - dc VAC Vacuum VCT Volume Control Tank VEL Velocity VENT Ventilation VERT Vertical VISC Viscosity VLV Valve E/I Voltage to Current VOL Volume WDS Waste Disposal System E/P Voltage to Pneumatic WGS Waste Gas System WPS Waste Processing System WT Weight WTR Water WTS Waste Treatment System XMTR Transmitter XS Transformer Switch ZS Position Switch S1-7.doc 1.7-11

SQN 1.8 TECHNICAL QUALIFICATIONS OF APPLICANT (HISTORICAL) The TVA power system is one of the largest in the United States with hydro, fossil and nuclear generating capability. 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. The Tennessee Valley Authority has been engaged in the business of designing, constructing, and operating large power-producing hydro and steam units for over 50 years. TVA's technical qualifications to construct and operate Sequoyah units 1 and 2 are evidenced by the skills and experience gained over many years in the power business. This experience is supplemented by the skills and experience of TVA's consultants and its contractors in assisting in the design, construction, and operation of the Sequoyah Nuclear Plant. TVA acts as its own engineer-constructor and as such has pioneered in erecting large generating units. Examples are the 1,150 megawatt electric (MWe) unit placed in operation at the Paradise Steam Plant; the 1,300 MWe units in operation at the Cumberland Steam Plant; the three 1,100 MWe units at the Browns Ferry Nuclear Plant; and one 1,170 MWe unit 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. TVA has an experienced, competent nuclear plant design organization, including a large number of engineers with many years of steam plant experience in the design and construction of large steam plants, including the design of the Browns Ferry (completed), Sequoyah (completed), and Watts Bar (Unit 1 completed), and Watts Bar Unit 2 and Bellefonte Nuclear Plants which are now in a deferred status. Hartsville, Phipps Bend, and Yellow Creek Nuclear Plants have been canceled. Much of TVA's experience has been gained from early and continuing participation in nuclear power studies. In 1946, TVA took part 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 system. In 1960, TVA agreed to operate the Experimental Gas-Cooled Reactor for the Atomic Energy Commission at Oak Ridge, Tennessee and developed a technical and operating staff. Many of these trained and experienced people were assigned to TVA engineering and operating organizations were directly involved in the planning, design, construction, and operation of the Sequoyah Nuclear Plant. S1-8.doc 1.8-1

SQN TABLE OF CONTENTS Section Title Page 2.0 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1-1 2.1.1 SITE LOCATION 2.1-1 2.1.2 SITE DESCRIPTION 2.1-1 2.1.2.1 Exclusion Area Control 2.1-1 2.1.2.2 Boundaries for Establishing Effluent Release Limits 2.1-2 2.1.2.3 The Restricted Area 2.1-2 2.1.3 POPULATION AND POPULATION DISTRIBUTION 2.1-2 2.1.3.1 Population Within Ten Miles 2.1-2 2.1.3.2 Population Within 50 Miles 2.1-3 2.1.3.3 Low Population Zone 2.1-3 2.1.3.4 Transient Population 2.1-3 2.1.3.5 Population Center 2.1-4 2.1.3.6 Public Facilities and Institutions 2.1-4 2.1.4 USES OF ADJACENT LANDS AND WATERS 2.1-4 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 2.2-1 2.2.1 LOCATION AND ROUTES 2.2-1 2.

2.2 DESCRIPTION

OF PRODUCTS 2.2-1 2.2.3 EVALUATIONS 2.2-2 2.2.3.1 Evaluation of Explosion Hazards 2.2-2 from Nearby Transportation Routes 2.2.3.2 Evaluation of Barge Impact with 2.2-4 the ERCW Intake Structure 2.2.3.3 Evaluation of Hazards from Air 2.2-7 Traffic 2.2.3.4 Evaluation of the Accidental Release of Toxic 2.2-7 Gases from Onsite Storage Facilities 2.2.3.5 Evaluation of the Accidental Release 2.2-7 of Toxic Gases from Offsite Storage Facilities 2.2.3.6 Evaluation of the Upstream Release of Corrosive 2.2-8 Liquids or Oils on the ERCW Intake Structure 2.2.3.7 Evaluation of the Potential for 2.2-8 Damage to Equipment or Structures Important to Reactor Safety in the Event of the Collapse of Cooling Towers S2-0toc.doc 2 -1

SQN TABLE OF CONTENTS Section Title Page 2.2.3.8 Evaluation of a Release on the Tennessee 2.2-8 River of Toxic or Flammable Materials on Plant Safety Features and Control Room Habitability 2.2.3.9 Evaluation of Potential Fire and 2.2-9 Smoke Hazard from Onsite Oil Storage Facilities 2.2.4 FOREST FIRES 2.2-9 2.

2.5 REFERENCES

2.2-9 2.3 METEOROLOGY 2.3-1 2.3.1 REGIONAL METEOROLOGY 2.3-1 2.3.1.1 Data Sources 2.3-1 2.3.1.2 General Climate 2.3-1 2.3.1.3 Severe Weather 2.3-2 2.3.2 LOCAL METEOROLOGY 2.3-3 2.3.2.1 Data Sources 2.3-4 2.3.2.2 Normal and Extreme Values of 2.3-4 Meteorological Parameters 2.3.2.3 Potential Influence of the Plant 2.3-7 and Its Facilities on Local Meteorology 2.3.2.4 Topographical Description 2.3-7 2.3.3 ON-SITE METEOROLOGICAL MEASUREMENT PROGRAM 2.3-8 2.3.3.1 Siting and Description of 2.3-8 Instruments 2.3.3.2 Data Acquisition System 2.3-9 2.3.3.3 Data Recording and Display 2.3-10 2.3.3.4 Equipment Servicing, Maintenance 2.3-11 and Calibration 2.3.3.5 Operational Meteorological Program 2.3-11 2.3.4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES 2.3-12 2.3.4.1 Objective 2.3-12 2.3.4.2 Calculations 2.3-12 2.3.5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES 2.3-16 2.3.5.1 Objective 2.3-16 2.3.5.2 Calculations 2.3-16 2.

3.6 REFERENCES

2.3-17 S2-0toc.doc 2 -2

SQN-24 TABLE OF CONTENTS Section Title Page 2.4 HYDROLOGIC ENGINEERING 2.4-1 2.4.1 HYDROLOGIC DESCRIPTION 2.4-1 2.4.1.1 Site and Facilities 2.4-1 2.4.1.2 Hydrosphere 2.4-1 2.4.2 FLOODS 2.4-5 2.4.2.1 Flood History (Historical) 2.4-5 2.4.2.2 Flood Design Considerations 2.4-5 2.4.3 PROBABLE MAXIMUM FLOOD (PMF) ON 2.4-7 STREAMS AND RIVERS 2.4.3.1 Probable Maximum Precipitation 2.4-7 2.4.3.2 Precipitation Losses 2.4-8 2.4.3.3 Runoff Model 2.4-9 2.4.3.4 Probable Maximum Flood Flow 2.4-11 2.4.3.5 Water Level Determinations 2.4-12 2.4.3.6 Coincident Wind Wave Activity 2.4-13 2.4.4 POTENTIAL DAM FAILURES (SEISMICALLY 2.4-15 AND OTHERWISE INDUCED) 2.4.4.1 Reservoir Description 2.4-15 2.4.4.2 Dam Failure Permutations 2.4-15 2.4.4.3 Unsteady Flow Analysis of Potential Dam Failures 2.4-30 2.4.4.4 Water Level at Plant Site 2.4-31 2.4.5 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING 2.4-31 (HISTORICAL INFORMATION) 2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING 2.4-31 (HISTORICAL INFORMATION) 2.4.7 ICE FLOODING AND LANDSLIDES (HISTORICAL 2.4-31 INFORMATION) 2.4.8 COOLING WATER CANALS AND RESERVOIRS 2.4-31 2.4.8.1 Canals 2.4-31 2.4.8.2 Reservoirs (Historical Information) 2.4-31 2.4.9 CHANNEL DIVERSIONS (HISTORICAL INFORMATION) 2.4-32 2.4.10 FLOODING PROTECTION REQUIREMENTS 2.4-32 2.4.11 LOW WATER CONSIDERATIONS 2.4-32 2.4.11.1 Low Flow in Rivers and Streams 2.4-32 2.4.11.2 Low Water Resulting from Surges, 2.4-33 Seiches, or Tsunamis S2-0toc.doc 2 -3

SQN-24 TABLE OF CONTENTS Section Title Page 2.4.11.3 Historical Low Water 2.4-33 2.4.11.4 Future Control 2.4-33 2.4.11.5 Plant Requirements 2.4-34 2.4.11.6 Heat Sink Dependability Requirements 2.4-34 2.4.12 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS 2.4-35 2.4.13 GROUNDWATER (HISTORICAL INFORMATION) 2.4-39 2.4.13.1 Description and Onsite Use 2.4-39 2.4.13.2 Sources 2.4-39 2.4.13.3 Accident Effects 2.4-40 2.4.13.4 Monitoring or Safeguard Requirements 2.4-41 2.4.13.5 Conclusions 2.4-41 2.4.14 TECHNICAL SPECIFICATIONS AND EMERGENCY 2.4-42 OPERATIONS REQUIREMENTS 2.4.15 REFERENCES 2.4-42 Appendix 2.4A FLOOD PROTECTION PLAN 2.4A-i 2.5 GEOLOGY AND SEISMOLOGY 2.5-1 2.5.1 BASIC GEOLOGIC AND SEISMIC DATA 2.5-1 2.5.1.1 Site Location and Scope of 2.5-1 Exploration 2.5.1.2 Physiography 2.5-1 2.5.1.3 Geologic History 2.5-1 2.5.1.4 Stratigraphy 2.5-2 2.5.1.5 Structure 2.5-2 2.5.1.6 Groundwater 2.5-5 2.5.1.7 Physical Character of the Rocks 2.5-5 2.5.1.8 Foundation Conditions 2.5-6 2.5.1.9 Physical Characteristics of the Soils 2.5-8 2.5.1.10 Detailed Safety-Related Criteria 2.5-9 and Computed Factors of Safety for the Materials Underlying the Foundations for Category I Structures 2.5.1.11 Compaction Criteria for Engineered 2.5-10 Backfill 2.5.2 VIBRATORY GROUND MOTION 2.5-12 2.5.2.1 Regional Tectonics 2.5-12 2.5.2.2 Site Area Tectonics 2.5-15 2.5.2.3 Seismic History 2.5-15 2.5.2.4 Site Seismic Evaluation 2.5-20 2.5.3 SURFACE FAULTING 2.5-22 S2-0toc.doc 2 -4

SQN TABLE OF CONTENTS Section Title Page 2.5.4 STABILITY OF SURFACE MATERIALS 2.5-22 2.5.4.1 Subsidence 2.5-22 2.5.4.2 Zones of Deformed or Weak Materials 2.5-23 2.5.4.3 Bedrock Stresses 2.5-23 2.5.5 STABILITY OF SUBSURFACE MATERIALS 2.5-23 2.5.5.1 Excavations and Backfill 2.5-23 2.5.5.2 Liquefaction Potential 2.5-23 2.5.5.3 Static Analysis 2.5-23 2.5.6 SLOPE STABILITY 2.5-25 2.5.6.1 Slope Characteristics 2.5-25 2.5.6.2 Design Criteria And Analysis 2.5-26 2.5.6.3 Compaction Specifications 2.5-27 2.

5.7 REFERENCES

2.5-27

2.6 CONCLUSION

S 2.6-1 S2-0toc.doc 2 -5

SQN LIST OF TABLES Number Title 2.1.1-1 Coordinates of Unit 1 Reactor Building Centerline 2.1.2-1 Deleted 2.1.2-2 Deleted 2.1.3-1 1970 Population Distribution Within Ten Miles of Site 2.1.3-2 1980 Population Distribution Within Ten Miles of Site 2.1.3-3 1985 Population Distribution Within Ten Miles of Site 2.1.3-4 1990 Population Distribution Within Ten Miles of Site 2.1.3-5 2000 Population Distribution Within Ten Miles of Site 2.1.3-6 2010 Population Distribution Within Ten Miles of Site 2.1.3-6a 2020 Population Distribution Within Ten Miles of Site 2.1.3-7 1970 Population Distribution Within Fifty Miles of Site 2.1.3-8 1980 Population Distribution Within Fifty Miles of Site 2.1.3-9 1985 Population Distribution Within Fifty Miles of Site 2.1.3-10 1990 Population Distribution Within Fifty Miles of Site 2.1.3-11 2000 Population Distribution Within Fifty Miles of Site 2.1.3-12 2010 Population Distribution Within Fifty Miles of Site 2.1.3-12a 2020 Population Distribution Within Fifty Miles of Site 2.1.3-13 1970 Estimated Peak Hour Recreation Visits Within Ten Miles of Site 2.1.3-14 1980 Estimated Peak Hour Recreation Visits Within Ten Miles of Site 2.1.3-15 1985 Estimated Peak Hour Recreation Visits Within Ten Miles of Site 2.1.3-16 1990 Estimated Peak Hour Recreation Visits With Ten Miles of Site 2.1.3-17 2000 Estimated Peak Hour Recreation Visits Within Ten Miles of Site 2.1.3-18 2010 Estimated Peak Hour Recreation Visits Within Ten Miles of Site 2.1.3-19 2020 Estimated Peak Hour Recreation Visits Within Ten Miles of Site S2-0toc.doc 2-6

SQN LIST OF TABLES Number Title 2.1.3-20 Educational Institutions in Vicinity of Sequoyah Nuclear Plant 1990-2020 2.1.4-1 Farm Oriented Land Use 2.1.4-2 Crops Harvested 2.2.2-1 Hazardous River Traffic that Passes Sequoyah Nuclear Plant, 1980-1990 (Tons) - U.S. Army Corps of Engineers Data 2.2.2-1a Hazardous River Traffic that Passes Sequoyah Nuclear Plant 1990 (TVA Survey) 2.2.2-1b Volunteer Army Ammunition Plant Storage and Transportation of Munitions 2.2.3-1 Barge Freight Traffic Passing Sequoyah Nuclear Plant Site Tennessee River Mile 484.5 2.2.3-2 Tennessee River Traffic Passing Sequoyah Nuclear Plant 2.2.3-3 Gasoline Barge Receipts at Port at Knoxville (In Net Tons) 2.3.2-1 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - January 1, 1972 - December 31, 1975 2.3.2-2 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - January (72-75) 2.3.2-3 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - February (72-75) 2.3.2-4 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - March (72-75) 2.3.2-5 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - April (72-75) 2.3.2-6 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - May (72-75) 2.3.2-7 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - June (72-75) 2.3.2-8 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - July (72-75) 2.3.2-9 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - August (72-75) S2-0toc.doc 2-7

SQN LIST OF TABLES Number Title 2.3.2-10 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - September (72-75) 2.3.2-11 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - October (72-75) 2.3.2-12 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - November (72-75) 2.3.2-13 Joint Percentage Frequencies of Wind Speed by Direction Disregarding Stability Class - December (72-75) 2.3.2-14 Wind Direction Persistence Data; Disregarding Stability; Sequoyah Nuclear Plant January 1, 1972 - December 31, 1975 2.3.2-15 Temperature; Sequoyah Nuclear Plant; April 2, 1971-March 31, 1972 2.3.2-16 Temperature; Chattanooga, Tennessee 2.3.2-17 Absolute Humidity; Sequoyah Nuclear Plant; April 2, 1971 - March 31, 1972 2.3.2-18 Relative Humidity; Sequoyah Nuclear Plant; April 2, 1971 - March 31, 1972 2.3.2-19 Precipitation; Friendship School, Tennessee; 1948 - 1967 2.3.2-20 Snowfall; Chattanooga, Tennessee 2.3.2-21 Heavy Fog; Chattanooga, Tennessee 2.3.2-22 Percent Occurrence of Atmospheric Stability; Sequoyah Nuclear Plant; January 1, 1972 - December 31, 1975 2.3.2-23 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class A - January 1, 1972 - December 31, 1975 2.3.2-24 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class B - January 1, 1972 - December 31,1975 2.3.2-25 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class C - January 1, 1972 - December 31, 1975 S2-0toc.doc 2-8

SQN-21 LIST OF TABLES Number Title 2.3.2-26 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class D - January 1, 1972 - December 31, 1975 2.3.2-27 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class E - January 1, 1972 - December 31, 1975 2.3.2-28 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class F - January 1, 1972 - December 31, 1975 2.3.2-29 Joint Percentage Frequencies of Wind Speed by Wind Direction for Stability Class G - January 1, 1972 - December 31, 1975 2.3.2-30 Percent of Observations in Each Stability Class Hourly Average and End of Hour Temperature Differences (T) (May 1975 - April 1976) 2.3.2-31 Joint Percentage Frequencies of Wind Direction and Wind speed for Different Stability Classes, Stability Class A (May 1975 - April 1976) 2.3.2-32 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class A (May 1975 - April 1976) 2.3.2-33 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class B (May 1975 - April 1976) 2.3.2-34 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class B (May 1975 - April 1976) 2.3.2-35 Joint Percentage Frequencies of Wind Direction and wind Speed for Different Stability Classes, Stability Class C (May 1975 - April 1976) 2.3.2-36 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class C (May 1975 - April 1976) 2.3.2-37 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class D (May 1975 - April 1976) 2.3.2-38 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class D (May 1975 - April 1976) 2.3.2-39 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class E (May 1975 - April 1976) 2.3.2-40 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class E (May 1975 - April 1976) 2.3.2-41 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class F (May 1975 - April 1976) S2-0toc.doc 2-9

SQN LIST OF TABLES Number Title 2.3.2-42 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class F (May 1975 - April 1976) 2.3.2-43 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class G (May 1975 - April 1976) 2.3.2-44 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class G (May 1975 - April 1976) 2.3.2-45 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class A (May 1975 - April 1976) 2.3.2-46 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class A (May 1975 - April 1976) 2.3.2-47 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class B (May 1975 - April 1976) 2.3.2-48 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class B (May 1975 - April 1976) 2.3.2-49 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class C (May 1975 - April 1976) 2.3.2-50 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class C (May 1975 - April 1976) 2.3.2-51 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class D (May 1975 - April 1976) 2.3.2-52 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class D (May 1975 - April 1976) 2.3.2-53 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class E (May 1975 - April 1976) 2.3.2-54 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class E (May 1975 - April 1976) 2.3.2-55 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class F (May 1975 - April 1976) 2.3.2-56 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class F (May 1975 - April 1976) 2.3.2-57 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class G (May 1975 - April 1976) S2-0toc.doc 2-10

SQN LIST OF TABLES Number Title 2.3.2-58 Joint Percentage Frequencies of Wind Direction and Wind Speed for Different Stability Classes, Stability Class G (May 1975 - April 1976) 2.3.4-1 Distances from Release Zones or Points to Exclusion Area Boundary 2.3.4-2 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at Exclusion Area Boundary Due to Ground-Level Releases from Release Zone 1 2.3.4-3 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at Exclusion Area Boundary Due to Ground-Level Releases from Release Zone 2 2.3.4-4 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at Exclusion Area Boundary Due to Ground-Level Releases from Release Zone 3 2.3.4-5 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at 556 Meters (Minimum Exclusion Area Boundary Distance) Due to Ground-Level Releases from Release Zone 1 2.3.4-6 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at 600 Meters (Minimum Exclusion Area Boundary Distance) Due to Ground-Level Releases from Release Zone 2 2.3.4-7 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at 509 Meters (Minimum Exclusion Area Boundary Distance) Due to Ground-Level Releases from Release Zone 3 2.3.4-8 Atmospheric Dispersion Factors Frequency Distribution; Calculated 1-Hour-Average Atmospheric Dispersion Factors at Outer Boundary of Low Population Zone Due to Ground-Level Releases from a Location Representative of Release Zone 1, Release Zone 2, and Release Zone 3 2.3.4-9 Atmospheric Dispersion Factors Frequency Distribution; Calculated 8-Hour-Average Atmospheric Dispersion Factors at Outer Boundary of Low Population Zone Due to Ground-Level Releases from a Location Representative of Release Zone 1, Release Zone 2, and Release Zone 3 2.3.4-10 Atmospheric Dispersion Factors Frequency Distribution; Calculated 16-Hour-Average Atmospheric Dispersion Factors at Outer Boundary of Low Population Zone Due to Ground-Level Releases from a Location Representative of Release Zone 1, Release Zone 2, and Release Zone 3 S2-0toc.doc 2-11

SQN-17 LIST OF TABLES Number Title 2.3.4-11 Atmospheric Dispersion Factors Frequency Distribution; Calculated 3-Day-Average Atmospheric Dispersion Factors at Outer Boundary of Low Population Zone Due to Ground-Level Releases from a Location Representative of Release Zone 1, Release Zone 2, and Release Zone 3 2.3.4-12 Atmospheric Dilution Factors Frequency Distribution; Calculated 26-Day-Average Atmospheric Dilution Factors at Outer Boundary of Low Population Zone Due to ground-Level Releases from a Location Representative of Release Zone 1, Release Zone 2, and Release Zone 3 2.3.4-13 Fifth Percentile Atmospheric Dispersion Factors (c/Q's) for Comparative Data

           - Hourly Average and End of Hour Temperature Differences (T) 2.3.4-14    Average Annual Dispersion Factors, c/Q (s/m3), Downwind Distance (miles) 2.4.1-1     Facts about Major TVA Dams and Reservoirs 2.4.1-2     Facts about Non-TVA Dam and Reservoir Projects 2.4.1-3     Flood Detention Capacity - TVA Projects above Sequoyah Nuclear Plant 2.4.1-4     Public and Industrial Surface Water Supplies Withdrawn from the 98.6 Mile Reach of the Tennessee River Between Dayton, Tennessee, and Meade Corporation, Stevenson, Alabama 2.4.1-5     Dam Safety Modification Status (Hydrologic) 2.4.3-1     Probable Maximum Storm Rainfall and Precipitation Excess 2.4.3-2     Unit Hydrograph Data 2.4.4-1     Floods from Postulated Seismic Failure of Upstream Dams 2.4.13-1    Well and Spring Inventory Within 2-Mile Radius of Sequoyah Nuclear Plant Site 2.4.13-2    Groundwater Supplies Within 20-Mile Radius of the Plant Site 2.5.1-1     Summary of In Situ Up-Hole Dynamic Testing Reactor Foundation Area 2.5.1-2     Summary of In Situ Cross-Hole Dynamic Testing Reactor Foundation Area 2.5.1-3     Equation for Dynamic Modulus of Elasticity 2.5.1-4     Summary of Grouting 2.5.1-5     Static and Dynamic Rock-Bearing Capacities for Rock Supported Category I Structures S2-0toc.doc                            2-12

SQN LIST OF TABLES Number Title 2.5.1-6 Soil-Bearing Capacities and Factors of Safety for Soil Supported Category I Structures 2.5.1-7 Summary of Earthfill Test Data - Density 2.5.1-8 Summary of Earthfill Test Data - Moisture Content 2.5.1-9 Sequoyah Nuclear Plant - Summary of In-Situ Soil Down-Hole - Dynamic Testing Diesel Generator Building 2.5.1-10 Sequoyah Nuclear Plant - Onsite Storage Facility Dynamic Soil Test Array SD-1 2.5.1-11 Sequoyah Nuclear Plant - Onsite Test Facility Dynamic Soil Test Array SD-3 2.5.1-12 Sequoyah Nuclear Plant - ERCW Pipeline In-Situ Down-Hole Soil Dynamics 2.5.1-13 Sequoyah Nuclear Plant - Additional Diesel Generator Building 2.5.1-14 Sequoyah Nuclear Plant - Primary Refueling Water Tanks Seismic Refraction Survey In-Situ Elastic Properties 2.5.2-1 Historical Earthquake Listing - 200 Mi. Radius Around - 85.1 W Longitude - 35.2 N Latitude, February 17, 1982 2.6-1 Sequoyah Nuclear Plant Site Characteristics S2-0toc.doc 2-13

SQN LIST OF FIGURES Number Title 2.1.1-1 Location of Sequoyah Nuclear Plant Site 2.1.1-2 Features Within 50 Miles 2.1.1-3 Features Within 10 Miles 2.1.2-1 General Site Plan 2.1.2-2 Site Boundary 2.1.3-1 Low Population Zone 2.1.4-1 Land Use in Plant Vicinity 2.1.4-2 Forest Types and Cover - Bradley and Hamilton Counties 2.3.1-1 Normal Sea Level Pressure Distribution over North America and the North Atlantic Ocean 2.3.1-2 Total Number of Forecast Days of High Meteorological Potential for Air Pollution in a 5 Year Period 2.3.2-1 Environmental Data Station Location 2.3.2-2 Climatological Data Sources 2.3.2-3 Wind Rose, 10 M Wind, All Stability Classes January 1, 1972 - December 31, 1975 2.3.2-4 Wind Rose, 10 M Wind, All Stability Classes January (72-75) 2.3.2-5 Wind Rose, 10 M Wind, All Stability Classes February (72-75) 2.3.2-6 Wind Rose, 10 M Wind, All Stability Classes March (72-75) 2.3.2-7 Wind Rose, 10 M Wind, All Stability Classes April (72-75) 2.3.2-8 Wind Rose, 10 M Wind, All Stability Classes May (72-75) 2.3.2-9 Wind Rose, 10 M Wind, All Stability Classes June (72-75) 2.3.2-10 Wind Rose, 10 M Wind, All Stability Classes July (72-75) 2.3.2-11 Wind Rose, 10 M Wind, All Stability Classes August (72-75) 2.3.2-12 Wind Rose, 10 M Wind, All Stability Classes September (72-75) S2-0toc.doc 2-14

SQN LIST OF FIGURES Number Title 2.3.2-13 Wind Rose, 10 M Wind, All Stability Classes October (72-75) 2.3.2-14 Wind Rose, 10 M Wind, All Stability Classes November (72-75) 2.3.2-15 Wind Rose, 10 M Wind, All Stability Classes December (72-75) 2.3.2-16 10 M Wind, 9 and 46 M Temp - Stability Class A January 1, 1972 - December 31, 1975 2.3.2-17 Wind Rose 10 M Wind, 9 and 46 M Temp - Stability Class B January 1, 1972

           - December 31, 1975 2.3.2-18    Wind Rose 10 Wind, 9 and 46 M Temp - Stability Class C January 1, 1972 -

December 31, 1975 2.3.2-19 Wind Rose 10 M Wind, 9 and 46 M Temp - Stability Class D January 1, 1972

           - December 31, 1975 2.3.2-20    Wind Rose 10 M Wind, 9 and 46 M Temp - Stability Class E January 1, 1972
           - December 31, 1975 2.3.2-21    Wind Rose 10 M Wind, 9 and 46 M Temp - Stability Class F January 1, 1972
           - December 31, 1975 2.3.2-22    Wind Rose 10 M Wind, 9 and 46 M Temp - Stability Class G January 1, 1972
           - December 31, 1975 2.3.2-23    Topography - Within 5 mile Radius (9 sheets) 2.4.1-1     Topographic Map, Plant Vicinity 2.4.1-2     Tennessee River Basin Mean Annual Precipitation, 30-yr Period, 1935 - 1964 2.4.1-3     Multiple - Purpose Reservoir Operations Chickamauga Project (14 sheets) 2.4.2-1     Flood Distribution Diagram, Chattanooga, TN 2.4.3-1     Probable Maximum March Isohyets (21,400-sq mi downstream), 1st 6 hours (in.)

2.4.3-2 Probable Maximum March Isohyets (7980 sq mi), 1st 6 hrs (in.) 2.4.3-3 Rainfall-Time Distribution, Adopted Standard Mass Curve S2-0toc.doc 2-15

SQN-17 LIST OF FIGURES Number Title 2.4.3-4 72-Hr March Probable Maximum Storm Depths (in) 2.4.3-5 Hydrologic Model Unit Areas 2.4.3-6 6 Hour Unit Hydrographs (11 sheets) 2.4.3-7 1973 Flood - Chickamauga Reservoir Unsteady Flow Model Verification 2.4.3-8 Steady State Model Verification, Watts Bar Dam Tailwater Rating Curve 2.4.3-9 Hydrologic Model Verification 1973 Flood 2.4.3-10 Hydrologic Model Verification 1963 Flood 2.4.3-11 Sequoyah Nuclear Plant Probable Maximum Flood Discharge 2.4.3-12 Sequoyah Nuclear Plant Probable Maximum Flood Elevation 2.4.3-13a General Grading for Site Drainage 2.4.3-14 ERCW Pump Station Location 2.4.3-15 Sequoyah Nuclear Plant NNW Wind Wave Fetch 2.4.3-16 Sequoyah Nuclear Plant NE Wind Wave Fetch 2.4.3-17 Topography Surrounding Diesel Generator Bldg and Cooling Towers 2.4.4-1 Watts Bar Dam Powerhouse & Spillway Results of Analysis for Operating Basis Earthquake 2.4.4-3 Spillway Gate Positions for 25 Year Flood 1/2 Probable Maximum Flood S2-0toc.doc 2-16

SQN LIST OF FIGURES Number Title 2.4.4-4 Powerhouse and Spillway Fort Loudoun Dam Results of Analysis for 1/2 SSE 2.4.4-5 Embankment Fort Loudoun Dam Results of Analysis for 1/2 SSE 2.4.4-6 Nonoverflow and Spillway Tellico Dam Results of Analysis Operating Basis Earthquake 2.4.4-7 Embankment Tellico Dam Results of Analysis for 1/2 SSE 2.4.4-8 Spillway and Nonoverflow Norris Dam Results of Analysis for 1/2 SSE 2.4.4-9 Norris Dam Analysis for 1/2 SSE and 1/2 Maximum Possible Flood 2.4.4-10 Spillway and Nonoverflow Cherokee Dam Results of Analysis for 1/2 SSE 2.4.4-11 Embankment Cherokee Dam Results of Analysis for 1/2 SSE 2.4.4-12 Cherokee Dam Assumed Condition of Dam After Failure - 1/2 SSE and 1/2 Maximum Possible Flood 2.4.4-13 Spillway and Overflow Douglas Dam Results of Analysis for 1/2 SSE 2.4.4-14 Saddle Dam No. 1 Douglas Dam Results of Analysis for 1/2 SSE 2.4.4-15 Douglas Dam Assumed Condition of Dam After Failure - 1/2 SSE and 1/2 Maximum Possible Flood S2-0toc.doc 2-17

SQN-17 LIST OF FIGURES Number Title 2.4.4-16 Fontana Dam - Assumed Condition of Dam After 1/2 SEE and 1/2 Maximum Possible Flood 2.4.4-17 Fontana Project - Concrete Strengthening of Blks 33, 34, and 35 2.4.4-18 1/2 SSE with Epicenter Within Area Shown 2.4.4-21 Seismic Flood Analysis Fontana Plus Hiwassee System in 1/2 SSE - Sequoyah Plant 2.4.4-22 Deleted by Amendment 6 2.4.4-23 Deleted by Amendment 6 2.4.4-24 Spillway Fort Loudoun Dam Results of Analysis for SSE 2.4.4-25 Embankment Fort Loudoun Dam Results of Analysis for SSE 2.4.4-26 Fort Loudoun Dam Assumed Condition of Dam After Failure - SSE Combined With 25 Year Flood 2.4.4-27 Tellico Dam - Tellico Project Assumed Condition of Dam After Failure - SSE Combined with 25 Year Flood 2.4.4-28 SSE Plus 25 Year Flood, Judged Condition of Dam After Failure, Norris Dam 2.4.4-29 SSE With Epicenter in North Knoxville Vicinity 2.4.4-30 Flood Hydrographs 2.4.4-31 SSE With Epicenter in West Knoxville Vicinity S2-0toc.doc 2-18

SQN-17 LIST OF FIGURES Number Title 2.4.4-37 General Plan, Elevation and Sections, Watts Bar 2.4.4-38 Topographic Map, General Area - Watts Bar Dam 2.4.4-39 Location, Plan and Section - Watts Bar 2.4.8-1 Grading Plan Intake Channel 2.4.13-1 Location of Wells and Springs Within 2 Mile Radius 2.4.13-2 Sequoyah Nuclear Plant Site Monitoring Well Locations and Generalized Water-Table Map 2.4A-5 Douglas PMF Failure Wave at Sequoyah Plant 2.5.1-1 Physiographic Map of Plant Area 2.5.1-2 Geologic and Tectonic Map of Plant Area 2.5.1-3 Geologic Investigations Geologic Map of Plant Site 2.5.1-4 Valley and Ridge Providence Seismic Reflection Profile 2.5.1-5 Geologic Sections W-30+00 through W-36+100 S2-0toc.doc 2-19

SQN LIST OF FIGURES Number Title 2.5.1-6 Geologic Sections W-38+00, A-A and B-B 2.5.1-7 Geologic Sections N-68+00 through W-74+00 2.5.1-8 Geologic Sections N-76+00 through N-80+00 2.5.1-9 Foundation Treatment First Stage Grouting 2.5.1-10 Foundation Treatment Second Stage Grouting 2.5.1-11 Reactor Buildings No. 1 and 2 Area, Looking South 2.5.1-12 Excavation and Backfill - Category I Structures 2.5.1-12a Excavation and Backfill - Category I Structures 2.5.1-12b Excavation and Backfill - Category I Structures 2.5.1-13 Essential Soil Investigations In-Situ 2.5.1-13a Essential Soil Investigations In-Situ 2.5.1-14 Standard Proctor Compaction Borrow Area 2.5.1-15 Standard Proctor Compaction Borrow Area 2.5.2-1 Location of Earthquake Epicenters 2.5.2-2 Map Showing the Extent of Earthquake Disturbances in the New Madrid Area in 1811-12 2.5.2-3 Isoseismals of the Charleston Earthquake 2.5.2-4 East Tennessee Earthquake of April 17, 1913 2.5.2-5 Earthquake of February 21, 1916 2.5.2-6 The Alabama Earthquake of October 18, 1916 2.5.2-7 Southern Appalachian Earthquake of October 1924 2.5.2-8 Appalachian Earthquake of November 2, 1928 2.5.2-9 Chattanooga Earthquake October 19, 1940 S2-0toc.doc 2-20

SQN LIST OF FIGURES Number Title 2.5.2-10 East Tennessee Earthquake of July 13, 1969 2.5.2-11 Comparison of Response Spectra for Safe Shutdown Earthquake, 1/2% Damping 2.5.2-12 Comparison of Response Spectra for Safe Shutdown Earthquake, 1% Damping 2.5.2-13 Comparison of Response Spectra for Safe Shutdown Earthquake, 2% Damping 2.5.2-14 Comparison of Response Spectra for Safe Shutdown Earthquake, 5% Damping 2.5.5-1 Sequoyah Nuclear Plant - Diesel Generator Building - Time Settlement Curves 2.5.6-1 Diesel Generator Bldg - Sequoyah Nuclear Plant 2.5.6-2 Section of Forebay and Intake Slope Sequoyah Nuclear Plant Pumping Station S2-0toc.doc 2-21

SQN-20 2.0 SITE CHARACTERISTICS Chapter 2 provides information on the Sequoyah Nuclear Plant site, its environs and environment, and presents the results of studies that have been made to evaluate the physical characteristics of the site which influence the safety-related design bases of the plant. The minimum exclusion and low population zone distances as defined by 10 CFR Part 100 are approximately 1825 feet and three miles respectively. The population center distance which is the distance to the nearest corporate limit of the city of Chattanooga, Tennessee, is approximately 7.5 miles southwest. 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 Site Location The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The coordinates of the plant site are given in Table 2.1.1-1. Figure 2.1.1-1 shows the site in relation to other TVA projects. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant. Refer to Figure 2.1.1-2 for the regional features within 50 miles of the site. 2.1.2 Site Description The Sequoyah Nuclear Plant site comprises approximately 525 acres (land above normal pool elevation of 683.0 ft MSL) which are owned, including mineral rights, by the United States and in the custody of TVA. A general plan of the plant layout is shown in Figure 2.1.2-1. The distance from the reactor building (containment) to the nearest point on the boundary of the exclusion area (minimum exclusion area distance) is approximately 1825 feet (556 meters). The site boundary is considered to be the boundary of the exclusion area. 2.1.2.1 Exclusion Area Control There are no residences, commercial operations, or public recreational areas within the Sequoyah Nuclear Plant exclusion area boundary shown in Figure 2.1.2-2. The Sequoyah Training Center is within the TVA exclusion area and outside the security barrier. No public railroads or major highways penetrate the exclusion area boundary. Two rural county roads, Igou Ferry and Stonesage, penetrate the western boundary of TVA property and run adjacent to it for a short distance before leaving the site. Igou Ferry Road connects with Hixson Pike which follows the western shore of Chickamauga Lake and joins state route 153 just north of Chickamauga Dam. The plant access road crosses Igou Ferry Road at the exclusion area boundary and eventually connects with US Highway 27 near Soddy-Daisy, Tennessee. TVA has absolute authority for the exclusion of personnel and property within the exclusion area which includes marking of the S2-1.doc 2.1-1

SQN boundaries per 10 CFR 73. The control of personnel access to the exclusion area during emergencies is discussed in the Radiological Emergency Plan for the Sequoyah Nuclear Plant. 2.1.2.2 Boundaries for Establishing Effluent Release Limits The effluent boundary (or unrestricted area boundary) is shown in Figure 2.1.2-2. The boundary of the Unrestricted Area (as defined in 10 CFR 20) is the same as the site boundary, but does not include the area over bodies of water. In accordance with the SQN Technical Specifications, limits for gaseous effluent releases are established for areas at or beyond the unrestricted area boundary using the methodology of the Offsite Dose Calculation Manual (ODCM). The distances from the plant to these areas are listed in Table 11.3.9-1 consistent with the ODCM. Routine releases of radioactivity meet the requirements of 10 CFR 20 and 10 CFR 50, Appendix I. 2.1.2.3 The Restricted Area An area inside the exclusion area boundary is designated as the Restricted Area (as defined in 10 CFR 20). Access to this area is controlled for the purpose of protection of individuals from exposure to radiation and radioactive materials. The restricted area boundary can be adjusted, or temporary restricted areas established, as necessary, for the purpose of radiation protection. 2.1.3 Population and Population Distribution Present and projected population information is contained in this section. Population data for 1985 are based on the Provisional Estimates of the Population of Counties, July 1, 1985. Population data for 1990 are based on the "1990 Census of Population" for Tennessee, North Carolina, Georgia, and Alabama. Projected population data are based on "County Projection to 2040" by the Regional Economic Analysis Division, Bureau of Economic Analysis, U.S. Department of Commerce, 1992. The allocation of county population into the various segments was based on a count of dwelling units from 1985 low-level aerial photography within ten miles of the site and census and 1:250,000 topographic maps for the remaining area. 2.1.3.1 Population Within 10 Miles Population is distributed rather unevenly within 10 miles of the Sequoyah Nuclear Plant site. Over 50 percent of the 1990 population was in only seven sectors of the 5- to 10-mile range. These sectors are from S to and including NW (going clockwise around the compass). This concentration is a reflection of suburban Chattanooga and the town of Soddy-Daisy. Resident population in the remaining area is sparse and scattered with the exception of the 4-5 WSW annular segment. This pattern is projected to continue in the future with 55 percent of the total 2020 population being contained in this same portion of the 10-mile area. In addition, the 3-4 WSW annular segment is also projected for significant growth. The 0-10 mile population distributions for 1970 through 2020 are given in Tables 2.1.3-1 through 2.1.3-6a and are keyed to the various distances and directions shown on Figure 2.1.1-3. S2-1.doc 2.1-2

SQN 2.1.3.2 Population Within 50 Miles Although the site is located in southeastern Tennessee, the area within a 50-mile radius of the site encompasses portions of northwestern Georgia, northeastern Alabama, and a small portion of southwestern North Carolina. The largest population concentration within 50 miles of the site is the city of Chattanooga, with a 1990 population of 152,466. The northernmost limits of the urbanization around Chattanooga are approximately four miles west-southwest of the plant site. Four smaller population centers (population of 10,000 to 50,000) are scattered around the area. The closest is Cleveland, Tennessee, about 13 miles east-southeast of the plant site with 1990 population of 30,354. In the 30- to 40-mile range are Dalton, Georgia, to the south-southeast (1990 population 21,761) and Athens, Tennessee, to the east-northeast (1990 population 12,054). McMinnville, Tennessee, with a 1990 population of 11,194, is 50 miles northwest of the plant site. In addition, the town of Soddy-Daisy (1990 pop. 8400) is located approximately 6 miles from the site. Development throughout the rest of the region consists primarily of smaller towns dispersed throughout low density rural development. Most of them serve as small retail or service centers for the surrounding farms, although a number are developing an industrial base. Tables 2.1.3-7 through 2.1.3-12a show the 0-50 mile population distributions for the year 1970 through 2020 for various distances and directions shown on Figure 2.1.1-2. 2.1.3.3 Low Population Zone The low population zone distance as defined in 10 CFR Part 100 has been chosen to be three miles (4,828 meters). The population of this area (2,005 in 1970) and the population density (71 people per square mile in 1970) are both low. In addition, this area is of such size that in the unlikely event of a serious accident there is a reasonable probability that appropriate measures could be taken to protect the health and safety of the residents. Specific provisions for the protection of this area were considered in the development of the Sequoyah Nuclear Plant site emergency plan. The present and projected population figures for this area are included in Tables 2.1.3-1 through 2.1.3-6. Features of the area within the low population zone distances are shown on Figure 2.1.3-1. 2.1.3.4 Transient Population Transient population within 10 miles of the plant is made up primarily of visitors to the various recreation facilities along the shoreline of the Chickamauga Reservoir. Figure 2.1.1-3 shows the location of the three primary public recreation facilities: Harrison Bay and Booker T. Washington State Parks and the Chester Frost County Park. In addition, there are many commercial marinas, group camps, and cottage developments as well as small formal and informal public access areas along the reservoir shoreline. Peak hour attendance at these facilities was estimated by the TVA Recreation Resources Branch and is shown in Tables 2.1.3-11 through 2.1.3-16 for various distances and direction. The attendance at the major facilities is distributed to various segments according to where specific activities are located within the total park. S2-1.doc 2.1-3

SQN The transient population on the site is very limited. The Sequoyah Energy Connection is less than one mile southwest of the plant and it accommodates visitor groups of up to about 75. This visitation is not reflected in Tables 2.1.3-13 through 2.1.3-19. 2.1.3.5 Population Center The nearest population center (as defined in 10 CFR Part 100) is Chattanooga, Tennessee, located as described previously. 2.1.3.6 Public Facilities and Institutions Schools are the only public institutions containing significant population concentrations within 10 miles of the site. Their names, locations, and the 1990, 1993, 1997, and projected enrollments are contained in Table 2.1.3-20. To project enrollments, TVA consulted with the Hamilton County and Bradley County school officials. 2.1.4 Uses of Adjacent Lands and Waters Land use in the vicinity of the proposed plant site can be examined best by dividing the area into four parts (see Figure 2.1.4-1): (1) the area west of Chickamauga Reservoir and north of the plant; (2) the area west of Chickamauga Reservoir, north of the city of Chattanooga, and southwest of the plant; (3) the area east of Chickamauga Reservoir and southeast of Harrison Bay and the Volunteer Army Ammunition Plant (VAA Plant); and (4) the area east of Chickamauga Reservoir and northeast of Harrison Bay and the VAA Plant. Area No. 1 With the exception of the community of Soddy-Daisy, the area west of Chickamauga Reservoir and north of the site is sparsely settled. Development consists of scattered dwellings with some associated small-scale farming. Public access areas, campgrounds, boat docks, and an occasional small residential subdivision have been developed along the reservoir shoreline in scattered locations. The Soddy, Possum, and Sale Creek embayments are especially popular with fishermen and family boaters. U.S. Highway 27 parallels the reservoir approximately five miles to the west. Soddy-Daisy, with a 1985 population of 8,400, is located along this highway about six miles from the plant. This area is projected to experience a number of changes by the year 2010. One that was recently completed is the upgrade of U.S. 27 into a major north-south highway connecting northern Hamilton County with downtown Chattanooga. It has replaced the old two lane road and reduced commuting time significantly. Much more residential development is forecast for this area because of that, but not to the point that population densities will be significant. Contributing to the projected development are two other proposals. First is the provision of sewer to part of the area, which would increase both the rate and density of growth. Second is a proposed east-west road crossing the lake just north of the Sale Creek embayment. It would connect Cleveland with highways in Sequatchie County. If built, it would stimulate development along its route and a major concentration of commercial and high-density residential at its S2-1.doc 2.1-4

SQN intersection with U.S. 27 if the proposed sewers are built. Another significant proposed land use is an industrial park between the nuclear plant and Hixson Pike. It too is dependent on the provision of sewers. It would likely house light manufacturing plants. Area No. 2 The area west of Chickamauga Reservoir between the Chattanooga city limits and the site has experienced considerable residential growth in the last few years. The area is characterized by considerable vacant land interspersed with high quality residential subdivisions. Much of the new residential development is concentrated between the Hixson and Dallas Hills communities and along the reservoir shoreline. Public recreation facilities are dominated by the 280-acre Chester Frost County Park (formerly Hamilton County Park) receiving over 250,000 visits annually. North Chickamauga Creek in the 9-10 mile range has been designated as a "greenway" with the development of trails and day use facilities near the mouth of the creek underway. Residential development is expected to advance steadily in this general area in the future because of the improvement to U.S. 27 discussed in Area 1. In summary, this area is considered a growth area in Hamilton County. As the population projections indicate, increases are expected throughout the area. In the past the tendency has been to concentrate along the reservoir shoreline. This trend is expected to continue; but, as the shoreline becomes developed, growth is expected to take place in the form of infilling throughout the entire area utilizing the now vacant land. Area No. 3 Until 1977, when explosives production ceased, the VAA Plant had been a significant barrier to growth in this area because of environmental problems. Since then, residential development has picked up in the area, especially in the vicinity of the lake. There is also substantial commercial and light industrial use along State Highways 58 and 153. This pattern of growth is expected to continue within the natural limitation of the area, which is primarily poor soil for septic tank drain fields. In addition, a significant portion of the VAA site is being marketed for use as an industrial park, which should also increase the development in this area. Sewers are projected for this area, which would increase the rate and density of residential development. The primary recreation feature is the Booker T. Washington State Park, which had 393,000 visits in 1987. Area No. 4 As in Area No. 3, much of this area also has been affected in the past by the VAA Plant, with residential development picking up in recent years. However, the basic character of the area is rural, with the exception of the Harrison Bay State Park in the two- to five-mile range along the eastern shoreline. In addition to numerous farms, there are scattered private cottages and houses in the vicinity of the park. Public campsites are also located at Skull Island and Grasshopper Creek Park. From 7 to 10 miles in the vicinity of the city of Cleveland, residential subdivisions have concentrated along existing roads. Also, Interstate 75 is causing readjustments in development through the area. S2-1.doc 2.1-5

SQN-19 At present, Area No. 4 is not a growth area for Chattanooga and sewers are not projected for most of the area. Therefore, due to the hilly terrain and poor soils for drain fields, future residential development is expected to be very low density. However, industrial development at the VAA plant, as mentioned previously, may have an impact in this area. Hamilton and Bradley Counties, Tennessee, fall within a 10-mile radius of the Sequoyah site, having a total land area of approximately 555,000 acres with 159,359 acres of this in farms or about 29 percent of the total land area. On the 1,367 farms in this area, 87,465 acres were found to be used as cropland. A breakdown of the farm oriented land use for each county is given in Table 2.1.4-1. Table 2.1.4-2 tabulates yield and associated land area for various harvested crops. As of 11-1-88, the number of dairy cows within a 5-mile radius of the plant site was 69. In general, the land adjacent to the plant site is suitable dairying land. A land use census is conducted annually by TVA to locate the nearest milk producing animals. In 1988 all animals were cows. A 1980 U.S. Forest Service survey of Tennessee indicates that approximately 51 percent of the land area in Bradley and Hamilton counties is forested and 49 percent is non-forested. These two counties contain 96,600 and 202,710 acres of forest respectively. Growing stock volume in the counties is estimated to be 335.3 million cubic feet, with 51.8 percent softwood and 48.2 percent hardwoods. The general extent and type of forest cover is shown in Figure 2.1.4-2. Chickamauga Reservoir is one of a series of TVA multipurpose reservoirs located on the mainstream of the Tennessee River. The primary project uses are for flood control, navigation and hydropower generation, although extensive secondary uses including industrial and public water supply, commercial and sport fishing, recreation, and disposal of treated wastewater have also developed. Chickamauga Reservoir, which extends from Chickamauga Dam (TRM 471.0) to Watts Bar Dam (TRM 529.9), has been classified by the Tennessee Division of Water Pollution Control for the following uses: municipal water supply, industrial water supply, fish and aquatic life, recreation, irrigation, livestock watering and wildlife, and navigation. The reservoir receives extensive use for these purposes. The historic water quality and aquatic ecology conditions of Chickamauga Reservoir were described in the final Environmental Statement for Sequoyah Nuclear Plant Units 1 and 2, TVA, February 13, 1974. On July 26, 1974 TVA submitted a Standard Form C Application to the Environmental Protection Agency (EPA) for a National Pollutant Discharge Elimination System permit (NPDES) for the nonradiological discharges from Sequoyah Nuclear Plant. On June 4, 1979, TVA received NPDES permit No. TN0026450 from the EPA for the nonradiological component of the discharges from Sequoyah Nuclear Plant. This permit is updated as required to maintain permits for nonradiological discharges from Sequoyah Nuclear Plant. The permit includes appropriate provisions for the implementation and reporting of instream preoperational and operational monitoring programs in Chickamauga Reservoir with respect to water quality and aquatic ecology. As required by the permit, copies of these reports are also submitted to NRC. The reports of instream monitoring programs submitted under the NPDES permit, both past and future, contain updating information on the water quality and aquatic ecology of Chickamauga Reservoir. A separate updating and reporting of the aquatic conditions of Chickamauga Reservoir outside of the established framework of the NPDES permit requirements is neither planned or warranted in the FSAR. S2-1.doc 2.1-6

SQN TABLE 2.1.1-1 SEQUOYAH NUCLEAR PLANT Coordinates of Unit 1 Reactor Building Centerline Latitude 35° 13' 35.65"N Longitude 85° 05' 28.17"W Universal Transverse Mercator N 3,899,640.62 E 673,718.24 Revised by Amendment 1 T211-1.doc

SQN TABLE 2.1.3-1 1970 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 890 - 15 50 10 5 810 NNE 545 - - 60 85 45 355 NE 390 - - - 45 30 315 ENE 650 - 15 - 100 130 405 E 540 - 25 20 85 70 340 ESE 1,225 10 65 65 135 80 870 SE 965 5 190 25 85 85 575 SSE 1,275 - 35 115 335 105 685 S 2,570 - 80 5 190 265 1,030 SSW 3,425 - 55 55 205 115 2,995 SW 2,535 - - 45 175 45 2,270 WSW 6,475 5 65 335 650 615 4,805 W 3,430 5 35 115 275 200 2,800 WNW 3,030 - 25 145 405 285 2,170 NW 3,965 10 40 185 210 200 3,320 NNW 1,235 10 80 15 40 145 945 Total 32,145 45 725 1,235 3,030 2,420 24,690 T213-1to20.doc

SQN TABLE 2.1.3-2 1980 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 730 - 15 40 10 5 660 NNE 440 - - 50 65 40 285 NE 315 - - - 40 25 250 ENE 555 - 15 - 80 105 355 E 505 - 20 15 70 55 345 ESE 1,195 10 50 50 110 65 910 SE 900 5 155 20 70 70 580 SSE 1,045 - 25 95 270 85 570 S 1,275 - 65 5 155 215 835 SSW 2,785 - 45 45 170 95 2,430 SW 2,860 - - 40 140 35 2,645 WSW 6,785 5 50 270 530 500 5,430 W 3,845 5 30 95 220 180 3,315 WNW 3,385 - 20 120 325 375 2,545 NW 4,930 10 35 150 165 220 4,350 NNW 1,160 10 60 10 35 160 885 Total 32,710 45 585 1,005 2,455 2,230 26,390 T213-1to20.doc

SQN TABLE 2.1.3-3 1985 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 2,045 20 41 175 76 62 1,671 NNE 870 0 30 73 136 62 573 NE 746 0 0 67 67 54 558 ENE 1,114 0 11 24 172 210 697 E 1,186 0 70 11 191 137 777 ESE 2,084 0 118 113 194 137 1,522 SE 1,186 0 129 272 118 152 1,165 SSE 3,171 0 73 320 500 430 1,848 S 3,494 0 67 143 229 547 2,508 SSW 5,878 0 32 81 288 116 5,361 SW 6,575 0 10 236 435 122 5,772 WSW 13,676 20 146 495 866 1,113 11,036 W 4,397 10 20 180 506 530 3,151 WNW 3,462 10 30 281 461 461 2,219 NW 3,142 50 80 225 438 259 2,090 NNW 2,038 10 202 80 71 171 1,504 Total 55,714 120 1,059 2,776 4,744 4,563 42,452 T213-1to20.doc

SQN TABLE 2.1.3-4 1990 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 2,195 28 52 212 85 65 1,753 NNE 1,036 0 36 88 160 75 677 NE 901 0 0 81 82 65 673 ENE 1,419 0 13 29 209 255 913 E 1,485 0 85 13 232 166 989 ESE 2,754 0 143 137 235 166 2,073 SE 2,469 0 157 329 143 187 1,653 SSE 3,719 0 88 388 607 516 2,120 S 3,658 0 82 173 277 663 2,463 SSW 7,471 0 39 98 349 140 6,845 SW 6,517 0 12 323 475 141 5,566 WSW 15,895 24 208 697 1,341 1,435 12,190 W 5,245 8 32 259 739 771 3,436 WNW 4,205 4 35 413 640 539 2,574 NW 3,802 67 118 318 625 312 2,362 NNW 2,460 4 290 114 74 214 1,764 Total 65,231 135 1,390 3,672 6,273 5,710 48,051 T213-1to20.doc

SQN TABLE 2.1.3-5 2000 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 2,289 29 54 221 89 68 1,828 NNE 1,080 0 38 92 167 78 706 NE 940 0 0 84 86 68 702 ENE 1,480 0 14 30 218 266 952 E 1,549 0 89 14 242 173 1,031 ESE 2,872 0 149 143 245 173 2,162 SE 2,575 0 164 343 149 195 1,724 SSE 3,878 0 92 405 633 538 2,211 S 3,814 0 86 180 289 691 2,568 SSW 7,791 0 41 102 364 146 7,138 SW 6,796 0 13 337 495 147 5,804 WSW 16,575 25 217 727 1,398 1,496 12,711 W 5,469 8 33 270 771 804 3,583 WNW 4,385 4 36 431 667 562 2,684 NW 3,965 70 123 332 652 325 2,463 NNW 2,565 4 302 119 77 223 1,839 Total 68,021 141 1,449 3,829 6,541 5,954 50,106 T213-1to20.doc

SQN TABLE 2.1.3-6 2010 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 2,360 30 56 228 91 70 1,885 NNE 1,114 0 39 95 172 81 728 NE 969 0 0 87 88 70 724 ENE 1,526 0 14 31 225 274 982 E 1,597 0 91 14 249 179 1,064 ESE 2,962 0 154 147 253 179 2,229 SE 2,655 0 169 354 154 201 1,778 SSE 3,999 0 95 417 653 555 2,280 S 3,934 0 88 186 298 713 2,649 SSW 8,034 0 42 105 375 151 7,361 SW 7,008 0 13 347 511 152 5,985 WSW 17,093 26 224 750 1,442 1,543 13,109 W 5,640 9 34 279 795 829 3,695 WNW 4,522 4 38 444 688 580 2,768 NW 4,089 72 127 342 672 336 2,540 NNW 2,645 4 312 123 80 230 1,897 Total 70,147 145 1,495 3,949 6,746 6,140 51,672 T213-1to20.doc

SQN TABLE 2.1.3-6a 2010 POPULATION DISTRIBUTION WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 2,418 31 57 234 94 72 1,931 NNE 1,141 0 40 97 176 83 746 NE 993 0 0 89 90 72 741 ENE 1,563 0 14 32 230 281 1,006 E 1,636 0 94 14 256 183 1,090 ESE 3,034 0 158 151 259 183 2,284 SE 2,720 0 173 362 158 206 1,821 SSE 4,097 0 97 427 669 568 2,335 S 4,030 0 90 191 305 730 2,713 SSW 8,230 0 43 108 384 154 7,541 SW 7,179 0 13 356 523 155 6,132 WSW 17,511 26 229 768 1,477 1,581 13,429 W 5,778 9 35 285 814 849 3,785 WNW 4,632 4 39 455 705 594 2,836 NW 4,188 74 130 350 689 344 2,602 NNW 2,710 4 319 126 82 236 1,943 Total 71,861 149 1,531 4,045 6,911 6,290 52,935 T213-1to20.doc

SQN TABLE 2.1.3-7 1970 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 14,550 890 3,425 1,860 2,570 5,805 NNE 19,970 545 6,055 3,915 4,685 4,770 NE 22,025 390 1,210 2,830 7,600 9,995 ENE 41,510 650 3,770 5,425 21,405 10,260 E 19,690 540 9,995 3,285 1,835 4,035 ESE 43,600 1,225 26,685 3,250 1,055 11,385 SE 13,265 965 4,960 3,135 1,845 2,360 SSE 48,495 1,275 6,075 8,590 29,210 3,345 S 47,810 1,570 9,840 9,785 19,000 7,615 SSW 137,590 3,425 79,150 34,630 13,825 6,560 SW 146,185 2,535 104,960 25,950 7,495 5,245 WSW 48,275 6,475 19,655 4,455 9,345 8,345 W 17,075 3,430 1,490 4,660 3,785 3,710 WNW 14,545 3,030 2,390 3,135 4,080 1,910 NW 14,320 3,965 980 1,365 725 7,285 NNW 10,110 1,235 540 2,780 1,545 4,010 Total 659,015 32,145 281,180 119,050 130,005 96,635 T213-1to20.doc

SQN TABLE 2.1.3-8 1980 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 15,605 730 3,560 2,030 2,535 6,750 NNE 20,805 440 6,485 4,120 4,705 5,055 NE 23,270 315 1,230 2,860 7,615 11,250 ENE 46,035 555 3,900 6,200 24,740 10,640 E 21,920 505 11,930 3,380 2,005 4,100 ESE 51,760 1,195 34,815 3,350 1,075 11,325 SE 15,040 900 6,835 3,140 1,795 2,370 SSE 56,420 1,045 6,840 9,005 36,080 3,450 S 51,060 1,275 9,565 9,895 22,290 8,035 SSW 156,825 2,785 90,575 42,330 14,695 6,440 SW 162,260 2,860 115,955 29,725 8,655 5,065 WSW 54,975 6,785 23,310 4,595 11,440 8,845 W 17,480 3,845 1,470 4,820 3,705 3,640 WNW 14,875 3,385 2,645 3,160 3,835 1,850 NW 17,880 4,930 1,050 1,460 765 9,675 NNW 10,060 1,160 510 2,725 1,555 4,110 Total 736,270 32,710 320,675 132,795 147,490 102,600 T213-1to20.doc

SQN TABLE 2.1.3-9 1985 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 21,308 2,045 4,922 3,190 2,310 8,841 NNE 31,222 870 9,507 4,365 7,350 9,130 NE 29,466 746 2,175 5,524 5,573 15,448 ENE 52,493 1,114 3,942 4,881 26,393 16,163 E 29,712 1,186 14,581 5,761 4,534 3,650 ESE 60,518 2,084 39,948 4,272 1,745 12,469 SE 27,161 1,836 4,977 4,548 12,881 2,919 SSE 63,290 3,171 10,711 7,829 31,660 9,920 S 70,268 3,494 20,067 18,800 17,723 10,184 SSW 159,215 5,878 84,597 42,513 16,248 9,979 SW 143,916 6,575 98,057 20,998 8,179 10,108 WSW 63,676 13,676 24,026 3,551 13,269 9,155 W 23,283 4,397 1,355 5,560 4,963 7,008 WNW 20,291 3,462 4,915 4,070 5,688 2,156 NW 21,140 3,142 1,230 1,490 1,096 14,182 NNW 12,847 2,038 445 2,910 2,515 4,939 Total 829,804 55,714 325,453 140,260 162,127 146,250 T213-1to20.doc

SQN TABLE 2.1.3-10 1990 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 21,471 2,195 4,390 2,665 2,641 9,580 NNE 31,190 1,036 9,280 4,399 7,206 9,269 NE 29,749 901 2,390 5,916 5,308 15,234 ENE 55,722 1,419 7,461 4,897 25,698 16,247 E 33,376 1,485 18,584 5,296 4,526 3,485 ESE 53,443 2,754 32,802 4,305 1,734 11,848 SE 23,655 2,469 5,659 6,099 3,970 5,458 SSE 76,949 3,719 10,496 10,471 41,756 10,507 S 93,648 3,658 38,376 21,859 20,136 9,619 SSW 163,242 7,472 87,613 40,958 16,818 10,381 SW 98,030 6,515 55,198 17,609 8,997 9,711 WSW 85,592 15,889 44,979 3,524 13,109 8,092 W 25,078 5,247 2,616 5,546 5,059 6,611 WNW 19,124 4,204 3,611 3,445 5,677 2,188 NW 22,599 3,802 1,801 2,015 1,164 13,817 NNW 14,273 2,460 839 3,055 2,646 5,274 Total 847,142 65,225 326,093 142,060 166,445 147,318 T213-1to20.doc

SQN TABLE 2.1.3-11 2000 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 23,320 2,201 4,954 2,856 2,860 10,450 NNE 34,058 1,036 10,595 4,679 7,667 10,081 NE 31,899 902 2,668 6,265 5,634 16,430 ENE 60,379 1,421 8,578 5,245 27,527 17,607 E 36,433 1,485 20,674 5,688 4,846 3,740 ESE 58,292 2,754 36,514 4,626 1,842 12,556 SE 26,081 2,469 6,314 6,775 4,414 6,108 SSE 85,780 3,719 11,818 11,774 46,792 11,678 S 103,675 3,658 42,248 24,566 22,584 10,618 SSW 178,503 7,472 96,253 45,246 18,356 11,176 SW 106,520 6,839 60,896 19,168 9,589 10,028 WSW 92,896 17,190 49,314 3,870 14,280 8,242 W 27,248 5,715 2,885 6,088 5,426 7,134 WNW 20,522 4,500 3,917 3,699 6,034 2,372 NW 24,507 4,144 1,960 2,176 1,222 15,004 NNW 15,114 2,515 966 3,286 2,802 5,546 Total 925,225 68,021 360,554 156,007 181,874 158,769 T213-1to20.doc

SQN TABLE 2.1.3-12 2010 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 24,711 2,206 5,385 3,009 3,028 11,082 NNE 36,232 1,036 11,600 4,893 8,022 10,681 NE 33,460 903 2,859 6,495 5,855 17,349 ENE 63,886 1,422 9,431 5,499 28,862 18,672 E 38,743 1,485 22,276 5,972 5,080 3,930 ESE 61,927 2,754 39,360 4,859 1,918 13,036 SE 27,870 2,469 6,817 7,270 4,729 6,585 SSE 92,224 3,719 12,806 12,726 50,436 12,537 S 111,202 3,658 45,208 26,632 24,354 11,350 SSW 189,612 7,472 102,822 48,274 19,331 11,713 SW 112,822 7,086 65,232 20,223 9,973 10,308 WSW 98,545 18,178 52,615 4,139 15,197 8,415 W 28,884 6,071 3,089 6,509 5,698 7,517 WNW 21,522 4,726 4,126 3,875 6,288 2,508 NW 25,933 4,405 2,074 2,295 1,261 15,899 NNW 15,780 2,557 1,064 3,475 2,925 5,759 Total 983,353 70,147 386,764 166,147 192,954 167,341 T213-1to20.doc

SQN TABLE 2.1.3-12a 2020 POPULATION DISTRIBUTION WITHIN FIFTY MILES OF SITE Miles from Site 0-10 10-20 20-30 30-40 40-50 Direction Total N 25,824 2,210 5,737 3,119 3,154 11,605 NNE 38,021 1,036 12,425 5,073 8,318 11,170 NE 34,872 904 3,050 6,738 6,077 18,103 ENE 66,776 1,424 10,096 5,719 30,013 19,524 E 40,611 1,485 23,516 6,229 5,286 4,094 ESE 64,776 2,754 41,562 5,071 1,991 13,398 SE 29,079 2,469 7,206 7,596 4,910 6,898 SSE 96,099 3,719 13,494 13,290 52,566 13,030 S 116,275 3,658 47,531 27,909 25,402 11,775 SSW 197,551 7,472 107,951 50,169 19,934 12,025 SW 117,867 7,284 68,724 20,954 10,250 10,654 WSW 103,157 18,975 55,273 4,337 15,894 8,678 W 30,194 6,358 3,249 6,820 5,914 7,852 WNW 22,333 4,908 4,292 4,020 6,499 2,614 NW 27,075 4,615 2,162 2,383 1,311 16,605 NNW 16,353 2,591 1,140 3,602 3,034 5,987 Total 1,026,862 71,861 407,408 173,028 200,554 174,010 T213-1to20.doc

SQN TABLE 2.1.3-13 1970 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 465 0 0 35 30 20 380 NNE 270 0 0 110 10 20 130 NE 20 0 20 0 0 0 0 ENE 130 0 130 0 0 0 0 E 30 0 30 0 0 0 0 ESE 10 5 10 0 0 0 0 SE 15 0 15 0 0 0 0 SSE 475 0 35 0 0 210 230 S 755 10 105 0 0 10 630 SSW 1,210 0 10 160 210 280 550 SW 1,655 0 50 155 305 870 275 WSW 10 0 0 0 10 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 0 0 0 0 0 0 0 NNW 195 0 0 0 40 155 0 Total 5,240 10 405 460 605 1,565 2,195 T213-1to20.doc

SQN TABLE 2.1.3-14 1980 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 593 0 0 43 40 25 485 NNE 346 0 0 140 13 25 168 NE 25 0 25 0 0 0 0 ENE 165 0 165 0 0 0 0 E 40 0 40 0 0 0 0 ESE 15 0 15 0 0 0 0 SE 20 0 20 0 0 0 0 SSE 608 0 45 0 0 270 293 S 964 13 135 0 0 13 803 SSW 1,541 0 13 205 270 358 695 SW 2,124 0 65 201 390 1,118 350 WSW 13 0 0 0 13 0 0 W 330 330 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 0 0 0 0 0 0 0 NNW 249 0 0 0 51 198 0 Total 7033 343 523 589 777 2,007 2,794 T213-1to20.doc

SQN TABLE 2.1.3-15 1985 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 453 0 0 0 0 35 418 NNE 217 0 0 3 0 3 211 NE 87 0 87 0 0 0 0 ENE 5 0 5 0 0 0 0 E 45 0 45 0 0 0 0 ESE 0 0 0 0 0 0 0 SE 124 0 124 0 0 0 0 SSE 8 0 0 0 0 0 8 S 731 0 73 0 0 328 330 SSW 2,502 0 147 206 276 213 1,660 SW 1,918 0 38 5 237 935 703 WSW 265 0 0 265 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 4 0 0 0 0 4 0 NNW 269 0 0 45 98 126 0 Total 6,628 0 519 524 611 1,644 3,330 T213-1to20.doc

SQN TABLE 2.1.3-16 1990 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 1,439 0 0 0 0 80 1,359 NNE 150 0 0 75 0 75 0 NE 412 0 412 0 0 0 0 ENE 87 0 87 0 0 0 0 E 46 0 46 0 0 0 0 ESE 0 0 0 0 0 0 0 SE 128 0 128 0 0 0 0 SSE 87 0 0 0 0 0 87 S 749 0 75 0 0 336 338 SSW 4,066 0 151 212 1,375 219 2,109 SW 3,637 0 468 512 243 1,140 1,274 WSW 272 0 0 272 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 87 0 0 0 0 87 0 NNW 277 0 0 46 101 130 0 Total 11,437 0 1,367 1,117 1,719 2,067 5,167 T213-1to20.doc

SQN TABLE 2.1.3-17 2000 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 1,571 0 0 0 0 87 1,484 NNE 401 0 0 82 0 82 237 NE 450 0 450 0 0 0 0 ENE 95 0 95 0 0 0 0 E 50 0 50 0 0 0 0 ESE 0 0 0 0 0 0 0 SE 140 0 140 0 0 0 0 SSE 95 0 0 0 0 0 95 S 818 0 82 0 0 367 369 SSW 4,441 0 165 232 1,502 239 2,303 SW 3,971 0 511 559 265 1,245 1,391 WSW 297 0 0 297 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 95 0 0 0 0 95 0 NNW 302 0 0 50 110 142 0 Total 12,726 0 1,493 1,220 1,877 2,257 5,879 T213-1to20.doc

SQN TABLE 2.1.3-18 2010 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITE Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 1,672 0 0 0 0 93 1,579 NNE 426 0 0 87 0 87 252 NE 479 0 479 0 0 0 0 ENE 101 0 101 0 0 0 0 E 53 0 53 0 0 0 0 ESE 0 0 0 0 0 0 0 SE 149 0 149 0 0 0 0 SSE 101 0 0 0 0 0 101 S 870 0 87 0 0 390 393 SSW 4,725 0 176 247 1,598 254 2,450 SW 4,226 0 544 595 282 1,325 1,480 WSW 316 0 0 316 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 101 0 0 0 0 101 0 NNW 321 0 0 53 117 151 0 Total 13,540 0 1,589 1,298 1,997 2,401 6,255 T213-1to20.doc

SQN TABLE 2.1.3-19 2020 ESTIMATED PEAK HOUR RECREATION VISITS WITHIN TEN MILES OF SITES Miles from Site 0-1 1-2 2-3 3-4 4-5 5-10 Direction Total N 1,752 0 0 0 0 97 1,655 NNE 446 0 0 91 0 91 264 NE 502 0 502 0 0 0 0 ENE 106 0 106 0 0 0 0 E 56 0 56 0 0 0 0 ESE 0 0 0 0 0 0 0 SE 156 0 156 0 0 0 0 SSE 106 0 0 0 0 0 106 S 912 0 91 0 0 409 412 SSW 4,954 0 184 259 1,675 267 2,569 SW 4,431 0 570 624 296 1,389 1,552 WSW 331 0 0 331 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 0 0 0 0 0 5 0 NNW 179 0 0 56 123 152 0 Total 13,931 0 1,665 1,361 2,094 2,253 6,558 T213-1to20.doc

SQN-21 TABLE 2.1.3-20 EDUCATIONAL INSTITUTIONS IN VICINITY OF SEQUOYAH NUCLEAR PLANT 1990-2020 School Location 1990 1993 1997 2000 2010 2020 Harrison Bay Vocational School 3-4 SE 473 400 401 434 462 485 McConnel Elementary School 3-4 WSW 836 895 751 855 909 954 Loftis Middle School 3-4 WSW 839 910 1000 1100 John Allen Elementary School 3-4 W 227 309 368 390 400 420 Snowhill Elementary School 4-5 SE 831 655 651 650 650 650 Big Ridge Elementary School 4-5 SW 851 720 569 600 700 800 Soddy-Daisy Elementary School 4-5 W 756 640 400 413 439 461 Soddy-Daisy High School 4-5 W 1580 1510 1607 1687 1800 2000 Daisy Elementary 4-5 W ---- 176 509 560 610 700 Sequoyah Vocational Center 4-5 W 600 600 635 650 700 770 McDonald Elementary School (Bradley County) 5-10 SE 175 161 Closed ---- ---- ---- Ooltewah High School 5-10 SSE 1561 1450 1569 1710 1880 2000 Wallace A. Smith Elementary School 5-10 S 496 614 670 695 770 847 Brown Junior High School 5-10 SSW 755 814 433 486 550 605 Central High School 5-10 SSW 1218 1046 1077 1176 1252 1313 Harrison Elementary School 5-10 SSW 809 563 583 866 922 967 Hixson High School 5-10 SW 1323 895 1130 1384 1473 1544 Falling Water Elementary School 5-10 WSW 259 220 326 330 340 357 Ganns-Middle Valley School 5-10 WSW 780 622 449 500 600 720 Mowbray Elementary School 5-10 WNW 98 74 Closed ---- ---- ---- Soddy-Daisy Middle School* 5-10 WNW 808 825 1607 1700 1870 2000 Soddy Elementary School 5-10 W 573 535 400 440 484 540 Total: 15,009 13,724 14,974 16,416 17,811 19,233

Name change--formerly Soddy-Daisy Junior High School T213-1to20.doc

SQN TABLES 2.1.4-1 FARM ORIENTED LAND USE LAND AND LAND IN FARMS Approximate Land Proportion County Land in Area in Farms in Farms

                     ---------------Ac----------------                                   -----pct----

Bradley 210,000 94,364 45.0 Hamilton 345,000 64,995 18.8 NUMBER AND AVERAGE SIZE OF FARM Average Size County All Farms of Farm

                                  --no.--                           ----Ac----

Bradley 754 125 Hamilton 613 106 LAND IN FARMS ACCORDING TO USE Woodland Including All Irrigated County Cropland Woodland Pasture Other Land Land

                                -----------------Ac----------------------------

Bradley 53,488 28,497 12,379 633 Hamilton 33,977 23,364 7,654 1,021 CROPLAND Harvested Cropland Used All Other County Cropland for Pasture Cropland

                       -------------------------Ac----------------------

Bradley 20,477 31,382 1,629 Hamilton 13,159 18,919 1,919 Source: 1982 Census of Agriculture T214-01to02.doc

SQN TABLES 2.1.4-2 CROPS HARVESTED Bradley County Hamilton County Yield Acres Yield Acres Field corn bu/Ac 77 1,482 71 1,057 Sorghum bu/Ac - - 63 45 Wheat bu/Ac 37 896 26 1,414 All other small grain N/A 291 N/A - Soybeans bu/Ac 34 1,005 22 2,026 Hay tons/Ac 1.8 15,661 1.6 8,596 Cotton bales/Ac - - - - Peanuts lbs/Ac - - - - Tobacco lbs/Ac 1,826 81 1,885 7 Vegetable, sweet corn, or melon N/A 50 N/A 87 Irish and sweet potatoes N/A 5 N/A 5 Berries N/A 10 N/A - Land in orchards N/A 311 N/A 147 Other crops N/A 685 N/A - Source: 1982 Census of Agriculture T214-01to02.doc

35" Revised by Amendment 13 BASE MAP MATERIALS PREPARED BY AND AVAILABLE FROM MAPPING SERVICES BRANCH.

                                                                          -----------------------------...1 83"                  82"

I MORGA

                                            -'1'*'"j, -----

OCOEE F A'N-N1 :>-, N

                                                              , I      -- ,.
                                                     --/ - ---

RRAY j I -* SEQUOYAH NUCLEAR PLANT

                                                                )
                                                                                     --                 \     FINAL SAFETY ANALYSIS REPORT IY FEATURES WITHIN 50 MILES n       .. 1                ...

FIGURE 2.1. 1*2

                                                             - -    .. _ : i , _ _ J " _ -
                                               !                                           ..  -L           '
                                    " V I t.._            PICKENS,*                                 -*     ,DAWS     I s                                                         L---,

I I _, ", 10 0 10 20 30 r-: SCALE OF MILES

SEQUOYAH NUCLEAR PLANT FINAL SAFE"TY ANALYSIS REPORT FEATURES WITHIN 10 MILES FIGURE 2.'\.1-3 0 2 3 4 BBB SCALE OF MILES

    -         SQN ProtectedArea Boundary (D Unit 1                (jJ)OeHerator Bldg. GJcoollng Tow&ra

@ Unit 2 @ on. & Pwr. SI.Fe, @Prolecled Area VehlcleEnlry @3 Turbine Bldg. 6) Deslgn,Serv.Bldgs. "SallyPort" @ AuxlllaryBldg. 63) Toacup (V\tntakeChnnnel G) Service Bldg. rj])Acceu Control Porlal@ProtecllveDlkD @controlBldg. Q§ERCWPumpSt. @ F I R E PUMP HOUSE C'\\J G) OfflceJCafeterlBldg.QJ)tnt*kePumpSI. @cnnd. Demlnll, Bldg. q]i Solnr Bldg.

                                               @J)FIRE OPERATIONS :::--:::::;,,r hi?, CENTER 3
                                                                                             \j
                                               'el'M U L TI-PURPOSE BUI L D ING

@Make-UpWtr.

                                               @INDEPENDENT SPENT FUEL STORAGE INSTALLATION

(@) Add. Dlaaelaen. Bldg (ISF SI )

                                               @ S K I M M ER WALL
                                               @ U N I T 1 OLD STEAM GENERATOR STORAGE FACILITY (OSGSF)                             SEQUOYAH NUCLEAR PLANT FINAL SAFETY
                                               @)LOW LEVEL RAD WASTE FACILITY                                                            ANALYSIS REPORT
                                               @ U N I T 2 OLD STEAM GENERATOR STORAGE FACILITY (OSGSF)                            FIGURE 2.1 .2-1
                                               @ F L E X EQUIPMENT STORAGE BUILDING (FESB)                                         GENERAL SITE PLAN (REVISED BY AMENDMENT 26)

I CAO MAINTAINED DRAWING I

EXCLUSION AREA -** - ROUTINE LIQUID RELEASE POINT DISCHARGE CANAL ROUTINE GASEOUS RELEASE POINT (elevation above plant grade) RELEASE POINT 1 Auiliary Building vent exhaust el. 107 n Shreld Bu i ding vent exhaust eL 121 _5 ft ./ RELEASE POINT 2 Service Buildtng vent exhaust el. 26 ft RELEASE POINT 3 OBA DOSE CALCULATION Condenser air ejector exhaust vent eL 93. 7 ft

                                                                                       *oao = m = :::::i*o o =:380 0 e :

SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT F I GU RE 2 . 1 . 2 - 2 SITE BOUNDARY (REVISED BY AMENDMENT 20) I CAD MAINTAINED DRAWING I

Best Available Historical Image r

LOW POPULATION ZONE FIGURE 2.1.~*

                               ** ::-:=b. !':=:
                   *-****~*

r r LAND USE IN PLANT VICINITY SCALE l:250,000 F I G U R E 2.1.4-l

KY TENN- \ k: 0 - ,

                             - - . , . a        ,

LEGEND *----- Pine _ _ _ _ ...

                                                        ,_       *" ' - .      S C Hardwood _ _                                          GA Mixed _ _    _

LOCATION MAP SEQUOYAH NUCLEAR PLANT FlNAL SAFETY ANALYSIS REPOIT FOREST TYPES AND COVER BRADLEY AND HAMILTON COUNTIES FIGURE 2.1.4-2 ICALEOf I U I 1 I 1 2 3 4 I I 7 I I RIPIM I J E-3 IIAYltlt

SQN 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES There are no industrial or military facilities within five miles of the Sequoyah Nuclear Plant site which would potentially pose a hazard to the safe operation of the plant. A discussion of the highway network in the vicinity of the plant site is contained in Section 2.1. Facilities of interest beyond five miles include the Volunteer Army Ammunition (VAA) Plant and the Dallas Bay Sky Park. Also, Federal Airway V333 passes directly over the site, and Chickamauga Lake is a commercially navigable waterway. The Chattanooga Airport is located approximately 14.5 miles from the plant site. These are the only facilities of potential significance to the safe operation of the plant, and based on the evaluations set forth below, these activities will pose no hazard. 2.2.1 Location and Routes Chickamauga Lake is a navigable waterway used by both commercial and recreational traffic. Through a series of locks and dams, commercial traffic can travel from Knoxville, upstream of the site to the mouth of the Tennessee River at the Ohio River. The Dallas Bay Sky Park is a general aviation airport located about 5.5 miles WSW of the plant. The Chattanooga Airport is a full-service commercial airport located about 14.5 miles SSW of the plant. The nearest boundary of the VAA Plant is about eight miles from the plant site. Figure 2.1.1-3 shows this relationship. The plant is in a stand-by mode and has not produced explosives since 1977. It is not expected to resume production unless there would be a national emergency. However, a small amount of munitions is stored on the site and shipped to and from the site by truck. There are no specific restrictions on the routes to be taken by trucks that would keep them away from the nuclear plant. Barges have never been used for shipping and they are not expected to be used in the future. Rail cars have been used in the past for explosives when the plant was in production but are not expected to be used in the future unless production resumes. (The nearest mainline railroad is about five and one-half miles west of the nuclear plant.) Also, the VAA plant currently contracts its facility to Raytheon Company, which utilizes the plant for final assembly of two air-to-ground missiles: The IR Maverick Missile and the SM-2 Standard Missile. The missiles are shipped to and from the site by truck. Trucks leaving the VAA follow Bonnie Oaks Drive to I-75 and proceed either North or South. West bound shipments exit onto I-24 West. 2.2.2 Description of Products Up to 44 training operations per day take place at the Dallas Bay Sky Park with an average of about

25. Many of them involve low-altitude maneuvers in the general vicinity of the plant.

Air traffic on or near Federal Airway V333 on the most recent peak traffic day at the Chattanooga Airport was 42. This includes both IFR (Instrument Flight Rules) and VFR (Visual Flight Rules) flights. They ranged in altitude from 2,000 to 15,000 feet. The type of aircraft which utilize Federal Airway V333 include: Cessna 152; Cessna 425; BA-31; DC-9; MD-80; Boeing 727; K-10; F-28; C-130; SW-3; BE-100; BE-200; and BE-90. S2-2.doc 2.2-1

SQN The data were for an 18 hour period on July 21, 1992, and reflect the peak traffic for the area of responsibility of the airport, not necessarily V333. Traffic during the six undocumented hours is likely to be very small. Air traffic at the Chattanooga Airport averages about 140 incoming flights per day. Under certain wind conditions, an estimated 35 - 40 percent will make an approach that takes them over or near the plant at an elevation of about 2500 feet above the ground. The SM-2 Standard Missile contains 285 pounds net explosive weight and is transported 18 to a truck. The IR Maverick Missile contains 362 pounds net explosive weight and is shipped 36 to a truck. The small munitions are stored in two magazines each designed to store 500,000 pounds of TNT. Table 2.2.2-1b shows the type of munitions stored on site; the typical amount stored on site; and the typical amount transported by truck. There is no set schedule for the shipment of the munitions. Table 2.2.2-1 shows the total amount of certain hazardous materials shipped past the Sequoyah Nuclear Plant from 1982 to 1992 on a yearly basis based on Corps of Engineers lock data. The product listed as gasoline on the table is actually RU250. In addition, data on chlorine shipments became available starting in 1990. Table 2.2.2-1a contains 1990 shipping data from a TVA survey of dock operators. Based on 1992 shipping data, chlorine is shipped at a rate of about one 1,100 ton barge every ten days; RU250 (gasoline) is no longer shipped; residual fuel oil is shipped at a rate of one three-barge tow every three months with about 1,500 tons per barge; and asphalt is shipped at a rate of about three barges per month with two 1,500-ton barges and one 3,000-ton barge. Variations in total yearly shipments occur by adjusting any or all of the three variables--shipping frequency, number of barges per tow, and barge size. 2.2.3 Evaluations 2.2.3.1 Evaluation of Explosion Hazards from Nearby Transportation Routes As indicated in Tables 2.2.3-1 and 2.2.3-2, certain hazardous materials are transported by river barge past the Sequoyah Nuclear Plant site. In addition, explosive materials are also transported over nearby railroad lines. Therefore, these materials were evaluated for their potential to damage the safety related structures of the plant. The materials include TNT, gasoline, liquid natural gas (LNG) and unspecified fertilizers. Table 1736 of AMCH-385-224 requires that 500,000 lb of TNT (maximum transported by rail) be stored at least 5,400 feet from any unbarricaded, inhabited building and that 400,000 lb of TNT be stored at least 2,550 feet from such building. These distances are much less than the nearest railroad (29,000 feet) or highway (39,000 feet) to Sequoyah over which large amounts of explosives can be transported. Thus, there is no potential for damage to the Sequoyah plant due to the transport of TNT from or storage of TNT at the VAA Plant. Table 2.2.3-3 indicates the amount of gasoline shipped past the Sequoyah site over the past 15 years. The gasoline supply for Knoxville is provided by pipeline. As of 1974 with the pipeline in S2-2.doc 2.2-2

SQN full operation no future gasoline barge shipments past the Sequoyah site are expected except in case of an emergency. The potential for damage to the Sequoyah plant from a gasoline barge explosion is considered to be negligible. In response to concerns raised by the ACRS, the possibility of a barge explosion in the vicinity of the new ERCW pumping station has been reviewed. Our response is as follows: (1) The ACRS identified liquid natural gas (LNG) as a substance to be considered in an exploding barge scenario. From our review of the barge shipments past Sequoyah for calendar year 1978, there were no shipments of LNG on the Tennessee River. It should be noted that barge shipments of LNG past Sequoyah are not likely since natural gas transportation is handled almost entirely by pipeline in this region. Therefore, we do not consider the potential for an exploding LNG barge near the new ERCW pumping station to be a credible event. (2) As indicated in Table 2.2.3-2, there were, in calendar year 1978, shipments of unspecified fertilizers past the Sequoyah Nuclear Plant. Hence, the possibility of an accidental explosion must be considered. In 1966, the U.S. Bureau of Mines issued a study entitled "Explosion Hazards of Ammonium Nitrate Under Fire Exposure," which examined the deflagration and detonation hazards associated with Ammonium Nitrate (AN). The study indicates: (a) Ordinary fertilizer-grade AN requires strong overpressures to initiate detonation within the mixture. (b) AN and AN-fuel mixtures were exposed to fire with no transition from deflagration to detonation being observed. (c) A combination of fire and overpressure results in transition to detonation. However, in 2 free-flowing beds of AN and AN-fuel mixtures, pressures as high as 8000 lb/in did not generate detonation. Only in experiments where the AN was not allowed to flow freely was transition to detonation observed in the AN-fuel mixture at pressures above 1000 lb/in2, but not with pure AN. (d) It was found that hot AN (under fire exposure) readily detonated when impacted with a high velocity projectile or shock wave. Explosions in storage and shipments of AN have apparently resulted only when nearby explosions or structure collapse have occurred concurrent with fire in the AN. (e) Gas detonations have been shown incapable of initiating detonation in AN mixtures. In general, fertilizers shipped on the Tennessee River employ diatomaceous earth and kaolin clay for anticaking dusts rather than using oil sealant, thus detonations are possible only in cargoes where fire and missiles or external detonation are present. Most bulk fertilizers with earth or clay mixtures will not burn without mixing a considerable amount of paper or flammable material into the fertilizer. S2-2.doc 2.2-3

SQN Based on the insensitivity to detonation exhibited by most common fertilizers, the unlikely sequence of events required for detonation must include: Barge collision, fire in the fertilizer cargo, and concurrent detonation or missile-inducing event. Therefore, given the low probability of a barge collision and the low percentage of fertilizer shipments on the Tennessee River, it is concluded that, because of the very low probabilities associated with the event, no hazard exists to the intake pumping station from the transportation of fertilizers by barge on the Tennessee River system. 2.2.3.2 Evaluation of Barge Impact with the ERCW Intake Structure The collision of a tow with the ERCW intake pumping station is considered to be an unlikely event. The intake structure is protected by location from collision with river traffic heading downstream for water surfaces up to elevation 705, which is 22 feet above maximum normal pool level and 15 feet above a flood condition equivalent to one-half the probable maximum flood. The probability per year of a collision with a drifting barge heading downstream is conservatively estimated to be 4.4 x 10-8. The probability of a collision involving a tow heading upstream has been determined to be 1.6 x 10-5/year. These probabilities were calculated using the event tree techniques (Reference 1) as described below and are believed to be conservative. Collision With River Traffic Heading Downstream

1. Probability of reaching or exceeding flood level 705. Because of the existence of an upstream protective dike with a top elevation of 700.0 as shown in Figure 2.1.2-1 the flood level has to be 705.0 or higher in order for a river vessel to go over the top of the dike and subsequently collide with the intake structure. The probability of a water surface reaching or exceeding flood level 705 is 4 x 10-6 in any given year.

2. Probability of random hit. The probability that a barge drifts, on a collision course, toward the intake structure depends on the relative sizes of river width and intake structure. Probability of random hit equals structure size divided by river width: P=67/6000 = 1.1 x 10-2. The width of the river at the plant site, based on a flood level of 705, was estimated conservatively from Figure 2.4.1-1. The length of the upstream exterior wall of the intake structure was used as the structure size in the computation.

3. Other considerations.

a. Mechanics of river flow. The Sequoyah Nuclear Plant is located on the convex bank of the river. According to flow theory and actual observations made on various rivers (Reference 2), surface-drifting subjects will never be able to reach the vicinity of the intake structure.

Water particles in a bend have a "transverse circulation"; particles near the surface move toward the concave bank and those at the bottom move toward the convex bank. Since the transverse circulation of water particles and the direction of the bend are related by the laws of fluid dynamics, the reversal of the direction of the transverse circulation is a condition almost impossible to exist.

b. Correlation between flood occurrence and river vessel release. Occurrence of a flood does not necessarily result in the release of a river vessel, and for any given level the probability of release is always less than one.

S2-2.doc 2.2-4

SQN

c. Probability of river vessel arrival. Even if a certain flood level were reached and a river vessel were released, the river vessel might not be able to arrive at the immediate upstream station of the intake structure due to the fluctuation of the flood level and the irregularity of the bank formation.

If only the probability of reaching flood level 705 and the probability of random hit are accounted for, the collision probability is then the product of the probabilities of the two individual events, yielding a probability of 4.4 x 10-8 collisions/year. This procedure is conservative because the consideration of river flow mechanics and chance of release and arrival of river vessel are not included in the computation. Therefore, river traffic-intake structure collision at the Sequoyah Nuclear Plant site is considered to be incredible. Collision With River Traffic Heading Upstream Tow operators on the Tennessee River have been required to be licensed by the U.S. Coast Guard since 1972. A requirement for this license is that they must abide by the Western Rivers Rules of the Road. These rules provide that only tows having radar may proceed during inclement weather while those not having radar must tie up. The U.S. Coast Guard has stated that the type of shoreline and mooring cells in the vicinity of Sequoyah Nuclear Plant afford excellent weather protection. The plant is located between Tennessee River Mile (TRM) 484 and 485; first class safety harbors are located near TRM 483 and 489. The Coast Guard has further stated that the present channel markings are more than sufficient for a prudent navigator. The pumping station is well outside the navigation channel (approximately 300 feet from the boundary) and a daymarker and light is located on the far side of the channel directly opposite the plant to guide upstream traffic away from the plant. Sequoyah Nuclear Plant is located on the convex bank of a bend in the Tennessee River Channel. Upstream tows attempting to cut short the navigation of the bend would have a difficult angle of approach to the pumping station. As addressed in the discussion for traffic heading downstream, tows losing power in the bend and drifting will drift toward the shoreline opposite the intake structure.

                             -5 The probability of 1.6 x 10 collisions/year was obtained using the following information. The calculation is believed to be conservative.

1. Data available for the years 1945-1979 was searched for barge groundings on the Chickamauga Reservoir. Of the 10 groundings found, 7 were not applicable because of grounding during inclement weather before 1972 or because of intentional grounding caused by loss of power. A range of 40.35 miles (40.35 x 5280 x 2 feet) of shoreline and a total of 19,674 tows during these years were involved. This yields a probability of grounding per tow per foot of shoreline on the reservoir of 3.6 x 10-10.

2. The target length of the intake structure susceptibility was conservatively taken as 200 feet. (The intake structure is 118 feet by 67 feet.) The average number of tows heading upstream past the intake structure during 1974 to 1979 was approximately 225 per year. The number of tows on the Chickamauga Reservoir reached a peak in 1970, but has been S2-2.doc 2.2-5

SQN roughly uniform during 1974 to 1979 and is believed to be a good indication of the expected number of tows for the next several years. The probability is therefore calculated as 3.6 x 10-10 groundings per tow per foot of shoreline x 200 feet of shoreline x 225 tows per year = 1.6 x 10-5 collisions/year. An evaluation of the navigation capabilities and requirements for navigation through this section of the river, mile 484 to 485, was conducted. This evaluation provides a strong qualitative rationale that the expected rate of occurrence of an upstream barge impact on the ERCW pumping station is very unlikely compared to the random probability of a tow grounding. TVA is confident that the real expected rate of occurrence of barge impact on the ERCW is far less than the calculated value of 1.6 x 10-5 events per year. TVA's understanding of the inadequately documented events has led to the belief that the calculated random probability of hitting a portside bank (tow traveling upstream) at the Sequoyah river location is conservative. The rationale for this belief is discussed below. Discussions with the U.S. Coast Guard revealed the following information about the potential for a barge tow to accidentally collide (direct impact or otherwise) with the ERCW pumping station. The certified barge tug pilot primarily navigates in the traditional "river-pilot" manner, which is by (1) experience, (2) line of sight to landmarks, (3) U.S. Corps of Engineers chart (updated annually), and (4) the Coast Guard Western Rules of the Road. However, the modern (1981) river tug pilot is generally equipped with depth finders (sonar fathomometers), range finding radars, electronics to define water and wind vectors, 2-way radio, and electronic status indication of operational systems. The development and upgrade of modern navigational aids, as well as a more reliable propulsion system, ensures an increasingly accurate, effective navigation of the river by barge pilots. In all weather, the position, without electronic aids, is known to less than 200 feet, and with navigational electronics, to less than 50 feet. On Chickamauga Reservoir, in the traverse by the Sequoyah Nuclear Plant, the position is very well defined because there are buoys every 0.2 mile on the port and starboard sides (a total of 14); there are five navigation lights; the river and riverbank topography is unusually distinctive; and there are distinctive landmarks (the Sequoyah cooling towers and power transmission lines). The radar equipped boat uses the transmission lines as the primary position locator. A river pilot going upstream by Sequoyah will choose to go on the starboard side because of courtesy (Western Rules of the Road) and because of the need to efficiently and safely navigate an "s" curve through this traverse. The upstream barge is surprisingly maneuverable. A barge can make a 180° change in course without emergency measures in about twice the length of tow (i.e., within 400 to 800 feet). An upstream barge can make a 90° controlled turn in less than 0.2 mile under typical conditions, i.e., current (2-1/2 knots), wind (10 knots), and power (single screw). If a tug loses propulsion in upstream traverse, he still has effective steerage for 1/4-1/2 mile (approximately 3-6 minutes, worst case). The pilot can make emergency stops by slipping an anchor or a spud. An upstream barge can easily be piloted to hit a target area 90° to port or starboard within 25 feet under bad conditions and within 5 feet under good conditions. Therefore, a certified river pilot, even in extremis (defined as 'must take emergency measures to avoid trouble or to ground his S2-2.doc 2.2-6

SQN tow'), can and would avoid the ERCW. The ERCW is a significant structure, which is well marked and lighted as a navigation hazard. In extremis, a pilot will select the best course of action from an economic and safety standpoint. And, in a traverse by the Sequoyah ERCW, he will most likely attempt a grounding on an underwater shoal to his starboard side (the Denny Bluff Shoal). The river barge pilot is a U.S. Coast Guard certified pilot, whose license is renewed annually and who has periodic physical and proficiency examinations. If a pilot is suspected of malfeasance, a suspension and relocation proceeding is conducted. No cases of malfeasance or of reported drunkenness have occurred on the north Tennessee River in the last five years. 2.2.3.3 Evaluation of Hazards from Air Traffic Traffic along Federal Airway V333 is so slight and passes at such an altitude (4000 feet minimum) so as to pose no hazard. 2.2.3.4 Evaluation of the Accidental Release of Toxic Gases from Onsite Storage Facilities Main control room habitability during a postulated hazardous chemical release at or near the plant has been evaluated (reference 3). This evaluation utilizes the approach outlined in Regulatory Guide 1.78 and concludes that the main control room habitability is not jeopardized by accidental release of chemicals stored on site. In addition, plant procedures maintain a list of these hazardous materials, their storage facilities, and quantities they are stored in. 2.2.3.5 Evaluation of the Accidental Release of Toxic Gases from Offsite Storage Facilities There are no industrial or military facilities where large quantities of toxic chemicals could be stored within a 5-mile radius of the plant. S2-2.doc 2.2-7

SQN 2.2.3.6 Evaluation of the Upstream Release of Corrosive Liquids or Oils on the ERCW Intake Structure Protection of the ERCW intake structure from corrosive liquids or oils, released upstream of the plant site, is provided by the mechanics of river flow. The intake structure is located on the inside convex bank of the river bend downstream of a dike rising to an elevation of approximately 700 feet (MSL). The dike coupled with the mechanics of river flow protects the structure. According to flow theory and actual observations made on various rivers, water particles in a bend have a "transverse circulation"; particles near the surface move toward the concave bank and those at the bottom move toward the convex bank. Hence, for normal river levels, the released material would be swept around the intake structure. In the event of liquids or oils reaching the intake structure, no significant effect should occur. Pumps take suction approximately 50 feet below the minimum normal water level and approximately 13 feet below the level anticipated in the event of downstream dam failure. Any oils or fluids which did enter the pumps would be highly diluted and in such a state would have a minimum effect on system piping losses and heat exchanger capabilities. 2.2.3.7 Evaluation of the Potential for Damage to Equipment or Structures Important to Reactor Safety in the Event of the Collapse of Cooling Towers As shown in Figure 2.1.2-1, the natural draft cooling towers are located a distance away from safety-related structures at least equal to the height of the towers above grade. Therefore, if the towers collapse, the function of the safety-related structures will not be impaired. Missiles resulting from flying debris will also not impair the safety-related structures as discussed in Chapter 3. 2.2.3.8 Evaluation of a Release on the Tennessee River of Toxic or Flammable Materials on Plant Safety Features and Control Room Habitability The shipping on the Tennessee River consists mainly of fuel oils, wood products and minerals. Chemicals represent only a minor percentage of the barge shipping by the Sequoyah Nuclear Plant. A list of the commodities shipped passed the Sequoyah Nuclear Plant in 1972 is presented in Table 2.2.3-1. On the average, seven tows per week consisting of three barges passed the Sequoyah site. Of the dangerous cargo traffic, one tow per week consisting of two barges passed the Sequoyah site on the average. The release of flammable or toxic materials on the river in the vicinity of the plant will have no effect on the plant safety features. The ERCW intake pumping station is protected against fire by virtue of design. Pump suction is taken from the bottom of the channel. All pumps and essential cables and instruments are protected from fire by being enclosed within concrete walls. Even if fuel oil from a spill should reach the intake pumping station, the oil would not have significant effect on the water intake system or the systems it serves. Entry of oil in the intake structure is unlikely since oil will float on water. Any oil that did enter the pumps would be highly diluted and in such a state would have a minor effect on system piping losses and heat exchanger capabilities. S2-2.doc 2.2-8

SQN-26 In the event of a release of dense smoke from combustion of flammable liquids in the direction of the control room, personnel in the MCR can manually initiate a CRI which will isolate the control room when a hazardous smoke concentration level is detected. (See sections 6.4 and 9.4.) The Control Room Air Cleanup System has high efficiency particulate filters and charcoal absorbers. A portion of the control room air recirculation flow is also passed through filters. Thus, the concentration of smoke will be maintained at a very low level. In addition, self-contained breathing apparatus will also be available. 2.2.3.9 Evaluation of Potential Fire and Smoke Hazard from Onsite Fuel Oil Storage Facilities The onsite storage facilities for diesel fuel oil are described in detail in Sections 9.5.4.1 and 9.5.4.2. The maximum amount of fuel oil stored at the plant is (1) 68,000 gallons in each of four storage tanks within the diesel generator building, (2) Two 550-gallon "day" tanks are also located within each diesel generator room. (3) Two storage tanks with a capacity of 71,000 gallons each are located south-southeast of the diesel generator building and (4) two 2,900 gallon FLEX diesel generator Day Tanks are located in the Additional Diesel Generator Building. The storage sites are approximately 260 and 300 meters from the control building, respectively. The oil storage tanks in the diesel generator building (DGB) are embedded in a concrete substructure of a Class I seismic building. The storage tanks and diesel generators are separated by thick concrete walls. Fire protection for the DGB is described in the fire protection report (see 9.5.1). A postulated fire involving the oil storage facilities which are located south-southeast of the diesel generator building should have no consequences other than the effects of dense smoke. These tanks are separated from other facilities and are surrounded by a high dike. Additional fuel oil storage tanks have been added to support Diverse and Flexible Coping Strategies (FLEX) operations. These include: (1) Two 185-gallon day tanks for the two 225-kVA diesel generators located on the Auxiliary Building roof (elevation 763.0), and (2) One 9,495-gallon fuel oil storage tank located in the yard south of the Auxiliary Building. Each diesel generator (including day tank) has fire suppression and detection. The fuel oil storage tank is a double wall UL2085 Fireguard tank and because it is located more than 50 feet from a critical building, fire detection and suppression are not required (NFPA 30). The Auxiliary Building roof and Control and Auxiliary Building walls which could be exposed to fire from either of these sources are all credited as 3-hour fire barriers. An evaluation of the hazard to personnel in the control room from a release of dense smoke is given in Section 6.4.1.2. 2.2.4 Forest Fires Further clearing has taken place since the time of plant construction. For the most part, the ground has been cleared for two thousand feet around the plant buildings. There are no wooded areas close enough to present a hazard from forest fires. 2.2.5 References

1. Atomic Energy Commission, WASH-1400-D, Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, 1974.

2. Kondratev, N. E., River Flow and River Channel Formation, Technical Services, U. S. Department of Commerce, 1959.

3. TIC-ECS-27, "Main Control Room Habitability During Hazardous Chemical Releases at or Near the Plant."

S2-2.doc 2.2-9

SQN TABLE 2.2.2-1 HAZARDOUS RIVER TRAFFIC THAT PASSES SEQUOYAH NUCLEAR PLANT 1982 - 1992 (TONS) U.S. ARMY CORPS OF ENGINEERS DATA COMMODITY 1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 2871 Nitrogenous 2,982 20,260 12,417 20,958 19,867 12,1234 11,636 7,591 8,988 NA NA Fertilizer 56216 Urea NA NA NA NA NA NA NA NA 8,988 35,569 24,657 Fertilizers 2911 Gasoline 0 0 0 0 3,287* 0 0 0 0 0 0 2914 Distilate 0 3,325 2,762 0 0 0 0 0 0 0 0 Fuel Oil 2915 Residual 14,223 0 31,008 43,469 21,849 0 25,487 13,375 16,205 NA NA Fuel Oil 33440 Fuel Oils NA NA NA NA NA NA NA NA 16,205 9,105 26,582 NEC 2819 Basic Chems 20,295 0 6,036 4,778 2,906 2,588 3,132 0 46,200 NA NA NEC 52210 Carbon NA NA NA NA NA NA NA NA 0 0 2,869 52224 Chlorine NA NA NA NA NA NA NA NA 46,200 34,100 38,500 TOTAL 37,500 23,585 52,223 69,205 47,909 14,722 40,255 20,966 71,393 77,774 92,608 NA More detailed and specific commodity codes became available in 1990. Duplicate entries are found in 1990 because the old commodity and the new were identical.

The actual product was RU250.

T222-01.doc

SQN Table 2.2.2-1a Hazardous River Traffic That Passes Sequoyah Nuclear Plant Calendar Year 1990 (TVA Survey Data) Asphalt- Five barges/month, two at 3,000 tons/barge and three at 1,500 tons/barge Caustic Soda- One barge/month, 1,400 tons/barge Chlorine- One barge every eight days, 1,100 tons/barge Phosphate- One barge every two months, 1,500 tons/barge Potash- One barge every two months, 1,500 tons/barge Residual Fuel Oil- Three barges every two months, 1,500 tons/barge Sulfate Potash- One barge every four months, 1,500 tons/barge Urea- Six barges per year (three in spring, three in fall), 1,500 tons/barge T222-1ato1b.doc

SQN Table 2.2.2-1b Volunteer Army Ammunition Plant Storage and Transport of Munitions Typical Amount Typical Amount Type of Quantity Stored on Site Shipped Munitions (per case) (cases)* (cases) 7.62 mm 800 rounds 60 15 (machine gun) 5.56 mm 1,680 rounds 30 9 (machine gun) 9 mm 2000 rounds 4 4 (pistol) Hand held 20 2 2 aluminum flares

All munitions stored in a magazine designed to store 500,000 pounds of TNT.

T222-1ato1b.doc

SQN TABLE 2.2.3-1 (Sheet 1) BARGE FREIGHT TRAFFIC PASSING SEQUOYAH NUCLEAR PLANT SITE TENNESSEE RIVER MILE 484.5 Calendar Year 1972 Commodity Net Tons Classed As Wheat 14,516 -- Manganese Ores and Concentrates 20,773 -- Nonferrous Metal Ores 32,110 -- Coal and Lignite 260,959 -- Limestone 826 -- Sand, Gravel, Crushed Rock 9,990 -- Nonmetallic Minerals, nec 38,364 -- Molasses 7,848 -- Pulpwood 234,017 -- Newsprint 89,383 -- Paper and Paperboard 2,912 -- Pulp, Paper, nec 751 -- Caustic Soda, Liquid,* 3,557 Corrosive Liquid Basic Chemicals and Products,* nec 26,471 Inflammable Compressed Miscellaneous Chemical Products* 7,650 Noninflammable Compressed Gas Gasoline* 126,378 Inflammable Liquid Kerosene* 879 Combustible Liquid Distillate Fuel Oil* 2,330 Combustible Liquid T223-1to3.doc

SQN TABLE 2.2.3-1 (Sheet 2) (Continued) BARGE FREIGHT TRAFFIC PASSING SEQUOYAH NUCLEAR PLANT SITE TENNESSEE RIVER MILE 484.5 Calendar Year 1972 Commodity Net Tons Classed As Residual Fuel Oil* 22,520 Combustible Liquid Asphalt Tar and Pitches* 104,696 Hazardous Lime 3,469 -- Misc. Nonmetallic Mineral Product 255 -- Slag 1,595 -- Iron and Steel Ingots 621 -- Iron and Steel Bars, Angles, etc. 1,379 -- Iron and Steel Plates and Sheets 2,395 -- a/* Ferroalloys 10,235 Hazardous Primary Iron and Steel Products, nec 864 -- Copper 8,496 -- Aluminum, Unworked 5,545 -- Machinery, except Electrical 1,854 -- Electrical Machinery 300 -- Nonferrous Metal Scrap 1,554 -- TOTAL 1,045,492 nec - not elsewhere classified

Considered dangerous cargo as set forth in Code of Federal Regulations, Title 46, Parts 146 to 149, revised as of January 1, 1969, pp. 24-27.

a/ If ferrochrome, ferromanganese, or ferrosilicon. Source: Corps of Engineers, Department of the Army. T223-1to3.doc

SQN TABLE 2.2.3-2 (Sheet 1) TENNESSEE RIVER TRAFFIC PASSING SEQUOYAH NUCLEAR PLANT (Tennessee River Mile 484.5) Calendar Year 1978 Code Commodity Net Tons 0107 Wheat 2,773 1011 Iron Ore 14,390 1061 Manganese Ore 152,043 1121 Coal 182,021 1411 Limestone 2,800 1491 Salt 146,036 2062 Molasses 7,985 2415 Pulpwood 317,407 2611 Pulp 32,039 2621 Newsprint 20,882 2631 Paper and Paperboard 7,141 2810 Caustic Soda 7,811 2819 Basic Chemicals, NEC 42,174

                 * (Methyl Methacrylate)                         (37,137) 2871              Nitrogenous Chemical Fertilizers                 4,825 2879              Fertilizers and Materials, NEC                   10,491

2915 Residual Fuel Oil 132,681
2918 Asphalt, Tar and Pitches 151,379 2920 Coke 14,640 T223-1to3.doc

SQN TABLE 2.2.3-2 (Sheet 2) (Continued) TENNESSEE RIVER TRAFFIC PASSING SEQUOYAH NUCLEAR PLANT (Tennessee River Mile 484.5) Calendar Year 1978 Code Commodity Net Tons 3291 Miscellaneous Nonmetallic Minerals 346 3312 Slag 2,918 3314 Iron and Steel Ingots 1,186 3315 Iron and Steel Bars 1,504 3316 Iron and Steel Plates 3,473 3318 Ferroalloys 2,800 3319 Primary Iron and Steel 35 3411 Fabricated Metal Products 125 3511 Machinery 575 3611 Electrical Machinery 150 3711 Motor Vehicles 235 3791 Miscellaneous Transportation Equipment 125 TOTAL 1,262,990 Source: Corps of Engineers, Department of the Army

Flammable liquids as classified in the "Code of Federal Regulations" T223-1to3.doc

SQN TABLE 2.2.3-3 Gasoline Barge Receipts at Port at Knoxville (In Net Tons) Year Net Tons 1960 219,452 1961 143,453 1962 203,625 1963 228,264* 1964 11,084 1965 16,773 1966 2,390 1967 45,079 1968 14,005 1969 36,831 1970 27,361 1971 157,743 1972 126,378 1973 36,506 1974 0**

Pipeline completed 12/63

              ** TVA estimate Source:    "Waterbore Commerce of United States Part II" Department of Army Corp. of Engineers T223-1to3.doc

SQN 2.3 METEOROLOGY 2.3.1 Regional Meteorology 2.3.1.1 Data Sources References used in describing the regional meteorology were the (1) general surface windflow patterns shown by the normal sea level pressure distribution (annual, February, July, and October) for North America and the North Atlantic Ocean--from the U.S. Atomic Energy Commission, ORO-99, A Meteorological Survey of the Oak Ridge Area, Weather Bureau, Oak Ridge, Tennessee, November 1953; (2) wind storm and thunderstorm occurrence--from (a) Local Climatological Data, "Annual Summary with Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Center, 1979, and (b) Severe Local Storm Occurrences, 1955-1967, ESSA Technical Memorandum WSTM FCST 12, U.S. Department of Commerce, Weather Bureau (now NWS), Silver Spring, Maryland, September l969; (3) tornado occurrence--from (a) "Tornado Occurrences in Tennessee, 1916-1964," John V. Vaiksnoras, State Climatologist, U.S. Department of Commerce, Weather Bureau, Nashville, Tennessee, May 5, 1965, (b) "Tornado Probabilities," H. C. S. Thom, Monthly Weather Review, Volume 91, Nos. 10-12, 1963, (c) discussion with John Vaiksnoras, State Climatologist for Tennessee, Nashville, Tennessee, August 3, 1972, (d) "Tornadoes of the United States," Snowden D. Flora, University of Alabama, November 1953, and (e) National Severe Storms Forecast Center tornado data, 1987 (4) air pollution potential--from Mixing Heights, Wind Speeds, and Potential for Urban Air Pollution Throughout the Contiguous United States, George C. Holzworth, Division of Meteorology, Environmental Protection Agency, Preliminary Document, May 10, 1971; and (5) precipitation--from (a) Precipitation in the Tennessee River Basin, TVA, Division of Water Control Planning, Hydraulic Data Branch, period of record 35 years (1935-1969), (b) Local Climatological Data, "Annual Summary with Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Center, 1979, (c) U.S. Army, Domestic Area Section, Glaze - Its Meteorology and Climatology, Geographical Distribution, and Economic Effects, Technical Report EP-105, Quartermaster Research and Engineering Center, Natick, Massachusetts, March 1959, and (d) Ostby, Frederick (Employee of U.S. Department of Commerce, NOAA, NWS, National Severe Storms Forecast Center, Kansas City, Missouri), telephone conversation with TVA meteorologist, Norris Nielsen, September 14, 1973. 2.3.1.2 General Climate The Sequoyah site is in the eastern Tennessee portion of the Southern Appalachian region which is dominated much of the year by the Azores-Bermuda anticyclonic circulation shown in the annual normal sea level pressure distribution (Figure 2.3.1-1). [1] This circulation over the southeastern United States is most pronounced in the fall and is accompanied by extended periods of fair weather and widespread atmospheric stagnation. [2] In winter, the normal circulation pattern becomes diffuse as the eastward moving migratory high and low pressure systems, associated with the midlatitude westerly current, bring alternating cold and warm air masses into the area with resultant changes in wind direction, wind speed, atmospheric stability, precipitation, and other meteorological elements. In summer, the migratory systems are less frequent and less intense, and the area is under the dominance of the western edge of the Azores-Bermuda anticyclone with a warm moist air influx from the Atlantic Ocean and the Gulf of Mexico. S2.3.doc 2.3-1

SQN-18 The terrain features of the region have some effect on the general climate. With the mountain ridge and valley terrain aligned northeast-southwest over eastern Tennessee, there is a definite bimodal upvalley-downvalley windflow in the lower 500 to 1000 feet during much of the year. The high Cumberland Plateau terrain, 1500 to 1800 feet above the valley elevation, tends to moderate many of the migratory storms which move from the west across the region. A detectable lake breeze circulation resulting from discontinuities in differential surface heating between land and water is not expected because of the relatively narrow width of the Tennessee River as it flows southwestward through the valley area. 2.3.1.3 Severe Weather Wind storms may occur several times a year, particularly during winter, spring, and summer with winds exceeding 35 mph and on occasion exceeding 60 mph. The records show the highest wind speed recorded in Chattanooga was 82 mph in March 1947. [3] The highest hourly wind speed recorded at the Sequoyah meteorological facility during the first year of operation, April 2, 1971 -March 31, 1972, was 40 mph. High wind may accompany moderate-to-strong cold frontal passages about 20 to 30 times a year with the maximum frequency in March and April. High wind may accompany thunderstorms, which occur on about 55 days a year with a maximum frequency in July [3]. The distribution of average monthly thunderstorm occurrences recorded during 1931-1979 at the Chattanooga National Weather Service Office is as follows: Jan. Feb. Mar. Apr. May June July Aug. Sep. Oct. Nov. Dec. Annual 1 2 4 5 7 l0 11 9 4 1 1 1 56 Severe storm data for 1955-1967 [4] show l0 occurrences of hail 3/4 inch or greater in diameter, 20 occurrences of wind storms with speeds of 50 knots or greater, and 15 occurrences of tornadoes in the one degree latitude-longitude square containing the site. If these severe storm occurrences are assumed to be exclusive of one another, it can be assumed that about 45 severe thunderstorms occurred in the one degree square in this 13-year period. The annual occurrence for the square would be about 3.5. A smaller annual occurrence would be expected for the immediate site area, which is much smaller than the one degree square for which these statistics apply. The probability of tornado occurrence is extremely low. Statistics show that during the 49-year period, 1916-1964, no tornadoes were reported in Hamilton County, where the Sequoyah site is located. [5] During the 1965-1986 period, three tornadoes were reported in the county. [18] During 1987-October 2002, seven tornadoes were reported in the county. [24] During 1955-1967, a total of 15 tornadoes was recorded for the one degree latitude-longitude square containing the site, for an annual occurrence of 1.15. [4] Using the principles of geometric probability described by H. C. S. Thom, [6] 2 his frequency data for that 1-degree square, and a tornado path size of 0.284 mi , [7] the probability of

                                                                -5 a tornado striking any point in the plant site area is 4.4 x 1O .

The National Severe Storms Forecast Center in Kansas City, Missouri calculated the tornado return probability for the Sequoyah site based on tornado occurrences within a 30 nautical mile (nm) radius during 1950-1986.[18] A circle of 30 nm radius has an area comparable to a one S2.3.doc 2.3-2

SQN-21 degree latitude-longitude square. Based on the 29 tornado occurrences with path size estimates in the 37-year period, the return probability is 1.635 x 10-4 and the mean return interval is 6,115 years. The annual tornado occurrence in the 30nm radius circle was 0.84 (based on 31 tornadoes reported) during that period. During the subsequent period spanning 1987 through October 2002, 23 tornadoes were reported in the same circle. [24] Thus, for the period spanning 1950 through October 2002, 54 tornadoes occurred for an annual occurrence of 1.02. Given the typically small path size of these tornadoes, the return probability and return interval given above should still be representative. Tornadoes in the eastern Tennessee area generally move northeasterly and cover an average surface path five miles long and one hundred yards wide. [7] Winds of 150 to 200 mph are common in the whirl and are estimated to occasionally reach 300 mph. [7,8] Days of high air pollution potential, shown in Figure 2.3.1-2, have been depicted by G. C. Holzworth, who presents an expected frequency of high meteorological potential for air pollution. [9] Over a five-year period, his data show that there were about thirty days, or about six days annually, that such conditions could have affected the site area, with most of the days occurring in the fall. The highest monthly average rainfall near the site area occurs during the winter and early spring months, with March usually having the greatest amount. [10] The maximum 24-hour rainfall reported near the plant site was 7.56 inches in August. High precipitation is also observed in July when air mass thunderstorm activity is common. Minimum precipitation occurs normally in October. The occurrence of snow, freezing rain, and ice storms in the mid-winter period is not uncommon. During 1931-1995, the maximum total monthly snowfall recorded at Chattanooga was 20.0 inches in March 1993. [25] The average annual snowfall for this period was 4.4 inches. The best estimate of the 100-year recurrence snowfall from a single storm is 14.5 inches which fell during a period from December 4, 1886 through December 6, 1886. [19] The maximum amount on the ground at any one time was 19 inches. This March 1993 24-hour storm was the maximum that occurred in 118 years of record at Chattanooga, Tennessee. No greater single storm or monthly amounts were observed in the southeastern Tennessee area around the plant site through July 2002. [26] The record depth of snow is below the maximum that the safety-related structures can withstand. Assuming the 20-inch snowfall was the depth on top of above ground structures, this equates to a snow load of 14.6 pounds per square foot compared to the design snow load of 20 pounds per square foot. Design criteria for the roofs of safety-related structures is given in Section 3.8. From 1917-18 to 1924-25, there were about three observations of ice storms heavy enough to damage telephone and telegraph lines in the Sequoyah site area. [ll] At least three and perhaps as many as six glaze storms occurred in the general area of the site from 1925-26 to 1952-53. There were about four glaze storms with ice thickness 1/4-inch or more during the period 1928-29 to 1936-37. Also, from 1939 to 1948, freezing rain or drizzle of a trace (0.01 inch) or more occurred on about two days a year. Hail storms of significant intensity (hailstones 3/4 inch or more in diameter) would likely never occur in the plant area. [7] The probability of occurrence of such a storm can be calculated using Thom's 2 tornado probability equation. [6] With a mean hail path area of two mi. (1/2 mi. by 4 mi.) [12], an annual occurrence (of hail 3/4 inch or more in diameter) of 0.77 [4], and an area of 3887 mi.2 for the one degree latitude-longitude square containing the site [6], the probability is calculated to be 3.96 x 10-4. Lightning strike density in the vicinity of the plant has been computed to be an average of about 8 ground strikes per square kilometer per year. [27] These are defined as cloud to ground strokes of lightning. S2.3.doc 2.3-3

SQN-18 2.3.2 Local Meteorology 2.3.2.1 Data Sources Most of the data used in this meteorological description were collected at the onsite meteorological facility (Environmental Data Station) in the four-year period from January 1, 1972 through December 31, 1975. Location of this facility with respect to the Sequoyah Nuclear Plant is shown in Figure 2.3.2-1. A one-year period (May 1, 1975 - April 30, 1976) of wind and temperature data was used for comparison of stability classifications based on hourly-average vertical temperature difference (WT) values with those based on end-of-hour WT values. This comparison was done to determine any effects on the stability class frequency distribution and the joint wind speed and wind direction frequency distributions by stability class resulting from the change in temperature recording procedure from an end-of-hour reading to an hourly-average value. Because of the limited period of onsite data, long-term fog and snowfall trends as well as supplementary temperature information were obtained from data records for the National Weather Service Office at Lovell Field, Chattanooga, located 14.5 miles south-southwest of the site (Figure 2.3.2-2). Precipitation data were obtained from a 20-year record from the TVA rain gauge station 685, Friendship School, Tennessee, located about 2.5 miles north-northeast of the plant site. 2.3.2.2 Normal and Extreme Values of Meteorological Parameters With the limited period of onsite data, it is not reasonable to discuss normal and extreme values of meteorological parameters measured onsite; instead, the data should point toward representative mean values of the local meteorological parameters. Therefore, normal and extreme values of parameters measured offsite should be more representative of long-term regional climate, although local site influences may not be reflected. Wind Direction Data from the 33-foot wind instruments at the permanent meteorological facility for the January 1972 - December 1975 period represent reasonably well the expected wind conditions in the plant site area. The annual and monthly patterns (Tables 2.3.2-1 through 2.3.2-13 and Figures 2.3.2-3 through 2.3.2-15) show the predominant directions from the northeast and southwest quadrants which reflect the orographic channeling effects of the northeast-southwest aligned valley-ridge terrain. For most of the months, but especially for the cooler months of the year, there is a weak secondary maximum of wind frequency from the northwest quadrant. This is most likely associated with post cold frontal winds, which are most likely during the optimum seasons (winter and early spring) for frequent migratory low pressure systems. Wind Direction Persistence The wind direction persistence1 analysis (based on the 33-foot (10-meter) data) shown in Table 2.3.2-14, gives the persistence for periods two hours or more from the given wind directions. The greatest persistence was from the north-northeast, which included the maximum of 33 S2.3.doc 2.3-4

SQN-18 hours. Persistence of 24 hours or more occurred with winds from the southwest, north, and northeast. The analysis shows that the occurrence of persistence periods lasting three hours or more is about 59 percent. For 12 hours or more, the occurrence is about four percent. Wind Speed The seasonal and annual occurrences of wind speed at the 33-foot tower level for all wind directions are shown in Tables 2.3.2-1 through 2.3.2-13 and Figures 2.3.2-3 through 2.3.2-15. The preponderance of winds from the northeast within the 0.6 to 3.4 mph wind speed range is most likely attributable to the anticyclonic circulation that dominates the eastern Tennessee region in the late summer and fall. Also, the identification of wind speeds less than 3.5 mph with stable anticyclonic flow is reflected in the high frequency of occurrence of this range in late summer and early fall--a period during which stable anticyclonic conditions are most common. On the other hand, these low wind speeds occur least often in winter and early spring--a period frequented by the passage of migratory low pressure systems. Wind speeds 7.5 mph and greater occurred most frequently with upvalley winds (from the southwest). These wind speeds occurred very infrequently with winds from the east-northeast, east, east-southeast, and southeast. The predominance of strong winds from the southwest may be attributable to the channeling of the southerly and southwesterly flow preceding the passage of cold fronts through the area. Winds greater than 7.5 mph were more frequent from November through April, with a maximum of about 32 percent in April; they occurred least often in July and August. 1 Persistent wind is defined in this analysis as a continuous wind from one of the 22-1/2 degree sectors (e.g., north-northeast) except that 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. Temperature A summary of the first year (April 2, 1971 - March 31, 1972) of onsite temperature data from the meteorological facility is shown in Table 2.3.2-15. The average annual temperature was 59.7°F with the range of monthly averages from 40.1°F in February to 75.5°F in August. The extreme maximum and minimum were 96.3°F and 2.9°F in June and January, respectively. Onsite temperature data compare reasonably well with the normal temperature records from the Chattanooga National Weather Service Office (Weather Bureau) shown in Table 2.3.2-16, although extremes of temperature from the one year of onsite data are somewhat conservative as compared to extremes for Chattanooga. [3] [25] Atmospheric Water Vapor The first year of onsite temperature and dew point data were used to compute mean and extreme values of absolute and relative humidity shown in Tables 2.3.2-17 and 2.3.2-18. The average annual absolute humidity was 9.7 g/m3 with the range of monthly averages from 16.2 g/m3 in June to 4.2 g/m3 in February. The extreme maximum was 22.3 g/m3 in June and the extreme minimum was 1 g/m3 in February. S2.3.doc 2.3-5

SQN-18 The average annual relative humidity was 66.5 percent with the range of monthly averages from 50.6 percent in April to 78.4 percent in October and December. The extreme maximum was 100 percent in March, June, September, November, and December, and the extreme minimum was 17 percent in April. Precipitation Precipitation patterns, based on a 20-year period (1948-1967) of data collection at the TVA rain gauge station 685, 2.5 miles north-northeast of the plant site, are shown in Table 2.3.2-19. [10] The data show that there was an average of 117 days annually with 0.01 inch or more of precipitation. The average monthly precipitation was 4.81 inches, with the maximum monthly average 6.76 inches occurring in March and the minimum monthly average 2.86 inches occurring in October. The extreme monthly maximum and minimum were 16.58 inches in November and 0.09 inch in October, respectively. This station was discontinued after 1972, but examination of records for 1968-1972 showed no changes in extremes. [28] Also, the extreme maximum and minimum values in Table 2.3.2-19 have not been exceeded at the Chattanooga airport station during the 1940-2002 period. [25] Snowfall does not occur often in the Sequoyah site area. Chattanooga snowfall data in Table 2.3.2-20 are considered representative. [25] The average annual snowfall was 4.4 inches and occurred mostly in December through March. The maximum 24-hour snowfall reported at Chattanooga was 20.0 inches in March 1993; the next highest was 10.2 inches in January 1988. Fog No observations of the frequency and intensity of fogs have been made in the site area. However, Chattanooga National Weather Service records (Table 2.3.2-21) indicate that heavy fogs (visibility of 1/4 mile or less) occurred on an average of 36 days annually with a maximum average monthly frequency of six days in October and a minimum average monthly frequency of two days from February through July. [3] Atmospheric Stability At the present time, atmospheric stability is calculated from the difference between the hourly-average temperature values from two levels. Prior to January 8, 1975, the temperature difference was calculated by a high speed digital computer that was programmed to convert the difference between the ambient temperature sensor resistances at any two instrument levels to a temperature difference value (WT). Before January 8, 1975, both temperature and temperature difference data were obtained from end-of-hour readings. Four years (January 1, 1972 - December 31, 1975) of onsite temperature difference data from the 33-and 150-foot (9- and 46-meter) tower levels of the permanent meteorological facility were categorized into seven atmospheric stability groups (Pasquill classes A through G). Table 2.3.2-22 shows that the Pasquill stability classes E, F, and G occurred about 72 percent of the time. The most stable class, G, occurred about seven percent of the time. The total occurrence of the least stable classes, A, B, and C, was about eight percent, while the neutral stability class, D, occurred about 20 percent of the time. S2.3.doc 2.3-6

SQN Joint percentage frequencies of wind direction and wind speed for the Pasquill stability classes A through G are summarized in Tables 2.3.2-23 through 2.3.2-29 and Figures 2.3.2-16 through 2.3.2-22. The most critical conditions, class G and wind speeds less than 3.5 mph (Table 2.3.2-29, Figure 2.3.2-22), occurred less than six percent of the time. Stability category G is most often associated with downvalley winds (from the north-northeast and northeast), with a secondary maximum associated upvalley winds (from the southwest and south-southwest). Annual frequencies for classes E and F (Tables 2.3.2-27 and 2.3.2-28) show respective frequencies of about 17 and 15 percent for wind speeds less than 3.5 mph. Using the same type of instrumentation, the capability for calculating hourly average T values (based on hourly-average temperature values) was established in January 1975. A special adjustment of the computer program developed for this purpose was made to also obtain instantaneous, end-of-hour T values for comparison with the hourly-average values. Table 2.3.2-30 provides the frequencies for hourly-average and end-of-hour stability classes (Pasquill A-G), and Tables 2.3.2-31 through 2.3.2-58 provide joint frequencies of wind direction and wind speed by stability class, each for hourly-average and end-of-hour T values. Summaries based on hourly-average and end-of-hour T values are presented for 33- to 150-foot T and 33-foot wind direction and wind speed data, and for 33- to 300-foot T and 300-foot wind direction and wind speed data. The same wind direction and wind speed data were used with the hourly-average and the end-of-hour T data. 2.3.2.3 Potential Influence of the Plant and its Facilities on Local Meteorology The presence and operation of the Sequoyah Nuclear Plant should have no noticeable effects on the local meteorology, with the exception of a slight increase in frequency, duration, and intensity of steam fogs forming at the river surface due to heated water releases through the diffusers. These fogs develop as a result of elevation of the dew point by the addition of moisture to the air from the water surface. Once this shallow fog moves on shore, the moisture source is cut off and the fog dissipates. Thus, the increased fogging should be confined within the boundaries of the Chickamauga Reservoir and should not affect long-term fog patterns in the surrounding area. This phenomenon has been observed frequently over the extended river and reservoir system within the Tennessee Valley Region. Based on previous experience with natural-draft cooling tower operation at the TVA Paradise Steam Plant, no adverse impact on the local meteorology is expected from the operation of supplemental natural-draft cooling towers at the Sequoyah Plant. Some minor effects may include increased atmospheric moisture, decreased solar radiation, and increased concentrations of aerosols related to the drift. However, the significance of these effects would be very difficult or impossible to measure. 2.3.2.4 Topographical Description The principal effect of the topography in the Sequoyah area on the diffusion of effluent releases is one of confinement to the downwind sectors of predominant wind. Figure 2.3.2-23, sheets 1-9, shows the topographic features within five miles and topographic cross sections in the 16 compass sectors. Annually, the majority of the releases of radioactive effluent would be S2.3.doc 2.3-7

SQN-24 dispersed within the northeasterly and southwesterly quadrants from the plant as a result of the upvalley-downvalley low-level wind. Therefore, relative ground-level concentrations would be expected to be higher in these sectors, particularly during periods of low wind and stable conditions. Also, with the relatively flat and undulating valley floor, there should be minimal discontinuity of the general low-level wind pattern from terrain roughness or irregularity. Furthermore, differences in the ambient thermal or stability structure in the area from differential surface heating between land and water should not cause significant alterations to the wind and stability patterns in the plant area. On rare occasions, slight buildup of effluent concentration could occur in the Cumberland escarpment area, about 15 miles to the northwest, where some geographically induced impingement or entrapment of the effluent might be expected. 2.3.3 On-Site Meteorological Measurement Program 2.3.3.1 Siting and Description of Instruments The Sequoyah 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 an Environmental Data Station (EDS), which houses the data collection and recording equipment. A system of lightning and surge protection circuitry with proper grounding is included in the facility design. This facility is located approximately 0.74 miles (1.2 kilometers) southwest of the Reactor Building and about 50 feet (15 meters) above plant grade (Figure 2.3.2-1). Rainfall is monitored from a rain gauge located approximately 55 feet from the tower. Data collected include: (1) wind speed and direction at 10, 46, and 91 meters (33, 150, and 300 feet), (2) temperature at 10, 46, and 91 meters; (3) dewpoint at 10 meters; and (4) rainfall at 1 meter (3 feet). More exact measurements heights for wind and temperature sensors are given in EDS Manual [Reference 20]. Elsewhere in this document, temperature and wind sensor heights are given as 10, 46, and 91 meters. Collection of onsite meteorological data at the Sequoyah Nuclear Plant commenced in April 1971 with measurements of wind speed and wind direction at 10 meter and 91 meters, temperature at 1, 10, 46, and 91 meters; and dewpoint and rainfall at 1 meter. Measurements of 46 meter wind speed/direction and 10 meter dewpoint began on August 6, 1976. Measurement of 1 meter dewpoint ended on January 9, 1979. Measurement of 1 meter temperature ended on January 10, 1979. The dewpoint sensor was moved to a separate tower on June 7, 1994. 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). [Reference 13] SENSOR HEIGHT (meters) DESCRIPTION Wind Direction and 10, 46, and 91 Ultrasonic Wind Sensor. Wind Speed S2.3.doc 2.3-8

SQN-23 SENSOR HEIGHT (meters) DESCRIPTION Temperature 10, 46, and 91 Platinum wire resistance temperature detector (RTD) with aspirated radiation shield. Dewpoint 10 Capacitive Humidity Sensor. Rainfall 1 Tipping bucket rain gauge. 2.3.3.2 Data Acquisition System The data acquisition system is located at the EDS and consists of meteorological sensors, a computer (with peripherals), and various interface devices. These devices send meteorological data to the plant, to the Central Emergency Control Center (CECC), and to enable callup for data validation and archiving offsite. S2.3.doc 2.3-9

SQN-24 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 satisfies the R.G. 1.23 accuracy requirements. A detailed listing of error sources for each parameter is included in the EDS manual. 2.3.3.3 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 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 of dewpoint, and hourly horizontal wind direction sigmas. Vector wind speed and direction are calculated along with arithmetic average wind speed. Selected data each 15 minutes and all data each hour are stored for remote data access. Data sent to the plant computer systems every minute includes 10, 46, and 91 meter values for wind speed, wind direction, and temperature. S2.3.doc 2.3-10

SQN-23 Data sent to the Central Emergency Control Center (CECC) computer in Chattanooga every 15 minutes includes 91-, 46-, and 10-meter wind direction, wind speed, and temperature values. These data are available from the CECC computer to other TVA and State emergency centers in support of the Radiological Emergency Plan (REP), including the Technical Support Center at Sequoyah. Remote access of meteorological data by the NRC is available through the CECC computer. Data are sent from the EDS to an offsite computer for validation, reporting, and archiving. 2.3.3.4 Equipment Servicing, Maintenance, and Calibration The meteorological equipment at EDS is kept in proper operating condition by staff that are trained and qualified for necessary tasks. 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 by a laboratory calibrated component. More frequent calibration intervals for individual components may be conducted, on the basis of the operational history of the component type. Detailed procedures are used and are referenced in the EDS Manual. 2.3.3.5 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. This phase of the meteorological data collection program will be continuous without major interruptions. The meteorological program has been developed to be consistent with guidance given in NRC Regulatory Guide 1.23 (Revision 1) and the reporting procedure in Regulatory Guide 1.21 (Revision 1). [Reference 14] The basic objective is to maintain data collection performance to assure at least 90 percent 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 maintenance 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. S2.3.doc 2.3-11

SQN-28 In the event that the onsite meteorological facility is rendered nonfunctional, or there is an outage of communications or data access systems; there is no fully representative offsite source of meteorological data for identification of atmospheric dispersion conditions. Therefore; TVA has prepared objective backup procedures to provide estimates 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 backup procedures. 2.3.4 Short-Term (Accident) Diffusion Estimates 2.3.4.1 Objective Two sets of atmospheric dilution factors (X/Q values) are currently used for accident releases modeled as ground level releases from the Sequoyah Nuclear Plant for specified time intervals and distances. The first set is based on one year (April 2, 1971 through March 31, 1972) of data from the Sequoyah permanent meteorological facility. Part of this set was used in the design accident dose calculations and is shown in Table 15A-2. The latest and most widely used set is based on four years (January 1972 through December 1975) of data (Tables 2.3.2-23 through 2.3.2-29). This data was used in Chapter 11. 2.3.4.2 Calculations Two mathematical models were used in estimating atmospheric dilution factors during postulated reactor accidents - one for the 1-hour and 8-hour (0-8 hours) averaging periods and the other for the 16-hour (8-24 hours), 3-day (1-4 days), and 26-day (4-30 days) averaging periods. Calculations with the two models utilize hourly values of wind direction, wind speed, and atmospheric stability (Pasquill classes A through G). Nomenclature A = minimum cross-sectional area of the Reactor Building (m2) c = an empirical constant used in defining the magnitude of the building wake (dimensionless) Q = source strength or effluent release rate (curies/sec) u = mean horizontal wind speed at 10 meters (m/sec) x = distance from effluent release point to point at which X/Q values are computed (m)

= 3.1416 S2.3.doc                                           2.3-12

SQN-20 y = Pasquill horizontal crosswind plume standard deviation (m) z = Pasquill vertical plume standard deviation (m) x = ground-level concentration (curies/m3) Model for the 1-Hour and 8-Hour Averaging Periods Atmospheric dilution factors were calculated for the 1-hour and 8-hour averaging periods using a Gaussian centerline building wake diffusion equation discussed in NRC Regulatory Guide 1.4 (Revision 2) [15] and Slade [16]: 1 X / Q= (1) ( Y Z + cA)u where cA is a building wake factor. Model for Averaging Periods Greater than 8 Hours Atmospheric dilution factors were calculated for the 16-hour, 3-day, and 26-day averaging periods using a Gaussian sector average building wake diffusion equation presented in NRC Regulatory Guide 1.4 (Revision 2): 2.032 X / Q= (2) Z xu For this model, it is assumed that sufficient time elapses to allow the plume to meander and uniformly spread across the 22-1/2-degree downwind sector. Locations for Which Atmospheric Dilution Factors Were Calculated and Effluent Release Zones Atmospheric dilution factors were calculated for two location categories: (1) exclusion area boundary, and (2) outer boundary of the Low Population Zone (LPZ). The effluent release zones for the Sequoyah Plant were defined for three locations (see Figure 2.1.2.-2): (1) Release Zone 1, the Auxiliary Building vent exhaust and the Shield Building vent exhaust; (2) Release Zone 2, the radioactive chemical hood exhaust; and (3) Release Zone 3, the condenser air ejector exhaust. Atmospheric Dilution Factors for the Exclusion Area Boundary Each release zone was considered individually in calculating atmospheric dilution factors at the exclusion area boundary. The distances from each effluent release zone to the intersections of the 16 compass-point directional sectors with the exclusion area boundary are shown in Table 2.3.4-1. S2.3.doc 2.3-13

SQN The hourly average wind speed and atmospheric stability were obtained for a given hour in the January 1972 - December 1975 data period. These data were used with equation (l) to calculate an atmospheric dilution factor corresponding to the exclusion area boundary distance for a particular release zone. This procedure was repeated for each release zone as frequently as there was valid hourly meteorological information available during the 48-month period. These calculations resulted in a list of hourly values for each of the three release zones which were tabulated into cumulative frequency distributions and are shown in Tables 2.3.4-2, 2.3.4-3, and 2.3.4-4 corresponding to Release Zones 1, 2, and 3, respectively. The 5th and 50th percentile and average values of the atmospheric dilution factors for each release zone were also computed and follow: One-Hour Atmospheric Dilution Factors At Exclusion Area Boundary (sec/m3) Release 5th 50th Zone Percentile Percentile Average 1 0.859 x 10-3 0.163 x 10-3 0.269 x 10-3 2 0.795 x 10-3 0.145 x 10-3 0.243 x 10-3 3 0.892 x 10-3 0.164 x 10-3 0.279 x 10-3 A more conservative approach consisted of using the above procedure except selecting the shortest distance from each release zone to the exclusion area boundary and calculating the atmospheric dilution factor for all directions using this fixed distance. The minimum distances as shown in Table 2.3.4-1 are 556 meters, 600 meters, and 509 meters for Release Zones 1, 2, and 3, respectively. The calculations resulted in a list of hourly values for each of the three release zones. These values were tabulated into cumulative frequency distributions as shown in Tables 2.3.4-5, 2.3.4-6, and 2.3.4-7, corresponding to Release Zones 1, 2, and 3, respectively. The 5th and 50th percentile and average atmospheric dilution factors follow: One-Hour Atmospheric Dilution Factors At Exclusion Area Boundary (sec/m3) Release 5th 50th Zone Percentile Percentile Average 1 0.147 x 10-2 0.234 x 10-3 0.396 x 10-3 2 0.130 x 10-2 0.215 x 10-3 0.365 x 10-3 3 0.162 x 10-2 0.258 x 10-3 0.435 x 10-3 S2.3.doc 2.3-14

SQN Atmospheric Dilution Factors for Outer Boundary of the LPZ Atmospheric dilution factors for the outer boundary of the LPZ were calculated by considering a single source or release zone that was assumed to be representative of the three actual release zones. Unlike the calculations for the actual exclusion area boundary in which distances changed with direction, the distance of 4828 meters was used for all calculations for the outer boundary of the LPZ. These values were calculated for averaging times of 1 hour, 8 hours, 16 hours, 3 days, and 26 days. All 1-hour average values were obtained by use of equation (1) and the hourly meteorological observations. The cumulative frequency distribution of these values is listed in Table 2.3.4-8. The 5th and 50th percentile and average values are also shown. For a given sector, the 8-hour average atmospheric dilution factor was obtained by averaging the hourly values. For a given 8-hour period, sixteen 8-hour averages were obtained--one for each compass-point sector. The average value selected to represent the given 8-hour period was the maximum of the sixteen. There were 35,057 8-hour periods from January 1, 1972 through December 31, 1975 where consecutive 8-hour periods overlapped for seven hours. An atmospheric dilution factor was not calculated for an 8-hour period unless there were at least four hours of valid meteorological observations during the period. After the values were computed for the valid 8-hour periods, they were summarized into the cumulative frequency distribution shown in Table 2.3.4-9. The average and 5th and 50th percentile statistics were also computed. All other averages (the 16-hour, 3-day, and 26-day averages) were treated in a fashion analogous to the 8-hour average except that equation (2) was used to calculate the atmospheric dilution factors. Tables 2.3.4-10, 2.3.4-11, and 2.3.4-12 summarize the cumulative frequency distributions of the values for the corresponding l6-hour, 3-day, and 26-day averaging periods, respectively. The 5th and 50th percentile and average values for each averaging period are included in the following table: Atmospheric Dilution Factor at Outer Boundary of LPZ (sec/m3) Averaging 5th 50th Time Percentile Percentile Average 1-hour 0.139 x 10-3 0.142 x 10-4 0.319 x 10-4 8-hour 0.539 x 10-4 0.980 x 10-5 0.169 x 10-4 16-hour 0.717 x 10-5 0.236 x 10-5 0.299 x 10-5 3-day 0.434 x 10-5 0.176 x 10-5 0.201 x 10-5 26-day 0.271 x 10-5 0.153 x 10-5 0.148 x 10-5 Data from the one-year period (May 1, 1975 through April 30, 1976) were used to compare atmospheric dilution factors obtained from stability classes determined from end-of-hour S2.3.doc 2.3-15

SQN-21 temperature measurements and those determined from hourly average temperature measurements. These data (Tables 2.3.2-31 through 2.3.2-44) include wind direction and wind speed at 33 feet (10 meters) above ground and temperature difference between the elevations of 33 and 150 feet (46 meters). Table 2.3.4-13 compares atmospheric dilution factors based on (1) hourly-average T data and (2) end-of-hour T data. The values presented for comparison are fifth percentile values for 1-hour and 8-hour periods at the minimum exclusion area boundary distance of 556 meters and for 8-hour, 16-hour, 3-day, and 26-day periods at the LPZ distance of 4828 meters. It is apparent from examination of the data tables that the differences between atmospheric dilution factors obtained from the data set containing hourly-average T and those obtained from the data set containing end-of-hour T are not significant. The joint frequencies of wind direction and wind speed by atmospheric stability class for 33- to 300-foot T and 300-foot wind data show even closer agreement than those based on 33- to 150-foot T and 33-foot wind data. Therefore, any calculations based on end-of-hour 33- to 300- foot T, or even 150- to 300-foot T, could be expected to be at least as representative of those based on hourly-average T as those for 33- to 150-foot T and 33-foot wind data presented in Table 2.3.4-13. 2.3.5 Long-Term (Routine) Diffusion Estimates 2.3.5.1 Objective In this section, calculated average annual atmospheric dispersion factors (X/Q values) are reported at specified distances for routine releases from the Sequoyah Nuclear Plant. A dispersion equation is applied which accounts for initial dilution of gaseous effluents in the building wake. Joint frequency distributions of wind direction and speed by atmospheric stability class based on onsite meteorological data collected during the period of January 1972 through December 1975 are used in the calculations. Joint frequency distributions are presented in Tables 2.3.2-23 through 2.3.2-29. 2.3.5.2 Calculations Average annual atmospheric dispersion factors are calculated for locations along 16 radial lines corresponding to the major compass points drawn from the center of the nuclear plant complex. Calculations in each of the 16 sectors are made for the site boundary and for the distances 1, 2, 3, 4, 5, 10, 15, 20, 30, 40, and 50 miles. Three effluent release zones are designated for calculating atmospheric dispersion factors at the site boundary (see Figure 2.1.2-2). These are as follows: Release Zone 1 - Auxiliary Building vent exhaust and Shield Building vent exhaust. Release Zone 2 - Radioactive chemical hood exhaust. Release Zone 3 - Condenser air ejector exhaust. S2.3.doc 2.3-16

SQN-21 In calculating the average annual atmospheric dispersion factors for the selected distances between 1 and 50 miles, it is assumed that gaseous effluents are released from a single point (the three release zones are not considered in these calculations). The distances to the unrestricted area boundary from this point are shown in Table 11.3.9-1. Atmospheric dispersion calculations are based on a building wake model described by Davidson [16,17]. The average annual atmospheric dispersion factor at any point of interest x is given by: wind stability speeds types X 2 1/ 2 1 f ij

        =                 i  j                  , SEC / m3 Qo                  W                     )

( z j Ui where W = 2p x/16, the sector width at downwind distances x, m, ui = wind speed i, m/s, fij= frequency with which wind speed ui occurs in the sector of interest during atmospheric stability class j, 1/2 cA ( z ) j = ( z )j 2 + the vertical standard deviation of the plume (modified for the effect of building wake dilution) at the distance x for stability class j, m, ( z ) j = Pasquill vertical standard deviation of the plume at the distance x for stability class j, m, c = parameter that relates the cross-sectional area of a building to the size of the turbulent wake caused by the building, A = minimum Reactor Building cross-sectional area, m2. In the expression for ( z ), c is assumed to be 0.5 and A is assumed to be 1,800 m2. Table 2.3.4-14 lists average annual atmospheric dispersion factors for the Sequoyah site. 2.3.6 References

1. U.S. Atomic Energy Commission, A Meteorological Survey of the Oak Ridge Area, ORO-99, Weather Bureau, Oak Ridge, Tennessee November 1953, page 377.

S2.3.doc 2.3-17

SQN-23

2. U.S. Atomic Energy Commission, A Meteorological Survey of the Oak Ridge Area, ORO-99, Weather Bureau, Oak Ridge, Tennessee, November 1953, page 192.

3. Local Climatological Data, "Annual Summary With Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Center, Asheville, North Carolina, 1979.

4. Severe Local Storm Occurrences, 1955-1967, ESSA Technical Memorandum WSTM FCST 12, U.S. Department of Commerce, Weather Bureau (now NWS), Silver Spring, Maryland, September 1969.

5. "Tornado Occurrences in Tennessee, 1916-1964," John V. Vaiksnoras, U.S. Department of Commerce, Weather Bureau, Nashville, Tennessee, May 5, 1965.

6. "Tornado Probabilities," H. C. S. Thom, Monthly Weather Review, Volume 91, Nos. 10-12, 1963, pp. 730-736.

7. Discussion with John Vaiksnoras, State Climatologist for Tennessee, Nashville, Tennessee, August 3, 1972.

8. "Tornadoes of the United States," Snowden D. Flora, University of Oklahoma Press, November 1953.

9. Mixing Heights, Wind Speeds, and Potential for Urban Air Pollution Throughout the Contiguous United States, George C. Holzworth, Division of Meteorology, Environmental Protection Agency, Preliminary Document, May 10, 1971.

10. Precipitation in the Tennessee River Basin, Tennessee Valley Authority, Division of Water Control Planning, Hydraulic Data Branch, period of record 35 Years (1935-1969).

11. Glaze - Its Meteorology and Climatology, Geographical Distribution, and Economic Effects, Technical Report EP-105, U.S. Army, Domestic Area Section, Quartermaster Research and Engineering Center, Environmental Protection Research Division, Natick, Massachusetts, March 1959.

12. Ostby, Frederick, employee of U.S. Department of Commerce, NOAA, NWS, National Severe Storms Forecast Center, Kansas City, Missouri, telephone conversation with TVA meteorologist, Norris Nielsen, September 14, 1973.

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

14. 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. S2.3.doc 2.3-18

SQN-23

15. Regulatory Guide 1.4, Revision 2, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," U.S. Atomic Energy Commission, Washington, D.C., June 1974.

16. Meteorology and Atomic Energy, D. H. Slade, ed., USAEC Report, TID-24190, July 1968.

17. Atmospheric Diffusion Experiments with SF6 Tracer Gas at Three Mile Island Nuclear Station Under Low Wind Speed Inversion Conditions, Pickard, Lowe, and Associates, Inc., The Research Corporation of New England, General Public Utilities Service Corporation, January 1972.

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

19. "Extremes of Snowfall: United States and Canada," Weatherwise, Published for American Meteorological Society by Weatherwise, Inc., December 1970.

20. Sequoyah Nuclear Plant Environmental Data Station Manual, Tennessee Valley Authority.

21. Deleted by Amend 15

22. Deleted by Amend 15

23. Deleted by Amend 23.

24. Storm Data, U.S. Department of Commerce, NOAA, National Climatic Data Center, Asheville, North Carolina, Volume 29, Number 1 through Volume 44, Number 10, January 1987 - October 2002.

25. Local Climatological Data, Annual Summary with Comparative Data, Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Data Center, Asheville, North Carolina, 2002.

26. Climatological Data, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Data Center, Asheville, North Carolina, Volumes 101-107, July (Number 7) Issues, 1996 - 2002.

27. Lighting Strike Density for the Contiguous United States From Thunderstorm Duration Records, NUREG/CR-3759, prepared by NOAAs National Severe Storms Laboratory, Norman, Oklahoma, for Division of Health, Siting, and Waste Management, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1984.

28. Precipitation in the Tennessee River Basin, Tennessee Valley Authority, Division of Water Control Planning, Hydraulic Data Branch, 1968 - 1972.

S2.3.doc 2.3-19

SQN TABLE 2.3.2-1 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.51 3.20 1.63 0.67 0.58 0.0 0.0 0.0 6.59 NNE 0.82 8.30 5.05 2.46 2.18 0.11 0.0 0.0 18.92 NE 0.48 3.86 2.59 1.01 0.83 0.06 0.0 0.0 8.83 ENE 0.42 1.58 0.39 0.09 0.0 0.01 0.0 0.0 2.49 E 0.50 0.80 0.11 0.03 0.02 0.01 0.0 0.0 1.47 ESE 0.33 0.45 0.07 0.02 0.01 0.02 0.0 0.0 0.90 SE 0.34 0.82 0.19 0.01 0.02 0.0 0.0 0.0 1.38 SSE 0.41 1.36 0.55 0.23 0.36 0.06 0.02 0.0 2.99 S 0.47 2.89 2.49 1.58 1.53 0.14 0.0 0.0 9.10 SSW 0.29 3.79 4.91 3.44 2.84 0.24 0.0 0.0 15.51 SW 0.30 3.55 4.79 3.02 1.93 0.20 0.02 0.0 13.81 WSW 0.24 1.68 1.19 0.66 0.69 0.16 0.02 0.0 4.64 W 0.21 0.78 0.47 0.35 0.44 0.06 0.01 0.0 2.32 WNW 0.27 0.70 0.36 0.34 0.51 0.03 0.0 0.0 2.21 NW 0.18 0.93 0.63 0.74 0.83 0.07 0.0 0.0 3.38 NNW 0.27 1.55 1.23 0.93 0.99 0.04 0.0 0.0 5.01 SUBTOTAL 6.04 36.24 26.65 15.58 13.76 1.21 0.07 0.0 99.55 TOTAL HOURS OF VALID WIND OBSERVATIONS 32338 TOTAL HOURS OF OBSERVATIONS 35064 RECOVERABILITY PERCENTAGE 92.2 TOTAL HOURS CALM 140 = 0.43 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 4.6 MPH T232-1to19.doc

SQN TABLE 2.3.2-2 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT JANUARY (72-75) 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 0.61 2.27 1.29 0.68 1.21 0.0 0.0 0.0 6.06 NNE 1.59 5.04 5.04 2.46 2.20 0.04 0.0 0.0 16.37 NE 0.68 4.81 2.77 0.95 2.27 0.27 0.0 0.0 11.75 ENE 0.34 1.25 0.30 0.11 0.0 0.0 0.0 0.0 2.00 E 0.45 0.87 0.15 0.27 0.04 0.0 0.0 0.0 1.78 ESE 0.38 0.49 0.0 0.0 0.0 0.0 0.0 0.0 0.87 SE 0.27 0.38 0.0 0.0 0.0 0.0 0.0 0.0 0.65 SSE 0.42 0.64 0.27 0.04 0.19 0.11 0.23 0.0 1.90 S 0.27 1.89 1.17 0.98 1.74 0.11 0.0 0.0 6.16 SSW 0.30 3.07 4.02 3.67 5.15 0.42 0.0 0.0 16.63 SW 0.30 3.45 5.49 3.45 2.65 0.68 0.0 0.0 16.02 WSW 0.30 2.01 1.55 0.87 1.29 0.42 0.0 0.0 6.44 W 0.15 0.83 0.42 0.45 0.42 0.0 0.0 0.0 2.27 WNW 0.11 0.42 0.30 0.08 0.38 0.04 0.0 0.0 1.33 NW 0.30 0.45 0.61 0.49 0.53 0.0 0.0 0.0 2.38 NNW 0.49 1.10 1.06 1.25 2.39 0.04 0.0 0.0 6.33 SUBTOTAL 6.96 28.97 24.44 15.75 20.46 2.13 0.23 0.0 98.94 TOTAL HOURS OF VALID WIND OBSERVATIONS 2640 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 88.7 TOTAL HOURS CALM 28 = 1.1 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 5.2 MPH T232-1to19.doc

SQN TABLE 2.3.2-3 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT FEBRUARY (72-75) 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 0.20 2.19 1.75 1.04 0.92 0.04 0.0 0.0 6.14 NNE 0.68 5.77 4.22 1.99 3.07 0.44 0.0 0.0 16.17 NE 0.48 4.62 2.91 0.96 1.15 0.36 0.0 0.0 10.48 ENE 0.48 2.35 0.52 0.28 0.04 0.08 0.0 0.0 3.75 E 0.56 0.80 0.20 0.12 0.16 0.08 0.0 0.0 1.92 ESE 0.28 0.56 0.12 0.12 0.12 0.28 0.0 0.0 1.48 SE 0.24 0.44 0.16 0.12 0.28 0.04 0.0 0.0 1.28 SSE 0.32 0.60 0.36 0.20 0.56 0.12 0.04 0.0 2.20 S 0.32 1.71 1.63 0.80 0.92 0.08 0.0 0.0 5.46 SSW 0.16 2.79 4.10 2.67 3.42 0.24 0.0 0.0 13.38 SW 0.28 3.07 4.54 3.82 2.99 0.56 0.0 0.0 15.26 WSW 0.20 1.83 1.55 1.12 0.60 0.12 0.0 0.0 5.42 W 0.12 0.60 0.44 0.64 0.76 0.04 0.0 0.0 2.60 WNW 0.28 0.44 0.52 0.76 1.27 0.04 0.0 0.0 3.31 NW 0.04 0.64 0.72 1.67 1.83 0.16 0.04 0.0 5.10 NNW 0.0 1.00 1.51 1.43 1.59 0.16 0.04 0.0 5.73 SUBTOTAL 4.64 29.41 25.25 17.74 19.68 2.84 0.12 0.0 99.68 TOTAL HOURS OF VALID WIND OBSERVATIONS 2511 TOTAL HOURS OF OBSERVATIONS 2712 RECOVERABILITY PERCENTAGE 92.6 TOTAL HOURS CALM 10 = 0.40 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 5.3 MPH T232-1to19.doc

SQN TABLE 2.3.2-4 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT MARCH (72-75) 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 2.09 1.70 0.85 0.57 0.0 0.0 0.0 5.39 NNE 0.39 5.87 4.85 1.95 2.94 0.14 0.0 0.0 16.14 NE 0.25 3.64 2.76 0.99 0.32 0.04 0.0 0.0 8.00 ENE 0.18 2.05 0.50 0.07 0.0 0.0 0.0 0.0 2.80 E 0.28 0.67 0.11 0.0 0.0 0.0 0.0 0.0 1.06 ESE 0.14 0.28 0.14 0.04 0.0 0.0 0.0 0.0 0.60 SE 0.18 0.32 0.18 0.0 0.0 0.0 0.0 0.0 0.68 SSE 0.25 0.67 0.46 0.42 0.67 0.07 0.04 0.0 2.54 S 0.42 1.45 1.27 1.49 3.89 0.42 0.0 0.0 8.94 SSW 0.21 2.58 3.93 3.61 5.80 0.88 0.0 0.0 17.01 SW 0.21 2.55 5.20 2.69 1.73 0.35 0.0 0.0 12.73 WSW 0.18 1.59 1.38 0.64 0.85 0.35 0.11 0.0 5.10 W 0.14 0.71 0.74 0.28 1.42 0.28 0.14 0.0 3.71 WNW 0.04 0.50 0.35 0.71 1.31 0.11 0.04 0.0 3.06 NW 0.04 0.88 0.64 1.45 2.16 0.21 0.0 0.0 5.38 NNW 0.21 1.13 1.95 1.63 1.70 0.18 0.0 0.0 6.80 SUBTOTAL 3.30 26.98 26.16 16.82 23.36 3.03 0.29 0.0 99.94 TOTAL HOURS OF VALID WIND OBSERVATIONS 2826 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 95.0 TOTAL HOURS CALM 2 = 0.07 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 5.7 MPH T232-1to19.doc

SQN TABLE 2.3.2-5 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT APRIL (72-75) 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 1.34 0.81 0.81 1.00 0.0 0.0 0.0 4.00 NNE 0.19 4.99 3.30 2.19 1.69 0.08 0.0 0.0 12.44 NE 0.12 4.41 2.49 1.69 2.26 0.04 0.0 0.0 11.01 ENE 0.19 1.53 0.19 0.12 0.0 0.0 0.0 0.0 2.03 E 0.15 0.73 0.12 0.0 0.0 0.0 0.0 0.0 1.00 ESE 0.23 0.12 0.12 0.0 0.0 0.0 0.0 0.0 0.47 SE 0.08 0.46 0.23 0.0 0.0 0.0 0.0 0.0 0.77 SSE 0.35 1.04 0.27 0.58 1.53 0.23 0.0 0.0 4.00 S 0.46 1.50 1.38 2.46 3.03 0.46 0.0 0.0 9.29 SSW 0.27 2.95 4.22 3.38 5.45 0.07 0.0 0.0 17.34 SW 0.15 2.23 4.87 3.68 5.87 0.46 0.15 0.0 17.41 WSW 0.04 1.61 1.34 0.92 1.65 0.73 0.12 0.0 6.41 W 0.04 0.31 0.42 0.61 0.69 0.31 0.0 0.0 2.38 WNW 0.08 0.54 0.73 0.50 1.27 0.12 0.0 0.0 3.24 NW 0.12 0.46 0.73 0.96 1.42 0.23 0.0 0.0 3.92 NNW 0.0 0.54 0.77 1.11 1.73 0.08 0.0 0.0 4.23 SUBTOTAL 2.51 24.76 21.99 19.01 27.59 3.81 0.27 0.0 99.94 TOTAL HOURS OF VALID WIND OBSERVATIONS 2606 TOTAL HOURS OF OBSERVATIONS 2880 RECOVERABILITY PERCENTAGE 90.5 TOTAL HOURS CALM 3 = 0.12 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 6.0 MPH T232-1to19.doc

SQN TABLE 2.3.2-6 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT MAY (72-75) 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 0.45 3.18 1.89 0.63 0.24 0.0 0.0 0.0 6.39 NNE 0.77 8.00 4.75 2.58 1.19 0.08 0.0 0.0 17.29 NE 0.52 3.35 2.79 1.29 0.56 0.04 0.0 0.0 8.51 ENE 0.31 1.75 0.66 0.03 0.0 0.0 0.0 0.0 2.75 E 0.49 1.36 0.21 0.0 0.0 0.0 0.0 0.0 2.06 ESE 0.52 0.52 0.07 0.0 0.0 0.0 0.0 0.0 1.11 SE 0.36 1.12 0.24 0.0 0.0 0.0 0.0 0.0 1.74 SSE 0.52 2.10 0.66 0.14 0.14 0.03 0.0 0.0 3.59 S 0.42 3.25 3.35 2.34 2.03 0.21 0.0 0.0 11.60 SSW 0.31 4.83 6.53 3.39 2.58 0.10 0.0 0.0 17.80 SW 0.10 4.40 4.02 2.27 1.22 0.10 0.03 0.0 12.14 WSW 0.17 1.50 1.12 0.49 0.42 0.03 0.0 0.0 3.73 W 0.31 0.66 0.45 0.21 0.07 0.0 0.0 0.0 1.70 WNW 0.31 0.63 0.24 0.21 0.14 0.0 0.0 0.0 1.53 NW 0.24 0.98 0.73 0.49 0.77 0.03 0.0 0.0 3.24 NNW 0.14 1.47 1.05 0.52 0.94 0.03 0.0 0.0 4.15 SUBTOTAL 5.96 39.16 28.76 14.59 10.30 0.53 0.03 0.0 99.33 TOTAL HOURS OF VALID WIND OBSERVATIONS 2863 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 96.2 TOTAL HOURS CALM 16 = 0.56 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 4.3 MPH T232-1to19.doc

SQN TABLE 2.3.2-7 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT JUNE (72-75) 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 0.55 3.19 1.46 0.24 0.0 0.0 0.0 0.0 5.44 NNE 1.26 7.60 3.94 2.36 1.06 0.04 0.0 0.0 16.26 NE 0.43 2.28 1.69 0.24 0.0 0.0 0.0 0.0 4.64 ENE 0.63 1.85 0.63 0.31 0.0 0.0 0.0 0.0 3.42 E 0.55 0.47 0.12 0.0 0.0 0.0 0.0 0.0 1.14 ESE 0.43 0.59 0.04 0.0 0.0 0.0 0.0 0.0 1.06 SE 0.39 1.38 0.12 0.0 0.0 0.0 0.0 0.0 1.89 SSE 0.43 1.46 1.14 0.16 0.16 0.0 0.0 0.0 3.35 S 0.71 4.05 3.78 2.44 1.18 0.04 0.0 0.0 12.20 SSW 0.35 5.75 6.26 4.76 1.42 0.04 0.0 0.0 18.58 SW 0.47 4.92 5.94 3.11 1.14 0.0 0.0 0.04 15.62 WSW 0.35 1.57 1.06 0.67 0.51 0.0 0.0 0.0 4.16 W 0.43 1.02 0.43 0.39 0.39 0.0 0.0 0.0 2.66 WNW 0.47 0.83 0.24 0.24 0.16 0.0 0.0 0.0 1.94 NW 0.08 0.67 0.83 0.67 1.02 0.0 0.0 0.0 3.27 NNW 0.39 1.34 1.26 0.51 0.31 0.0 0.0 0.0 3.81 SUBTOTAL 7.92 38.97 28.94 16.10 7.35 0.12 0.0 0.04 99.44 TOTAL HOURS OF VALID WIND OBSERVATIONS 2541 TOTAL HOURS OF OBSERVATIONS 2880 RECOVERABILITY PERCENTAGE 88.2 TOTAL HOURS CALM 14 = 0.55 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 4.0 MPH T232-1to19.doc

SQN TABLE 2.3.2-8 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT JULY (72-75) 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 0.25 4.46 1.55 0.18 0.07 0.0 0.0 0.0 6.51 NNE 0.68 9.72 4.50 1.76 0.50 0.0 0.0 0.0 17.16 NE 0.18 1.62 1.98 0.68 0.0 0.0 0.0 0.0 4.46 ENE 0.25 1.44 0.43 0.07 0.0 0.0 0.0 0.0 2.19 E 0.47 0.79 0.0 0.0 0.0 0.0 0.0 0.0 1.26 ESE 0.22 0.68 0.07 0.0 0.0 0.0 0.0 0.0 0.97 SE 0.43 1.73 0.47 0.0 0.0 0.0 0.0 0.0 2.63 SSE 0.40 2.20 0.90 0.25 0.11 0.0 0.0 0.0 3.86 S 0.79 5.11 3.92 0.97 0.40 0.0 0.0 0.0 11.19 SSW 0.40 5.94 8.32 4.43 0.86 0.0 0.0 0.0 19.95 SW 0.29 4.86 5.83 3.38 1.12 0.0 0.0 0.0 15.48 WSW 0.40 1.94 0.90 0.29 0.04 0.0 0.0 0.0 3.57 W 0.25 1.26 0.32 0.18 0.0 0.0 0.0 0.0 2.01 WNW 0.32 1.26 0.43 0.25 0.07 0.0 0.0 0.0 2.33 NW 0.25 1.98 0.65 0.22 0.0 0.0 0.0 0.0 3.10 NNW 0.22 2.38 0.54 0.18 0.0 0.0 0.0 0.0 3.32 SUBTOTAL 5.80 47.37 30.81 12.84 3.17 0.0 0.0 0.04 99.99 TOTAL HOURS OF VALID WIND OBSERVATIONS 2778 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 93.3 TOTAL HOURS CALM 0 = 0.00 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 3.7 MPH T232-1to19.doc

SQN TABLE 2.3.2-9 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT AUGUST (72-75) 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 0.45 5.35 1.40 0.35 0.03 0.0 0.0 0.0 7.58 NNE 1.08 12.81 5.39 2.27 0.59 0.0 0.0 0.0 22.14 NE 0.42 2.97 2.27 0.21 0.17 0.0 0.0 0.0 6.04 ENE 0.59 1.47 0.35 0.03 0.0 0.0 0.0 0.0 2.44 E 0.56 0.77 0.07 0.0 0.0 0.0 0.0 0.0 1.40 ESE 0.35 0.38 0.0 0.0 0.0 0.0 0.0 0.0 0.73 SE 0.21 1.33 0.14 0.0 0.0 0.0 0.0 0.0 1.68 SSE 0.35 1.92 0.84 0.10 0.14 0.0 0.0 0.0 3.35 S 0.42 3.92 4.02 2.52 0.45 0.0 0.0 0.0 11.33 SSW 0.17 4.83 6.33 3.95 0.94 0.0 0.0 0.0 16.22 SW 0.42 4.58 3.81 3.29 0.87 0.0 0.0 0.0 12.97 WSW 0.31 2.03 1.01 0.21 0.14 0.0 0.0 0.0 3.70 W 0.31 0.87 0.24 0.10 0.0 0.0 0.0 0.0 1.52 WNW 0.56 0.98 0.21 0.0 0.0 0.0 0.0 0.0 1.75 NW 0.28 1.22 0.35 0.35 0.03 0.0 0.0 0.0 2.23 NNW 0.38 2.62 1.29 0.42 0.03 0.0 0.0 0.0 4.74 SUBTOTAL 6.86 48.05 27.72 13.80 3.39 0.0 0.0 0.0 99.82 TOTAL HOURS OF VALID WIND OBSERVATIONS 2858 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 96.0 TOTAL HOURS CALM 1 = 0.03 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 3.6 MPH T232-1to19.doc

SQN TABLE 2.3.2-10 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT SEPT. (72-75) 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 0.99 5.27 1.99 0.77 0.52 0.0 0.0 0.0 9.54 NNE 0.92 12.04 6.15 2.98 3.98 0.07 0.04 0.0 26.18 NE 0.52 3.50 2.25 0.70 0.33 0.04 0.0 0.0 7.34 ENE 0.44 1.10 0.33 0.0 0.0 0.0 0.0 0.0 1.87 E 0.85 0.85 0.15 0.04 0.0 0.0 0.0 0.0 1.89 ESE 0.44 0.44 0.11 0.04 0.0 0.0 0.0 0.0 1.03 SE 0.70 1.25 0.33 0.0 0.0 0.0 0.0 0.0 2.28 SSE 0.48 1.77 0.63 0.04 0.07 0.0 0.0 0.0 2.99 S 0.63 3.83 3.53 1.66 1.07 0.0 0.0 0.0 10.72 SSW 0.29 3.35 4.71 2.84 0.74 0.0 0.0 0.0 11.93 SW 0.33 2.69 4.31 1.91 0.66 0.0 0.0 0.0 9.90 WSW 0.44 1.55 0.63 0.22 0.0 0.0 0.0 0.0 2.84 W 0.29 0.81 0.29 0.0 0.04 0.0 0.0 0.0 1.43 WNW 0.63 0.88 0.18 0.07 0.04 0.0 0.0 0.0 1.80 NW 0.33 1.33 0.22 0.26 0.11 0.0 0.0 0.0 2.25 NNW 0.37 2.25 1.88 0.74 0.37 0.0 0.0 0.0 5.61 SUBTOTAL 8.65 42.91 27.69 12.27 7.93 0.11 0.04 0.0 99.60 TOTAL HOURS OF VALID WIND OBSERVATIONS 2716 TOTAL HOURS OF OBSERVATIONS 2880 RECOVERABILITY PERCENTAGE 94.3 TOTAL HOURS CALM 12 = 0.44 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 3.9 MPH T232-1to19.doc

SQN TABLE 2.3.2-11 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT OCTOBER (72-75) 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 1.69 4.31 2.06 0.71 0.45 0.0 0.0 0.0 9.22 NNE 1.20 11.55 6.90 3.30 3.83 0.26 0.0 0.0 27.04 NE 1.01 5.63 2.81 1.05 0.34 0.0 0.0 0.0 10.84 ENE 0.75 1.91 0.15 0.0 0.0 0.0 0.0 0.0 2.81 E 0.71 0.98 0.04 0.0 0.0 0.0 0.0 0.0 1.73 ESE 0.49 0.45 0.0 0.0 0.0 0.0 0.0 0.0 0.94 SE 0.79 0.53 0.08 0.0 0.0 0.0 0.0 0.0 1.40 SSE 0.86 1.28 0.34 0.30 0.15 0.0 0.0 0.0 2.93 S 0.34 3.49 2.10 0.75 0.34 0.0 0.0 0.0 7.02 SSW 0.41 3.86 2.63 1.50 0.56 0.0 0.0 0.0 8.96 SW 0.41 3.75 4.09 2.21 0.60 0.0 0.0 0.0 11.06 WSW 0.23 1.95 1.28 0.83 0.49 0.0 0.0 0.0 4.78 W 0.19 1.13 0.60 0.41 0.15 0.0 0.0 0.0 2.48 WNW 0.34 0.60 0.23 0.34 0.04 0.0 0.0 0.0 1.55 NW 0.23 0.49 0.56 0.56 0.11 0.0 0.0 0.0 1.95 NNW 0.56 1.58 0.90 0.71 0.30 0.0 0.0 0.0 4.05 SUBTOTAL 10.21 43.49 24.77 12.67 7.36 0.26 0.0 0.0 98.76 TOTAL HOURS OF VALID WIND OBSERVATIONS 2666 TOTAL HOURS OF OBSERVATIONS 2976 RECOVERABILITY PERCENTAGE 89.6 TOTAL HOURS CALM 34 = 1.28 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 3.9 MPH T232-1to19.doc

SQN TABLE 2.3.2-12 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT NOVEMBER (72-75) 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 0.48 2.85 2.15 0.85 0.37 0.0 0.0 0.0 6.70 NNE 0.70 8.66 6.77 3.18 2.81 0.22 0.0 0.0 22.34 NE 0.55 5.11 3.44 1.44 1.41 0.07 0.0 0.0 12.02 ENE 0.44 1.07 0.48 0.04 0.0 0.0 0.0 0.0 2.03 E 0.55 0.78 0.18 0.0 0.0 0.0 0.0 0.0 1.51 ESE 0.33 0.26 0.18 0.0 0.0 0.0 0.0 0.0 0.77 SE 0.22 0.26 0.18 0.0 0.0 0.0 0.0 0.0 0.66 SSE 0.30 0.92 0.37 0.18 0.41 0.15 0.0 0.0 2.33 S 0.37 1.92 1.70 1.70 1.78 0.22 0.0 0.0 7.69 SSW 0.33 2.07 3.29 3.74 3.70 0.07 0.0 0.0 13.20 SW 0.37 2.48 4.29 2.85 2.00 0.07 0.0 0.0 12.06 WSW 0.11 1.15 1.48 0.78 0.92 0.07 0.0 0.0 4.51 W 0.11 0.33 0.67 0.48 0.67 0.0 0.0 0.0 2.26 WNW 0.04 0.44 0.26 0.26 0.92 0.04 0.0 0.0 1.96 NW 0.07 0.81 1.04 0.92 1.04 0.15 0.0 0.0 4.03 NNW 0.26 1.52 1.29 1.18 0.96 0.0 0.0 0.0 5.21 SUBTOTAL 5.23 30.63 27.77 17.60 16.99 1.06 0.0 0.0 99.28 TOTAL HOURS OF VALID WIND OBSERVATIONS 2703 TOTAL HOURS OF OBSERVATIONS 2880 RECOVERABILITY PERCENTAGE 93.9 TOTAL HOURS CALM 18 = 0.67 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 4.9 MPH T232-1to19.doc

SQN TABLE 2.3.2-13 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY DIRECTION DISREGARDING STABILITY CLASS SEQUOYAH NUCLEAR PLANT DECEMBER (72-75) 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 0.23 1.56 1.44 1.03 1.63 0.0 0.0 0.0 5.89 NNE 0.42 7.00 4.64 2.47 2.47 0.04 0.0 0.0 17.04 NE 0.57 4.56 2.89 2.02 1.25 0.0 0.0 0.0 11.29 ENE 0.42 1.25 0.11 0.08 0.0 0.0 0.0 0.0 1.86 E 0.34 0.49 0.0 0.0 0.0 0.0 0.0 0.0 0.83 ESE 0.15 0.57 0.04 0.0 0.0 0.0 0.0 0.0 0.76 SE 0.23 0.57 0.11 0.0 0.0 0.0 0.0 0.0 0.91 SSE 0.27 0.60 0.30 0.30 0.19 0.0 0.0 0.0 2.66 S 0.49 2.43 1.83 0.80 1.52 0.11 0.0 0.0 7.18 SSW 0.30 3.23 4.30 3.27 3.54 0.11 0.0 0.0 14.75 SW 0.27 3.57 5.21 3.73 2.62 0.27 0.0 0.0 15.67 WSW 0.08 1.41 1.03 0.99 1.52 0.27 0.0 0.0 5.30 W 0.11 0.76 0.57 0.46 0.72 0.11 0.0 0.0 2.73 WNW 0.04 0.87 0.68 0.68 0.61 0.0 0.0 0.0 2.88 NW 0.15 1.10 0.57 0.91 0.99 0.08 0.0 0.04 3.84 NNW 0.23 1.52 1.29 1.56 1.67 0.04 0.0 0.0 6.31 SUBTOTAL 4.30 32.49 25.01 18.30 18.73 1.03 0.0 0.04 99.90 TOTAL HOURS OF VALID WIND OBSERVATIONS 2630 TOTAL HOURS OF OBSERVATIONS 2952 RECOVERABILITY PERCENTAGE 89.1 TOTAL HOURS CALM 2 = 0.08 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 5.1 MPH T232-1to19.doc

SQN TABLE 2.3.2-14 (Sheet 1) WIND DIRECTION PERSISTENCE DATA DISREGARDING STABILITY SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 LOST RECORD(%)= 7.77 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 2 190 277 205 82 39 18 38 86 253 333 360 123 62 58 94 138 14 2370 5804 100.00 3 99 163 106 23 10 10 9 33 107 187 179 45 21 26 38 54 9 1119 3434 59.17 4 47 135 66 11 3 0 5 11 80 120 128 33 17 10 20 25 1 712 2315 39.89 5 20 89 33 6 2 1 3 3 43 77 87 21 8 10 17 22 2 444 1603 27.62 6 10 65 27 3 1 0 0 0 29 57 53 11 3 1 9 15 1 285 1159 19.97 7 13 45 14 1 1 0 0 5 20 51 43 6 1 3 7 14 0 224 874 15.06 8 9 40 18 0 0 0 0 4 8 29 18 3 4 1 5 10 0 149 650 11.20 9 6 36 10 1 0 0 0 1 8 25 15 3 1 1 2 8 0 117 501 8.63 10 3 32 8 0 0 0 0 0 6 16 10 0 0 0 3 3 0 81 384 6.62 11 0 29 7 1 0 0 0 0 4 10 5 1 1 0 3 2 0 63 303 5.22 12 0 17 8 1 0 0 0 0 5 12 5 2 0 0 2 2 0 54 240 4.14 13 3 16 1 0 0 0 0 0 2 11 6 0 0 0 0 0 0 39 186 3.20 14 0 15 3 0 0 0 0 0 3 6 7 0 0 0 0 1 0 35 147 2.53 15 0 9 2 0 0 0 0 0 1 4 3 0 1 0 0 1 0 20 112 1.93 16 0 6 3 0 0 0 0 0 0 3 4 0 0 0 1 1 0 18 92 1.59 17 0 11 3 0 0 0 0 0 1 2 1 0 0 0 2 0 0 20 74 1.27 18 0 8 0 0 0 0 0 0 0 3 1 0 0 1 0 0 0 13 54 0.93 19 0 5 1 0 0 0 0 0 1 1 1 0 0 0 0 1 0 10 41 0.71 20 0 3 1 0 0 0 0 0 0 3 0 1 0 0 0 0 0 8 31 0.53 21 0 2 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 4 23 0.40 22 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 19 0.33 23 0 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 19 0.33 24 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 17 0.29 25 0 2 1 0 0 0 0 0 0 0 1 0 0 0 0 0 0 4 16 0.28 26 0 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 12 0.21 27 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 10 0.17 28 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 10 0.17 29 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 9 0.16 30 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 7 0.12 31 0 2 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 6 0.10 32 0 2 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 3 4 0.07

>32           0      1    0   0  0  0    0      0       0        0       0        0  0   0   0   0   0     1     1    0.02 TOTAL        401    1015 519 129 56 29  55     143     572      951     928     249 119 111 203 297 27   5804 T232-1to19.doc

SQN TABLE 2.3.2-14 (Sheet 2) (Continued) WIND DIRECTION PERSISTENCE DATA DISREGARDING STABILITY SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 LOST RECORD(%)= 7.77 PERSISTENCE WIND DIRECTION (HOURS) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW CALM MAXIMUM PERSISTENCE 29 33 26 12 7 5 5 9 12 21 32 20 15 18 17 19 6 (HOURS) 50.0% 3 4 3 2 2 2 2 2 3 3 3 3 2 2 3 3 2 80.0% 4 8 6 3 3 3 3 3 5 6 5 4 4 4 5 5 3 90.0% 6 12 8 5 4 3 4 4 7 9 7 6 5 5 7 7 5 99.0% 10 25 17 11 7 5 5 8 14 17 15 12 11 9 16 15 6 99.9% 29 32 26 12 7 5 5 9 21 21 32 20 15 18 17 17 6 METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant NOTE: Persistent wind is defined in this analysis as WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL a wind blowing continuously from one of the named 22-1/2o sectors (i.e., north-northwest) except that it is not considered to be interrupted if it departs from that sector for one hour and then returns, or if there are up to two hours of missing data followed by a continued directional persistence. T232-1to19.doc

SQN Table 2.3.2-15 TEMPERATURE* Sequoyah Nuclear Plant April 2, 1971-March 31, 1972 Avg. Temp. Avg. Max. Temp. Avg. Min. Temp. Extreme Max Extreme Min.

                                     °           °                   ° Month                                  F           F                   F                     Temp. °F   Temp. °F Dec.                                 49.0      56.2                    42.3                   72.0          23.3 Jan.                                 42.7      52.2                    33.5                   71.3           2.9 Feb.                                 40.1      49.7                    30.8                   74.8          15.2 Winter                         43.9      52.7                    35.5                   74.8           2.9 Mar.                                 48.7      59.3                    38.6                   75.8          26.4 Apr.                                 59.2      72.8                    45.9                   86.0          33.1 May                                  64.6      75.8                    54.2                   84.9          38.2 Spring                         57.5      69.3                    46.2                   86.0          26.4 June                                 75.4      86.7                    66.6                   96.3          55.3 July                                 75.4      83.4                    68.7                   90.8          61.8 August                               75.5      86.1                    68.0                   91.4          59.7 Summer                         75.4      85.4                    67.7                   96.3          55.3 Sept.                                72.4      82.8                    63.6                   95.1          53.4 Oct.                                 64.7      74.9                    57.3                   87.0          43.1 Nov.                                 48.8      58.8                    41.0                   78.0          29.2 Fall                           61.9      72.1                    53.9                   95.1          29.2 Annual                         59.7      69.8                    50.8                   96.3           2.9

Temperature instrument 4 feet above ground.

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SQN-18 Table 2.3.2-16 TEMPERATUREa,d (Chattanooga, Tennessee) Avg. Temp.b Avg. Max. Temp.b Avg. Min. Temp.b Extreme Max.c Extreme Min.c Month °F °F °F Temp. °F Temp. °F Dec. 41.2 50.9 31.4 78 -2 Jan. 40.2 49.9 30.5 78 -10 Feb. 42.9 53.4 32.3 79 1 Winter 41.4 51.4 -- -- -- Mar. 49.8 61.2 38.4 87 8 Apr. 60.5 72.9 48.1 93 25 May. 68.5 81.0 56.0 99 34 Spring 59.6 71.7 -- -- -- June 76.0 87.5 64.5 104 41 July 78.8 89.5 68.1 106 51 Aug. 78.0 89.0 67.0 105 50 Summer 77.6 88.7 -- -- -- Sept. 71.9 83.4 60.4 102 36 Oct. 60.8 73.5 48.1 94 22 Nov. 48.9 60.7 37.1 84 4 Fall 60.5 72.5 -- -- -- Annual 59.8 71.1 48.5 106 -10 a. Local Climatological Data, "Annual Summary with Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Center, Asheville, N.C., 1979. b. Based on record for 1941-1970. c. Period of record 63 years, through 2002. d. Local Climatological Data, Annual Summary With Comparative Data, Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Data Center, Asheville, M.C., 2002. T232-1to19.doc

SQN Table 2.3.2-17 ABSOLUTE HUMIDITY* Sequoyah Nuclear Plant April 2, 1971-March 31, 1972 Avg. A. H. Avg. Max. A. H. Avg. Min. A. H. Extreme Max. A. H. Extreme Min. A. H. Month g/m3 g/m3 g/m3 g/m3 g/m3 Dec. 7.6 9.3 6.0 15.8 1.2 Jan. 5.4 7.1 3.8 15.4 1.1 Feb. 4.2 5.2 2.7 12.2 1.0 Winter 5.7 7.2 4.2 15.8 1.0 Mar. 5.9 8.0 4.3 12.7 1.5 Apr. 6.3 7.8 5.0 12.2 2.7 May 9.6 11.7 7.8 17.3 3.3 Spring 7.3 9.2 5.7 17.3 1.5 June 16.2 18.7 14.2 22.3 9.9 July 14.1 15.8 12.6 18.5 10.0 Aug. 13.9 15.9 12.2 19.6 8.7 Summer 14.7 16.8 13.0 22.3 8.7 Sept. 14.6 17.2 12.0 21.8 8.0 Oct. 12.4 14.7 10.3 19.6 5.6 Nov. 6.4 8.4 5.2 18.2 2.1 Fall 11.1 13.4 9.2 21.8 2.1 Annual 9.7 11.7 8.0 22.3 1.0

Computed from dry bulb and dew point temperature measurements 4 feet above ground.

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SQN Table 2.3.2-18 RELATIVE HUMIDITY* Sequoyah Nuclear Plant April 2, 1971-March 31, 1972 Avg. R. H. Avg. Max. R. H. Avg. Min. R. H. Extreme Max. R. H. Extreme Min. R. H. Month (percent) (percent) (percent) (percent) (percent) Dec. 78.4 89.6 62.6 100.0 34.8 Jan. 65.0 79.9 50.1 93.9 22.5 Feb. 59.8 74.2 43.5 95.3 22.1 Winter 67.7 81.2 52.1 100.0 22.1 Mar. 63.8 83.4 43.4 100.0 21.9 Apr. 50.6 75.8 26.8 86.6 17.0 May 62.2 82.5 40.9 95.1 18.4 Spring 58.9 80.5 37.0 100.0 17.0 June 74.4 90.1 51.3 100.0 34.5 July 64.3 73.7 51.6 78.8 37.2 Aug. 63.3 72.7 47.2 85.3 33.8 Summer 67.3 78.8 50.0 100.0 33.8 Sept. 73.1 84.0 53.2 100.0 32.1 Oct. 78.4 89.0 61.7 99.3 37.8 Nov. 65.3 79.6 50.4 100.0 28.0 Fall 72.2 84.2 55.1 100.0 28.0 Annual 66.5 81.2 48.6 100.0 17.0

Computed from dry bulb and dew point temperature measurements 4 feet above ground.

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SQN Table 2.3.2-19 PRECIPITATION* (Friendship School, Tennessee) 1948-1967 Days with Monthly Extreme Extreme Max. In 0.01 Inch Average Monthly Max. Monthly Min. 24 Hrs. Month or More (inches) (inches) (inches) (inches) Dec. 10 5.40 12.15 0.82 3.02 Jan. 12 5.99 13.61 2.35 3.88 Feb. 11 5.82 11.41 2.43 3.08 Winter 33 17.21 Mar. 12 6.76 15.22 2.60 6.08 Apr. 10 4.70 10.88 1.18 2.62 May 9 3.87 7.53 1.41 2.75 Spring 31 15.33 June 9 4.16 7.20 0.59 2.60 July 11 5.34 11.31 0.74 2.98 Aug. 10 3.91 8.01 1.90 7.56 Summer 30 13.41 Sept. 7 4.02 15.40 0.83 4.27 Oct. 7 2.86 9.63 0.09 2.24 Nov. 9 4.86 16.58 0.95 3.21 Fall 23 11.74 Annual 117 57.69

TVA Raingage Station 685, Friendship School, Tennessee, located about 2-1/2 miles north-northeast of Sequoyah Landing site; period of record 20 years since station activation April 30, 1948.

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SQN-18 Table 2.3.2-20 SNOWFALLa,b (Chattanooga, Tennessee) Month Mean Total Maximum Total Maximum Total in 24 Hours Jan. 1.8 10.2 10.2 Feb. 1.2 10.4 8.7 Mar. 0.7 20.0 20.0 Apr. 0.1 2.8 2.8 May T T T June T T T July 0 0 0 Aug. 0 0 0 Sept. 0 0 0 Oct. T T T Nov. 0.1 2.8 2.8 Dec. 0.6 9.1 8.9 Annual 4.4 a. Local Climatological Data, "Annual Summary With Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Data Center, Asheville, N.C., 2002. b. Period of record, 1931-1996. T232-20to22.doc

SQN Table 2.3.2-21 HEAVY FOG (Chattanooga, Tennessee) Mean No. of Days Month With Heavy Fogc Dec. 3 Jan. 3 Feb. 2 Winter 8 Mar. 2 Apr. 2 May 2 Spring 6 June 2 July 2 Aug. 3 Summer 7 Sept. 4 Oct. 6 Nov. 4 Fall 14 Annual 36 a. Local Climatological Data, "Annual Summary With Comparative Data," Chattanooga, Tennessee, U.S. Department of Commerce, NOAA, National Climatic Center, Asheville, N.C., 1979. b. Heavy fog is defined as fog reducing the visibility to 1/4 mile or less. c. Period of record 49 years, through 1979. Rounding to whole days results in one-day difference between the sum of the monthly averages and the annual average. T232-20to22.doc

SQN Table 2.3.2-22 PERCENT OCCURRENCE OF ATMOSPHERIC STABILITY* Sequoyah Nuclear Plant January 1, 1972 - December 31, 1975 Pasquill Vertical Temperature Percent Stability Class Difference ( T)** Occurrence** A T -1.9°C/100 m 2.91 B -1.9 < T -1.7°C/100 m 1.24 C -1.7 < T -1.5°C/100 m 3.78 D -1.5 < T -0.5°C/100 m 19.91 E -0.5 < T 1.5°C/100 m 44.36 F 1.5 < T 4.0°C/100 m 20.79 G T > 4.0°C/100 m 6.93 Total 99.92

*Temperature instruments 9 and 46 meters above ground.

- Valid T = 91.33 percent of total hours in period; percent occurrences are percentages of valid T occurrences.

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SQN TABLE 2.3.2-23 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS A (DELTA T<=-1.9 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.03 0.04 0.04 0.0 0.0 0.0 0.13 NNE 0.0 0.04 0.19 0.20 0.16 0.01 0.0 0.0 0.60 NE 0.0 0.08 0.20 0.15 0.13 0.0 0.0 0.0 0.56 ENE 0.0 0.03 0.03 0.01 0.0 0.0 0.0 0.0 0.07 E 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.0 0.01 ESE 0.0 0.01 0.01 0.0 0.0 0.01 0.0 0.0 0.03 SE 00 0.01 0.02 0.0 0.0 0.0 0.0 0.0 0.03 SSE 0.0 0.01 0.03 0.02 0.02 0.01 0.0 0.0 0.09 S 0.0 0.01 0.04 0.06 0.05 0.01 0.0 0.0 0.17 SSW 0.0 0.01 0.09 0.18 0.16 0.01 0.0 0.0 0.45 SW 0.0 0.04 0.12 0.10 0.09 0.02 0.0 0.0 0.37 WSW 0.0 0.02 0.03 0.03 0.02 0.02 0.0 0.0 0.12 W 0.0 0.01 0.0 0.01 0.02 0.0 0.0 0.0 0.04 WNW 0.0 0.0 0.0 0.0 0.01 0.01 0.0 0.0 0.02 NW 0.0 0.01 0.01 0.01 0.05 0.01 0.0 0.0 0.09 NNW 0.0 0.01 0.0 0.02 0.08 0.01 0.0 0.0 0.12 SUBTOTAL 0.01 0.31 0.80 0.83 0.83 0.12 0.0 0.0 2.90 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS A 958 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS A 934 TOTAL HOURS CALM 4 = 0.01 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 6.5 MPH T232-23to29.doc

SQN TABLE 2.3.2-24 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS B (-1.9< DELTA T<=-1.7 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.0 0.01 0.03 0.0 0.0 0.0 0.05 NNE 0.0 0.02 0.10 0.10 0.08 0.0 0.0 0.0 0.30 NE 0.0 0.03 0.12 0.04 0.02 0.0 0.0 0.0 0.21 ENE 0.0 0.01 0.02 0.0 0.0 0.0 0.0 0.0 0.03 E 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.0 0.01 ESE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SE 00 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 SSE 0.0 0.01 0.01 0.0 0.01 0.0 0.0 0.0 0.03 S 0.0 0.03 0.01 0.03 0.03 0.0 0.0 0.0 0.10 SSW 0.0 0.01 0.03 0.07 0.09 0.0 0.0 0.0 0.20 SW 0.0 0.01 0.06 0.06 0.05 0.0 0.0 0.0 0.18 WSW 0.0 0.0 0.01 0.0 0.01 0.0 0.0 0.0 0.02 W 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 WNW 0.0 0.0 0.01 0.0 0.02 0.0 0.0 0.0 0.03 NW 0.0 0.0 0.0 0.0 0.03 0.0 0.0 0.0 0.03 NNW 0.0 0.0 0.0 0.01 0.02 0.0 0.0 0.0 0.03 SUBTOTAL 0.0 0.15 0.38 0.32 0.39 0.0 0.0 0.0 1.24 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS B 416 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS B 411 TOTAL HOURS CALM 1 < 0.01 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 6.4 MPH T232-23to29.doc

SQN TABLE 2.3.2-25 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS C (-1.7< DELTA T<=-1.5 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.03 0.03 0.02 0.0 0.0 0.0 0.09 NNE 0.0 0.08 0.25 0.21 0.22 0.0 0.0 0.0 0.76 NE 0.0 0.10 0.31 0.09 0.07 0.0 0.0 0.0 0.57 ENE 0.0 0.05 0.03 0.01 0.0 0.0 0.0 0.0 0.09 E 0.0 0.02 0.02 0.0 0.0 0.0 0.0 0.0 0.04 ESE 0.0 0.01 0.01 0.01 0.0 0.0 0.0 0.0 0.03 SE 00 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 SSE 0.0 0.02 0.04 0.0 0.03 0.0 0.0 0.0 0.09 S 0.0 0.04 0.07 0.09 0.07 0.02 0.0 0.0 0.29 SSW 0.0 0.04 0.16 0.27 0.24 0.04 0.0 0.0 0.75 SW 0.0 0.05 0.13 0.20 0.12 0.02 0.0 0.0 0.52 WSW 0.0 0.02 0.02 0.05 0.03 0.01 0.01 0.0 0.14 W 0.0 0.01 0.01 0.01 0.02 0.01 0.0 0.0 0.06 WNW 0.0 0.0 0.01 0.02 0.02 0.0 0.0 0.0 0.05 NW 0.01 0.0 0.0 0.02 0.05 0.01 0.0 0.0 0.09 NNW 0.0 0.01 0.04 0.03 0.09 0.01 0.0 0.0 0.18 SUBTOTAL 0.01 0.47 1.14 1.04 0.98 0.12 0.01 0.0 3.77 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS C 1237 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS C 1214 TOTAL HOURS CALM 2 = 0.01 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 6.3 MPH T232-23to29.doc

SQN TABLE 2.3.2-26 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS D (-1.5< DELTA T<=-0.5 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.24 0.22 0.16 0.17 0.0 0.0 0.0 0.80 NNE 0.06 0.73 1.03 0.84 0.78 0.07 0.0 0.0 3.51 NE 0.02 0.76 0.88 0.42 0.42 0.05 0.0 0.0 2.55 ENE 0.01 0.21 0.11 0.03 0.0 0.0 0.0 0.0 0.36 E 0.01 0.12 0.03 0.02 0.01 0.0 0.0 0.0 0.19 ESE 0.01 0.06 0.02 0.0 0.0 0.0 0.0 0.0 0.09 SE 0.0 0.12 0.08 0.0 0.0 0.0 0.0 0.0 0.20 SSE 0.0 0.15 0.15 0.05 0.06 0.01 0.01 0.0 0.43 S 0.01 0.31 0.53 0.38 0.25 0.02 0.0 0.0 1.50 SSW 0.01 0.44 1.25 0.95 0.70 0.07 0.0 0.0 3.42 SW 0.01 0.47 1.17 1.03 0.52 0.03 0.01 0.0 3.24 WSW 0.0 0.22 0.34 0.18 0.21 0.07 0.01 0.0 1.03 W 0.01 0.06 0.08 0.10 0.19 0.02 0.01 0.0 0.47 WNW 0.01 0.06 0.05 0.11 0.18 0.01 0.0 0.0 0.42 NW 0.0 0.08 0.08 0.22 0.31 0.03 0.0 0.0 0.72 NNW 0.01 0.15 0.14 0.25 0.36 0.02 0.0 0.0 0.93 SUBTOTAL 0.18 4.18 6.16 4.74 4.16 0.40 0.04 0.0 19.86 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS D 6567 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS D 6345 TOTAL HOURS CALM 16 = 0.05 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 5.8 MPH T232-23to29.doc

SQN TABLE 2.3.2-27 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS E (-0.5< DELTA T<=1.5 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.23 1.26 0.83 0.39 0.27 0.0 0.0 0.0 2.98 NNE 0.31 2.83 2.46 1.07 0.92 0.03 0.0 0.0 7.62 NE 0.15 1.03 0.71 0.31 0.18 0.01 0.0 0.0 2.39 ENE 0.12 0.48 0.16 0.04 0.0 0.0 0.0 0.0 0.80 E 0.14 0.24 0.05 0.01 0.01 0.0 0.0 0.0 0.45 ESE 0.09 0.11 0.01 0.01 0.01 0.01 0.0 0.0 0.24 SE 0.10 0.37 0.06 0.01 0.01 0.0 0.0 0.0 0.55 SSE 0.11 0.58 0.24 0.13 0.23 0.04 0.02 0.0 1.35 S 0.17 1.33 1.49 0.91 1.05 0.08 0.0 0.0 5.03 SSW 0.10 1.67 2.32 1.67 1.45 0.11 0.0 0.0 7.32 SW 0.17 1.59 2.07 1.30 0.99 0.10 0.0 0.0 6.22 WSW 0.13 0.87 0.55 0.35 0.40 0.06 0.0 0.0 2.36 W 0.10 0.42 0.28 0.21 0.22 0.03 0.0 0.0 1.26 WNW 0.14 0.37 0.22 0.19 0.27 0.02 0.0 0.0 1.21 NW 0.10 0.50 0.37 0.43 0.38 0.02 0.0 0.0 1.80 NNW 0.15 0.80 0.68 0.57 0.40 0.01 0.0 0.0 2.61 SUBTOTAL 2.31 14.45 12.50 7.60 6.79 0.52 0.02 0.0 44.19 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS E 14624 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS E 14146 TOTAL HOURS CALM 54 = 0.17 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 4.8 MPH T232-23to29.doc

SQN TABLE 2.3.2-28 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS F ( 1.5< DELTA T<=4.0 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.22 1.42 0.45 0.04 0.0 0.0 0.0 0.0 2.13 NNE 0.35 3.69 0.86 0.05 0.0 0.0 0.0 0.0 4.95 NE 0.22 1.19 0.29 0.01 0.0 0.0 0.0 0.0 1.71 ENE 0.16 0.41 0.03 0.0 0.0 0.0 0.0 0.0 0.60 E 0.22 0.23 0.0 0.0 0.0 0.0 0.0 0.0 0.45 ESE 0.13 0.19 0.02 0.0 0.0 0.0 0.0 0.0 0.34 SE 0.15 0.24 0.02 0.0 0.0 0.0 0.0 0.0 0.41 SSE 0.16 0.38 0.07 0.03 0.01 0.0 0.0 0.0 0.65 S 0.18 0.80 0.30 0.10 0.06 0.0 0.0 0.0 1.44 SSW 0.13 1.15 0.73 0.26 0.12 0.0 0.0 0.0 2.39 SW 0.10 1.03 0.87 0.29 0.13 0.0 0.0 0.0 2.42 WSW 0.09 0.47 0.20 0.04 0.01 0.0 0.0 0.0 0.81 W 0.07 0.20 0.07 0.01 0.0 0.0 0.0 0.0 0.35 WNW 0.10 0.24 0.07 0.01 0.0 0.0 0.0 0.0 0.42 NW 0.05 0.30 0.15 0.06 0.01 0.0 0.0 0.0 0.57 NNW 0.09 0.53 0.35 0.05 0.01 0.0 0.0 0.0 1.03 SUBTOTAL 2.42 12.47 4.48 0.95 0.35 0.0 0.0 0.0 20.67 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS F 6718 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS F 6637 TOTAL HOURS CALM 39 = 0.12 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 3.0 MPH T232-23to29.doc

SQN TABLE 2.3.2-29 JOINT PERCENTAGE FREQUENCIES OF WIND SPEED BY WIND DIRECTION FOR STABILITY CLASS G (DELTA T > 4.0 C/100 M) SEQUOYAH NUCLEAR PLANT JAN 1, 72 - DEC 31, 75 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 0.28 0.08 0.0 0.0 0.0 0.0 0.0 0.41 NNE 0.10 0.95 0.19 0.0 0.0 0.0 0.0 0.0 1.24 NE 0.08 0.70 0.11 0.0 0.0 0.0 0.0 0.0 0.89 ENE 0.13 0.40 0.02 0.0 0.0 0.0 0.0 0.0 0.55 E 0.12 0.17 0.01 0.0 0.0 0.0 0.0 0.0 0.30 ESE 0.10 0.07 0.0 0.0 0.0 0.0 0.0 0.0 0.17 SE 0.09 0.07 0.0 0.0 0.0 0.0 0.0 0.0 0.16 SSE 0.15 0.20 0.0 0.0 0.0 0.0 0.0 0.0 0.35 S 0.09 0.37 0.04 0.01 0.0 0.0 0.0 0.0 0.51 SSW 0.06 0.45 0.30 0.02 0.01 0.0 0.0 0.0 0.84 SW 0.03 0.40 0.40 0.04 0.0 0.0 0.0 0.0 0.87 WSW 0.01 0.10 0.06 0.0 0.0 0.0 0.0 0.0 0.17 W 0.03 0.08 0.02 0.0 0.0 0.0 0.0 0.0 0.13 WNW 0.01 0.03 0.01 0.0 0.01 0.0 0.0 0.0 0.06 NW 0.01 0.05 0.03 0.0 0.0 0.0 0.0 0.0 0.09 NNW 0.02 0.08 0.03 0.0 0.0 0.0 0.0 0.0 0.13 SUBTOTAL 1.08 4.40 1.30 0.07 0.02 0.0 0.0 0.0 6.87 TOTAL HOURS OF VALID STABILITY OBSERVATIONS 32723 TOTAL HOURS OF STABILITY CLASS G 2203 TOTAL HOURS OF VALID WIND DIRECTION-WIND SPEED-STABILITY CLASS G 2202 TOTAL HOURS CALM 18 = 0.06 percent ALL COLUMNS AND CALM TOTAL 100 PERCENT OF JOINT VALID OBSERVATIONS METEOROLOGICAL FACILITY located 1.2 km southwest of Sequoyah Nuclear Plant STABILITY BASED ON LAPSE RATE MEASURED BETWEEN 9.25 and 45.99 meters WIND SPEED AND DIRECTION MEASURED AT THE 9.73 METER LEVEL MEAN WIND SPEED = 2.5 MPH T232-23to29.doc

SQN Table 2.3.2-30 Sequoyah Nuclear Plant - Percent of Observations in Each Stability Class - Hourly-Average and End-of-Hour Temperature Differences (T) (May 1975-April 1976) 150' - 33' T 300' - 33' T Vs. 33' Wind Data Vs. 300' Wind Data Stability Class Hourly-Average End-of-Hour Hourly-Average End-of-Hour A 1.73 3.23 0.14 0.62 B 3.20 2.96 0.89 1.12 C 2.25 2.26 2.37 2.61 D 19.24 18.00 33.55 32.63 E 41.97 42.48 41.17 41.21 F 21.56 20.22 15.06 14.80 G 9.96 10.89 6.71 6.92 Joint Recovery Rate 97.4% 97.4% 97.1% 97.1% (Wind Direction, Wind Speed, and T) Number of Hours of 4979 4898 3808 3705 Inversion T Total Hours of 8620 8621 8589 8590 Valid T Percent Frequency of 57.8% 56.8% 44.3% 43.1% Hours of Inversion T (Inversion/Total x 100) T232-30.doc

SQN TABLE 2.3.2-31 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS A DELTA T<=-1.9 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 1975 - APRIL 30, 1976 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 0.01 0.08 0.06 0.0 0.0 0.0 0.15 NNE 0.0 0.02 0.14 0.27 0.23 0.0 0.0 0.0 0.66 NE 0.0 0.01 0.20 0.21 0.09 0.0 0.0 0.0 0.51 ENE 0.0 0.0 0.06 0.0 0.0 0.0 0.0 0.0 0.06 E 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 ESE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SE 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 SSE 0.0 0.01 0.02 0.01 0.02 0.0 0.0 0.0 0.06 S 0.0 0.0 0.0 0.0 0.04 0.0 0.0 0.0 0.04 SSW 0.0 0.0 0.0 0.01 0.05 0.02 0.0 0.0 0.08 SW 0.0 0.0 0.01 0.01 0.01 0.04 0.0 0.0 0.07 WSW 0.0 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.02 W 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.01 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 0.01 0.0 0.0 0.0 0.01 NNW 0.0 0.0 0.0 0.01 0.02 0.0 0.0 0.0 0.03 SUBTOTAL 0.0 0.05 0.47 0.62 0.53 0.06 0.0 0.0 1.73 CALM = 0.0 154 STABILITY CLASS A OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 151 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 154 STABILITY CLASS A OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-32 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS A DELTA T<=-1.9 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.09 0.05 0.0 0.0 0.0 0.18 NNE 0.0 0.07 0.26 0.19 0.28 0.01 0.0 0.0 0.81 NE 0.0 0.09 0.27 0.20 0.13 0.0 0.0 0.0 0.69 ENE 0.0 0.06 0.09 0.0 0.0 0.0 0.0 0.0 0.15 E 0.0 0.05 0.05 0.0 0.0 0.0 0.0 0.0 0.10 ESE 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 SE 0.0 0.02 0.04 0.0 0.0 0.0 0.0 0.0 0.06 SSE 0.0 0.02 0.04 0.0 0.02 0.0 0.0 0.0 0.08 S 0.0 0.04 0.02 0.09 0.04 0.0 0.0 0.0 0.19 SSW 0.0 0.0 0.06 0.08 0.15 0.04 0.0 0.0 0.33 SW 0.0 0.02 0.11 0.13 0.05 0.02 0.0 0.0 0.33 WSW 0.0 0.0 0.0 0.02 0.0 0.0 0.0 0.0 0.02 W 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 WNW 0.01 0.0 0.0 0.0 0.05 0.0 0.0 0.0 0.06 NW 0.0 0.0 0.0 0.01 0.12 0.0 0.0 0.0 0.13 NNW 0.0 0.0 0.02 0.01 0.04 0.0 0.0 0.0 0.07 SUBTOTAL 0.01 0.40 1.00 0.82 0.93 0.07 0.0 0.0 3.23 CALM = 0.0 279 STABILITY CLASS A OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 276 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 279 STABILITY CLASS A OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-33 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS B

                                                                        -1.9< DELTA T< =-1.7 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.02 0.04 0.07 0.01 0.0 0.0 0.14 NNE 0.0 0.08 0.29 0.15 0.20 0.0 0.0 0.0 0.72 NE 0.0 0.09 0.32 0.08 0.09 0.01 0.0 0.0 0.59 ENE 0.0 0.04 0.04 0.0 0.0 0.0 0.0 0.0 0.08 E 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.0 0.03 ESE 0.0 0.02 0.02 0.0 0.0 0.0 0.0 0.0 0.04 SE 0.0 0.02 0.04 0.0 0.0 0.0 0.0 0.0 0.06 SSE 0.0 0.02 0.01 0.01 0.01 0.0 0.0 0.0 0.05 S 0.0 0.0 0.02 0.08 0.04 0.0 0.0 0.0 0.14 SSW 0.0 0.02 0.13 0.09 0.28 0.07 0.0 0.0 0.59 SW 0.0 0.04 0.05 0.08 0.05 0.01 0.0 0.0 0.23 WSW 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.03 W 0.0 0.0 0.0 0.01 0.05 0.0 0.0 0.0 0.08 WNW 0.0 0.0 0.01 0.01 0.04 0.0 0.0 0.0 0.06 NW 0.0 0.0 0.0 0.02 0.12 0.0 0.0 0.0 0.14 NNW 0.0 0.02 0.02 0.05 0.15 0.0 0.0 0.0 0.24 SUBTOTAL 0.0 0.37 0.98 0.64 1.11 0.10 0.0 0.0 3.20 CALM = 0.0 277 STABILITY CLASS B OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 276 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 277 STABILITY CLASS B OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-34 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS B

                                                                        -1.9< DELTA T<=-1.7 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.02 0.0 0.06 0.0 0.0 0.0 0.08 NNE 0.0 0.08 0.13 0.16 0.12 0.0 0.0 0.0 0.49 NE 0.0 0.15 0.28 0.07 0.08 0.0 0.0 0.0 0.58 ENE 0.0 0.01 0.02 0.0 0.0 0.0 0.0 0.0 0.03 E 0.0 0.02 0.0 0.0 0.0 0.0 0.0 0.0 0.02 ESE 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 SE 0.0 0.02 0.02 0.0 0.0 0.0 0.0 0.0 0.04 SSE 0.0 0.01 0.06 0.0 0.01 0.0 0.0 0.0 0.08 S 0.0 0.0 0.08 0.09 0.01 0.0 0.0 0.0 0.18 SSW 0.0 0.02 0.15 0.15 0.29 0.01 0.0 0.0 0.62 SW 0.0 0.01 0.11 0.18 0.13 0.01 0.0 0.0 0.44 WSW 0.0 0.0 0.02 0.04 0.0 0.01 0.0 0.0 0.07 W 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.03 WNW 0.0 0.0 0.0 0.01 0.04 0.0 0.0 0.0 0.05 NW 0.0 0.0 0.0 0.01 0.14 0.0 0.0 0.0 0.05 NNW 0.0 0.0 0.0 0.06 0.13 0.0 0.0 0.0 0.19 SUBTOTAL 0.0 0.32 0.90 0.79 0.92 0.03 0.0 0.0 2.96 CALM = 0.0 258 STABILITY CLASS B OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 256 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 258 STABILITY CLASS B OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-35 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS C

                                                                        -1.7<DELTA T<=-1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.02 0.02 0.02 0.0 0.0 0.0 0.07 NNE 0.0 0.02 0.07 0.09 0.05 0.01 0.0 0.0 0.24 NE 0.0 0.09 0.12 0.05 0.04 0.0 0.0 0.0 0.30 ENE 0.0 0.05 0.05 0.0 0.0 0.0 0.0 0.0 0.10 E 0.0 0.04 0.02 0.0 0.0 0.0 0.0 0.0 0.06 ESE 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 SE 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 SSE 0.0 0.02 0.07 0.01 0.0 0.0 0.0 0.0 0.10 S 0.0 0.02 0.02 0.05 0.04 0.01 0.0 0.0 0.14 SSW 0.0 0.0 0.12 0.16 0.20 0.01 0.0 0.0 0.49 SW 0.0 0.0 0.09 0.15 0.16 0.0 0.0 0.0 0.40 WSW 0.0 0.0 0.0 0.01 0.02 0.01 0.0 0.0 0.04 W 0.0 0.0 0.02 0.01 0.01 0.0 0.0 0.0 0.04 WNW 0.0 0.0 0.04 0.01 0.04 0.0 0.0 0.0 0.09 NW 0.0 0.0 0.0 0.0 0.08 0.0 0.0 0.0 0.08 NNW 0.0 0.0 0.0 0.01 0.07 0.0 0.0 0.0 0.08 SUBTOTAL 0.0 0.25 0.66 0.57 0.73 0.04 0.0 0.0 2.25 CALM = 0.0 196 STABILITY CLASS C OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 195 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 196 STABILITY CLASS C OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-36 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS C

                                                                        -1.7< DELTA T<=-1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.04 0.01 0.02 0.0 0.0 0.0 0.08 NNE 0.0 0.05 0.14 0.22 0.08 0.01 0.0 0.0 0.50 NE 0.0 0.09 0.15 0.09 0.05 0.01 0.0 0.0 0.39 ENE 0.0 0.02 0.0 0.0 0.0 0.0 0.0 0.0 0.02 E 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 ESE 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.0 0.01 SE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SSE 0.0 0.02 0.01 0.01 0.0 0.0 0.0 0.0 0.04 S 0.0 0.01 0.06 0.06 0.02 0.0 0.0 0.0 0.15 SSW 0.0 0.02 0.12 0.19 0.09 0.01 0.0 0.0 0.43 SW 0.0 0.04 0.08 0.11 0.06 0.0 0.0 0.0 0.29 WSW 0.0 0.04 0.05 0.01 0.04 0.0 0.0 0.0 0.14 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 0.01 0.02 0.0 0.0 0.0 0.03 NW 0.0 0.0 0.01 0.01 0.07 0.0 0.0 0.0 0.09 NNW 0.0 0.01 0.01 0.0 0.05 0.0 0.0 0.0 0.07 SUBTOTAL 0.0 0.33 0.68 0.72 0.50 0.03 0.0 0.0 2.26 CALM = 0.0 196 STABILITY CLASS C OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 195 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 196 STABILITY CLASS C OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-37 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS D

                                                                        -1.5< DELTA T<=-0.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.29 0.21 0.27 0.0 0.0 0.0 0.95 NNE 0.0 0.51 0.81 0.64 0.40 0.05 0.0 0.0 2.41 NE 0.0 0.88 0.68 0.26 0.19 0.0 0.0 0.0 2.01 ENE 0.0 0.23 0.08 0.0 0.0 0.0 0.0 0.0 0.31 E 0.0 0.15 0.04 0.0 0.0 0.0 0.0 0.0 0.19 ESE 0.0 0.08 0.02 0.0 0.0 0.0 0.0 0.0 0.10 SE 0.0 0.13 0.07 0.0 0.0 0.0 0.0 0.0 0.20 SSE 0.0 0.22 0.25 0.09 0.05 0.0 0.0 0.0 0.61 S 0.0 0.28 0.85 0.64 0.16 0.02 0.0 0.0 1.95 SSW 0.0 0.42 1.31 1.09 0.86 0.01 0.0 0.0 3.69 SW 0.01 0.48 1.52 1.59 0.39 0.0 0.0 0.0 3.99 WSW 0.0 0.18 0.30 0.19 0.22 0.01 0.0 0.0 0.90 W 0.0 0.06 0.14 0.05 0.05 0.0 0.0 0.0 0.30 WNW 0.0 0.04 0.01 0.09 0.18 0.0 0.0 0.0 0.32 NW 0.0 0.06 0.09 0.12 0.15 0.0 0.0 0.0 0.42 NNW 0.0 0.05 0.12 0.21 0.50 0.01 0.0 0.0 0.89 SUBTOTAL 0.01 3.95 6.58 5.18 3.42 0.10 0.0 0.0 19.24 CALM = 0.0 1656 STABILITY CLASS D OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 1645 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1656 STABILITY CLASS D OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-38 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS D

                                                                        -1.5< DELTA T< =-0.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.26 0.23 0.32 0.01 0.0 0.0 1.01 NNE 0.02 0.74 0.98 0.55 0.40 0.05 0.0 0.0 2.74 NE 0.0 0.67 0.55 0.22 0.15 0.0 0.0 0.0 1.59 ENE 0.01 0.27 0.11 0.0 0.0 0.0 0.0 0.0 0.39 E 0.0 0.13 0.06 0.0 0.0 0.0 0.0 0.0 0.19 ESE 0.0 0.06 0.02 0.0 0.0 0.0 0.0 0.0 0.06 SE 0.0 0.13 0.07 0.0 0.0 0.0 0.0 0.0 0.20 SSE 0.01 0.18 0.21 0.12 0.05 0.0 0.0 0.0 0.57 S 0.0 0.32 0.76 0.42 0.19 0.02 0.0 0.0 1.71 SSW 0.0 0.49 1.22 0.78 0.74 0.06 0.0 0.0 3.29 SW 0.01 0.40 1.29 1.26 0.33 0.04 0.0 0.0 3.33 WSW 0.0 0.16 0.26 0.18 0.21 0.0 0.0 0.0 0.81 W 0.0 0.07 0.12 0.09 0.08 0.0 0.0 0.0 0.36 WNW 0.0 0.06 0.07 0.08 0.16 0.0 0.0 0.0 0.37 NW 0.0 0.11 0.08 0.07 0.15 0.0 0.0 0.0 0.41 NNW 0.0 0.09 0.13 0.20 0.53 0.0 0.0 0.0 0.95 SUBTOTAL 0.05 4.07 6.19 4.20 3.31 0.18 0.0 0.0 18.00 CALM = 0.0 1548 STABILITY CLASS D OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 1536 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1548 STABILITY CLASS D OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-39 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS E

                                                                        -0.5< DELTA T<= 1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.08 1.25 0.99 0.76 0.58 0.01 0.0 0.0 3.67 NNE 0.08 2.40 2.31 1.05 1.20 0.05 0.01 0.0 7.10 NE 0.04 0.78 0.49 0.20 0.12 0.01 0.0 0.0 1.64 ENE 0.11 0.53 0.11 0.01 0.0 0.0 0.0 0.0 0.76 E 0.06 0.32 0.07 0.0 0.0 0.0 0.0 0.0 0.45 ESE 0.04 0.15 0.01 0.0 0.0 0.0 0.0 0.0 0.20 SE 0.08 0.51 0.05 0.0 0.0 0.0 0.0 0.0 0.64 SSE 0.02 0.83 0.22 0.20 0.28 0.02 0.0 0.0 1.57 S 0.04 1.51 1.71 0.81 1.90 0.07 0.0 0.0 5.04 SSW 0.06 1.89 2.26 1.65 1.13 0.05 0.0 0.0 7.04 SW 0.04 1.37 1.86 0.99 0.49 0.07 0.0 0.0 4.82 WSW 0.02 0.78 0.50 0.20 0.27 0.02 0.0 0.0 1.79 W 0.02 0.55 0.30 0.16 0.07 0.01 0.0 0.0 1.11 WNW 0.04 0.36 0.16 0.12 0.11 0.0 0.0 0.0 0.79 NW 0.09 0.71 0.46 0.51 0.34 0.04 0.0 0.0 2.15 NNW 0.07 0.86 0.79 0.84 0.63 0.0 0.0 0.0 3.19 SUBTOTAL 0.89 14.80 12.29 7.50 6.12 0.35 0.01 0.0 41.96 CALM = 0.01 3630 STABILITY CLASS E OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 3592 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 3630 STABILITY CLASS E OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-40 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS E

                                                                       -0.5< DELTA T<= 1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.11 1.34 1.04 0.76 0.55 0.01 0.0 0.0 3.81 NNE 0.06 2.52 2.09 1.08 1.16 0.04 0.01 0.0 7.02 NE 0.06 0.91 0.54 0.20 0.12 0.01 0.0 0.0 1.84 ENE 0.08 0.43 0.12 0.01 0.0 0.0 0.0 0.0 0.64 E 0.06 0.33 0.01 0.0 0.0 0.0 0.0 0.0 0.40 ESE 0.05 0.19 0.01 0.0 0.0 0.0 0.0 0.0 0.25 SE 0.12 0.47 0.05 0.0 0.0 0.0 0.0 0.0 0.64 SSE 0.04 0.02 0.27 0.20 0.25 0.02 0.0 0.0 1.60 S 0.02 1.48 1.66 0.86 0.92 0.07 0.0 0.0 5.01 SSW 0.08 1.81 2.33 1.79 1.25 0.05 0.0 0.0 7.31 SW 0.04 1.39 1.90 1.19 0.53 0.05 0.0 0.01 5.11 WSW 0.04 0.71 0.50 0.19 0.27 0.04 0.0 0.0 1.75 W 0.02 0.51 0.34 0.13 0.08 0.01 0.0 0.0 1.09 WNW 0.06 0.37 0.15 0.13 0.09 0.0 0.0 0.0 0.80 NW 0.09 0.65 0.46 0.51 0.33 0.04 0.0 0.0 2.08 NNW 0.08 0.85 0.68 0.85 0.64 0.01 0.0 0.0 3.11 SUBTOTAL 1.01 14.84 12.15 7.90 6.19 0.35 0.01 0.01 42.46 CALM = 0.02 3667 STABILITY CLASS E OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 3634 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 3667 STABILITY CLASS E OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-41 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS F 1.5< DELTA T<= 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 1.88 0.53 0.05 0.01 0.0 0.0 0.0 2.56 NNE 0.16 4.06 1.09 0.02 0.0 0.0 0.0 0.0 5.33 NE 0.07 0.90 0.18 0.04 0.0 0.0 0.0 0.0 1.19 ENE 0.06 0.36 0.05 0.0 0.0 0.0 0.0 0.0 0.47 E 0.12 0.30 0.0 0.0 0.0 0.0 0.0 0.0 0.42 ESE 0.09 0.26 0.0 0.0 0.0 0.0 0.0 0.0 0.35 SE 0.15 0.37 0.02 0.0 0.0 0.0 0.0 0.0 0.54 SSE 0.25 0.67 0.07 0.06 0.01 0.0 0.0 0.0 1.06 S 0.11 0.91 0.44 0.05 0.02 0.0 0.0 0.0 1.53 SSW 0.12 1.39 0.74 0.34 0.09 0.0 0.0 0.0 2.68 SW 0.02 1.10 0.60 0.20 0.05 0.0 0.0 0.0 1.97 WSW 0.08 0.47 0.11 0.02 0.0 0.0 0.0 0.0 0.68 W 0.06 0.21 0.05 0.04 0.0 0.0 0.0 0.0 0.36 WNW 0.14 0.27 0.05 0.01 0.01 0.0 0.0 0.0 0.48 NW 0.02 0.42 0.21 0.07 0.01 0.0 0.0 0.0 0.73 NNW 0.07 0.72 0.34 0.05 0.01 0.0 0.0 0.0 1.19 SUBTOTAL 1.61 14.29 4.48 0.95 0.21 0.0 0.0 0.0 21.54 CALM = 0.02 1852 STABILITY CLASS F OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 1843 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1852 STABILITY CLASS F OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-42 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS F 1.5< DELTA T<= 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 1.59 0.42 0.07 0.02 0.0 0.0 0.0 2.17 NNE 0.20 3.58 1.19 0.04 0.05 0.0 0.0 0.0 5.06 NE 0.06 0.71 0.22 0.05 0.0 0.0 0.0 0.0 1.04 ENE 0.07 0.35 0.02 0.0 0.0 0.0 0.0 0.0 0.44 E 0.13 0.27 0.02 0.0 0.0 0.0 0.0 0.0 0.42 ESE 0.12 0.23 0.02 0.0 0.0 0.0 0.0 0.0 0.37 SE 0.12 0.34 0.01 0.0 0.0 0.0 0.0 0.0 0.47 SSE 0.16 0.68 0.06 0.05 0.05 0.0 0.0 0.0 1.00 S 0.12 0.89 0.43 0.08 0.02 0.01 0.0 0.0 1.55 SSW 0.08 1.36 0.63 0.35 0.09 0.0 0.0 0.0 2.51 SW 0.01 1.02 0.68 0.15 0.06 0.0 0.0 0.0 1.92 WSW 0.07 0.50 0.09 0.02 0.01 0.0 0.0 0.0 0.69 W 0.08 0.19 0.05 0.04 0.0 0.0 0.0 0.0 0.34 WNW 0.07 0.20 0.06 0.01 0.0 0.0 0.0 0.0 0.34 NW 0.01 0.41 0.19 0.11 0.01 0.0 0.0 0.0 0.73 NNW 0.06 0.67 0.39 0.04 0.0 0.0 0.0 0.0 1.16 SUBTOTAL 1.41 12.99 4.48 1.01 0.31 0.01 0.0 0.0 20.21 CALM = 0.01 1739 STABILITY CLASS F OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 1728 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1739 STABILITY CLASS F OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-43 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS G DELTA T > 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.06 0.41 0.13 0.01 0.0 0.0 0.0 0.0 0.61 NNE 0.07 1.75 0.50 0.02 0.0 0.0 0.0 0.0 2.34 NE 0.12 0.72 0.11 0.01 0.0 0.0 0.0 0.0 0.96 ENE 0.15 0.48 0.0 0.0 0.0 0.0 0.0 0.0 0.63 E 0.21 0.29 0.0 0.0 0.0 0.0 0.0 0.0 0.50 ESE 0.19 0.11 0.02 0.0 0.0 0.0 0.0 0.0 0.32 SE 0.07 0.12 0.0 0.0 0.0 0.0 0.0 0.0 0.19 SSE 0.09 0.40 0.0 0.0 0.0 0.0 0.0 0.0 0.49 S 0.09 0.71 0.05 0.0 0.0 0.0 0.0 0.0 0.85 SSW 0.02 0.98 0.51 0.0 0.0 0.0 0.0 0.0 1.51 SW 0.02 0.44 0.56 0.04 0.0 0.0 0.0 0.0 1.06 WSW 0.01 0.12 0.02 0.0 0.0 0.0 0.0 0.0 0.15 W 0.02 0.04 0.01 0.0 0.0 0.0 0.0 0.0 0.07 WNW 0.02 0.06 0.01 0.0 0.01 0.0 0.0 0.0 0.10 NW 0.0 0.06 0.01 0.01 0.0 0.0 0.0 0.0 0.08 NNW 0.0 0.08 0.0 0.0 0.0 0.0 0.0 0.0 0.08 SUBTOTAL 1.14 6.77 1.93 0.09 0.01 0.0 0.0 0.0 9.94 CALM = 0.02 855 STABILITY CLASS G OCCURRENCES OUT OF TOTAL 8620 VALID TEMPERATURE DIFFERENCE READINGS 855 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 855 STABILITY CLASS G OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-44 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS G DELTA T > 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.08 0.56 0.20 0.0 0.0 0.0 0.0 0.0 0.82 NNE 0.04 1.73 0.42 0.01 0.0 0.0 0.0 0.0 2.20 NE 0.11 0.85 0.08 0.01 0.0 0.0 0.0 0.0 1.05 ENE 0.15 0.54 0.01 0.0 0.0 0.0 0.0 0.0 0.70 E 0.20 0.32 0.0 0.0 0.0 0.0 0.0 0.0 0.52 ESE 0.15 0.12 0.01 0.0 0.0 0.0 0.0 0.0 0.28 SE 0.07 0.20 0.01 0.0 0.0 0.0 0.0 0.0 0.28 SSE 0.15 0.44 0.01 0.01 0.0 0.0 0.0 0.0 0.61 S 0.09 0.69 0.08 0.0 0.0 0.0 0.0 0.0 0.86 SSW 0.04 1.00 0.56 0.01 0.0 0.0 0.0 0.0 1.61 SW 0.04 0.55 0.55 0.05 0.0 0.0 0.0 0.0 1.19 WSW 0.01 0.13 0.02 0.0 0.0 0.0 0.0 0.0 0.16 W 0.02 0.08 0.01 0.0 0.0 0.0 0.0 0.0 0.11 WNW 0.06 0.09 0.0 0.0 0.01 0.0 0.0 0.0 0.16 NW 0.0 0.08 0.04 0.01 0.0 0.0 0.0 0.0 0.13 NNW 0.0 0.12 0.05 0.01 0.01 0.0 0.0 0.0 0.19 SUBTOTAL 1.19 7.50 2.05 0.11 0.02 0.0 0.0 0.0 10.87 CALM = 0.02 934 STABILITY CLASS G OCCURRENCES OUT OF TOTAL 8621 VALID TEMPERATURE DIFFERENCE READINGS 933 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 934 STABILITY CLASS G OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 150 FEET ABOVE GROUND WIND INSTRUMENTS 33 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-45 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS A DELTA T<=-1.9 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.02 0.0 0.0 0.0 0.02 NNE 0.0 0.0 0.0 0.0 0.0 0.04 0.0 0.0 0.04 NE 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.03 ENE 0.0 0.0 0.0 0.0 0.02 0.0 0.0 0.0 0.02 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 0.01 0.0 0.0 0.01 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 0.0 0.0 0.0 0.0 0.0 0.0 WSW 0.0 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.01 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 0.01 0.0 0.0 0.01 SUBTOTAL 0.0 0.0 0.0 0.0 0.07 0.07 0.0 0.0 0.14 CALM = 0.0 13 STABILITY CLASS A OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 13 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 13 STABILITY CLASS A OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-46 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS A DELTA T<=-1.9 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.01 0.02 0.0 0.0 0.03 NNE 0.0 0.0 0.0 0.0 0.06 0.04 0.0 0.0 0.10 NE 0.0 0.0 0.01 0.0 0.06 0.04 0.0 0.0 0.11 ENE 0.0 0.0 0.01 0.0 0.06 0.0 0.0 0.0 0.07 E 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.0 0.01 ESE 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 SE 0.0 0.0 0.02 0.0 0.0 0.0 0.0 0.0 0.02 SSE 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.01 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 0.05 0.0 0.01 0.0 0.06 SW 0.0 0.01 0.0 0.01 0.02 0.01 0.0 0.0 0.05 WSW 0.01 0.0 0.0 0.01 0.04 0.0 0.0 0.0 0.06 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 0.02 0.0 0.0 0.02 NW 0.0 0.0 0.0 0.0 0.01 0.01 0.0 0.0 0.02 NNW 0.0 0.0 0.0 0.0 0.04 0.01 0.01 0.0 0.06 SUBTOTAL 0.01 0.01 0.05 0.03 0.35 0.15 0.02 0.0 0.62 CALM = 0.0 54 STABILITY CLASS A OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 54 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 54 STABILITY CLASS A OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-47 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS B

                                                                        -1.9< DELTA T<=-1.7 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.01 0.05 0.06 0.0 0.0 0.12 NNE 0.0 0.0 0.0 0.0 0.15 0.02 0.01 0.0 0.18 NE 0.0 0.0 0.0 0.02 0.11 0.07 0.0 0.0 0.20 ENE 0.0 0.0 0.0 0.01 0.02 0.0 0.0 0.0 0.03 E 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ESE 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.03 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 0.02 0.0 0.0 0.02 SSW 0.0 0.0 0.0 0.0 0.02 0.05 0.01 0.0 0.08 SW 0.0 0.0 0.0 0.0 0.01 0.02 0.0 0.0 0.03 WSW 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.04 0.04 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 0.01 0.04 0.0 0.0 0.05 NNW 0.0 0.0 0.0 0.0 0.05 0.04 0.01 0.0 0.10 SUBTOTAL 0.0 0.0 0.03 0.05 0.42 0.32 0.03 0.04 0.89 CALM = 0.0 78 STABILITY CLASS B OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 77 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 78 STABILITY CLASS B OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-48 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS B

                                                                         -1.9 < DELTA T<=-1.7 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.05 0.02 0.0 0.0 0.07 NNE 0.0 0.0 0.01 0.0 0.11 0.02 0.02 0.0 0.16 NE 0.0 0.0 0.02 0.04 0.08 0.06 0.0 0.0 0.20 ENE 0.0 0.0 0.02 0.01 0.0 0.0 0.0 0.0 0.03 E 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 ESE 0.0 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.02 SE 0.0 0.0 0.02 0.0 0.0 0.0 0.0 0.0 0.02 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 0.01 0.01 0.02 0.01 0.0 0.05 SSW 0.0 0.0 0.0 0.02 0.07 0.14 0.0 0.01 0.24 SW 0.0 0.0 0.0 0.04 0.07 0.08 0.0 0.0 0.19 WSW 0.0 0.0 0.01 0.0 0.0 0.0 0.0 0.02 0.03 W 0.0 0.0 0.0 0.0 0.01 0.0 0.0 0.0 0.01 WNW 0.0 0.0 0.0 0.0 0.0 0.02 0.0 0.0 0.02 NW 0.0 0.0 0.0 0.0 0.01 0.04 0.0 0.0 0.05 NNW 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.03 SUBTOTAL 0.0 0.0 0.09 0.13 0.43 0.41 0.03 0.03 1.12 CALM = 0.0 100 STABILITY CLASS B OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 99 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 100 STABILITY CLASS B OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-49 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS C

                                                                        -1.7 < DELTA T<=-1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.01 0.09 0.08 0.06 0.0 0.24 NNE 0.0 0.0 0.05 0.02 0.18 0.09 0.01 0.0 0.35 NE 0.0 0.01 0.02 0.02 0.22 0.16 0.02 0.0 0.45 ENE 0.0 0.01 0.04 0.02 0.02 0.0 0.0 0.0 0.09 E 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.0 0.02 ESE 0.0 0.0 0.0 0.02 0.0 0.0 0.0 0.0 0.02 SE 0.0 0.0 0.02 0.0 0.0 0.0 0.0 0.0 0.02 SSE 0.0 0.0 0.01 0.01 0.0 0.0 0.0 0.0 0.02 S 0.0 0.0 0.0 0.0 0.04 0.01 0.01 0.01 0.07 SSW 0.0 0.0 0.0 0.01 0.14 0.21 0.04 0.02 0.42 SW 0.0 0.0 0.0 0.02 0.13 0.14 0.01 0.0 0.30 WSW 0.0 0.0 0.0 0.0 0.02 0.0 0.01 0.0 0.03 W 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.03 WNW 0.0 0.0 0.0 0.0 0.01 0.05 0.0 0.0 0.06 NW 0.0 0.0 0.0 0.0 0.06 0.08 0.01 0.0 0.15 NNW 0.0 0.01 0.0 0.0 0.02 0.07 0.0 0.0 0.10 SUBTOTAL 0.0 0.04 0.15 0.13 0.95 0.90 0.17 0.03 2.37 CALM = 0.0 208 STABILITY CLASS C OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 208 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 208 STABILITY CLASS C OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-50 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS C

                                                                       -1.7< DELTA T<= -1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.06 0.04 0.0 0.26 NNE 0.0 0.01 0.05 0.04 0.23 0.12 0.0 0.0 0.45 NE 0.0 0.05 0.01 0.07 0.11 0.14 0.04 0.0 0.42 ENE 0.0 0.0 0.01 0.01 0.0 0.0 0.01 0.0 0.03 E 0.0 0.0 0.05 0.01 0.0 0.0 0.0 0.0 0.06 ESE 0.0 0.01 0.01 0.02 0.0 0.0 0.0 0.0 0.04 SE 0.0 0.0 0.02 0.01 0.0 0.01 0.0 0.0 0.04 SSE 0.0 0.01 0.0 0.02 0.02 0.0 0.0 0.0 0.05 S 0.0 0.0 0.0 0.0 0.02 0.01 0.0 0.0 0.03 SSW 0.0 0.02 0.01 0.05 0.13 0.15 0.01 0.01 0.38 SW 0.0 0.0 0.06 0.09 0.22 0.06 0.0 0.0 0.43 WSW 0.0 0.0 0.01 0.02 0.07 0.0 0.01 0.01 0.12 W 0.0 0.02 0.0 0.01 0.0 0.01 0.0 0.0 0.04 WNW 0.0 0.0 0.0 0.0 0.0 0.05 0.0 0.0 0.05 NW 0.0 0.0 0.0 0.0 0.0 0.06 0.01 0.0 0.07 NNW 0.0 0.0 0.0 0.0 0.02 0.12 0.0 0.0 0.14 SUBTOTAL 0.0 0.12 0.23 0.39 0.94 0.79 0.12 0.02 2.61 CALM = 0.0 225 STABILITY CLASS C OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 225 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 225 STABILITY CLASS C OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-51 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS D

                                                                      -1.5< DELTA T<=-0.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.13 0.25 0.22 0.68 0.96 0.29 0.01 2.55 NNE 0.0 0.29 0.55 0.74 1.63 0.84 0.14 0.0 4.19 NE 0.0 0.50 0.60 0.56 0.90 0.55 0.09 0.0 3.20 ENE 0.0 0.32 0.38 0.20 0.19 0.01 0.11 0.0 1.21 E 0.0 0.21 0.25 0.08 0.05 0.02 0.01 0.0 0.62 ESE 0.0 0.18 0.12 0.05 0.04 0.0 0.0 0.0 0.39 SE 0.0 0.12 0.33 0.04 0.02 0.01 0.0 0.0 0.52 SSE 0.0 0.18 0.27 0.14 0.11 0.12 0.0 0.0 0.82 S 0.0 0.38 0.36 0.28 0.45 0.46 0.22 0.04 2.19 SSW 0.0 0.34 0.93 0.81 1.91 1.00 0.21 0.05 5.25 SW 0.01 0.25 1.34 1.29 2.06 0.46 0.08 0.04 5.53 WSW 0.0 0.22 0.59 0.49 0.54 0.26 0.07 0.0 2.17 W 0.01 0.16 0.11 0.09 0.25 0.21 0.07 0.02 0.92 WNW 0.0 0.04 0.05 0.05 0.28 0.25 0.05 0.0 0.72 NW 0.0 0.04 0.09 0.08 0.47 0.64 0.13 0.04 1.49 NNW 0.0 0.05 0.08 0.12 0.63 0.70 0.20 0.0 1.78 SUBTOTAL 0.03 3.41 6.30 5.24 10.21 6.49 1.67 0.20 33.55 CALM = 0.0 2873 STABILITY CLASS D OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 2857 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 2873 STABILITY CLASS D OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-52 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS D

                                                                       -1.5< DELTA T<=-0.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.09 0.23 0.20 0.61 1.02 0.32 0.01 2.49 NNE 0.0 0.30 0.61 0.75 1.63 0.88 0.20 0.0 4.37 NE 0.0 0.48 0.56 0.57 1.05 0.57 0.11 0.0 3.34 ENE 0.0 0.30 0.38 0.22 0.16 0.01 0.07 0.0 1.14 E 0.0 0.23 0.19 0.07 0.05 0.02 0.01 0.0 0.57 ESE 0.01 0.18 0.12 0.06 0.04 0.0 0.0 0.0 0.41 SE 0.0 0.13 0.27 0.01 0.01 0.0 0.0 0.0 0.42 SSE 0.0 0.18 0.27 0.09 0.08 0.08 0.0 0.0 0.70 S 0.0 0.41 0.34 0.28 0.36 0.47 0.20 0.04 2.10 SSW 0.0 0.27 1.00 0.74 1.79 1.04 0.21 0.05 5.10 SW 0.0 0.26 1.30 1.14 1.88 0.46 0.08 0.05 5.17 WSW 0.0 0.16 0.57 0.46 0.42 0.25 0.08 0.0 1.94 W 0.01 0.12 0.12 0.08 0.27 0.22 0.08 0.02 0.92 WNW 0.0 0.05 0.05 0.05 0.30 0.19 0.05 0.0 0.69 NW 0.0 0.06 0.07 0.08 0.49 0.64 0.11 0.02 1.47 NNW 0.0 0.07 0.05 0.13 0.66 0.69 0.20 0.0 1.80 SUBTOTAL 0.03 3.29 6.13 4.93 9.80 6.54 1.72 0.19 32.63 CALM = 0.0 2800 STABILITY CLASS D OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 2785 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 2800 STABILITY CLASS D OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-53 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS E

                                                                       -0.5< DELTA T<= 1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.06 0.23 0.22 0.27 0.89 0.70 0.13 0.0 2.50 NNE 0.0 0.41 0.84 0.89 2.11 1.10 0.22 0.04 5.61 NE 0.01 0.46 0.67 0.73 1.10 0.27 0.18 0.02 3.44 ENE 0.01 0.33 0.29 0.08 0.18 0.06 0.0 0.0 0.95 E 0.01 0.14 0.14 0.08 0.11 0.02 0.0 0.0 0.50 ESE 0.02 0.23 0.06 0.07 0.0 0.01 0.0 0.0 0.39 SE 0.01 0.21 0.12 0.06 0.05 0.02 0.0 0.0 0.47 SSE 0.02 0.27 0.14 0.11 0.35 0.23 0.07 0.0 1.19 S 0.02 0.47 0.36 0.39 0.96 1.15 0.39 0.12 3.86 SSW 0.04 0.41 1.30 1.29 2.93 2.41 0.49 0.07 8.94 SW 0.01 0.43 1.11 1.27 2.20 0.71 0.25 0.05 6.03 WSW 0.05 0.38 0.52 0.46 0.75 0.20 0.05 0.0 2.41 W 0.02 0.13 0.15 0.25 0.25 0.15 0.04 0.0 0.99 WNW 0.01 0.18 0.09 0.09 0.30 0.08 0.0 0.0 1.75 NW 0.0 0.14 0.18 0.15 0.52 0.35 0.09 0.0 1.43 NNW 0.0 0.26 0.16 0.16 0.76 0.35 0.02 0.0 1.71 SUBTOTAL 0.29 4.68 6.35 6.35 13.46 7.81 1.93 0.30 41.17 CALM = 0.0 3542 STABILITY CLASS E OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 3515 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 3542 STABILITY CLASS E OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-54 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS E

                                                                       -0.5< DELTA T<= 1.5 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.32 0.23 0.33 0.93 0.68 0.13 0.0 2.67 NNE 0.0 0.39 0.76 0.82 2.16 1.04 0.16 0.04 5.37 NE 0.01 0.49 0.66 0.68 1.01 0.26 0.15 0.02 3.28 ENE 0.01 0.32 0.27 0.06 0.20 0.09 0.02 0.0 0.97 E 0.0 0.13 0.16 0.07 0.09 0.02 0.0 0.0 0.47 ESE 0.01 0.22 0.06 0.06 0.0 0.01 0.0 0.0 0.36 SE 0.01 0.20 0.13 0.06 0.06 0.04 0.0 0.0 0.50 SSE 0.02 0.27 0.12 0.13 0.33 0.28 0.07 0.0 1.22 S 0.01 0.41 0.38 0.38 1.00 1.13 0.41 0.13 3.85 SSW 0.04 0.45 1.24 1.31 2.99 2.39 0.50 0.07 8.99 SW 0.02 0.42 1.10 1.38 2.25 0.74 0.25 0.05 6.21 WSW 0.05 0.43 0.48 0.56 0.76 0.21 0.04 0.0 2.53 W 0.01 0.15 0.16 0.22 0.26 0.13 0.02 0.0 0.95 WNW 0.01 0.14 0.07 0.08 0.28 0.11 0.0 0.0 0.69 NW 0.0 0.12 0.20 0.15 0.53 0.35 0.12 0.01 1.48 NNW 0.0 0.26 0.19 0.16 0.71 0.33 0.02 0.0 1.67 SUBTOTAL 0.25 4.72 6.21 6.45 13.56 7.81 1.89 0.32 41.21 CALM = 0.0 3542 STABILITY CLASS E OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 3516 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 3542 STABILITY CLASS E OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-55 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS F 1.5< DELTA T<= 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.15 0.30 0.49 0.13 0.0 0.0 1.26 NNE 0.01 0.21 0.40 0.50 1.24 0.36 0.01 0.0 2.73 NE 0.0 0.18 0.42 0.41 0.23 0.0 0.0 0.0 1.24 ENE 0.01 0.06 0.09 0.08 0.06 0.06 0.0 0.0 0.36 E 0.01 0.05 0.05 0.02 0.01 0.0 0.0 0.0 0.14 ESE 0.0 0.02 0.05 0.0 0.0 0.0 0.0 0.0 0.07 SE 0.0 0.06 0.02 0.04 0.01 0.02 0.0 0.0 0.15 SSE 0.0 0.13 0.12 0.01 0.14 0.09 0.0 0.0 0.49 S 0.0 0.25 0.19 0.12 0.61 0.19 0.0 0.0 1.36 SSW 0.01 0.20 0.29 0.40 1.20 0.35 0.01 0.0 2.46 SW 0.01 0.22 0.53 0.64 0.79 0.09 0.0 0.0 2.28 WSW 0.01 0.20 0.27 0.42 0.26 0.04 0.0 0.0 1.20 W 0.02 0.07 0.11 0.13 0.20 0.01 0.0 0.0 0.54 WNW 0.01 0.07 0.01 0.02 0.01 0.02 0.0 0.0 0.14 NW 0.0 0.06 0.05 0.05 0.02 0.02 0.0 0.0 0.20 NNW 0.01 0.12 0.09 0.08 0.11 0.02 0.0 0.01 0.44 SUBTOTAL 0.10 2.09 2.84 3.22 5.38 1.40 0.02 0.01 15.06 CALM = 0.0 1294 STABILITY CLASS F OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 1288 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1294 STABILITY CLASS F OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-56 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS F 1.5< DELTA T< = 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.14 0.14 0.28 0.48 0.12 0.0 0.0 1.18 NNE 0.01 0.20 0.42 0.53 1.09 0.39 0.01 0.0 2.65 NE 0.0 0.11 0.43 0.39 0.28 0.0 0.0 0.0 1.21 ENE 0.01 0.11 0.11 0.06 0.07 0.02 0.0 0.0 0.38 E 0.02 0.05 0.04 0.04 0.02 0.0 0.0 0.0 0.17 ESE 0.0 0.02 0.02 0.0 0.0 0.0 0.0 0.0 0.04 SE 0.0 0.08 0.04 0.05 0.01 0.01 0.0 0.0 0.19 SSE 0.0 0.12 0.13 0.02 0.16 0.07 0.0 0.0 0.50 S 0.01 0.29 0.20 0.13 0.63 0.21 0.01 0.0 1.48 SSW 0.01 0.23 0.26 0.41 1.13 0.29 0.02 0.0 2.35 SW 0.01 0.21 0.52 0.54 0.74 0.11 0.01 0.0 2.14 WSW 0.0 0.19 0.30 0.30 0.26 0.04 0.0 0.0 1.09 W 0.02 0.08 0.09 0.12 0.18 0.02 0.0 0.0 0.51 WNW 0.01 0.09 0.04 0.04 0.02 0.01 0.0 0.0 0.21 NW 0.0 0.07 0.05 0.02 0.05 0.04 0.0 0.0 0.23 NNW 0.02 0.12 0.11 0.05 0.12 0.04 0.0 0.01 0.47 SUBTOTAL 0.14 2.11 2.90 2.98 5.24 1.37 0.05 0.01 14.80 CALM = 0.0 1270 STABILITY CLASS F OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 1262 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 1270 STABILITY CLASS F OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.2-57 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS G DELTA T > 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.04 0.06 0.15 0.28 0.01 0.0 0.0 0.56 NNE 0.0 0.06 0.11 0.25 0.29 0.14 0.0 0.0 0.85 NE 0.01 0.07 0.05 0.05 0.01 0.0 0.0 0.0 0.19 ENE 0.0 0.06 0.02 0.01 0.01 0.0 0.0 0.0 0.10 E 0.0 0.04 0.01 0.01 0.0 0.0 0.0 0.0 0.06 ESE 0.01 0.07 0.0 0.0 0.0 0.0 0.0 0.0 0.08 SE 0.01 0.09 0.02 0.02 0.0 0.01 0.0 0.0 0.15 SSE 0.01 0.02 0.02 0.08 0.0 0.0 0.0 0.0 0.13 S 0.01 0.16 0.21 0.13 0.33 0.01 0.01 0.0 0.86 SSW 0.01 0.22 0.25 0.32 0.73 0.21 0.0 0.0 1.74 SW 0.0 0.11 0.19 0.21 0.45 0.07 0.0 0.0 1.03 WSW 0.0 0.11 0.08 0.06 0.02 0.0 0.0 0.0 0.27 W 0.01 0.08 0.06 0.01 0.05 0.0 0.0 0.0 0.21 WNW 0.01 0.07 0.06 0.02 0.0 0.0 0.01 0.0 0.17 NW 0.0 0.04 0.01 0.01 0.01 0.0 0.0 0.0 0.07 NNW 0.02 0.09 0.04 0.05 0.04 0.0 0.0 0.0 0.24 SUBTOTAL 0.12 1.33 1.19 1.38 2.22 0.45 0.02 0.0 6.71 CALM = 0.0 581 STABILITY CLASS G OCCURRENCES OUT OF TOTAL 8589 VALID TEMPERATURE DIFFERENCE READINGS 574 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 581 STABILITY CLASS G OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "HOURLY AVERAGE TEMPERATURE" T232-31to58.doc

SQN TABLE 2.3.2-58 JOINT PERCENTAGE FREQUENCIES OF WIND DIRECTION AND WIND SPEED FOR DIFFERENT STABILITY CLASSES* STABILITY CLASS G DELTA T > 4.0 DEG. C/100M SEQUOYAH NUCLEAR PLANT METEOROLOGICAL FACILITY MAY 1, 75 - APRIL 30, 76 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 0.04 0.07 0.13 0.30 0.02 0.0 0.0 0.57 NNE 0.0 0.07 0.09 0.27 0.32 0.12 0.0 0.0 0.87 NE 0.01 0.09 0.05 0.05 0.02 0.0 0.0 0.0 0.22 ENE 0.0 0.04 0.02 0.05 0.01 0.0 0.0 0.0 0.12 E 0.0 0.04 0.01 0.01 0.0 0.0 0.0 0.0 0.06 ESE 0.01 0.07 0.02 0.0 0.0 0.0 0.0 0.0 0.10 SE 0.01 0.07 0.02 0.02 0.0 0.01 0.0 0.0 0.13 SSE 0.01 0.02 0.05 0.07 0.0 0.01 0.0 0.0 0.16 S 0.01 0.14 0.22 0.12 0.36 0.01 0.0 0.0 0.86 SSW 0.01 0.19 0.26 0.30 0.74 0.22 0.0 0.0 1.72 SW 0.0 0.12 0.20 0.23 0.47 0.05 0.0 0.0 1.07 WSW 0.0 0.12 0.07 0.07 0.06 0.0 0.0 0.0 0.32 W 0.02 0.07 0.05 0.05 0.05 0.0 0.0 0.0 0.24 WNW 0.01 0.07 0.06 0.02 0.0 0.0 0.01 0.0 0.17 NW 0.0 0.02 0.01 0.04 0.0 0.0 0.0 0.0 0.07 NNW 0.01 0.08 0.04 0.07 0.04 0.0 0.0 0.0 0.24 SUBTOTAL 0.11 1.25 1.24 1.50 2.37 0.44 0.01 0.0 6.92 CALM = 0.0 599 STABILITY CLASS G OCCURRENCES OUT OF TOTAL 8590 VALID TEMPERATURE DIFFERENCE READINGS 592 VALID WIND DIRECTION - WIND SPEED READINGS OUT OF TOTAL 599 STABILITY CLASS G OCCURRENCES ALL COLUMNS AND CALM TOTAL 100 PERCENT OF NET VALID READINGS

METEOROLOGICAL FACILITY LOCATED .74 MILES SW OF SEQUOYAH NUCLEAR PLANT TEMPERATURE INSTRUMENTS 33 AND 300 FEET ABOVE GROUND WIND INSTRUMENTS 300 FEET ABOVE GROUND "END OF HOUR TEMPERATURE READINGS" T232-31to58.doc

SQN TABLE 2.3.4-1 DISTANCES FROM RELEASE ZONES OR POINTS TO EXCLUSION AREA BOUNDARY Sequoyah Nuclear Plant Distance From Distance From Distance From Release Zone 1a Release Zone 2b Release Zone 3c Sector (Meters) (Meters) (Meters) N 945 899 899 NNE 732 732 732 NE 701 863 701 ENE 556 600 556 E 564 604 564 ESE 610 692 610 SE 640 811 640 SSE 701 899 701 S 869 1049 869 SSW 983 1125 975 SW 1280 1372 1256 WSW 914 936 823 W 671 823 524 WNW 655 619 509 NW 663 637 524 NNW 732 710 771 a. Release Zone 1 - Auxiliary building vent exhaust and shield building vent exhaust. b. Release Zone 2 - Radioactive chemical hood exhaust. c. Release Zone 3 - Condenser air ejector exhaust. T234-1.doc

SQN TABLE 2.3.4-2 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT EXCLUSION AREA BOUNDARY DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 1* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.900E 0.999E-02 1 0.00 0.00 0.800E 0.899E-02 2 0.01 0.01 0.700E 0.799E-02 2 0.01 0.02 0.600E 0.699E-02 8 0.03 0.04 0.500E 0.599E-02 3 0.01 0.05 0.400E 0.499E-02 30 0.09 0.14 0.300E 0.399E-02 39 0.12 0.27 0.200E 0.299E-02 120 0.38 0.64 0.100E 0.199E-02 906 2.84 3.48 0.900E 0.999E-03 324 1.02 4.50 0.800E 0.899E-03 390 1.22 5.72 0.700E 0.799E-03 545 1.71 7.43 0.600E 0.699E-03 834 2.62 10.05 0.500E 0.599E-03 1198 3.76 13.80 0.400E 0.499E-03 1867 5.85 19.66 0.300E 0.399E-03 2782 8.72 28.38 0.200E 0.299E-03 3966 12.44 40.82 0.100E 0.199E-03 7864 24.66 65.48 0.900E 0.999E-04 1272 3.99 69.47 0.800E 0.899E-04 1236 3.88 73.34 0.700E 0.799E-04 1471 4.61 77.96 0.600E 0.699E-04 1415 4.44 82.40 0.500E 0.599E-04 1234 3.87 86.26 0.400E 0.499E-04 1050 3.29 89.56 0.300E 0.399E-04 750 2.35 91.91 0.200E 0.299E-04 661 2.07 93.98 0.100E 0.199E-04 673 2.11 96.09 0.900E 0.999E-05 52 0.16 96.26 0.800E 0.899E-05 61 0.19 96.45 0.700E 0.799E-05 72 0.23 96.67 0.600E 0.699E-05 60 0.19 96.86 0.500E 0.599E-05 69 0.22 97.08 0.400E 0.499E-05 106 0.33 97.41 0.300E 0.399E-05 122 0.38 97.79 0.200E 0.299E-05 187 0.59 98.38 0.100E 0.199E-05 239 0.75 99.13

    <= 0.999E-06                                                        278               0.87             100.00 TOTALS                                                             31889             100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.859E-03 SEC/M3, 50TH PERCENTILE= 0.163E-03 SEC/M3, AVERAGE= 0.269E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 1 - Auxiliary building vent exhaust and shield building vent.

T23402.doc

SQN TABLE 2.3.4-3 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT EXCLUSION AREA BOUNDARY DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 2* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.800E 0.899E-02 1 0.00 0.00 0.700E 0.799E-02 2 0.01 0.01 0.600E 0.699E-02 7 0.02 0.03 0.500E 0.599E-02 5 0.02 0.05 0.400E 0.499E-02 18 0.06 0.10 0.300E 0.399E-02 26 0.08 0.19 0.200E 0.299E-02 126 0.40 0.58 0.100E 0.199E-02 766 2.40 2.98 0.900E 0.999E-03 245 0.77 3.75 0.800E 0.899E-03 373 1.17 4.92 0.700E 0.799E-03 470 1.47 6.39 0.600E 0.699E-03 710 2.23 8.62 0.500E 0.599E-03 939 2.94 11.57 0.400E 0.499E-03 1641 5.15 16.71 0.300E 0.399E-03 2643 8.23 24.94 0.200E 0.299E-03 3878 12.16 37.10 0.100E 0.199E-03 7483 23.47 60.56 0.900E 0.999E-04 1295 4.06 64.62 0.800E 0.899E-04 1336 4.19 68.81 0.700E 0.799E-04 1490 4.67 73.49 0.600E 0.699E-04 1547 4.85 78.34 0.500E 0.599E-04 1565 4.91 83.24 0.400E 0.499E-04 1360 4.26 87.51 0.300E 0.399E-04 1010 3.17 90.68 0.200E 0.299E-04 817 2.56 93.24 0.100E 0.199E-04 778 2.44 95.68 0.900E 0.999E-05 62 0.19 95.87 0.800E 0.899E-05 76 0.24 96.11 0.700E 0.799E-05 67 0.21 96.32 0.600E 0.699E-05 74 0.23 96.55 0.500E 0.599E-05 75 0.24 96.79 0.400E 0.499E-05 70 0.22 97.01 0.300E 0.399E-05 129 0.40 97.41 0.200E 0.299E-05 184 0.58 97.99 0.100E 0.199E-05 219 0.69 98.68

    <= 0.999E-06                                               422                    1.32                 100.00 TOTALS                                                    31889                  100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.795E-03 SEC/M3, 50TH PERCENTILE= 0.145E-03 SEC/M3, AVERAGE= 0.243E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 2 - Radioactive chemical hood exhaust.

T234-03.doc

SQN TABLE 2.3.4-4 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT EXCLUSION AREA BOUNDARY DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 3* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.100E 0.199E-01 1 0.00 0.00 0.900E 0.999E-02 1 0.00 0.01 0.800E 0.899E-02 2 0.01 0.01 0.700E 0.799E-02 1 0.00 0.02 0.600E 0.699E-02 5 0.02 0.03 0.500E 0.599E-02 19 0.06 0.09 0.400E 0.499E-02 26 0.08 0.17 0.300E 0.399E-02 63 0.20 0.37 0.200E 0.299E-02 176 0.55 0.92 0.100E 0.199E-02 972 3.05 3.97 0.900E 0.999E-03 294 0.92 4.89 0.800E 0.899E-03 421 1.32 6.21 0.700E 0.799E-03 524 1.64 7.86 0.600E 0.699E-03 849 2.66 10.52 0.500E 0.599E-03 1194 3.74 14.26 0.400E 0.499E-03 1819 5.70 19.97 0.300E 0.399E-03 2806 8.80 28.77 0.200E 0.299E-03 3981 12.48 41.25 0.100E 0.199E-03 7836 24.57 65.82 0.900E 0.999E-04 1253 3.93 69.75 0.800E 0.899E-04 1221 3.83 73.58 0.700E 0.799E-04 1449 4.54 78.12 0.600E 0.699E-04 1415 4.44 82.56 0.500E 0.599E-04 1222 3.83 86.39 0.400E 0.499E-04 1051 3.30 89.69 0.300E 0.399E-04 705 2.21 91.90 0.200E 0.299E-04 665 2.09 93.99 0.100E 0.199E-04 683 2.14 96.13 0.900E 0.999E-05 54 0.17 96.30 0.800E 0.899E-05 62 0.19 96.49 0.700E 0.799E-05 58 0.18 96.67 0.600E 0.699E-05 69 0.22 96.89 0.500E 0.599E-05 58 0.18 96.07 0.400E 0.499E-05 102 0.32 97.39 0.300E 0.399E-05 131 0.41 97.80 0.200E 0.299E-05 196 0.61 98.42 0.100E 0.199E-05 238 0.75 99.16

    <= 0.999E-06                                              267                     0.84                  100.00 TOTALS                                                   31889                   100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.892E-03 SEC/M3, 50TH PERCENTILE= 0.164E-03 SEC/M3, AVERAGE= 0.279E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 3 - Condenser air ejector exhaust.

T234-04.doc

SQN TABLE 2.3.4-5 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT 556 METERS (MINIMUM EXCLUSIVE AREA BOUNDARY DISTANCE) DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 1* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.900E 0.999E-02 18 0.06 0.06 0.400E 0.499E-02 82 0.26 0.31 0.300E 02 - 0.399E-02 103 0.32 0.64 0.200E 0.299E-02 346 1.09 1.72 0.100E 0.199E-02 1963 6.16 7.88 0.900E 0.999E-03 649 2.04 9.91 0.800E 0.899E-03 700 2.20 12.11 0.700E 0.799E-03 810 2.54 14.65 0.600E 0.699E-03 1319 4.14 18.78 0.500E 0.599E-03 1514 4.75 23.53 0.400E 0.499E-03 2327 7.30 30.83 0.300E 0.399E-03 3063 9.61 40.43 0.200E 0.299E-03 4622 14.49 54.93 0.100E 0.199E-03 8358 26.21 81.14 0.900E 0.999E-04 1050 3.29 84.43 0.800E 0.899E-04 835 2.62 87.05 0.700E 0.799E-04 748 2.35 89.39 0.600E 0.699E-04 643 2.02 91.41 0.500E 0.599E-04 483 1.51 92.93 0.400E 0.499E-04 359 1.13 94.05 0.300E 0.399E-04 381 1.19 95.25 0.200E 0.299E-04 357 1.12 96.37 0.100E 0.199E-04 397 1.24 97.61 0.900E 0.999E-05 55 0.17 97.78 0.800E 0.899E-05 87 0.27 98.06 0.700E 0.799E-05 91 0.29 98.34 0.600E 0.699E-05 130 0.41 98.75 0.500E 0.599E-05 166 0.52 99.27 0.400E 0.499E-05 132 0.41 99.68 0.300E 0.399E-05 84 0.26 99.95 0.200E 0.299E-05 16 0.05 100.00 0.100E 0.199E-05 1 0.00 100.00

    <= 0.999E-06                                                            0            0.00               100.00 TOTALS                                                              31889           100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.147E-02 SEC/M3, 50TH PERCENTILE= 0.234E-03 SEC/M3, AVERAGE= 0.396E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 1 - Auxiliary building vent exhaust and shield building vent.

T234-05.doc

SQN TABLE 2.3.4-6 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT 600 METERS (MINIMUM EXCLUSION AREA BOUNDARY DISTANCE) DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 2* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.800E 0.899E-02 18 0.06 0.06 0.400E 0.499E-02 59 0.19 0.24 0.300E 0.399E-02 50 0.16 0.40 0.200E 0.299E-02 261 0.82 1.22 0.100E 0.199E-02 1715 5.38 6.59 0.900E 0.999E-03 566 1.77 8.37 0.800E 0.899E-03 621 1.95 10.32 0.700E 0.799E-03 842 2.64 12.96 0.600E 0.699E-03 1143 3.58 16.54 0.500E 0.599E-03 1574 4.94 21.48 0.400E 0.499E-03 2424 7.60 29.08 0.300E 0.399E-03 2915 9.14 38.22 0.200E 0.299E-03 4422 13.87 52.09 0.100E 0.199E-03 8359 26.21 78.30 0.900E 0.999E-04 1067 3.35 81.65 0.800E 0.899E-04 1054 3.31 84.95 0.700E 0.799E-04 944 2.96 87.91 0.600E 0.699E-04 707 2.22 90.13 0.500E 0.599E-04 655 2.05 92.18 0.400E 0.499E-04 417 1.31 93.49 0.300E 0.399E-04 391 1.23 94.72 0.200E 0.299E-04 427 1.34 96.05 0.100E 0.199E-04 381 1.19 97.25 0.900E 0.999E-05 64 0.20 97.45 0.800E 0.899E-05 68 0.21 97.66 0.700E 0.799E-05 87 0.27 97.94 0.600E 0.699E-05 102 0.32 98.26 0.500E 0.599E-05 157 0.49 98.75 0.400E 0.499E-05 202 0.63 99.38 0.300E 0.399E-05 137 0.43 99.81 0.200E 0.299E-05 57 0.18 99.99 0.100E 0.199E-05 3 0.01 100.00

    <= 0.999E-06                                                 0                     0.0                  100.00 TOTALS                                                    31889                  100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.130E-02 SEC/M3, 50TH PERCENTILE= 0.215E-03 SEC/M3, AVERAGE= 0.365E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 2 - Radioactive chemical hood exhaust.

T234-06.doc

SQN TABLE 2.3.4-7 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT 509 METERS (MINIMUM EXCLUSION AREA BOUNDARY DISTANCE) DUE TO GROUND-LEVEL RELEASES FROM RELEASE ZONE 3* SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 80.100E 0.199E-01 18 0.06 0.06 0.500E 0.599E-02 59 0.19 0.24 0.400E 0.499E-02 50 0.16 0.40 0.300E 0.399E-02 160 0.50 0.90 0.200E 0.299E-02 429 1.35 2.25 0.100E 0.199E-02 2329 7.30 9.55 0.900E 0.999E-03 421 1.32 10.87 0.800E 0.899E-03 830 2.60 13.47 0.700E 0.799E-03 816 2.56 16.03 0.600E 0.699E-03 1324 4.15 20.18 0.500E 0.599E-03 1914 6.00 26.18 0.400E 0.499E-03 2466 7.73 33.92 0.300E 0.399E-03 3004 9.42 43.34 0.200E 0.299E-03 5067 15.89 59.23 0.100E 0.199E-03 7962 24.97 84.20 0.900E 0.999E-04 821 2.57 86.77 0.800E 0.899E-04 709 2.22 88.99 0.700E 0.799E-04 596 1.87 90.86 0.600E 0.699E-04 533 1.67 92.53 0.500E 0.599E-04 341 1.07 93.60 0.400E 0.499E-04 351 1.10 94.70 0.300E 0.399E-04 339 1.06 95.77 0.200E 0.299E-04 283 0.89 96.65 0.100E 0.199E-04 437 1.37 98.02 0.900E 0.999E-05 74 0.23 98.26 0.800E 0.899E-05 102 0.32 98.58 0.700E 0.799E-05 123 0.39 98.96 0.600E 0.699E-05 126 0.40 99.36 0.500E 0.599E-05 101 0.32 99.67 0.400E 0.499E-05 73 0.23 99.90 0.300E 0.399E-05 28 0.09 99.99 0.200E 0.299E-05 2 0.01 100.00 0.100E 0.199E-05 1 0.00 100.00

    <= 0.999E-06                                                0                     0.0                    100.00 TOTALS                                                   31889                  100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.162E-02 SEC/M3, 50TH PERCENTILE= 0.258E-03 SEC/M3, AVERAGE= 0.435E-03 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND

Release Zone 3 - Condenser air ejector exhaust.

T234-07.doc

SQN TABLE 2.3.4-8 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 1-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT OUTER BOUNDARY OF LOW POPULATION ZONE DUE TO GROUND-LEVEL RELEASES FROM A LOCATION REPRESENTATIVE OF RELEASE ZONE 1, RELEASE ZONE 2, AND RELEASE ZONE 3 SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.100E 0.199E-02 18 0.06 0.06 0.500E 0.599E-03 20 0.06 0.12 0.400E 0.499E-03 62 0.19 0.31 0.300E 0.399E-03 91 0.29 0.60 0.200E 0.299E-03 342 1.07 1.67 0.100E 0.199E-03 1734 5.44 7.11 0.900E 0.999E-04 338 1.06 8.17 0.800E 0.899E-04 575 1.80 9.97 0.700E 0.799E-04 602 1.89 11.86 0.600E 0.699E-04 968 3.04 14.90 0.500E 0.599E-04 1059 3.32 18.22 0.400E 0.499E-04 1754 5.50 23.72 0.300E 0.399E-04 1799 5.64 29.36 0.200E 0.299E-04 2793 8.76 38.12 0.100E 0.199E-04 6560 20.57 58.69 0.900E 0.999E-05 1118 3.51 62.19 0.800E 0.899E-05 1438 4.51 66.70 0.700E 0.799E-05 1413 4.43 71.13 0.600E 0.699E-05 1518 4.76 75.89 0.500E 0.599E-05 1618 5.07 80.97 0.400E 0.499E-05 1485 4.66 85.63 0.300E 0.399E-05 1196 3.75 89.38 0.200E 0.299E-05 887 2.78 92.16 0.100E 0.199E-05 654 2.05 94.21

    <= 0.999E-06                                           1847                        5.79                 100.00 TOTALS                                                31889                       100.00 PERCENT OF THE POSSIBLE 35064 HOURLY OBSERVATIONS WHICH WERE VALID = 90.95 5TH PERCENTILE= 0.139E-03 SEC/M3, 50TH PERCENTILE= 0.142E-04 SEC/M3, AVERAGE= 0.319E-04 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND T234-08.doc

SQN TABLE 2.3.4-9 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 8-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT OUTER BOUNDARY OF LOW POPULATION ZONE DUE TO GROUND-LEVEL RELEASES FROM A LOCATION REPRESENTATIVE OF RELEASE ZONE 1, RELEASE ZONE 2, AND RELEASE ZONE 3 SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.300E 0.399E-03 8 0.03 0.03 0.200E 0.299E-03 32 0.12 0.15 0.100E 0.199E-03 203 0.76 0.91 0.900E 0.999E-04 71 0.27 1.17 0.800E 0.899E-04 126 0.47 1.65 0.700E 0.799E-04 182 0.68 2.23 0.600E 0.699E-04 380 1.42 3.75 0.500E 0.599E-04 545 2.04 5.79 0.400E 0.499E-04 881 3.29 9.08 0.300E 0.399E-04 1723 6.44 15.52 0.200E 0.299E-04 2944 11.01 26.53 0.100E 0.199E-04 6078 22.73 49.27 0.900E 0.999E-05 985 3.68 52.95 0.800E 0.899E-05 1124 4.20 57.15 0.700E 0.799E-05 1377 5.15 62.30 0.600E 0.699E-05 1475 5.52 67.82 0.500E 0.599E-05 1767 6.61 74.43 0.400E 0.499E-05 1926 7.20 81.63 0.300E 0.399E-05 2031 7.60 89.23 0.200E 0.299E-05 1726 6.45 95.68 0.100E 0.199E-05 960 3.59 99.27 0.900E 0.999E-06 39 0.15 99.42 0.800E 0.899E-06 46 0.17 99.59 0.700E 0.799E-06 29 0.11 99.70 0.600E 0.699E-06 29 0.11 99.81 0.500E 0.599E-06 18 0.07 99.87 0.400E 0.499E-06 11 0.04 99.91 0.300E 0.399E-06 11 0.04 99.95 0.200E 0.299E-06 3 0.01 99.97 0.100E 0.199E-06 2 0.01 99.97

    <= 0.999E-06                                             7                       0.03                   100.00 TOTALS                                                26739                     100.00 PERCENT OF THE POSSIBLE 35057 8-HOUR OBSERVATIONS WHICH WERE VALID = 76.27 5TH PERCENTILE= 0.539E-04 SEC/M3, 50TH PERCENTILE= 0.980E-05 SEC/M3, AVERAGE= 0.169E-04 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND T234-09.doc

SQN TABLE 2.3.4-10 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 16-HOUR-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT OUTER BOUNDARY OF LOW POPULATION ZONE DUE TO GROUND-LEVEL RELEASES FROM A LOCATION REPRESENTATIVE OF RELEASE ZONE 1, RELEASE ZONE 2, AND RELEASE ZONE 3 SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.300E 0.399E-04 26 0.09 0.09 0.200E 0.299E-04 61 0.22 0.32 0.100E 0.199E-04 439 1.60 1.92 0.900E 0.999E-05 151 0.55 2.47 0.800E 0.899E-05 272 0.99 3.46 0.700E 0.799E-05 513 1.87 5.33 0.600E 0.699E-05 842 3.07 8.39 0.500E 0.599E-05 1313 4.78 13.18 0.400E 0.499E-05 2167 7.89 21.07 0.300E 0.399E-05 3694 13.46 34.53 0.200E 0.299E-05 6680 24.34 58.86 0.100E 0.199E-05 9097 33.14 92.00 0.900E 0.999E-06 619 2.26 94.26 0.800E 0.899E-06 573 2.09 96.35 0.700E 0.799E-06 388 1.41 97.76 0.600E 0.699E-06 286 1.04 98.80 0.500E 0.599E-06 161 0.59 99.39 0.400E 0.499E-06 99 0.36 99.75 0.300E 0.399E-06 61 0.22 99.97 0.200E 0.299E-06 8 0.03 100.00

    <= 0.999E-07                                            0                          0.0                 100.00 TOTALS                                               27450                        100.00 PERCENT OF THE POSSIBLE 35049 16-HOUR OBSERVATIONS WHICH WERE VALID = 78.32 5TH PERCENTILE= 0.717E-05 SEC/M3, 50TH PERCENTILE= 0.236E-05 SEC/M3, AVERAGE= 0.299E-05 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND T234-10.doc

SQN TABLE 2.3.4-11 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 3-DAY-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT OUTER BOUNDARY OF LOW POPULATION ZONE DUE TO GROUND-LEVEL RELEASES FROM A LOCATION REPRESENTATIVE OF RELEASE ZONE 1, RELEASE ZONE 2, AND RELEASE ZONE 3 SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.100E 0.199E-04 33 0.13 0.13 0.900E 0.999E-05 2 0.01 0.14 0.800E 0.899E-05 65 0.26 0.40 0.700E 0.799E-05 104 0.42 0.82 0.600E 0.699E-05 112 0.45 1.27 0.500E 0.599E-05 366 1.47 2.75 0.400E 0.499E-05 850 3.42 6.17 0.300E 0.399E-05 1883 7.59 13.76 0.200E 0.299E-05 6107 24.61 38.37 0.100E 0.199E-05 12251 49.36 87.73 0.900E 0.999E-06 1157 4.66 92.39 0.800E 0.899E-06 836 3.37 95.76 0.700E 0.799E-06 512 2.06 97.82 0.600E 0.699E-06 229 0.92 98.75 0.500E 0.599E-06 168 0.68 99.42 0.400E 0.499E-06 124 0.50 99.92 0.300E 0.399E-06 19 0.08 100.00

    <= 0.999E-07                                            0                        0.0                   100.00 TOTALS                                               24818                      100.00 PERCENT OF THE POSSIBLE 34993 3-DAY OBSERVATIONS WHICH WERE VALID = 70.92 5TH PERCENTILE= 0.434E-05 SEC/M3, 50TH PERCENTILE= 0.176E-05 SEC/M3, AVERAGE= 0.201E-05 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND T234-11.doc

SQN TABLE 2.3.4-12 ATMOSPHERIC DISPERSION FACTORS FREQUENCY DISTRIBUTION CALCULATED 26-DAY-AVERAGE ATMOSPHERIC DISPERSION FACTORS AT OUTER BOUNDARY OF LOW POPULATION ZONE DUE TO GROUND-LEVEL RELEASES FROM A LOCATION REPRESENTATIVE OF RELEASE ZONE 1, RELEASE ZONE 2, AND RELEASE ZONE 3 SEQUOYAH NUCLEAR PLANT (BASED ON DATA COLLECTED AT THE METEOROLOGICAL STATION FROM JAN 1, 1972 THROUGH DEC 31, 1975) ATMOSPHERIC DISPERSION FACTORS FREQUENCY CUMULATIVE (SEC/M3) (NO. OF OBSERVATIONS) PERCENT PERCENT 0.300E 0.399E-05 354 1.61 1.61 0.200E 0.299E-05 2554 11.60 13.20 0.100E 0.199E-05 17288 78.50 91.71 0.900E 0.999E-06 1390 6.31 98.02 0.800E 0.899E-06 363 1.65 99.67 0.700E 0.799E-06 73 0.33 100.00

    <= 0.999E-07                                           0                            0.0                     100.00 TOTALS                                               22022                         100.00 PERCENT OF THE POSSIBLE 34441 26-DAY OBERSERVATIONS WHICH WERE VALID = 63.94 5TH PERCENTILE = 0.271E-05 SEC/M3, 50TH PERCENTILE= 0.153E-05 SEC/M3, AVERAGE= 0.148E-05 SEC/M3 TEMPERATURE INSTRUMENTS LOCATED 46 AND 9 METERS ABOVE GROUND WIND INSTRUMENTS LOCATED 10 METERS ABOVE GROUND T234-12.doc

SQN Table 2.3.4-13 Sequoyah Nuclear Plant - Fifth Percentile Atmospheric Dispersion Factors (/Q's) for Comparative Data - Hourly-Average and End-of-Hour Temperature Differences (T) (May 1975-April 1976)* Minimum Exclusion Boundary Distance (556 meters) Period Hour-Average T End-of-Hour T 1-hour 0.978 x 10-3 0.985 x 10-3 8-hour 0.392 x 10-3 0.389 x 10-3 Low Population Zone (LPZ) Distance (4828 meters) Period Hour-Average T End-of-Hour T 8-hour 0.494 x 10-4 0.484 x 10-4 16-hour 0.613 x 10-5 0.612 x 10-5 3-day 0.360 x 10-5 0.351 x 10-5 26-day 0.267 x 10-5 0.254 x 10-5

     *Wind direction and wind speed measured at 33 feet above ground. Temperature measured at 33 and 150 feet above ground.

T234-13.doc

SQN TABLE 2.3.4-14 SEQUOYAH NUCLEAR PLANT AVERAGE ANNUAL DISPERSION FACTORS,1 /Q, (s/m3) Downwind Distances (miles) Sector 1 2 3 4 5 10 15 20 30 40 50 N 0.2386E-05 0.8903E-06 0.4990E-06 0.3318E-06 0.2423E-06 0.9330E-07 0.5432E-07 0.3733E-07 0.2231E-07 0.1563E-07 0.1193E-07 NNE 0.3358E-05 0.1246E-05 0.6963E-06 0.4621E-06 0.3370E-06 0.1292E-06 0.7507E-07 0.5151E-07 0.3071E-07 0.2149E-07 0.1638E-07 NE 0.3160E-05 0.1169E-05 0.6523E-06 0.4325E-06 0.3152E-06 0.1207E-06 0.7003E-07 0.4803E-07 0.2861E-07 0.2001E-07 0.1625E-07 ENE 0.1324E-05 0.4874E-06 0.2713E-06 0.1796E-06 0.1309E-06 0.4998E-07 0.2899E-07 0.1988E-07 0.1184E-07 0.8283E-08 0.6314E-08 E 0.6960E-06 0.2585E-06 0.1446E-06 0.9600E-07 0.7007E-07 0.2691E-07 0.1565E-07 0.1075E-07 0.6423E-08 0.4499E-08 0.3434E-08 ESE 0.7180E-06 0.2661E-06 0.1486E-06 0.9861E-07 0.7194E-07 0.2760E-07 0.1605E-07 0.1103E-07 0.6585E-08 0.4613E-08 0.3521E-08 SE 0.8539E-06 0.3141E-06 0.1748E-06 0.1158E-06 0.8432E-07 0.3221E-07 0.1869E-07 0.1282E-07 0.7638E-08 0.5343E-08 0.4073E-08 SSE 0.1301E-05 0.4778E-06 0.2656E-06 0.1757E-06 0.1279E-06 0.4883E-07 0.2832E-07 0.1942E-07 0.1157E-07 0.8098E-08 0.6175E-08 S 0.2338E-05 0.8796E-06 0.4945E-06 0.3294E-06 0.2410E-06 0.9313E-07 0.5434E-07 0.3741E-07 0.2241E-07 0.1573E-07 0.1202E-07 SSW 0.5847E-05 0.2192E-05 0.1231E-05 0.8188E-06 0.5983E-06 0.2304E-06 0.1343E-06 0.9237E-07 0.5521E-07 0.3870E-07 0.2955E-07 SW 0.2629E-05 0.9936E-06 0.5602E-06 0.3736E-06 0.2735E-06 0.1057E-06 0.6163E-07 0.4238E-07 0.2534E-07 0.1776E-07 0.1356E-07 WSW 0.1264E-05 0.4918E-06 0.2811E-06 0.1891E-06 0.1393E-06 0.5467E-07 0.3212E-07 0.2220E-07 0.1336E-07 0.9408E-08 0.7207E-08 W 0.1031E-05 0.4016E-06 0.2296E-06 0.1544E-06 0.1137E-06 0.4464E-07 0.2623E-07 0.1814E-07 0.1092E-07 0.7692E-08 0.5894E-08 WNW 0.6277E-06 0.2446E-06 0.1398E-06 0.9406E-07 0.6927E-07 0.2720E-07 0.1599E-07 0.1105E-07 0.6658E-08 0.4690E-08 0.3594E-08 NW 0.7777E-06 0.2973E-06 0.1684E-06 0.1127E-06 0.8273E-07 0.3221E-07 0.1886E-07 0.1301E-07 0.7811E-08 0.5492E-08 0.4203E-08 NNW 0.1316E-05 0.5079E-06 0.2893E-06 0.1942E-06 0.1428E-06 0.5588E-07 0.3278E-07 0.2264E-07 0.1361E-07 0.9581E-08 0.7337E-08

1. Based on data collected at the meteorological station from January 1, 1972 through December 31, 1975.

T234.14.doc

r SEQUOYAH NUCLEAR PLANT r FINAL SAFETY ANALYSIS REPORT Figure 2.3.1-1 Normal Sea Level Pressure Distribution Over North America and the Nortf, *. Atlantic Ocean

r-n-- I j

I I I

       -                             0 I

I I I

SEQUOYAH SITE l5 r

SEQUOYAH NUCLEAR PLANT FINALSAFETY ANALYSIS REPORT Figure 2.3.1-2 Total Number of Forecast-Days of High Meteorological Potential .f_ar A i r Pollution in a 5 Year Period

                       ))
                    ; "- ***    *Locust Hill SEQUOYAH NUCLEAR PLANT r                .*-*v:    .
    ,:§=>                               ***
                    )     CHICKAMAUGA         LAKE METEOROLOGICAL FACILJTY (Environmental Data Station)

With 300-Foot Meteorological Tower SEQUOYAH NUCLEAR PLANT r FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-1 Environmental Data Station Location

r N FRIENDSHIP SCHOOL SEQ UOYAH SITE METEOROLOGICAL TOWER r SCALE Ol-" MILES 2 0 2 4 6 IBBHHB j CHATTANOOGA r AIRPORT SEQUOYAH NUCLEAR PLANT FINAL SAFE-TY ANALYSIS REPORT Figure 2.3.2-2 Climatological Data Sources

N WINO SPEED (HPHJ

                        >-21.S l2.S-18.1 1.s-12.

s.s-1.* E J.S-S.i

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January 1, 72 - Dec 31, 75

N

             \IINO SPEED     (MPHI 18.S-2f.4 12.5-18.-t 1.s-12. 1 5.5-7.-t E                         J.S-li.4 I .S-3.4 B.0-1,4 Revised by Amendment 1 SE(!UOYAII NUCLEAR PLAN1' F I W1 L SArETY ANALYSIS REPORT F i g u r e 2. 3. 2-4 Wind Rose 10 M Wind All Stability Classes January (72-75)

N WINO SPEED (MPHJ 18,5*2*.* 12.5-18, 4 1.s-12.* s.s-,.* E 1,5-J.f 111.0-1 ** Revised by Amendment 1 SEQUOYAH NUCI.EAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-5 Wind Rose 10 M Wind A11 Stability Classes V I February (72-75) 0

         \./IND SPEED lMPHl 18.S-2f.1 12.5-18,1 7.5-12,f S.5-7.

E S.S-5. 1.5-:S. e.lS-1. Revised by Amendment 1 EQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-6 Wind Rose 10 M Wind All Stability Classes IV Mardi (72-75) 0 w

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l N WINO SPEED (MPHI l8.S-21,i 12.s-,e ...

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I.S-J. 0.6-1.1 Revised by Amendment l SEQUOYAII NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-8 Wind Rose 10 M Wind All Stability Classes IIJ May (72-75) 0

1 WINO SPEED IMPH) 18.S-21.1 12.5-18,1

        "'                    7.5-12,1 5.5-7.1 E                     5.5-5.4 I.5-l.1 0.l'i-1. 1 Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-9    Wind Rose 10 M Wind s All Stability Classes IV June (72-75) 0 ID

N WINO SPEED {MPHt 18.5-2'1.t 12.s-,a.t

       '"                    7.S-12,t S.S-7. 4 E                     I.S-S. 4 1,5-1.4 8.G-1,t Revised by Amendment 1 SEQUOYAH NUCLF.AR PLANT FINAL SAFUTY ANALYSIS REPORT Figure 2.3,2-10     Wind Rose 10 M Wind All Stability Classes
      .July (72-75)

        \./IND SPEED fMPH) 18,5-21.1 12,5-18.4 7,5-12.4 5,5-7.4 E                    5.5-5,4 I .S-5.1 e.0-1.1 Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.3.2-11 Wind Rose 10 M Wind All Stability Classes August (72-75)

WINO SPEED lHPHI 18.S-21. 4 12.S-IB. i 1.s-12.* S.S-7. 4 E S.S-5.1 I.S-l.* 8.8-1. 4 Revised by Amendment ...1 SEQUOYAH NUCLEAR PLANT FINAi. SAl ETY ANALYSIS REPOlt'f Figure 2.3.2-12 Wind Rose 10 M Wind All Stability Classes 2-75 N

N WIND SPEED IMPHl 19.S-2-t. 12.S-18. 7 .5-12, 4 5.S-7.1 E !.S-5.1 I .S-3. 1 e.0-1.1 Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT 11lNAL SAFETY ANALYSIS REPORT ngure 2.3.2-13 Wind Rose 10 M Wtnd All Stability Classes October (72-75)

N WINO SPEED IMPH) 12.S*IB.' BJ S.ii-7.1 E J.S*S.i I .S-3.1 8.6-1, 4 Revised by Amendment l SEQUOYAH NUCLEAR PLANT n N A L SAFETY ANALYSIS REPORT Figure 2.3.2-14 10 M Wind 5 All Stability Classes

 ' Nove her (72-75)

WIND SPEED fMPH> 18.5-2-f.-f 12.S-18.1 7 .S-12. 4 S,S-7,1 E 5.5-S.1 t .S-J. 4 Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT l igure 2.3.2-15 10 M Wind All Stability Classes N Decrnnher (72-75) N

N WIND SPEED (MPH) 18.5-21.1 7.5-12. 11' 5,S-7, E !I. s-s ... I.S-5,i e.e-* .* Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT FINAL SA1"ETY ANALYSIS REPORT Figure 2.J,2-16 10 M Wind, 9 & 4 6 M Temp Stability Class A Jan 1, 72 - Dec 31, 75 N N w

N WINO SPEED IMPH) 18.S-2-t.i 12.S-18. 7.S-12.4 s.s-1.1 E 3.5-5,1 1.5-.5. i Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT PINAi. SAFETY ANALYSIS REPORT Figure 2.3.2-17 Wind Rose 10 M Wind, 9 & 46 M Temp Stability Class B Jan 1, 72 - Dec 31. 75 N N U1

WINO SPEED CHPHI 18.5-21.1 12.s-,a.1 7.S-12,1 s.s-1.1 E J.5-S.4 1.5-5.i 8.6-1.4 Revised by Amendment 1 SEQUOYAH NUCLEAR Pl.ANT F.INAL SAFETY ANALYSIS REPORT Figure 2.3.2-18 Wind Rose 10 M Wind, 9 & L16 M 1'emp Stability Class C s Jan 1 1 72 - Dec 31, 75

N WINO SPEED lHPHl 18.5-21.i 7.5-12.f 11 S.S-7.f E 3.S-5.4 t.S-J.f e.0-1.1 Revised by Amendment 1 SEQUOYAH NUCl,EAR PLANT l INAL SAFETY ANALYSIS REPOltT Figure 2.3.2-19 Wind Rose 10 M Wind, 9 & 4 6 M Temp Stability Class D Jan I, 72 - Dec 31 1 75 N

WIND SPEED (MPH) 18.S-21.1 12.5-18. 7 ,S-12. s.s-1. 1 E 1.s-s.1 1.5-l.1 1.6-1 ... Revised by Amendment l SEQUOYAH NUCLEAR PLANT l*'INAL SAFE1'Y ANALYSIS REPOR'f Figure 2.3.2-20 Wi d Rose 10 M Wind. 9 & lt6 M Temp s Stability Class E Jan 1, 72 - Dec 31. 75

WINO SPEED (HPHl 18.5*21.1 12.5-18.4 7.S-12,4 11 5.5-7.i E I.S-S.4 1.s-s.1 1.a-1.1 Revised by Amendment 1 SEQUOYAH NUCl.l AR PI.AN1' FINAi. SAFETY ANAl,YSIS RIW01t1' Figure 2.3.2-21 Wlnd Rose 10 M Wind, 9 . & 46 M 'l'cmp Stability Class F Jon 1, 72 - Dec 31, 75 w w

N WIND SPEED (MPH> 18.5-21.1 12.S-18.-t 1.s-12. 71 S.S-1. E J.S-S.1 1.s-:s.1 1.6-1.1 Revised by Amendment 1 SEQUOYAH NUCLEAR PLANT FINAL SAl ETY ANALYSIS REPORT Figure 2.3.2-22 Wind Rose 10 M Wind. 9 & 46 M l'emp Stability Class C N J a n 1 . 72 - Dec 31. 75 w 1 1 .,1

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IV U1 w Topography l l i t h i n F i v e M i l e Radius

SQN-17 2.4 HYDROLOGIC ENGINEERING 2.4.1 Hydrologic Description 2.4.1.1 Site and Facilities The location of key plant structures and their relationship to the original site topography are shown on Figure 2.1.2-1. The structures which have safety-related equipment and systems are indicated on this figure and are tabulated below, along with the elevation of major exterior accesses. Number of Structure Access Accesses Elevation Intake pumping (1) Stairwell entrance 1 705.0 structure (2) Access hatches 6 705.0 (3) Cable tunnel 1 690.0 Auxiliary and (1) Railroad access opening 1 706.0 control buildings (2) Doors to turbine building 2 706.0 (3) Doors to turbine building 2 732.0 (4) Doors to turbine building 2 685.0 (5) Personnel lock to SB 1 690.0 (6) General vent or intake 2 714 (7) Doors to AEB and MSVV 4 714 Shield building (1) Personnel lock (watertight) 1 691.0 (2) Equipment hatch 1 730.0 (3) Personnel lock 1 732.0 Diesel generator (1) Equipment access door 4 722.0 building (2) Personnel access door 1 722.0 (3) Emergency exit 4 722.0 (4) Emergency exit 1 740.5 ERCW intake (1) Access door 1 725.0 pumping station (2) Trash sluice 1 723.5 (3) Deck drainage (sealed for flood) 1 720.0 Exterior accesses are also provided to each of the class IE electrical systems manholes and handholes at elevations varying from 700 to 724 feet MSL, depending upon the location of each structure. The relationship of the plant site to the surrounding area can be seen in Figures 2.1.2-1 and 2.4.1-1. It can be seen from these figures that significant natural drainage features of the site have not been altered. Local surface runoff drains into the Tennessee River. 2.4.1.2 Hydrosphere The Sequoyah Nuclear Plant (SQN) site comprises approximately 525 acres on a peninsula on the western shore of Chickamauga Lake at Tennessee River Mile (TRM) 484.5. As shown by Figure 2.4.1-1, the site is on high ground with the Tennessee River being the only potential source of flooding. S2-4.doc 2.4-1

SQN-17 The Tennessee River above SQN site drains 20,650 square miles. The drainage area at Chickamauga Dam, 13.5 miles downstream, is 20,790 square miles. Three major tributaries--Hiwassee, Little Tennessee, and French Broad 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 Tennessee River and its Clinch and Holston River tributaries 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. The drainage pattern is shown on Figure 2.1.1-1. About 20 percent of the watershed rises above elevation 3000 with a maximum elevation of 6,684 at Mt. Mitchell, North Carolina. The watershed is about 70 percent forested with much of the mountainous area being 100 percent forested. The climate of the watershed is humid temperate. Mean annual precipitation for the Tennessee Valley is shown by Figure 2.4.1-2. Above Chickamauga Dam, annual rainfall averages 51 inches and varies from a low of 40 inches at sheltered locations in the mountains to high spots of 85 inches on the southern and eastern divide. Rainfall occurs relatively evenly throughout the year. See Section 2.3 for a discussion of rainfall. Major flood-producing storms are of two general types; the cool-season, winter type, and the warm-season, hurricane type. Most floods at SQN, however, have been produced by winter-type storms in the months of January through early April. Watershed snowfall is relatively light, averaging only about 14 inches annually above the plant. The maximum average annual snowfall of 63 inches occurs at Mt. Mitchell, the highest point east of the Mississippi River. The overall snowfall average above the 3,000-foot elevation, however, is only 22 inches annually. Individual snowfalls are normally light, with an average of 13 snowfalls per year. Snowmelt is not a factor in maximum flood determinations. Chickamauga Dam, 13.5 miles downstream, affects water surface elevations at SQN. Normal full pool elevation is 683.0 feet. At this elevation the reservoir is 58.9 miles long on the Tennessee River and 32 miles long on the Hiwassee River, covering an area of 35,400 acres, with a volume of 628,000 acre-feet. The reservoir has an average width of nearly 1 mile, ranging from 700 feet to 1.7 miles. At SQN, the reservoir is about 3,000 feet wide with depths ranging between 12 feet and 50 feet at normal pool elevation. The Tennessee River above Chattanooga, Tennessee, is one of the best 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, 20 miles downstream from SQN. There are 20 major reservoirs in the TVA system upstream from the plant, 13 of which have substantial reserved flood detention capacity during the main flood season. Table 2.4.1-1 lists pertinent data for TVA's major dams prior to modifications made by the Dam Safety Program (see Table 2.4.1-5). In addition, there are six major dams owned by the Aluminum Company of America (ALCOA). The ALCOA reservoirs often contribute to flood reduction but were ignored in this analysis because they do not have dependable reserved flood detention capacity. The locations of these dams and the minor dams, Nolichucky and Walters (Waterville Lake), are shown on Figure 2.1.1-1. Table 2.4.1-2 lists pertinent data for the major and minor ALCOA dams and Walters Dam. The flood detention capacity reserved in the TVA system varies seasonally, with the greatest amounts during the flood season. Figure 2.4.1-3, containing 14 sheets, shows tributary and main river reservoir seasonal operating guides for those reservoirs having major influence on SQN flood S2-4.doc 2.4-2

SQN-17 flows. Table 2.4.1-3 shows the flood control reservations at the multiple-purpose projects above SQN at the beginning and end of the winter flood season and in the summer. Assured system detention capacity above the plant varies from 5.6 inches on January 1 to 4.5 inches on March 15, decreasing to 1.0 inch during the summer and fall. Actual detention capacity may exceed these amounts, depending upon inflows and power demands. Flood control above SQN is provided largely by 11 tributary reservoirs. Tellico Dam is counted as a tributary reservoir because it is located on the Little Tennessee River, although, because of canal connection with Fort Loudoun Dam, it also functions as a main river dam. On March 15, near the end of the flood season, these provide a minimum of 4,436,000 acre-feet of detention capacity, equivalent to 5.8 inches on the 14,476 square-mile area they control. This is 90 percent of the total available above Chickamauga Reservoir. The two main river reservoirs, Fort Loudoun and Watts Bar, provide 490,000 acre-feet, equivalent to 1.5 inches of detention capacity on the remaining area above the plant. Daily flow volumes at the plant, for all practical purposes, are represented by discharges from Chickamauga Dam with drainage area of 20,790 square miles, only 140 square miles more than at the plant. Momentary flows at the nuclear plant may vary considerably from daily averages, depending upon turbine operations at Watts Bar Dam upstream and Chickamauga Dam downstream. There may be periods of several hours when there are no releases from either or both Watts Bar and Chickamauga Dams. Rapid turbine shutdown at Chickamauga may sometimes cause periods of up-stream flow in Chickamauga Reservoir. Based upon discharge records since closure of Chickamauga Dam in 1940, the average daily streamflow at the plant is 32,600 cfs. The maximum daily discharge was 223,200 cfs on May 8, 1984. Except for two special operations on March 30 and 31, 1968, when discharge was zero to control milfoil, the minimum daily discharge was 700 cfs on November 1, 1953. Flow data for water years 1951-1972 indicate an average rate of about 27,600 cfs during the summer months (May-October) and about 38,500 cfs during the winter months (November-April). Flow durations based upon Chickamauga Dam discharge records for the period 1951-1972 are tabulated below. Average Daily Percent of Time Discharge, cfs Equaled or Exceeded 5,000 99.6 10,000 97.7 15,000 93.3 20,000 84.0 25,000 69.3 30,000 46.8 35,000 31.7 Channel velocities at SQN average about 0.6 fps under normal winter conditions. Because of lower flows and higher reservoir elevations in the summer months, channel velocities average about 0.3 fps. As listed on Table 2.4.1-4, there are 23 surface water users within the 98.6-mile reach of the Tennessee River between Dayton, TN and Stevenson, AL. These include fifteen industrial water supplies and eight public water supplies. The industrial users exclusive of SQN withdraw about 497 million gallons per day from the Tennessee River. Most of this water is returned to the river after use with varying degrees of contamination. S2-4.doc 2.4-3

SQN-17 The public surface water supply intake (Savannah Valley Utility District), originally located across Chickamauga Reservoir from the plant site at TRM 483.6, has been removed. Savannah Valley Utility District has been converted to a ground water supply. The nearest public downstream intake is the East Side Utility (formerly referred to as U.S. Army, Volunteer Army Ammunition Plant). This intake is located at TRM 473.0. Groundwater resources in the immediate SQN site are described in Section 2.4.13. 2.4.1.3 TVA Dam Safety Program Most of the dams upstream from SQN were designed and built before the hydrometerological approach to spillway design had gained its current level of acceptance. Spillway design capacity was generally less than would be provided today. The original FSAR analyses were based on the existing dam system before dam safety modifications were made and included failure of some upstream dams from overtopping. In 1982, TVA officially began a safety review of its dams. The TVA Dam Safety Program was designed to be consistent with Federal Guidelines for Dam Safety and similar efforts by other Federal agencies. Technical studies and engineering analyses were conducted and physical modifications implemented to ensure the hydrologic and seismic integrity of the TVA dams and demonstrate that TVAs dams can be operated in accordance with Federal Emergency Management Agency (FEMA) guidelines. Table 2.4.1-5 provides the status of TVA Dam Safety hydrologic modifications as of 1998. These modifications enable these projects to safely pass the probable maximum flood. The remaining hydrologic modifications planned for Bear Creek Dam and Chickamauga Dam will not affect SQN in any manner which might invalidate the reanalysis described below. In 1997-98, TVA reanalyzed the nuclear plant design basis flood events. The purpose of the reanalysis was to evaluate the effects of the hydrologic dam safety modifications on the flood elevations and response times in the SQN FSAR and to confirm the adequacy of the plant flood plans. The following methods and assumptions were applied to the reanalysis:

1. The computer programs and modeling methods were the same as previously used and documented in the FSAR.

2. Probable maximum precipitation, time distribution of precipitation, precipitation losses and reservoir operating procedures were unchanged from the original analysis.

3. The original stability analyses and postulated seismic dam failure assumptions were conservatively assumed to occur in the same manner and in combination with the same previously postulated rainfall events. No credit was taken for the 1988 post-tensioning of Fontana and Melton Hill Dams to prevent seismic failure. Nor was any credit taken for Dam Safety seismic evaluations of Norris, Cherokee, Douglas, Fort Loudon, Tellico, Hiwassee, Apalachia, and Blue Ridge Dams which demonstrated their structural integrity for a seismic event with a return period of approximately 10,000 years.

4. The planned modification of Chickamauga Dam (armoring the embankment to permit overtopping) was conservatively assumed to have been implemented for the purpose of calculating flood effects. Under present existing conditions, the Chickamauga embankment would be severely eroded in the overtopping PMF event and the maximum flood elevation at SQN would be lower than that with the planned modification.

S2-4.doc 2.4-4

SQN-17 2.4.2 Floods 2.4.2.1 Flood History (Historical) The nearest location with extensive formal flood records is 20 miles downstream at Chattanooga, Tennessee, where continuous records are available since 1874. Knowledge about significant floods extends back to 1826, based upon newspaper and historical reports. Flood flows and stages at Chattanooga have been altered by TVA's reservoir system beginning with the closure of Norris Dam in 1936 and reaching essentially the present level of control in 1952 with closure of Boone Dam, the last major dam with reserved flood detention capacity constructed above Chattanooga. Tellico Dam provides additional reserved flood detention capacity; however, the percentage increase in total detention capacity above the Watts Bar site is small. Thus, for practical purposes, flood records for the period 1952 to date can be considered representative of prevailing conditions. Figure 2.4.2-1 shows the known flood experience at Chattanooga in diagram form. The maximum known flood under natural conditions occurred in 1867. This flood reached elevation 690.5 at SQN. The maximum flood under present-day regulation reached elevation 687.9 at the site on May 9, 1984. The following table lists the highest floods at SQN: Elevation, Discharge, Date Feet cfs Before Regulation March 11, 1867 690.5 450,000 March 1, 1875 686.2 405,000 April 3, 1886 684.5 385,000 March 7, 1917 680.0 335,000 April 5, 1920 676.5 270,000 Since Present Regulation February 3, 1957 683.7 180,000 March 13, 1963 684.8 205,000 March 18, 1973 687.0 219,000 May 9, 1984 687.9 250,000 2.4.2.2 Flood Design Considerations TVA has planned the SQN project to conform with regulatory position 2 of Regulatory Guide 1.59. The types of events evaluated to determine the worst potential flood included (1) Probable Maximum Precipitation (PMP) on the total watershed and critical subwatersheds, including seasonal variations and potential consequent dam failures and (2) dam failures in a postulated Safe Shutdown Earthquake (SSE) or one-half SSE with guide specified concurrent flood conditions. The computed maximum stillwater flood level in the reservoir at the plant site from any cause is elevation 719.6. Maximum level including wave height is 722.4. This elevation would result from the probable maximum precipitation critically centered on the watershed and a 45-mile-per-hour overwater wind, from the most critical direction coincident with the peak of the resulting flood. S2-4.doc 2.4-5

SQN-27 Other rainfall floods will also exceed plant grade, elevation 705, and will necessitate plant shutdown. Flood warning criteria and forecasting techniques have been developed to assure that there will always be adequate time to shut the plant down and be ready for floodwaters above plant grade and are described in Subsections 2.4.10 and 2.4.14, and Appendix 2.4A. Seismic and concurrent flood events could create flood levels which would exceed plant grade. The maximum elevation reached in such an event is elevation 707.9, 2.9 feet above plant grade and 11.7 feet below the controlling event probable maximum flood (PMF), excluding wind-wave considerations. In all such events there is adequate time for safe plant shutdown after the seismic event and before plant grade would be crossed. The emergency protective measures and warning criteria are described in Subsections 2.4.10 and 2.4.14, and Appendix 2.4A. Most safety-related building accesses are located at elevation 706 or above. The accesses below elevation 706 are within the powerhouse and will not be exposed to floodwater until plant grade is exceeded. Therefore, the structures are protected from flooding prior to the end of the shutdown period. Drainage to the Tennessee River has been provided to accommodate runoff from the probable maximum precipitation on the local area of the plant site. Specific analysis of Tennessee River flood levels resulting from oceanfront surges and tsunamis is not required because of the inland location of the plant. Snowmelt and ice jam considerations are also unnecessary because of the temperate zone location of the plant. Flood waves from landslides into upstream reservoirs required no specific analysis, in part because of the absence of major elevation relief in nearby upstream reservoirs and because the prevailing thin soils offer small slide volume potential compared to the available detention space in reservoirs. All safety-related facilities, systems, and equipment are housed in structures which provide protection from flooding for all flood conditions up to plant grade at elevation 705. For the condition where flooding exceeds plant grade, as described in Subsections 2.4.3 and 2.4.4, all equipment required to maintain the plant safely during the flood, and for 100 days after the beginning of the flood, is either designed to operate submerged, located above the maximum flood level, or otherwise protected. Safety-related facilities, systems, and equipment located in the containment structure are protected from flooding by the shield building. All accesses and penetrations below the maximum flood level in the shield building are designed and constructed as essentially water tight elements. The turbine, control, and auxiliary building will be allowed to flood. Wind wave run-up during the PMF at the diesel generator building reaches elevation 721.8 which is 0.2 feet below the operating floor. Consequently, wind wave run-up will not impair the safety function of systems in the diesel generator building. The accesses and penetrations below this elevation in the diesel generator building are designed and constructed to minimize leakage into the buildings. Redundant sump pumps are provided within the building to remove minor leakage. Protective measures are taken to ensure that all safety-related systems and equipment in the Emergency Raw Cooling Water (ERCW) pump station will remain functional when subjected to the maximum flood level. S2-4.doc 2.4-6

SQN-17 Class IE electrical cables, located below the Probable Maximum Flood (PMF) plus wind-wave activity and required in a flood, are designed for submerged operation. Structures housing safety-related facilities, systems, and equipment are protected from flooding during a local PMF by the slope of the plant yard. The yard is graded so that the surface runoff will be carried to Chickamauga Reservoir without exceeding the elevation of the external accesses given in Paragraph 2.4.1.1 except those at the intake pumping station whose pumps can operate submerged. 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers The guidance of Appendix A of Regulatory Guide 1.59 was followed in determining the PMF. Plant surface drainage was evaluated and found capable of passing the local probable maximum storm without reaching or exceeding the critical floor elevation 706, as further described in 2.4.3.5. Evaluation of seasonal and areal variations of probable maximum storms showed that the probable maximum Tennessee River flood level at the plant would be caused by a sequence of storms occurring in March centered in the mountains, east of the plant. The flood crest at the plant would be augmented by the failure of the west saddle dike at Watts Bar Dam upstream. The estimated maximum discharge is 1,236,000 cfs. The probable maximum elevation at the plant is 719.6, excluding any wind-wave effects, and excluding any lower flood level due to failure of Chickamauga Dam downstream. 2.4.3.1 Probable Maximum Precipitation Probable maximum precipitation (PMP) for the Tennessee River watershed above SQN has been defined for TVA by the Hydrometeorological Branch of the National Weather Service in Hydrometeorological Report No. 41 Reference [1]. Two basic storm positions were evaluated. One would produce maximum rainfall over the total watershed. The other would produce maximum rains in the part of the basin downstream from major TVA tributary reservoirs, hereafter referred to as the 7,980-square-mile storm. Snowmelt is not a factor in generating maximum floods at the plant site. Controlling PMP depths for 21,400-square-mile and 7,980-square-mile areas are tabulated below. These storms would occur in March. Depths for other months would be less. Depth, Inches 72-Hour Main Storm Sq. Miles Antecedent Storm 6-Hour 24-Hour 72-Hour 21,400 6.7 5.03 11.18 16.78 7,980 8.1 7.02 14.04 20.36 Two possible isohyetal patterns producing the total area depths are presented in Report No. 41. The one critical to this study is the "downstream pattern" shown in Figure 2.4.3-1. The isohyetal pattern for the 7,980-square-mile storm is shown in Figure 2.4.3-2. The pattern is not orographically fixed and can be moved parallel to the long axis northeast and southwest along the Valley. A 72-hour storm three 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. S2-4.doc 2.4-7

SQN-26 Potential storm amounts differing by seasons were analyzed in sufficient number to make certain that the March storms would be controlling. Enough centerings were investigated to assure that a most critical position was used. Storms producing PMP above upstream tributary dams, whose failure has the potential to create maximum flood levels, were evaluated in the original FSAR analysis. Dam safety modifications at upstream tributary dams have eliminated these potential failures and subsequent plant site flood levels. A standard time distribution pattern was adopted for all storms based upon major observed storms transposable to the Tennessee Valley and in conformance with the usual practice of Federal agencies. The adopted distribution is shown on Figure 2.4.3-3. The critical probable maximum storm was determined to be a total basin storm with downstream orographically fixed pattern (Figure 2.4.3-1) which would follow an antecedent storm commencing on March 15. Translation of the PMP from Report No. 41 to the basin results in an antecedent storm producing an average precipitation of 6.4 inches in three days, followed by a three-day dry period, and then by the main storm producing an average precipitation of 16.5 inches in three days. Figure 2.4.3-4 is an isohyetal map of the maximum three-day PMP. Basin rainfall depths are given in Table 2.4.3-1. To evaluate the local plant drainage system for a PMP event, Hydrometeorological Report No. 56 was used to calculate a 1-hr storm totaling 16.21 inches. Three different temporal distributions were applied to the model, with peak intensity of 2.81 inches/5-min = 33.72 in/hr shifted between early, middle, and late occurrence. Depths for each 5-minute increment of the controlling late peak distribution were 0.68, 0.78, 0.87, 0.97, 0.97, 1.17, 1.26, 1.36, 1.55, 2.81, 2.04, and 1.75. Rainfall on plant building roofs was assumed to discharge to the ground surface. 2.4.3.2 Precipitation Losses Precipitation losses in the probable maximum storm are estimated with multivariable relationships used in the day-to-day operation of the TVA system. These relationships, developed from a study of storm and flood records, relate the amount of precipitation excess (and hence the precipitation loss) to the week of the year, an antecedent precipitation index (API), and geographic location. The relationships are such that the loss subtraction from rainfall to compute precipitation excess is greatest at the start of the storm and decreases to no subtraction when the storm rainfall totals from 7 to 16 inches. Precipitation losses become zero in the late part of extreme storms. For this probable maximum flood analysis, median moisture conditions as determined from past records were used to determine the API at the start of the storm sequence. The antecedent storm is so large, however, that the precipitation excess computed for the later main storm is not sensitive to variations in adopted initial moisture conditions. The precipitation loss in the critical probable maximum storm totals 4.13 inches, 2.30 inches in the antecedent storm amounting to 36 percent of the 3-day 6.44-inch rainfall, and 1.83 inches in the main storm amounting to 11 percent of the 3-day, 16.46 inch rainfall. Table 2.4.3-1 displays the API, rain, and precipitation excess for each of the 45 subwatersheds of the hydrologic model for the SQN probable maximum flood. No precipitation loss was applied in the probable maximum storm on the local area used to test the adequacy of the site drainage system and roofs of safety-related structures. Runoff was made equal to rainfall. S2-4.doc 2.4-8

SQN-17 2.4.3.3 Runoff Model The runoff model used to determine Tennessee River flood hydrographs at SQN is divided into 45 unit areas. Unit hydrographs are used to compute flows from these areas. The unit area flows are combined with appropriate time sequencing or channel routing procedures to compute inflows into the most upstream reservoirs, which in turn are routed through the reservoirs, using standard techniques. Resulting outflows are combined with additional local inflows and carried downstream using appropriate time sequencing or routing procedures, including unsteady flow routing. Figure 2.4.3-5 shows unit areas of the watershed upstream from SQN. The runoff model used in this updated FSAR differs from that used previously because of refinements made in some elements of the model during PMF studies for other nuclear plants and those made from information gained from the 1973 flood, the largest that has occurred during present reservoir conditions. Changes are identified when appropriate in the text. They include both additional and revised unit hydrographs and additional and revised unsteady flow stream course models. Unit hydrographs were developed for each unit area from maximum flood hydrographs either recorded at stream gauging stations or estimated from reservoir headwater elevation, inflow, and discharge data. The number of unit areas has been increased from 34 used previously to 45. The differences include:

1. Use of the model developed for the Phipps Bend study which combined the two unit areas for Watauga River (Sugar Grove and Watauga local) into one unit area and divided the Cherokee to Gate City unit area into two unit areas (Surgoinsville local and Cherokee local below Surgoinsville);

2. Use of the model developed for the Clinch River Breeder Reactor which increased the unit areas on the Clinch River from 3 to 11 and the Watts Bar local from 1 to 2;

3. Changes to add an unsteady flow model for the Fort Loudoun-Tellico Dam complex which included dividing the lower Little Tennessee River unit area into two unit areas (Fontana to Chilhowee and Chilhowee to Tellico), and the Fort Loudoun local unit area into three unit areas (French Broad River local, Holston River local and Fort Loudoun local); and

4. Combining the two unit areas above Ocoee No. 1 (Ocoee No. 1 and Ocoee No. 3) into one unit area (Ocoee No. 1 to Blue Ridge).

In addition, eight of the unit graphs have been revised. Figure 2.4.3-6, which contains 11 sheets, shows the unit hydrographs. Table 2.4.3-2 contains essential dimension data for each unit hydrograph and identification of those hydrographs which are new or revised. Tributary reservoir routings, except for Tellico, were made using the Goodrich semigraphical method and flat pool storage conditions. Main river reservoir and Tellico routings were made using unsteady flow techniques. This differs from the previous submission in that:

1. An unsteady flow model has been added for the Fort Loudoun-Tellico complex, and

2. The Chickamauga unsteady flow model has been revised using the 1973 flood data and results from the HEC-2 backwater computer program.

S2-4.doc 2.4-9

SQN-17 In the original study, the failure wave hydrograph of the mouth of the Hiwassee River was approximated for the postulated failures of Hiwassee, Apalachia and Blue Ridge dams as described in section 2.4.4.2.1. In the 1998 reassessment, an unsteady flow model developed during the dam safety studies was used as an adjunct to route the Hiwassee, Apalachia and Blue Ridge failures in the one half SSE. The model was verified by comparing model elevations in a state of steady flow with elevations computed by the standard-step method. This was done for steady flows ranging from 25,000 cfs to 1,000,000 cfs. Unsteady flow routings were computer-solved with a mathematical model based on the equations of unsteady flow, [3]. Boundary conditions prescribed were inflow hydrographs at the upstream boundary, local inflow, and headwater discharge relationships at the downstream boundary based upon normal operating rules, or based upon rated curves when geometry controlled. The unsteady flow mathematical model for the 49.9-mile-long Fort Loudoun Reservoir was divided into twenty-four 2.08-mile reaches. The model was verified at three gauged points within Fort Loudoun Reservoir using 1963 and 1973 flood data. The unsteady flow model was extended upstream on the French Broad and Holston Rivers to Douglas and Cherokee Dams, respectively. The French Broad and Holston River unsteady flow models were verified at one gaged point each at mile 7.4 and 5.5, respectively, using 1963 and 1973 flood data. The Little Tennessee River was modeled from Tellico Dam, mile 0.3, through Tellico Reservoir to Chilhowee Dam at mile 33.6, and upstream to Fontana Dam at mile 61.0. The model for Tellico Reservoir to Chilhowee Dam was tested for adequacy by comparing its results with steady-state profiles at 1,000,000 and 2,000,000 cfs computed by the standard-step method. Minor decreases in conveyance in the unsteady flow model yielded good agreement. The average conveyance correction found necessary in the reach below Chilhowee Dam to make the unsteady flow model agree with the standard-step method was also used in the river reach from Chilhowee to Fontana Dam. The Fort Loudoun and Tellico unsteady flow models were joined by a canal unsteady flow model. The canal was modeled with five equally-spaced cross Sections at 525-foot intervals for the 2,100-foot-long canal. The unsteady flow routing model for the 72.4-mile-long Watts Bar Reservoir was divided into thirty-four 2.13-mile reaches. The model was verified at two gauged points within the reservoir using 1963 flood data. The unsteady flow mathematical model for the total 58.9-mile-long Chickamauga Reservoir was divided into twenty-eight 2.1-mile reaches providing twenty-nine equally-spaced grid points. The grid point at mile 483.62 is nearest to the plant, mile 484.5. The unsteady flow model was verified at four gauged points within Chickamauga Reservoir using 1973 flood data. This differs from the previous submission in that the 1973 flood was added for verification, replacing the 1963 flood. The 1973 flood occurred during preparation of the FSAR and therefore, was not available for verification. The 1973 flood is the largest which has occurred since closure of South Holston Dam in 1950. Comparisons between observed and computed stages in Chickamauga Reservoir are shown in Figure 2.4.3-7. It is impossible to verify the models with actual data approaching the magnitude of the probable maximum flood. The best remaining alternative was to compare the model elevations in a state of steady flow with elevations computed by the standard step method. This was done for steady flows ranging up to 1,500,000 cfs. An example shown by the rating curve of Figure 2.4.3-8 shows the good agreement. S2-4.doc 2.4-10

SQN-17 The watershed runoff model was verified by using it to reproduce the March 1963 and March 1973 floods; the largest recorded since closure of South Holston Dam. This differs from the previous submission in that the 1973 flood was added for verification, replacing the 1957 flood. Observed volumes of precipitation excess were used in verification. Comparisons between observed and computed outflows from Watts Bar and Chickamauga Dams for the 1973 and 1963 floods are shown in Figures 2.4.3-9 and 2.4.3-10, respectively. From a study of the basic units of the predicting system and its response to alterations in various basic elements, it is concluded that the model serves adequately and conservatively to determine maximum flood levels. 2.4.3.4 Probable Maximum Flood Flow The probable maximum flood discharge at SQN was determined to be 1,236,000 cfs. The hydrograph of this flood is shown in Figure 2.4.3-11. This flood would result from the total basin downstream orographically fixed storm pattern, Figure 2.4.3-4, more completely described in Section 2.4.3.1. The dam safety modification to Fort Loudon, Tellico, and Watts Bar Dams enable them to safely pass the PMF. The west saddle dike at Watts Bar Dam would be overtopped and breached. Chickamauga would be overtopped but was assumed not to fail as a failure would reduce the flood level at the site. In the original FSAR analysis, the flood would overtop and breach the earth embarkments of Fort Loudon, Tellico, and Watts Bar Dams upstream. A second candidate storm is the 7,980-square-mile storm centered at Bulls Gap, Tennessee, 50 miles northeast of Knoxville, shown in Figure 2.4.3-2. The flood from this storm would overtop and breach the west saddle dike at Watts Bar Dam. The flood from the 7,980-square-mile storm is the less critical storm and would produce a probable maximum discharge less than from the total basin storm. The previous PMF evaluations considered candidate situations involving upstream tributary dams Douglas and Watauga. These two situations were shown at that time to be non-governing. Dam safety modifications have since eliminated the potential failures of these dams. Therefore, these two candidate situations have been eliminated. Reservoir routings started at median observed elevations for the mid-March large area PMP storms. Median levels were reevaluated using operating experience for:

1. The total project period, or

2. The five-year period, 1972-1976, for those projects whose operating guides were changed in 1971.

Because of the wet years of 1972-1975 and the operating guide changes, median elevations were higher for 8 of the 13 tributary reservoirs where routing is involved. Normal reservoir operating procedures were used in the antecedent storm. These used turbine and sluice discharge in the tributary reservoirs. Turbine discharges are not used in the main river reservoirs after large flood flows develop because head differentials are too small. Normal operating procedures were used in the principal storm, except that turbine discharge was not used in either the tributary or main river dams. S2-4.doc 2.4-11

SQN-26 All gates were determined to be operable without failures during the flood. Gates on main river dams would be fully raised, thus requiring no additional operations by the last day of the storm, which is before the structures and access roads would be inundated. Median initial reservoir elevations were used at the start of the storm sequence used to define the PMF to be consistent with statistical experience and to avoid unreasonable combinations of extreme events. As a result, 53 percent of the total reserved system flood detention capacity was occupied at the start of the main flood. This is considered to be amply conservative. The statement made in the PSAR and subsequent versions of the FSAR that 67 percent of the reserved system detention capacity was occupied at the start of the main storm was in error. The correct percentage was 33. The remaining reserved system detention capacity was 67 percent. This erroneous statement was first made in the PSAR and was copied in subsequent statements where the routings were the same. In the revised analysis submitted in Amendment 51, all reservoirs are higher or about the same elevation at the beginning of the main storm as a result of the revised starting levels explained in Section 2.4.3.4 of the FSAR. This conservative change results in 53 percent of the total reservoir system detention capacity being occupied at the start of the main flood rather than 33 percent in previous studies. Neither the initial reservoir levels nor the operating rules would have significant effect on maximum flood discharges and elevations at the plant site because spillway capacities, and hence, uncontrolled conditions, were reached early in the flood. The procedures used to determine if and when an overtopped earth embankment would fail and the procedures for computing the effect of such failures are described in 2.4.4.2 and 2.4.4.3. In testing the adequacy of the yard drainage system, to safely pass the site PMP, all underground drains were assumed clogged and the surface drainage to be full. 2.4.3.5 Water Level Determinations The elevation hydrograph of the controlling PMF, cresting at elevation 719.6, is shown on Figure 2.4.3-12. Computation of both the probable maximum discharge hydrograph (Figure 2.4.3-11) and the corresponding elevation hydrograph was accomplished concurrently using the unsteady flow techniques described in Section 2.4.3.3. The less critical total area storm-producing PMP depths on the 7,980-square-mile watershed would produce crest elevation 718.9 at the plant site. Maximum water levels at buildings expected to result from the local plant PMP were determined using a transient flow (unsteady flow) model with hydraulically connected storage areas. Much of the plant site is flat, particularly at the switchyards, and a single flow path is not well defined. A transient model with interconnected storage areas very roughly approximates a two-dimensional model using one-dimensional methods by providing multiple simultaneous outlet paths for the exterior areas adjacent to plant buildings. The separate watershed subareas and flowpaths are shown on Figure 2.4.3-13a. S2-4.doc 2.4-12

SQN-26 The western plant site was evaluated as six interconnected storage areas with four primary weir-flow outlets and one connected transient flow stream-course model. Runoff from the western plant site will flow: Northwest to a channel along the main plant tracks and then across the main access highway (Area 7); to the West through a parking lot (Areas 6A, 6C, and 6E connected to transient flow model); Southwest through the vehicle barrier system directly to Chickamauga Lake (Area 6E); or South through the vehicle barrier system to the Yard Drainage and other Ponds (Area 6C). The maximum water surface elevations in Areas 6A and 6BS are below critical floor elevation 706. The eastern plant site was evaluated as three interconnected storage areas with three weir-flow outlets and two connected transient flow stream-course models. Runoff from the eastern plant site will flow: North around the West and East ends of the multipurpose building to the intake channel (Area 5 connected to two transient flow models); South to the Condenser Circulating Water Discharge Channel (Areas 4 and 6D); or Southwest into the western plant site (Area 6D into 6C). The maximum water surface elevations in Areas 4, 5, and 6D are below critical floor elevation 706. Underground drains were assumed clogged throughout the storm. For fence sections, the Mannings n value was doubled to account for increased resistance to flow and the potential for debris blockage. The only stream adjacent to SQN is the Tennessee River. There are no streams within the site. The 1 percent-chance floodplain of the Tennessee River at the site is delineated on Figure 2.4.3-14. Details of the analyses used in the computation of the 1-percent-chance flood flow and water elevation are described in a study made by TVA for the Federal Insurance Administration (FIA) and published in February 1979 [5]. The only structures located in the 1-percent-chance floodplain are transmission towers, the intake pumping station skimmer wall, and the ERCW pump station deck. The ERCW pumps are located on the pump station deck at elevation 720.5, well above the 1-percent-chance flood level. These structures are shown on Figure 2.4.3-14. The structures that are located in the floodplain will not alter flood flows or elevations. The 20,650-square-mile drainage area is not altered and the reduction in flow area at the site is infinitesimal and at the fringe of the flooded area. The site will be well maintained and any debris generated from it will be minimal and will present no problem to downstream facilities. 2.4.3.6 Coincident Wind-Wave Activity Some wind waves are likely when the probable maximum flood crests at SQN. The flood would be near its crest for a day beginning about 2-1/2 days after cessation of the probable maximum storm. The day of occurrence would most likely be in the month of March or possibly the first week in April. A conservatively high velocity of 45 miles per hour over water was adopted to associate with the probable maximum flood crest. A 45-mile- per-hour overwater velocity exceeds maximum March one-hour velocities observed in severe March windstorms of record in a homogeneous region as reported by the Corps of Engineers [6]. That a 45-mile-per-hour overwater wind is conservatively high, is supported also by an analysis of March day maximum winds of record collected at Knoxville and Chattanooga, Tennessee. The records analyzed varied from 30 years at Chattanooga to 26 years at Knoxville, providing samples ranging from 930 to 806 March days. The recorded fastest mile wind on each March day was used S2-4.doc 2.4-13

SQN-17 rather than hourly data because this information is readily available in National Weather Service publications. Relationships to convert fastest mile winds to winds of other durations were developed from Knoxville and Chattanooga wind data contained in USWB Form 1001 and the maximum storm information contained in Technical Bulletin No. 2 [6]. From the wind frequency analysis it was determined that the 45-mile-per-hour overwater wind for the critical minimum duration of 20 minutes had an 0.1 percent chance of occurrence on any given March day. The probability that this wind might occur on the specific day that the probable maximum flood would crest is extremely remote. Even assuming that the flood was to crest once during the 40-year plant life,

                                                                                     -6 the probability of the wind occurring on that particular day is in the order of 1 x 10 .

TVA estimates that the probability of the flood and wind occurring in a given year on the same day to be in the order of 1 x 10-11 to 1 x 10-13. Computation of wind waves was made using the procedures of the Corps of Engineers [7]. The critical directions were from the north-northwest and northeast with effective fetches of 1.7 and 1.5 miles, respectively. For the 45-mile-per-hour wind, 99.6 percent of the waves approaching the plant would be less than 4.2- and 4.0-foot-high crest to trough for the 1.7- and 1.5-mile fetches as shown on Figures 2.4.3-15 and 2.4.3-16. Maximum water surfaces in the reservoir approaching the plant would be 2.8 and 2.7 feet above the maximum computed level or elevations 722.4 and 722.3, respectively. The maximum water level attained due to the PMF plus wind-wave activity is elevation 723.8 at the ERCW pump station and the nuclear island structures (shield, auxiliary, and control building). The wind waves approaching the Diesel Generator Building and cooling towers break before reaching the structures due to the shallow depth of water. The topography surrounding these structures is such that the wind waves will break on a steeper slope (4H:1V) than the slope immediately adjacent to the structures. This is shown by Figure 2.4.3-17. The runup estimates are calculated on the basis that the incoming wind waves break before reaching the structure and then reform for a shallower water depth. This reformed wave then approaches the structure. The runups are lower than the maximum reservoir level due to the small wave height for the reformed wave, the shallow water, and the very shallow slope before reaching the structures. Wind-wave runup coincident with the maximum flood level for the diesel generator building and cooling towers is elevation 721.8. The level inside structures that are allowed to flood is elevation 720.1. The flood elevations used as design bases are given in Section 2.4A.1.1. Dynamic Effect of Waves

1. Nonbreaking Waves The dynamic effect of nonbreaking waves on the walls of safety- related structures was investigated using the Rainflow Method [8]. As a result of this investigation, concrete and reinforcing stresses were found to be within allowables.

2. Breaking Waves The dynamic effect of breaking waves on the walls of safety-related structures was investigated using a method developed by D. D. Gaillard and D. A. Molitar. The concrete and reinforcing stresses were found to be less than the allowable stresses using this method.

S2-4.doc 2.4-14

SQN-17

3. Broken Waves The dynamic effect of broken waves on the walls of safety-related structures was investigated using a method proposed by the U.S. Army Coastal Engineering Research Center [7]. This method of design yielded concrete and reinforcing stresses within allowable limits.

All safety-related structures are designed to withstand the static and dynamic effects of the water and waves as stated in Section 2.4.2.2. 2.4.4 Potential Dam Failures (Seismically and Otherwise Induced) There are 20 major dams above SQN. These were examined individually and in groups to determine if failure might result from a seismic event and if such failure or failures occurring concurrently with storm runoff would create critical flood levels at the plant. Two situations were examined: (1) a one-half Safe Shutdown Earthquake (SSE) as defined in Subsection 2.5.2, imposed concurrently with one-half the probable maximum flood and (2) a Safe Shutdown Earthquake (SSE) as defined in Subsection 2.5.2, imposed concurrently with a 25-year flood. Neither of these conditions would create levels greater than the hydrologic probable maximum flood at SQN, described previously in 2.4.3. Details of the dam failure analysis are discussed in Section 2.4.4.2, Dam Failure Permutations. Failure of Chickamauga Dam, downstream, can affect cooling water supplies at the plant. Consequently for conservatism, an arbitrary failure was imposed. This resulting condition would not be critical to plant operation, as discussed in Section 2.4.11.6. 2.4.4.1 Reservoir Description Characteristics of dams that influence river conditions at SQN are contained in Tables 2.4.1-1 and 2.4.1-2. Their location with respect to the plant is shown on Figure 2.1.1-1. Seismic safety criteria were not incorporated in the design of dams upstream from SQN, except Tellico and Norris. Those projects having a potential to influence plant flooding levels were examined, as described in Section 2.4.4.2. Elevation-storage relationships and seasonally varying storage allocations in the major projects are shown on the 14 sheets of Figure 2.4.1-3. 2.4.4.2 Dam Failure Permutations The plant site and upstream reservoirs are located in the Southern Appalachian Tectonic Province and, therefore, subject to moderate earthquake forces with possible attendant failure. All upstream dams, whose failure has the potential to cause flood problems at the plant, were investigated to determine if failure from seismic or hydrologic events would endanger plant safety. Potential failures from both seismic and hydrologic events and the resulting consequences are discussed in this section. It should be clearly understood that these studies have been made solely to ensure the safety of SQN against failure by floods caused from excessive rainfall or by the assumed failure of dams due to seismic forces. To assure that safe shutdown of SQN is not impaired by flood waters, TVA has in these studies added conservative assumptions to conservative assumptions to be able to show that the plant can be safely controlled even in the event that all these unlikely events occur in just the proper sequence. TVA is of the strong opinion that the chances of the assumed events occurring approach zero probability. S2-4.doc 2.4-15

SQN-17 By furnishing this information, TVA does not infer or concede that its dams are inadequate to withstand great floods and/or earthquakes that may be reasonably expected to occur in the TVA region under consideration. TVA has a program of inspection and maintenance carried out on a regular schedule to keep its dams safe. Instrumentation of the dams to help keep check on their behavior was installed in many of the dams during original construction. Other instrumentation has been added since and is still being added as the need may appear or as new techniques become available. In short, TVA has confidence that its dams are safe against catastrophic destruction by any natural forces that could be expected to occur. 2.4.4.2.1 Seismic Failure Analysis Seismic failure analysis consisted of the following:

1. Determination of the water level at the plant during one-half the PMF with full reservoirs if its crests were augmented by flood waves from the postulated failure of upstream dams during a one-half SSE.

2. Determination of the water level at the plant during a 25-year flood with full reservoirs if its crests were augmented by flood waves from the postulated failure of upstream dams during a Safe Shutdown Earthquake (SSE).

The one-half SSE identified in condition 1 is defined in FSAR Section 2.5.2.4 as having a peak horizontal acceleration value of 0.09 g at the rock foundation. The discussion in Section 2.5.2.4 shows the extreme conservatism contained in the analysis. In the 1998 reanalysis all potentially critical seismic events involving dam failure upstream of the plant site were reevaluated. The six events included the postulated one-half SSE failure of (1) Norris, (2) Fontana, (3) Cherokee-Douglas, and (4) Fontana-Hiwassee-Apalachia-Blue Ridge during one-half the PMF; and the postulated SSE failure of (5) Norris-Cherokee-Douglas and (6) Norris-Douglas-Fort Loudoun-Tellico during a 25 year flood. Seismic failure of upstream dams during nonflood periods pose no threat to the plant. Summary A summary of the results of the seismic analysis is given in Table 2.4.4-1. SQN and upstream dams are located as shown on Figure 2.1.1-1. The highest flood level at SQN from different seismic dam failure and flood combinations would be elevation 707.9 from simultaneous failure of Fontana Dam on the Little Tennessee River and Hiwassee, Blue Ridge, and Apalachia Dams on the Hiwassee River during a one-half safe shutdown earthquake coincident with one-half the PMF. This includes improvements resulting from modifications performed for the Dam Safety Program. Wind waves could raise the elevation to 709.6 in the reservoir. Runup could reach elevation 710.4 on a 4:1 slope to elevation 712.8 on a vertical wall in shallow (4.9 feet) water, and to elevation 710.4 on a vertical wall in deep water. Only one other seismic dam failure combination with coincident floods could cause elevations above plant grade. S2-4.doc 2.4-16

SQN-17 Plant safety would be assured by shutdown prior to these floods crossing plant grade, elevation 705, using the warning system described in Appendix 2.4A. The effect of postulated seismic bridge failure and resulting failure of spillway gate anchors at Watts Bar and Fort Loudoun Dams would not create a safety hazard at SQN. Procedures Concrete Structures The standard method of computing stability is used. The maximum base compressive stress, average base shear stress, the factor of safety against overturning, and the shear strength required for a shear-friction factor of safety of 1 are determined. To find the shear strength required to provide a safety factor of 1, a coefficient of friction of 0.65 is assigned at the elevation of the base under consideration. As stated in Section 2.4.1.2, all of the original stability analyses and postulated dam failure assumptions in the 1998 reanalyses were conservatively assumed to occur in the same manner and in combination with the same postulated rainfall events. The analyses for earthquake are based on the static analysis method as given by Hinds [10] with increased hydrodynamic pressures determined by the method developed by Bustamante and Flores [11]. These analyses include applying masonry inertia forces and increased water pressure to the structure resulting from the acceleration of the structure horizontally in the upstream direction and simultaneously in a downward direction. The masonry inertia forces are determined by a dynamic analysis of the structure which takes into account amplification of the accelerations above the foundation rock. No reduction of hydrostatic or hydrodynamic forces due to the decrease of the unit weight of water from the downward acceleration of the reservoir bottom is included in this analysis. Waves created at the free surface of the reservoir by an earthquake are considered of no importance. Based upon studies by Chopra [12] and Zienkiewicz [13], it is our judgment that before waves of any significant height have time to develop, the earthquake will be over. The duration of earthquake used in this analysis is in the range of 20 to 30 seconds. Although accumulated silt on the reservoir bottom would dampen vertically traveling waves, the effect of silt on structures is not considered. There is only a small amount of silt now present, and the accumulation rate is slow, as measured by TVA for many years [14]. Embankment Embankment analysis was made using the standard slip circle method, except for Chatuge and Nottely Dams where the Nemark method for the dynamic analysis of embankment slopes was used. The effect of the earthquake is taken into account by applying the appropriate static inertia force to the dam mass within the assumed slip circle. In the analysis, the embankment design constants used, including the sheer strength of the materials in the dam and the foundation, are the same as those used in the original stability analysis. S2-4.doc 2.4-17

SQN-17 Although detailed dynamic soil properties are not available, a value for seismic amplification through the soil has been assumed based on previous studies pertaining to TVA nuclear plants. These studies have indicated maximum amplification values slightly in excess of two for a rather wide range of shear wave velocity to soil height ratios. For these analyses, a straight-line variation is used with an acceleration at the top of the embankment being two times the top of rock acceleration. Flood Routing The runoff model described in Section 2.4.3.3, which includes unsteady flow models for critical reservoirs and river reaches, was used to reevaluate plant site flood levels resulting from the postulated SSE and one-half SSE dam failure combinations. The remaining events produced plant site flood levels sufficiently lower than the controlling events and were not evaluated. Reservoir operating procedures used were those applicable to the season and flood inflows. This section was revised with a major rearrangement to locate the controlling events evaluated in the 1998 analysis first and the non-controlling events, which were not re-calculated later. The non-controlling events are left in the SAR for history. One-half SSE Concurrent With One-Half the Probable Maximum Flood Previous evaluations have been made which determined flood levels at SQN for potentially critical events. Re-evaluations made later using the updated runoff model described in Section 2.4.3.3 and including the Dam Safety Program modifications did not determine flood levels for those events which were previously shown to clearly not be controlling. The 1998 analysis for determining the effects of the Dam Safety Program modifications determined that non-flood related seismic dam failure events clearly pose no threat to the plant. Flood levels were determined for six combined seismic/flood events. Only two of these controlling seismic/flood events would exceed plant grade. These two events consist of multiple dam failures on (1) Little Tennessee/Hiwassee, and (2) Clinch/Upper Tennessee rivers with flood levels at SQN of El. 707.9 and 706, respectively. The following is detailed descriptions of the potentially critical controlling events including reevaluated flood levels, followed by brief descriptions of the non-controlling failure events previously evaluated. Multiple Failures Although considered, as discussed in the following paragraphs, TVA believes that multiple dam failures are an extremely unlikely event. TVA's search of the literature reveals no record of failure of concrete dams from earthquake. The postulation of an SSE of 0.18 g acceleration is a very conservative upper limit in itself (as stated in Section 2.5.2). In addition, the SSE must be located in a very precise region to have the potential for multiple dam failures. SSE - In order to fail three dams--Norris, Cherokee, and Douglas--the epicenter of a SSE must be confined to a relatively small area, the shape of a football, about 10 miles wide and 20 miles long. In order to fail four dams--Norris, Douglas, Fort Loudoun, and Tellico--the epicenter of an SSE must be confined to a triangular area with sides of approximately 1 mile in length. However, as an extreme upper limit the above two combinations of dams are postulated to fail as well as the combinations of (1) Fort Loudoun, Tellico, and Fontana; (2) Fontana and Douglas; and (3) Fontana and the six Hiwassee River dams. The 1998 re-analysis determined that only the first two combinations are controlling and need to be considered. Only the Norris-Cherokee-Douglas event would exceed plant grade elevation. S2-4.doc 2.4-18

SQN-17 One-half SSE - Attenuation studies of the one-half SSE show that there are three combinations of simultaneous failures of more than one dam which need to be considered with respect to SQN safety which are discussed below. These are (1) Cherokee-Douglas, (2) Fontana-Hiwassee-Apalachia-Blue Ridge, and (3) Hiwassee-Apalachia-Blue Ridge-Ocoee No 1.-Nottely. The 1998 re-analysis determined that only the first two combinations are controlling and need to be considered. Only the Fontana-Hiwasse-Apalachia-Blue Ridge event would exceed plant grade. The following descriptions are first for the controlling events for which flood levels were calculated for the 1998 reanalysis, followed by the non-controlling events which were not re-analyzed in 1998. One-half SSE Concurrent With One-Half the Probable Maximum Flood (Controlling Events)

1. Norris Dam Results of the Norris Dam stability analyses for a typical spillway block and a typical non-overflow section of maximum height are shown on Figure 2.4.4-8. Because only a small percentage of the spillway base is in compression, this structure is judged to fail. The high non-overflow section with a small percentage of the base in compression and with high compressive and shearing stresses is also judged to fail.

Figure 2.4.4-9 shows the likely condition of the dam after failure. Based on stability analyses, the non-overflow blocks remaining in place are judged to withstand the one-half SSE. Blocks 33-44 are judged to fail by overturning. The location of the debris is not based on any calculated procedure of failure because it is believed that this is not possible. It is TVA's judgment, however, that the failure mode shown is one logical assumption; and, although there may be many other logical assumptions, the amount of channel obstruction would probably be about the same. The discharge rating for this controlling, debris section was developed from a 1:150 scale hydraulic model at the TVA Engineering Laboratory and was verified closely by mathematical analysis. In the hydrologic routing for this failure, Melton Hill Dam was postulated to fail when the flood wave reached headwater elevation 804, based on structural analysis. The headwater at Watts Bar Dam would reach elevation 758.1, 8.9 feet below top of dam. The west saddle dike at Watts Bar Dam would be overtopped and breached. A complete washout of the dike was assumed. The resulting water level at the nuclear plant site is 698.1, 6.9 feet below plant grade 705.

2. Fontana Dam Fontana Dam was assumed to fail in the one-half SSE, although no stability analysis was made.

Fontana is a high dam constructed with three longitudinal contraction joints in the higher blocks. A structural defect in Fontana Dam was found in October of 1972 and consists of a longitudinal crack in three blocks in the curved portion at the left end of the dam (see Figure 2.4.4-16). Strengthening of these blocks by post-tensioning and grouting of the cracks was completed in October 1973 (see Figure 2.4.4-17). Only these three blocks are cracked, and there is no evidence that any other portion of the dam is weakened. S2-4.doc 2.4-19

SQN-17 Studies and tests, undertaken with the concurrence of a board of private consulting engineers, indicate that this cracking was caused by a longitudinal thrust created by a combination of long-time concrete growth and expansion due to temperature rise in the summer months. This thrust tends to push the curved blocks upstream. The studies and tests will continue until there is established a basis for design of permanent measures to control the future behavior of the dam. The strengthening work has reestablished the structural integrity of the cracked blocks. Although the joints are keyed and grouted, it is possible that the grouting was not fully effective. Consequently, there is some question as to how this structure will respond to the motion of a severe earthquake. To be conservative, therefore, it is assumed that Fontana Dam will not resist the one-half SSE without failure. Figure 2.4.4-16 shows the part of Fontana Dam judged to remain in its original position after failure and the assumed location of the debris of the failed portion. The location of the debris after failure is one logical assumption based on a failure of the dam at the longitudinal contraction joints. There may be other logical assumptions, but the amount of channel obstruction would probably be about the same. The higher blocks 9-27, containing either two or three longitudinal joints, are assumed to fail. Right abutment blocks 1-8 and left abutment blocks 28 and beyond were judged to be stable for the following reasons:

1. Their heights are less than one-half the maximum height of the dam.

2. None of these blocks have more than one longitudinal contraction joint, and some have no longitudinal joints.

3. The back slope of Fontana Dam is one on 0.76, which the original stability analysis shows is flatter than that required for stability for the normal static loadings.

Although not investigated, it was assumed that Nantahala Dam, upstream from Fontana and Santeetlah on a downstream tributary, and the three ALCOA dams, downstream on the Little Tennessee River, Cheoah, Calderwood, and Chilhowee, would fail along with Fontana in the one-half SSE. Instant vanishment was assumed. Tellico and Watts Bar Dam spillway gates would be operable during and after the one-half SSE. Failure of the bridge at Fort Loudoun Dam would render the spillway gates inoperable in the wide-open position. The flood wave would overtop Tellico Dam and its saddle dikes. Transfer of water into Fort Loudoun would occur but would not be sufficient to overtop the dam or to prevent failure of Tellico Dam. Tellico was postulated to completely fail. Watts Bar headwater would reach elevation 761.3, 5.7 feet below top of dam. The Watts Bar west saddle dike would be overtopped and breached. A complete washout of the dike was assumed. The elevation at the plant site would be 702.8, 2.2 feet below plant grade.

3. Cherokee-Douglas The simultaneous failure of Cherokee and Douglas Dams could occur when the one-half SSE is located midway between the dams which are just 15 miles apart.

Results of the Cherokee Dam stability analysis for a typical spillway block are shown in Figure 2.4.4-

10. Based on this analysis, the spillway is judged stable at the foundation base elevation 900.

Analyses made for other elevations above elevation 900, but not shown in Figure 2.4.4-10, indicate the resultant of forces falls outside the base at elevation 1010. The spillway is assumed to fail at that elevation. S2-4.doc 2.4-20

SQN-17 The non-overflow dam is embedded in fill to elevation 981.5 and is considered stable below that elevation. However, stability analysis indicates failure will occur above the fill line. The powerhouse intake is massive and backed up by the powerhouse. Therefore, it is judged able to withstand the one-half SSE without failure. Results of the analysis for the highest portion of the south embankment are shown on Figure 2.4.4-11. The analysis was made using the same shear strengths of material as were used in the original analysis and shows a factor of safety of 0.85. Therefore, the south embankment is assumed to fail during the one-half SSE. Because the north embankment and saddle dams 1, 2, and 3 are generally about one-half or less as high as the south embankment, they are judged to be stable for the one-half SSE. Figure 2.4.4-12 shows the assumed condition of the dam after failure. All debris from the failure of the concrete portion is assumed to be located downstream in the channel at elevations lower than the remaining portions of the dam, and therefore, will not obstruct flow. Results of the Douglas Dam original stability analysis for a typical spillway block are shown in Figure 2.4.4-13. The upper part of the Douglas spillway is approximately 12 feet higher than Cherokee, but the amplification of the rock surface acceleration is the same. Therefore, based on the Cherokee analysis, it is judged that the Douglas spillway will fail at elevation 937, which corresponds to the assumed failure elevation of the Cherokee spillway. The Douglas non-overflow dam is similar to that at Cherokee and is embedded in fill to elevation 927.5. It is considered stable below that elevation. However, based on the Cherokee analysis, it is assumed to fail above the fill line. The abutment non-overflow blocks 1-5 and 29-35, being short blocks, are considered able to resist the one-half SSE without failure. The powerhouse intake is massive and backed up downstream by the powerhouse. Therefore, it is considered able to withstand the one-half SSE without failure. Results of the original analysis of the saddle dam shown on Figure 2.4.4-14 indicate a factor of safety of one. Therefore, the saddle dam is considered to be stable for the one-half SSE. Figure 2.4.4-15 shows the portions of the dam judged to fail and the portions judged to remain. All debris from the failed portions is assumed to be located downstream in the channel at elevations lower than the remaining portions of the dam and, therefore, will not obstruct flow. These failures, in conjunction with one-half the probable maximum flood, would overtop Fort Loudon for only 6 hours, but would not fail the dam. At Watts Bar the west saddle dike would be overtopped and breached. A complete washout of the dike was assumed. Crest level at SQN would be elevation 701.1, 3.9 feet below plant.

4. Fontana, Hiwassee, Apalachia, and Blue Ridge Dams Fontana, Hiwassee, Apalachia, and Blue Ridge Dams could fail when the one-half SSE is located within the football-shaped area shown in Figure 2.4.4-18.

This event produces maximum ground accelerations of 0.09 g at Fontana, 0.09 g at Hiwassee, 0.07 g at Apalachia, 0.08 g at Chatuge, 0.05 g at Nottely, 0.03 g at Ocoee No. 1, 0.04 g at Blue Ridge, 0.04 g at Fort Loudoun and Tellico, and 0.03 g at Watts Bar. Failure is postulated for Fontana and Hiwassee for an earthquake epicenter located anywhere within the football-shaped area shown on S2-4.doc 2.4-21

SQN-17 Figure 2.4.4-18. Ground accelerations shown for the various dams are maximum that could occur for epicenters located at various points in the described area and would not occur simultaneously. Fort Loudoun, Tellico, and Watts Bar Dams and spillway gates would remain intact. The degree of Fontana failure and likely position of debris are judged to be comparable to that shown for single failure in Figure 2.4.4-16. Hiwassee, Apalachia, and Blue Ridge Dams were assumed to completely disappear. Chatuge was judged not to fail as the acceleration is less than for the one-half SSE centered at the dam. Nottely Dam is a rockfill dam with large central impervious rolled fill core. The maximum attenuated ground acceleration at Nottely is only 0.054 g. A field exploration boring program and laboratory testing program of samples obtained in a field exploration was conducted. During the field exploration program, standard penetration tests blow counts were obtained on both the embankment and its foundation materials. Both static and dynamic (cyclic) triaxial shear tests were made. This information was used in the Newmark Method of Analysis. The "Newmark Method of Analysis" (Newmark, N. M., "Effects of Earthquake on Dams of Embankments," Geotechnique 15:140-141, 156, 1965) utilizing the information obtained from the testing program was used to determine the structural stability of Nottely Dam. We conclude Nottely Dam can easily resist the attenuated ground acceleration of 0.054 g with no detrimental damage. Ocoee No. 1 Dam is a concrete gravity structure. The maximum attenuated ground acceleration is 0.03 g. The 0.03 g with the proper amplification was used to analyze the structural stability of structures at Ocoee No. 1. The method of analysis used was the same as described previously under "Procedures, Concrete Structures." The analysis shows low stresses with good factors of safety against sliding and overturning. We conclude the dam will not fail. In the original analysis, the failure wave hydrograph was approximated for the Hiwassee River at its mouth for the failures of Hiwassee, Apalachia and Blue Ridge Dams. In the 1998 re-analysis an unsteady flow model described in Section 2.4.3.3 developed during the dam safety studies was used as an adjunct to route the Hiwassee, Apalachia and Blue Ridge failures. In the simultaneous failure of Fontana, Hiwassee, Apalachia, and Blue Ridge Dams, the Fontana failure wave would overtop and fail the Tellico embankments. Transfer of water into Fort Loudoun would occur but would not be sufficient to overtop the dam or to prevent failure of Tellico. Tellico was postulated to completely fail. Watts Bar headwater would reach elevation 761.3, 5.7 feet below top of dam. The west saddle dike at Watts Bar would be overtopped. A complete washout of the dike down to ground elevation was assumed. This flood wave combined with that of Hiwassee, Blue Ridge, and Apalachia Dams would produce a maximum flood level at the plant site of 707.9, 2.9 feet above 705 plant grade. This is the highest flood resulting from any combination of seismic and concurrent flood events. The stage hydrograph at the plant site is shown on Figure 2.4.4-21. SSE Concurrent With 25-Year Flood (Controlling Events)

5. Norris, Cherokee, and Douglas Norris, Cherokee, and Douglas Dams were also postulated to fail simultaneously. Figure 2.4.4-29 shows the location of an SSE, and its attenuation, which produces 0.15 g at Norris, 0.09 g at Cherokee and Douglas, 0.08 g at Fort Loudoun and Tellico, 0.05 g at Fontana, and 0.03 g at Watts Bar. Fort Loudoun, Tellico, and Watts Bar have been judged not to fail for the one-half SSE (acceleration value of 0.09 g) (see following discussion of non-controlling events). The bridge at Fort Loudoun Dam, however, might fail under 0.08 g forces, falling on any open gates and on gate-hoisting machinery. Trunnion anchor bolts of open gates would fail and the gates would be washed downstream, leaving an open spillway. Closed gates could not be opened. The most conservative S2-4.doc 2.4-22

SQN-17 assumption was used that at the time of the seismic event on the upstream tributary dams, the crest of the 25-year flood would likely have passed Fort Loudoun and flows would have been reduced to turbine capacity. Hence spillway gates would be closed. As stated before, it is believed that multiple dam failure is extremely remote, and it seems reasonable to exclude Fontana on the basis of being the most distant in the cluster of dams under consideration. For the postulated failures of Norris, Cherokee, and Douglas, the portions judged to remain and debris arrangements are as given in Figures 2.4.4-9, 2.4.4-12, and 2.4.4-15, respectively. The SSE will produce the same postulated failures of Cherokee and Douglas Dams as were described for the one-half SSE. For Norris under SSE conditions, blocks 31-45 (883 feet of length) are judged to fail. The resulting debris downstream would occupy a greater span of the valley cross section than would the debris from the one-half SSE but with the same top level, elevation 970. Figure 2.4.4-28 shows the part of the dam judged to fail and the location and height of the resulting debris. The discharge rating for this controlling debris section was developed from a 1:150 scale hydraulic model at the TVA Engineering Laboratory and was verified closely by mathematical analysis. The somewhat more extensive debris in SSE failure restricts discharge slightly compared to one-half SSE failure conditions. The flood for the postulated failure combination would overtop and breach Fort Loudoun Dam. Although transfer of water into Tellico would occur, it would not be sufficient to overtop the dam. At Watts Bar Dam the headwater would reach 764.9, 2.1 feet below the top of the earth embankment of the main dam. However, the west saddle dike at Watts Bar Dam would be overtopped and breached. Resulting water surface at SQN would reach elevation 706. This is 1.0 foot higher than plant grade. This is the highest flood resulting from any combination of SSE seismic and flood events. The flood elevation Flow and stage hydrographs at the plant site is shown on Figure 2.4.4-30.

6. Norris, Douglas, Fort Loudoun, and Tellico Norris, Douglas, Fort Loudoun, and Tellico Dams were postulated to fail simultaneously. Figure 2.4.4-31 shows the location of an SSE, and its attenuation, which produces 0.12 g at Norris, 0.08 g at Douglas, 0.12 g at Fort Loudoun and Tellico, 0.07 g at Cherokee, 0.06 g at Fontana, and 0.04 g at Watts Bar. Cherokee is judged not to fail at 0.07 g; Watts Bar has previously been judged not to fail at 0.09 g; and, for the same reasons as given above, it seems reasonable to exclude Fontana in this failure combination. For the postulated failures of Norris, Douglas, Fort Loudoun, and Tellico, the portions judged to remain and the debris arrangements are as given in Figures 2.4.4-9, 2.4.4-15, 2.4.4-26, and 2.4.4-27, respectively. For analysis purposes, Fort Loudoun and Tellico were postulated to fail completely as the portions judged to remain are relatively small.

The SSE will produce the same postulated failure of Douglas Dam as was described for the one-half SSE. Results of the stability analysis for Fort Loudoun Dam are shown on Figure 2.4.4-24. Because the resultant of forces falls outside the base, a portion of the spillway is judged to fail. Based on previous modes of failure for Cherokee and Douglas, the spillway is judged to fail above elevation 750 as well as the bridge supported by the spillway piers. The results of the slip circle analysis for the highest portion of the embankment are shown on Figure 2.4.4-25. Because the factor of safety is less than one, the embankment is assumed to fail. S2-4.doc 2.4-23

SQN-17 No analysis was made for the powerhouse under SSE. However, an analysis was made for the one-half SSE with no water in the units, a condition believed to be extremely remote to occur during the one-half SSE. Because the stresses were low and a large percentage of the base was in compression, it is considered that the addition of water in the units would be a stabilizing factor, and the powerhouse is judged not to fail. Figure 2.4.4-26 shows the condition of the dam after assumed failure. All debris from the failure of the concrete portions is assumed to be located in the channel below the failure elevations. No structural analysis was made for Tellico Dam failure in the SSE. Because of the similarity to Fort Loudoun, the spillway and entire embankment are judged to fail in a manner similar to Fort Loudoun. Figure 2.4.4-27 shows after failure conditions with all debris assumed located in the channel below the failure elevation. This postulated failure combination results in Watts Bar headwater elevation 758.9, 8.1 feet below above the top of the embankment of the main dam. The west saddle dike at Watts Bar Dam would be overtopped and breached. A complete washout of the dike was assumed. The resulting water level at SQN would be elevation 699.3, 5.7 feet below plant grade 705. One-half SSE Concurrent With One-Half the Probable Maximum Flood (Non-controlling Events-Historical)

1. Watts Bar Dam Stability analyses of Watts Bar Dam powerhouse and spillway sections result in the judgment that these structures will not fail. The analyses show low stresses with about 38 percent of the spillway base in compression and about 42 percent of the powerhouse base in compression. Results are given in Figure 2.4.4-1. Dynamic analysis of the concrete structures resulted in the determination that the base acceleration is amplified at levels above the base.

The slip circle analysis of the earth embankment section results in a factor of safety of 1.52, and the embankment is judged not to fail. Results are given in Figure 2.4.4-2. Normally for the condition of peak discharge at the dam for one-half the PMF, the spillway gates would be in the wide open position (Figure 2.4.4-3). But, analysis of the bridge structure for forces resulting from a one-half SSE, including amplification of acceleration results in the determination that the bridge could fail as a result of shearing the anchor bolts. The downstream bridge girders could strike the spillway gates. The impact of the girders striking the gates could fail the bolts which anchor the gate trunnions to the pier anchorages allowing the gates to fall. The flow over the spillway crest would be the same as that prior to bridge and gate failure. Hence, bridge failure will cause no adverse effect on the flood. A potentially severe condition is the bridge falling when most spillway gates would be closed. The gate hoisting machinery would be inoperable after being struck by the bridge. As a result, the flood would crest with the gates closed and the bridge deck and girders lying on top of the spillway piers. Analysis of the concrete portions of the dam for the headwater for this condition shows that they will not fail. Flood levels at SQN for all the conditions described above is safely below plant grade elevation 705. S2-4.doc 2.4-24

SQN-17

2. Fort Loudoun Dam Stability analyses of Fort Loudoun Dam powerhouse and spillway sections result in the judgment that these structures will not fail. The analyses show low base stresses, with near two-thirds of the base in compression. Results are given in Figure 2.4.4-4.

Slip circle analysis of the earth embankment results in a factor of safety of 1.26, and the embankment is judged not to fail. Results are given in Figure 2.4.4-5. The spillway gates and bridge are of the same design as those at Watts Bar Dam. Conditions of failure during a one-half SSE are the same, and no problems are likely. Coincident failure at Fort Loudoun and Watts Bar does not occur. For the potentially critical case of Fort Loudoun bridge failure at the onset of the main portion of one-half the probable maximum flood flow into Fort Loudoun Reservoir, it was found that the Watts Bar inflows are much less than the condition resulting from simultaneous failure of Cherokee and Douglas.

3. Tellico Dam No part of Tellico Dam is judged to fail. Results of the stability analyses for a typical non-overflow block and a typical spillway block are shown in Figure 2.4.4-6. The result of the stability analysis of the earth embankment is shown in Figure 2.4.4-7 and indicates a factor of safety of 1.28.

4. Cherokee Dam No hydrologic results are given for the single failure of Cherokee Dam because the simultaneous failure of Cherokee and Douglas is more critical.

5. Douglas Dam No hydrologic results are given for the single failure of Douglas Dam because the simultaneous failure of Cherokee and Douglas is more critical.

6. Hiwassee River Dams Hiwassee Dam was assumed to fail in the one-half SSE. No hydrologic results are given for the single failure of Hiwassee Dam because its simultaneous failure with other dams is more critical.

7. Apalachia Apalachia Dam was assumed to fail in the one-half SSE. No hydrologic results are given for the single failure of Apalachia Dam because its simultaneous failure with other dams is more critical.

8. Blue Ridge Blue Ridge Dam was assumed to fail in the one-half SSE. No hydrologic results are given for the single failure of Blue Ridge Dam because its simultaneous failure with other dams is more critical.

S2-4.doc 2.4-25

SQN-17

9. Ocoee No. 1 Ocoee No. 1 Dam was assumed to fail in the one-half SSE. No hydrologic results are given for the single failure of Ocoee No. 1 Dam because its simultaneous failure with other dams is more critical.

10. Nottely Nottely Dam was assumed to fail in the one-half SSE. No hydrologic results are given for the single failure of Nottely Dam because its simultaneous failure with other dams is more critical.

11. Chatuge Chatuge Dam is a homogeneous, impervious rolled-fill dam. With the epicenter of the one-half SSE located at the dam, the maximum ground acceleration at Chatuge is 0.09 g. Ground accelerations of this magnitude should have no detrimental effects on a well-constructed compacted earthfill embankment. We know of no failures of compacted earth embankment slopes from earthquake motions. Failures to date have been associated with other liquefaction of hydraulic fill embankments of liquefaction of loose granular foundation materials. The rolled embankment materials in Chatuge are not sensitive to liquefaction. To verify these conclusion analysis using the "Newmark Method for the Dynamic Analysis of Embankment Slopes" (Newmark, N. M., "Effects of Earthquake on Dams of Embankments," Geotechnique 15:140-141, 156, 1965) was made to determine the structural stability of Chatuge. We conducted a field exploration boring program and laboratory testing program of samples obtained in the field exploration. During the field exploration program, standard penetration tests blow counts were obtained on both the embankment and its foundation materials. Both static and dynamic (cyclic) triaxial shear tests were made. This information was used in the Newmark Method of Analysis. We concluded from the Analysis that the Chatuge Dam can easily resist the ground acceleration of 0.09 g with no detrimental damage.

12. Hiwassee, Apalachia, Blue Ridge, Ocoee No. 1, and Nottely Hiwassee, Apalachia, Blue Ridge, Ocoee No.1, and Nottely Dams could fail when the one-half SSE is critically located. All five dams were assumed to completely disappear in this event. Resulting crest level at SQN would be below plant grade 705.

SSE Concurrent With 25-Year Flood (Non-controlling Events - Historical)

1. Watts Bar Dam A reevaluation was not made for Watts Bar Dam for SSE conditions. A previous evaluation had determined that even if the dam is arbitrarily removed instantaneously, the level at the nuclear plant site would be below plant grade.

2. Fort Loudoun Dam No hydrologic routing for the single failure of Fort Loudoun, including the bridge structure, is made because its simultaneous failure with Tellico and Fontana, as well as with Tellico, Norris, and Douglas, are controlling.

3. Tellico Dam No routing for the single failure of Tellico is made for the reasons given above for Fort Loudoun.

S2-4.doc 2.4-26

SQN-17

4. Norris Dam This postulated single failure would result in peak headwater at Watts Bar below the top of the earth portions of the dam. Routing was not carried further because it was evident that flood levels at the plant site would be considerably lower than for the Norris failure in the one-half SSE combined with the one-half PMF.

5. Hiwassee River Dams Considered Separately No structural analyses were made for Chatuge, Nottely, Blue Ridge, Ocoee No. 1, Hiwassee, and Apalachia in the SSE. Instead, all six dams were postulated to fail completely.

No routing for the failure of the six Hiwassee dams alone is made because their simultaneous failure with Fontana is considered as discussed earlier in this subparagraph.

6. Cherokee, Douglas, and Fontana Considered Separately The SSE will produce the same postulated failures of Cherokee, Douglas, and Fontana Dams as were described for the one-half SSE. None of these failures need to be carried downstream, however, because elevations would be lower than the same failures in one-half the probable maximum flood.

7. Fort Loudoun, Tellico, and Fontana An SSE centered between Fontana and the Fort Loudoun-Tellico complex was postulated to fail these three dams. The four ALCOA dams downstream from Fontana and Nantahala, an ALCOA dam, upstream were also postulated to fail completely in this event. Watts Bar Dam and spillway gates would remain intact, but failure of the roadway bridge was postulated which would render the spillway gates inoperable. At the time of seismic failure, discharges would be small in the 25-year flood. For conservatism, Watts Bar gates were assumed inoperable in the closed position after the SSE event.

This event would result in a flood level at the nuclear plant site below 705 plant grade.

8. Douglas and Fontana Douglas and Fontana were postulated to fail simultaneously. The location of an SSE required to fail both dams would produce 0.14 g at Douglas, 0.09 g at Fontana, 0.07 g at Cherokee, 0.05 g at Norris, 0.06 g at Fort Loudoun and Tellico, and 0.03 g at Watts Bar. For the postulated failures of Douglas and Fontana, the portions judged to remain and the debris arrangements are as given in Figures 2.4.4-15 and 2.4.4-16. Fort Loudoun, Tellico, and Watts Bar have previously been judged not to fail for the OBE (0.09 g). The bridge at Fort Loudoun Dam, however, might fail under 0.06 g forces, falling on gates and on gate hoisting machinery. Fort Loudoun gates were assumed inoperable in the closed position following the SSE event. Resulting water surface at SQN would be below plant grade.

9. Fontana and Hiwassee River Dams Fontana and six Hiwassee River dams--Hiwassee, Apalachia, Chatuge, Nottely, Blue Ridge, and Ocoee No. 1--were postulated to fail simultaneously. For the postulated failure of Fontana, the portion judged to remain and the debris arrangements are as given in Figure 2.4.4-16. The six Hiwassee dams were assumed to fail completely. Fort Loudoun, Tellico, and Watts Bar are judged not to fail with all gates operable. The Fontana surge combined with that of the six Hiwassee River dams would reach an elevation at the plant site below the plant grade.

S2-4.doc 2.4-27

SQN-17 2.4.4.2.2 Hydrologic Failure Analysis All upstream and downstream dams which could have significant influence on flood levels at SQN were examined for potential failure during all flood conditions, which would have the potential to produce maximum plant flood levels including the dam PMF at the individual upstream dams. Concrete sections were examined for overturning and horizontal shear and sliding. Spillway gates were examined for stability at potentially critical water levels and against failure from being struck by water borne objects. Locks and lock gates were examined for stability, and earth embankments were examined for erosion due to overtopping. During the SQN PMF, the only failure would be the west saddle dike at Watts Bar. Chickamauga Dam would be overtopped but was conservatively assumed not to fail. Concrete Section Analysis For concrete dam sections, comparisons were made between the original design headwater and tailwater levels and those that would prevail in the PMF. If the overturning moments and horizontal forces were not increased by more than 20 percent, the structures were considered safe against failure. All upstream dams passed this test except Douglas, Fort Loudoun, and Watts Bar. Original designs showed the spillway sections of these dams to be most vulnerable. These spillway sections were examined in further detail and judged to be stable. Spillway Gates During peak PMF conditions the radial spillway gates of Fort Loudoun and Watts Bar Dams will be wide open with flow over the gates and under the gates. For this condition both the static and dynamic load stresses in the main structural members of the gate will be less than the yield stress by a factor of three. The stress in the trunnion pin is less than the allowable design stress by a factor greater than 10. The trunnion pin is prevented from dislodgment by a key into the gate anchorage assembly and fitting into a slot in the pin. The gates were also investigated for the condition when rising headwater level first begins to exceed the bottom of the gates in the wide-open position. This condition produces the largest forces tending to rotate the radial gates upward. In the wide-open position the gates are dogged against steel gate stops anchored to the concrete piers. The stresses in the gate stop members are less than the yield stress of the material by a factor of 2. It is concluded that the above-listed margins are sufficient to provide assurance also that the gates will not fail as a result of additional stresses which may result from possible vibrations of the gates acting as orifices. Waterborne Objects Consideration has been given to the effect of water borne objects striking the spillway gates and bents supporting the bridge across Watts Bar Dam at peak water level at the dam. The most severe potential for damage would be by a barge which has been torn loose from its moorings and floats into the dam. Should the barge approach the spillway portion of the dam end on, one bridge bent could be failed by the barge and two spillway gates could be damaged and possibly swept away. The loss of one bridge bent will not collapse the bridge because the bridge girders are continuous members and the stress in S2-4.doc 2.4-28

SQN-17 the girders will be less than the ultimate stress for this condition of one support being lost. Should two gates be swept away, the nappe of the water surface over the spillway weir would be such that the barge would be grounded on the tops of the concrete spillway weirs and provide a partial obstruction to flow comparable to unfailed spillway gates. Hence the loss of two gates from this cause will have little effect on the peak flow and elevation. Should the barge approach the spillway portion broadside, two and possibly three bridge bents could be failed. For this condition, the bridge would collapse on the barge and the barge would be grounded on the tops of the spillway weirs. This would be probable because the approach velocity of the barge would be from 4-to-7 miles per hour and the bottom of the barge would be about six inches above the tops of the weirs. For this condition the barge would be grounded before striking the spillway gates because the gates are about 20 feet downstream from the leg of the upstream bridge bents. Lock Gates The lock gates at Fort Loudoun, Watts Bar, and Chickamauga were examined for possible failure with the conclusion that no potential for failure exists because the gates are designed for a differential hydrostatic head greater than that which exists during the probable maximum flood. Embankment Breaching In the 1998 reanalysis, the only embankment failure would be the west saddle dike at Watts Bar Dam. Chickamauga Dam, downstream of the plant, would be overtopped but was assumed not to fail. This is conservative as failure of Chickamauga Dam would slightly lower flood elevations at the plant. The adopted relationship to compute the rate of erosion in an earth dam failure is that developed and used by the Bureau of Reclamation in connection with its safety of dams program [16]. The expression relates the volume of eroded fill material to the volume of water flowing through the breach. The equation is: Qsoil

                              = Ke  x Q water where Qsoil = Volume of soil eroded in each time period Qwater = Volume of water discharged each time period K     = Constant of proportionality, 1 for the soil and discharge relationships in this study e     = Base of natural logarithm system b

X = tan d H Where b= Base length of overflow channel at any given time H= Hydraulic head at any given time d = Developed angle of friction of soil material. A conservative value of 13 degrees was adopted for materials in the dams investigated. S2-4.doc 2.4-29

SQN-17 Solving the equation, which was computerized, involves a trial and error procedure over short depth and time increments. In the program, depth changes of 0.1 foot or less are used to keep time increments to less than one second during rapid failure and up to about 350 seconds prior to breaching. The solution of an earth embankment breach begins by solving the erosion equation using a headwater elevation hydrograph assuming no failure. Erosion is postulated to occur across the entire earth section and to start at the downstream edge when headwater elevations reached a selected depth above the dam top elevation. Subsequently, when erosion reaches the upstream edge of the embankment, breaching and rapid lowering of the embankment begins. Thereafter, computations include headwater adjustments for increased reservoir outflow resulting from the breach. Watts Bar West Saddle Dike Embankment Failure Figure 2.4.4-37 is a general plan of Watts Bar showing elevations and sections. Figure 2.4.4-38 is a topographic map of the general vicinity of Watts Bar Dam. Figure 2.4.4-39 is a general plan and section of the west saddle dike. The west saddle dike was examined and found subject to failure from overtopping. This failure was assumed to be a complete washout and add to the discharge from Watts Bar Dam. Some verification for the breaching computational procedures illustrated above was obtained by comparison with actual failures reported in the literature and in informal discussion with hydrologic engineers. These reports show that overtopped earth embankments do not necessarily fail. Earth embankments have sustained overtopping of several feet for several hours before failure occurred. An extreme example is Oros earth dam in Brazil [17] which was overtopped to a depth of approximately 2.6 feet along a 2,000-foot length for 12 hours before breaching began. Once an earth embankment is breached, failure tends to progress rapidly, however. How rapidly depends upon the material and headwater depths during failure. Complete failures computed in this and other studies have varied from about one-half to six hours after initial breaching. This is consistent with actual failures. Chickamauga Embankment Failure In the original analysis, the failure of earth embankments at Chickamauga Dam, 13.5 miles downstream from SQN, reduced reduce flood levels at the plant by 0.9 feet. Future embankment improvements are planned for Chickamauga Dam, which if implemented, would prevent failure. Therefore, although overtopped in the PMF, the dam was assumed not to fail in determining flood elevations at the plant. This assumption is conservative. 2.4.4.3 Unsteady Flow Analysis of Potential Dam Failures Unsteady flow routing techniques were used to evaluate plant site flood levels wherever their inherent accuracy was needed. For PMF determinations unsteady flow models described in Section 2.4.3.3 were used. For routing floods from postulated seismically induced dam failures of tributary dams, additional unsteady flow models were used as adjuncts to those described in Section 2.4.3.3. Unsteady flow techniques were applied in Norris Reservoir. The Norris Reservoir model was developed in sufficient detail to define the manner in which the reservoir would supply and sustain outflow following postulated dam failure. The model was verified by comparing its routed headwater level in the one-half PMF with those using storage-routing techniques. Headwater level agreed within a foot, and the model was considered adequate for the purpose. S2-4.doc 2.4-30

SQN-17 Unsteady flow techniques were also applied in Cherokee, Douglas, and Fontana Reservoirs. The reservoir models were developed in sufficient detail to define the manner in which the reservoirs would supply and sustain outflow following postulated dam failure. 2.4.4.4 Water Level at Plant Site Maximum water level at the plant from different postulated combinations of seismic dam failures coincident with floods would be elevation 707.9, excluding wind wave effects. It would result from the one-half SSE failure of Fontana, Hiwassee, Apalachia, and Blue Ridge Dams coincident with one-half the probable maximum flood. March wind with one percent exceedance probability over the 1.4-mile effective fetch from the critical north-northwest direction is 26 miles per hour over land. This would cause reservoir waves to reach elevation 709.6. Runup could reach elevation 710.4 on a smooth 4:1 slope, elevation 712.8 on a vertical wall in shallow (4.9 feet) water, and elevation 710.4 on a vertical wall in deep water. 2.4.5 Probable Maximum Surge and Seiche Flooding (HISTORICAL INFORMATION) Chickamauga Lake level during nonflood conditions could be no higher than elevation 685.44, top of gates, and is not likely to exceed elevation 682.5, normal summer level, for any significant time. No conceivable hurricane or cyclonic-type winds could produce the over 20 feet of wave height required to reach plant grade elevation 705. 2.4.6 Probable Maximum Tsunami Flooding (HISTORICAL INFORMATION) Because of its inland location, SQN is not endangered by tsunami flooding. 2.4.7 Ice Flooding and Landslides (HISTORICAL INFORMATION) Because of the location in a temperate climate, significant amounts of ice do not form on the Tennessee Valley rivers and lakes. SQN is in no danger from ice flooding. Flood waves from landslides into upstream reservoirs pose no danger because of the absence of major elevation relief in nearby upstream reservoirs and because the prevailing thin soils offer small slide volume potential compared to the available detention space in reservoirs. 2.4.8 Cooling Water Canals and Reservoirs (HISTORICAL INFORMATION) 2.4.8.1 Canals The intake channel, as shown in Figure 2.1.2-1, referenced in paragraph 2.4.1.1, is designed for a flow of 2,250 cfs. At minimum pool (elevation 675), as shown in Figure 2.4.8-1, this flow is maintained at a velocity of 2.7 fps. The protection of the intake channel slopes from wind-wave activity is afforded by the placement of riprap, shown in Figure 2.4.8-1, in accordance with TVA Design Standards, from elevation 665 to elevation 690. The riprap is designed for a wind velocity of 45 mph. 2.4.8.2 Reservoirs (HISTORICAL INFORMATION) Chickamauga Reservoir provides the cooling water for SQN. This reservoir and the extensive TVA system of upstream reservoirs, which regulate inflows, are described in Table 2.4.1-1. The location in an area of ample runoff and the extensive reservoir system assures sufficient cooling waterflow for the plant. S2-4.doc 2.4-31

SQN-21 2.4.9 Channel Diversions (HISTORICAL INFORMATION) Channel diversion is not a potential problem for the plant. There are now no channel diversions upstream of SQN that would cause diverting or rerouting of the source of plant cooling water, and none are anticipated in the future. The floodplain is such that large floods do not produce major channel meanders or cutoffs. Carbon 14 dating of material at the high terrace levels shows that the Tennessee River has essentially maintained its present alignment for over 35,000 years. The topography is such that only an unimaginable catastrophic event could result in flow diversion above the plant. 2.4.10 Flooding Protection Requirements Assurance that safety-related facilities are capable of surviving all possible flood conditions is provided by the discussions given in Paragraph 2.4.2.2, Section 3.4, Section 3.8, and Appendix 2.4A. The plant is designed to be shutdown and remain in a safe shutdown condition for any rainfall flood exceeding plant grade, up to the "design basis flood" discussed in Subsection 2.4.3, and for lower, seismic-caused floods discussed in Subsection 2.4.4. Any rainfall flood exceeding plant grade will be predicted at least 27 hours in advance by TVA's Reservoir Operations. Warning of seismic failure of key upstream dams will be available at the plant at least 27 hours before a resulting flood surge would reach plant grade. Hence, there is adequate time to prepare the plant for any flood. See Appendix 2.4A for a detailed presentation of the flood protection plan. 2.4.11 Low Water Considerations Because of its location on Chickamauga Reservoir, maintaining minimum water levels at SQN is not a problem. The high rainfall and runoff of the watershed and the regulation afforded by upstream dams assure minimum flows for plant cooling. 2.4.11.1 Low Flow in Rivers and Streams The targeted minimum water level at SQN is elevation 675, which corresponds to the lower bound of the winter operating zone for Chickamauga Reservoir. On rare occasions, the water level may be slightly lower (.1 or .2 tenths of a foot) for a brief period of time (hours) due to hydropower peaking operations at Chickamauga and Watts Bar Dams during the winter season. A minimum elevation of 675 must be maintained in order to provide the prescribed commercial navigation depth in Chickamauga Reservoir. The Preferred Alternative Reservoir Operating Policy was designed to provide increased recreation opportunities while avoiding or reducing adverse impacts on other operating objectives and resource areas. Under the Preferred Alternative, TVA will no longer target specific summer pool elevations at 10 tributary storage reservoirs. Instead, TVA tends to manage the flow of water through the system to meet operating objectives. TVA will use weekly average system flow requirements to limit the drawdown of 10 tributary reservoirs (Blue Ridge, Chatuge, Cherokee, Douglas, Fontana, Nottely, Hiawassee, Norris, South Holston, and Watauga) June 1 through Labor Day to increase recreation opportunities. For four main stem reservoirs (Chickamauga, Guntersville, Wheeler, and Pickwick), summer operating zones will be maintained through Labor Day. For Watts Bar Reservoir, the summer operating zone will be maintained through November 1. Weekly average system minimum flow requirements from June 1 through Labor Day, measured at Chickamauga Dam, are determined by the total volume of water in storage at the 10 tributary reservoirs compared to the seasonal total tributary system minimum operating guide (SMOG). If the S2-4.doc 2.4-32

SQN-27 volume of water in storage is above the SMOG, the weekly average system minimum flow requirement will be increased each week from 14,000 cfs (cubic feet per second) the first week of June to 25,000 cfs the last week of July. Beginning August 1 and continuing through Labor Day, the weekly average flow requirement will be 29,000 cfs. If the volume of water in storage is below the SMOG curve, 13,000 cfs weekly average minimum flows will be released from Chickamauga Dam between June 1 and July 31, and 25,000 cfs weekly average minimum flows will be released from August 1 through Labor Day. Within these weekly averages, TVA has the flexibility to schedule daily and hourly flows to best meet all operating objectives, including water supply for TVAs thermal power generating plants. Flows may be higher than these stated minimums if additional releases are required at tributary or main river reservoirs to maintain allocated flood storage space or during critical power situations to maintain the integrity and reliability of the TVA power supply system. In the assumed event of complete dam failure of the south embankment of Chickamauga Dam resulting in a breach width of 400 feet, with the Chickamauga pool at elevation 681, the water surface at SQN will begin to drop within one hour and will fall to elevation 641 about 60 hours after failure. TVA will begin providing steady releases of at least 14,000 cfs at Watts Bar within 12 hours of Chickamauga Dam failure to assure that the water level recession at SQN does not drop below elevation 641. The estimated minimum river flow requirement for the ERCW system is only 45 cfs.

Reference:

Programmatic Environmental Impact Statement, TVA Reservoir Operations Study, Record of Decision, May 2004. 2.4.11.2 Low Water Resulting From Surges, Seiches, or Tsunamis Because of its inland location on a relatively small, narrow lake, low water levels resulting from surges, seiches, or tsunamis are not a potential problem. 2.4.11.3 Historical Low Water From the beginning of stream gauge records at Chattanooga in 1874 until the closure of Chickamauga Dam in January 1940, the lowest daily flow in the Tennessee River at SQN was 3,200 cfs on September 7 and 13, 1925. The next lowest daily flow of 4,600 cfs occurred in 1881 and also in 1883. Since January 1942, low flows at the site have been regulated by TVA reservoirs, particularly by Watts Bar and Chickamauga Dams. Under normal operating conditions, there may be periods of several hours daily when there are no releases from either or both dams, but average daily flows at the site have been less than 5,000 cfs only 0.65 percent of the time and have been less than 10,000 cfs, 5.19 percent of the time. On March 30 and 31, 1968, during special operations for the control of watermilfoil, there were no releases from either Watts Bar or Chickamauga Dams during the two-day period. The previous minimum daily flow was 700 cfs on November 1, 1953. TVA no longer conducts special operations for the control of water milfoil on Chickamauga Reservoir. Since January 1940, water levels at the plant have been controlled by Chickamauga Reservoir. Since then, the minimum level at the dam was 673.3 on January 21, 1942. TVA no longer routinely conducts pre-flood drawdowns below elevation 675 at Chickamauga Reservoir and the minimum elevation in the past 20 years (1987 - 2006) was 674.97 at Chickamauga head water. 2.4.11.4 Future Control Future added controls which could alter low flow conditions at the plant are not anticipated because no sites that would have a significant influence remain to be developed. S2-4.doc 2.4-33

SQN-24 2.4.11.5 Plant Requirements 2.4.11.5.1 Two-Unit Operation The safety related water supply systems requiring river water are: the essential raw cooling water (ERCW) (Subsection 9.2.2), and that portion of the high-pressure fire-protection system (HPFP) (Subsection 2.4A.4.1) supplying emergency feedwater to the steam generators. The fire/flood mode pumps are submersible pumps located in the CCW intake pumping station. The CCW intake pumping station sump is at elevation 648. The entrances to the suction pipes for the fire/flood mode pumps are at elevation 651 feet 0 inches which is 32 feet and 24 feet, respectively, below the maximum normal water elevation of 683.0 and the normal minimum elevation of 675.0 for the reservoir. Abnormal reservoir level is 670 feet with a technical specification limit of 674 ft. For flow requirements of the HPFP during engineering safety feature operation (Reference 22). The ERCW pump sump in this independent station is at elevation 625.0, which is 58.0' below maximum normal water elevation, 50.0' below minimum normal water elevation, and 16' below the 641 minimum possible elevation of the river. Since the ERCW pumping station has direct communication with the river for all water levels and is above probable maximum flood, the ERCW system for two-unit plant operation always operates in an open cooling cycle. 2.4.11.6 Heat Sink Dependability Requirements The ultimate heat sink, its design bases and its operation, under all normal and credible accident conditions is described in detail in Subsection 9.2.5. As discussed in Subsection 9.2.5, the sink was modified by a new essential raw cooling water (ERCW) pumping station before unit 2 began operation. The design basis and operation of the ERCW system, both with the original ERCW intake station and with the new ERCW intake station, is presented in Subsection 9.2.2. As described in these sections, the new ERCW station is designed to guarantee a continued adequate supply of essential cooling water for all plant design basis conditions. This position is further assured since additional river water may be provided from TVA's upstream multiple-purpose reservoirs, as previously discussed during Low Flow in Rivers and Streams. 2.4.11.6.1 Loss of Downstream Dam The loss of downstream dam will not result in any adverse effects on the availability of water to the ERCW system or these portions of the original HPFP supplying emergency feedwater to the steam generator. Loss of downstream dam reduces ERCW flow about 7% to the component cooling and containment spray heat exchangers. ERCW flow does not decrease below that assumed in the analysis (analyzed as 670 to 639) until more than two hours after the peak containment temperature and pressure occurs. (See Section 6.2.1.3.4.) 2.4.11.6.2 Adequacy of Minimum Flow The cooling requirements for plant safety-related features are provided by the ERCW system. The required ERCW flow rates under the most demanding modes of operation (including loss of downstream dam) as given in Subsection 9.2.2 are contained in TVA calculations and flow diagrams. Two other safety-related functions may require water from the ultimate heat sink; these are fire protection water (refer to Subparagraph 2.4.11.6.3) and emergency steam generator feedwater (refer to Subsection 10.4.7). These two functions have smaller flow requirements than the ERCW systems. Consequently, the relative abundance of the river flow, even under the worst conditions, assures the availability of an adequate water supply for all safety-related plant cooling water requirements. S2-4.doc 2.4-34

SQN-21 River operations methodology for maintaining UHS temperatures are discussed in Monitoring and Moderating Sequoyah Ultimate Heat Sink, Reference 21. 2.4.11.6.3 Fire-Protection Water Refer to the Fire Protection Report discussed in Section 9.5.1. 2.4.12 Environmental Acceptance of Effluents The ability of surface waters near SQN, located on the right bank near Tennessee River Mile (TRM) 484.5, to dilute and disperse radioactive liquid effluents accidentally released from the plant is discussed herein. Routine radioactive liquid releases are discussed in Section 11.2. The Tennessee River is the sole surface water pathway between SQN and surface water users along the river. Liquid effluent from SQN flows into the river from a diffuser pond through a system of diffuser pipes located at TRM 483.65. An accidental, radioactive liquid effluent release from SQN would enter the Tennessee River after it reached the diffuser pond and entered the diffuser pipes. The contents of the diffuser pond enter the diffuser pipes and mix with the river flow upon discharge. The diffusers are designed to provide rapid mixing of the discharged effluent with the river flow. The flow through the diffusers is driven by the elevation head difference between the diffuser pond and the river [1](McCold 1979). Descriptions of the diffusers and SQN operating modes are given in Paragraph 10.4.5.2. Flow is discharged into the diffuser pond via the blowdown line, ERCW System (Subsection 9.2.2) and CCW System (Subsection 10.4.5). A layout of SQN is given in Figures 2.1.2-1 and 2.1.2-2. Two pipes comprise the diffuser system and are set alongside each other on the river bottom. They extend from the right bank of the river into the main channel. The main channel begins near the right bank of the river and is approximately 900 feet wide at SQN [1] (McCold, 1979). Each diffuser pipe has a 350-foot section through which flow is discharged into the river. The downstream diffuser leg discharges across a section 0 to 350 feet from the right bank of the main channel. The upstream diffuser leg starts at the end of the downstream diffuser leg and discharges across a section 350 to 700 feet from the right bank of the main channel. The two diffusers therefore provide mixing across nearly the entire main channel width. The river flow near SQN is governed by hydro power operations of Watts Bar Dam upstream (TRM 529.9) and Chickamauga Dam downstream (TRM 471.0). The backwater of Chickamauga Dam extends to Watts Bar Dam. Peaking hydro power operations of the dams cause short periods of zero (i.e., stagnant) and reverse (i.e., upstream) flow near the plant. Effluent released from the diffusers during these zero and reverse flow periods will not concentrate near the plant or affect any water intake upstream. The maximum flow-reversal during 1978-1981 were not long enough to cause discharge from the diffusers to extend upstream to the SQN intake [2] (El-Ashry, 1983), which is the nearest intake and located at the right bank near TRM 484.7. Moreover, the warm buoyant discharge from the diffusers will tend toward the water surface as it mixes the river flow and away from the cooler, denser water found near the intake opening below the skimmer wall. The intake opening extends the first 10 feet above the riverbed elevation of about 631 feet mean sea level (MSL). The minimum flow depth at the intake is approximately 45 feet [3] (Ungate and Howerton, 1979). There are no other surface water users between the diffusers and this intake. Subsection 2.4.13 discusses groundwater movement at SQN. Effluent released through the diffusers will have no impact on SQN groundwater sources along the banks of the river. Paragraph 2.2.3.8 discusses the effect on plant safety features from flammable or toxic materials released in the river near SQN. The predominant transport and effect of a diffuser release is along the main channel and in the downstream direction. The nearest downstream surface water intake is located along the left bank at TRM 473.0 (Table 2.4.1-4). S2-4.doc 2.4-35

SQN-21 A mathematical analysis is used to estimate the downstream transport and dilution of a contaminant released in the Tennessee River during an accidental spill at SQN. Only the main channel flow area without the adjacent overbank regions is considered in the analysis. The mathematical analysis of a potential spill scenario can involve: (1) a slug release, which can be modeled as an instantaneous release; (2) a continuous release, which can be modeled as a steady-state release; (3) a bank release, which can be modeled as a vertical line source; and (4) a diffuser release, which can be modeled either as a vertical line or plane source, depending on the width of the diffuser with respect to the channel width. The following assumptions are used in the mathematical analyses to compute the minimum dilution expected downstream from SQN and, in particular, at the nearest water intake.

1. Mixing calculations are based on unstratified steady flow in the reservoir. River flow, Q, is assumed to be 27,474 cubic feet per second (cfs), which is equalled or exceeded in the reservoir approximately 50 percent of the time (Paragraph 2.4.1.2). Because various combinations of the upstream and downstream hydro power dam operations can create upstream flows past SQN, a minimum flow is not well defined. Larger (smaller) flows will decrease (increase) the travel time to the nearest intake but cause less than an order of magnitude change in the calculated dilution.

2. Because the SQN diffusers and the nearest downstream water intake are on opposite banks of the river, and the diffusers extend across most of the main channel width, an analysis using a diffuser release (rather than a bank release) is selected to yield a lesser (i.e., more conservative) dilution at the intake. Thus, the accidental spill is modeled as a vertical plane source across the width of the main channel.

3. The contaminant concentration profile from a slug release is assumed to be Gaussian (i.e.,

normal) in the longitudinal direction.

4. The contaminant is conservative, i.e., it does not degrade through radioactive decay, chemical or biological processes, nor is it removed from the reservoir by adsorption to sediments or by volatilization.

5. The transport of the contaminant is described using the motion of the river flow, i.e., the contaminant is neutrally buoyant and does not rise or sink due to gravity.

The main channel and dynamic, flow-dependent processes of the reservoir reach between SQN and the first downstream water intake are modeled as a channel of constant rectangular cross section with the following constant geometric, hydraulic and dispersion characteristics. Longitudinal distance, x = 10.6 miles Average water surface elevation = 678.5 feet MSL (Figure 2.4.1-3 (1)) Average width, W = 1175 feet Average depth, H = 50 feet Average velocity, U (= Q/(W H)) = 0.468 feet per second (fps) Average travel time (for approximate peak contaminant), t (= x/U) = 1.4 days S2-4.doc 2.4-36

SQN-21 Manning coefficient n (surface roughness) = 0.03 Longitudinal dispersion parameter, alpha = 200 where: alpha = Ex / (H u) Ex = constant longitudinal dispersion coefficient (square feet per second) u = shear velocity (fps) = gRS g = acceleration due to gravity = 32.174 ft/s2 R = hydraulic radius (ft) S = slope of the energy line (ft/ft) The average width and depth were estimated from measurements of 9 cross sections in the reach [4] (TVA) [5] (TVA). For wide channels (i.e., large width-to-depth ratio), the hydraulic radius can be approximated as the average depth. The value of alpha = 200 is on the conservative (i.e., low) side [6] (Fischer, et al., 1979). The value of the Manning coefficient n is representative for natural rivers [7] (Chow, 1959). The equation used to describe the maximum downstream activity (or concentration), C, at a point of interest due to an instantaneous plane source release of volume V is [8] (Guide 1.113): C V

Co W H 4 E x t (2.4.12-1) where: Co = initial activity (or concentration) in the plant of the released contaminant

 = 3.14156 Any consistent set of units can be used on each side of Equation 2.4.12-1 (e.g., C and Co in mCi/ml; V in cf; W and H in ft; Ex in ft2/s; t in s).

The term, C/Co, is the relative (i.e., dimensionless) activity (or concentration) and its reciprocal is the dimensionsless dilution factor. Equation 2.4.12-1 simplifies to C/Co = 8.3E-10

V (V expressed in cubic feet (cf)) when the parameters are substituted and the Manning equation [7] (Chow, 1959) is used in the definition of the shear velocity, u. In the substitution, u = 0.028 ft/s and Ex = 282.1 ft2/s.

S2-4.doc 2.4-37

SQN-21 The equation used to describe the maximum downstream concentration at a point of interest due to a continuous plane source release rate, Qs, where Qs << Q, is [8] (Guide 1.113): (2.4.12-2) C Q

                 = s Co          Q Any consistent set of units can be used on each side of Equation 2.4.12-2 (e.g., C and Co in mCi/ml; Qs and Q in cfs).

Equation 2.4.12-2 simplifies to C/Co = 3.64E-05

Qs (Qs expressed in cfs) for Q = 27,474 cfs.

Examples of quantities and concentrations of potential contaminant releases and the use of Equations 2.4.12-1 and 2.4.12-2 follow. Because Co is defined as the in-plant activity (or concentration) and not that of the diffuser release, an estimate of the dilution of liquid waste occurring in the diffuser pond and diffuser pipes is not needed. This is because the flow available for dilution in the plant (e.g., CCW and ERCW) is taken from and returned to the river. Only effluent extraneous to the river flow requires consideration in the analyses to calculate the dilution. More information on the possible means which liquid waste from the plant enters the diffuser pond is contained in Subsection 10.4.5. The largest outdoor tanks whose contents flow into the diffuser pond are the two condensate storage tanks (Paragraph 11.2.3.1), which each have an overflow capacity of 398,000 gallons. Liquid waste that reaches the diffuser pond enters the Tennessee River through the diffuser system. The diffuser pond is approximately 2000 feet long and 500 feet wide with a depth that, although it depends on the Chickamauga Reservoir elevation, averages about 10 feet [9] (McIntosh, et al., 1982). The design flow residence time of the pond is approximately one hour (i.e., diffuser design flow is 2,480 cfs at maximum plant capacity [3] [Ungate and Howerton, 1979]). For example, assume an instantaneous plane source release into the Tennessee River of the contents of one condensate storage drain tank. Assume the full 398,000 gallon (53,210 cf) volume contains Iodine-131 (I-131) at an activity of 1.5E-06 mCi/gm (Table 10.4.1-1). From Equation 2.4.12-1, the activity, C, at the first downstream water intake would be 6.6E-11 mCi/gm, which is within the acceptable limit [10] (CFR) for soluble I-131. For a continuous plane source release, assume the contents of the 398,000 gallon (53,210 cf) floor drain tank leak out steadily over a 24-hour period. The effective release rate is 0.6 cfs at an activity of 1.5E-06 mCi/gm. The expected activity at the first downstream water intake would be 3.4E-11 mCi/gm using Equation 2.4.12-2 and is within the acceptable limit [10] (CFR) for soluble I-131. REFERENCES (for Section 2.4.12 only) [1] McCold, L. N. (March 1979), "Model Study and Analysis of Sequoyah Nuclear Plant Submerged Multiport Diffuser," TVA, Division of Water Resources, Water System Development Branch, Norris, TN, Report No. WR28-1-45-103. [2] El-Ashry, Mohammed T., Director of Environmental Quality, TVA, February 1983 letter to Paul Davis, Manager, Permit Section, Tennessee Division of Water Quality Control, SEQUOYAH NUCLEAR PLANT---NPDES PERMIT NO. T0026450. [3] Ungate, C. D., and Howerton, K. A. (April 1978; revised March 1979), "Effect of Sequoyah Nuclear Plant Discharges on Chickamauga Lake Water Temperatures," TVA, Division of Water Management, Water Systems Development Branch, Norris, TN, Report No. WR28-1-45-101. S2-4.doc 2.4-38

SQN-21 [4] TVA, Chickamauga Reservoir Sediment Investigations, Cross Sections, 1940-1961, Division of Water Control Planning, Hydraulic Data Branch. [5] TVA, Measured Cross Sections of Chickamauga Reservoir, 1972, Flood Protection Branch. [6] Fischer, H. B., List, E. J., Koh, R.C.Y., Imberger, J., Brooks, N. H. (1979), Mixing in Inland and Costal Waters, Academic Press, New York. [7] Chow, V. T. (1959) Open-Channel Hydraulics, McGraw-Hill, New York. [8] United States Nuclear Regulatory Commission, Office of Standards Development, Regulatory Guide 1.113 (April 1977), "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," Revision 1. [9] McIntosh, D. A., Johnson, B. E. and Speaks, E. B. (October 1982), "A Field Verification of Sequoyah Nuclear Plant Diffuser Performance Model: One-Unit Operation," TVA, Office of Natural Resources, Division of Air and Water Resources, Water Systems Development Branch, Norris, TN, Report No. WR28-1-45-110. [10] 10 CFR Part 20, Appendix B, Table II, Column 2. [11] TVA SQN Calculation SQN-SQS2-0242, SQN Site Iodine-131 Release Concentration in Tennessee River. 2.4.13 Groundwater (HISTORICAL INFORMATION) 2.4.13.1 Description and Onsite Use The peninsula on which SQN is located is underlain by the Conasauga Shale, a poor water-bearing formation. About 2,000 feet northwest of the plant site, the trace of the Kingston Fault separates this outcrop area of the Conasauga Shale from a wide belt of Knox Dolomite. The Knox is the major water bearing formation of eastern Tennessee. Groundwater in the Conasauga Shale occurs in small openings along fractures and bedding planes; these rapidly decrease in size with depth, and few openings exist below a depth of 300 feet. Groundwater in the Knox Dolomite occurs in solutionally enlarged openings formed along fractures and bedding planes and also in locally thick cherty clay overburden. There is no groundwater use at SQN. 2.4.13.2 Sources The source of groundwater at SQN is recharged by local, onsite precipitation. Discharge occurs by movement mainly along strike of bedrock, to the northeast and southwest, into Chickamauga Lake. Rises in the level of Chickamauga Lake result in corresponding rises in the water table and recharge along the periphery of the lake, extending inland for short distances. Lateral extent of this effect varies with local slope of the water table, but probably nowhere exceeds 500 feet. Lowering levels of Chickamauga Lake results in corresponding declines in the water table along the lake periphery, and short-term increase in groundwater discharge. When SQN was initially evaluated in the early 1970s, it was in a rural area, and only a few houses within a two-mile radius of the plant site were supplied by individual wells in the Knox Dolomite (see Table 2.4.13-1, Figure 2.4.13-1). Because the average domestic use probably does S2-4.doc 2.4-39

SQN-25 not exceed 500 gallons per day per house, groundwater withdrawal within a two-mile radius of the plant site was less than 50,000 gallons per day. Such a small volume withdrawal over the area would have essentially no effect on areal groundwater levels and gradients. Although development of the area has increased, public supplies are available and overall groundwater use is not expected to increase. Public and industrial groundwater supplies within a 20 mile radius of the site in 1985 are listed in Table 2.4.13-2. The area groundwater gradient is towards Chickamauga Lake, under water table conditions, and at a gradient of less than 120 feet per mile. The water table system is shallow, the surface of which conforms in general to the topography of the land surface. Depth to water ranges from less than 10 feet in topographically low areas to more than 75 feet in higher areas underlain by Knox Dolomite. Figure 2.4.13-2 is a generalized water-table map of SQN, based on water level data from five onsite observation wells, and in private wells adjacent to the site in April 1973, and also based on surface resistivity measurements of depth to water table made in 1972. Because permeability across strike in the Conasauga Shale is extremely low, and nearly all water movement is in a southwest-northeast direction, along strike, the Conasauga-Knox Dolomite Contact is a hydraulic barrier, across which only a very small volume of water could migrate in the event large groundwater withdrawals were made from the adjacent Knox. Although some water can cross this boundary, the permeability normal to strike of the Conasauga is too low to allow development of an areally extensive cone of depression. Groundwater recharge occurs to the Conasauga Shale at the plant site. Recharge water moves no more than 3,000 feet before being discharged to Chickamauga Lake. 2.4.13.3 Accident Effects Design features in SQN further protect groundwater from contamination. Category I structures in the SQN facility are designed to assure that all system components perform their designed function, including maintenance of integrity during earthquake. Buildings in which radioactive liquids could be released due to the equipment failure, overflow, or spillage are designed to retain such liquids even if subject to an earthquake equivalent to the safe shutdown earthquake. Outdoor tanks that contain radioactive liquids are designed so that if they overflow, the overflow liquid is redirected to the building where the liquid is collected in the radwaste system. Two outdoor tanks that contain low concentrations of radioactivity at times overflow to yard drains which discharge into the diffuser pond. Overflow liquid is discharged near the discharge diffuser. The capacity for dispersion and dilution of contaminants by the groundwater system of the Conasauga Shale is low. Dispersion would occur slowly because water movement is limited to small openings along fractures and bedding planes in the shale. Clay minerals of the Conasauga Shale do, however, have a relatively high exchange capacity, and some of the radioactive ions would be absorbed by these minerals. Any ions moving through the groundwater system eventually would be discharged to Chickamauga Lake. S2-4.doc 2.4-40

SQN-23 The Conasauga Shale is heterogeneous and anisotropic vertically and horizontally. Water-bearing characteristics change abruptly within short distances. Standard aquifer analyses cannot be applied, and meaningful values for permeability, time of travel, or dilution factors cannot be obtained. Bedrock porosity is estimated to be less than 3 percent based on examination of results of exploratory core drilling. It is known from experience elsewhere in this region that water movement in the Conasauga Shale occurs almost entirely parallel to strike. Subsurface movement of a liquid radwaste release at the plant site would be about 1,000 feet to the northeast or about 2,000 feet to the southwest before discharge to Chickamauga Lake. Time of travel can only be estimated as being a few weeks for first arrival, a few months for peak concentration arrival, and perhaps two or more years for total discharge. The computed mean time of travel of groundwater from SQN to Chickamauga Lake is 303 days. No radwaste discharge would reach a groundwater user. At the nearest point, the reservation boundary lies 2,200 feet northwest of the plant site, across strike. Groundwater movement will not occur from the plant site in this direction across this distance. During initial licensing, the radionuclide concentrations were determined for both groundwater and surface water movement to the nearest potable water intake (Savannah Valley Utility District, which is no longer in service) and found to be of no concern (see Safety Evaluation Report, March 1979, Section 2.4.4 Groundwater). 2.4.13.4 Monitoring or Safeguard Requirements SQN is on a peninsula of low-permeability rock; the groundwater system of the site is essentially hydraulically isolated and potential hazard to groundwater users of the area is minimal. The environmental radiological monitoring program is addressed in Section 11.6. Monitor wells 1, 2, 3, and 4 were sampled and analyzed for radioactivity during the period from 1976 through 1978. Well 5 was not monitored because of insufficient flow. An additional well (Well 6) was drilled in late 1978 downgradient from the plant and a pump sampler installed. Wells 1, 2, 4, and 5 are each 150 feet deep, Well 6 is 250 feet deep, and Wells L6 and L7 are 75-80 feet deep. All of the wells are cased in the residuum and open bore in the Conasauga Shale. 2.4.13.5 Conclusions SQN was designed to provide protection of groundwater resources by preventing the escape of the leaks of radionuclides. Site soils and underlying geology provide further protection in that they retard the movement of water and attenuate any contaminants that would be released. All groundwater movement is toward Chickamauga Lake. The Knox Dolomite is essentially hydraulically separated from the Conasauga Shale; therefore, offsite pumping, including future development, should have little effect upon the groundwater table in the Conasauga Shale at the plant. Even though the potential for accidental contamination of the groundwater system is extremely low, the radiological monitoring program will provide ample lead times to mitigate any offsite contamination. S2-4.doc 2.4-41

SQN-17 As a consequence of the geohydrologic conditions that remain unchanged from evaluations conducted in the 1970s, the information in Chapter 2.4.13 Groundwater is historical and should not be subject to updating revisions. 2.4.14 Technical Requirements and Emergency Operation Requirements Emergency flood protection plans, designed to minimize impact of floods above plant grade on safety-related facilities, are described in Appendix 2.4A. Procedures for predicting rainfall floods, arrangements to warn of upstream dam failure floods, and lead times available and types of action to be taken to meet related safety requirements for both sources of flooding are described therein. The Technical Requirements Manual specify the action to be taken to minimize the consequences of floods. 2.4.15 References

1. U.S. Weather Bureau, "Probable Maximum and TVA Precipitation Over The Tennessee River Basin Above Chattanooga," Hydrometeorological Report No. 41, 1965.

2. U.S. Weather Bureau, "Probable Maximum and TVA Precipitation for Tennessee River Basins Up To 3,000 Square Miles in Area and Duration to 72 Hours," Hydrometeorological Report No.

45, 1969.

3. Garrison, J. M., Granju, J. P., and Price, J. T., "Unsteady Flow Simulation in Rivers and Reservoirs," Journal of the Hydraulics Division, ASCE, Vol. 95, No. HY5, Proceedings Paper 6771, September 1969, pp. 15559-1576.

4. PSAR, Phipps Bend Nuclear Plant, Docket Nos. 50-553, 50-554.

5. Tennessee Valley Authority, "Flood Insurance Study, Hamilton County, Tennessee, (Unincorporated Areas)," Division of Water Resources, February 1979.

6. U.S. Army Engineering, Corps of Engineers, Omaha, Nebraska, "Severe Windstorms of Record,"

Technical Bulletin No. 2, Civil Works Investigations Project CW-178 Freeboard Criteria for Dams and Levees, January 1960.

7. U.S. Army Corps of Engineers, "Computation of Freeboard Allowances for Waves in Reservoirs,"

Engineering Engineer Technical Letter No. 1110-2-8, August 1966.

8. U.S. Army Coastal Engineering Research Center, "Shore Protection, Planning, and Design," 3rd Edition, 1966.

9. Reference removed per Amendment 6.

10. Hinds, Julian, Creager, William P., and Justin, Joel D., "Engineering For Dams," Vol. II, Concrete Dams, John Wiley and Sons, Inc., 1944.

11. Bustamante, Jurge I., Flores, Arando, "Water Pressure in Dams Subject to Earthquakes," Journal of the Engineering Mechanics Division, ASCE Proceedings, October 1966.

12. Chopra, Anil K., "Hydrodynamic Pressures on Dams During Earthquakes," Journal of the Engineering Mechanics Division, ASCE Proceedings, December 1967.

S2-4.doc 2.4-42

SQN-24

13. Zienkiewicz, O. C., "Hydrodynamic Pressures Due to Earthquakes," Water Power, Vol. 16, September 1964, pp. 382-388.

14. Tennessee Valley Authority, "Sedimentation in TVA Reservoirs," TVA Report No. 0-6693, Division of Water Control Planning, February 1968.

15. Reference removed per Amendment 6.

16. Cristofano, E. A., "Method of Computing Erosion Rate for Failure of Earthfill Dams," Engineering and Research Center, Bureau of Reclamation, Denver 1966.

17. "The Breaching of the Oros Earth Dam in the State of Ceara, North-East Brazil," Water and Water Engineering, August 1960.

18. NRC letter to TVA dated December 8, 1989, "Chickamauga Reservoir Sediment Deposition and Erosion - Sequoyah Nuclear Plant, Units 1 and 2."

19. Programmatic Environmental Impact Statement, TVA Reservoir Operations Study, Record of Decision, May 2004.

20. Updated Predictions of Chickamauga Reservoir Recession Resulting from Postulated Failure of the South Embankment at Chickamauga Dam; TVA River System Operations and Environment, Revised June 2004 (B85 070509 001).

21. Monitoring and Moderating Sequoyah Ultimate Heat Sink, June 2004, River System Operations and Environment, River Operations, River Scheduling (B85 070509 001).

22. SQN Calculation MDQ0026970001A, High Pressure Fire Protection Supply to the Steam Generators for Flood Mode Operation.

S2-4.doc 2.4-43

SQN-17 TABLE 2.4.1-1 FACTS ABOUT MAJOR TVA DAMS AND RESERVOIRS (HISTORICAL INFORMATION) Main River State Type Max. Length Drainage Length Area Lake Elevation Lake Volume (acre-feet) Useful Construction River of Height (Feet) area above of Lake of Lake (feet above sea level) Controlled Started Projects Dam (Feet) dam (miles) at Full Ordinary Top of Storage (d) (sq. mi.) Pool Ordinary Top of Fall Minimum Gates (Ac-Fl) (acres) Minimum Gates Pool (g) Elevation Elevation Kentucky Tenn. Ky. CGE 206 8,422 40,200 184.3 160,300 354 375 359 2,121,000 6,129,000 4,008,000 7-1-38 Pickwick Landing Tenn. Tenn. CGE 113 7,715 32,820 52.7 43,100 408 418 414 688,000 1,105,000 417,000 3-8-35 Wilson (f) Tenn. Ala. CG 137 4,535 30,750 15.5 15,500 504.5 507.88 507.5 582,000 641,000 59,000 4-14-18 Wheeler Tenn. Ala. CG 72 6,342 29,590 74.1 67,100 550 556.3 556 720,000 1,071,000 351,000 11-21-33 Guntersville Tenn. Ala. CGE 94 3,979 24,450 75.7 67,900 592 505.44 595 379.700 1,052,000 172,300 12-4-35 Nickajack (e) Tenn. Tenn. CGE 83 3,767 21,870 46.3 10,900 632 635 634 221.600 254,600 33,000 454 Chickamauga Tenn. Tenn. CGE 129 5,800 20,790 58.9 35,400 675 685.44 682.5 392.000 739.000 347,000 1-13-36 Watts Bar Tenn. Tenn. CGE 112 2,960 17,310 72.4 39,000 735 745 741 796.000 1,175,000 379,000 7-1-39 Ft Loudon Tenn. Tenn. CGE 122 4,190 9,550 55.0 14,600 807 815 813 282.000 393,000 111,000 7-8-40 TRIBUTARIES Tims Ford Elk Tenn. E&R 170 1,470 529 34 10,700 860 895 888 294.000 617,000 323,000 3-28-66 Appalachia Hiwassee N.C. CG 150 1,308 1,018 9.8 1,100 1,272 1,280 1,280 48.600 57,500 8,900 7-17-41 Hiwassee Hiwassee N.C. 307 1,376 968 22 6,090 1,415 1,528.5 1,524.5 71.800 434,000 362,200 7-15-36 Chatuga Hiwassee N.C. E 144 2,850 189 13 7,050 1,860 1,928 1,927 18.400 240,500 222,100 7-17-41 Ocoee No. 1 (f) Ocoee Tenn. CG 135 840 595 7.5 1,890 818.9 837.65 837.65 53.500 87,300 33,800 810 Ocoee No. 2 (f) Ocoee Tenn. RFT 30 450 516 ----- ----- ----- 1,115 1,115 ----- ----- ----- 5---12 Ocoee No. 3 Ocoee Tenn. CG 110 612 496 7 621 1,112 1,425 1,435 790 4,650 3,860 7-17-41 Blue Ridge (f) Toccoa Ga. E 167 1,000 232 10 3,290 1,590 1,691 1,690 12.500 196,500 184,000 11--25 (b) Nettely Nettely Ga. E&R 184 2,300 214 20 4,180 1,690 1,779 1,779 12.700 174,300 161,600 7-17-41 Melton Hill Clinch Tenn. CG 103 1,020 3,343 44 5,690 790 796 795 94.500 126,000 31,500 9-6-60 Norris Clinch Tenn. CGE 265 1,860 2,912 72 34,200 930 1,034 1,020 290,000 2,555,000 2,265,000 10-1-33 Tellico Little T. Tenn. CGE 108 3,238 2,627 33.2 16,500 807 815 813 321,300 447,300 126,000 3-15-67 Fontana Little T. N.C. CG 480 2,365 1,571 29 10,640 1,525 1,710 1,708 295,000 1,448,000 1,153,000 1-1-42 Douglas French Bread Tenn. CGE 202 1,705 4,541 43.1 30,400 920 1,092 1,000 84,500 1,490,000 1,105,500 2-2-42 Cherokee Holston Tenn. CGE 175 6,760 3,428 59 30,300 989 1,075 1,073 83,600 1,544,000 1,160,400 8-1-40 Fort Patrick S. Fork Holston Henry Tenn. CG 95 737 1,903 10.3 872 1,258 1,263 1,263 22,700 26,900 4,290 5-14-51 Boone S. ForkHolston Tenn. CGE 160 1,532 1,840 17.3 4,400 1,330 1,385 1,385 45,000 193.400 148,400 8-29-50 South Holston S. Fork Holston Tenn. E&R 285 1,600 703 24.3 7,580 1,616 1,742 1,729 121.400 764.000 642,600 8-4-47 (c) Watauga Watauga E&R 318 900 468 16.7 6,430 1,815 1,975 1,959 52,300 677,000 624,700 7-22-46 (c) Great Falls (f) (in Caney Cumberland Valley) Fork Tenn. CG 92 800 1,675 22 2,100 780 405.30 805.30 14,600 51,600 37,000 -15 TOTALS 638,353 8,621,490 23,732,359 15,110,860 PUMPED STORAGE Tenn. Tenn. E&R 230 ---------- --------- 520 1,530 ------- 1,672 2,000 37,800 35,400 7-6-70 Racoon Mountain

a. Foundation to operating deck. e. Nickajack Dam replaced the old Hales Bar Dam 6 miles upstream.

b. Construction discontinued early in 1926; resumed in March 1929. f. Acquired: Wilson by transfer from U. S. Corps of Engineers in 1933; Ocoee No. 1, Ocoee No. 2, Blue Ridge, and Great Falls by purchase from TEP Co. In 1939. Subsequent to acquisition, TVA heightened and installed additional units at Wilson.

c. Initial construction started February 16, 1942; temporarily discontinued to conserve critical materials during war. g. Full Pool Elevation is the normal upper level to which the reservoirs may be filled. Where storage space is available above this level, additional filling may be made as needed for flood control.

d. Abbreviations: CG - Concrete gravity dams. CGE - Concrete gravity with earth embankments. E - Earth fill.

E&R - Earth and rock fill. RFT - Rock-filled timber. T241-1

SQN-17 Table 2.4.1-2 FACTS ABOUT NON-TVA DAM AND RESERVOIR PROJECTS (HISTORICAL INFORMATION) Area Length Usefula Drainage Miles Maximum of of Storage ALCOA Area Above Height, Length Lake, Lake, Acre- Construction Projects River Sq. Miles Mouth Feet Feet Acres Miles Feet Started Major Dams Calderwood Little Tenn 1,856 43.7 232 916 536 8 1,570 1928 Cheoah Little Tenn 1,608 51.4 225 750 595 10 1,850 1916 Chilhowee Little Tenn 1,976 33.6 91 1,373 1,690 8.9 6,564 1955 Nantahala Nantahala 108 22.8 250 1,042 1,605 4.6 126,000 1930 Santeetlan Cheoah 176 9.3 212 1,054 2,863 7.5 133,290 1926 Thorpe West Fork (Glenville) Tuckasegee 36.7 9.7 150 900 1,462 4.5 67,100 1940 Minor Dams Bear Creek East Fork Tuckasegee 75.3 4.8 215 740 476 4.6 4,536 1952 Cedar Cliff East Fork Tuckasegee 80.7 2.4 165 600 121 2.4 698 1950 Mission (Andrews) Hiwassee 292 106.1 50 390 61 1.46 157 1924 Queens Creek Queens Creek 3.58 1.5 78 382 37 0.5 490 1947 Wolf Creek Wolf Creek 15.2 1.7 180 810 176 2.2 6,909 1952 East Fork East Fork Tuckasegee 24.9 10.9 140 385 39 1.4 906 1952 Tuckasegee West Fork Tuckasegee 54.7 3.1 61 254 9 0.5 35 1949 Walters (Carolina P&L) Pigeon 455 38.0 200 00000 870 340 5.5 20,500 a. Volume between elevations of top of gates and maximum drawdown. T241-2.doc

SQN-17 Table 2.4.1-3 Flood Detention Capacity TVA Projects Above Sequoyah Nuclear Plant Storage Reserved for Flood Control in Acre - Feet* January 1 March 15 Summer Project Elev. (Ft) Storage Elev. (Ft) Storage Elev. (Ft) Storage Tributary Douglas 940 1,251,000 958 1,021,300 994 237,500 Watauga 1940 223,000 1951.5 155,900 1959 108,500 South Holston 1702 290,200 1713 220,100 1729 106,100 Boone 1358 92,400 1369 60,400 1382.5 10,800 Cherokee 1030 1,011,800 1042 807,800 1071 118,100 Fontana 1644 580,000 1644 580,000 1703 73,400 Norris 985 1,473,000 1000 1,113,000 1020 512,000 Hiwassee 1465 270,200 1482 216,100 1521 35,000 Chatuge 1912 93,000 1916 73,300 1926 13,900 Nottely 1745 100,000 1755 79,100 1777 12,300 Tellico 809 92,000 809 92,000 813 32,000 Main River Fort Loudoun 809 85,700 809 85,700 813 30,000 Watts Bar 737 312,100 737 312,100 741 165,000 Total 5,874,400 4,816,800 1,454,600

2001 Conditions T241-3.doc

SQN-17 Table 2.4.1-4 PUBLIC AND INDUSTRIAL SURFACE WATER SUPPLIES WITHDRAWN FROM THE 98.6 MILE REACH OF THE TENNESSEE RIVER BETWEEN DAYTON TENNESSEE AND MEADE CORP. STEVENSON ALA. (HISTORICAL INFORMATION) Approximate Distance From Site Plant Name Use (MGD) Location (River Miles) Type Supply City of Dayton 1.780 TRM 503.8 R 19.1 (Upstream) Municipal Cleveland Utilities Board 5.030 TRM 499.4 L 37.6 (Upstream) Municipal Hiwassee RM 22.9 Bowaters Southern Paper 80.000 TRM 499.4 L 37.4 (Upstream) Industrial Hiwassee RM 22.7 & Potable Hiwassee Utilities 3.000 TRM 499.4 L 37.2 (Upstream) Municipal Hiwassee RM 22.5 Olin Corporation 5.000 TRM 499.4 L 37.0 (Upstream) Industrial Hiwassee RM 22.3 & Potable Soddy-Daisy Falling Water U.D. 0.927 TRM 487.2 R 7.1 (Upstream) Municipal Soddy Cr. 4.6 Plus 2 Wells Sequoyah Nuclear Plant 1615.680 TRM 484.7 R 0.0 Industrial East Side Utility 5.000 TRM 473.0 L 11.7 (Downstream) Municipal Chickamauga Dam # TRM 471.0 13.7 (Downstream) Industrial DuPont Company 7.200 TRM 469.9 R 14.8 (Downstream) Industrial Tennessee-American Water 40.930 TRM 465.3 L 19.4 (Downstream) Municipal Rock-Tennessee Mill 0.510 TRM 463.5 R 21.2 (Downstream) Industrial Dixie Sand and Gravel 0.035 TRM 463.2 R 21.5 (Downstream) Industrial Chattanooga Missouri Portland Cement 0.100 TRM 456.1 R 28.6 (Downstream) Industrial Signal Mountain Cement 2.800 TRM 454.2 R 30.5 (Downstream) Industrial Racoon Mount. Pump Stor. 0.561 TRM 444.7 L 40.0 (Downstream) Industrial Signal Mountain Cement 0.200 TRM 433.3 R 51.4 (Downstream) Industrial Nickajack Dam # TRM 424.7 60.0 (Downstream) Industrial South Pittsburg 0.900 TRM 418.0 R 66.7 (Downstream) Municipal Penn Dixie Cement 0.00001 TRM 417.1 R 67.6 (Downstream) Industrial Bridgeport 0.600 TRM 413.6 R 71.1 (Downstream) Municipal Widows Creek Stream Plant 397.440 TRM 407.7 R 77.0 (Downstream) Industrial Mead Corporation 4.400 TRM 405.2 R 79.5 (Downstream) Industrial

        # Water usage is not metered T241-4.doc

SQN-17 TABLE 2.4.1-5 Sheet 1 of 2 DAM SAFETY MODIFICATION STATUS (HYDROLOGIC) DAM *DAM MODIFICATION Year Completed Main River Dams Fort Loudon-Tellico Fort Loudon Dam embarkment was raised 3.25 with a concrete wall to elevation 833.25. A 2000-foot 1989 uncontorolled spillway with crest at elevation 817 was added at Tellico Dam. Watts Bar Embankment of main dam was raised 10 feet with earthfill/concrete wall to elevation 767. West 1997 Saddle Dike was not modified. Top of saddle dike remains at elevation 757. Nickajack South embankment was raised 5 feet with earthfill/concrete wall to elevation 657. A 1900-foot 1992 roller-compacted concrete overflow dam with top at elevation 634 was added below the north embankment. Guntersville Embankments were raised 7.5 feet with earthfill and concrete walls to elevation 617.5. 1996 T241-5.doc

SQN-17 TABLE 2.4.1-5 Sheet 2 of 2 DAM SAFETY MODIFICATION STATUS (HYDROLOGIC) DAM *DAM MODIFICATION Year Completed Tributary Dams Little Bear Creek Embankment was raised 4.5 feet. 1998 Beech Embankment was raised 4.5 feet with earthfill to elevation 475.5. 1992 Blue Ridge Three (3) additional spillway bays were added in 1982. Embankment was raised 7 feet with 1995 earthfill/concrete wall to elevation 1713, and a 320-foot uncontrolled spillway with crest at elevation 1691 was added in 1995. Boone Embankment was raised 8.5 feet with earthfill to elevation 1408.5. 1984 Cedar Creek Embankment was raised 5.5 feet with concrete wall to elevation 605. 1997 Chatuge Embankment was raised 6.5 feet with earthfill to elevation 1946.5. 1986 Cherokee A portion (600 feet) of the non-overflow dam was raised 7.75 feet to elevation 1089.75. 1982 Douglas A portion of the non-overflow dam was raised 13.5 feet to elevation 1022.5, and eight saddle dams 1988 were raised 6.5 feet with earthfill to elevation 1023.5. Nottely Embankment was raised 13.5 feet with rockfill to elevation 1807.5 1988 Upper Bear Creek Embankment was raised 4 feet with concrete wall to elevation 817. 1997 Watauga Embankment was raised 10 feet with rockfill to elevation 2012. 1983 Fontana Dam post-tensioned. 1988 Melton Hill Dam post-tensioned. 1988

These dam safety modifications enable these projects to safely pass the probable maximum flood (PMF).

Note: Plans are to armor the embankment at Chickamauga and Bear Creek Dams to permit overtopping. T241-5.doc

SQN Table 2.4.3-1 (Sheet 1) PROBABLE MAXIMUM STORM RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Index Rain, Pe,a Rain, Pe,b No. Area Inches Inches Inches Inches

1. Asheville 6.44 2.99 17.40 14.72

2. Newport, French Broad 6.44 4.04 18.50 16.51

3. Newport, Pigeon 6.44 4.04 19.30 17.31

4. Embreeville 6.44 4.04 15.10 13.11

5. Nolichucky Local 6.44 4.04 15.50 13.51

6. Douglas Local 6.44 4.86 17.10 15.88

7. Little Pigeon River 6.44 4.04 20.90 18.91

8. French Broad Local 6.44 4.19 18.60 16.81

9. South Holston 6.44 4.52 12.30 10.70

10. Watauga 6.44 4.04 13.30 11.31

11. Boone Local 6.44 4.04 14.10 12.11

12. Fort Patrick Henry 6.44 4.86 14.40 13.18

13. Gate City 6.44 4.86 12.30 11.08

14. Surgoinsville Local 6.44 4.86 14.60 13.38

15. Cherokee Local below Surgoinsville 6.44 4.86 15.80 14.58

16. Holston River Local 6.44 4.52 17.10 15.50

17. Little River 6.44 4.04 21.50 19.51

18. Fort Loudoun Local 6.44 4.04 17.60 15.61

19. Needmore 6.44 2.99 21.20 18.52

20. Nantahala 6.44 2.99 21.50 18.82

21. Bryson City 6.44 2.99 19.10 16.42

22. Fontana Local 6.44 2.99 20.70 18.02

23. Little Tennessee Local -

Fontana to Chilhowee Dam 6.44 2.99 24.00 21.32

24. Little Tennessee Local -

Chilhowee to Tellico Dam 6.44 4.04 21.00 19.01

25. Watts Bar Local above Clinch River 6.44 4.04 15.80 13.81

26. Norris Dam 6.44 4.86 13.80 12.58

27. Coal Creek 6.44 4.52 14.60 13.19

28. Clinch Local 6.44 4.52 14.90 13.49

29. Hinds Creek 6.44 4.52 15.30 13.89

30. Bullrun Creek 6.44 4.68 15.70 14.29 T243-1.doc

SQN Table 2.4.3-1 (Sheet 2) (Continued) PROBABLE MAXIMUM STORM RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Index Rain, Pe,a Rain, Pe,b No. Area Inches Inches Inches Inches

31. Beaver Creek 6.44 4.52 16.10 14.69

32. Clinch Local (5 areas) 6.44 4.52 15.30 13.89

33. Local above mile 16 6.44 4.52 15.30 13.89

34. Poplar Creek 6.44 4.52 14.90 13.49

35. Emory River 6.44 4.52 13.10 11.69

36. Local Area at Mouth 6.44 4.52 14.90 13.49

37. Watts Bar Local below Clinch River 6.44 4.52 14.40 12.99

38. Chatuge 6.44 2.99 21.40 18.72

39. Nottely 6.44 2.99 19.10 16.42

40. Hiwassee Local 6.44 2.99 18.90 16.22

41. Apalachia 6.44 2.99 17.90 15.22

42. Blue Ridge 6.44 2.99 22.10 19.42

43. Ocoee No. 1, Blue Ridge to Ocoee No. 1 6.44 4.04 18.30 16.31

44. Lower Hiwassee 6.44 4.19 15.20 13.41

45. Chickmauga Local 6.44 4.52 14.50 13.09 Average above Watts Bar Dam 6.44 4.20 16.34 14.56 Average above Chickamauga Dam 6.44 4.14 16.46 14.63 a.

Adopted API prior to antecedent storm, 1.0 inch. b. Computed API prior to main storm, 3.65 inches. T243-1.doc

SQN Table 2.4.3-2 UNIT HYDROGRAPH DATA Unit Drain Area, Duration, Q C T W W T AREA Name Sq. Miles Hours p p p 50 75 B 1 French Broad River at Asheville 945 6 15,000 .27 14 35 12 166 2 French Broad River, Newport to Asheville 913 6 35,000 .53 12 12 7 108 3 Pigeon River at Newporta 666 6 26,600 .56 12 11 6 78 4 Nolichucky River at Embreeville 805 6 27,300 .58 14 14 9 82 5 Nolichucky Local 378 6 10,600 .40 12 16 9 87 6 Douglas Locala 832 6 47,930 .27 6 8 6 60 7 Little Pigeon River at Sevierville 353 6 15,600 .62 12 10 6 102 8 French Broad River Localb 207 6 7,500 .51 12 11 8 60 9 South Holston 703 6 16,000 .53 18 24 17 100 10 Wataugab 468 6 17,700 .53 12 13 7 84 11 Boone Locala 669 6 22,890 .16 6 13 8 90 12 Fort Patrick Henry 63 6 3,200 .40 8 8 6 64 13 North Fork Holston River near Gate Citya 672 6 12,260 .60 24 33 25 108 14 Surgoinsville Localb 299 6 10,280 .48 12 13 9 66 15 Cherokee Local below Surgoinsvilleb 554 6 18,750 .48 12 14 7 66 16 Holston River Localb 289 6 6,800 .55 18 22 15 96 17 Little River at Mouthb 379 4 11,730 .68 16 14 8 96 18 Fort Loudoun Localb 323 6 20,000 .29 6 10 6 36 19 Little Tennessee River at Needmore 436 6 9,130 .49 18 23 12 126 20 Nantahala 91 6 3,770 .45 10 12 7 70 21 Tuckasegee River at Bryson City 655 6 26,000 .43 10 12 7 58 22 Fontana Local 389 6 16,350 .46 10 9 5 94 23 Little Tennessee River Local, Fontana-Chilhoweeb 406 6 16,900 .58 12 9 5 84 24 Little Tennessee River Local Chilhowee-Tellico Damb 650 6 17,000 .61 18 21 11 72 25 Watts Bar Local above Clinch Riverb 293 6 11,300 .30 8 9 7 84 26 Norris Dam 2912 6 43,300 .07 6 15 8 118 27 Coal Creekb 36.6 2 2,150 .64 8 9 5 40 28 Clinch Localb 22.25 2 1,350 .10 2 8 5 34 29 Hinds Creekb 66.4 2 3,620 .68 9 7 5 54 30 Bull Run Creekb 104 2 2,400 .47 14 21 14 84 31 Beaver Creekb 90.5 2 2,600 .58 14 14 10 88 32 Clinch Locals (5 areas)b 111.25 2 1,350 .10 2 8 5 34 33 Local above mi. 16b 37 2 4,490 .95 6 4 3 46 34 Poplar Creekb 136 2 2,800 .61 20 25 13 88 35 Emory River at Mouthb 865 6 34,000 .37 9 13 8 87 36 Local area at Mouthb 32 2 3,870 .95 6 3 2 46 37 Watts Bar Local below Clinch Riverb 427 6 16,300 .36 9 9 7 84 38 Chatuge Dama 189 6 13,570 .34 6 6 5 54 39 Nottely Dama 215 6 13,500 .29 6 5 4 80 40 Hiwassee Local 564 6 13,800 .36 12 18 12 124 41 Apalachia Local 50 6 2,900 .54 9 6 4 90 42 Blue Ridge Dama 232 6 11,920 .24 6 7 4 54 43 Ocoee No. 1 to Blue Ridgeb 363 6 17,000 .37 8 11 7 36 44 Lower Hiwassee 1087 6 32,500 .93 23 16 10 136 45 Chickamauga Locala 780 6 32,000 .38 9 14 7 36 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 a = Revised b = New T243-2.doc

SQN-17 Table 2.4.4-1 FLOODS FROM POSTULATED SEISMIC FAILURES OF UPSTREAM DAMS Plant Grade is Elevation 705 One-Half SSE Failures With One-Half Probable Maximum Flood Elevation

1. Norris a, b 698.1 b, c, d, e

2. Fontana 702.8

3. Cherokee-Douglas b, f 701.1

4. Fontana-Hiwassee-Apalachia-Blue Ridge b, e 707.9 SSE Failures With 25-Year Flood g,h

5. Norris-Cherokee-Douglas 706.0

6. Douglas-Fort Loudoun-Tellico b 699.3

a. Melton Hill fails from failure wave.

b. Watts Bar West Saddle Dike fails from failure wave.

c. Includes failure of five Alcoa dams - Nantahala upstream, Santeetlah on a downstream tributary; and Cheoah, Calderwood and Chilhowee downstream.

d. Fort Loudoun gates fail in open position.

e. Tellico fails from failure wave.

f. Failure wave overtops but does not fail Fort Loudoun.

g. Fort Loudoun gates blocked in closed position from failure of bridge. Failure wave would overtop and breach Watts Bar West Saddle Dike.

h. Failure wave overtops and fails Fort Loudoun.

t244-1.doc

SQN-17 Table 2.4.13-1 (Sheet 1) WELL AND SPRING INVENTORY WITHIN 2-MILE RADIUS OF SEQUOYAH NUCLEAR PLANT SITE (HISTORICAL INFORMATION) Estimated Map Well Elevation, Feet Well Ident. Location Depth, Water Dia., No. Latitude Longitude Feet Ground Surface Feet Remarks 1 35°13'34" 85°06'09" -- 725 -- .5 Serves 2 families; submersible 2 35°13'23" 85°06'12" 75 720 685 .5 Submersible pump 3 35°13'30" 85°06'47" 116 745 -- .5 Submersible pump 4 35°13'58" 85°05'45" 42 700 696 3.0 5 35°14'15" 85°06'25" -- 680 -- .5 1/4-hp pump 6 35°14'34" 85°06'46" 85 720 -- 15 Submersible pump 7 35°14'35" 85°06'52" 65 720 670 2.5 3/4-hp pump 8 35°14'36" 85°06'57" 73 735 687 .5 1/3-hp pump 9 35°15'06" 85°06'32" 27 780 761 5.0 Bucket 10 35°14'46" 85°06'16" 110 720 -- .5 Submersible 11 35°14'55" 85°06'15" -- 725 -- - 12 35°14'53" 85°06'13" 77 800 -- .5 13 35°14'52" 85°06'13" -- 800 -- - Summer home 14 35°14'50" 85°06'12" -- 800 -- - Summer home 15 35°14'45" 85°06'14" 50 720 680 .5 16 35°14'44" 85°06'18" 275 795 525 .5 1-hp submersible pump 17 35°14'45" 85°06'22" -- 740 -- .5 1-hp pump 18 35°14'21" 85°05'30" -- 695 -- - 19 35°14'26" 85°05'27" 200 695 -- .5 1-hp pump 20 35°14'34" 85°05'29" 150 695 -- .5 1/2-hp pump 21 35°14'31" 85°05'29" -- 695 -- .5 22 35°14'29" 85°05'29" 110 690 -- .5 1-hp pump 23 35°14'23" 85°05'32" 85 700 -- .75 1-hp jet pump 24 35°14'22" 85°05'40" -- 695 -- .5 Serves 2 familes; 1-hp pump 25 35°14'24" 85°05'46" 52 710 680 .5 3/4-hp pump 26 35°14'28" 85°05'45" 130 740 620 .5 27 35°14'26" 85°05'41" 90 740 710 .5 28 35°14'32" 85°05'44" 141 740 650 .5 29 35°14'34" 85°05'44" -- 735 -- - Summer home 30 35°14'38" 85°05'41" 58 700 670 .5 1/3-hp pump 31 35°14'41" 85°05'41" -- 720 -- .5 32 35°14'45" 85°05'46" -- 715 -- - 33 35°14'43" 85°05'47" -- 720 -- - 34 35°14'41" 85°05'48" -- 695 -- - Summer home 35 35°14'39" 85°05'50" 48 695 650 .5 1-hp pump 36 35°14'39" 85°05'53" 60 700 -- .5 Submersible pump 37 35°14'40" 85°05'58" -- 695 653 .5 1-hp pump 38 35°14'41" 85°05'56" 50 695 655 .5 3/4-hp pump 39 35°14'35" 85°05'54" -- 700 -- - Summer home 40 35°14'36" 85°05'57" -- 700 -- - 41 35°14'37" 85°06'01" -- 715 -- - Summer home 42 35°14'33" 85°05'02" 223 720 530 .5 NOTE: The information in this table is historic and not subject to updating revisions. T2413-1.doc

SQN-17 Table 2.4.13-1 (Sheet 2) (Continued) WELL AND SPRING INVENTORY WITHIN 2-MILE RADIUS OF SEQUOYAH NUCLEAR PLANT SITE (HISTORICAL INFORMATION) Estimated Map Well Elevation, Feet Well Ident. Location Depth, Water Dia., No. Latitude Longitude Feet Ground Surface Feet Remarks 43 35°14'46" 85°05'54" 65 695 655 .5 3/4-hp pump 44 35°14'47" 85°05'54" 95 705 655 .5 45 35°14'48" 85°05'53" -- 700 -- - Summer home 46 35°14'50" 85°05'53" 257 695 665 .5 1-hp submersible pump 47 35°14'52" 85°05'48" -- 710 -- - Summer home 48 35°15'04" 85°05'56" -- 725 -- - Summer home 49 35°15'06" 85°06'02" -- 720 -- - Summer home 50 35°15'06" 85°06'05" 90 705 625 .5 Submersible pump 51 35°14'58" 85°06'06" -- 695 -- - Summer home 52 35°15'01" 85°06'02" 65 720 680 .5 3/4-hp pump 53 35°14'47" 85°05'57" 46 700 670 .5 2 familes; 1-hp pump 54 35°14'42" 85°06'01" 48 695 675 .5 1/2-hp pump 55 35°14'41" 85°06'02" -- 695 -- - Summer home 56 35°14'40" 85°06'03" -- 695 -- - Summer home 57 35°14'37" 85°06'08" 155 690 670 .5 1-hp pump 58 35°14'34" 85°06'09" -- 695 -- - 59 35°14'23" 85°05'53" -- 760 -- .5 Submersible pump 60 35°14'49" 85°05'58" -- 705 -- - 61 35°13'01" 85°04'41" -- 720 -- - Summer home 62 35°13'18" 85°04'24" -- 845 -- .5 1-hp pump 63 35°13'19" 85°04'23" 206 845 645 .5 1/2-hp pump 64 35°13'33" 85°04'19" 50 720 680 .5 1-hp pump 65 35°13'49" 85°04'14" 100 720 640 .5 Servies clubhouse, 15 houses 66 35°13'57" 85°03'55" 175 741 -- .6 1-hp pump 67 35°13'53" 85°03'49" 100 738 690 .5 1-hp submersible pump 68 35°13'50" 85°03'52" 133 720 675 .5 1/2-hp pump 69 35°13'48" 85°03'43" 85 736 -- .5 1-hp pump 70 35°13'43" 85°03'38" 80 780 -- .5 1-hp pump 71 35°13'37" 85°03'36" 130 800 715 .5 1-hp pump 72 35°13'38" 85°03'43" -- 800 -- - Well not used 73 35°13'16" 85°03'30" 227 880 680 .5 Submersible pump 74 35°13'09" 85°03'41" 397 900 820 .5 2-hp pump 75 35°12'47" 85°03'58" 190 860 800 .5 Serves 2 families; submersible 76 35°13'03" 85°04'17" -- 720 -- - Summer home 77 35°13'05" 85°04'10" 90 740 670 .5 1/2-hp pump 78 35°12'50" 85°04'13" 85 760 -- .5 1-hp pump 79 35°12'45" 85°03'59" 190 880 -- .5 Serves 2 families; 1-hp pump 80 35°12'26" 85°04'07" 290 860 -- .5 Serves 5 families; submersible NOTE: The information in this table is historic and not subject to updating revisions. T2413-1.doc

SQN-17 Table 2.4.13-1 (Sheet 3) (Continued) WELL AND SPRING INVENTORY WITHIN 2-MILE RADIUS OF SEQUOYAH NUCLEAR PLANT SITE (HISTORICAL INFORMATION) Estimated Map Well Elevation, Feet Well Ident. Location Depth, Water Dia., No. Latitude Longitude Feet Ground Surface Feet Remarks 81 35°12'20" 85°04'33" 265 940 -- .5 Submersible pump 82 35°12'15" 85°04'34" 250 965 735 .5 1-hp submersible pump 83 35°12'24" 85°04'35" 305 965 665 .5 Submersible pump 84 35°12'22" 85°05'05" 135 740 690 .5 1-hp pump 85 35°12'21" 85°05'08" 120 740 -- .5 Serves 2 families; 3/4-hp jet pump 86 35°12'17" 85°05'06" 190 800 -- .5 3/4-hp submersible pump 87 35°12'23" 85°05'09" -- 740 -- .5 1-hp pump 88 35°12'16" 85°05'12" 55 740 720 2.5 Bucket 89 35°12'07" 85°05'09" 251 775 700 .5 Serves 2 families; 3/4-hp pump 90 35°11'54" 85°04'56" 170 980 -- .5 1/2-hp pump 91 35°12'19" 85°05'20" 125 740 705 .5 Submersible pump 92 35°12'22" 85°05'33" -- 725 -- - Summer home 93 35°12'22" 85°05'35" -- 700 -- - 1-hp pump 94 35°12'22" 85°05'36" -- 705 -- - Summer home 95 35°12'20" 85°05'44" -- 700 -- - Summer home 96 35°12'04" 85°05'56" 160 700 -- .5 Serves 5 families; 1-hp pump 97 35°12'04" 85°05'59 65 700 -- .5 House and cottage; 1-hp pump NOTE: The information in this table is historic and not subject to updating revisions. T2413-1.doc

SQN-17 Table 2.4.13-2 (Sheet 1) GROUND WATER SUPPLIES WITHIN 20-MILE RADIUS OF THE PLANT SITE (HISTORICAL INFORMATION) Approximate Average Distance Daily Use From Sitea Location Owner mgd Source (Miles)

1. Chattanooga Kay's Ice Cream Company 0.0400 Well 20.4

2. Chattanooga Selox, Inc. 0.0250 Well 21.0

3. Chattanooga Stainless Metal Products 0.0100 Well 16.4

4. Chattanooga American Cyanamid 0.0727 Well 21.0

5. Chattanooga Dixie Yarns, Inc. 0.5350 Wells (2) and Tennessee-American 13.3 Water Company

6. Chattanooga Scholze Tannery 0.1560 Wells (2) and Tennessee-American 24.0 Water Company

7. Chattanooga Southern Cellulose 4.0000 Well (1) and Tennessee-American 24.2 Products, Inc. 0.1000 Water Company

8. Chattanooga Alco Chemical Corporation 0.2300 Well (1) and Tennessee-American --

Water Company

9. Chattanooga Chattem Drug and Chemical 0.8500 Wells (3) and Tennessee-American 24.0 0.2380 Water Company

10. Chattanooga Cumberland Corporation 0.2380 Well (1) and Tennessee-American 17.4 0.0150 Water Company

11. Chattanooga Bacon Trailer Park Well --

12. Dunlap Bethel Church of Christ Well 20.0

13. Dayton Blue Water Trail and Well 19.0 Campground

14. Cleveland Cohulla Baptist Church Well 9.5

15. Dayton Crystal Springs Recreation Spring 19.0 Area

16. Georgetown Eastview School Well 9.5

17. Dayton Fort Bluff Youth Camp Well 19.0

18. Dayton Frazier Elementary School Well 19.0

19. Birchwood Grasshopper Church of God Well 11.3 NOTE: The information in this table is historic and not subject to updating revisions.

T2413-2.doc

SQN-17 Table 2.4.13-2 (Sheet 2) GROUND WATER SUPPLIES WITHIN 20-MILE RADIUS OF THE PLANT SITE (HISTORICAL INFORMATION) Approximate Average Distance Daily Use From Sitea Location Owner mgd Source (Miles)

20. Dayton Hastings Mobile Home Park Spring 19.0

21. Ooltewah High Point Baptist Church Well 10.0

22. Dayton Lake Richland Apartments Well 19.0

23. Dayton Laurelbrook Sanitarium School .017 Wells (7) 19.0

24. Cleveland Labanon Baptist Church Well 13.5

25. Cleveland Mt. Carmel Baptist Church Well 13.5

26. Sale Creek Mt. Vernon Baptist Church Well 11.0

27. Dayton Mt. Vista Mobile Home Park Wells (2) 19.0

28. Dayton New Bethel Methodist Church Well 19.0

29. Cleveland New Friendship Baptist Church Well 13.5

30. Dayton Ogden Baptist Church Well 19.0

31. Dunlap Old Union Water System Spring 20.0

32. Dunlap P.A.W., Inc. #2 Well 20.0

33. Cleveland Red Clay State Historic Area Well 13.5

34. Chattanooga Riverside Catfish House Well 25.0

35. Cleveland Robert Allen Well 13.5

36. Dayton Salem Baptist Church Well 19.0

37. Dunlap Sequatchie-Bledsoe VO- Well 20.0 Training

38. Dayton Seventh Day Adventist Church Well 19.0

39. Chattanooga Shamrock Motel Well 20.1

40. Dayton Sinclair Packing House Well 19.0

41. Dunlap Stonecave Institute Water 0.0064 Spring 20.0 System

42. Dunlap Old Union Water System Spring 20.0

43. Sale Creek Sale Creek Marina Well 11.0 Multiboating NOTE: The information in this table is historic and not subject to updating revisions.

T2413-2.doc

SQN-17 Table 2.4.13-2 (Sheet 3) GROUND WATER SUPPLIES WITHIN 20-MILE RADIUS OF THE PLANT SITE (HISTORICAL INFORMATION) Approximate Average Distance Daily Use From Sitea Location Owner mgd Source (Miles)

44. Sale Creek Sale Creek P.U.A. - TVA Well 11.0

45. Sale Creek Sale Creek Utility District 0.204 Wells (2) 10.8

46. Graysville Graysville Water Supply 0.220 Wells (2) 15.0

47. Graysville Graysville Nursing Home Well 15.0

48. Dayton Dayton Golf & CC % Mokas Well 19.0

49. Birchwood Birchwood School Well 11.3

50. Cleveland Cassons Grocery Water System 0.0170 Well 19.7

51. Cleveland Black Fox School Well 13.5

52. Cleveland Blue Springs Baptist Church Well 13.5

53. Cleveland Blue Springs School Well 13.5

54. Cleveland Bradley Limestone, Div. of 0.2400 Well 13.5 Dalton Rock Product Co.

55. Cleveland Hardwick Stone Company 0.1130 Well 13.5

56. Cleveland Cleveland-Tenn. Enamel 0.2240 Well 13.5

57. Cleveland Magic Chef, Inc. 0.4200 Spring 13.5

58. Hamilton Savannah Valley U.D. 0.720 Wells (2) 5.0 County

59. Hamilton Eastside Utility District 3.0130 Wells (3) and Tennessee American 7.9 County 0.0920 Water Company

60. Hamilton Hixson Utility District 4.0000 Cave Springs (3) and Tennessee 12.9 County 0.3330 American Water Company

61. Soddy Union Fork Bakewell, U.D. 0.192 Wells (3) and Sale Creek 9.8 0.0010 Utility District

62. Hamilton Walden's Ridge, U.D. 0.471 Wells (2) 17.4 County

63. Hamilton Container Corporation of 1.9200 Well 22.0 County America

64. Hamilton Dave L. Brown Company 0.0200 Well --

County NOTE: The information in this table is historic and not subject to updating revisions. T2413-2.doc

SQN-17 Table 2.4.13-2 (Sheet 4) GROUND WATER SUPPLIES WITHIN 20-MILE RADIUS OF THE PLANT SITE (HISTORICAL INFORMATION) Approximate Average Distance Daily Use From Sitea Location Owner mgd Source (Miles)

65. Hamilton De Sota, Inc. 0.0750 Well --

County

66. Hamilton Hamilton Concrete Products 0.0050 Spring 24 County

67. Cleveland Thompson Spring Baptist Well 13.5 Church

68. Dayton Vaughn Trailer Park Well 19.0

69. Dayton Walden's Ridge Baptist Well 19.0 Church

70. Dayton Walden's Ridge Elementary Well 19.0 School

71. Cleveland White Oak Baptist Church Well 13.5

72. Bradley Bockman Childrens Home Well 10.2 County

73. Catoosa Catoosa County U.D. Well 19.0 County a River mile distance from differences (TRM 483.6) for supplies taken from the Tennessee River channel; radial distance to other supplies.

NOTE: The information in this table is historic and not subject to updating revisions. T2413-2.doc

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5 15 25 5 15 25 5 15 25 5 15 25 5 15 2 5 5 15 25 5 15 2 5 5 15 2 5 5 15 25 5 15 2 5 5 15 2 5 JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV E,EC Gage zero = 6 2 1 . 1 2 (1929 genl. adj) NOTES: Complete to :>ct. 1, 1985 U S W B f l o o d stage = 3 0 feet T h e stages s h o w n f o r the 1867 f l o o d and the highest f l o o d m 1875 are f r o m w e l l authenticated f l o o d marks. Drainage area = 2 1 , 4 0 0 sq m i A l l o t h e r stages, except i n 1 8 7 4 and 1875, are r e p o r t e d ( ) o r estimated ( * ) crests f r o m observations at the same d a t u m and location, W a l n u t Street. T h e stage readings i n 1874 and 1 8 7 5 were based o n a d i f f e r e n t d a t u m , b u t t h e estimated crests were corrected t o be c o m p a r a b l e t o later records. Stages since O c t o b e r 22, 1 9 1 3 are n o t HISTORICAL c o m p a r a b l e w i t h earlier ones because o f the b a c k w a t e r e f f e c t o f Hales B a r D a m , 3 5 miles downstream. A change Figure 2.4.2-1 F l o o d Distribution i n H a l e s B a r S p i l l w a y i n 1948, the closure o f N i c k a j a c k D a m i n December 1967, and subsequent r e m o v a l o f Hales B a r D a m f u r t h e r affected Chattanooga stages, m a k i n g later stages incomparable t o earlier periods. Diagram Chattanooga, Tn. S i n c e M a r c h 4, 1936, w h e n upstream regulation began, b o t h c o m p u t e d natural ( o ) and reported crests are s h o w n

                                                                                                                                                                 - o n the y e a r l y chart. These natural crests are based o n c o n d i t i o n s when T V A was established, and hence are R e v i s e d b y Amendment 17 c o m p a r a b l e t o stages f r o m 1913 t o 1948. O n l y natural crests since M a r c h 1936 are s h o w n o n the seasonal diagram.

I 1

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                                                                  !I Watershed sh6wn is Tenqeaeee Valley abovelchattanooga.

HISTORICAL

I Ptgure 2.4.3-1 Probable maximum Mlir h lohyet* (2l.400*1q. mi. downetream).

l*t 6 hour* (In.) Revised by Amendment 17

waterehed shown ia Tennessee Valley above Chattanooga. Storm arbitrarily ceAtered at M!Ohee,

+-

poses. t Tennessee* tor illustrative pur* t** HISTORICAL Flgure 2.4.3-2 Probable 111ut11m March t*ohyete (7980 sq.

t . ) , lat 6 houre (in.)

Revised by Amendment 17 J J

r Best Available Historical Image l 80 r,. C 11:C 60 I ..J f e I r L

     ... 40 20 o..__ _ _ _ _.__ _ _ _ _ _ _ _ _           ""-_ _ _ _ _ _ _ _ _ _ _ _ _           ___.

l2 24 16 48 60 TIME- HOUR$ RAINFALL-TIM£ DISTRIBUTION ADOPTED STANDARD MASS CURVE HISTORICAL Revised.by Amendment 17

9.8 Best Available Historical Image HISTORICAL Revised bY Amendment 17 9,8 1-'1 gurii 2 .I, ._J-11

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           -----               AREA t , FRENCH BROAD RIVER AT ASHEVILLE. 91'5 SQ. Ml.

lREA 2, FREMCH BROAD RIVER. NEWPORT TO ASHEVILLE. 913 SQ. Ml.

           -------             AREA 3, PIGEON RIVER AT NEWPORT. 666 SQ. Ml.
           -         --        AREA ti-.NOLl CHUCKY Rt VER AT EMBREEVI LLE. 805 SQ. Ml.
           -       ---         AREA S . NOLICHUCKY' LOCAL. 378 SQ. Ml.

HISTORICAL Revised by Amendment 17

                                                                         &*HOUR UNlT HYDROGHAPHS SHEET l OF II FIGURE 2 . 4 . 3

6

501------+----+------+---- ----""t---- r ru 401--i ----4-----+------+-----+.-----r-i--- g

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T S M E - OAYS LEGEND:

                  ----        AREA 6. DOUGLAS LOCAL. 8 3 2 SQ. Ml.
                  -      -AREA     7  LITTLE PIGEOH RIVER. 35-3 50. MJ-.
                  ------AREA       8. FRENCH BROAD RIVER LOCAL, 207 sq. Ml.
                  ---AREA          9, SOUTH HOLSTON DAMt 7 0 3 SQ. Ml.

HISTORICAL 6-HOUR UNITHYOROGRAPHS l SHEET 2 OF Tf

                                                               ;_,. _____       F1_G_u_R_1:z._4 __3_-_s___      _ _ _ , j II .

Revised by Amendment 17

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2 3 4 6 TIME-DAY S LEGEND:

                    -----              AREA   I0 ,  WATAUGA DAM, i<<i8 sq. Ml *
                    --         --      AREA   I I,  BOONE LOCAL, 669 SQ. Mf.
                    ---      - - - - AREA     J2 . FORT PATRICK HEMRY LOCAL, 63 SQ. M l .
                    --        - -      AREA   13,   N. F. HOLSTON R. NR GATE CJTY, 6 7 2 SQ. HI.

r HISTORICAL Revised by Amendment 17

        *--                                                                       &-HOUR UNtT HYDROGRAPHS SHEET J OF II FIGURE 2 . 4 . 3 - 6

25 j 20 r r U') LL <.) I 1 --*--- I\\ l 0 ' I 0 15 0 I r I I i - et: f Jr\ ! -r-JO 1I \i <.) (I) [.'"

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_____,_ I 0 0 2 3 4 5 6 TIME-DAYS LEGEND:

      ----               AREA       Ill-. SURGOIMSYILLE LOCAL :- 299 SQ. Ml.
      --        -        AREA       IS.      CHEROKEE LOCAL BELOW SURGOINSV1LL£, SSEJ SQ. Ml.
       ------            AREA       16,      HOLSTON RIVER LOCAL. 289 SQ. Ml.
      -       - -         AREA      17       LI-TTL£ RIVER AT MOUTH, 379 SQ. Ml.
      ----               AREA 18, FORT LOUDOUN LOCAL. 323              sq. Ml.

HISTORICAL

Revised by Amendment 17

                                                                               &-HOUR UNIT HYDROGRAPHS SHEET 4 0 F II FIGURE 2 . 4 . 3 - 6

30,----,----,----..----,,_------- 25 I \

              * ( I ) 20 I t I     '

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                            -        - -              AREA 19.

TtME-DAYS LITTLE TENNESSEE R. AT NEEIM>RE. q36 SQ. Ml.

                            --        ---             AREA 20 7 NAJITAHALA. 91 SQ. ML.
                            -     --                  AREA 21 t TUCKASEGEE R. AT BRYSON CITY, 655 SQ. Ml.
                            ----                      AREA 22,   FONTANA LOCAL. 389 SQ. Ml.

r HISTORICAL Revised by Amendment 17

                                                                                                   &-HOUR UNIT HYDRO GRAPHS SHEETS O f II FIGURE 2 . 4 . ' 3 - 6

I 50r-----r----,----...,...----,-----T---- I 40t-t-;------,r-----i ----;--------t------i-----1 I f' I t-- - - .t - - =-.r - - - - +-- - - - - - - - - ...J

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5 6 LE6EID: AREA 23. LITTLE TEllNESSEE R. LOCAL, FONTANA TO CH l LHOWE£. ll06 SQ. MI *

                             --       -        AREA 211,, LITTLE TENNESSEE R. LOCAL, CHI LHOWE£ TO 1 0 TELLl CO DAM, 6 5 0 SQ. Ml.
                        * ----                  AREA 2 5 , W A m BAR LOCAL AIOYE CLIIICH RIVER. 293 SQ. Ml.
                             -*-               AREA 2 6 NORRIS DAM, 2912 SQ. Ml.

HISTORICAL Revised by Amendment 17 I-HOUR UNIT HYDRDGRAPHS l SHEET&OF U FIGURE 2 . 4 . 5 - 6

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                       **    \      I                 -...:*
                             **..'*t-.
                                  * ** :i:-,

I

                                                                            *--.. . *r---,;;;. -

2 3 4 5 6 i J M E - DAYS LEGEND:

           ------                       AREA T l ,                COAL CREEK, 3 6 . 6 SQ. Ml.
           -----                        AREA 29,.                 HINDS CREEK, 66.IJ SQ, Ml.
           ---*-                        AREA 3 0 ,                BULL.RUN CREEK, IOIJ SQ. Ml.
           ----                         AREA 3 1 ,                 BEAVER CREEXI' 9 0 . 5 SQ. Ml.
           -*-*****--*-                  AREAS 28                 AND 3 2 , a.INCH RIVER LOCAL AREAS, 22. 2 SQ. 'Ml.

r HISTORICAL Revised by Amendment 17 2-HOUR UNIT HYDROGRAPHS SHEET7 OF II FIGURE Z.4.l-6.

l 5--------------i-----..------,.---- ----. 4 1-1-----1------+----- +-----+- -----!:------t g 31-1 --...;..----+------;. -----+------, i- - - - O l\ ;I l

                                                                           ;            j l
'                 I a:                I \
               /        \
             ,- I \ \

21--'-+--/.*- * ':.-----+-----+*-----+------+----i i c5 I I \ I \

                               \ ..

I I ...... ,

                                                  -.j,.

i . . . ........... 2 3 4 5 6 TIME-DAYS LEGEMD:

     ----                AREA 33, LOCAL AREA AIOVE Ml. 16, 37 SQ. Ml.
     -----               AREA SIJ, POPLAR CREEi., 136 SQ. NI.
     -        --         AREA 36, LOCAL AREA AT MOUTH, 32 SQ. Mt.

HISTORICAL Revised by Amendment 17 2-HOUR UNIT HYDROGRAPHS SHEET 8 OF IJ FIGURE 2 . 4 . 3 - 6

C 35

             "           I I

l I g0 2 0 I i I I (\ I l

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2 3 4 s 6 TIME - DAYS LE6EJID:

         ---AREA             35. EMORY RI VER AT MOUTH. 865 Sf>. Ml.
         -     -AREA         3 7 , WATTS BAR LOCAL BELOW CLINCH RIVER, 1127 SQ. Ml.

HISTORICAL Revised by Amendment 11

                                                                     &-HOUR UNIT HYDROGRAPHS S1tEET 9 OF II FIGUR£ t.4.3 - 6

35 I 30 25

                                                        - ]-                   l             II 020                                                       I               . i 0

I ! w I a: 15 '

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            '                                                                  l l '-

tl\--..:::-- ~t-----,-----; ____ I I I 0 0 2 3 4 5 6 Tl E -DAYS LEGEND:

           ---                            lREA S8, CH.lTUGE DAM, 190 sq. Ml.

AREA 3 9 , MOTTELY DAM, 215 SQ. Ml.

           - -          -     -          AREA lJO,. H l WASSEE LOCAL. 5 6 SQ. MI *
           -          * -                AREA I

APALACH I.A. SO SQ. M l .

HISTORICAL 7 l Revised by Amendment

                                                                                   &-HOUR UNIT H'fDROGRAPHS SHEET 10 OF II FIGURE 2 . 4 . 3 - & **

35 r 30 t t /1 25 I\/

                \ r I
                       \

I I ti) I&.. u I 0 20 0 \ 2 l I

  &I.I                     1 a:

C,

                             \
                              \
                               \

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LEGEND:

           ----        AREA         2,  I W E RIDGE DAM, 232 SQ. Ml.
           --          AREA        113. OCOE£ NO. I TO B1.UE RIDGE DAM. 363 SQ. Ml.
           --     --   AREA         . LOWER HIWASSEE LOCAL. 1087 SQ. Ml.
           -     *-    AR£A        q5,  Cit I CKAMAUGA LOCAL, 780 SQ. MI

HISTORICAL Revised by Amendment 17

                                                                           &-HOUR UN1T MYDROGRAPHS SHEET 11 OF 11 FIGURE 2..4. " 3 - 6

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                                                                                           -                             ;:-;,-...,CMICUMAUGA DIICffAIIR                        MU*4?1.0                                                       i I      I          I                I           l          I                                             J l4 l*IS*1' IZ        14
                                                                                                                                   ?a 1&     24        12
                                                                                                                                                   >17-TS 24
                                                                                                                                                                                 **11 IZ          t4               IZ 3-f9*1'S 24           12 S-iD-7.S
                                                                                                                                                                                                                                                     -14 FfGURE 2,4.3-T 1'73 Fl.OOO*CHfCKAMAUGA RESERVOIR UNSTEADY FLOW MOCEL VERIFICATION HISTORICAL Revised by Amendment 17

(

  , 00  ,------------.------.-----t--------'--r---,---

P O S T - CHICKAMAUGA 7401-----t---------1...------+ - - - FAILURE CONDITIONS 73 P R E - CHICKAMAUGA o --------FAILURE CONDITIONS I I UNSTEADY FLOW ij 7 2 0 1-----1--M_O_D_E_L-tC_O_M_PU_t_l\T_I__.O_NS ---::aoS-*"'---- - + - - - - - - - - - - ....,......- ... RANGE OF HISTORICAL , ...... STANDARD STEP BACKWATER

   ?IO        KNOWN FLOODS             --            COMPUTATIONS 7001--------H" -----1 ---f-----t------t------t----1-----t 690 - - - -- - - - - --'----- - - - - - - - - ---- - ---- - - - - - - -

200 400 600 800 1000 1200 1400 1600 DISCHARGE -1000 CFS STEADY - STATE MODEL VERIFICATION WATTS B*AR DAM TAILWATER RATING CURVE J FIGUR r- 2 . 4 . 3 - 8 HISTOhJL Revised by Amendment 17 .J

300 0 ) u Cl jIl 200 0 C C 0 0, 0

tOO 0:

C a, 0 15 16 17 18 19 20 MARCH

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        =-=

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0 15 15 IT 18 19 20 MARCH HYDROL0GIC MODEL VERIFICATIO N-1973 FLOOD FIGURE 2 : 4 . 3 - 9 HISTORICAL Revised by Amendment I 7 t

300 0 . 0 g oafE D i,. I I 200 l - -COMPUTED r-7--(-*- Ill 0 a:: C \.. ---- l

             .,c lJ u
             =

100

                         ----,1-Vi       l I

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                               ]1                                     I I

I i l i ' 0 II 12 13 14 15 IG !7 18 MARCH 300 u 0 I 200 a flJ g lOO 0 II IZ 13 14 15 16 17 18 MAifCH HYDROLOGtC MODEL VERIFICATION - 1963 FLOOD F*GURE 2 . 4 . 3 - 1 0 HISTORICAL Revised by Amendment 17

1 1,400,000 - . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - , Best Available Historical Image 1,200,000 * * * * * * *

1,000,000 000,000 * .. * * * * * *
ca

        .c u

en 600,000 400,000 - * - * - - * *

I - *** *'-" - J .

200,000 * * * * * * * *

O-l-----1-----1-----1-----lr-----1-------1--------1--------1------ -------1----

3/21 3/22 3/23 3/24 3/25 3/26 3/27 3/26 3/29 3/30 3/31 Sequoyah Nuclear Plant Probable Maximum Flood Discharge Figure 2. 4. 3-11 Revised by Amendment 17

725 - . - - - - - - . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - --- --------,------- ....------------ **-- 720 * .* * * * * * * * * , * . . - . . . . . ..- - ** r 715* * * * * * * * *

1 * * * * * * * * * .....

710 * ,. * * * * * * *

c 705 *-- Plant Grade Elevation 705 J ......... ..

0 700 * * * * * * * * *

695 * * * * * * * * *
690 , .... . . . ... . *
r 1* * . *,* -... - . . - ,-
1 * * * * * *

                                                                                                                                                                     * , f .   .   .   .   .   .   . . .  .

680 3/21 3/22 3/23 3/24 3/25 3/26 3/27 3/28 3/29 3/30 3/31 Sequoyah Nuclear Plant Probable Maximum Flood Elevation Revised by Amendment 17 Figure 2.4.3-12

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                              ,.                                                                THIS DRAWING WAS CREATED FROM TVA
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                                                        =o       o=                             LIDAR/CAD DATA CONTAINED IN CALCULATION CDQ0000002D13000057 REV. I ATTACHMENT 3
                                     ' "                                                        -------------------- RMM o=
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                                                      "    AREA SC                                                 .CZ:      PMP FLOW DIRECTION
                                                                                                                          \

Q SCIII.Et 1 N C H " 1DD FEET SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 2.4.3-13A GENERAL GRADING FOR SITE DRAINAGE (REVISED BY AMENDMENT 26) I CAD MAINTAINED DRAWING I

I,

               ,,,-!..,...,-J    E R C W PUMP STATION I
                                 \
                                  \

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                                      µ " ° SKIMMER WALL 1*4 CHANCE FLOOD ELEVATION AI I

N ERCW p ump Station Location _FIGURE 2 *4 .tJ- _; 14 419

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Figure 2.4.3-16

SfCTKJN A I ,Z(J' SECTION

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i UNITS - ....._ ' I

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              ;1 i

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                                                                                   / 7 S ' l Force TYPICAL              8LOCK- SPILLWAY
                                    -                             610'                                                                                    T f f=.5' a7            11 1 I 111111 ii I Ill i 111111 I I -

TW+ HJ/V_-JW Hff= - NOT.5-5: 5_ 9 I OB E a,r//2--7,1d,lre ;,*Jer t/a UPLIFT DIAGRAM El .50 fore e s a s s limed d5 0 0 9 g UPLIFT PRESSURE ASSUMED TO horizonta/ly anc' 0 0 6 g ACT ON 100% OF 8ASE AREA i/er -I1cdi* v a f fhc b a s e N o t e A: r- a n d a m p l ! f i e d up fl-ie

                                                                                                                                                                                                                                                           *s + r u c f u r e .

ll;e powerhouse and sp1/hvay 2. Splllwa_y 9 a f e s were sfrocft1r s are well keyed 1"t-Jto rhe rock roundaf1t!')r1. The r o c k a s s u m e d open l'or f h i s f"orm,;;fior;s are seyere(y t°c>lded analysi=:.. wifh fhe dtp qe,r;eral/_y in a d/Jsfr direction J/aryinq rrom /{J fo 40 °. Q A11y ra1!t1re J"YOU/d r e q w r e cross /;ed shear o r the rock. Rock or /-IJ/s !3ASE P L A N A T El (;;50.0 type /2<Js cross b e d s h e a r s f r e n q f h r 32. 79'

I iI I' mwch 1:.-1ex-cess or f l ; a f r e q d fe,r aFac/-or o f ' S a r e f y o r / . !f____

I!') !' ! I I 1 li T

                                                                                                                                                                                                '              52.2{a 1 4 4 - - - - - - -

1 I 1 I l l . _ _ L3AS£ PR£SSUR£S 57'

                                                                                                         *r'ormu/a,                                                 't-.

Shear, sJ fhcJf is r e q d ror Q:./ 1"s CcJ/cu/i:5fed from shear - rr1cfie'Jn Scale 1"=40' O.; 0* 66

                                                                                                                                         ; : ; s 4. A  is  assumed              ro     b e e a f i re. f;as             area.

BASE PRES3UR£5 PO/IVl:RJ-I0U3£ &SPILLWAY IAvq ; s R e q d l l IMR Ir rticr:JI R£SUL T5 OFANALYSIS FOR f" :EH ! II . f max 1FS,7i!o- ...:,heal" on L. // 1.5hear; ,,; tor 0 4 "'" C PlnnP-A-A OP£RJJTING 811515 [ARTI-IQUAK£ 12 207 20,232 0.9<"a f"L7.

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                                                                                   .l 9 P5 /

(2!.5psi, ; I:'

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                                                                                                               ,,               *47.* 2 p51 50801< 1374931( 2.37 24pst !7.2psi I

Nate A

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CJUl.0 * . *..(i.ti _______ . ! TKD. . . _ --*--- tNC.l*[U

D . . ,

          ,-...,_._-,------r--1                                                                                                                                                                                                                                         FIGURE 2.4.4-1 Revised by Amendment 6.

( ( ( 760 __ LJ

                                                                                                                                        . d:-!.....,__

i

                                                                                                                                                   !:._-£;_1_15-
                                                                                                                                                                                       .   - -l"1-i Best Available Historical Image
                                                                                                                                                                 *f       12 3 Noh,s:

LAm1/y$1) w.t3 made us;n9 11te 140 *- -.L- --- sf4nd41rd slip e,"r-cle mofhod. I. ror tne oriqinc,I 11161/ify

                                        ,,.e, si, .$40 W11IIJ Silr Compvfotions E*23*2* o n d       ..

i uo;::: --*-**--* -- -**

                                      "Ht,porl Oil Soil Tests dnd Sld/Jll,"fy                                         J An.g(rsi,, It,    /Jor ct1rfh                                                   '-I Err1iunt'rmenf

f&,:,orl IJ,. 9*115.

                                     .1 Sl;e,vr s/reng/'nt of m11ferit1ls st1me - , , U'Jfld i n ori9"1dl fJJ'hJl'Y,515.                                           - * - - - - -                     - *

u - - - J - - - - 1 0.°'9 ock round11lion cl.

ASSUMED EARTJ.IQUAK£ /NU/TIA FORCS 0 aos9 a 1q o.J!J9 oe9 rw C'I ?135 SECTION N-N ( S T A .30 1- 0 i ) HISTORICAL F R O M ORKJtNAl ANALYSIS Revised by Amendment J7 Sci!lle I*.!JO' FfJtU'e 1 ,4, 4-2

                                                                                                                                                                      !n ank nt V tls hnr UJh, RllS1JlL1J u( Aualydo fuf    an $$£

WATTS BAR EL 752.0:? rw El. 694. s 2 5 YEAR FLOOD SCAI.E r*80' r EL 7 5 2 . 0 , NW 74.7 TW EL 713.S Et. 7/.3.0 j PROBABLE MAX FLOOD SCALE I"* 8 0 # r Figure 2.4. -3 Spillway Gate Positions fo 25-Year Flood - 1/2 Probable Ma.. ictum Flood Revised by Amendment 17

t-eg ,E/822 f / f f El 8/7 I I

                                                                               -1 5/Jdr,S 1/2,-;l /s r qd -For O=I is c"lctJl tl      F/-4m .shevr-l'ridion a-formulr1 ., - tJ.6S EY..SA       .,A is Z#                                               10.5..S' assumed le; be! Afire o/*eo.

a;

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                                                                      ..J ;(-,JI- Sheor .sfress s-, ret1tl /"<Jr                                 £:JASE PLAN El 7aJ I

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                                 - - - . - -. - -A- - - . - - - - - - , - ; z : : : z : : - : : : V , eir-cI y          !H v9   Is-Fo,-r0::./
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            - . .L - - - - - - - - l -

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     '2_,76.?Ir' 2?,.f4 II(                 ft.Ii          '/.f.6 ,IJSY Z6 p s i   81(        /3dZK,    0.75"             lr3.)0p5J) ;ti .J9,osi        1.26      17,PS/

I Figure 2.4.4 Powerhouse and Sp llway F o r t Loud:-"'1 Dam I Results o f Analys .s for 1/2 SSE

                                                                                                                . i -I 8...;;.'()-r--;li-="tJP O r PAM £f / 8.JO I
                                                                                                                                                                 ----*-7 8.10       . ---* - - - - + - - - -

Best Available Historical Image 8A --t-----lf--L-J_- * - ' - - - - ' - - - - . - 09 0.()29 0.0 9 /JOl,gOaJgO.IIJg 0.129 fJ./4,().l&g 0.189 ASSUMED ACC£1.elM71QN NfJTES: I Analysts w,u m,tk 11siNJ Nte sf ol'tl .sl/o ,i,clo ,,,_11,o;_

2. Sited,. .11-,-,nvlll.r a ,.,,,,,Mis

                                                                                                                                             .SMld as vst1,' b tlf"1'qindl d11t1l)'sis.
                                      *--=--*

TWEI ?9.rJ ..

                                                   /
                                                       /_
                                                            "/   ]j rel 7.Jo                                                 HISTORICAL Revised hy Amendment 17 TYPICAL EM8ANKNENT SECTION                                                        F!Jurc:2,4,4-5 EmbankllC!nt, Fort Loudoun Dao, Results 0£ ARlliyata for 1/2 SS$

Spillway 9oles o,,oen 0.46q El 799. 3 ) ,. 17 rTffEI 79L_ TWE/793 07 I II I I I I _E/7,?0 El 7/7!: 675' A I /lydrc;srdr'/C l-/,{./n"4y,/Jah//C i tJ.09'] f-- rec. rc1rci=fh To --Accd rd/fi?/l Cdrr/2 ,,vo,A'e_ (}r/Jdm TYJJICAL SPILLWAY SECTIO;V TPY-1-Jiff-TM= 5./3 K;{,,cadh ue rc v. 4 f Tff= 475%, I:: i!:Ii I! Ii 11:1II:;:: Ii. i. i. Ette-f/J,'j/:::6.25 3/4. UPLIFT LJIA6RAM El 7 / 7 UPL;r l Pl?ESSVl?E ASSuMELJ Ta ACT ON / 0 0 % or l3A.5- AREA

                                                                                                                                                                                                                  !VOTES:

I Ye;-/*[d/ .,_. e/.f c*d /,?,rJ c f /'lO/laYerf/oy-/ i l df!cl sp1//Wd_y a / ..!dse a:,-Sv'/l/ed f be O.0 6 q, B_/ d//'.10r .,c, a/'ie1 *ysi: ;, Jn_ p i//r_ ,/h :1,

                                                                                                ;if- S/2eat;s ., f/2af is ret;'d ror 0=/ /s                                                                       of' dcce/c, r J l"*,_    *.-, '-.,b::;,,re Me,b,;;.:,e WJs c a/culaled rrom shear- rr1c f/o/J                                                                           d /6'//l :-,.,,,d re:-/;(! ,!,,?q d i l":::1.; tor l/;t rorrm.Jla .,, 0= 0.652 V-,iSA,A is n                                                                                                             //c/lcYerf'low :,ec t-ion dno' 0 , / ? q a l
                                                                                                                          .ZII                                                                                     file c?,O Tor i/1<' *p.1//,1-Vd, r .

assumed to be en fire: area,

                                                                                                                                                                                                             ,2,,1/c?riz:c?/J,d/      dC-Cel r al - r/of f!Onaverfk,,,.,;

d/Jd S/J/1/Wd / d r' hd-;e .,-"'5st/m e d /-o ,6 e

                                                                                                -if S h e a r .sfress., s,, r e q d f' r Q= I                                                                    c, 9q. 8_ y dy//o/l)iC a/J.:1lysl_.,_,cJhlpltf1i.;_,1;,

675' j con.s1deri/Jq porl/o/J or base ir; 90.66 1 o f Jcce/ r,, -,-. aOtPYe f/,e i;a ':. C0/7,if)/"'essin ir;sfead or enfire d-<_-'//JJ!iied rv /J 0.499 a / /I; 1.-, - rc,, i!Je r,o,'?:Yer flow -;ec-1/n an<?"C. 46 q base area. BASE PLAN AT El 7 1 7 cJf- f!Jer,::,o &,rfile s,,oiJlw&J

3. S,r:;1I/J¥ci/ qu/-e-, assvmed ope!? rcr

                                                                                                                                                                                                                   /1!1\ c,/,'7,i.-/ *, * ,

11

                                                                                                                                                                                                        ; G ..8_' _ _ _ _S_c_a_ l_ e_ _/ _=4 0 ,
                                                                                                                                                                                                   .i-1.t/                                                                       _

L. 79 q.3' f3ASE PRESSURE NONOP'c-l?FLOJ1/&'S.PILLWA Y

,_:         }** lj,rj,1,v\                 \          \ _*_                                                                                                                                                            q*.:_-5Ui TS OF4NAL YSIS rc,q

'" ., \, "*'. '. *" . 1.- . J A..-q 5 Reqd .ZMRJlerfical OPERAT/IVG !3ASIS EA/r,T/10tl4)( Z..Y Zfl Shear, 5 For Q.:c I f max F.S.=- _11 /1 .Shear cm Zmo_Plane A-A ISU... I I . f t * - - - * - - * * * * * . .- - TELLICO DAM 16p s / 43P5 /i'-7'f 40p.s DnN/.f-f J ****-------------***- CNII.D . . .

                                                                 / 602)( 97151(.' 0.72                                                                                         1    / . /6       40psi

'°"' . . TIC.I . . . . . (NOHllll

f. 9ps// FIGURE 2 . 4 . 4 - 6 Revised by Amendment 6.

EL 8d0 T&POF OAM EL 8d0 EL 8 2 0 - -- - -+- -

                                                                     - t - - - - - - - - . - - - - - - - - - - - +!- -   l----l--- --t--*-  +- -    I---
                                                                                                                                              'RIZ EL 8 0 0                                                                           VRECTi
                                                    $'.)                                I  YERTICAL F'ACTOI? OF SAFETY,                        EL 7&0                           /JIRECTION--
            ;I::::_*:./. 2& (Ayeraqe orR &s fesf.s)

EL 7 6 0 EL 7 4 0 EL 7 2 0 011 (}.(Jtl9 0.04q 0.0690.0:!9 O.KJ 01.2!( 0149 0./69 0.I8g ASSUME£) ACCELEl?ATION EL 830 NOTES-' TWEL 7.3 --- -\ssu.MEt> SAT//l?AT/ON I.Analy is was made us1nq the I LINE f sfant:/41rd shj, c1r,d6 method. 2 .Shedr l r e n g f/1$ of' ma ferials same /JS used in ori9incJ/ i an,alysis., Revised by Amendment l TYPIC4L EMBANKMENTSECTION Figure 2.4.4-7 Embankment - _Tellico Dam Results' of Analysis for l/2 SSE

                                                                                                                                              -    - - -        = - - -

0.7S(l - 0.6.fg _ _ _ E._I_J._'06/.2__ _ _ _ _ _ _ _ _i 1"9-*

                                                                                                                                                                                ---=-E( 1 0 6 ! . i _ _ _               r---1
                                                                                                                                                                             *O
                                                                  /IW El 10.36 HWc//Od6 Iner f;,:, r--c Due        <!.

To Edr f-lu;,v.lre -----=::,.,.....'--._

                                  . I Tl'VE/851
                                                                                                                                                                .c1eoo L        'A      210*

J FJ1o:r,te:1t:r.o.st,1fic 7VPICAL SPILL WAY .SECTION rw3.19 Ksr 1! !1

                                            !1:! llt8IIIW*        HJY=I-<< 7 sKsr TW-=-dl9Ast HW-=l-73 Ksr UPLIFT 0/AGl?AM El              00                                                                                            UPLIFT 0/AGRMf c/lKJO                     M:JTES:

1./PLIFT Pl?e55tll?E ASSilMELJ lJPLIFT PNES5tll'(c ASSLJ 1.Yer-llc,1I dCCeMr<1h"'1 or nt:JntJflt!!rfll,,dlfd TtJ ACT ON JOO%OrBASE AREA spillWdy afh c dSSII"'*"' f o k a,. TO ACT 0 # / a ? OF BASE A/feA .iy tiyn.,11k tN'J.lysis, ,.,,pliFlc.<1l1iM; *! dccelcr,1h11n .-e Nie Ails* wcr., rI I dn-1r11>inetl fo /Je 0.1 , ,1-I- f f f'r Mentll'lt'NfYF*,r 3<<N -,,,/ -1

f f'!Je t.,p -,.,. fll* spilll'flay. i 2.Ho,izonlal acceter,1li,,,n d /10'10Nfihl

                                                                                                                                                                                                     .,,,,/ ,pillWdY df bM* iJUmed fob*

1-- ?!P' J *s/Jeo,; fi>tJf is req-lor t)=/ 15 0.() 'I 9. /Jy dyn,Mt ,1Nl)'s;s 1 i _,,,./iFictl'li,n of ;,a::ele rt!lli"1 <Mo.,.-

                            .BASE PLAN AT E/800                                                                                                                                                      fht! 1¥:se Ht!ls defern,in<<I fc 6.

ct!J/cul tl!d /i-"m 5Aeal"- lhdio,, O.U of the fbP lir" fht! /10!10wr!'b# r-49. 21 Frmv/8, a 0. 5 zy-, SA 1 ,4 IS secfit:tn d'1ti tJ. ?S9 al- IJe lqo it l"h* spi//lllldY-

                 - L l[J.)D1"'"L BASE PRESSUliE 16tJ. 8      .J                       **                            N ass"ou,d lo h enlir* dreo.

She.:,,;s, re'l(I for G-=/

                                                                                                                                                                                                .s ..5,P1tlll'*.J1' g4/es t!Jsd
                                                                                                                                                                                                     -M<::S    ysi.
                                                                                                                                                                                                                                         .:w,en r 4c"i C#nsitlerin<J po,tkn or /Jase in
                 !                                                                         compres.si,:,n(nc >'el1Si"") ,. 11-eod of entire htlse area.

XY XH U l A"" s -f7d Shea,;s F,,. Q.r/ Vm. MN Ytlr-r,c*I F . S = - SN.tr on Ellto lnJJ A-A zy ZH

                                                                                                                                  %,'I v

A ' s Aatl'i

                                                                                                                                        .SM!M",S 'For a*I       ,,,,.,      rM, l'*f' IC*/

Fs-- - .Sliur ,,, Mo :AlA ...... A - A 85'psi ,i2psi* 95psi 9psi* l,h'IJ,11si /.OJ 5Hpsi llK ;lf,¥8TK 118 it-'-'e} (/71,si}** psi 1.25 217psi 2/t!J/ K 278 1( I.P K Al2: 'i{,(ISpyJ-l!i Figure 2.4.4-8 Spillway & Nonover low Norris Dam, Results of Analysis for 1/2 SSE

I I __ L -£. fcNlca;re A s s t t m e cO l'/I I rai/ 1/'J o r i q posit" r e m a i f J Si n o r i gp o s i i i d n d

                                                                                                 .swy I                                        Approx 4 s s v m e < Ih c a l i e m or d e i J r i s i o c / 2 - 1 n n e l a f ' k r /;;1/ure
                                          )

I I I A s s 11m ed µ , s , I / P n o f " d e / J ri s i n d>t9 1M

                                                                                                                                                 >        I af'l-e r f',;)/11r"                                 SECT/ONA-A
                                                                                                                                                         .ASS/,/,1/E /) P ( ) S / T I O t V O r , ? I / L e i ) S P I L U ' Y A Y S
                                                                                                                                                                                                                                       !AIJ!AR}

( / l ( ) / l t J Y E J ' ; r (JIIJI f f J A M cC P O W E . & l / , Q / 5 £ , A S S / , / M c O 665' A 970 Apprc.x E / A s s v m e dl o c a f i o ' 1 of" c/2a/)l')e/ t l e c r i si n i l a r l e r T d " 'r c A ....J ELEJ/ATION Norri *. Dam, A n a l y s i s f or 2 SSE & One H a l f F i g u i1 2 . 4 . 4 - 9 Maxi m Possible Flood

                                                                                                  ---         E / 1 0- 2 IIW/OT5 lnt:!r fia Force OU<!

Tc £*rf/Jt?Vdlr *- L - F--- .--i - . _T...,..W 9--:c6-c0r---c:- _ _ _

                =7+==------Tf=                                      - - - ;:- - - - - L 6

I / .(_ ... - - - - - - -

                                                       /01,( ,,
       "--L.......i....::...L...-_:.....u..l                              _ ___..__ _
                                                        -,,,+-*-E._/..,.;900           _ _......___...__.__....._-:-- --                                                       NOTES-*
                                                                                                                                                                              /.J'er/ica/ t!ICCelert11t1'<,n     or   e i,.

1,-: -- Hy rcs/- lc spi//lfdf 4#' HJt! />dsJ M"SU/11 f., 0 . A /3tJ9/' __, Force. h)'"O"r. Fcrce/)u. 0 C) 9 g- 0. 0 6 g . B y dylldmid ,1n;&*s1 To c...-/-!lv,., e. Accf!!l*ral'iot1 O f M1p/i/°/ef#1'1M , I ' adceler4lkv113001'1/J 0*1n Ou To tJSe /Jase WdS ,111-eJmin-,tl , i J

TYPICAL SPILLWAY SECTIQIVrw IIW:_TW E,,H,,,,u,IA-.,

0-1/ g . I - M p. i

2. h'oriz,:,nf ,1CCet,,._f-n 111' .

T f f = . 3 . 7 5 3/4 * ((i I *!' 'i'

                                                                                * -*,    flff=ltJ.9./                          3/4*                                                                           I
                                                                                                                                                                               $pi,fWI.>' 111 M e /Jas

d-SSum11d .,,..., b
0 . 0 , 9 . B / ;_,,;,: .:MIYJ.&is, ampltfictdlNJ OT */-,r,:,-1-ion l!J611Y* L>ds* w4s ,leftr,11in,ul U P L ! FT 0 / A G l ? A M E l 9 0 0 to b e a . s - s 9 crl niie f<¥>-

U P L / F T P R E S S U R E ASSUMEO TO .3..5,Pi/lWtlf ( l d f " s ass} t:I o,,"" f'*r A C T ON 1 0 0 % O F B A S E AREA 1-liJs don.lyris. 13'5. 9 / ' BASE P L A N A T E l JtJO

S h e a r , s , IJ,a/- i s r e 9 d l b r 0 = 1 is C < 3 / c v l a l e qr r o m s h e a r - /rli:/-1on f o r m u l a., ' 2
0 . 6 . 5 % / / . , . S - I' A i s N

a s s u m e d ' ,,.., 6 e A h r e ar<!a. 88.2/' ;fShet11r s f r e s s., s r " e q d ' r o r Q , 1 Ct:Jnsu:lertnq portion o / '. b d s e i n 8ASE PReSSt//?E c c m , P r es s i o , ; 1 i u fe c f 'e d i r t J

                                                                                                              / J d s e r,rt d-Figure zlt..4-10 Spillway        _.&  Nonoverflow Cherokee Dam Results f Analysis f o r 1/2        SE

FAClOII IYSArErY, /".S.* O B S 0 9 00.09 O.°"; 0 9 a OA4, a oN9 01&9 al!l9 ASSVMfO ACCELE/UTfO,V

                                                        ....:.....:;_---:i-:-

NOTJf"S: I.An,1/rsis was ""9.1,: vsln9 f/,d st,;n<lt!lrd sli,D circlo nx,Mod.

2. Siu,.,, sfren9Ms: ot" ndi!.)("ls

                                            ¢,.;>6 *; c,o                                              S
                                                                                                       .<SnT-e d.S t1sdd 1"1' ortg,irJPI 1 Y, *.57.SPCr                                               dn<1/ysis.

WEAKEST C/RtU" El. ,<!JO HISTORICAL Revl,.d hy Amcnd111cn1 17 TYPICAL. EMSAMti1fENT s e c Tltw F l j u r e 2.,.,-11 EnihnkJDOnt, Cherokc"I D:a11 1 Results o f Anill)'sb for 1/2 SSE

                                                                                                                                              .It   .............

IAOOI.C OAM I

                                                                                                               .900'1

&:cP

                    ,:,AR r .'

rv._:-\..'E. FlC'W [.,C;*;,1,;y '11-,'£ A M ELcVATICt A,*_ A .t.,,_,1., .S,.C.ILLWAY TO /;4IL r

                                                                                                     / ST
                                                                                                                 £* if'j,O:;

r* , - * [._

                                                     -         I
                                                                                                    -:,\:

i- -- *r':,-**. ":--. t*,.,_,:.

            . * - - . : . , ;,c.* . . . . . - - , . '
                                                -*----,-y_      -                   I                          ...
..     .............,..                   ., '                 -                -     StJq'      2a          j SECTION                   8-B                                                     SECTION   .  -

(._ r Figure 2 4 4-12 Che r o k c e . Dam,

                                                                                                                                              . . Assumed Contlit10 . l l O f D*         '1m. Af t '-'*r Failure? /2 SSE and 1 : Pu'-siLlc l / L. "bximum, Flood

11.S,ttdr,S, M#I 11, F1r Q*I is Cdf I',.,_ Vie.Ir .,,.,.fiolt form,,/*, Q* 9-.1!]};-'.s.1.,' is

    , l , s , m , , d A,       ;,,,     ,,. ,,,.

' s,, ., slrtt1s,4 , r.,. tJ,t cus11ui,,f1 ,,,,,, ,, of/Jws* ,;.

(A -lttllli ol"ttlfll,tt l1st1 dnt,J. HISTORICAL Revised by Amcndnicm 17 FIOURE2.4.4-IJ Spillway and o.. rnow DouglasD1111 Resulls o f Analy*ls for1/2SSE

Best Available Historical Image J()JS filCr 0,:- SAFErV I F.S.

I.O 995 9

i lo;

                                                                                                           'JSS
                                                              '                                            9.3S' I                                                ,,1_4---__,_-+-_ _,,__ _ e - . - ........- - - . . - - - - 1 I

09 oo O/J#t fJ"4t/ 0M9 tit?y 1JJ.1v o. a4 o 18g

                                                                                                                            ,4SSf.NE'I) ACCELE'l?Arlt N
                                                 ,I I                                                                                                 NOTc.S:

iA,,,/lysis w*.s ,,,.,,,,. 11.slllg file .sldlltl{vtf .slip ciM/c 11>0/A,-#! 2.Slu,,-r .sl'rt1119/-lis JMIW,il(s

                                                                                                                                              .Smtltl d$ , , , , , , in ,-,i9,-;,;J dn.lJ,s1s, I

I

HISTORICAL Rcvisedb)'An,cnclu.,...17 blllbl ltbJ 11,Q SAbOL£ DAM No. I f l l l l *

2.4,t*l4 Saddle D** .... 1 Douala* Da:nj Re*ults o( Anolyoo for 1/2 SSE

I I I I I I I I I I PLAN B-B NOTc A: All debris FrCMra;/eo' par I-ion of" d.!Jm judg f" i>e .bekJ,.-

                                               /'atlure ekrafton.s.

NCTE8: P()werhouse .Irrtke blocks / - I 7 and N:,noyerf'/ow L)a,, blccks h - S DOWNSTREAM ELEVATION und 2'J -35 jvd9ed t11relfhlin in oriqiMI position. ( Figure 2.4.4-li Douglas Dam, Assumed Condition of D m*.*After Failure, 1/2 S E.. and 1/2 Maximum Po ,sible Flood

                  'I ll4M I,
        @j / ; :
        *

I
I

( _ A S S U M E D lOCATIONOF DEBRIS IN CHANNEL AFTCR FAILURE * , I * $£CT/ON A-A ASS(.! rO POSITION OF' FAILED l)A;,; to *l l I Figute 2.4.4-16

ASSUMiD I.OCAT/O/f OI' D£/JRIS font na Dam - Assumed Condition of Dam After 1/2 SSE and 1/2 Maximum IN CHANNEL AFTER FAILUIIE Possible Flood ELEVATION

_(ASSUMED AFTER /:'A/LURE)

= =

                                                                   !1                                                                                                                                       c:aKlffl -        ,t;IU.!Cill Mffli=

it, !'/

                                           -r=z;_                                                                                                                                     l ,,."

I. M,L _ , : : : u n : " - * - - l'Ul'AiHIO. ef E*1al11"l 1-*"1J?'-f!. >>JC&.* ! 14 I .111ra. f.">>11!. iCS11aen wt1:11'!'1!1.1

l :.:A *.u:,

                                                                                                                                                                                                                 --.CS*i:g                            ,\lllfl"- ; ; i , n , - . f l ' l l IFJf-iWJ. , . l,:j ; - a IOil
                                                                                                                =
                                                                                                                  =
                                                                                                                  -=
                                                                                                                   -_   __:_:_:-:_:-:-r_:-_:_     ::_::_-:_::_:----_-"-_    -: _
                                                                                                                                                                                                                              =::::c;;                                                                                  t
                                                                                             -= -,

- r:

c_ ::-c: = f rr._::a : :- : : :: =

= =

- z.

a . "'9,'.0_,:).__ ;,LQ o !;a: : Ill

                                                                                                                                                                                                                     -.ii.eJI! 1'fl r , C,H!!II l r l t t U i f !I: i'Ct iS1H i-11.
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I.

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CNIPB-.S.'Uftm -

11, rf!mr.

1*.

1. TIil e,;U, (11'..,."-IIUu. Cl 1 \tltfMW W g* i!l

gJ-iSJU,.i:. 1il

                                                                                                                                                                                                               " ' "                1lfT1n11 r:t8fWL

t.

ttiU. t . n W a . 1 ffl!t1i!ff tat :,w-ua n i:: M ?(d

                                                                                                                                                                                                                , w rEl'flU . l l l l f ' O U >>.ttJ?JI Mioll                               a,           .-,::- ";it "ir.l <<"' :;JS_
                                                                                                                                                                                                               . . . . . '1f( f f i l H " " '

f l l l fflf 11Q.1ill tt.":U Willl 1 h 1 IL.."f!--9

11 1 JtfT t , ' " " ' 9 - TU 1fll:U fW:.q i;; !"£ c t " : - ' l'7 ,"P,r

                                                                                                                                                                         * , ,   * . .j . LC L ...... 0 tKl'till I I                t --\.--            ttl_ ""'!         Cff'. ;;-: -                .:i -!:!,.j    l
                                                                                                                                                                                                                                -

m *nc*&WJ. m t!.i'--.,.;e.

                                                                                                                                              --; - - ,...JIS.o
                                                                                                                                                                                                        .JS
                                                                                                                                                                                       .£.P- ,u,.a ELEVATION Cl,.._..

0 4 - N O J -:-lf, . I

a:.4LE -r,o*

                                                                                                                                                                                                                                                                 !W' ;""r       ,:, 1;:r..._....o SECTION f ' - 8 & C - C C*C.U _ _           _

D-D Figure 2.4.4-17 Fontana Project Concrete Strength-ening of Blk 33, 34, and 35

r SCALE IN MILES 0 10 20 RHRRSI FIGURE 2.4.4-18 1/2 SSE WITH -EPICENTER WITHIN REA SHOWN 435

710 -r-------------------------------- ----------------- ---, 705 ......_ Plenl Grade Elevation 705 - - ,J . . . . . . . . - ** * ,. .*. . . . . . . , .. _,_ - .. .' 700

        * * - * - - ,* . . . . . . -. i * * * *                         .                  .
                                                                  * - l' ... - - ... --. - -
                                                                                                                                                         - - - .. - t * - - * - * -** - - -

C 0 695 690 * - - - * - * -* * * - - - - - \ . -..*. ... *. . . . J * - * - * - * . . . I * * * * * . . ., . * * * * *

I * * . .* .......... ' - .* -. - ..

685 0 10 20 30 40 60 60 70 80 90 100 110 120 130 140 150 Hours Since Fontana, HlwaHee, Apalachla, and Blue Ridge Failure Seismic Flood Analysis - Fontana, Hlwassee, Apalachla and Blue Ridge OBE Failure in 1/2 PMF - Sequoyah Plant Figure 2.4.4-21 Revised by Amendment 17

4 fe I-

.Spillway 9afe s o/'eo

0. 92 9 1// rh'aF-,.ce A,e NOT ES:

T, E.,.,.tlu vMe .......,,'1-----.c I. // r t/cti/ acce lerdt i o r sp1iV""}' a l b.st1 a s s u m d t " . b

IRq.a B y dyna mic .,,;; ys a1nplil'ic11/4m , , ;ccw.,,;1-iH, aDDY* f/2e L)t;.,sd d l rmin *tl l o l>e a:,e4 9ol 2.hbr lze,n fal accd cw or s,oil/Nt:iy 41' IJ.se c , s s
o 'I" /Jc a I t!f 9'* tJ)' dyl'ldmic s s-

                                                                                                                     'I I,..

TYPICAL

,4 A:75. 0 5 ,

SPIL LWA Y SEC TION

                                                                                                                     -l
                                                                                                                           /lyclrosrahc Fm-= I-IJ,,/ral}f1dmic}

Farce Pue ../ att!J 9 Aecelera//e>n 01 £Jam

                                                                                                                                                                           -r.*. _ '() e T . _,, .,,.
                                                                                                                                                                                        .rrnq.  . .,

ampli1ic .s,fl n cf" (icceltlrdl'i,,, cib Ye lh ba.s ws, lb &.11rth e/df( !'rmi n*d l o ce ,-2c1f np. 1: , Ii' '"

I : ""'. .,'. *., . ,. ,.

3. Spill way 9 << f s assti,,,r, d* o , , . "

                     ,,..,.,..,.,. ,.,.,. . rTTTT 'r'T'TTT nnT                                                                    TW+ IIW-rw TJIY.:11S N 3/4*l i          1 !1 11!1/ UllI l!: !l l/ /m-1r, 1/ 1 //I!

I

                                                                           '""'"Iii 1111/ .:= 7..3/ -

4 74'et T<V t h i s ,u*,Jfysis. t/PLI.F T /JIAGRAM E l 7tXI UPL IFT Pl?ESSO/?E ASSLIMcO TtJ A C T O N / 0 0 % a F BAS E AREA

                                                                                                                                                   ..¥ Shea.t; flit2f / s r e p d for {!;.: / i s I                                                                                                                                    calcu/'91-e-d fro/7) s h e a r - Frie l/on 1e1rmvl" 0.65" .!"Y sA., A                ,s Z/1 asst/ m d f c 6 e

n l h
e area .

                                                                                                                                                **f"br b a s e       f E l 7a? r sulra nr
                                                                /t:?5.05'                                           J                                  T&//s oUTsi o'e  6'9se Vt? r           SSE t3ASE PLA N£/ 700 B A S E P R E S S U R E :K-

Ay9

                                              .zj /                                       Sreqd I-I                                                                                       Z,#N J/           ZH                                        shear                     ror Q=I F m,a,r F Mo                                                                                                    .

ll;;si 14RSt1,r53,11( /.6,2 qt.,

                                                                                       .25psi

I' 0.9 Figure 2.4.4-24 Spillway, Fort oudoun Dam, Results of Analysis for SSE

tJSO 8.30 Best Available Historical Image l'.SO 7j0 09 ,()'../ , ,- M 9 aM, llMg a.c>Os, (U,tg o. 1).1,!, a,16 g AS Sf/MEO ACC£l.ERA r/ON NOTES: t An,1/y3 is - s m M usin9 fl,a sfdnthrd sljo circ:k method*

                                                                                                                .2. Sltft,,- 3fr,m9Ms oF mi,l rit,ls s - c1.S 1JSt1tlinqrigilldl
                                                                                                                    ,sn.,Jys/.s.

HISTORICAL Revised by Amendment 17 TYPICAL E'MBAM<N(Ett"( SECTION Figure 2. 4, 4-25 Embankment, Fort J,oudoun Dam, Results o f A n a l y s i s For SSE

                          \                              /         1                   1;'//I!
                           \           II             I
                                                       / r l ,
                                                                                      ' I I - _! - - - _,_ ,({_)'_II f;l1 I          I I         I,o/ I I

y i I l"-I'f1II

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I A..Jj / - - 7I - , I --/ ::, ... .._,_

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                                                / . //   /  //         '-

I / '--

                                               . / /

Sp//lwoy judged to

                                                                                                         ""                                       fe11'/ e,6ove £ / 750.0          .J:. : r£1&22.o
                                                                                                            '                                     (see Note d:-:.4 A)             / ,, ,. , )

c; 750. 0 PLAN ,_\

                                                                                                                                                               . * *  ***ii. .
                                                                                                                                                                         \I *          ".

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485

APPENDIX 2.4A FLOOD PROTECTION PLAN S2-4Atoc.doc 2.4A-i

SQN TABLE OF CONTENTS Section Title Page APPENDIX 2.4A FLOOD PROTECTION PLANT 2.4A.1 INTRODUCTION 2.4A-1 2.4A.1.1 Design Basis Flood 2.4A-1 2.4A.1.2 Combinations of Events 2.4A-2 2.4A.1.3 Post Flood Period 2.4A-2 2.4A.1.4 Localized Floods 2.4A-2 2.4A.2 PLANT OPERATION DURING FLOODS ABOVE GRADE 2.4A-2 2.4A.2.1 Flooding of Structures 2.4A-2 2.4A.2.2 Fuel Cooling 2.4A-3 2.4A.3 WARNING PLAN 2.4A-5 2.4A.3.1 Rainfall Floods 2.4A-6 2.4A.3.2 Seismic Dam Failure Floods 2.4A-6 2.4A.4 PREPARATION FOR FLOOD MODE 2.4A-6 2.4A.4.1 Reactors Initially Operating at Power 2.4A-7 2.4A.4.2 Reactor Initially Refueling 2.4A-7 2.4A.4.3 Plant Preparation Time 2.4A-7 2.4A.5 EQUIPMENT 2.4A-7 2.4A.5.1 Equipment Qualification 2.4A-8 2.4A.5.2 Temporary Modification and Setup 2.4A-8 2.4A.5.3 Electric Power 2.4A-8 2.4A.5.4 Instrument Control, Communication and Ventilation Systems 2.4A-9 2.4A.6 SUPPLIES 2.4A-9 2.4A.7 PLANT RECOVERY 2.4A-9 2.4A.8 BASIS FOR FLOOD PROTECTION PLAN IN RAINFALL FLOODS 2.4A-9 2.4A.9 BASIS FOR FLOOD PROTECTION PLAN IN SEISMIC-CAUSED DAM FAILURES 2.4A-14 S2-4Atoc.doc 2.4A-ii

SQN-17 LIST OF TABLES Number Title 2.4A-2 Critical Cases - Seismic - Caused Dam Failures Time Between Seismic Event and Selected Plantsite Flood Elevation LIST OF FIGURES Number Title 2.4A-2 Flow Diagram - Flood Protection Provisions - With New ERCW Intake Station in Operation - Natural Convection Cooling (Unit 1 Shown - Unit 2 Similar) 2.4A-3 Flow Diagram - Flood Protection Provisions - With New ERCW Intake Station in Operation - Open Reactor Cooling (Unit 1 Shown - Unit 2 Similar) 2.4A-4 Sequoyah Nuclear Plant Flood Protection Plan Basis for Safe Shutdown for Plant Flooding S2-4Atoc.doc 2.4A-iii

SQN-17 APPENDIX 2.4A FLOOD PROTECTION PLAN 2.4A.1 Introduction This appendix describes the methods by which the Sequoyah Nuclear Plant will be made capable of tolerating floods above plant grade without jeopardizing public safety. Since flooding of this magnitude, as explained in section 2.4, is most unlikely, extreme steps are considered acceptable including actions that create or allow extensive economic damage to the plant. The actions described herein will be implemented for floods ranging from slightly below plant grade, to allow for wave runup, to the Design Basis Flood (DBF). 2.4A.1.1 Design Basis Flood The DBF is the calculated upper limit flood that includes the probable maximum flood (PMF) plus the wave runup caused by a 45-mile-per-hour overwater wind; this is discussed in subsection 2.4.3.6. The table below gives representative levels of the DBF at different plant locations. Design Bases Flood (DBF) Levels Probable maximum flood (still reservoir) 719.6 DBF runup on vertical external, unprotected walls 723.8 DBF surge level within flooded structures 720.1 The lower flood elevations listed above are actual DBF elevations and are not normally used for the purpose of design but are typically used in plant procedures including procedures which direct plant actions in response to postulated DBF. For purposes of designing the flood protection for systems, structures, and components, the following higher elevations should be used thus ensuring additional margin has been included in the development of design analysis. Design Analysis Flood Levels Maximum still reservoir 723.5 Runup on vertical external, unprotected walls 729.5 Surge level within flooded structures 724.0 See FSAR References 2.4A.10-1 and 2.4A.10-2. In addition to level considerations, plant flood preparations will cope with the "fastest rising" flood which is the calculated flood that can exceed plant grade with the shortest prediction notice. Reservoir levels for large floods in the Tennessee Valley can be predicted well in advance. A minimum of 27 hours, divided into two stages, is provided for safe plant shutdown by use of this prediction capability. Stage I, a minimum of 10 hours long, will commence upon a prediction that flood-producing conditions might develop. Stage II, a minimum of 17 hours long, S2-4app.doc 2.4A-1

SQN-27 will commence on a confirmed estimate that conditions will provide a flood. This two-stage scheme is designed to prevent excessive economic loss in case a potential flood does not fully develop. 2.4A.1.2 Combinations of Events Because floods above plant grade, earthquakes, tornadoes, or design basis accidents, including a loss-of-coolant accident (LOCA), are individually very unlikely, a combination of a flood plus any of these events or the occurrence of one of these during the flood recovery time or of the flood during the recovery time after one of these events is considered incredible. Surges from seismic failure of upstream dams, however, can exceed plant grade, but to lower DBF levels, when imposed coincident with wind and certain floods. A minimum 27 hours of warning is assured so that ample time is available to prepare the plant for flooding. 2.4A.1.3 Post Flood Period Because of the improbability of a flood above plant grade, no detailed procedures will be established for return of the plant to normal operation unless and until a flood actually occurs. If flood mode operation (subsection 2.4A.2) should ever become necessary, it will be possible to maintain this mode of operation for a sufficient period of time (100 days) so that appropriate recovery steps can be formulated and taken. The actual flood waters are expected to recede below plant grade within 1 to 6 days. 2.4A.1.4 Localized Floods Localized plant site flooding due to the probable maximum storm (subsection 2.4.3) will not enter vital structures or endanger the plant. Plant shutdown will be forced by water ponding on the switchyard and around buildings, but this shutdown will not differ from a loss of offsite power situation as described in Chapter 15. The other steps described in this appendix are not applicable to this case. 2.4A.2 Plant Operation During Floods Above Grade "Flood mode" operation is defined as the set of conditions described below by means of which the plant will be safely maintained during the time when flood waters exceed plant grade (elevation 705) and during the subsequent period until recovery (subsection 2.4A.7) is accomplished. 2.4A.2.1 Flooding of Structures Only the Reactor Building, the Diesel Generator Building (DGB), and the Essential Raw Cooling Water Intake Station will be maintained essentially water tight during the flood mode. Walls and penetrations are designed to withstand all static and dynamic forces imposed by the DBF. The Reactor Buildings protect SSCs contained within that are required for Flood Mode operations. All penetrations below the Design Analysis Flood level of elevation 724 have been sealed with seals, which are tested to withstand hydrostatic forces generated by the Design Basis Flood. Analysis demonstrates the acceptability of minor leakage through the seals into the annulus. The lowest floor of the DGB is at elevation 722 with its doors on the uphill side facing away from the main body of flood water. This elevation is lower than the previous DBF elevation of 722.6. The 1998 reanalysis determined the still water elevation to be 719.6, with wind wave runup at the DGB to elevation 721.8. Therefore, flood levels do not exceed floor elevation of 722. The entrances into S2-4app.doc 2.4A-2

SQN-17 safety-related areas and all mechanical and electrical penetrations into safety-related areas are sealed to prevent major leakage into the building for water up to the PMF, including wave runup. Due to the 1998 reanalysis this only applies to below grade features. Redundant sump pumps are provided within the building to remove minor leakage. The Essential Raw Cooling Water (ERCW) intake station is designed to remain fully functional for floods up to the PMF, including wind-wave runup. The deck elevation (elevation 720) is below the PMF plus wind wave runup, but it is protected from flooding by the outside walls. The traveling screen wells extend above the deck elevation up to the design basis surge level. The wall penetration for water drainage from the deck in nonflood conditions is below the DBF elevation, but it is designed for sealing in event of a flood. All other exterior penetrations of the station below the PMF are permanently sealed. Redundant sump pumps are provided on the deck and in the interior rooms to remove rainfall on the deck and water seepage. All other structures, including the service, turbine, auxiliary, and control buildings, will be allowed to flood as the water exceeds their grade level entrances. All equipment, including power cables, that is located in these structures and required for operation in the flood mode is either above the DBF or designed for submerged operation. 2.4A.2.2 Fuel Cooling Spent Fuel Pit Fuel in the spent fuel pit will be cooled by the normal Spent Fuel Pit Cooling (SFPC) System. The pumps are located on a platform at elevation 721 which is above the surge level of 720.1. During the flood mode of operation, heat will be removed from the heat exchangers by ERCW instead of component cooling water. As a backup to spent fuel cooling, water from the Fire Protection (FP) System can be dumped into the spent fuel pool, and steam removed by the area ventilation system. Reactors Residual core heat will be removed from the fuel in the reactors by natural circulation in the Reactor Coolant (RC) system. Heat removal from the steam generators will be accomplished by adding river water from the FP System (subsection 9.5.1) and relieving steam to the atmosphere through the power relief valves. Primary system pressure will be maintained at less than 500 lb/in2g by operation of the pressurizer relief valves and heaters. This low pressure will lessen leakage from the system. Secondary side pressure will be maintained at or below 90 psig by operation of the steam line relief valves. An analysis has been performed to ensure that the limiting atmospheric relief capacity would be sufficient to remove steam generated by decay heat. At times beyond approximately 10 hours following shutdown of the plant two relief valves have sufficient capacity to remove the steam generated by decay heat. Since a minimum of 27 hours flood warning is available it is concluded that the plant could be safely shutdown and decay heat removed by operation of only two relief valves. Reference FSAR 2.4A.10-1. S2-4app.doc 2.4A-3

SQN-21 The main steam power operated relief valves will be adjusted to maintain the steam pressure at or below 90 psig. If this control system malfunctions, then the controls in the main control room can be utilized to operate the valves in an open-closed manner. Also, a manual loading station and the relief valve handwheel provide additional backup control for each relief valve. The secondary side steam pressure can be maintained for an indefinite time by the means outlined above. The cooling water flow paths conform to the single failure criteria as defined in FSAR Section 3.1.1. In particular, all active components of the secondary side feedwater supply and ERCW supply are redundant and can therefore tolerate a single failure in the short or long term. A passive failure, consistent with the 50 gpm loss rate specified in FSAR Section 3.1.1, can be tolerated for an indefinite period without interrupting the required performance in either supply. If one or both reactors are open to the containment atmosphere as during the refueling operations, then the decay heat of any fuel in the open unit(s) and spent fuel pit will be removed in the following manner. The refueling cavity will be filled with borated water (approximately 2000 ppm boron concentration) from the refueling water storage tank. The SFPC System pump will take suction from the spent fuel pit and will discharge to the SFPC System heat exchangers. The SFPC System heat exchanger output flow will be directed by a piping connection to the Residual Heat Removal (RHR) System heat exchanger bypass line. The tie-in locations in the SFPC System and the RHR System are shown in Figures 9.1.3-1 and 5.5.7-1, respectively. This connection will be made using prefabricated, in- position piping which is normally disconnected. During flood mode preparations, the piping will be connected using prefabricated spool pieces. Prior to flooding, valve number 78-513 (refer to Figure 9.1.3-1) and valves FCV 74-33, and 74-35 (refer to Figure 5.5.7-1) will be closed; valves HCV 74-36, 74-37, FCV 74-16, 74-28, 63-93, and 63-94 (refer to Figure 5.5.7-1 and 6.3.1-1) will be opened or verified open. This arrangement will permit flow through the RHR heat exchangers and the four normal cold leg injection paths to the reactor vessel. The water will then flow downward through the annulus, upward through the core (thus cooling the fuel), then exit the vessel directly into the refueling cavity. This results in a water level differential between the spent fuel pit and the refueling cavity with sufficient water head to assure the required return flow through the 20-inch diameter fuel transfer tube thereby completing the path to the spent fuel pit. Except for a portion of the RHR System piping, the only RHR System components utilized below flood elevation are the RHR System heat exchangers. Inundation of these passive components will not degrade their performance for flood mode operation. After alignment, all valves in this cooling circuit located below the maximum flood elevation will be disconnected from their power source to assure that they remain in a safe position. The modified cooling circuit for open reactor cooling will be assured of two operable SFPC System pumps (a third pump is available as a backup) as well as two SFPC System heat exchangers. Also, the large RHR System heat exchangers are supplied with essential raw cooling water during the open reactor mode of fuel cooling; these heat exchangers provide an additional heat sink not available for normal spent fuel cooling. Fuel coolant temperature calculations, assuming conservative heat loads and the most limiting, single active failure in the SFPC System, indicate that the coolant temperatures are acceptable. S2-4app.doc 2.4A-4

SQN-23 The temperatures can be maintained at a value appreciably less than the fuel pit temperature calculated for the nonflood spent fuel cooling case when assuming the loss of one equipment train. As further assurance, the open reactor cooling circuit was aligned and tested, during pre-operational testing, to confirm flow adequacy. Normal operation of the RHR System and SFPC System heat exchangers will confirm the heat removal capabilities of the heat exchangers. High spent fuel pit temperature will cause an annunciation in the MCR, thus indicating equipment malfunction. Additionally, that portion of the cooling system above flood water will be frequently inspected to confirm continued proper operation. For either mode of reactor cooling, leakage from the Reactor Coolant System will be collected, to the extent possible, in the reactor coolant drain tank; nonrecoverable leakage will be made up from supplies of clean water stored in the four cold leg accumulators, the pressurizer relief tank, the cask decontamination tank, and the demineralized water tank. If these sources prove insufficient, the FP System can be connected to the Auxiliary Charging System (subsection 9.3.5) as a backup. Whatever the source, makeup water will be filtered, demineralized, tested, and borated, as necessary, to the normal refueling concentration, and pumped by the Auxiliary Charging System into the reactor (see Figures 2.4A-2 and 2.4A-3). Power Electric power will be supplied by the onsite diesel generators starting at the beginning of Stage II or when offsite power is lost, whichever occurs first (subsection 2.4A.5.3). Cooling of Plant Loads Plant cooling requirements, with the exception of the FP System which must supply feedwater to the steam generators, will be met by the ERCW System (refer to subsection 9.2.2). Plant Water Supply The plant water supply is thoroughly discussed in subsection 9.2.2. The following is a summary description of the water supply provided for use during flooded plant conditions. The ERCW station is designed to remain fully functional for all floods up to and including the DBF. The CCW intake forebay will provide a water supply for the fire/flood mode pumps. If the flood approaches DBF proportions, there is a remote possibility that Chickamauga Dam will fail. Such an event would leave the Sequoyah Plant CCW intake forebay isolated from the river as flood water recedes below EL 665. Should this event occur, the CCW forebay has the capacity of retained water to supply two steam generators in each unit and provide spent fuel pit with evaporation makeup flow until CCW forebay inventory makeup is established. The ERCW station is designed to be operable for all plant conditions and includes provisions for makeup to the forebay. Reference FSAR 2.4A.10-1. 2.4A.3 Warning Plan Plant grade elevation 705 can be exceeded by both rainfall floods and seismic-caused dam failure floods. A warning plan is needed to assure plant safety from these floods. S2-4app.doc 2.4A-5

SQN-17 2.4A.3.1 Rainfall Floods Protection of the Sequoyah Plant from the low probability rainfall floods that might exceed plant grade depends on a flood warning issued by TVA's River Operations as described in Section 2.4A.8. With TVA's extensive climate monitoring and flood predicting systems and flood control facilities, floods in the Sequoyah area can be reliably predicted well in advance. The Sequoyah Nuclear Plant flood warning plan will provide a minimum preparation time of 27 hours including a 3 hour margin for operation in the flood mode. Four additional, preceding hours will provide time to gather data and produce the warning. The warning plan will be divided into two stages--the first a minimum of 10 hours long and the second of 17 hours--so that unnecessary economic penalty can be avoided while adequate time is ensured for preparing for operation in the flood mode. The first stage, Stage I, of shutdown will begin when there is sufficient rainfall on the ground in the upstream watershed to yield a projected plant site water level of 697 in the winter months (October 1 through April 15) and 703 in the summer (April 16 through September 30). This assures that the additional time required is available when shutdown is initiated. The water level of 703 (two feet below plant grade) will allow margin so that waves due to high winds cannot disrupt the flood mode preparation. Stage I will allow preparation steps causing some damage to be sustained but will withhold major economic damage until the Stage II warning assures a forthcoming flood above grade. The plant preparation status will be held at Stage I until either Stage II begins or TVA's River Operations determines that flood waters will not exceed elevation 703 at the plant. The Stage II warning will be issued only when enough rain has fallen to predict that elevation 703 is likely to be exceeded. 2.4A.3.2 Seismic Dam Failure Floods Protection of the Sequoyah plant from flood waves generated by seismically caused dam failures which exceed plant grade depends on TVAs River Operation organization to identify when a critical combination of dam failures and floods exist. There are nine upstream dams whose failure, in combination coincident with certain storm conditions, would cause a flood to exceed plant grade. These dams are Norris, Cherokee, Douglas, Fort Loudoun, Fontana, Hiwassee, Apalachia, Blue Ridge, and Tellico. 2.4A.4 Preparation for Flood Mode At the time the initial flood warning is issued, the plant may be operating in any normal mode. This means that either or both units may be at power or either unit may be in any stage of refueling. S2-4app.doc 2.4A-6

SQN-27 2.4A.4.1 Reactors Initially Operating at Power If both reactors are operating at power, Stage I and then, if necessary, Stage II procedures will be initiated. Stage I procedures will consist of a controlled reactor shutdown and other easily revokable steps such as moving supplies necessary to the flood protection plan above the DBF level and making temporary connections and load adjustments on the onsite power supply. Stage II procedures will be the less easily revokable and more damaging steps necessary to have the plant in the flood mode when the flood exceeds plant grade. The fire/flood mode pumps may supply auxiliary feedwater for reactor cooling (Reference 3). Other essential plant cooling loads will be transferred from the component cooling water to the ERCW System (subsection 9.2.2). Radioactive Waste System (Chapter 11) and CVCS tanks, which are susceptible to flotation will be secured by either filling the tanks during flood preparations or by opening the tanks to allow floodwaters to enter, tanks which are adequately anchored to prevent floatation are exempt from these requirements. Some power and communication lines running beneath the DBF and not designed for submerged operation will require disconnection. Batteries beneath the DBF will be disconnected. 2.4A.4.2 Reactor Initially Refueling If time permits, fuel will be removed from the unit(s) undergoing refueling and placed in the spent fuel pit; otherwise fuel cooling will be accomplished as described in subsection 2.4A.2.2. If the refueling canal is not already flooded, the mode of cooling described in subsection 2.4A.2.2 requires that the canal be flooded with borated water from the refueling water storage tank. If the flood warning occurs after the reactor vessel head has been removed or at a time when it could be removed before the flood exceeds plant grade, the flood mode reactor cooling water will flow directly from the vessel into the refueling cavity. If the warning time available does not permit this, then the upper head injection piping will be disconnected above the vessel head to allow the discharge of water through the four upper head injection standpipes. Additionally, it is required that the prefabricated piping be installed to connect the RHR and SFPC Systems, and that ERCW be directed to the secondary side of the RHR System and SFPC System heat exchangers. 2.4A.4.3 Plant Preparation Time All steps needed to prepare the plant for flood mode operation can be accomplished within 24 hours of receipt of the initial warning that a flood above plant grade is possible. An additional 3 hours are available for contingency margin before wave runup from the rising flood might enter the buildings. Site grading and building design prevent any flooding before the end of the 27 hour preflood period. 2.4A.5 Equipment Both normal plant components and specialized flood-oriented supplements will be utilized in coping with floods. All such equipment required in the flood mode is either located above the DBF or is within a nonflooded structure or is designed for submerged operation. Systems and components needed only in the preflood period are protected only during that period. S2-4app.doc 2.4A-7

SQN 2.4A.5.1 Equipment Qualification To ensure capable performance in this highly unlikely but rigorous, limiting design case, only high quality components will be utilized. Active components are redundant or their functions diversely supplied. Since no rapidly changing events are associated with the flood, repairability offers reinforcement for both active and passive components during the long period of flood mode operation. Equipment potentially requiring maintenance will be accessible throughout its use, including components in the Diesel Generator Building. 2.4A.5.2 Temporary Modification and Setup Normal plant components used in flood mode operation and in preparation for flood mode operation may require modification from their normal plant operating configuration. Such modification, since it is for a limiting design condition and since extensive economic damage is acceptable, will be permitted to damage existing facilities for their normal plant functions. However, most alterations will be only temporary and nondestructive in nature. For example, the switchover of plant cooling loads from the component cooling water to the ERCW System will be done through valves and a prefabricated spool piece, causing little system disturbance or damage. Equipment especially provided for the flood design case includes both permanently installed components and more portable apparatus that will be emplaced and connected into other systems during the preflood period. Detailed procedures to be used under flood mode operation have been developed and are incorporated in the plant's Abnormal Operating Instructions. 2.4A.5.3 Electric Power Because there is a possibility that high winds may destroy powerlines and disconnect the plant from offsite power at any time during the preflood transition period, only onsite power will be used once Stage II of the preparation period begins. While most equipment requiring alternating current electric power is a part of the permanent emergency onsite power system, other components will be temporarily connected, when the time comes, by prefabricated jumper cables. All loads that are normally supplied by onsite power but are not required for the flood will be switched out of the system during the preflood period. Those loads used during the preflood period but not during flood mode operation will be disconnected when they are no longer needed. During the preparation period, all power cables running beneath the DBF level, except those especially designed for submerged operation, will be disconnected from the onsite power system. Similarly, direct current electric power will be disconnected from unused loads and potentially flooded lines. Charging will be maintained for each battery by the onsite alternating current power system as long as it is required. Batteries that are beneath the DBF will be disconnected during the preflood period when they are no longer needed. S2-4app.doc 2.4A-8

SQN 2.4A.5.4 Instrument Control, Communication and Ventilation Systems All instrument, control, and communication lines that will be required for operation in the flood mode are either above the DBF or within a nonflooded structure or are designed for submerged operation. Unneeded cables that run below the DBF will be disconnected to prevent short circuits. Redundant means of communications are provided between the central control area (the main and auxiliary control rooms) and all other vital areas that might require operator attention, such as the Diesel Generator Building. Instrumentation is provided to monitor all vital plant parameters such as the reactor coolant temperature and pressure and steam generator pressure and level. Control of the pressurizer heaters and relief valves and steam generator feedwater flow and atmospheric relief valves will ensure continued natural circulation core cooling during the flood mode. All other important plant functions will be either monitored and controlled from the main control area or, in some cases where time margins permit, from other points in the plant that are in close communication with the main control area. Ventilation, when necessary, and limited heating or air-conditioning will be maintained for all points throughout the plant where operators might be required to go or where required by equipment heat loads. 2.4A.6 Supplies All equipment and most supplies required for the flood are on hand in the plant at all times. Some supplies will require replenishment before the end of the period in which the plant is in the flood mode. In such cases supplies on hand will be sufficient to last through the short time (subsection 2.4A.1.3) that flood waters will be above plant grade and until replenishment can be supplied. For instance, there is sufficient diesel generator fuel available at the plant to last for 3 or 4 weeks; this will allow sufficient margin for the flood to recede and for transportation routes to be reestablished. 2.4A.7 Plant Recovery The plant is designed to continue safely in the flood mode for 100 days even though the water is not expected to remain above plant grade for more than 1 to 6 days. After recession of the flood, damage will be assessed and detailed recovery plans developed. Arrangements will then be made for reestablishment of offsite power and removal of spent fuel. The 100-day period provides more than adequate time for the development of procedures for any maintenance, inspection, or installation of replacements for the recovery of the plant or for a continuation of flood mode operations in excess of 100 days. A decision based on economics will be made on whether or not to regain the plant for power production. In either case, detailed plans will be formulated after the flood, when damage can be accurately assessed. 2.4A.8 Basis For Flood Protection Plan In Rainfall Floods Summary Large Tennessee River floods can exceed plant grade elevation 705 at Sequoyah Nuclear Plant. Plant safety in such an event requires shutdown procedures which may take 24 hours to S2-4app.doc 2.4A-9

SQN-17 implement. TVA flood forecast procedures will provide at least 27 hours of warning before river levels reach elevation 703. Use of elevation 703, 2 feet below plant grade, provides enough freeboard to prevent waves from 45-mile-per-hour, overwater winds from endangering plant safety during the final hours of shutdown activity. For conservatism the fetches calculated for the PMF (Figures 2.4.3-15 and 2.4.3-16) were used to calculate maximum wind wave additive to the reservoir surface at elevation 703 feet msl. The maximum wind additive to the reservoir surface would be 2.8 feet and would not endanger plant safety during the final hours of shutdown. This is due to the long shallow approach and the waves breaking at the perimeter road (elevation 705 feet msl). After the waves break there is not sufficient depth or distance between the perimeter road and the safety-related facilities for new waves to be generated. Forecast will be based upon rainfall already reported to be on the ground. Different target river level criteria are needed for winter use and for summer use to allow for seasonally varied reservoir levels and rainfall potential. To be certain of 27 hours for preflood preparation, warnings of floods with the prospect of reaching elevation 703 must be issued early; consequently, some of the warnings may later prove to have been unnecessary. For this reason preflood preparations are divided into two stages. Stage I steps, requiring 10 hours, would be easily revokable and cause minimum damage. The estimated probability is less than 0.0026 that a Stage I warning will be issued during the 40-year life of the plant. Additional rain and streamflow information obtained during Stage I activity will determine if the more damaging steps of Stage II need to be taken with the assurance that at least 17 hours will be available before elevation 703 is reached. The estimated probability is less than 0.0010 that shutdown will need to continue into Stage II during plant life. Flood forecasting to assure adequate warning time for safe plant shutdown during floods will be by River Operations of River System Operations. TVA Forecast System (HISTORICAL INFORMATION) TVA has in constant use an extensive, effective system to forecast flow and elevation as needed in the Tennessee River Basin. This permits efficient operation of the reservoir system and provides warning of when water levels will exceed critical elevations at selected, sensitive locations. Elements of the present (2001) forecast system above Sequoyah Nuclear Plant include the following:

1. One hundred sixty (160) rain gages measure rainfall, with an average density of 165 square miles per rain gage. Of these gages 112 are owned by TVA, 35 are owned by the National Weather Service (NWS), 7 are owned by the United States Geological Service (USGS), 2 are owned by the United States Corps of Engineers (USACE), and 4 are owned by Alcoa. Most of these gages are tipping buckets collector type and the transmission of the data is either by satellite or telephone. At some of the gages located at hydroplants, the data is manually read.

S2-4app.doc 2.4A-10

SQN-17 Information normally is received daily from the gages at 6 a.m. and at least every 6 hours during flood periods. Close interval rainfall reports can be obtained from a majority of the gages if needed.

2. Streamflow data are received for 35 gages from 16 TVA gages amd 19 USGS gages. These gages trasmit their data either by satellite or telephone or both. Discharge data are received from 26 hydroplants. Of these plants, 25 also transmit headwater elevation data, and 13 transmit tailwater elevation data. Therefore, steamflow data are available from 61 locations. Streamflow data are received daily at 8 a.m. and at least every 2 hours if needed during flood operations.

3. Weather forecasts including quantitative precipitation forecasts are received four times daily and at other times when changes are expected.

4. Computer programs which translate rainfall into streamflow based on current runoff conditions and which permit a forecast of flows and elevations based upon both observed and predicted rainfall. Two separate computers are utilized and are designed to provide backup for each other.

One computer is used primarily for data collection, with the other used for executing forecasting programs for reservoir operations. The time interval between receiving input data and producing a forecast is less than 4 hours. Forecasts normally cover at least a 8-day period. As effective as the forecast system already is, it is constantly being improved as new technology provides better methods to interrogate the watershed during floods and as the watershed mathematical model and computer system are improved. Also, in the future, improved quantitative precipitation forecasts may provide a more reliable early alert of impending major storm conditions and thus provide greater flood warning time. The TVA forecast center is manned 24 hours a day. Normal operation produces two forecasts daily, one by 12 noon based on data collected at 6 a.m. Central time, and the second by 4 a.m. based on data collected at midnight Central Time. When serious flood situations demand, forecasts are produced every 4 hours. Basic Analysis To develop a forecast procedure to assure safe shutdown of Sequoyah Nuclear Plant for flooding, 17 hypothetical PMP storms, including their antecedent storms, were analyzed. They enveloped potentially critical seasonal variations and time distributions of rainfall. To be certain that fastest rising flood conditions were included, the effects of varied time distribution of rainfall were tested by alternatively placing the maximum daily PMP on the first, the middle, and the last day of the 3-day main storm. In each day the maximum 6-hour depth was placed during the second interval except when the maximum daily rain was placed on the last day. Then the maximum 6-hour amount was placed in the last 6 hours. The procedures used to compute flood flows and elevations are described in subsections 2.4.3.1, 2.4.3.2, and 2.4.3.3. Some flood events were analyzed using earlier versions of the watershed model described in subsection 2.4.3.3. Those events which established important elements of the warning system or those where the present model might produce significant differences in warning times have been reevaluated. Events reevaluated have been noted either in tables or figures where appropriate. S2-4app.doc 2.4A-11

SQN-17 The warning system is based on those storm situations which resulted in the shortest time interval between watershed rainfall and elevation 703, thus assuring that this elevation could be predicted at least 27 hours in advance. Hydrologic Basis for Warning System A minimum of 27 hours has been allowed for preparation of the plant for operation in the flood mode. An additional 4 hours for communication and forecasting computations are provided to translate rain on the ground to river elevations at the plant. Hence the warning plan must provide 31 hours from arrival of rain on the ground until critical elevation 703 could be reached. The 27 hours allowed for shutdown at the plant are utilized for a minimum of 10 hours of Stage I preparation and an additional 17 hours for Stage II preparation. This 27 hour allocation includes a 3-hour margin. Although river elevation 703, 2 feet below plant grade to allow for wind waves, is critical during final stages of plant shutdown for flooding, lower forecast target levels are used in most situations to assure that the 27 hours preflood transition interval will always be available. The target river levels differ with season. During the October 1 through April 15 "winter" season, Stage I shutdown procedures will be started as soon as target river elevation 697 has been forecast. Shutdown will be carried to completion if and when target river elevation 703 has been forecast. Corresponding target river elevation for the April 16 through September 30 "summer" season is 703. The one target river elevation in the summer season permits waiting to initiate shutdown procedures until enough rain is on the ground to forecast reaching critical elevation 703; shutdown would then be initiated and carried to completion. Inasmuch as the hydrologic procedures and target river elevations have been designed to provide adequate shutdown time in the fastest rising flood, longer times will be available in other floods. In such cases there will be a waiting period after the Stage I 10-hour shutdown activity during which activities shall be in abeyance until it is predicted from recorded rainfall that Stage II shutdown should be implemented or it is determined from weather conditions that plant operation can be resumed. Resumption of plant operation following Stage I shutdown activities will be allowable only after flood levels and weather conditions have returned to a condition in which 27 hours of warning will again be available. River Scheduling of River Operations prepares at least an 8-day water level forecast seven days per week for Tennessee River locations. During prospective flooding conditions forecasts can be prepared 4 times a day so that warnings for Sequoyah will assure that 27 hours always will be available to shut down the plant and prepare it for flooding. Hydrologic Basis for Target Stages Figure 2.4A.-4, in four parts, shows how target forecast flood elevations at the Sequoyah plant have been determined to assure adequate warning times. The floods shown are the fastest S2-4app.doc 2.4A-12

SQN-17 rising floods at the site which are produced by the 21,400-square-mile PMP with downstream centering described in subsection 2.4.3.1. The storms are the main PMP amounts and have been preceded 3 days earlier by a 3-day storm having 40 percent of the main storm rainfall. This has caused soil moisture to be high and reservoirs to be well above seasonal levels when the main storm begins. Figure 2.4A.-4 (A, B, and C) shows the winter PMP which could produce the fastest rising flood which would cross plant grade and variations caused by changed time distribution. The fastest rising flood occurs during a PMP when the 6-hour increments increase throughout the storm with the maximum 6 hours occurring in the last period. Figure 2.4A-4 (B) shows the essential elements of this storm which provides the basis for the warning scheme. In this flood 9.2 inches of rain would have fallen 31 hours (27 + 4) prior to the flood crossing elevation 703 and would produce elevation 697 at the plant. Hence, any time rain on the ground results in a predicted plant stage of 697 a Stage I shutdown warning will be issued. Examination of Figure 2.4A.-4 (A and C) shows that following this procedure in these noncritical floods would result in a lapsed time of 42 and 44 hours between when 9.2 inches had fallen and the flood would cross critical elevation 703. An additional 2.2 inches of rain must fall promptly for a total of 11.4 inches of rain to cause the flood to cross critical elevation 703. In the fastest rising flood, Figure 2.4A.-4 (B), this rain would have fallen in the next 5 hours. A Stage II warning would be issued within the next 4 hours. Thus, the Stage II warning would be issued 5 hours after issuance of a Stage I warning and 22 hours before the flood would cross critical flood elevation 703. In the slower rising floods, Figure 2.4A.-4 (A and C), the time between issuance of a Stage I warning and when the 11.4 inches of rain required to put the flood to elevation 703 would have occurred is 6 and 10 hours respectively. This would result in issuance of a Stage II warning not less than 4 hours later or 32 and 30 hours respectively before the flood would reach elevation 703. The summer flood shown by Figure 2.4A.-4 (D), with the maximum 1-day rain on the last day provides controlling conditions when reservoirs are at summer levels. At a time 31 hours (27 + 4) before the flood reaches elevation 703, 11 inches of rain would have fallen. This 11 inches of rain, under these runoff conditions, would produce critical elevation 703, so this level becomes both the Stage I and Stage II target. The above criteria all relate to forecasts which use rain on the ground. In actual practice quantitative rain forecasts, which are already a part of daily operations, would be used to provide advance alerts that need for shutdown may be imminent. Only rain on the ground, however, is included in the procedure for firm warning use. S2-4app.doc 2.4A-13

SQN-17 Because the above analyses have used fastest possible rising floods at the plant, all other floods will allow longer warning times than required for all physical plant shutdown activity. In summary, the predicted target levels which will assure adequate shutdown times are: Forecast Flood Elevations at Sequoyah For For Season Stage I Shutdown Stage II Shutdown Winter (October 1-April 15) 697 703 Summer (April 16-September 30) 703 703 Communications Reliability (HISTORICAL INFORMATION) Communication between projects in the TVA power system is via (a) TVA owned microwave network, (b) Fiber-Optic System, and (c) by commercial telephone. In emergencies, additional communication links are provided by Transmission Power Supply radio network. The four networks provide a high level of dependability against emergencies. The hydrologic network for the watershed above Sequoyah that would be available in flood emergencies if commecial telephone communications is lost include 138 rainfall gages (24 at power installations and 114 satellite and file transfer gages) and 47 streamflow gages (26 at hydroplants, 20 satellite gages, and 1 file transfer gage). River Scheduling is linked to the TVA power system by all four communication networks. The data from the satellite gages are received via a data collection platform-satellite computer system located in the River Schedulings office. These are so distributed over the watershed that reasonable flood forecasting can be done from this data while the balance of data is being secured from the remaining hydrologic network stations. The preferred, complete coverage of the watershed, employ 160 rainfall and 61 streamflow locations above the Sequoyah plant. Involved in the communications link to these locations are routine radio, radio satellite, and commercial telephone system networks. In an emergency, available radio communications would be called upon to assist. The various networks proved to be capable in the large floods of 1957, 1963, 1973, 1984, 1994, and 1998 of providing the rain and streamflow data needed for reliable forecasts. 2.4A.9 Basis for Flood Protection Plan in Seismic-Caused Dam Failures Floods resulting from combined seismic and flood events can exceed plant grade, thus requiring emergency measures. The 1998 reanalysis showed that only two combinations of seismic dam failures coincident with a flood would result in floods above plant grade: (1) failure of Fontana, Hiwassee, Apalachia, and Blue Ridge Dams in the one-half SSE concurrent with a 1/2 PMF, (2) SSE failure of Norris, Cherokee, and Douglas concurrent with a 25 year flood. As shown in Table 2.4.4-1 all other potentially critical candidates would create flood levels below plant grade elevation 705. Dam failure during non-flood periods would not present a problem at the plant. The reanalysis showed that failure in a non-flood period and at summer flood guide levels in the most critical dam failure combination (SSE failure of Norris, Cherokee and Douglas) would produce a maximum elevation of 703.6 at the plant, 1.4 feet below plant grade. All other combinations in non-flood periods would produce elevations much lower. S2-4app.doc 2.4A-14

SQN-17 The time from seismic occurrence to arrival of failure surge at the plant is adequate to permit safe plant shutdown in readiness for flooding. Table 2.4A-2 lists the time between the postulated seismic event and when the flood wave would exceed plant grade elevation 705 and elevation 703. Use of elevation 703 provides a margin for possible wind wave effects. The warning plan for safe plant shutdown is based on the fact that a combination of critically centered large earthquake and rain produced flood conditions must coincide before the flood wave from seismically caused dam failures will cross plant grade. In flood situations, an extreme earthquake must be precisely located to fail three or more major dams before a flood threat to the site would exist. The combination producing the shortest time interval between seismic event and plant grade crossing is a one-half SSE located so as to fail Fontana, Hiwassee, Apalachia, and Blue Ridge Dams during the one-half PMF. The time between the seismic event and the resulting flood wave crossing plant grade elevation 705 is 40 hours. The time to elevation 703, which allows a margin for wind wave considerations, is 35 hours. The event producing the next shortest time interval to elevation 703 involves the SSE failure of Norris, Cherokee, and Douglas during the 25-year flood resulting in a time interval of 63 hours. The warning system utilizes TVA's flood forecast system to identify when flood conditions will be such that seismic failure of critical dams could cause a flood wave to exceed elevation 703 at the plant site. Two levels of warning will be provided: (1) an early warning will be issued to SQN whenever a dam failure has occurred or is imminent for any single critical dam; or it appears from rain and flood forecasts that a critical situation may develop and (2) a flood warning or alert to begin preparation for plant shutdown when a critical situation exists that will result in the flood level to exceeding plant grade. A Stage I flood warning is declared once failure of critical dams has been confirmed and flood conditions are such that the flood surge will exceed plant grade. It shall be issued at least 27 hours before the flood level exceeds elevation 703 at the site. A Stage II flood warning will be issued at least 17 hours before the flood level exceeds elevation 703 at the site. Communication will be established and maintained during these two levels of warning to assure the 27 hour flood preparation period. Any prolonged interruption of communication or failure to confirm that a critical case has not occurred will result in the initiation of flood preparation at the plant site. The flood preparation shall continue until completion, unless communication is re-established and the site is notified that a critical case has not occurred. Communications between the plant, dams, power system control center, and River Operations at Knoxville, Tennessee, are provided by microwave networks, fiber-optic network, radio networks, and commercial telephone service. S2-4app.doc 2.4A-15

SQN-24 2.4A.10 References

1. SQN-DC-V-1.1, Design of Reinforced Concrete Structures Design Criteria

2. SQN-DC-V-12.1, Flood Protection Provisions Design Criteria

3. SQN-DC-V-43.0, High Pressure Fire Protection Water Supply System S2-4app.doc 2.4A-16

SQN-17 TABLE 2.4A-2 CRITICAL CASES - SEISMIC CAUSED DAM FAILURES TIME BETWEEN SEISMIC EVENT AND SELECTED PLANTSITE FLOOD ELEVATION Time in Hours Between Event and Plantside Elevation Dam Failed 703 705 One-half SSE failures with one-half probable maximum flood

1. Norris (2) (1)

2. Cherokee-Douglas (2) (1)

3. Fontana (2) 46 (1)

4. Fontana-Hiwassee-Apalachia-Blue Ridge 35 40 SSE failures with 25-year flood

5. Norris-Cherokee-Douglas 63 70

6. Norris-Douglas-Fort Loudoun-Tellico (2) (1)

(1) Elevation 705 not reached (2) Elevation 703 not reached T24A-2.doc

                                                                                                                                                                          / l.4"roin.on grour;,d wtou/d put reservoir to elev. 1 OJ -procaed W!fh sfoqe I [ s/7u/Oown Within                         Histroical These graphs would be 4 hours.\.                         i                    impacted by the safety modifications to I

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I -- SEQUOYl\H NUCLEAR PLANT FLOOD PROTECTION PLA~,J 6 sr) **-----*----*--------------------~:....--------::-=---' 20 21 22 23 24 25 26 27 BA S I S F 0 R S ,6- F E S H U TD 0 Wt'~ FOR PL/\NT FLOODING MAY Figure 2.4A- 4 HISTORICAL Revised by Amendment 17

SQN 2.5 GEOLOGY AND SEISMOLOGY 2.5.1 Basic Geologic and Seismic Data 2.5.1.1 Site Location and Scope of Exploration The Sequoyah plant site lies in Hamilton County, Tennessee, on a peninsula extending from the right shore into Chickamauga Lake between river miles 484 and 485 (Figure 2.5.1-1). The site first was explored in 1953. Twenty-nine holes were drilled into rock while 17 were fishtailed to the top of sound rock. From September 1968 to February 1969 additional holes were drilled to fill in a 100-foot grid in the control and auxiliary building area, and in the reactor areas, with holes drilled at the intake structure and other locations in the general plant area. In addition to obtaining information on the foundation conditions, the holes in the reactor areas were used for dynamic seismic investigations. During September and October 1969 a third drilling program was carried out to further investigate the reactor, control and auxiliary areas on a 50-foot spacing, and to examine the condition of the Kingston fault northwest of the plant site. For further details see ref. 84. 2.5.1.2 Physiography The Sequoyah site is located in the Appalachian Valley subregion of the Valley and Ridge Province of the Appalachian Highlands (Figure 2.5.1-1). Physiographically, this subregion is characterized by long narrow ridges and somewhat broader intervening valleys having a northeast-southwest trend. The ridges are roughly parallel and fairly evenly topped. They are developed in areas underlain by resistant sandstones and the more siliceous limestones and dolomites. The valleys have been excavated in the areas underlain by easily weathered shales and the more soluble limestone formations. In the vicinity of the Sequoyah site, the Tennessee River, prior to the impoundment of Chickamauga Lake, had entrenched its course to elevation 640. The small tributary Valley floors slope from the river up to around elevation 800, while the crests of the intervening ridges range between 900 and 1000 feet in elevation. 2.5.1.3 Geologic History The Sequoyah area lies near the western border of what was the active part of the Appalachian geosyncline during most of the Paleozoic era. During this time, the area was below sea level and more than 20,000 feet of sedimentary rocks were deposited. At the end of the Paleozoic era, some 250,000,000 years ago, the area was uplifted and subjected to compressive forces acting from the southeast. Folds developed which were compressed tightly, overturned to the northwest, and finally broken by thrust faults along their axial planes. The resultant structure, there- fore, is characterized by a series of overlapping linear fault blocks which dip to the southeast. Since this period of uplift, the area apparently has been above sea level and has been subjected to numerous cycles of erosion. This erosion accentuated the underlying geologic structure by differential weathering of the more resistant and less resistant strata resulting in the development of parallel ridges and valleys which are characteristic of the region. S2-5.doc 2.5-1

SQN 2.5.1.4 Stratigraphy Conasauga Formation The bedrock at the site is the Conasauga formation of Middle Cambrian age. In this region, the Conasauga is composed of interbedded limestone and shale in varying proportions. The shale, where fresh and unweathered, is dark gray, banded, and somewhat fissile in character. The limestone is predominantly light gray, medium grained to coarse crystalline to oolitic, with many shaly partings. A statistical analysis of the cores obtained from the site area indicates a ratio of 56 percent shale to 44 percent limestone. Farther to the southeast, higher in the geologic section, the amount of limestone increases in exposures along the shore of the lake. 2.5.1.5 Structure The controlling features of the geologic structure at the Sequoyah plant site are the Kingston Thrust fault and a major overturned anticline which resulted from the movement along the fault. This fault lies about a mile northwest of the plant site (Figure 2.5.1-2) and can be traced for 75 miles northeastward and 70 miles southwestward. The fault dips to the southeast, under the plant site, and along it steeply dipping beds of the Knox dolomite have been thrust over gently dipping strata of the Chickamauga limestone. The distance from the plant site, about one mile, and the dip of the fault, 30 degrees or more, will carry the plane of the fault at least 2000 feet below the surface at the plant site. The major overturned anticline results in the Conasauga formation at the plant site resting upon the underlying Knox dolomite which normally overlies it (Figure 2.5.1-3). As a result of the ancient structural movement of the fault and major fold, the Conasauga formation at the plant site is highly folded, complexly contorted, and cut by many very small subsidiary faults and shears. The general strike of these beds are N 30 degrees E and the overall dip is to the southeast, but the many small tightly folded, steeply pitching anticlines and synclines result in many local variations to the normal trend. In some of the drill cores, small faults and shears were noted intersecting the bedding at various angles. These dislocations are the result of shearing along the limbs of the minor folds which developed contemporaneously with the major movement along the Kingston fault. The Kingston fault is only one of the several lengthy thrust faults which characterize the geologic structure of the Appalachian Valley, a part of the "Valley and Ridge" physiographic province. A study of any one of these faults involves a consideration of the major structural features of the Valley as a whole. Structurally, the Appalachian Valley in eastern Tennessee is characterized very largely by a series of overlapping linear fault blocks of northeast-southwest strike and southeast dips. Most studies have attributed the deformation in the Southern Appalachians to the Appalachian orogeny at the end of the Paleozoic era. It has been assumed that the major tectonic structures have been inactive since the cessation of the orogenic movement. The duration of this orogenic epoch cannot be determined precisely in the Southern Appalachians since the Pennsylvanian strata are the youngest rocks known to have been affected. That some deformation continued after the major faults had attained their present development is attested by folded and faulted thrust sheets. These late structures may represent the final phase of the orogeny. S2-5.doc 2.5-2

SQN The only undeformed materials occurring in the Valley as mappable units are the unconsolidated materials: alluvial deposits, including the high level terrace deposits as well as the recent floodplain alluvium, and the residuum that nearly everywhere mantles bedrock. The alluvium along the Tennessee River and its tributaries ranges in age from less than a decade at the top up to several tens of thousand years at the base. The higher terrace deposits are much older than the lower terraces. The high level terraces have been considered as Pleistocene (King, 1949, page 89) or even older. The residuum which blankets the bedrock in the Appalachian Valley ranges in thickness from a feather's edge up to a maximum of a hundred feet or more. The age range within a thick accumulation of residuum has not been determined, but the oldest part of the residuum may be of Paleocene or even later Upper Cretaceous age. In several areas of the Valley, masses of bauxite occur in association with brown iron ores and lignite in the thick residuum over limestones and dolomites. The bauxite and the associated materials accumulated in the sinks or sink-like depressions. Bridge (1950, page 194) considers these deposits to be late Paleocene. The following quotation is from Rodgers: "The age of the residuum is even less definite. Weathering is going on and presumably some residuum is being formed now, yet some residuum was apparently already present when the bauxite-bearing clay bodies formed in their sinkholes." Thus it has probably been forming virtually throughout Cenozoic time, though perhaps at a greater rate at certain times, such as those of little stream erosion, than at others. Several lines of evidence suggest a time of particularly intensive chemical decay and activity during or after the formation of the "Valley Flood Peneplain" in the Appalachian Valley, perhaps in the earlier Cenozoic (King and others, 1944, pages 24-25, 59; Rogers, 1948, pages 15, 40; King, 1949, pages 82-83; Bridge, 1950). As indicated above, the age of the various unconsolidated materials in the Appalachian Valley of eastern Tennessee can be at best only estimated in very general terms. The bedrock and its structures are concealed very largely by these materials. The lack of any evidence of faulting, creep, or renewed movement in the unconsolidated materials even along the major tectonic faults indicates that there has been no movement along these faults for a very long time. This is true of the Kingston fault and all of the other numerous faults in the area. No formal trenching or age dating was attempted at the Sequoyah plant. The evidence previously cited is related to general observations and the field mapping experience of dozens of geologists for the past 100 years. None of the reports published by geologists working in east Tennessee mention any evidence of actual observations of displacement of surface features which relate to fault movement in historic time. More positive evidence comes from a branch of the Kingston fault called the Missionary Ridge Fault. The Missionary Ridge fault is a branch, or subsidiary, fault of the Kingston fault (Rodgers 1953, page 130-131, Plate 15, Figure 10). It runs northwest from the Kingston fault and has a total length of approximately 25 miles extending southwestward from the point where it diverges from the Kingston fault, 3 miles southwest of the Sequoyah site, and dying out in northwest Georgia (Hardeman, 1966; Butts and Gildersleeve, 1948). Along most of its length Cambro-Ordovician Knox dolomite and limestone are thrust over Middle and Upper Ordovician Chickamauga limestone. Near its southern terminus Knox is thrust over the Silurian Red Mountain formation. S2-5.doc 2.5-3

SQN The Missionary Ridge fault crosses the Tennessee River just upstream from Chickamauga Dam. In 1848 a railroad tunnel was driven through Missionary Ridge in Chattanooga and in the process the tunnel crossed the Missionary Ridge fault. The lining of this tunnel was inspected in 1974 and no cracking of the lining, offset along joints, or other signs of structural defects were found that would indicate any evidence or movement along the Missionary Ridge fault in the last 125 years. Three other vehicular tunnels through Missionary Ridge were also inspected and no structural indications of possible fault movement were found. TVA has drilled through some of the major faults in eastern Tennessee. Diamond core borings at Chickamauga Dam (1935-1936) went through the Missionary Ridge fault and the cores through the fault zone came out unbroken. The fault was not simply "healed" or recemented with secondary deposits of calcite or dolomite, but was a very tight contact along which apparently pulverized material had recrystallized. The recrystallization and solidification of the material along the fault plane indicated that this material had not been disturbed by renewed movements for an unknown, but apparently very long, period. Until recently, no indication of how long a period since the last movement was available. In studies for the Clinch River Breeder Reactor Plant, Law Engineering obtained similar material from the Copper Creek fault, one of the same family of faults as the Kingston and Missionary Ridge faults in east Tennessee, and obtained radiometric dates of 280 to 290 million years, +/- 10 million years. The results of these tests indicate that the last movements on these faults occurred during the late Paleozoic. Core borings have been made through at least one other major thrust fault in eastern Tennessee. It was reported to be "solid" similar to that through the Missionary Ridge fault. Although light earthquakes occasionally occur in the Valley of eastern Tennessee, there has not been a single instance in which the surface was deformed. The shocks are of "normal" focus, 15 to 20 km, but even at such shallow depths, the hypocenters are in the crystalline basement rock well below the sedimentary rocks. As previously stated, a study of any one of our major thrust faults involves a consideration of all the other similar faults. Many of the geologists who have spent years doing geologic work in eastern Tennessee believe that the several named faults are merely branches of a single nearly flat sole fault developed in some relatively incompetent formation just above the crystalline basement. Some, if not all, of the thrust sheets flatten out with depth, and some of them are cut through by erosion. It was not until early 1974 that definitive evidence was released to support the "thin-skinned" hypothesis. At that time Geophysical Services Incorporated published an advertising brochure describing reflection seismic data they had available for sale. The example of a reflection profile used in their brochure was made along U.S. Highway 70 from near Kingston, Tennessee, to the vicinity of Knoxville, Tennessee. This profile essentially at right angles to the regional strike is reproduced in Figure 2.5.1-4. The vertical scale of this profile is represented in seconds. This indicates the double travel time necessary for the shock wave to descend to the reflector and return to the surface. Assuming a S2-5.doc 2.5-4

SQN wave velocity of 20,000 ft/s, the times indicated equate to depths in thousands of feet. The "thin-skinned" tectonic structure of the upper strata, above the 1.5 second (15,000 foot) line, is clearly indicated. The depth of approximately 15,000 feet to basement strata in this area is confirmed by gravity and magnetic data (Watkins, 1964). The significance of the confirmation of "thin-skinned" tectonics in the area in relation to the geologic and seismic considerations of the Sequoyah plant lies in the fact that data now exist to show the separation of faults cropping out at the surface from geologic structures in the basement at a depth of approximately 15,000 feet or 4.5 km. This means that earthquakes with hypocenters at depths of five or more kilometers cannot be associated with faults cropping out at the surface even though the epicenter (surface projection of the hypocenter) falls on or near the trace of the fault. The evidence available from all of the geologic studies that have been made suggests that all of the Appalachian Valley faults, including the Kingston fault, are inactive. In the voluminous literature on the geologic structure of the Southern Appalachians, there is no mention of the possibility that any of the faults may still be potentially active. 2.5.1.6 Groundwater See Section 2.4.13. 2.5.1.7 Physical Character of the Rocks Unconfined compressive strength determinations were made on seven core samples from the Sequoyah site. The results of these tests gave compressive strengths varying from 16,794 lb/in2 and 11,936 lb/in2 for limestone and 5758 lb/in2 for shale. Seismic methods were used to determine the dynamic moduli of the foundation. The results of this work are explained below. Seismic measurements were made in boreholes located in the two proposed reactor foundations. The purpose of these measurements was to determine the dynamic modulus of elasticity, E, for these foundations so that an earthquake design criteria could be established. Laboratory velocity measurements of core samples were not made because the varying changes in rock types would not give valid results. The bedrock in which the seismic measurements were made is the Conasauga formation of middle Cambrian age. It is composed of inter-bedded lime- stone and shale in varying proportions. The shale, when unweathered, is dark gray to green, and somewhat fissile in character. In its weathered state it is very soft and in some cases has some of the characteristics of clay. The limestone is predominantly light gray, medium to coarse crystalline, oolitic, with many shaly partings and calcite healed fractures. The rock is badly contorted with dips ranging from 5 degrees to 90 degrees. Results of the Dynamic Testing Program Tables 2.5.1-1 and 2.5.1-2 give the results of the seismic studies that were made for each of the two reactor foundations. The average density of the rock is approximately 170 lb/ft3. Density values from representative core samples were established at 170 lb/ft3 and 169 lb/ft3. S2-5.doc 2.5-5

SQN Tables 2.5.1-1 and 2.5.1-2 give the up-hole and cross-hole velocity measurements by which the E was calculated from formulae shown on Table 2.5.1-3. The difference in the values is thought to be attributed primarily to the changes in dip and rock type for each borehole. The average up-hole modulus for both reactor foundations is 4.2 x 10E6 and for the cross-hole modulus it is 4.4 x 10E6 lb/in2. 2.5.1.8 Foundation Conditions As shown on Figures 2.5.1-5 through 2.5.1-8, bedrock was mantled by a varying thickness of residual material derived from the weathering of the underlying shale and limestone. As would be expected in a foundation composed of alternating strata of different composition and competency, the configuration of the bedrock surface was irregular. The strike of the rock strata is approximately parallel to the centerline of the reactors. Preliminary excavation down to 18 inches above design grade resulted in a series of alternating ridges of harder limestone separated by troughs underlain by the softer shale trending across the plant area. The last 18 inches were removed by careful and controlled means so as to limit breakage below the design grade to a minimum. Once foundation grade was reached, the area was carefully cleaned and then inspected jointly by engineers and geologists to determine what, if any, additional material needed to be removed because of weathering or shattering by blasting. After the final excavation was approved, the area was covered either by a coating of thick grout or a fill pour of concrete to prevent breakdown of the shale interbeds due to prolonged exposure. Observation of rock exposed in the foundation areas, examination of cores, and investigations of the walls of exploratory holes with a borehole television camera all indicated that solution cavities or caves are not a major problem in the foundation. Verified cavities generally were limited to the upper few feet or rock where solution developed in limestone beds near the overburden-rock interface. Practically all of this zone was above design grade and was removed. Inspection of other areas of nonrecovery of core at greater depths by the borehole television equipment proved that so-called cavities as reported by the drillers were in fact interbeds of shale that had been ground between overlying and underlying harder limestone strata. In the walls of the holes the camera showed solid shale in these nonrecovery areas. Large solution cavities are not to be expected in formations such as the Conasauga which are made up of interbedded limestone and shale. The insolubility of the shale precludes the development of large openings. Inspection of the walls of the exploratory holes with television disclosed thin, less than 0.05 foot, near-horizontal openings in some of the limestone beds. At the corresponding position, the drill cores showed unweathered breaks. These open partings are interpreted as "relief joints" developed by unloading either from erosion or excavation. The majority were found in the upper few feet of rock, but some were observed as deep as 131 feet below the rock surface. A consolidation grouting program was carried on from February 18, 1970 through June 15, 1970 in the foundation areas for the Reactor, Auxiliary, and Control Buildings at the Sequoyah Nuclear Plant. The extent of the area treated is shown on Figures 2.5.1-9 and 2.5.1-10. S2-5.doc 2.5-6

SQN The purpose of this program was twofold. The first was to consolidate near-surface fractures predominantly caused by blasting and excavation. The second was to treat any localized open joints, bedding planes, fractures, or isolated small cavities that pre-construction exploratory drilling indicated might be present to a depth of 45 feet below the design foundation grade. In the excavated area the contact between the residual material and essentially unweathered rock occurs at an average elevation of 680. The highest design level for the plant foundation grade under the Class I structures is at elevation 665. As a result, the preliminary excavation averaged a minimum of 15 feet in rock. Over most of the area the rock was suitable for foundation purposes at elevation 665. In two areas, however, additional rock had to be excavated to remove localized pockets of deeper weathering. These zones were confined in two synclinal areas which crossed the excavation parallel with the north- south baseline. The axis of one lies approximately 70 feet plant east of the baseline and the axis of the other is approximately 140 feet plant west of the baseline. These trough-like synclines had channeled ground- water movement toward and along their axes with the result that weathering had progressed deeper in these areas. Generally, less than 10 feet of additional rock had to be removed from the synclinal zones to obtain a satisfactory foundation; however, in the vicinity of W 140; S 220, on the south side of the Auxiliary Building, as much as 30 feet of weathered rock was removed. The limits of the synclinal areas are reflected on Figure 2.5.1-10 as zones of appreciable grout take. Elsewhere in the foundation area grout takes were minimal. This treatment program was approached in the same manner as a consolidation grouting program under a major dam. Grout crews with experience in grouting dam foundations were used, and the onsite technical direction of the program was performed by a member of the Geologic Branch who had previously supervised grouting operations at major dams. All grouting was done in strict accordance with TVA specification G-26, Pressure Grouting of Rock Foundations with Portland Cement. While the grouting was in progress, the program was reviewed in the field at least weekly by a senior member of the Geologic Branch. Prior to the start of any grouting, it was proposed to excavate the foundation area to be treated to a depth of two feet below required design grade. In practice, due to the irregularities of the rock foundation, this overexcavation varied from a minimum of 18 inches to a maximum of nearly 30 feet. As each section of the foundation was prepared, it was inspected and approved by a joint team consisting of representatives of the Division of Construction, the Division of Engineering Design, and the Geologic Branch. When the area was released by the inspection team, fill concrete was poured up to the design foundation grade. This fill pour acted as a grout cap, protected the shale strata in the bedrock from any tendency to slake or ravel due to prolonged exposure, and provided a good working surface for the grouting operations. The data contained in columns 3 and 5 of Table 2.5.1-4 indicate the tightness of the foundation. As shown in column 3, in the primary holes--those drilled over the entire area on a 20-foot grid--only 11 percent of the 10-foot-deep holes and 23 percent of the 45-foot-deep holes accepted any grout. This confirms the assumption made from the evaluation of the exploratory drilling, that grout takes would be confined to localized areas. Further confirmation is supplied by the relatively low percentage of holes with grout takes in the subsequent series of split S2-5.doc 2.5-7

SQN spaced holes. Normally, it would be expected that a high percentage of the split-spaced holes, especially the secondary holes, would accept grout because they were drilled in areas shown by the primary holes to require further treatment. Although these percentages were higher than for the primary holes, they never exceeded 50 percent and usually were less than 40 percent. A layout of the investigative programs for the other category I structures is presented as Figure 2.5.1-11. Sections of Category I structures supported on soil, piles, or caissons are provided on Figures 2.5.1-12,-12a, and -12b. The ERCW piping and conduit support slab which is founded on piles to rock is shown in section on FSAR Figure 3.8.4-9. The sections show general details of excavation and backfill limits for the Category I structures as well as the type of foundation. The classifications of borrow materials are discussed in Subsection 2.5.1.11. The Sequoyah foundation was completed prior to the time Atomic Energy Commission (AEC) began requesting commitments to produce geologic maps of the foundation. Therefore, detailed data such as were presented for the Watts Bar Nuclear Plant are not available. There are available several hundred photographs of the rock foundation. TVA has submitted by letter a series of photographs which give the best representation of the overall foundation. In addition to the photographs, quality assurance forms were included which indicate approval of rock conditions prior to all concrete subpours in the Reactor, Auxiliary, and Control Building areas. Rock inspections were made by a senior geologist and by senior design engineers who initiated the forms. 2.5.1.9 Physical Characteristics of Soils 2.5.1.9.1 Static Physical Characteristics of Soils A soils exploration program was conducted at the plant site to determine the static physical characteristics of the soils. Standard penetration split-spoon borings and undisturbed borings were made. Figure 2.5.1-13 shows the location of all borings made at the site for in situ soil sampling and testing. Graphic logs of all borings are kept on file by TVA. 2.5.1.9.2 Dynamic Characteristics of Soils In situ soil dynamic studies were made at the plant site to obtain data for computation of elastic moduli for earthquake design criteria. The areas investigated at the site were the Diesel Generator Building, the Low Level Radwaste Storage Facilities, the ERCW pipeline, the Additional Diesel Generator Building, and the Primary Water Storage Tank.

1. Diesel Generator Building Down-hole seismic surveys and a seismic refraction survey were performed. The results are tabulated on Table 2.5.1-9.

S2-5.doc 2.5-8

SQN

2. Low Level Radwaste Storage Facilities Both compressional and shear wave velocities were obtained through a series of cross-hole and down-hole measurements. The results are tabulated on Table 2.5.1-10 and 2.5.1-11.

3. Essential Raw Cooling Water Pipeline Down-hole seismic surveys were made. The results are tabulated on Table 2.5.1-12.

4. Additional Diesel Generator Building Cross-hole and down-hole seismic surveys were performed. The results are tabulated on Table 2.5.1-13.

5. Primary Refueling Water Tanks Seismic refraction surveys were made. The results are tabulated on Table 2.5.1-14.

2.5.1.10 Detailed Safety-Related Criteria and Computed Factors of Safety For the Materials Underlying the Foundations for Category I Structures

1. Category I Rock-Supported Structures The allowable rock-bearing pressure for sustained loading was determined based on the strength and stratigraphy of the foundation rock. The result using the physical characteristics of the foundation rock as described in section 2.5.1.7, and the geologic characteristics given in section 2.5.1.4 provided a reasonable bearing pressure. The allowable rock-bearing capacity is less than the ultimate bearing capacity by a factor of 2.5.

Table 2.5.1-5 lists the structures which are constructed with a base slab directly on rock. The table shows the allowable static and dynamic bearing pressures.

2. Category I Structures Supported by H-Piles or Caissons to Rock There are four Category I structures founded on piles or caissons. The structures are the East Steam Valve Room, the Waste Packaging Area, the Condensate Demineralizer Waste Evaporator Building, and the ERCW piping and conduit support slab in the ERCW pumping station access dike.The East Steam Valve Rooms were backfitted with caissons into rock after experiencing some settlement.

The Waste Packaging Area, the Condensate Demineralizer Waste Evaporator Building, and the ERCW piping and conduit support slab in the ERCW pumping station access dike are all supported on H-piles founded on rock.

3. Category I Soil-Supported Structures The allowable soil-bearing capacity for sustained loading is determined using the general shear failure formula, developed by Terzaghi and modified by Meyerhof.

S2-5.doc 2.5-9

SQN The allowable bearing pressure for sustained loads is less than the ultimate bearing by at least a factor of three. For dynamic loading the soil-bearing pressure is permitted to exceed the allowable for sustained loading. In no instance is the ratio of the ultimate soil-bearing pressure to the allowable soil pressure less than two. Table 2.5.1-6 contains a summary of the allowable soil-bearing capacities and factors of safety for the soil-supported Category I structures.

4. Category I Embankments See Subsection 2.5.6.

2.5.1.11 Compaction Criteria for Engineering Backfill 2.5.1.11.1 Earthfill Prior to and during construction, borrow investigations were made. These investigations were made on an as needed basis. The borrow samples were tested by the central materials laboratory according to ASTM D-698 to develop compaction control curves. The compaction curves were divided into subclasses, and these compaction curves are shown on Figures 2.5.1-14 and -15. These curves were used by the project laboratory to control compaction of earthfill at the site. At Sequoyah Nuclear Plant, Type A backfill was placed around all Category I structures. This material, which was selected earth placed in not more than 6-inch layers, has a minimum required compaction of 95 percent of the maximum dry density at optimum moisture content. The limits of excavation and the backfill around the Category I structures are shown in Figures 2.5.1-12,-12a, and -12b. Tables 2.5.1-7 and 2.5.1-8 are a summary of field control tests on Type A backfill. 2.5.1.11.2 Granular Fill Crushed Stone Fill A free draining granular fill material, consisting of crushed stone or sand and gravel, was placed below or next to Category I structures. This material was obtained commercially from off-site sources. The granular fill was suitable for compaction to a dense, stable mass and consisted of sound, durable particles which are graded within the following limits: Percent by Weight Passing Minimum Maximum 1-1/4-inch sieve 100 1-inch sieve 95 100 S2-5.doc 2.5-10

SQN Percent by Weight Passing Minimum Maximum 3/4-inch sieve 70 100 3/8-inch sieve 50 85 No. 4 sieve 33 65 No. 10 sieve 20 45 No. 40 sieve 8 25 No. 200 sieve 0 10 The material was free of disintegrated stone, soft friable particles, shale, salt, alkali, organic matter, or an adherent coating and reasonably free of thin, flat, or elongated pieces. The granular fill material was used; for structural support, to replace earthfill as a backfill material around piping or conduits during wet weather, and to provide a working base above wet soil. The material, when used for structural support, or replacement for earthfill, was compacted to a required relative density as determined by ASTM D 2049. When used for structural support, such as for the refueling water storage tank (Figure 2.5.1-12b), an average relative density of 85 percent or greater with a minimum relative density of 80 was required. When used as a replacement for earthfill, a relative density between 70 and 85 percent was required. Limestone Sand Fill A granular fill material that meets the gradation requirements of ASTM C 33 was used as backfill material around the ERCW piping along the piping alignment from the intake Pumping Station to the ERCW Pumping Station access dike. The gradation limits for the material are: Percent by Weight Passing Minimum Maximum 3/8" sieve 100 No. 4 sieve 95 100 No. 8 sieve 80 100 No. 16 sieve 50 85 No. 30 sieve 25 60 No. 50 sieve 10 30 No. 100 sieve 2 10 The granular fill was compacted to an average relative density of 75 percent or greater, with a minimum relative density of 70 percent as determined by ASTM D 2049. 2.5.1.11.3 Crushed Rock A crushed rock material that meets the gradation requirements shown below was used to S2-5.doc 2.5-11

SQN construct the core of the ERCW access dike and the material was also used for remedial treatment in local areas. The gradation limits for the material are: Passing Percent by Weight Minimum Maximum 3-inch sieve 95 100 2-inch sieve 25 55 1-1/2-inch sieve 0 15 1-inch sieve 0 2 The material consisted of sound durable particles; free of soft friable particles, shale, salt, organic matter, or an adherent coating (other than dust); and reasonably free of thin, flat or elongated pieces. ERCW Access Dike The ERCW Access Dike as shown on Figure 3.8.4-9 connects the ERCW Pumping Station Access Cells with the shore. The dike core was placed by end dumping the rockfill material between the shore and the access cells up to elevation 676.75 (1.75 feet above normal minimum reservoir level). Compaction was obtained using a vibratory roller. Above elevation 676.75, between the access cells and the shore, the rockfill material was placed in lifts and compacted using the same vibrating roller. Remedial Treatment The rockfill material was used in several locations at the site to improve the soil. This was generally done where moisture caused the soil to be unsatisfactory as a base for earthfill placement. The material was used in a limited area at the refueling water tank pipe tunnel. The material was placed in approximate 6-inch loose layers and rolled into the soil. If the required stiffness for the placement of earthfill was achieved, lifts of earth- fill or crushed stone fill were placed. If the required stiffness was not achieved, then additional lifts of the material were placed and rolled to obtain the desired stiffness. If shearing or pumping accurred in placement of the first lift, additional lifts of the material were placed as necessary. 2.5.2 Vibratory Ground Motion The lithologic, stratigraphic, and structural conditions at the site and in the surrounding area and the geologic history of the region have been discussed previously in Paragraphs 2.5.1.3, 2.5.1.4, and 2.5.1.5, and will not be repeated here. The static and dynamic engineering properties of the materials underlying the site are described in Paragraphs 2.5.1.7 through 2.5.1.9. 2.5.2.1 Regional Tectonics The fact that Pennsylvanian strata were involved in the deformation of the Valley and Ridge province in the Southern Appalachian area has in the past been taken as conclusive evidence that the structural features of the Appalachian system were formed near the end of the Paleozoic S2-5.doc 2.5-12

SQN Era. This has been termed the "Appalachian Revolution." This late Paleozoic orogeny, however, may have been only one of many movements, and in fact may have been a relatively mild concluding phase. The orogenic and tectonic history of the southern Appalachian geosyncline is composite. The lower part, up to about the middle of the Ordovician, is a thick mass of carbonates with sandstone at the base. These deposits indicate a time of crustal quiescence, with slow sinking of the area of deposition, and low marginal lands. The succeeding clastics, laid down in Middle Ordovician and later times, express a radical change in the environment of the geosyncline. The source of the sediments was now from the southeast and was probably orogenic in origin. In the southern Appalachians, the first orogenic movement indicated by the sediments of the geosyncline took place in Middle Ordovician time. This is somewhat earlier than the late Ordovician and early Silurian Taconian movements of the northern Appalachians, but may be considered a phase of the Taconian orogeny. To the southeast is a thick mass of shales and sandstones of Middle Ordovician Age, succeeded by red sandstones and siltstones, probably also Middle Ordovician. Farther northwest, all the Middle Ordovician is limestone, but the Upper Ordovician includes shales and red beds. These beds are topped by cleanly washed, quartzose Silurian sandstones, a post-orogenic deposit. Orogenic movements at about this time in the metamorphic and plutonic belt on the southeast are suggested by radioactive determinations which indicate that some of the pegmatites of that area are of Ordovician Age. Acadian, or late Devonian and early Mississippian, orogeny of the northern Appalachians seems to be poorly represented in the southern Appalachians. Slight early Mississippian movements, possibly a late phase of the Acadian orogeny, are expressed by clastic rocks of early Mississippian Age. However, Middle Paleozoic time in the southern Appalachians seems to have been mainly one of quiescence and readjustment, following the Ordovician orogeny. The next period of orogeny suggested by the sediments of the Valley and Ridge province probably took place in late Mississippian and early Pennsylvanian time, or at about the same time as the Wichita orogeny west of the Mississippi Embayment. Deposits of late Mississippian and early Pennsylvanian age thicken markedly southwestward along the Valley and Ridge province and reach their climax in the southeastern belts of outcrop in Alabama. If these thick late Mississippian and early Pennsylvanian deposits are related to orogeny, that orogeny must have occurred in the region southeast of the present belts of outcrop, for the deposits lie with apparent conformity on the beds beneath and share with them the strong folding and faulting of the Valley and Ridge province. No Paleozoic deposits younger than the Pottsville are present southwest of West Virginia and Kentucky. There may have been Arbuckle movements of late Pennsylvanian and early Permian age, and there may have been also Appalachian movements of late Permian age. Since the end of the Paleozoic the southern Appalachian mountain system has stood as a positive area and has undergone profound erosion. The present topography is the result of differential weathering of strata of varying resistance. The more durable units underlie the higher areas and the valleys are cut in softer formations. This differential erosion in the Valley and S2-5.doc 2.5-13

SQN Ridge Province has accentuated the long northeast-southwest trending series of fault belts that developed in the Paleozoic and have remained quiescent since. The Valley and Ridge Province from Roanoke, Virginia, southwestward is characterized by a series of overlapping linear fault blocks of northeast-southwest strike and southeast dip. Along the southeast margin of the province the Lower Cambrian and Pre-Cambrian strata have moved northwestward along the Great Smoky fault as much as 20 to 30 mi. as evidenced by exposures of Upper Cambrian and Ordovician strata in windows eroded through the thrust plate far southeast of the present mountain front. While this was happening, the less competent strata to the northwest were shingled into a series of imbricate thrust plates. The soles of these plates are normally incompetent shales in or below the Middle Cambrian Rome formation. On the present surface as many as 10 of these sheets can be defined across the Valley and Ridge Province in Tennessee. Most geologists familiar with the area now believe that there are two to four "master thrusts," such as the Pulaski, Saltville, and Pine Mountain, and others are subsidiary branches off the major faults. It is also believed that these faults do not extend into the basement but are a series of decollements developed in some relatively incompetent formation above the crystalline basement. There is no geologic evidence indicating that any of these faults could be considered to be "active" faults; that is, still undergoing movement. On the contrary, all geologic evidence points to the fact that they have not moved since the close of the Paleozoic era. Drainage patterns are controlled by the relative competency or incompetency of the strata crossed by the streams and do not indicate offsets where crossing faults. There is no evidence of creep, faulting, or renewed movement in the unconsolidated residual or alluvial deposits overlying the fault traces nor any observable offset of Plio-Pleistocene high level alluvial terraces. In exploration for various sites in the TVA area, some of these major fault planes have been intersected by exploratory drill holes. As an example, during the exploration for Chickamauga Dam near Chattanooga, Tennessee, cores across the Missionary Ridge fault were recovered unbroken. The fault was not simply "healed" or recemented with secondary deposits of calcite or dolomite, but was a very tight contact along which apparently pulverized material had recrystallized. In another instance at the Tellico Project near Knoxville, Tennessee, the Knoxville fault was cored in 10 holes and again the core across the fault was recovered unbroken although the stratigraphic displacement is in the neighborhood of 10,000 feet and the lateral displacement can be measured in miles. The evidence available from all of the geologic studies that have been made indicates that all of the thrusts in the Valley and Ridge Province are inactive. In the voluminous literature on the geologic structure of the southern Appalachians, there is no mention of the possibility that any of the faults may still be potentially active. Although light earthquakes occasionally occur in the region, there has not been a single instance where the surface has been deformed. These shocks are all of "normal" focus, 15-20 km deep, but even at these relatively shallow depths the hypocenters are well into the crystalline basement rocks far below the 5 km maximum thickness of the sedimentary cover. For this reason, any map showing epicenters of earthquakes in this area plotted in relation to fault traces gives an erroneous impression, for any such map drawn to a reasonable scale will show some epicenters falling near or on some of the relatively closely spaced thrust faults to which they are in no way related. S2-5.doc 2.5-14

SQN 2.5.2.2 Site Area Tectonics In recognition of the fact that sites in the southern Appalachians cannot reasonably be tied to any one "tectonic structure," NRC (formally AEC) in the preliminary evaluation of the Sequoyah Nuclear Plant defined a "Southern Appalachian Tectonic Province." This province is bounded on the east by the western margin of the Piedmont Province; on the west by the western limits of the Cumberland Plateau; on the south by the overlap of the Gulf Coastal Plain Province; and on the north by the re-entrant in the Valley and Ridge Province near Roanoke, Virginia. The limits of the province are shown on Figure 2.5.2-1. Under this concept accelerations at the site will be determined by assuming that the largest historic earthquake known in the province occurred adjacent to the site. For the Sequoyah site, this earthquake would be the May 31, 1897 quake in Giles County, Virginia, which had a reported epicentral intensity of MM VIII. In the specific site area there is no physical evidence of disturbance of surficial materials during prior earthquakes. Minor dislocations and shears in the substrata are directly related to movements along the major thrust faults which moved in the Paleozoic and have been "fossilized" since that time. The majority of these are healed and recemented although they do serve as loci for near-surface development of solution and cavities in the limestone strata. 2.5.2.3 Seismic History The evaluation of the earthquake hazard at the Sequoyah site involves a consideration of the known seismic history of a large surrounding area. By plotting the epicenters of hundreds of earthquake shocks, the areas of continuing seismic activity become apparent. The more active areas are described in the following summary.

1. Mississippi Valley, especially the New Madrid region of Arkansas, Kentucky, Missouri, and Tennessee. This region has been active seismically since the appearance of the white man and very probably long before that. A few great earthquakes and thousands of light to moderately strong shocks have been centered in the Mississippi Valley. Light to moderate shocks are still occurring at an average frequency of a few per year. The New Madrid region is more than 250 miles northwest of the Sequoyah site.

2. The Lower Wabash Valley of Illinois and Indiana. This area has been the center of several moderately strong earthquakes, some of which were felt as far south as Nashville, Tennessee. It is about 260 miles northwest of the Sequoyah site.

3. Charleston area, South Carolina. One of the country's greatest earthquakes was centered in the Charleston area. Earlier, many light to moderate shocks had been centered in the area long before the great earthquake, and the activity has continued to the present time. Charleston is more than 300 miles east of the Sequoyah site.

4. The Appalachian Mountains of eastern Tennessee and western North Carolina. The mountain belt of eastern Tennessee and western North Carolina is a region of continuing minor activity.

Light to moderate shocks occur at an average frequency of one or two per year. The activity is not uniform, as periods of several shocks per year are followed by longer periods of no perceptible shocks. This region is centered more than 50 miles to the east of the Sequoyah site. S2-5.doc 2.5-15

SQN In addition to these areas, shocks of light to moderate intensity have occurred at numerous other localities in the southeastern states at various distances from the Sequoyah site. At many of these localities, only a few light to moderate shocks from widely scattered epicenters are known. A few such shocks have occurred to the north and east of Huntsville, Alabama. Numerous light shocks have occurred in Knoxville and its environs. An annotated list of the earthquakes which have either affected the Sequoyah area or were centered somewhere near the area is presented below. In each case, the maximum intensity, or that applicable to the Sequoyah area, is assessed in terms of the modified Mercalli scale. 1811, December 16: 36.6° N - 89.6° W 1812, January 23: 36.6° N - 89.6° W 1812, February 7: 36.6° N - 89.6° W These were the strongest shocks of the great series of earthquakes of 1811-1812 centered in the Mississippi Valley and known collectively as the New Madrid earthquake. This series consisted of thousands of individual shocks, many of which were strong. The three strongest shocks had an intensity of XII in their epicentral areas, and were felt over an area of about 2,000,000 square miles. Topographic changes were effected over an area of 3000 to 5000 square miles in the Mississippi Valley. The three great shocks and many of the other strong shocks were felt in the Sequoyah area, where some of them may have attained intensities as high as VI or VII (Figure 2.5.2-2). 1843, January 4: 35.2° N - 90° W. A severe earthquake centered in the Mississippi Valley was felt over some 400,000 square miles in a 12-state area. Chimneys were thrown down in Memphis, Nashville, and St. Louis. Although the intensity was perhaps as high as in the epicentral area, it is not known to have attained damaging intensities in Alabama. This shock was perceptibly felt over the entire Tennessee Valley and may have had an intensity as high as V or VI in the Sequoyah area. 1861, August 31: A strong earthquake, thought to have been centered in Virginia, was felt from Washington, D.C., southward to Wilmington, North Carolina, and westward to Knoxville, Cincinnati, and Louisville. At Knoxville it was described as a "heavy shock" which "alarmed the encamped military very much." It may have affected the Sequoyah area at an intensity of III or IV. 1886, August 31: 32.9° N - 80.0° W. The great Charleston, South Carolina, earthquake was felt over the entire eastern U.S. Its maximum intensity in the epicentral area was X, but in eastern Tennessee it was perhaps between VI and VII, as shown on Figure 2.5.2-3. 1886, September 1: A shock reported at Chattanooga was believed to be an aftershock of the Charleston earthquake, many of which were felt in Tennessee. 1892, December 2: A very perceptible earthquake shock was felt in Chattanooga from Hill City (now north Chattanooga) to Missionary Ridge. According to contemporary reports, the motion was from north to south. Doors in houses flew open, piles of lumber were upset, coal at chutes rolled down, and water vibrated. These effects were reportedly limited to an area of 6.25 square miles, but a larger area probably was affected. S2-5.doc 2.5-16

SQN 1895, October 31: 37.0° N - 89.4° W. A strong earthquake centered at Charleston, Missouri, affected an area of 1,000,000 square miles in 23 states. It threw down chimneys and damaged buildings at various places in the Mississippi Valley, including Memphis, Tennessee. The earthquake was felt over the entire Tennessee Valley, but it was of low intensity in eastern Tennessee. 1897, May 31: 37.3° N - 80.7° W. A strong earthquake centered in Giles County, Virginia, was felt over an area of more than 250,000 square miles. It was felt throughout eastern Tennessee as far west as Tullahoma, but did not attain damaging intensities outside the epicentral area. 1902, May 29: A "strong shock" (intensity V) shook houses and awakened sleepers in Chattanooga. 1902, October 18: 35.0° N - 85.3° W. A moderate shock affected some 1,500 square miles in Georgia and Tennessee. It was felt from Dalton to Chattanooga. The maximum intensity was IV-V, but it is not known to have been felt as far to the northeast as the Sequoyah plant site. 1904, March 4: 35.7° N - 83.5° W. The epicenter of this earthquake was between Maryville and Sevierville, but the disturbance was felt along the mountain front over a distance of 90 to 100 miles. The shock affected an area of about 5,000 square miles, but the intensity was nowhere above V and over much of the felt area it was much lower. 1913, April 17: 35.3° N - 84.2° W. This moderately strong earthquake was felt over an area of about 3,500 square miles in eastern Tennessee, western North Carolina, northern Georgia, and northwestern South Carolina. The intensity was higher (V-VI) along the major axis of the affected area between Ducktown and Kiser. As shown by the map (Figure 2.5.2-4), the earthquake was not felt in the Sequoyah area, but it was felt some miles away. 1913, May 2: A light shock of several seconds duration was felt near Madisonville, Tennessee. This shock, intensity III, was centered nearly 50 miles from the plant site. 1914, January 23: 35.60 N - 84.50 W. A sharp local shock (V) was felt at Niota and Sweetwater, some 35 miles from the plant site. 1916, February 21: 35.50 N - 82.50 W. The strong earthquake, intensity VII, was centered in the mountains of western North Carolina. It affected an area of 500,000 square miles in the Carolinas, Georgia, Tennessee, Alabama, Kentucky, and Virginia. It was felt over nearly all of Tennessee, but was most severe in the mountains of eastern Tennessee. Chimneys were damaged at Sevierville and plaster was shaken from walls at Bristol, Morristown, and Knoxville. At Memphis, there was considerable motion in the higher stories of buildings. The earthquake affected the Sequoyah area at intensities between III and IV (Figure 2.5.2-5). 1916, October 18: 33.50 N - 86.20 W. A strong earthquake centered near Easonville, Alabama, was felt over an area of 100,000 square miles in a seven-state area. About two-thirds of Tennessee was affected by this earthquake, but there was no damage in the state. The S2-5.doc 2.5-17

SQN disturbance was felt strongly at Chattanooga, Nashville, Waynesboro, Carthage, Sparta, McMinville, Lewisburg, and other points in central Tennessee. A light shock was noticed in Knoxville and Clinton. At the Sequoyah plant site, the intensity was not more than IV (Figure 2.5.2-6). 1918, June 21: 36.10 N - 84.10 W. Centered near Lenoir City, this moderate shock (IV-V) affected an area of 3000 square miles. It is not known to have affected the Sequoyah area. 1920, December 24: 36.00 N - 85.00 W. A moderately strong shock was felt at a number of localities in eastern Tennessee including Rockwood, Glen Alice, Spring City, Harriman, Decatur, and Crossville. Many sleepers were awakened and the entire village of Glen Alice was aroused. This earthquake, with a maximum intensity of V, was centered about 45 miles from the Sequoyah plant site and is not known to have affected the site area. 1921, December 15: An earthquake of "considerable intensity" was felt along the western portion of the Appalachian Valley from Kingston and Rockwood to Decatur and Dayton and as far eastward as Athens. The maximum intensity was V, but the shock is not known to have been felt any nearer to Sequoyah than Dayton. 1924, October 20: 35.0° N - 82.6° W. A strong earthquake (V-VI) centered in Pickens County, South Carolina, was felt over 56,000 square miles in the Carolinas, Georgia, Tennessee, Virginia, and Florida. Although buildings were strongly shaken in the epicentral area, there was little damage. The intensity in eastern Tennessee was nowhere greater than III. At the Sequoyah plant site, the intensity was less than II (Figure 2.5.2-7). 1927, October 8: A moderately strong earthquake was felt in all parts of Chattanooga and suburban areas, including north Chattanooga, East Ridge, Lookout Mountain, Signal Mountain, St. Elmo, and Red Bank. The shock was felt in small and large buildings. Lights trembled and loose objects were disturbed. Other mild shocks were reported within a few hours following this shock. The shock is not known to have been felt in the Sequoyah area. 1928, November 2: 35.8° N - 82.8° W. A strong earthquake centered in the mountains of Madison County, North Carolina, was felt over an area of 40,000 square miles in a six-State area. The maximum intensity was VII, but in Tennessee the intensity diminished from VI along the state line to extinction somewhere in central Tennessee. At the Sequoyah plant site, the intensity was less than III (Figure 2.5.2-8). 1930, August 30: 35.9° N - 84.4° W. This earthquake was felt at Kingston, Lenoir City, Lawnville, Oliver Springs, and other points west and southwest of Knoxville. The maximum intensity was V. This shock is not known to have affected the Sequoyah site area perceptibly. 1938, March 31: An earthquake centered in the mountains in the Little Tennessee Basin was widely felt in Tennessee and North Carolina. In Tennessee it was felt at Copperhill, Parksville, Knoxville, and Sweetwater where the intensities ranged from III to I. The shock is not known to have affected any part of Tennessee west of Sweetwater. S2-5.doc 2.5-18

SQN 1940, October 19: An earthquake which shook houses and rattled loose objects awoke thousands of sleepers in Chattanooga. It affected some 1,100 square miles in Tennessee and Georgia. It was felt as far north as Charleston and Birchwood but at very low intensities (Figure 2.5.2-9). 1941, September 8: An earthquake was felt throughout Chattanooga and as far west as Jasper. It was especially strong in the Lookout Mountain area where walls vibrated, loose objects rattled, and glassware was broken. This earthquake is not known to have been felt upstream from Chattanooga. 1945, June 13: This shock, centered near Cleveland, Tennessee, where the intensity was V, was felt over an area of 4,000 square miles in southeastern Tennessee and northwestern Georgia. It was felt north-eastward to Knoxville, southwestward to Chattanooga, and southeastward to Blue Ridge, Georgia. The felt area of this shock was never mapped, but the shock may have affected the Sequoyah area at an intensity of III or less. 1946, April 6: Another light shock was felt at Cleveland, Tennessee. This shock was not reported felt outside of the city. 1947, December 27: A light earthquake (IV) felt in Chattanooga, Tennessee; and Fort Oglethorpe, Rossville, Ringgold, and Boynton, Georgia, affected an area of 300 miles. It was centered east of the Missionary Ridge fault, where houses shook, loose objects rattled and piano wires popped. The shock is not known to have been felt any nearer to Sequoyah than Chattanooga. 1954, January 22: A light earthquake was felt over much of McMinn County from Athens to Etowah and Englewood. It is not known to have been felt outside of the county. 1957, June 23: 35° 54' N - 84° 14' W. A light local earthquake was felt in western Knox County and nearby sections of Anderson and Loudon Counties. At Dixie Lee Junction and in neighboring communities, people were awakened by the "jumping" of houses and the rattling of loose objects. 1959, June 12: 35° 21' N - 84° 20' W. A light earthquake was felt over an area of 900 square miles in eastern Tennessee and western North Carolina. It was most strongly felt at Tellico Plains and Mount Vernon where an intensity of IV was attained. 1960, April 15: 35.8° N - 83.9° W. A shock of intensity V, centered near Knoxville, Tennessee, was felt over a 1,300 square mile area. It was not reported as felt in the Sequoyah area. 1966, August 24: 35.9° N - 83.9° W. This shock of intensity IV, centered near Knoxville, Tennessee, was not felt in the Sequoyah area. 1968, November 9: 38.0° N - 88.5° W. This earthquake, centered in southern Illinois, with an epicentral intensity of VII was felt over a 400,000 square mile area in 23 states, including Tennessee, and in Canada. In the Sequoyah area it had an approximate intensity between II and III. S2-5.doc 2.5-19

SQN 1969, July 13: 36.1° N - 83.7° W. The epicenter of this intensity IV shock was located northeast of Knoxville, Tennessee. This shock was not felt in the Sequoyah area (Figure 2.5.2-10). 1969, November 20: 37.4° N - 81.0° W. This intensity V shock, with its epicenter in southern West Virginia, was not reported felt in the Sequoyah area. 1971, July 12: 35.9° N - 84.3°. A light local tremor (MM III-IV) was felt at 10:00 p.m. in the Knoxville-Oak Ridge area. It was not felt in the Sequoyah area. A list of all seismic events to 1982 and within a 200-miles radius of the plant site is presented as Table 2.5.2-1. The seismic history of the southeastern U.S. has been known for only about a century and a half, but so far as can be determined from the records the Sequoyah site is as stable seismically as any area in the State. Great distant earthquakes have affected the area with intensities equal to or greater than the maximum intensities of the several shocks centered within 50 or 60 miles of the site. Of the 40 earthquakes identified in the foregoing annotated list, only 12 are positively known to have been felt at Sequoyah. Of these, four were centered in the Mississippi Valley, one at Charleston, South Carolina, one in Alabama, one in Illinois, and five at various centers in east Tennessee, Virginia, and western North Carolina. In addition to these, it is probable that a few other shocks might have affected the area at very low intensities. On Figure 2.5.2-1, epicenters of all historic quakes within 120 miles of the Sequoyah site and all epicenters of historic quakes with MM intensities of V or greater up to and beyond 250 miles from the site are plotted. 2.5.2.4 Site Seismic Evaluation The known seismic history of the southeastern United States suggests that the earthquake hazard is negligible at the Sequoyah site. There are no active faults in the vicinity of the site and there is no physical evidence of any seismic activity at the site. There have been several shocks in the general area including two shocks of intensity MM V centered within 15 and 20 miles of the site. However, the nearest known epicenter of damaging intensity (MM VII) is 100 miles northeast of the site. The maximum intensity to have been felt at the site in the recorded history of the area is probably MM V and certainly no more than MM VI. On the basis of present knowledge, the maximum historic felt intensity was derived from major earthquakes centered at distant points, especially in the Mississippi Valley. There is continuing seismic activity in the Mississippi Valley and the possibility of another great earthquake in the New Madrid region cannot be discounted. An earthquake of intensity MM X to MM XII at New Madrid might be felt at Sequoyah with an intensity of MM V or MM VI. There is no known correlation between earthquakes observed in the region and any surficial tectonic structures. The site lies in the Southern Appalachian tectonic province as defined during the preliminary evaluation of the Sequoyah Nuclear Plant site. This province is bounded on the east by the western edge of the Piedmont Province; on the west by the western limits of the Cumberland Plateau; on the south by the overlap of the Gulf Coastal Plain Province; and on The north by the re-entrant in the Valley and Ridge Province near Roanoke, Virginia (Figure 2.5.2-1). S2-5.doc 2.5-20

SQN The maximum historic quake reported in this province was assigned an intensity of MM VIII although there is reason to believe it should have been rated as MM VII. It occurred in Giles County, Virginia, in 1897. Although this earthquake occurred 285 miles northeast of the site, this intensity is assumed to occur at the site for the purpose of defining the Safe Shutdown Earthquake (SSE). The maximum acceleration for an intensity of this level is estimated to be 0.14 g. This peak acceleration has been estimated from empirical relationships which are based almost exclusively on data obtained on overburden and hence provide some margin of conservatism for a rock site (seismic site studies indicate a shear wave velocity of 7,000 ft/s). Initially, it was felt the Housner spectrum for maximum top of rock acceleration of 0.14 g for the SSE best represented the historic seismic threat at the site, i.e., large shocks at long distances. This information was submitted to TVA's consultant (Weston Geophysical Research, Incorporated) for their review. TVA's consultant agreed that the maximum ground acceleration values were conservative but felt the Housner spectra did not give sufficient weight to the effect of close earthquakes. TVA's consultant recommended a spectrum reflecting more energy in the 5 to 10 Hz frequency range, and his recommendations were accepted by TVA. Another consultant was contracted to produce such a spectrum and a set of four artificial earthquake records whose average response would approximate this spectrum. During the course of the Sequoyah PSAR review, a special meeting was called on November 13, 1969 to discuss earthquake design criteria. AEC structural and geological-seismological consultants for Sequoyah were present. At this meeting, AEC's geological-seismological consultants took the position that maximum top of rock accelerations should be 0.18 g for the SSE. AEC's structural consultants stated that 0.18 g coupled with a Housner spectrum would be considered satisfactory as a minimum design basis. TVA stated that it would use the arithmetically averaged response spectra generated by four artificial records previously mentioned after the high frequency end had been raised to coincide with the 0.18 g Housner spectra. The structural consultants agreed that if TVA wished to use these records, which give more conservative results, this would certainly be acceptable to them. Accordingly, the plant is designed so that all structures, systems, and components important to safety will remain functional when subjected to an SSE having maximum horizontal acceleration of 0.18 g and maximum vertical ground acceleration of 0.12 g. 10 CFR Part 100, Appendix A, 1971, allowed the utilities to independently select the g-level for the Operating Basis Earthquake (OBE). Accordingly, TVA selected 0.00g as the OBE. The regulations required, however, the establishment of a "1/2 SSE" which was based on a g-level of 1/2 of the SSE. The 1/2 SSE for Sequoyah was therefore 0.09g (i.e., 1/2 of the 0.18g maximum horizontal ground acceleration). The seismic design basis for Sequoyah Nuclear Plant is the 0.18 g modified Housner spectrum discussed above. However, in the course of their review for the operating license, NRC requested additional information concerning the seismic design basis. This culminated in the development of a site specific response spectrum. This spectrum represents the 84th percentile of 13 actual earthquake recordings and has a peak acceleration of 0.22 g. This site specific spectrum was used for evaluation of present designs and not as a design basis. The development of the site specific spectrum is presented in the following reports. S2-5.doc 2.5-21

SQN

1. Justification of the Seismic Design Criteria Used for the Sequoyah, Watts Bar, and Bellefonte Nuclear Plants - Phase I, TVA, April 1978.

2. Justification of the Seismic Design Criteria Used for the Sequoyah, Watts Bar, and Bellefonte Nuclear Plants - Phase II, TVA, August 1978.

3. Prediction of strong motions for Eastern North America on the Basis of Magnitude, Weston Geophysical Report for TVA, August 1978.

4. Earthquake Ground Motion Study in the Vicinity of the Sequoyah Nuclear Power Plant, Weston Geophysical Report for TVA, February 1979.

5. Justification of the Seismic Design Criteria Used for the Sequoyah, Watts Bar, and Bellefonte Nuclear Plants - Phase II - Responses to NRC Questions 1 to 6, TVA, June 1979.

Therefore, as a result of the development of the site specific response spectrum in 1979, an SSE of 0.22g has been considered. 10 CFR Part 100, Appendix A, 1973, regulations no longer require a 1/2 SSE; however, applicants are required to select an OBE equal to at least 1/2 of the SSE unless supporting data are presented to clearly justify otherwise. TVA presented such data (reports 2 and 5, above) and justified an OBE of 0.09g, less than 1/2 of the present site specific SSE of 0.22g and the same as the 1/2 SSE used in early seismic analyses. Figures 2.5.2-11 through 2.5.2-14 illustrate the relationship between the minimum design response spectra and the actual site seismic design response spectra for the SSE for all damping ratios used in the design of rock-supported structures. 2.5.3 Surface Faulting The lithologic, stratigraphic, and structural conditions at the site and in the surrounding area and the geologic history of the region have been discussed previously in Paragraphs 2.5.1.3, 2.5.1.4, and 2.5.1.5, and will not be repeated here. 2.5.4 Stability of Surface Materials 2.5.4.1 Subsidence Most major Category I structures are founded on bedrock and no subsidence is to be expected. In most instances the weight of rock removed in foundation excavation equals or exceeds the weight imposed by the structure. Sufficient exploratory drilling has been done to assure there are no karstic solution zones underlying the plant that would allow collapse. Any small solution areas below foundation grade have been grouted in the routine course of construction. No mining or extensive groundwater withdrawal, either of which might allow subsidence, occurs in the area. Loads imposed by the plant structures are not of sufficient magnitude to develop compaction 2 subsidence in material having compressive strengths ranging from 5,000 to 15,000 lb/in . No regional warping is known in the southern Appalachian area of sufficient magnitude to impose unequal stresses on the plant structures. S2-5.doc 2.5-22

SQN 2.5.4.2 Zone of Deformed or Weak Material Sufficient exploration was done prior to final location of the individual structures to insure that weak or deformed zones are not present in the foundation areas. Any minor defects that were disclosed during excavation were treated appropriately as a standard construction procedure. 2.5.4.3 Bedrock Stresses No specific investigations of residual stress accumulations in the foundation strata were made. Experience at numerous previous major construction projects in the region has shown that this is not a consideration. Such stress effects as "popping," rock bursts, and foundation "heaving" were not observed during foundation evacuation. 2.5.5 Stability of Subsurface Materials 2.5.5.1 Excavations and Backfill Excavations and backfill are described in Paragraph 2.5.1.11. 2.5.5.2 Liquefaction Potential The liquefaction potential of all slopes and soil deposits were evaluated by using empirical rules based on observed performance and by comparing the soil conditions and earthquake characteristics at the site with similar sites that have liquefied. The empirical rules used are based on the Japanese experience during the Niigata earthquake. It was observed that the following general conditions could cause liquefaction:

1. The percentage of silt and clay-size particles should be less than 10 percent.

2. The particle diameter at 60 percent passing should be between 0.2 mm and 1.0 mm.

3. The uniformity coefficient should be between 2 and 5.

4. The blow count from Standard Penetration Tests should be less than 15.

Using these rules there were no soils which indicated potential liquefaction. A comparison of the soil conditions and the earthquake characteristics at the site with similar sites that have liquefied indicated that there were no potentially liquefiable soils at the site. 2.5.5.3 Static Analysis 2.5.5.3.1 Settlement Analysis Soil supported Category I structures were investigated to determine the amount of settlement each would undergo. Settlement calculations were made for the Diesel Generator Building and the Low Level Radwaste Storage Facility. S2-5.doc 2.5-23

SQN Diesel Generator Building The Diesel Generator Building (DGB) had a net increase in load on the soil. The settlement calculations contain several conservative assumptions which make the estimated value of settlement an upper bound. As a result of these conservative assumptions, the settlement actually experienced is less than estimated. A time-settlement rate was not determined for the original calculations, as we were committed to waiting for settlement to stabilize. We determined that settlement had stabilized sufficiently in the first two years (see Figure 2.5.5-1). Low Level Radwaste Storage Facility The Low Level Radwaste Storage (LLRW) Facility is located in an area that underwent significant changes during the construction of the plant. Initially, the area served as a borrow source, and material was excavated to approximately the final grade for the LLRW facility. The area was then used for a yard storage area and later as a storage area for spoil material. Prior to its use for the LLRW facility, the spoil material and some additional in situ material were removed to reach final grade. The maximum net increase in soil pressure due to the LLRW facility above the original overburden load was 0.32 tons/ft2. The resultant theoretical settlement due to the imposed load was less than the allowable settlement. A settlement monitoring program for the LLRW facility has been established and is described in section 2.5.5.3.2. 2.5.5.3.2 Settlement Monitoring Settlement monitoring programs were developed for the Diesel Generator Building, the East Steam Valve Rooms, the Low Level Radwaste Storage Facility, and the ERCW Support Slab and Pumping Station. Settlement programs were not developed for the Waste Packing Area and the Condensate Demineralizer Waste Evaporator Building. The details of each program or the reasons for not developing a settlement program are given below. Diesel Generator Building - This soil supported structure was monitored for settlement. It has a uniform bearing pressure of 1400 lb/ft2. Settlement monuments were placed at each corner of the structure. Readings were started in January 1973 and read monthly until January 1974 and then quarterly until January 1975. No readings were then made until April 1979. Based on available data and our past experience, there are no adverse trends being exhibited; settlements are not significant; and there has been no adverse structural performance. Settlement readings will no longer be reported for this structure. The construction period for the DGB extended from June 1972 to September 1973. The base slab and the first lift of the exterior walls were constructed before the settlement markers were placed and the first settlement readings were taken. The electrical conduit connections were made between November 1974 and January 1975. The piping connections were made after July 1978. East Steam Valve Room - This structure was originally supported on soil but due to excessive settlement was underpinned with caissons. The caissons were completed between February and S2-5.doc 2.5-24

SQN August 1976. Negligible settlement has occurred since the caissons were installed. Because of this excellent performance, a continued settlement monitoring program is not warranted. The electrical conduit connections were made between May 1978 and the present. The piping connections were made between September 1977 and October 1978. All of these were installed after the caissons were in place. ERCW Support Slab and Pump Station - The ERCW support slab is supported on piles driven to rock. The ERCW pumping station is supported on rock. A settlement monitoring program was developed for both of these features. The survey markers were read monthly from June 1979 to March 1980, semiannually from March 1980 to September 1981, and annually from September 1981 to September 1984. Negligible settlement was found during the monitoring program. The settlement monitoring program was discontinued in September 1984 after 5 years of monitoring. Waste Packaging Area and Condensate Demineralizer Waste Evaporator Building - These structures are supported on piles driven to rock. No settlement monitoring program was developed for these structures. Since the piles are driven to rock, there is no need to monitor settlement. The supporting piles were driven to rock before placement of the foundation mat. For the Waste Packaging Area, the piles were completed in October 1975, and the electrical conduit connections were made between January 1977 and December 1978. There is no Category I piping for this building. For the Condensate Demineralizer Waste Evaporator Building, the piles were completed in June 1977. The piping connections were made in August 1978. Low Level Radwaste Storage Facility Each storage module has four individual compartments with each compartment being composed of five unit cells. The storage modules are designed for a total settlement of 9 inches, a differential settlement of 4 inches over an individual storage compartment, and a differential settlement of 4 inches between individual compartments. Settlement monitoring points are established on each corner of each compartment of each module and settlements are recorded annually until settlement has essentially ceased. 2.5.6 Slope Stability 2.5.6.1 Slope Characteristics 2.5.6.1.1 Slopes at Diesel Generator Building and Cooling Towers The Diesel Generator Building and Cooling Towers are located on a gently sloping hillside southeast of the main plant area. A cross section of the hillside is shown in Figure 2.5.6-1. The soil properties are obtained as described in Paragraph 2.5.1.9. S2-5.doc 2.5-25

SQN The R-test strengths of the soil are used in the seismic pseudo-static stability analyses. The soil properties used in the seismic pseudostatic stability analyses are shown in Figure 2.5.6-1. 2.5.6.1.2 Condenser Cooling Water Pumping Station Intake Channel Slopes The intake channel shown in Figure 2.1.2-1 is located on the north side of the main plant area. The side slopes of both the approach channel and the forebay area are cut on a 3.5 horizontal to 1 vertical slope. Typical cross sections of the approach channel and forebay slopes are shown in Figure 2.4.8-1. The side slopes in the forebay area are Category I slopes and are constructed to remain stable for the most critical design conditions. Enough water is retained in the forebay for plant shutdown using a closed mode of operation and therefore the approach channel slopes are not designed as Category I slopes. The soil properties used in the seismic pseudostatic stability analysis of the side slopes are shown in Figure 2.5.6-2. See paragraph 2.5.1.9 for additional information on the soil properties. 2.5.6.1.3 Dike Slopes at the ERCW Pumping Station The dike leading to the ERCW pumping station on Chickamauga Reservoir shown in Figure 2.1.2-1 is located northeast of the main plant across the embayment from the condenser cooling water supply pumping station. The dike has Category I slopes and is designed to remain stable for the most critical design conditions. 2.5.6.2 Design Criteria and Analysis 2.5.6.2.1 Design Criteria and Analysis of Slopes at Diesel Generator Building and Cooling Towers The seismic stability analysis of the hillside is performed assuming circular failure arcs using the Modified Swedish Method with Slices and a Newmark analysis. Horizontal and vertical seismic accelerations are used in the analyses. The accelerations for the Safe Shutdown Earthquake in the soil deposit and on these soil-supported structures are obtained as discussed in Paragraphs 3.7.1.6 and 3.7.2.1. The worst location for failure is a section which includes the Diesel Generator Building since it is the heaviest structure and has the largest seismic forces acting on it. The water table in the soil deposit is conservative assumed at elevation 705.0. The factor of safety during a Safe Shutdown Earthquake must be greater than 1.0. Several circular failure arcs are considered to determine the location of the critical arc. The critical failure arc is shown in Figure 2.5.6-1. A Newmark analysis is performed for this critical failure arc. The Newmark analysis shows that the Design Basis Earthquake will not induce sliding along this failure arc. From these analyses it is concluded that the hillside will be stable during a Safe Shutdown Earthquake. S2-5.doc 2.5-26

SQN 2.5.6.2.2 Design Criteria and Analysis of (Condenser Cooling Water Pumping Station) Intake Channel Slopes The side slopes of the forebay portion of the intake channel are designed and constructed such that they remain stable for the most critical design condition, the occurrence of a Safe Shutdown Earthquake coincident with a sudden drawdown of the reservoir water level. The stability analyses of the slopes were performed assuming circular failure planes using the Modified Swedish Method with Slices. Horizontal and vertical seismic coefficients were used in the analyses. The accelerations for the Safe Shutdown Earthquake in the soil deposit were obtained as discussed in Paragraph 3.7.1.6. Several circular failure planes were considered and the minimum factor of safety was found to be 1.31. This failure plane is shown in Figure 2.5.6-2. In addition a level ledge with a 15-foot-minimum width extends from the toe of the slide slopes to the edge of the forebay. This precludes the spillage of material into the forebay from a localized slippage of the slope. 2.5.6.2.3 Design Criteria and Analyses of Dike Slopes at the ERCW Pumping Station The Category I slopes of the dike leading to the ERCW pumping station are designed such that they remain stable for the most critical design condition; the occurrence of a Safe Shutdown Earthquake coincident with normal reservoir level. The dike is also designed to remain stable during the PMF and subsequent drawdown. The stability analysis of the slopes were performed using wedge analysis techniques. Pseudo-static analyses were used in all the seismic evaluations. Horizontal seismic coefficients were used in these analyses. The accelerations in the dike from the Safe Shutdown Earthquake were obtained as discussed in paragraph 3.7.1.6. The minimum factor of safety was determined to be 1.22. Calculations were also performed to approximate the deformations which might be expected to occur as a result of stresses caused by a seismic event. This calculation considered the effect of vertical acceleration. The resulting deformations were shown to have no significant effect on the buried ERCW pipes. 2.5.6.3 Compaction Specifications See Paragraph 2.5.1.11. 2.5.7 References

1. Bollinger, G. A., Harding, S. T., Langer, C. J., 1974, Alcoa Maryville Tennessee, Earthquake Swarm of November-December, 1973, Geological Society of America, Abstracts with Programs, Vol. 6, No. 4, p. 336.

S2-5.doc 2.5-27

SQN

2. Bridge, Josiah, 1950, Bauxite Deposits of the Southeastern United States, 1949 Symposium on Mineral Resources of the Southeastern United States, University of Tennessee Press, Knoxville, Tennessee, pp. 170-201.

3. Butts, Charles, and Gildersleeve, Benjamin, 1948, Geology and Mineral Resources of the Paleozoic Area in Northwest Georgia, Georgia Department of Mines, Mining, and Geology, Bulletin 54, p. 176.

4. Campbell, M. R., 1899, Bristol Quadrangle, USGS Geologic Atlas, Folio 59.

5. Colton, G. W., 1970, The Appalachian Basin - Its Depositional Sequences and Their Geologic Relationships, Studies of Appalachian Geology, Central and Southern, John Wiley and Sons, pp. 5-47.

6. Cooper, B. M., 1961, Grand Appalachian Field Excursion, Geologic Society of America, 74th Annual Meeting, Guidebook Trip 1, p. 187.

7. 1964, Relation of Stratigraphy to Structure in the Southern Appalachians, in Tectonics of the Southern Appalachians, VPI Memoir 1, pp. 81-114.

8. Dietz, R. S., 1972, Geosynclines, Mountains, and Continent - Building, Scientific American, Vol. 226, No. 3, pp. 30-38.

9. Fenneman, N. M., 1938, Physiography of the Eastern United States, McGraw-Hill Company, New York, New York.

10. Ferguson, H. W. and Jewell, W. B., 1951, Geology and Barite Deposits of the DelRio District, Cocke County, Tennessee, Tennessee Division of Geology, Bulletin 57, p. 235.

11. Grant, L. F., Concept of Curtain Grouting Evaluation, Journal of Soil Mechanics Division of the ASCE, Vol. 90, SM 1, (1964).

12. Gwinn, V.E., 1964, Thin-skinned Tectonics in the Plateau and Northwestern Valley and Ridge Provinces of the Central Appalachians, Geological Society of America Bulletin, Vol. 75, pp. 863-899.

13. Hack, J. T., 1965, Geomorphology of the Shenandoah Valley, Virginia and West Virginia, USGS Professional Paper 484, p. 84.

14. 1966, Interpretation of Cumberland Escarpment and Highland Rim, South Central Tennessee and Northeastern Alabama, USGS Professional Paper 524-C, p. 16.

15. Hardeman, W. D., et al, 1966, Geologic Map of Tennessee, Tennessee Division of Geology, Nashville, Tennessee, four sheets.

16. Harris, L. D., 1965, The Clinchport Thrust Fault - A Major Structural Element of the Southern Appalachian Mountains, USGS Professional Paper 525-3, pp. 49-55.

S2-5.doc 2.5-28

SQN

17. 1970, Details of Thin-Skinned Tectonics in Parts of Valley and Ridge and Cumberland Plateau Provinces of the Southern Appalachians, Studies of Appalachian Geology, Central and Southern, John Wiley & Sons, pp. 161-173.

18. Hayes, C. W., 1891, The Overthrust Faults of the Southern Appalachians, Geological Society of America Bulletin, Vol. 2, pp. 141-154.

19. 1892, Report on the Geology of North-Eastern Alabama, and Adjacent Portions of Georgia and Tennessee, Geological Survey of Alabama, Bulletin 4, p. 85.

20. 1894, Ringgold Quadrangle, USGS Geologic Atlas, Folio No. 2.

21. 1894a, Kingston Quadrangle, USGS Geologic Atlas, Folio No. 4.

22. 1894b, Chattanooga Quadrangle, USGS Geologic Atlas,Folio No. 6.

23. 1895, Stevenson Quadrangle, USGS Geologic Atlas, Folio No. 19.

24. 1895a, Cleveland Quadrangle, USGS Geologic Atlas, Folio No. 20.

25. 1895b, Pikeville Quadrangle, USGS Geologic Atlas, Folio No. 21.

26. 1895c, The Southern Appalachians, National Geographic Society Monograph, Vol. 1, No. 10, pp. 305-336.

27. 1899, Physiography of the Chattanooga District, in Tennessee, Georgia, and Alabama, USGS 19th Annual Report, Pt. 2, pp. 1-58.

28. Hayes, C. W. and Campbell, M. R., 1894, Geomorphology of The Southern Appalachians, National Geographic Magazine, Vol. 6, pp. 63-126.

29. Keith, Arthur, 1895, Knoxville Quadrangle , USGS Geologic Atlas, Folio No. 15.

30. 1896, Some Stages of Appalachian Erosion, Geological Society of American Bulletin, Vol. 7, pp. 519-525.

31. 1896a, Loudon Quadrangle, USGS Geologic Atlas, Folio No. 25.

32. 1896b, Morristown Quadrangle, USGS Geologic Atlas, Folio No. 27.

33. 1896c, Briceville Quadrangle, USGS Geologic Atlas, Folio No. 33.

34. 1901, Maynardville Quadrangle, USGS Geologic Atlas, Folio No. 75.

35. 1902, Topography and Geology of the Southern Appalachians, U.S., 57th Congress, Senate Document No. 84, App. B, pp. 111-122.

36. 1902a, Folded Faults in the Southern Appalachians, Science Vol. 15, pp. 822-823.

S2-5.doc 2.5-29

SQN

37. 1903, Cranberry Quadrangle, USGS Geologic Atlas, Folio No. 90.

38. 1904, Asheville Quadrangle, USGS Geologic Atlas, Folio No. 116.

39. 1905, Greeneville Quadrangle, USGS Geologic Atlas, Folio No. 118.

40. 1905a, Mt. Mitchell Quadrangle, USGS Geologic Atlas, Folio No. 124.

41. 1907, Nantahala Quadrangle, USGS Geologic Atlas, Folio No. 143.

42. 1907a, Roan Mountain Quadrangle, USGS Geologic Atlas, Folio No. 151.

43. 1923, Outlines of Appalachian Structure, Geological Society of America Bulletin, Vol. 34, pp.

304-380.

44. 1927, Great Smoky Overthrust, Geological Society of America Bulletin, Vol. 38, pp. 154- 155.

45. Kelleberg, J. M., and Benziger, C. P., 1953, Preliminary Geologic Investigations for Steam Plant H - Soddy Daisy Site, TVA Report.

46. King, P. B., 1949, The Floor of the Shenandoah Valley, American Journal of Science, Vol. 247, pp. 73-93.

47. 1950, Tectonic Framework of the Southeastern United States, American Association of Petroleum Geologists Bulletin, Vol. 34, pp. 635-671.

48. 1955, A Geologic Section Across the Southern Appalachians, Geological Society of America, 68th Annual Meetings, Guidebook, pp. 332-373.

49. 964, Further Thoughts on Tectonic Framework of Southeastern United States, in Tetonics of the Southern Appalachians, VPI Memoir 1, pp. 5-31.

50. 1964a, Geology of the Great Smoky Mountains, Tennessee, USGS Professional Paper 349-c, p.148

51. 1969, The Tectonics of North America, USGS Professional Paper 628, p. 45.

52. King, P. B. and Ferguson, H. W., 1960, Geology of Northernmost Tennessee, USGS Professional Paper 311, p. 136.

53. King, Phillip B., et al, 1944, Geology and Manganese Deposits Northeastern Tennessee.

Tennessee Division of Geology, Bulletin 52, pp. 24-25, 59.

54. Milici, R. C., 1962, The Structural Geology of the Harriman Corner, Roane County, Tennessee, American Journal of Science, Vol. 260, pp. 787-793.

S2-5.doc 2.5-30

SQN

55. 1963, Low-Angle Overthrust Faulting, as Illustrated by the Cumberland Plateau-Sequatchie Valley Fault System, American Journal of Science, Vol. 261, pp. 815-825.

56. 1967, The Physiography of Sequatchie Valley and Adjacent Portions of the Cumberland Plateau, Tennessee Division of Geology Report of Investigations, No. 22.

57. 1968, Mesosoic and Cenozoic Physiographic Development of the Lower Tennessee River - in Terms of the Dynamic Equilibrium Concept, Journal of Geology, Vol. 76, pp. 472-479.

58. 1970, The Allegheny Structural Front in Tennessee and its Regional Tectonic Implications, American Journal of Science, Vol. 268, pp. 127-141.

59. Miller, R. L., 1962, The Pine Mountain Overthrust at the Northeast End of the Powell Valley Anticline, Virginia, USGS Professional Paper 450-D, pp. 69-72.

60. Morrill, B. J., Risavich, F., 1974, Preliminary Report of Maryville, Tennessee, Earthquake of November 30, 1973, Manuscript Report, USGS Seismic Engineering Unit, p. 7.

61. Neuman, R. B. and Nelson, W. H., 1965, Geology of the Western Great Smoky Mountains, Tennessee, USGS Professional Paper 349-D, p. 81.

62. Owens, J. P., 1970, Post-Triassic Tectonic Movements in the Central and Southern Appalachians as Recorded by Sediments of the Atlantic Coastal Plain, Studies of Appalachian Geology, Central and Southern, John Wiley and Sons, pp. 417-427.

63. Rodgers, John, 1948, Geology and Mineral Deposits of Bumpus Cove, Unicoi and Washington Counties, Tennessee. Tennessee Division of Geology, Bulletin 54, pp. 15, 40.

64. 1949, Evolution of Thought on Structure of Middle and Southern Appalachians, American Association of Petroleum Geologists Bulletin, Vol. 33, pp. 1643-1654.

65. 1950, Mechanics of Appalachian Folding as Illustrated by Sequatchie Anticline, Tennessee and Alabama, American Association of Petroleum Geologists Bulletin, Vol. 34, pp. 672-681.

66. 1953, Geologic Map of East Tennessee with Explanatory Text, Tennessee Division of Geology, Bulletin 58, Part II, p. 168.

67. 1953a, The Folds and Faults of the Appalachian Valley and Ridge Province, Kentucky Geological Survey, Series 9, Special Publication 1 pp. 150-166.

68. 1963, Mechanics of Appalachian Foreland Folding in Pennsylvania and West Virginia, American Association of Petroleum Geologists Bulletin, Vol. 47, pp. 1527-1536.

69. 1964, Basement and No-Basement Hypotheses in the Jura and the Appalachian Valley and Ridge, in Tectonics of the Southern Appalachians, VPI Memoir 1, pp. 71-80.

70. 1967, Chronology of Tectonic Movements in the Appalachian Region of Eastern North America, American Journal of Science, Vol. 265, pp. 408-427.

S2-5.doc 2.5-31

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71. 1970, The Tectonics of the Appalachians , Wiley- Interscience, New York, p. 271.

72. Safford, J. M., 1856, A Geological Reconnaissance of the State of Tennessee, First Biennial Report, Nashville, p. 164.

73. 1859, On Some Points in American Geological History, American Journal of Science, 2nd Series, Vol. 27, pp. 140-141.

74. 1869, Geology of Tennessee, Nashville, Tennessee.

75. Stearns, R. G., 1954, The Cumberland Plateau Overthrust and Geology of the Crab Orchard Mountains Area, Tennessee, Tennessee Division of Geology, Bulletin 60, p. 47.

76. 1955, Low-Angle Overthrusting in the Central Cumberland Plateau, Tennessee, Geological Society of America Bulletin, Vol. 66, pp. 615-628.

77. Swingle , G. D., 1961, Structural Geology Along the Eastern Cumberland Escarpment, Tennessee, Tennessee Division of Geology, Report of Investigations 13, p. 46.

78. Troost, Gerrard, 1837, Fourth Geologic Report to the Twenty-First General Assembly of the State of Tennessee, Nashville, Tennessee.

79. 1840, Fifth Geologic Report to the Twenty-First General Assembly of the State of Tennessee, Nashville, Tennessee.

80. 1841, Sixth Geologic Reports to the Twenty-First General Assembly of the State of Tennessee, Nashville, Tennessee.

81. Watkins, J. S., 1964, Regional Geologic Implications of the Gravity and Magnetic Fields of a Part of Eastern Tennessee and Southern Kentucky, USGS Professional Paper 516-A, p. 17.

82. Willis, Bailey, 1893, The Mechanics of Appalachian Structure, USGS 13th Annual Report, Part 2, pp. 211-281.

83. Wilson, C. W., Jr., and Stearns, R. G., 1958, Structure of the Cumberland Plateau, Tennessee, Geological Society of America Bulletin, Vol. 69, pp. 1283-1296.

84. CEB REPORT 97-01, Geotechnical and Soils Investigations for the Sequoyah Nuclear Plant, RIMS No. B39 970811 001.

S2-5.doc 2.5-32

SQN Table 2.5.1-1

SUMMARY

OF IN SITU UP-HOLE DYNAMIC TESTING REACTOR FOUNDATION AREA Vp Vp Compressional Shear Young's Density Velocity Velocity Vp Poisson's Modulus 6 Station Geophone Shot lbs/cu ft ft/sec ft/sec Vs Ratio psi, 10 Number Elevation Elevation Rock Type and Dip Calculated Measured Measured Ratio Calculated Calculated W26+84 677.2 627.2 Limestone with 12% 170 13,550 7,450 1.8 0.28 5.3 N70+58 shale, 60°-70° W27+50 672.9 629.9 Limestone with 20% 170 9,736 4,873 2.0 0.33 2.4 N69+90 shale, 45°-55° W27+50 676.9 635.9 Limestone, scattered 170 11,714 5,616 2.1 0.35 3.2 N70+58 shale partings, 50° W27+50 675.6 630.6 Limestone with 15% 170 11,842 7,258 1.6 0.18 4.8 N71+23 shale, 45°-50° W27+85 664.8 622.8 Limestone with 14% 170 8,400 -- -- -- -- N68+50 shale, 70°-85° W28+16 678.9 627.9 Limestone with 25% 170 12,500 7,083 1.8 0.28 4.5 N70+58 shale, 60°-80° W28+50 642.6 601.6 Limestone with 5% 170 15,185 -- -- -- -- N67+75 shale, 50°-70° W28+50 668.2 628.2 Limestone with 6% 170 10,444 5,437 1.9 0.31 2.8 N68+40 shale, 45°-65° W28+50 674.6 634.6 Limestone with 10% 170 12,903 6,557 2.0 0.31 5.8 N69+06 shale, 40°-60° W29+15 661.0 621.0 Limestone with 5% 170 13,333 6,993 1.9 0.31 4.7 N68+50 shale, 5°-90° Note: A valid shear velocity measurement could not be established for two stations. T251-1.doc

SQN Table 2.5.1-2 (Sheet 1)

SUMMARY

OF IN SITU CROSS-HOLE DYNAMIC TESTING REACTOR FOUNDATION AREA Sequoyah Nuclear Plant Vp Vp Compressional Shear Young's Between Density Velocity Velocity Vp Poisson's Modulus Geophone Shot Hole lbs/cu ft ft/sec ft/sec Vs Ratio psi, 106 Station Station Elevation Calculated Measured Measured Ratio Calculated Calculated Type Rock W26+84 W27+50 665 170 11,470 -- -- -- -- Limestone with inter-N70+58 N70+58 bedded shale W27+50 W27+50 665 170 18,649 -- -- -- -- Limestone, with inter-N69+90 N70+58 bedded shale W27+50 W27+50 665 170 18,659 9,697 1.9 0.31 9.3* Limestone with inter-N71+23 N70+50 bedded shale W27+85 W28+50 665 170 14,114 7,155 2.0 0.33 4.9 Limestone with inter-N68+50 N69+06 bedded shale W27+85 W28+50 665 170 12,286 -- -- -- -- Limestone with inter-N68+50 N67+75 bedded shale W28+16 W27+50 665 170 12,226 -- -- -- -- Limestone with inter-N70+58 N70+58 bedded shale W28+50 W27+85 665 170 11,799 -- -- -- -- Limestone with inter-N68+40 N68+50 bedded shale W28+50 W28+50 643 170 15,403 7,143 2.2 0.37 4.9 Limestone with inter-N68+40 N67+75 bedded shale

Note: Young's modulus value 9.3 x 106 is considered abnormally high for this type rock, and should be omitted when averaging. The average value is 4.4 x 106 psi as shown at the end of section 2.5.1.7.

T251-2.doc

SQN Table 2.5.1-2 (Sheet 2) (Continued)

SUMMARY

OF IN SITU CROSS-HOLE DYNAMIC TESTING REACTOR FOUNDATION AREA Sequoyah Nuclear Plant Vp Vp Compressional Shear Young's Between Density Velocity Velocity Vp Poisson's Modulus Geophone Shot Hole lbs/cu ft ft/sec ft/sec Vs Ratio psi, 106 Station Station Elevation Calculated Measured Measured Ratio Calculated Calculated Type Rock W28+50 W28+50 665 170 13,983 -- -- -- -- Limestone with inter-N68+40 N69+06 bedded shale W28+50 W29+15 661 170 14,255 6,700 2.1 0.35 4.7 Limestone with inter-N68+40 N68+50 bedded shale W28+50 W28+50 665 170 12,000 5,860 2.0 0.33 3.6 Limestone with inter-N69+06 N67+75 bedded shale W29+15 W27+85 665 170 13,436 -- -- -- -- Limestone with inter-N68+50 N68+50 bedded shale W29+15 W28+50 665 170 11,583 6,300 1.8 0.28 3.9 Limestone with inter-N68+50 N67+75 bedded shale Note: A valid shear velocity measurement could not be established for seven stations. T251-2.doc

SQN Table 2.5.1-3 EQUATION FOR DYNAMIC MODULUS OF ELASTICITY (Vp )2 (1 + ) (1 - 2) E = 144 g (1 - ) Where E = Dynamic modulus of elasticity (psi) Vp = Compressional wave velocity (ft/sec)

= Poisson's Ratio g = Gravitational constant of 32.2 ft/sec
  = Unit Weight (lbs/ft3)

EQUATION FOR POISSON'S RATIO 1 V 2p

          -1 2  V 2s 
=

V 2p 2 -1 V s Where

= Poisson's Ratio Vp = Compressional wave velocity (ft/sec)

Vs = Shear wave velocity (ft/sec) T251-3.doc

SQN Table 2.5.1-4

SUMMARY

OF GROUTING First Stage Grouting (holes drilled 10 feet into rock) (1) (2) (3) (4) (5) Holes with % Holes Unit Take Holes Drilled Take With Take Bags of Cement (Bags/Foot of Hole) Primary 333 38 11.4% 471 1.24 Secondary 71 11 15.1% 105 0.95 Third Series 16 1 6.3% 1 0.10 Total 420 50 577 Average --- --- 11.9% --- 1.15 Second Stage Grouting (holes drilled 45 feet into rock) (1) (2) (3) (4) (5) Holes with % Holes Unit Take Holes Drilled Take with Take Bags of Cement (Bags/Foot of Hole) Primary 220 51 23.2% 528 0.23 Secondary 93 35 37.6% 420 0.27 Third Series 109 49 44.9% 448 0.20 Fourth Series 63 21 33.3% 171 0.18 Fifth Series 44 12 27.2% 81 0.15 Total 529 168 1648 Average --- --- 31.8% --- 0.22 Total bags of cement injected. . . . . 2225 Total bags of cement-backfill. . . . . . 681 Total bags of cement-waste . . . . . . 643 Total bags of cement used . . . . . . 3549 T251-4

SQN TABLE 2.5.1-5 STATIC AND DYNAMIC ROCK-BEARING CAPACITIES FOR ROCK SUPPORTED CATEGORY I STRUCTURES(1) Static Bearing Dynamic Bearing Structure Allowable Allowable (lb/in2) (lb/in2) Shield 500 Adequate Auxiliary-Control 500 Adequate Additional Equipment 500 Adequate Intake Pump Station 500 Adequate Intake Pump Station 500 Adequate Retaining Wall ERCW Pump Station 500 1500 ERCW Pump Station 500 1500 Access Dike Cells (1) Base slab on rock. T251-05.doc

SQN TABLE 2.5.1-6 SOIL-BEARING CAPACITIES AND FACTORS OF SAFETY FOR SOIL SUPPORTED CATEGORY I STRUCTURES Sustained Loads Dynamic Loads Allowable Allowable Soil Factor Soil Bearing(1) of Bearing(2) lb/ft2 Safety lb/ft2 Diesel Generator 2,500 3,000 Building Refueling Water Storage 6,000 6,000 Tank Foundations

1. The factor of safety for the allowable soil bearing capacity for sustained loads is at least 3.0.

2. The factor of safety for the allowable soil bearing capacity for dynamic loads is at least 2.0.

T251-06.doc

SQN Table 2.5.1-7 Revised by Amendment 13 Standard

SUMMARY

OF EARTHFILL TEST DATA- DENSITY Compaction Project Sequoyah Nuclear Plant Feature Type "A" Backfill - Category I Period 1-19-71 To 5-1-78 Test No. _ _ _ _ To _ _ _ Prepared Fill Quantity: Period ____ yd 3 To Date _ _ __ by

                               -**. *      ~~~~ / ~:v /b---_,;..;.;;............-~rh...1.._s....-P..-e..;;.r_i_o....d~C,_um_......~""'c,.....tim..._-;r-.._,.--,......;;F-;;;.=.:::.::..,......_...-.i
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                                                                                -*--:                                                                                                               .I Totals
                                                                                                                                                           *.~:
                                                                                                                                                                                                       -TIU.$-
                                                                  -_ .... -_ -*                                                                                                     ?:rev.              -p.-i;,..;r.-...2. To Date I 98.8 96.4
                                                                                                                                                                                                                                 -2.4 iOO ___ - ._         94                     96.                           98                    100           .:- 102                 104 :-*-                  106         108               110-                  ll2   114 I
                                                                                                      . I                                                                                                            .*:.

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Percent Compact:io~ (Ydf tYdL)xlOO

SQN Revised by Amendment 13 Table 2.5.1-8

SUMMARY

OF EARTHFILL TEST DATA - MOISTURE CONTENT Standard Compaction Project Sequoyah Nuclear Plant Feature Type "A" Backfill Period 1-19-71 To 5-1-78 Test No. To Prepared Fill Quantity: Period yd3 To Date by

SQN Table 2.5.1-9 SEQUOYAH NUCLEAR PLANT

SUMMARY

OF IN-SITU SOIL DOWN-HOLE DYNAMIC TESTING DIESEL GENERATOR BUILDING Vp Vs Compressional Shear Zone Velocity Velocity Density Poisson's Modulus Modulus Depth ft/sec ft/sec lbs/cu ft Ratio psi, 103 psi, 103 Location Station Elevation Measured Measured Assumed Calculated Calculated Calculated Diesel 760E,129S 733.3-728.3 1471 631 100 0.39 8.6 23.8 Generator 728.3-728.3 2500 1,235 100 0.34 32.9 88.1 Building 708.3-673.3 6242 955 100 0.49 19.7 58.6 1.All holes were drilled by a truck-mounted auger. Note: 2.State 760E, 129S was not augered to refusal. SEQUOYAH NUCLEAR PLANT SEISMIC REFRACTION SURVEY IN-SITU ELASTIC PROPERTIES Vp Vs Compressional Shear Shear Velocity Velocity Density Poisson's Modulus Zones ft/sec ft/sec lbs/cu ft Ratio psi 103

Measured Calculated Assumed Assumed Calculated 1 1400 672 100 0.35 9.7 1400 571 100 0.4 7.0 1400 422 100 0.45 3.8 2 2900 1393 100 0.35 41.9 2900 1183 100 0.4 30.2 2900 874 100 0.45 16.5 3 7987 3836 100 0.35 317.5 7987 3260 100 0.4 229.3 7987 2408 100 0.45 125.0
For zone locations see Figure 2.5.1-10 Calculation Reference 841861022007 T251-09.doc

SQN Table 2.5.1-10 SEQUOYAH NUCLEAR PLANT ONSITE STORAGE FACILITY DYNAMIC SOIL TEST ARRAY SD-1 Summary of Cross-Hole Data (preferred arrival times) Poisson's Poisson's Elevation Vp Vp Vs Ratio Ratio (feet) Range Average Range Average Range Average 740 2880 - 3420 3060 1120 - 1160 1120 .40 - .44 .42 735 2820 - 3000 2910 920 - 1010 960 .43 - .45 .44 730 3680 - 4040 3910 780 - 900 850 .47 - .48 .48 725 3940 - 4360 4140 830 - 900 880 .47 - .48 .48 720 4000 - 4220 4140 880 - 1020 960 .46 - .48 .47 715 3660 - 4000 3870 810 - 1260 1090 .43 - .48 .46 710 N/A 3280 N/A 840 N/A .46 Summary of Downhole Data Elevation Vp Vs Poisson's (feet) fps fps Ratio 745.8- 2040 760 .42 736.0 736.0- 5240 760 .49 710.0 Calculation

Reference:

B41861022011 T251-10.doc

SQN Table 2.5.1-11 (Sheet 1) SEQUOYAH NUCLEAR PLANT ONSITE STORAGE FACILITY DYNAMIC SOIL TEST ARRAY SD-3 Crosshole Survey Average Average Compressional Shear Young's Shear Bulk Elevation Source and Velocity Velocity Poisson's Modulus Modulus Modulus Source and Receiver (ft/sec) (ft/sec) Ratio PSI x 104 PSI x 104 PSI x 105 Density Receiver Depth (measured) (measured) (calculated) (calculated) (calculated) (calculated) (lb.ft3) 736 5 1806 843 .36 4.71 1.73 0.56 113 731 10 2314 847 .42 4.97 1.75 1.07 113 726 15 2866 803 .46 4.58 1.57 1.79 113 721 20 3202 790 .47 4.46 1.52 2.30 113 716 25 3390 758 .47 4.13 1.40 2.61 113 711 30 3719 733 .48 3.88 1.31 3.20 113 706 35 3545 842 .47 5.08 1.73 2.83 113 701 40 3486 772 .47 4.28 1.45 2.77 113 696 45 3545 785 .47 4.43 1.50 2.86 113 691 50 3947 834 .48 5.01 1.70 3.57 113 686 55 3110 944 .45 6.29 2.17 2.07 113 681 60 3885 1008 .46 7.25 2.48 3.35 113 676 65 4065 1069 .46 8.15 2.48 3.66 113 672 69 1181 671 70 4065 666 75 4950 661 80 4950 656 85 4950 652 89 5657 Note:

1. Shear Wave velocities could not be obtained below elevation due to the difference in borehole depths.

2. The average compressional and shear wave velocities are calculated by averaging the measured velocities for the 25.4-, 19.8- and 14.6-foot distances.

3. The average compressional wave velocities below elevation 691 are calculated by averaging the measured velocities for the 19.8- and 14.6-foot distances. Hole C was blocked below this elevation, therefore no data could be obtained.

4. The density is a representative value determined from laboratory testing of soil samples taken near the array.

Calculation Reference CEB810515025 T251-11.doc

SQN Table 2.5.1-11 (Sheet 2) SEQUOYAH NUCLEAR PLANT ONSITE STORAGE FACILITY DYNAMIC SOIL TEST ARRAY SD-3 Downhole Survey Compressional Shear Young's Shear Bulk Travel Velocity Velocity Poisson's Modulus Modulus Modulus Elevation Path (ft/sec) (ft/sec) Ratio PSI x 104 PSI x 104 PSI x 105 Density Receiver Distance (measured) (measured) (calculated) (calculated) (calculated) (calculated) (lb.ft3) 736 11.1 1850 792 .39 4.24 1.53 0.63 113 731 14.1 2014 783 .41 4.22 1.49 0.79 113 726 18.0 2571 818 .44 4.71 1.63 1.39 113 721 22.3 2787 825 .45 4.82 1.66 1.67 113 716 26.9 2988 815 .46 4.73 1.62 1.96 113 711 31.5 3511 810 .47 4.71 1.60 2.79 113 706 36.4 3309 808 .47 4.67 1.59 2.46 113 701 41.2 3169 777 .47 4.32 1.47 2.25 113 696 46.0 3285 807 .47 4.66 1.59 2.42 113 691 50.9 3393 783 .47 4.40 1.49 2.61 113 686 55.9 3493 810 .47 4.71 1.60 2.76 113 681 60.8 3377 844 .47 5.09 1.74 2.55 113 676 65.7 3457 864 .47 5.34 1.82 2.67 113 671 70.7 3927 906 .47 5.89 2.00 3.49 113 666 75.6 3780 910 .47 5.93 2.02 3.21 113 661 80.6 3838 937 .47 6.28 2.14 3.30 113 656 85.5 3886 909 .47 5.92 2.01 3.41 113 651 90.5 3934 932 .47 6.22 2.12 3.49 113 Compressional Shear Poisson's Young's Shear Bulk Zones Velocity Velocity Ratio Modulus Modulus Modulus Density 741-736 1850 783 .39 4.16 1.49 0.64 113 736-691 4480 783 .48 4.44 1.49 4.69 113 691-651 4480 1275 .46 11.54 3.96 4.36 113 Note:

1. The density is a representative value determined from laboratory testing of soil samples taken near the array.

Calculation Reference CEB810515025 T251-11.doc

SQN Table 2.5.1-12 SEQUOYAH NUCLEAR PLANT ERCW PIPELINE IN-SITU DOWN-HOLE SOIL DYNAMICS UNSATURATED SOIL Dynamic Dynamic Compressional Shear Shear Young's Velocity Velocity Modulus Modulus Ft./Sec. Ft./Sec. PSI x 103 PSI x 103 Measured Calculated Calculated Calculated Average 3173 1523 49.2 132.8 Minimum 1585 761 12.5 33.8 Maximum 3888 1867 75.2 203.10 SATURATED SOIL 4005 1207 31.4 91.2 Calculated Reference B41861022009 T251-12.doc

SQN Table 2.5.1-13 (Sheet 1) SEQUOYAH NUCLEAR PLANT ADDITIONAL DIESEL GENERATOR BUILDING Summary of Cross-Hole Data for 19.6- and 24.4-foot Travel Paths Elevation Vp VP Vs Vs Poisson's Ratio Poisson's Ratio (feet) Range Average Range Average Range Average 715 1970 1970 890 - 930 0.33 -0.37 0.36 710 1880 - 1960 1930 920 - 1060 990 .27 - .36 .32 705 1850 1870 920 - 1120 1035 .21 - .34 .28 700 1920 - 2220 2070 905 - 1080 990 .27 - .40 .35 695 2180 - 2220 2215 1030 - 1085 1095 .33 - .36 .34 690 2880 2900 1100 - 1210 1165 .39 - .41 .40 685 3015 - 3470 3350 1350 - 1420 1435 .36 - .41 .39 680 4445 - 4900 4830 1510 - 1690 1635 .42 - .45 .44 675 4665 - 5315 5035 1720 - 1780 1790 .42 - .44 .43 670 5600 - 6110 5825 1835 - 2035 1945 .42 - .45 .44 665 5435 - 5765 5605 1880 - 1920 1870 .43 - .44 .44 660 5600 - 5695 5895 1745 - 1920 1890 .43 - .45 .44 655 5600 - 5695 5895 1920 - 1985 2055 .43 - .44 .43 650 5555 - 5600 5640 1920 - 2070 2060 .42 - .43 .42 648 N/A 5960 N/A 2070 N/A .43 Notes:

1. Averages calculated from all velocities (minimum, preferred, and maximum) at each elevation. These averages were used to calculate the Poisson's Ratio average.

2. The ranges are from preferred arrival times at each elevation.

Calculation Reference 41861022012 T251-13.doc

SQN Table 2.5.1-13 (Sheet 2) SEQUOYAH NUCLEAR PLANT ADDITIONAL DIESEL GENERATOR Building Summary of Cross-Hole Data for 15.2 Foot Travel Distance (preferred arrival times) Elevation Vp Vs Poisson's (feet) (fps) (fps) Ratio 715.0 1925 975 0.33 710.0 2350 1040 .38 705.0 2550 1230 .35 700.0 2925 1600 .29 695.0 3800 2200 .25 690.0 4110 2340 .26 685.0 3800 2110 .28 680.0 5040 2550 .33 675.0 6050 3050 .33 670.0 6040 2440 .40 655.0 6040 2560 .39 660.0 6050 2540 .39 655.0 6050 2500 .40 650.0 6000 2440 .40 646.5 5000 2330 .36 Summary of Downhole Data Elevation Vp Vs Poisson's (feet) (fps) (fps) Ratio 720-700 2375 940 0.41 700-640 5350 2075 .41 Calculation Reference B41861022012 T251-13.doc

SQN TABLE 2.5.1-14 SEQUOYAH NUCLEAR PLANT PRIMARY REFUELING WATER TANKS SEISMIC REFRACTION SURVEY IN-SITU ELASTIC PROPERTIES Vp Vs Compressional Shear Shear Velocity Velocity Density Poisson's Modulus ft/sec ft/sec lbs/cu ft Ratio psi (103)

Zones Measured Calculated Assumed Assumed Calculated One 2150 1033 110 0.35 25.3 2150 878 110 0.4 18.3 2150 648 110 0.45 9.9 Two 3250 1561 110 0.35 57.8 3250 1326 110 0.4 41.8 3250 980 110 0.45 22.8 Zone one - Between elevations 705.0 and 696.9 Zone two - Between elevations 696.5 and 679.1.

Surface elevation 705.0 Top of rock 679.1, as computed from the refraction survey. Calculation Reference B41861022008 T251-14.doc

SQN Table 2.5.2-1 (Sheet 1) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

1. 1776 Nov 5 IV Jackson Co.,NC 35.4 83.2

2. 1817 Dec 11 IV SC-GA 0.0 0.0

3. 1817 Dec 12 <IV KY 0.0 0.0

4. 1825 Mar 19 Columiba,TN 35.6 87.0

5. 1828 Mar 10 IV Southwestern VA 0.0 0.0

6. 1829 <IV Andrews,NC 35.2 83.8

7. 1843 Aug 9 IV Columbia,TN 35.6 87.0

8. 1844 Jun <IV Jackson Co.,NC 35.2 83.1

9. 1844 Nov 28 VI Knoxville,TN 36.0 83.9

10. 1848 <IV McDowell Co.,NC 35.7 82.0

11. 1851 Aug 11 V Asheville,NC 35.6 82.6

12. 1852 Oct 12 <IV Clinton,GA 33.0 83.5

13. 1852 Oct 23 <IV Clinton,GA 33.0 83.5

14. 1854 Feb 13 <IV Manchester,KY 37.2 83.8

15. 1860 Jan 20 NC-SC-GA 0.0 0.0

16. 1872 Jun 17 IV Milledgeville,GA 33.1 83.2

17. 1874 Feb 22 V McDowell Co.,NC 35.7 82.1

18. 1875 Jul 29 <IV Milledgeville,GA 33.1 83.2

19. 1875 Nov 2 IV Washington, GA 33.7 82.7

20. 1875 Nov 12 <IV Knoxville,TN 36.0 83.9

21. 1876 Jan 23 <IV McDowell Co.,NC 35.7 82.0

22. 1877 Apr 26 <IV Franklin,NC 35.2 83.4

23. 1877 May 25 <IV Knoxville,TN 36.0 83.9

24. 1877 Jun 3 <IV Stanford,KY 37.5 84.7

25. 1877 Oct 9 <IV Hendersonville,NC 35.3 82.5

26. 1877 Nov 16 IV Knoxville,TN 36.0 83.9

27. 1878 Nov 23 <IV Murphy,NC 35.1 84.0

28. 1880 Jan 28 <IV McDowell Co.,NC 35.7 82.0

29. 1882 Oct 15 <IV Murphy,NC 35.1 84.0

30. 1883 Jan 1 IV Ashwood,TN 35.6 87.1

31. 1884 Jan <IV McDowell Co.,NC 35.7 82.0

32. 1884 Mar 31 <IV Milledgeville,GA 33.1 83.2

33. 1884 Apr 30 <IV Ogreeta,NC 35.2 84.2

34. 1884 <IV Elk Mt.,NC 35.7 82.5

35. 1884 Aug 25 IV Knoxville,TN 36.0 83.9

36. 1886 Feb 5 IV Valley Head,AL 34.6 85.6

37. 1888 Mar 17 <IV Jonesboro,TN 36.3 82.5

38. 1889 Jun 7 IV Benton Co.,TN 35.9 88.1

39. 1889 Sep 28 <IV Parksville,TN 35.1 84.6

40. 1892 Dec 2 V Chattanooga,TN 35.0 85.3

41. 1895 Jul 27 Savannah,TN 35.2 88.3

42. 1898 Mar 30 <IV Mt. Hermon,KY 36.8 85.8

43. 1898 Jun 6 <IV Richmond,KY 37.8 84.3

44. 1902 May 29 IV Chattanooga,TN 35.0 85.3

45. 1902 Oct 18 V Chattanooga,TN 35.0 85.3

46. 1904 Mar 5 <IV Maryville,TN 35.8 84.0

47. 1909 Oct 8 <IV Dalton,GA 34.8 85.0

48. 1911 Apr 22 <IV Hendersonville,NC 35.3 82.5 49 1912 Oct 23 <IV Macon,GA 32.8 83.6

50. 1912 Dec 7 <IV West Springs,SC 34.8 81.8

51. 1913 Jan 1 VII West Springs,SC 34.8 81.8

52. 1913 Mar 13 <IV Calhoun,GA 34.5 85.0 T252-1.doc

SQN Table 2.5.2-1 (Sheet 2) (Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

53. 1913 Mar 28 VI Knoxville,TN 36.0 83.9

54. 1913 Apr 17 V Madisonville,TN 35.5 84.4

55. 1913 May 2 <IV Madisonville,TN 35.5 84.4

56. 1913 Aug 3 IV Knoxville,TN 36.0 83.9

57. 1914 Jan 24 IV Sweetwater,TN 35.6 84.5

58. 1914 Mar 5 IV Central GA 33.5 84.0

59. 1915 Jan 14 IV Briston,TN 36.6 82.2

60. 1915 Oct 29 IV Marshall,NC 35.8 82.7

61. 1916 Feb 21 VII Waynesville,NC 35.5 83.0

62. 1916 Mar 2 IV Anderson,SC 34.5 82.7

63. 1916 Oct 18 VII Irondale,AL 33.5 86.7

64. 1916 Nov 4 IV Birmingham,AL 33.5 86.8

65. 1917 Jan 2 IV McMillan,TN 36.6 83.9

66. 1917 Jan 25 Jefferson City,TN 36.1 83.5

67. 1917 Mar 5 Knoxville,TN 36.0 83.9

68. 1917 Mar 27 V Jefferson City,TN 36.1 83.5 69 1917 Apr 19 <IV southwestern VA 0.0 0.0

70. 1918 Jan 17 IV Knoxville,TN 36.0 83.9

71. 1918 Jun 22 IV Lenoir City,TN 35.8 84.3

72. 1920 Apr 7 II 36.3 88.2

73. 1920 Dec 24 IV Glen Alice,TN 35.8 84.7

74. 1921 Jul 15 V Mendota,VA 36.7 82.3

75. 1921 Sep 2 IV Statesville,TN 36.0 86.1

76. 1921 Dec 15 IV Glen Alice,TN 35.0 84.7

77. 1922 Mar 30 <IV Farmington,TN 35.5 86.7

78. 1922 Mar 30 <IV Arcadia,TN 36.6 82.5

79. 1923 Oct 18 IV Hendersonville,NC 35.3 82.5

80. 1924 Jan 1 IV Greenville,SC 34.8 82.4

81. 1924 Oct 20 IV Pickens,SC 34.9 82.7

82. 1924 Nov 13 V Bristol,VA 36.6 82.2

83. 1926 Jul 8 VII McDowell Co.,NC 35.7 82.0

84. 1927 Jun 16 IV Scottsboro,AL 34.7 86.0

85. 1927 Jul 20 V Knoxville,TN 36.0 83.9

86. 1927 Oct 8 IV Chattanooga,TN 35.0 85.3

87. 1928 Mar 7 IV Columbia,TN 35.6 87.0

88. 1928 Nov 3 VII Hot Springs,NC 35.9 82.8

89. 1928 Nov 20 IV Hot Springs,NC 35.9 82.8

90. 1929 Oct 28 IV Due West,SC 34.3 82.4

91. 1930 Aug 30 V Kingston,TN 35.9 84.5 92 1930 Oct 16 VI Knoxville,TN 36.0 83.9

93. 1930 Dec 10 Due West,SC 34.3 82.4

94. 1931 Apr 1 Hopkinsville,KY 36.9 87.5

95. 1931 May 5 VI Birmingham,AL 33.5 86.8

96. 1931 Nov 27 <IV Nashville,TN 36.2 86.8

97. 1935 Jan 1 V GA-NC 35.1 83.6

98. 1936 Jan 1 <IV Blue Ridge,GA 34.9 84.3

99. 1938 Mar 31 IV Tapoco,NC 35.5 84.0 100. 1939 May 5 V Anniston,AL 33.7 85.8 101. 1939 Jun 24 IV Huntsville,AL 34.7 86.6 102. 1940 Oct 19 IV Ryall Springs,TN 35.0 85.1 103. 1940 Dec 25 IV Hot Springs,NC 35.9 82.8 104. 1941 Mar 4 <IV Rockford,TN 35.9 83.9 T252-1.doc

SQN Table 2.5.2-1 (Sheet 3) (Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 105. 1941 May 10 IV Asheville,NC 35.6 82.6 106. 1941 Sep 8 IV Lookout Mt.,TN 35.0 85.4 107. 1945 Jun 14 V Cleveland,TN 35.2 84.9 108. 1946 Apr 7 IV Cleveland,TN 35.2 84.9 109. 1947 Jun 6 IV Knoxville,TN 36.0 83.9 110. 1947 Dec 28 IV Ryall Springs,TN 35.0 85.1 111. 1948 Feb 10 VI Wells Springs,TN 36.4 84.0 112. 1949 Sep 17 V Pennington Gap,VA 36.8 83.0 113. 1950 Jun 19 IV Tapoco,NC 35.5 84.0 114. 1952 Feb 6 V Birmingham,AL 33.5 86.8 115. 1952 Jun 11 VI Johnson City,TN 36.3 82.4 116. 1953 Nov 10 IV Knoxville,TN 36.0 83.9 117. 1953 Dec 5 IV Knoxville,TN 36.0 83.9 118. 1954 Jan 1 IV Hazard,KY 37.2 83.2 119. 1954 Jan 2 VI Hazard,KY 37.2 83.2 120. 1954 Jan 14 IV Knoxville,TN 36.0 83.9 121. 1954 Jan 23 IV Etowah,TN 35.3 84.5 122. 1955 Jan 6 IV Bristol,TN 36.6 82.2 123. 1955 Jan 12 IV Maryville,TN 35.8 84.0 124. 1955 Jan 25 IV Knoxville,TN 36.0 83.9 125. 1956 Jan 5 IV Due West,SC 34.3 82.4 126. 1956 May 19 IV Due West,SC 34.3 82.4 127. 1956 May 27 IV Due West,SC 34.3 82.4 128. 1956 Sep 7 VI Maynardville,TN 36.2 83.8 129. 1956 Sep 9 IV College Grove,TN 35.8 86.7 130. 1957 Jan 25 IV Middlesboro,KY 36.6 83.7 131. 1957 Apr 23 VI Birmingham,AL 33.5 86.8 132. 1957 May 13 VI McDowell Co.,NC 35.7 82.0 133. 1957 Jun 23 IV Dixie Lee Junction,TN 35.9 84.2 134. 1957 Jul 2 VI Asheville,NC 35.6 82.6 135. 1957 Nov 7 <IV Powell,TN 36.0 84.0 136. 1957 Nov 24 VI Bryson City,NC 35.4 83.4 137. 1958 May 16 IV Asheville,NC 35.6 82.6 138. 1958 Oct 20 IV Anderson,SC 34.5 82.7 139. 1959 Jun 13 IV Tellico Plains,TN 35.4 84.3 140. 1959 Aug 12 VI Meridianville,AL 34.8 86.6 141. 1960 Jan 3 IV Spruce Pine,NC 35.9 82.1 142. 1960 Feb 9 VI Edneyville,NC 35.4 82.4 143. 1960 Apr 15 IV Maryville,TN 35.8 84.0 144. 1963 Apr 11 IV Greenville,SC 34.8 82.4 145. 1963 Nov 14 <IV Nashville,TN 36.2 86.8 146. 1963 Dec 5 <IV Beechmont,KY 37.2 87.0 147. 1963 Dec 15 <IV Beechmont,KY 37.2 87.0 148. 1964 Jan 20 IV Pensacola,NC 35.8 82.3 149. 1964 Feb 18 V Mentone,AL 34.6 85.6 150. 1964 Mar 13 IV Haddock,GA 33.0 83.4 151. 1964 Jul 28 <IV Inskip,TN 36.0 84.0 152. 1964 Oct 13 Knoxville,TN 36.0 83.9 153. 1965 Apr 7 McCormick,SC 33.9 82.3 154. 1965 Nov 8 <IV Canton,GA 34.2 84.5 155. 1966 Aug 24 IV Maryville,TN 35.8 84.0 156. 1969 May 5 GA-SC Border 33.9 82.50 T252-1.doc

SQN Table 2.5.2-1 (Sheet 4) (Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 157. 1969 Jul 13 V Knoxville,TN 36.0 83.9 158. 1969 Jul 24 Knoxville,TN 36.0 83.9 159. 1969 Dec 13 IV SC-NC Border 35.0 83.0 160. 1971 Jul 13 IV Kingston,TN 35.9 84.5 161. 1971 Jul 13 VI Newry,SC 34.7 82.9 162. 1971 Oct 9 V Gatlinburg,TN 35.7 83.5 163. 1973 Nov 30 VI Maryville,TN 35.8 84.0 164. 1974 Aug 2 V McCormick Co., SC 33.9 82.5 165. 1974 Oct 8 Clark Hill Reservoir,SC 34.0 82.3 166. 1974 Nov 5 Clark Hill,SC 33.7 82.2 167. 1974 Dec 3 Mt. Carmel,SC 34.0 82.5 168. 1975 Feb 10 Gatlinburg,TN 35.7 83.5 169. 1975 May 2 Oakdale,TN 36.0 84.6 170. 1975 May 14 Oak Ridge,TN 36.0 84.3 171. 1975 Jun 24 IV Fayette,AL 33.7 87.8 172. 1975 Aug 29 VI Palmerdale,AL 33.8 86.6 173. 1975 Oct 18 IV Jocassee Lake Dam,SC 34.9 83.0 174. 1975 Nov 7 Samantha,AL 33.4 87.6 175. 1975 Nov 25 IV Salem,SC 34.9 83.0 176. 1976 Jan 19 VI Knox Co.,KY 36.9 83.8 177. 1976 Feb 4 VI Conasauga,TN 35.0 84.7 178. 1976 Apr 15 V Sacramento,KY 37.4 87.3 179. 1977 Jul 27 V Athens,TN 35.4 84.6 180. 1978 Mar 1 III near Huntsville,AL 34.4 86.6 181. 1978 Oct 27 near Jasper,AL 33.8 87.5 182. 1979 Jan 19 IV Newry,SC 34.7 82.9 183. 1979 Aug 13 V near Cleveland,TN 35.2 84.4 184. 1979 Aug 26 VI Tamasee,SC 34.9 83.1 185. 1979 Sep 12 V Maryville,TN 35.8 84.0 186. 1980 Mar 23 IV Narrows,KY 37.6 86.7 187. 1980 Apr 21 Maryville,TN 35.8 84.0 188. 1980 Jun 25 IV Maryville,TN 35.8 84.0 189. 1980 Jul 12 III near Horse Branch,KY 37.3 87.0 T252-1.doc

SQN Table 2.5.2-1 (Sheet 5) (Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON

87. 1928 Mar 7 IV Columbia,TN 35.6 87.0 7.

11. IV

92. 1930 Oct 16 VI Knoxville,TN 36.0 83.9

13. VI

26. <IV

29. <IV

32. IV

41. IV

59. VI

62. IV 73.

76. IV

92. V 116. IV 123. IV 124. IV 127. IV 131. IV 159.

164. V 165.

83. 1926 Jul 8 VII McDowell Co.,NC 35.7 82.0

15. <IV

27. <IV

34. <IV

37. <IV 139. VI 134. 1957 Jul 2 VI Asheville,NC 35.6 82.6

16. V 112. IV 144. IV

13. 1852 Oct 23 <IV Clinton,GA 33.0 83.5

17. <IV

16. 1872 Jun 17 IV Milledgeville,GA 33.1 83.2

24. <IV

38. <IV

79. 1923 Oct 18 IV Hendersonville,NC 35.3 82.5

31. <IV

54. <IV

29. 1882 Oct 15 <IV Murphy,NC 35.1 84.0

33. <IV 149. 1964 Feb 18 V Mentone,AL 34.6 85.6

42. IV

45. 1902 Oct 18 V Chattanooga,TN 35.0 85.3

46. V

50. IV

93. IV 163. 1973 Nov 30 VI Maryville,TN 35.8 84.0

52. <IV 130. IV 150. IV T252-1.doc

SQN Table 2.5.2-1 (Sheet 6) (Continued) SEQUOYAH PLANT HISTORICAL EARTHQUAKE LISTING 200 MILE RADIUS AROUND 85.1 W LON 35.2 N LAT YEAR MONTH DAY INTENSITY LOCATION NLAT WLON 162. IV 192. V 194. 195. IV

51. 1913 Jan 1 VII West Springs,SC 34.8 81.8

56. <IV

54. 1913 Apr 17 V Madisonville,TN 35.5 84.4

61. <IV

82. 1924 Nov 13 V Bristol,VA 36.6 82.2

65. IV 129. IV 138. 1958 Oct 20 IV Anderson,SC 34.5 82.7

68. IV 131. 1957 Apr 23 VI Birmingham,AL 33.5 86.8

70. IV 102. VI 121. V

68. 1917 Mar 27 V Jefferson City,TN 36.1 83.5 72.

177. 1976 Feb 4 VI Conasauga,TN 35.0 84.7

82. IV 144. 1963 Apr 11 IV Greenville,SC 34.8 82.4

87. IV

88. 1928 Nov 3 VII Hot Springs,NC 35.9 82.8

96. IV 110. IV 127. 1956 May 27 IV Due West,SC 34.3 82.4

97. IV 100.

132. IV 133. IV

91. 1930 Aug 30 V Kingston,TN 35.9 84.5 167. IV 145. 1963 Nov 14 <IV Nashville,TN 36.2 86.8 103. <IV 113. 1950 Jun 19 IV Tapoco,NC 35.5 84.0 106. IV 110. 1947 Dec 28 IV Ryall Springs,TN 35.0 85.1 109. IV 107. 1945 Jun 14 V Cleveland,TN 35.2 84.9 115. IV 119. 1954 Jan 2 VI Hazard,KY 37.2 83.2 125. IV 151. 1964 Jul 28 <IV Inskip,TN 36.0 84.0 142. <IV 147. 1963 Dec 15 <IV Beechmont,KY 37.2 87.0 153. <IV 164. 1974 Aug 2 V McCormick Co.,SC 33.9 82.5 163.

161. 1971 Jul 13 VI Newry,SC 34.7 82.9 189. IV 162. 1971 Oct 9 V Gatlinburg,TN 35.7 83.5 175. 1975 Nov 25 IV Salem,SC 34.9 83.0 180. IV T252-1.doc

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F'HYSIOGRAPMY BY SCALE: N. M. FENNEMAN 0 Figure 2.5.1-1 Physiographic Map of Plant Area (464K33)

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SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Figure 2.5.l-3 GEOLOGIC INVESTIGATIONS EJKnoa~*....S..,........ '. I GEOLOGIC MAP OF PLANT SITE f- I EJ /.~,...*--**_,

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                         ')/                                                                      FINAL SAFETY ANALYSIS REPORT
                        '                                                                Figure 2. 5.1-' "I U1  ""v                                                                                   VALLEY AND RIDGE PROVINCE SEISMIC.REFLECTION PROFILE 00 U1 (822A2128)

TOO . t!l'S SECTION W-.'10*00

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                                                                                                                                     ,..                                                                                  :".:.*-I     ,,  f Figure 2.5.1-5                Geologic Sections.

W-30+00 through W-36+100 (822Kll80-1)

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676 660 660 SECTION W-381-00 no 1'00 no SECTION . B-B SCALES:

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                                   ~f/#116-                                                                                                                                                                     SEQUOYAH NUCLEAR PLANT
                                                                                                                                                                                                                     ..       FINAL SAFETY ANALYSIS REPORT.

FIGURE 2 *.5.1-6. GELCGIC SECTIONS W-38+00, AA, AND B-B ( 822Kll80-2)

I SECTION N-u~oo 71!5 7()() 67.~ SECTION N-/2+00 NOTES: _, Tll* INarr llO/id I - on Illa ut:lion* ._.,,,,., l/N 71!5 ..01.'t#Hd ,,11:1 ln1111 Ill# 1n1-,IMtltl..t s/llJltl and

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~ DENOTE ( REVIS ED BY AMEN DMEN EXCEP T AS NOTED AND BACKFILL STRUCTURES 2 3 5 6 7 8 9 10 FSAR F !GURE MAINTAINED
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SHEET 3 FIGURE 2.5.1-12b Revised by Amendment 13
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.a, SOIL BORING FDR SPUT-sPO&W SAMl'i.IM3 @ SOii. IJORIM; FOR SPIJT-SPOON MO VNDIS7l/RIJED SAMPUNG e SOIL BORl/13 FOR UNOJSTURBEO SAMPLING (SOIL PERM£1181UTY) 0 BORING ADVANCED BY nsHTAJL/Nt:
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NOTES: l COORDINATCS AT lHE INTERSEeTkJN or 111£ N-S AND E-W iJA!iCLINE3 ARE TE/111£SSEC lAM8EllT a:xJINJINATES. 100 0 /OD SCALE - FEET SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT FIGURE 2.5.1-13 ESSENTIAL SOIL INVESTIGATIONS IN SITU (REVISED BY AMENDMENT 13)
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Figure 2.5.2-3 Isoseismals of the Charleston Earthquake 597
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GEORGIA 0 IC 20 30 MILES Figure 2.5.2-4 East Tennessee Earthquake of April 17, 1913 599
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L (/) F Figure 2.5.2-6 L 4.. The A~abama Earthquake of October 18, 1916 l' )0 .. - - qg* 8'. 603 Jsoo;et~o: or the Al3.b.am:a e~hqg,."lJce ot October lEl. 1916.
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TENNESSEE ALABAMA GEORGlA 0 TRENTON DAL TON After Brill Figure 2. 5. 2-9 Chattanooga Earthquake October 19, 1940 609
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Figure 2. 5. 2-11 Comparison of Response Spectra for Safe Shutdown Earthquake, 1/23 Damping 613
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PERIOD (SECS) Figure 2.5.2-12 Comparison of Response Spectra for Safe Shutdown* Earthquake, 1% Damping 615
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Figure 2.5.2-14 Comparison of Response Spectra for Safe Shutdown Earthquake, 5% Damping 619
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. TIME {DAYS)
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Figure 2.5.6-2 Section of Forebay and Int~ke Slope Sequoyah Nuclear e1ant Pumping Statibn
SQN
2.6 CONCLUSION
S The various significant characteristics of the Sequoyah Nuclear Plant site which have, to varying degrees, influenced plant design and operating plans, are shown in Table 2.6-1. The foregoing discussions have shown how these various site characteristics have influenced the plant design and how this design has taken into consideration the site related aspects of the NRC Regulatory Guides, based on the material in Sections 2.1 through 2.5 and the following considerations, it is concluded that the Sequoyah Nuclear Plant site meets the Reactor Siting Criteria of 10 CFR Part 100:
l. The site, consisting of approximately 525 acres, provides a minimum exclusion distance of approximately 1824 feet.

2. There are no residences on the site.

3. The population density and use characteristics of the environs are compatible with the location of a nuclear plant. The low population zone and population center distances are approximately 3 and 7.5 miles, respectively.

4. The geological, seismological, hydrological, and meteorological characteristics of the site and environs are considered suitable for its intended uses and have been considered in the plant design and operating plans.

5. As analyzed in Chapter 15, the radiation doses to the public at the exclusion distance and low population zone distance, under postulated hypothetical accident conditions, are well within the reference values of 10 CFR Part 100.
S2-6.doc 2.6-1
SQN Table 2.6-1 SEQUOYAH NUCLEAR PLANT SITE CHARACTERISTICS
1. Site location Approximately 7.5 miles northeast of Chattanooga, TN at TN River Mile (TRM) 484.5

2. Site area 525 acres

3. Exclusion distance 1824 feet

4. Low population zone distance three miles

5. Population center distance 7.5 miles

6. Elevation of plant grade 705 feet MSL

7. Tennessee River normal maximum pool elevation 682.5 feet MSL

8. Design basis flood 726.8 feet MSL

9. Population density, 0-10 mile, 1970 census 102.32 persons/sq. mile

10. Distance from diffuser discharge to the nearest downstream 10.7 miles drinking water intake

11. Tectonic province of site Southern Appalachian

12. Maximum historical earthquake MM VIII

13. Safe Shutdown Earthquake (SSE) 0.18 g horizontal peak accelerations 0.12 g vertical

14. Depth of soil overburden Ranges from 3.2 to 45 feet

15. Bedrock at site Interbedded limestone and shale
-5
16. Tornado probability 4.4 x 10

17. Average annual air temperature 59.7°F

18. Average annual precipitation 58 inches

19. Chickamauga Reservoir Maximum 30°C surface temperature Minimum 2°C

20. Intake location TRM 484.8

21. Diffuser location TRM 483.7 T261-1.doc
SQN TABLE OF CONTENTS Section Title Page 3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1-1 3.1.1 SINGLE FAILURE CRITERIA 3.1-1 3.1.2 OVERALL REQUIREMENTS 3.1-2 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, 3.2-1 AND COMPONENTS 3.2.1 SEISMIC QUALIFICATIONS 3.2-1 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION 3.2-1 (FLUID COMPONENTS) 3.2.2.1 Class A 3.2-1 3.2.2.2 Class B 3.2-2 3.2.2.3 Class C 3.2-3 3.2.2.4 Class D 3.2-3 3.2.2.5 Relationship of Applicable Codes 3.2-4 to Safety Classification for Mechanical Components 3.2.2.6 Nonnuclear Safety Class (NNS) 3.2-4 3.
2.3 REFERENCES
3.2-5 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.3.1 WIND LOADINGS 3.3-1 3.3.1.1 Design Wind Velocity 3.3-1 3.3.1.2 Basis for Wind Velocity Selection 3.3-1 3.3.1.3 Vertical Velocity Distribution and Gust Factor 3.3-1 3.3.1.4 Determination of Applied Force 3.3-1 3.3.2 TORNADO LOADINGS 3.3-1 3.3.2.1 Applicable Design Parameters 3.3-1 3.3.2.2 Determination of Forces on Structures 3.3-2 3.3.2.3 Ability of Category I Structures to 3.3-3 Perform Despite Failure of Structures Not Designed for Tornado Loads 3.
3.3 REFERENCES
3.3-4 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 FLOOD ELEVATIONS 3.4-1 S3-0toc.doc 3-1
SQN-26 TABLE OF CONTENTS Section Title Page 3.4.2 FLOOD FORCE APPLICATION 3.4-1 3.4.3 FLOOD PROTECTION 3.4-1 3.5 MISSILE PROTECTION 3.5-1 3.5.1 MISSILE BARRIER AND LOADINGS 3.5-4 3.5.2 MISSILE SELECTION 3.5-4 3.5.2.1 Control Rod Drive Mechanism 3.5-4 3.5.2.2 Valves 3.5-4 3.5.2.3 Temperature and Pressure Sensing 3.5-4 Assemblies 3.5.2.4 Pressurizer Heaters 3.5-5 3.5.2.5 Electrical Cables 3.5-5 3.5.2.6 Steel Containment Structure 3.5-5 3.5.2.7 Shield Building 3.5-6 3.5.2.8 Ice Condenser Containment System 3.5-6 3.5.2.9 Emergency Core Cooling System 3.5-6 3.5.2.10 Containment Spray System 3.5-7 3.5.2.11 Containment Isolation System 3.5-7 3.5.2.12 Diesel Generator Building 3.5-8 3.5.2.13 Control Bay 3.5-8 3.5.2.14 Spent Fuel Pool 3.5-8 3.5.2.15 The Auxiliary Building 3.5-9 3.5.2.16 The Control Building 3.5-9 3.5.2.17 Intake Structure 3.5-9 3.5.2.18 Components Not Credible Sources 3.5-9 of Missiles Inside Containment 3.5.2.19 Missiles From Main Turbine 3.5-10 3.5.2.20 Nearby Site Activity Generated 3.5-10 Missiles 3.5.2.21 Tornado-Generated Missiles 3.5-11 3.5.2.22 Condensate Storage Tanks 3.5.11 3.5.3 SELECTED MISSILES 3.5-11 3.5.4 BARRIER DESIGN PROCEDURES 3.5-13 3.5.5 MISSILE BARRIER FEATURES 3.5-17 3.
5.6 REFERENCES
3.5-24 3.6 PROTECTION AGAINST EFFECTS ASSOCIATED 3.6-1 WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 SYSTEMS IN WHICH DESIGN BASIS PIPING 3.6-1 BREAKS OCCUR 3.6.1.1 Main Reactor Coolant Piping System 3.6-1 3.6.1.2 Other Piping Systems 3.6-1 S3-0toc.doc 3-2
SQN TABLE OF CONTENTS Section Title Page 3.6.2 DESIGN BASIS PIPING BREAK CRITERIA 3.6-6 3.6.2.1 Main Reactor Coolant Piping System 3.6-6 3.6.2.2 Other Pipe Systems (Inside Containment) 3.6-6 3.6.3 DESIGN LOADING COMBINATIONS 3.6-11 3.6.3.1 Main Reactor Coolant Piping System 3.6-11 3.6.3.2 Other Piping Systems (Inside Containment) 3.6-11 3.6.4 DYNAMIC ANALYSES 3.6-13 3.6.4.1 Main Reactor Coolant System 3.6-13 3.6.4.2 Other Piping Systems (Inside Containment) 3.6-14 3.6.5 PROTECTIVE MEASURES 3.6-18 3.6.5.1 Main Reactor Coolant Piping System 3.6-18 3.6.5.2 Other Piping Systems 3.6-20 3.6.6 ASSUMPTIONS 3.6-20 3.6.6.1 Plant Operating Mode 3.6-20 3.6.6.2 Offsite Power 3.6-20 3.6.6.3 Unintended Operation of Equipment 3.6-20 3.6.6.4 Operator Response 3.6-20 3.6.7 SYSTEM EVALUATION 3.6-20 3.6.7.1 General 3.6-21 3.6.7.2 Objective 3.6-21 3.6.7.3 Guiding Philosophy 3.6-21 3.6.7.4 Exclusions 3.6-21 3.6.7.5 Evaluation Procedure 3.6-22 3.6.7.6 Main Steam 3.6-22 3.6.7.7 Main Feedwater 3.6-23 3.6.8 WELDS 3.6-24 3.6.9 REFERENCE 3.6-24 3.7 SEISMIC DESIGN 3.7-1 3.7.1 SEISMIC DESIGN FOR STRUCTURES 3.7-1 3.7.1.1 Design Response Spectra 3.7-1 3.7.1.2 Design Response Spectra Derivation 3.7-1 3.7.1.3 Critical Damping Values 3.7-1 3.7.1.4 Bases for Site Dependent Analysis 3.7-1 3.7.1.5 Soil-Supported Category I Structures 3.7-1 3.7.1.6 Soil-Structure Interaction 3.7-1 3.7.2 SEISMIC SYSTEM ANALYSIS 3.7-2 3.7.2.1 Category I Systems and Components 3.7-2 Supplied by Westinghouse 3.7.2.2 Category I Structures Listed 3.7-10 S3-0toc.doc 3-3
SQN-16 TABLE OF CONTENTS Section Title Page 3.7.2.3 Seismic Analysis Methods for Category I 3.7-23 Structures 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7-27 3.7.3.1 Determination of Number of Earthquake Cycles 3.7-27 3.7.3.2 Basis for Frequency Selection 3.7-28 3.7.3.3 Modal Response Combinations (TVA Analysis) 3.7-28 3.7.3.4 Modal Response of Closely Spaced 3.7-29 Frequencies (NSSS Analysis) 3.7.3.5 Equivalent Static Loads 3.7-29 3.7.3.6 Seismic Analysis of System Piping 3.7-29 3.7.3.7 Basis for Computing Combined Response 3.7-33 3.7.3.8 Amplified Seismic Response 3.7-33 3.7.3.9 Use of Alternate Dynamic Analysis 3.7-33 3.7.3.10 Modal Period Variation 3.7-33 3.7.3.11 Torsional Effects of Eccentric Masses 3.7-33 3.7.3.12 Buried Seismic Category I Piping Systems 3.7-34 3.7.3.13 Interaction of Other Piping with 3.7-34 Category I Piping 3.7.3.14 Field Location of Supports and 3.7-35 Restrains 3.7.3.15 Seismic Analysis for Fuel Elements, 3.7-35 Control Rod Assemblies, and Control Rod Drives 3.7.4 SEISMIC INSTRUMENTATION PROGRAM 3.7-35 3.7.4.1 Comparison with NRC Regulatory 3.7-36 Guide 1.12 3.7.4.2 Location and Description of 3.7-36 Instrumentation 3.7.4.3 Control Room Operator Notification 3.7-37 3.7.4.4 Controlled Shutdown Logic 3.7-38 3.7.4.5 References 3.7-39 3.7.5 SEISMIC DESIGN CONTROL MEASURES 3.7-39 3.7.5.1 Westinghouse Control Measures 3.7-39 3.7.5.2 TVA Control Measures 3.7-40 3.
7.6 REFERENCES
3.7-41 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1 3.8.1 CONCRETE CONTAINMENT 3.8-1 3.8.1.1 Shield Building 3.8-1 3.8.1.2 Applicable Codes, Standards and 3.8-2 Specifications 3.8.1.3 Loads and Loading Combinations 3.8-4 3.8.1.4 Design and Analysis Procedures 3.8-7 S3-0toc.doc 3-4
SQN TABLE OF CONTENTS Section Title Page 3.8.1.5 Structural Acceptance Criteria 3.8-10 3.8.1.6 Materials, Quality Control and 3.8-12 Special Construction Techniques 3.8.1.7 Testing and Inservice Surveillance 3.8-15 Requirements 3.8.2 STEEL CONTAINMENT SYSTEM 3.8-15 3.8.2.1 Description of the Containment 3.8-15 3.8.2.2 Applicable Codes, Standards, and 3.8-15 Specifications 3.8.2.3 Loads and Loading Combinations 3.8-24 3.8.2.4 Design and Analysis Procedure 3.8-30 3.8.2.5 Structural Acceptance Criteria 3.8-37 3.8.2.6 Design Loading Combinations and Stress Limits 3.8-40 3.8.2.7 References 3.8-41 3.8.3 CONCRETE INTERIOR STRUCTURES 3.8-41 3.8.3.1 Description of the Interior Structures 3.8-41 3.8.3.2 Applicable Codes, Standards and 3.8-47 Specifications 3.8.3.3 Loads and Loading Combinations 3.8-47 3.8.3.4 Design and Analysis Procedures 3.8-50 3.8.3.5 Structural Acceptance Criteria 3.8-63 3.8.3.6 Materials, Quality Control, and Special 3.8-67 Construction Techniques 3.8.3.7 Testing and Inservice Surveillance 3.8-71 Requirements 3.8.3.8 Environmental Effects 3.8-72 3.8.4 OTHER CATEGORY I STRUCTURES 3.8-73 3.8.4.1 Description of the Structures 3.8-73 3.8.4.2 Applicable Codes, Standards, and 3.8-82 Specifications 3.8.4.3 Loads and Loading Combinations 3.8-83 3.8.4.4 Design and Analysis Procedures 3.8-85 3.8.4.5 Structural Acceptance Criteria 3.8-95 3.8.4.6 Materials, Quality Control, and 3.8-98 Special Construction Techniques 3.8.4.7 Testing and Inservice Surveillance 3.8-100 Requirements 3.8.5 FOUNDATIONS AND CONCRETE SUPPORTS 3.8-101 3.8.5.1 Description of Foundation and Supports 3.8-101 3.8.5.2 Applicable Codes, Standards, and 3.8-103 Specifications 3.8.5.3 Loads and Loading Combinations 3.8-104 3.8.5.4 Design and Analysis Procedure 3.8-104 3.8.5.5 Structural Acceptance Criteria 3.8-105 3.8.5.6 Materials, Quality Control, and 3.8-107 Special Construction Techniques S3-0toc.doc 3-5
SQN TABLE OF CONTENTS Section Title Page 3.8.6 CATEGORY 1(L) CRANES 3.8-107 3.8.6.1 Polar Cranes 3.8-107 3.8.6.2 Auxiliary Building Crane 3.8-111 APPENDIX 3.8A SHELL TEMPERATURE TRANSIENTS 3.8A-1 APPENDIX 3.8B CONTAINMENT VESSEL PENETRATIONS 3.8B-1 APPENDIX 3.8C CONTAINMENT ANCHORAGE 3.8C-1 APPENDIX 3.8D COMPUTER PROGRAMS USED IN STRUCTURAL 3.8D-1 ANALYSIS APPENDIX 3.8E DESIGN PROCEDURE FOR REINFORCED 3.8E-1 CONCRETE BLOCK WALLS 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.9.1 DYNAMIC SYSTEM ANALYSIS AND TESTING 3.9-1 3.9.1.1 Vibration Operational Test Program 3.9-1 3.9.1.2 Dynamic Testing Procedures 3.9-3 3.9.1.3 Dynamic System Analysis Methods 3.9-5 for Reactor Internals 3.9.1.4 Correlation of Tests and Analytical 3.9-14 Results for Reactor Internals 3.9.1.5 Analysis Methods Under LOCA Loadings 3.9-15 3.9.1.6 Analytical Methods for ASME Code 3.9-16 Class 1 NSSS Components 3.9.2 SAFETY CLASS B, C, AND D FLUID COMPONENTS 3.9-17 3.9.2.1 Plant Conditions and Design Loading 3.9-17 Combinations 3.9.2.2 Design Loading Combinations 3.9-17 3.9.2.3 Inelastic Deformation 3.9-18 3.9.2.4 Design and Installation Criteria, 3.9-18 Pressure Relieving Devices 3.9.2.5 Stress Levels for Category I Components 3.9-19 3.9.2.6 Field Run Piping 3.9-29 3.9.2.7 Interim Acceptance Criteria 3.9-30 3.9.3 NSSS COMPONENTS NOT COVERED BY ASME CODE 3.9-31 3.9.3.1 Core and Internals Integrity 3.9-31 Analysis (Mechanical Analysis) 3.9.3.2 Reactor Internals Response Under 3.9-32 Blowdown and Seismic Excitation 3.9.3.3 Acceptance Criteria 3.9-33 3.9.3.4 Methods of Analysis 3.9-35 3.9.3.5 Blowdown Forces Due to Cold and Hot 3.9-35 Leg Break S3-0toc.doc 3-6
SQN-31 TABLE OF CONTENTS Section Title Page 3.9.3.6 Methods and Results of Blowdown 3.9-38 Analysis (Mechanical) 3.9.3.7 Control Rod Drive Mechanisms 3.9-39 3.9.3.8 Evaluation of Reactor Internals for 3.9-39 Large Branch Line Breaks 3.9.4 ADDITIONAL SUPPORT REQUIREMENTS 3.9-41 3.9.4.1 Support Welds 3.9-41 3.9.4.2 Allowable Loads for U-Bolts 3.9-41 and Unistruct Type Clamps 3.
9.5 REFERENCES
3.9-41 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3.10-1 AND ELECTRICAL EQUIPMENT 3.10.1 SEISMIC DESIGN CRITERIA 3.10-1 3.10.1.1 Instrumentation 3.10-1 3.10.1.2 Electrical Equipment 3.10-3 SEISMIC ANALYSES, TESTING PROCEDURES, 3.10-4 AND RESTRAINT MEASURES Instrumentation 3.10-4 Support Structures 3.10-4 3.
10.3 REFERENCES
3.10-8 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3.11-1 ELECTRICAL EQUIPMENT 3.11.1 EQUIPMENT IDENTIFICATION 3.11-1 3.11.2 ENVIRONMENTAL DESIGN AND ANALYSES 3.11-1 3.11.2.1 Environmental Design Criteria 3.11-1 3.11.2.2 Environmental Design Criteria for 3.11-2 ESF Equipment 3.11.2.3 Environmental Design of ESF Components 3.11-2 3.11.2.4 Environmental Design of 10 CFR 50.49 Scope 3.11-2 Equipment 3.11.3 LOSS OF VENTILATION 3.11-2 3.12 CONTROL OF HEAVY LOADS 3.12-1 3.
12.1 INTRODUCTION
/LICENSING BACKGROUND 3.12.1 3.12.2 SAFETY BASIS 3.12-1 3.12.3 SCOPE OF HEAVY LOAD HANDLING SYSTEMS 3.12-2 S3-0toc.doc 3-7
SQN-26 TABLE OF CONTENTS Section Title Page 3.12.4 CONTROL OF HEAVY LOADS PROGRAM 3.12-2 3.12.4.1 SQN Commitments in Response to 3.12-2 NUREG-0612, Section 5.1.1 3.12.4.2 Reactor Pressure Vessel Head (RPVH) 3.12-5 Lifting Procedures 3.12.5 SAFETY EVALUATION 3.12-7 3.13 FLEX RESPONSE SYSTEM 3.13-1 3.13.1 FLEX RESPONSE SYSTEM MITIGATION OF BEYOND 3.13-1 DESIGN BASIS EXTERNAL EVENTS 3.13.2 REFERENCE 3.13-1 S3-0toc.doc 3-8
SQN-18 LIST OF TABLES Number Title 3.2.1-1 Category I Structures 3.2.1-2 Summary of Criteria - Mechanical System Components (Excluding Piping) 3.2.1-3 Electrical Power System Equipment Designed to Operate During and After a "Safe Shutdown Earthquake" 3.2.2-1 Summary of Codes and Standards for Requirements for Sequoyah Mechanical System Components Excluding Piping 3.2.2-2 Summary of Codes and Standards for Requirements for Sequoyah Piping 3.2.2-3 Non-Nuclear Safety Classifications 3.5.1-1 Missiles 3.5.2-1 Missile Characteristics 3.5.5-1 Tornado Missile Spectrum for Category I Structures - Original Design 3.5.5-2 Sequoyah Nuclear Plant - Tornado Missile Spectrum A for Category I Structures 3.5.5-3 Tabulation of Walls and Roofs of Category I Structures Which are Less Than 2 Feet Thick 3.5.5-4 Sequoyah Nuclear Plant - Tornado Missile Spectrum B for ERCW Pumping Station 3.5.5-5 Sequoyah Nuclear Plant - Tornado Missile Spectrum D for Diesel Generator Equipment Doors 3.6.1-1 Piping Systems Inside Containment Where Energy Classifications Differ from Regulatory Guide 1.46 Definition 3.6.2-1 Postulated Design Basis Break Location for LOCA Analysis 3.6.7-1 Summary of Combined Stresses at Break Locations for Main Steam Lines 3.6.7-2 Checklist of Protection Provided Against Unacceptable Consequences of Main Steam and Feedwater Line Ruptures 3.6.7-3 Summary of Combined Stresses at Break Locations for Main Feedwater Lines S3-0toc.doc 3-9
SQN LIST OF TABLES Number Title 3.7.1-1 Category I Structures of the Original Plant Design 3.7.1-2 Periods for Spectral Values 3.7.1-3 Damping Ratios Used in Analysis of Category I Structures, Systems, Components, and Soil for Structures Listed in Table 3.7.1-1 3.7.1-3A Damping Ratios Used in Analysis of Category 1 Equipment, Components, and Their Supports for Structures Listed in Table 3.7.1-1 3.7.1-4 Soil Supported Category I Structures 3.7.1-5 Pile and Caisson Supported Category I Structures 3.7.2-1 Summary of Varied Parameter Ranges Used in the Seismic Analysis of the Ice Condenser Basket Support Frame 3.7.2-2 Category I Structures Affected by Concrete Modulus Change for Structures Listed in Table 3.7.1-1 3.7.2-3 Deleted per Amendment 6 3.7.2-4 Deleted per Amendment 6 3.7.2-5 Deleted per Amendment 6 3.7.2-6 Shield Building Element and Mass Point Properties 3.7.2-7 Shield Building Periods of Natural Modes of Vibration 3.7.2-8 Deleted per Amendment 6 3.7.2-9 Deleted per Amendment 6 3.7.2-10 Interior Concrete Structure Element Properties 3.7.2-11 Interior Concrete Structure Mass Point Properties 3.7.2-12 Interior Concrete Structure Periods for Natural Modes of Vibration 3.7.2-13 Steel Containment Vessel Element Properties 3.7.2-14 Steel Containment Vessel Mass Point Properties 3.7.2-15 Steel Containment Vessel Periods for Natural Modes of Vibrations S3-0toc.doc 3-10
SQN LIST OF TABLES Number Title 3.7.2-16 Auxiliary-Control Building Element Properties 3.7.2-17 Auxiliary-Control Building Mass Point Properties 3.7.2-18 Auxiliary-Control Building Periods for Natural Modes of Vibration 3.7.2-19 Element Properties Additional Equipment Building Unit 1 3.7.2-20 Element Properties Additional Equipment Building Unit 2 3.7.2-21 Mass Point Properties Additional Equipment Building Unit 1 3.7.2-22 Mass Point Properties Additional Equipment Building Unit 2 3.7.2-23 Normal Modes of Vibration Additional Equipment Building Unit 1 3.7.2-24 Additional Equipment Building - Unit 2 3.7.2-25 Pumping Station Element Properties 3.7.2-26 Pumping Station Mass Point Properties 3.7.2-27 Pumping Station Periods for Natural Modes of Vibration 3.7.2-28 ERCW Pumping Station Element Properties for Mathematical Model 3.7.2-29 ERCW Pumping Station Mass Point Properties for Mathematical Model 3.7.2-30 ERCW Pumping Station Normal Modes of Vibration 3.7.2-31 Diesel-Generator Building Element Properties 3.7.2-32 Diesel-Generator Building Mass Point Properties 3.7.2-33 Diesel-Generator Building Periods for Normal Modes of Vibration 3.7.2-34 Waste Packaging Area Element Properties 3.7.2-35 Waste Packaging Area Mass Point Properties 3.7.2-36 Waste Packaging Area Stiffness of Soil Springs 3.7.2-37 Condensate Demineralizer Waste Evaporator Building Stiffness for Soil Springs S3-0toc.doc 3-11
SQN-24 LIST OF TABLES Number Title 3.7.2-38 Condensate Demineralizer Waste Evaporator Building Element Properties 3.7.2-39 Condensate Demineralizer Waste Evaporator Building Mass Point Properties 3.7.2-40 Condensate Demineralizer Waste Evaporator Building Natural Periods of the Structural Model 3.7.2-41 East Steam Valve Room Element Properties 3.7.2-42 East Steam Valve Room Element Properties 3.7.2-43 East Steam Valve Room Mass Point Properties 3.7.2-44 East Steam Valve Room Spring Constants for Combined Caisson-Soil System 3.7.2-45 East Steam Valve Room Frequency Comparison for Three Soil Cases Horizontal Motion 3.8.1-1 Loading Combinations and Allowable Stresses for the Shield Building 3.8.1-2 Shield Building Equipment Hatch Doors and Sleeves Loads, Loading Combinations, and Allowable Stresses 3.8.2-1 Allowable Stress Criteria - Containment Vessel 3.8.2-2 Loading Combinations for Various Plant Conditions 3.8.3-1 Loading Combinations and Allowable Stresses for the Interior Concrete Structure 3.8.3-2 Loading Combinations and Load Factors 3.8.3-2A Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6) 3.8.3-3 Seals Between Upper and Lower Compartments 3.8.3-4 Personnel Access Doors in Crane Wall 3.8.3-5 Ice Condenser Allowable Limits 3.8.3-6 Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criteria 3.8.3-7 Equipment Access Hatch S3-0toc.doc 3-12
SQN LIST OF TABLES Number Title 3.8.3-8 Escape Hatch - Divider Barrier Floor - Load Combinations - Allowable Stresses 3.8.3-9 Air Return Duct Penetration 3.8.3-10 Maximum Stress - Summary (DBA x 1.2) Per Table 3.8.3-1 Criteria 3.8.3-11 Selection of Steels in Relation to Prevention of Non-Ductile Fracture of Ice Condenser Components 3.8.4-1 Auxiliary Control Building Concrete Structure Loads, Loading Combinations, and Allowable Stresses 3.8.4-2 Auxiliary Control Building Structural Steel Loads, Loading Conditions, and Allowable Stresses 3.8.4-3 Condenser Cooling Water Intake Pumping Station - Loading Cases, Allowable Stresses, Factors of Safety and Material Properties 3.8.4-4 Retaining Walls - Loading Cases, Allowable Stresses, Factors of Safety, and Material Properties 3.8.4-5 Control Room Shield Doors - Loads, Loading Combinations, and Allowable Stresses 3.8.4-6 Auxiliary Building Railroad Access Hatch Covers - Loads, Loading Combinations, and Allowable Stresses 3.8.4-7 Railroad Access Door - Loads, Loading Combinations, and Allowable Stresses 3.8.4-8 Manways in RHR Sump Valve Room - Loads, Loading Combinations, and Allowable Stresses 3.8.4-9 Pressure Confining Personnel Doors - Loads, Loading Combinations, and Allowable Stresses 3.8.4-10 Diesel Generator Building Doors and Bulkheads - Loads, Loading Combinations, and Allowable Stresses 3.8.4-11 Diesel Generator Building - Loads, Loading Combinations, Allowable Stresses and Material Properties 3.8.4-12 Primary and Refueling Water Pipe Tunnels - Loads, Loading Combinations, Allowable Stresses, and Material Properties S3-0toc.doc 3-13
SQN-25 LIST OF TABLES Number Title 3.8.4-13 Class 1E Electrical Systems Structures - Loads, Loading Combinations, Allowable Stresses, and Material Properties 3.8.4-14 East Steam Valve Room - Loads, Load Combinations, Allowable Stresses, Factors of Safety and Material Properties 3.8.4-15 East Steam Valve Room - Structural Steel - Loading Combinations and Allowable Stresses for Structural Steel 3.8.4-16 ERCW Pumping Station Loads and Loading Combinations 3.8.4-17 Refueling Water Storage Tank Foundation 3.8.4-18 Spent Fuel Pool Gates Loads and Loading Combinations 3.8.6-1 Polar Cranes - Loads, Loading Combinations, and Allowable Stresses 3.8.6-2 Deleted 3.9.2-1 Codes and Other Criteria Governing the Analysis of TVA Class B, C, and D Components 3.9.2-2 Design Loading Combinations for Group Classes B, C, and D Components 3.9.2-3 Safety Class B, C, and D Component Loading Conditions and Stress Limits 3.9.2-4 Loading Combinations and Stress Limits for Safety Class B, C, and D Piping 3.9.2-5 Loading Combinations and Stress/Loading Limits for Safety Class B, C, and D Supports 3.9.3-1 Maximum Deflections for Reactor Internals Under Blowdown and Seismic Excitation 3.10.2-1 Load Combinations and Allowables for Cable Tray Supports 3.11.1-1 Electrical and Mechanical Equipment Required to Function During and/or After an Accident S3-0toc.doc 3-14
SQN-31 LIST OF FIGURES Number Title 3.3.2-1 Variations of Differential Pressure and Tangential Plus Translational Velocity as a Function of the Distance from the Center of the Tornado 3.5.2-1 Ice Condenser Lower Inlet Door Opening, Typical Missile Trajectory Orientation 3.5.2-2 Location of Main Steam and Main Feedwater Containment Isolation Valves at Azimuth 0 3.5.2-3 Location of Main Steam and Main Feedwater Containment Isolation Valves at Azimuth 180 3.5.4-1 Comparison of Missile Formulas 3.5.4-2 Comparison of Missile Formulas 3.5.4-3 Depth of Missile Penetration for Tornado 3.5.4-4 Depth of Missile Penetration for Tornado 3.6.1-1 High Energy Classification Per Sequoyah Nuclear Plant Pipe Rupture Analysis Definition and Per Regulatory Guide 1.46 Definition 3.6.2-1 Location of Postulated Breaks 3.6.4-2 Jet Expansion Models 3.6.7-1 Steam Generators 1 and 4 Postulated Break Locations and Fixes 3.6.7-2 Steam Generators 2 and 3 Postulated Break Locations and Fixes 3.7.1-1 Mathematical Model for Soil-Structure Interaction 3.7.1-2 ERCW Access Dike 3.7.2-1 Deleted 3.7.2-2 Deleted 3.7.2-3 Model of Horizontal Lattice Frame Structure 3.7.2-4 Group of Six Interconnected Lattice Frames S3-0toc.doc 3-15
SQN LIST OF FIGURES Number Title 3.7.2-5 Typical Model of Lattice Frame 3.7.2-6 Typical Multi-Level Horizontal Dynamic Model of Lattice Frame Basket Assembly 3.7.2-7 Lattice Frame Ice Basket Gap 3.7.2-8 Typical Displacement Time Histories for 12 ft. Basket With End Supports - Pluck Test 3.7.2-9 Typical Crane Wall Acceleration 3.7.2-10 Typical Crane Wall Velocity 3.7.2-11 Typical Crane Wall Displacement 3.7.2-12 Typical Ice Basket Displacement Response 3.7.2-13 Typical Ice Basket Impact Force Response 3.7.2-14 Typical Crane Wall Panel Load Response 3.7.2-15 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-16 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-17 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-18 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-19 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-20 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-21 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-22 Nonlinear Dynamic Model Results Wall Panel Seismic Load Versus Gap Size 3.7.2-23 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size S3-0toc.doc 3-16
SQN LIST OF FIGURES Number Title 3.7.2-24 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-25 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-26 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-27 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-28 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-29 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-30 Nonlinear Dynamic Model Results Seismic Impact Load Versus Gap Size 3.7.2-31 Original Accelerogram 3.7.2-32 Accelerogram After Integration and Differentiation 3.7.2-33 Nonlinear Dynamic Model 3.7.2-34 Section Through Reactor Shield Building Looking West, Lumped Mass Model for Dynamic Analysis 3.7.2-35 Flowchart of Operations for Response of the Dome 3.7.2-36 Shell Model for Dome Analysis - Shield Building 3.7.2-37 Lumped Mass Model for Dome Analysis - Shield Building 3.7.2-38 Reactor Building, Interior Concrete Structure Sectional Elevation Looking West, Lumped Mass Model for Dynamic Analysis 3.7.2-39 Steel Containment Vessel, Lumped Mass Model for Dynamic Analysis 3.7.2-40 Steel Containment Vessel, Finite Element Model 3.7.2-41 Sectional Elevation of Auxiliary Control Building Lumped Mass Model for Dynamic Analysis S3-0toc.doc 3-17
SQN LIST OF FIGURES Number Title 3.7.2-42 Additional Equipment Building - Unit 1 3.7.2-43 Additional Equipment Building - Unit 2 3.7.2-44 Sectional Elevation of Intake Pumping Station, Lumped Mass Model for Dynamic Analysis 3.7.2-45 ERCW Pumping Station - Model for Dynamic Analysis Sectional Elevation 3.7.2-46 ERCW Pumping Station - Key Plan 3.7.2-47 Sectional Elevation of Diesel Generator Building, Lumped Mass Model for Dynamic Analysis 3.7.2-48 Sectional Elevation of Waste Packaging Area, Lumped Mass Model for Dynamic Analysis 3.7.2-49 Response Acceleration Spectra 3.7.2-50 Condensate Demineralizer Waste Evaporator Building - Mathematical Model for Dynamic Analysis 3.7.2-51 Averaged Ground Response Spectrum 3.7.2-52 Concrete Caisson 3.7.2-53 East Valve Room - Lumped Mass and Spring Model 3.7.2-54 Interior Concrete Structure N-S Translational Motion, Translational Mode 1 3.7.2-55 Interior Concrete Structure N-S Translational Motion, Translational Mode 2 3.7.2-56 Deleted by Amendment 6 3.7.2-57 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 1 3.7.2-58 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 2 3.7.2-59 Interior Concrete Structure E-W Translation Plus Torsion, Translational Mode 3 S3-0toc.doc 3-18
SQN LIST OF FIGURES Number Title 3.7.2-60 Interior Concrete Structure E-W Translational Plus Torsion, Translational Mode 4 3.7.2-61 Interior Concrete Structure E-W Translational Plus Torsion, Translational Mode 5 3.7.2-62 Interior Concrete N-S Translation Response, Maximum Acceleration, Safe Shutdown Earthquake 3.7.2-63 Interior Concrete N-S Translation Response, Maximum Deflection, Safe Shutdown Earthquake 3.7.2-64 Interior Concrete N-S Translation Response, Maximum Shear, Safe Shutdown Earthquake 3.7.2-65 Interior Concrete N-S Translation Response, Maximum Bending Moment, Safe Shutdown Earthquake 3.7.2-66 Interior Concrete E-W Translation Plus Torsion Response, Maximum Acceleration, Safe Shutdown Earthquake 3.7.2-67 Interior Concrete E-W Translation Plus Torsion Response, Maximum Deflection, Safe Shutdown Earthquake 3.7.2-68 Interior Concrete E-W Translation Plus Torsion Response, Maximum Shear, Safe Shutdown Earthquake 3.7.2-69 Interior Concrete E-W Translation Plus Torsion Response, Maximum Bending Moment, Safe Shutdown Earthquake 3.7.2-70 Interior Concrete E-W Translation Plus Torsion Response, Maximum Angular Acceleration, Safe Shutdown Earthquake 3.7.2-71 Interior Concrete E-W Translation Plus Torsion Response, Maximum Angular Displacement, Safe Shutdown Earthquake 3.7.2-72 Interior Concrete E-W Translation Plus Torsion, Response Maximum Torque, Safe Shutdown Earthquake 3.7.2-73 Response Acceleration Spectrum, Reactor Building Interior Concrete, NS Mass Point No. 1, Safe Shutdown Earthquake 3.7.2-74 Response Acceleration Spectrum, Reactor Building Interior Concrete, EW Mass Point No. 1, Safe Shutdown Earthquake 3.7.2-75 Response Acceleration Spectrum Reactor Building Interior Concrete, NS Mass Point No. 6, Safe Shutdown Earthquake S3-0toc.doc 3-19
SQN-18 LIST OF FIGURES Number Title 3.7.2-76 Response Acceleration Spectrum Reactor Building Interior Concrete, EW Mass Point No. 6, Safe Shutdown Earthquake 3.7.2-77 Flow Chart for Development of Floor Response Spectra 3.7.2-78 Comparison of Time History and Response Spectrum Response 3.7.2-79 Replacement Steam Generator One-Stick Model 3.8.1-1 Temperature Gradient 3.8.2-1 Structural Steel Containment Vessel 3.8.2-2 Layout of Containment Shell 3.8.2-3 Containment Panel Analysis - Finite Element Model 3.8.2-4 Pressure - Time Function of Local Detonation 3.8.2-5 Graphical Representation of Pressure Function 3.8.2-6 Hydrogen Explosion - Containment Panel Analysis - Finite Element Model of Shell Segment 3.8.2-7 Structural Steel Containment Vessel Anchor Bolt Plan and Base Details 3.8.2-8 Steel Containment Vessel, Lumped Mass Model for Dynamic Analysis 3.8.2-9 Steel Containment Vessel, Finite Element Model 3.8.2-10 Steel Containment Vessel, Finite Element Model 3.8.2-11 Expansion Bellows For Personnel Locks 3.8.3-1 Seals Between Ice Condenser and Containment Vessel Arrangement 3.8.3-2 Temperature Gradient 3.8.4-1 Concrete Floor Design Data 3.8.4-2 D. G. Building Doors and Bulkhead Arrangement 3.8.4-3 Concrete-Pipe Tunnels and Tank FDNS Outline 3.8.4-4 Concrete-Pipe Tunnels and Tank FDNS Outline S3-0toc.doc 3-20
SQN LIST OF FIGURES Number Title 3.8.4-5 Category 1 Yard Electrical Manholes and Handholes 3.8.4-6 Concrete-Manholes and Handholes - Outline 3.8.4-7 Concrete - ERCW Pumping Station and ERCW Channel 3.8.4-8 Concrete - ERCW Pumping Station 3.8.4-9 Concrete ERCW Skimmer Wall and Underwater Dam 3.8.4-10 Concrete ERCW Discharge Box 3.8.4-11 Spent Fuel Pool Gate 3.8.6-1 RB 175 Ton Polar Cranes 3.8.6-2 RB 175 Ton Polar Cranes, Trolley 3.8.6-3 RB 175 Ton Polar Cranes, Trolley 3.8.6-4 RB 175 Ton Polar Cranes, Bridge 3.8.6-5 RB 175 Ton Polar Cranes 3.8.6-6 AB 125 Ton Crane 3.8.6-7 AB 125 Ton Crane Trolley 3.8.6-8 AB 125 Ton Crane Trolley 3.8.6-9 AB 125 Ton Crane Bridge 3.8.6-10 AB 125 Ton Crane Limit Switch 3.8.6-11 AB 125 Ton Crane Mechanical Stop 3.8.6-12 Sectional Elevation View Through the Spent Fuel Pit 3.9.1-1 Upper Internals Assembly 3.9.1-2 Upper Internal Support Model 3.9.1-3 Computer Geometry Plot of Lower Internals Support Model 3.9.1-4 Lower Internals Support Structure Comparison Between Experimental and Theoretical Vertical Deflections S3-0toc.doc 3-21
SQN LIST OF FIGURES Number Title 3.9.1-5 Reactor Vessel and Internals Vibration 3.9.1-6 Thermal Shield, Mode Shape n=4 Obtained from Shaker Test 3.9.1-7 Thermal Shield, Maximum Amplitude of Vibration During Preoperational Tests 3.9.1-8 Time-History Dynamic Solution for LOCA Loading 3.9.2-1 Reactor Coolant Piping Pressurizer Surge Line S3-0toc.doc 3-22
SQN 3.0 DESIGN CRITERIA STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1.1 Single Failure Criteria Each of the Engineered Safety Features is designed to tolerate a single failure during the period of recovery following an incident without loss of its protection functions. This period of recovery consists of two segments; the short-term period and the long-term period. During the short-term period, the single failure is limited to a failure of an active component to complete its function as required. Should the single failure occur during the long-term period rather than the short term, the Engineered Safety-Related System is designed to tolerate an active failure or a passive failure without loss of its protective function. The following definitions are applicable to terms that pertain to the single failure criterion: Period of Recovery. The time necessary to bring the plant to safe shutdown and regain access to faulted equipment. The recovery period is the sum of the short-term and long-term periods defined below. Note that safe shutdown is used here in lieu of cold shutdown. SQN is a hot standby plant, not a cold shutdown plant. (
Reference:
NUREG 0011 & supplement 1; and NUREG 1232, Volume 2, pg. 2-7 ). Incident. Any natural or accidental event of infrequent occurrence and its related consequences which affect the plant operation and require the use of engineered safeguards systems. Such events, which are analyzed independently and are not assumed to occur simultaneously, include the loss-of-coolant accident, steam line ruptures, steam generator tube ruptures, etc. A loss of offsite power may be an isolated occurrence or may be concurrent with any event requiring engineered safeguards systems use. Short Term. The time immediately following the incident during which automatic actions are performed, system responses are checked, type of incident is identified and preparations for long-term recovery operation are made. The short term is the first 24 hours following initiation of engineered safeguards system operations. Long Term. The remainder of the recovery period following the short term. In comparison with the short term where the main concern is to remain within Nuclear Regulatory Commission specified site criteria, the long-term period of operation involves bringing the plant to safe shutdown conditions where faulted equipment can be accessed and repaired. Note that safe shutdown is used here in lieu of cold shutdown. SQN is a hot standby plant, not a cold shutdown plant. (
Reference:
NUREG 0011 & supplement 1; and NUREG 1232 volume 2, pg. 2-7 ). Active Failure. The failure of a powered component such as a piece of mechanical equipment, component of the electrical supply system or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a motor-operated valve to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan or diesel generator to start, etc. S3-01.doc 3.1-1
SQN Equipment moving spuriously from the proper safeguards position without signal, such as a motor-operated valve inadvertently shutting at the moment it is required, is not considered credible. Passive Failure. The structural failure of a static component which limits the components effectiveness in carrying out its design function. When applied to a fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gal/min for 30 minutes. Such leak rates are consistent with limited cracks in pipes, sprung flanges, valve packing leaks or pump seal failures. 3.1.2 Overall Requirements The Sequoyah Nuclear Plant was designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits published in July, 1967. The Sequoyah construction permit was issued in May, 1970. This FSAR, however, addresses the NRC General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971. Each criterion is followed by a discussion of the design features and procedures which meet the intent of the criteria. Any exception to the 1971 GDC resulting from the earlier commitments is identified in the discussion of the corresponding criterion. References to other sections of the FSAR are given for system design details. Criterion 1 - Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A Quality Assurance Program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit. Compliance Discussions related to the applicable codes, design criteria and standards used in the design of particular systems are contained in the appropriate FSAR sections and in Tables 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.2-1 and 3.2.2-2. The Quality Assurance Program conforms to the requirements of 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plant." Details of the program are given in Chapter 17 and the TVA Nuclear Quality Assurance Plan. Criterion 2 - Design Bases for Protection Against Natural Phenomena. Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and S3-01.doc 3.1-2
SQN seiches without loss of capability to perform their safety function. The design bases for these structures, systems, and components shall reflect:
1. Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated,

2. Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena, and

3. The importance of the safety functions to be performed.
Compliance The structures, systems, and components important to safety are designed to either withstand the effects of natural phenomena without loss of capability to perform their safety functions, or to fail in the safest condition. Those structures, systems, and components vital to the shutdown capability of the reactor are designed to withstand the maximum probable natural phenomenon expected at the site, determined from recorded data for the site vicinity, with appropriate margin to account for uncertainties in historical data. Appropriate combinations of normal, accident, and natural phenomena structural loadings are considered in the plant design. The nature and magnitudes of the natural phenomena considered in the design of this plant are discussed in subsections 2.3, 2.4, and 2.5. Subsections 3.2 through 3.11 discuss the design of the plant in relationship to natural events. Seismic and safety classifications, as well as other pertinent standards and information, are given in the sections listed above and those sections discuss individual structures and components. Criterion 3 - Fire Protection. Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat-resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire-fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components. Compliance The plant is designed to minimize the probability of fires and explosions, and in the event of such occurrences to minimize the potential effects of such events to plant safety-related equipment and personnel. Prime consideration is given to these requirements throughout the design process by providing for the duplication and physical separation of components in plant design and the use of materials classified as noncombustible and/or fire resistant wherever practical in all areas of the plant. Equipment and facilities for fire protection, including detection, alarm, and S3-01.doc 3.1-3
SQN extinguishment are provided to protect both plant equipment and personnel from fire, explosion, and the resultant release of toxic vapors. Fire-fighting systems are designed to assure that their rupture or inadvertent operation will not impair systems important to safety. All portions of the Fire Protection Systems necessary to protect safety-related equipment in Class I structures are designed to seismic requirements. All systems are designed and installed in accordance with the applicable requirements as described in the Fire Protection Report (see 9.5.1). The Fire Protection System is designed such that a failure of any component of the system or inadvertent operation:
1. Will not cause a nuclear accident or significant release of radioactivity to the environment.

2. Will not impair the ability of equipment to safely shut down and isolate the reactor or limit the release of radioactivity to the environment in the event of a postulated accident.
The Fire Protection Systems for the Sequoyah Nuclear Plant are discussed in the Fire Protection Report. Criterion 4 - Environmental and Missile Design Bases. Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCA. These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. Compliance Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. The associated environmental parameters are identified and incorporated in the design requirements and specifications. Particular care is given to the extreme environmental conditions associated with major incidents such as a loss of coolant. Required equipment and instrumentation are identified, environmental conditions such as temperature, pressure, humidity, irradiation, etc., are calculated, and the effects of the latter on the former are evaluated either analytically or experimentally. The dynamic effects associated with an accident are carefully identified and assurance is given that the structures and systems (including engineered safeguards) assumed undamaged in the total assessment of the accident consequences are suitably protected. Emergency core cooling components are austenitic stainless steel or equivalent corrosion-resistant material and hence are compatible with the containment atmosphere over the full range of exposure during the postaccident conditions. Where vital components cannot be located away from potential missiles, protective walls and slabs, local missile shielding, and restraining devices are provided to protect the containment and engineered safety feature components within the containment against damage from missiles generated by the equipment failures associated with the DBA. S3-01.doc 3.1-4
SQN The environmental design of safety-related items is discussed in subsection 3.8 on the design of structures, subsections 6.2.2 and 6.2.3 on containment heat removal and air purification and subsection 9.4 on ventilation systems. Safety-related systems and components use the input from these sections for design as discussed in subsection 3.11. The missile and environmental protection given each system is discussed with the individual system in Chapters 3 through 11. Criterion 5 - Sharing of Structures, Systems, and Components. Structures, systems, and components important to safety shall not be shared between nuclear power units unless it is shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. Compliance The two units share several structures and systems, many of which have no safety function. The structures important to safety are the Auxiliary/Control Building (subsection 3.8), Diesel Generator Building (subsection 3.8), CCW Pumping Station (subsection 3.8), the ERCW pumping station (subsection 3.8), and a few miscellaneous structures. Shared safety-related systems include the ERCW (subsection 9.2), component cooling water (subsection 9.2), fire protection (subsection 9.5), fuel handling/storage and cooling (subsection 9.1), fuel oil storage (subsection 9.5), preferred and emergency electric power (subsections 8.2 and 8.3, respectively), chemical and volume control (subsection 9.3), condensate (subsection 9.2), radioactive waste (Chapter 11), Gas Treatment System ( subsection 6.2 ), and Control and Auxiliary Building Ventilation Systems (subsections 6.4 and 9.4). The Vital Direct-Current Power System is shared to the extent that a few loads (e.g., the vital inverters) in one nuclear unit are energized by the direct-current power channels assigned primarily to power loads of the other unit. In no case does the sharing inhibit the safe shutdown of one unit while the other unit is experiencing an accident. All shared systems are sized for all credible initial combinations of normal and accident states for the two units, with appropriate isolation to prevent an accident condition in one unit from carrying into the other. Criterion 10 - Reactor Design. The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. Compliance The reactor core with its related coolant, control, and protection systems is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The Reactor Trip System is designed to actuate a reactor trip, when necessary, for any anticipated combination of plant conditions, to ensure that fuel design limits are not exceeded. The core design, together with reliable process and decay heat removal systems, provides for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations, including the effects of the loss of reactor coolant flow, trip of the turbine-generator, loss of normal feedwater, and loss of power. S3-01.doc 3.1-5
SQN Chapter 4 discusses the design bases and design evaluation of reactor components. Chapter 5 discusses the Reactor Coolant System. The details of the Reactor Trip and Engineered Safety Features Actuation Systems design and logic are discussed in Chapter 7. This information supports the accident analyses presented in Chapter 15. Criterion 11 - Reactor Inherent Protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity. Compliance A negative reactivity coefficient is a basic feature of core nuclear design as discussed in Chapter 4. Criterion 12 - Suppression of Reactor Power Oscillations. The reactor core and associated coolant, or, control and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed. Compliance Power oscillations of the fundamental mode are inherently eliminated by the negative Doppler and nonpositive moderator temperature coefficient of reactivity. Oscillations, due to xenon spatial effects, in the radial, diametral and azimuthal overtone modes are heavily damped due to the inherent design and due to the negative Doppler and nonpositive moderator temperature coefficients of reactivity. Oscillations, due to xenon spatial effects, in the axial first overtone mode may occur. Assurance that fuel design limits are not exceeded by xenon axial oscillations is provided as a result of reactor trip functions using the measured axial power imbalance as an input. Oscillations, due to xenon spatial effects, in axial modes higher than the first overtone, are heavily damped due to the inherent design and due to the negativ}}

ML23349A014

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1.0 INTRODUCTION

1.1 INTRODUCTION

2.6 CONCLUSION

SUMMARY

SUMMARY

7.1 INTRODUCTION

8.1 INTRODUCTION

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1.0 INTRODUCTION

1.1 INTRODUCTION

5.6 REFERENCES

1.0 INTRODUCTION

1.1 INTRODUCTION

2.2 DESCRIPTION

2.5 REFERENCES

3.6 REFERENCES

5.7 REFERENCES

2.6 CONCLUSION

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SUMMARY

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2.6 CONCLUSION

2.3 REFERENCES

3.3 REFERENCES

5.6 REFERENCES

7.6 REFERENCES

9.5 REFERENCES

10.3 REFERENCES

12.1 INTRODUCTION

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