ML20004E426
| ML20004E426 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/11/1981 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML20004E423 | List: |
| References | |
| NUDOCS 8106120168 | |
| Download: ML20004E426 (103) | |
Text
__ _ _. _ -
SUMMARY
OF CHANGES IN AMENDMENT NO. 23
- 1. A List of Effective Pages is provided.
- 2. Pages from Section No. 8.0 have been reissued since they were forwarded with Amendment No. 22 marked improperly as Amendment No. 6.
- 3. Information is provided concerning Reactor Coolant Sampling and Reactor Building Atmospheric Sampling.
J
- 4. Errors on the containment isolation tables are corrected.
- 5. Certain commitment dates have been deleted since the commitment dat?s have 3
been provided in response to NUREG-0737 by letter dated January 23, 1981 (TLL-680).
- 6. Additional information justifying the Pressurizer Safety Valve position moni-tors is provided.
- 7. Certain revised information is provided concerning the GPU Nuclear organiza-tion.
- 8. Operator Guidelines for steam generator filling have been added to the re-
- sponse to Question No. 55 of Supplement No. 1, Part No. 1.
O 9. Revised analytical assumptions are added for Appendix SA analyses.
I I
i 810612 0 l6%$
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. ~ . . . .. - - - _ - _ . - . . _ _ . . . _ .__ - . =.
I g- TABLE OF EFFECTIVE PAGES.
s TMI-1 RESTART REPORT SECTION NO. 1: SECTION NO. 2: (Cont'd.)
Page No. Am. No. Page No. Am. No.
1-1 21 2.1-31 23 1-2 21 2.1-32 21 2.1-33 21 2.1-34 21 SECTION NO. 2: 2.1-35 21 2.1-36 21 Page No. Am. No. 2.1-37 21
- 2.1-38 22 2.1-1 3 2.1-38a 22 2.1-2 23 2.1-38b 22 2.1-3 18 2.1-38c 22 2.1-4 18 2.1-38d 22 i 2.1-5 23 2.1-38e 22
- 2.1-6 22 2.1-38f 22 2.1-7 22 2.1-38g 22 2.1-7a 22 2.1-38h 22 2.1-7b 22 2.1-38i 22
.l
() 2.1-7c 2.1-8 22 21 2.1-39 2.1-40 23 23 2.1-9 21 2.1-41 23 2.1-10 21 2.1-42 23 2.1-11 21 2.1-43 23 2.1-12 21 2.1-44 23 2.1-13 21 2.1-45 23 2.1-14 21 2.1-46 23 2.1-15 21 2.1-47 23 l 2.1-16 21 2.1-48' 23 l 2.1-16a 21 SECTION NO 2 - TABLES:
2.1-17 22 2.1-18 22 Table No. Am. No.
2.1-19 5 .
2.1-20 22 2.1-1 23 2.1-21 22 2.1-2 23 2.1-22 22 2.1-3 23 2.1-23 22 2.1-4 13 2.1-24 13 2.1-5 12 2.1-25 22 2.1-6 13 2.1-26 22 2.1-7 22 2.1-27 22 2.1-28 22 2.1-29 22 2.1-29a 22
) 2.1-29b 22
('~/
\- 2.1-29c 22 j 2.1-30 23
- ,, 'Iment numbers not noted on pages. Am. 23 l _
s SECTION NO. 2 - FIGURES: SECTION NO. 3 - TABLES:
Figure No. Am. No. Table No. Am. No.
2.1-1 5 3.1-1 5 (Page 1) 2.1-2 5 15 (Page 2) 2.1-3 5 3.1-2 10 (10 Pages) 2.1-4 22 2.1-5 6 2.1-6 21 2.1-7 21 SECTION NO. 4:
2.1-8 5 2.1-9 5 Page No. Am. No.
2.1-10 5 2.1-11 21 i 19-2.1-12 22 11 19 2.1-13 22 111 19 2.1-14 23 iv 19 2.1-15 23 y 19 2.1-16 23 vi 19 4-1 19 4.1-1 19 4.1-2 19 SECTION NO. 2 - APPENDIX 2A: 4.1-3 19 4.1-4 19 Description Am. No. 4.1-5 19 4.2-1 19
- 1. BAW-1603 23- 4.2-2 19
- 2. TDR-240, Rev. 0 23 4.2-3 19 4.2-4 19 4.2-5 19 4.2-6 19 SECTION NO. 2 - APPENDIX 2B: 4.2-7 19 4.3-1 19 Description Am. No. 4.3-2 19 4.4-1 19 Cylindrical Source 15 4.4-2 19 Shielding Equations 4.4-3 19 Using Compact Func- 4.4-4 19 tions and Including 4.4-5 19 Buildup 4.4-6 19 4.4-7 19 4.4-8 19 4.4-9 19
~SECTION NO. 3: 4.4-10 19 4.5-1 19 Page No. Am. No. 4.5-2 19 4.5-3 19 3-1 5 4.5-4 19
. 3-2 5 4.5-5 19 3-3 5 4.5-6 19 4.5-7 19 4.5-8 19
- Amendment numbers not noted on page.
Am. 23
SECTION NO. 4: (Cont'd.) SECTION NO. 4: (Cont'd.)
Page No. Am. No. Page No. Am. No.
4.5-9 19 4,7-8 19 4.5-10 19 4.7-9 19 4.5-11 19 4.7-10 19 4.5-12 19 4.7 11 19 4.5-13 19 4.7 12 19 4.5-14 19 4.7-13 19 4.5-15 19 4.7 14 19 4.5-16 19 4.7 15 19 4.5-16a 19 4.7-16 19 4.5-17 19 4.7-17 19 4.5-18 19 4.7-18 19 4.5-19 19 4.7-19 19 4.5-20 19 4.7-20 19 4.5-21 19 4.7 21 19 4.5-22 19 4.7,22 19 4.5-23 19 4.7-23 19 4 5-24 19 4.8-1 19 4.5-25 19 4.82 19 4.6-1 19 4.8-3 19 4.6-2 19 4.8-4 19 4.6-3 19 4.8-5
() 4,6-4 4.6-5 19 19 4.8 6 4.8,7 19 19 19 4.6-6 19 4.8-8 19 4.6-7 19 4.8-9 19 4.6-8 19 4.8-10 19 4.6-9 19 4.9-1 19 4.6-10 19 4.10,1 19 4.6-11 19 4.10,2 19 4.6-12 19 4.10-3 19 4.6-13 19 4.6-14 19 4.6-15 19 SECTION NO. 4 - FIGURES:
4.6-16 19 4.6-17 19 Figure No. Am. No.
4.6-18 19 4.6-19 19 1 19 4.6-20 19 2 19 4.6-21 19 3 19 4.6-22 19
- 4 19 4.7-1 19 5 19 4.7-2 19 6 19 4.7-3 19 7 19 4.7-4 19 8 19 4.7-5 19 9 19
- 4. 7-6 19 10 19
() 4.7-7 19
- Amendment number not noted on page.
Am. 23
SECTION NO. 4 - FIGURES: (Cont'd.) SECTION NO. 4 - APPENDIXES: (Cont'd.)
Figure No. Am. No. Am. No.
11 19 Appendix C 19 (31 pgs.)
12 19 Appendixes D - I 19 (1 pg.)
13 19 14 19 15 19 SECTION NO. 5:
16 19 17 19 Page No. Am. No.
18 19 19 19 5.1-1 20 20 19 5.1-2 20 21 19 5.1-3 20 22 19 5.2-1 20 23 19 5.2-2 20 24 19 5.2-3 20 l 25 19 5.2-4 20 4
26 19 5.2-5 20
, 27 19 5.2-6 23 5.2-7 20 5.2-8 23 SECTION NO. 4 - TABLES: 5.2-9 20 5.2-10 20 O. Table No. Am. No. 5.2-11 20 5.2-12 20 1 19 5.2-13 20 2 19 5.2-14 20 3 19 5.2-15 20 4 19 5.2-16 20 5 19 5.2-17 20 6 19 5.2-18 20 7 19 5.2-19 20 8 19 5.2 20 20 9 19 5.2-21 20 10 19 5.2-22 20 11 19 5.2-23 20 12 19 5.2-24 20 13 19 5.2-25 20 14 19 5.2-26 20 15 19 5.2-27 20 16 19 5.2-28 20 17 19 5.2-29 20 18 19 5.2-30 20 19 19 5.2 31 20 20 19 5.2-32 20 5.2-33 20 5.2-34 20 SECTION NO. 4 - APPENDIXES: 5.2-35 20
() Am. No.
5.2-36 5.2-37 20 20 5.2-38 20 Appendix A 19 5.2-39 20 Appendix B 19 (25 pgs.)
- Amendment number not noted on page. Am. 23
() SECTION NO. 5: (Cont'd.) SECTION NO. 5: (Cont'd,)
Page No. Am. No. Page No. Am. No.
5.2-40 20 5.3 1 20 5.2-41 20 5.3 2 20 5.2-42 20 5.3-3 20 5.2-43 20 5.3-4 20 5.2-44 20 5.3-5 20 5.2-45 20 5.3-6 20 5.2-46 20 5.3-7 20 5.2-47 20 5.3-8 20 5.2-48' 20 5.3-9 20 5.2-49 20 5.3-10 20 5.2-50 20 5.3-11 20 5.2-51 20 5.3-12 20 5.2-52 20 5.3-13 20 5.2-53 20 5.3-14 20 5.2-54 20 5.3-15 20 5.2-55 20 5.3-16 20 5.2-56 20 5.3-17 20 5.2-57 20 5.3-18 20 5.2-58 20 5.3-19 22 5.2-59 20 5.3-20 22 5.2-60 20 5.3-21 22
() 5.2-61 5.2-62 20 20 5.3-22 5.3-23 22 22 5.2-63' 20 5.3-24 .22 5.2-64 20 5.3-25 ~22 5.2-65 20 5.3-26 22 5.2-66 20 5.3-27 20 5.2-67 20 5.3-28 20 5.2-68 20 5.3-29 20 l
5.2-69 20 5.3-30 20 5.2-70 20 5.2t 71 20 5.2-72 20 SECTION NO. 5 - FIGURES:
5.2-73 20 5.2-74 20 Figure No. Am. No.
5.3-1 23 SECTION NO. 5 - FIGURES:
Figure No. Am. No. SECTION NO. 5: (Cont'd.)
5.2-1 20 Page No. Am. No.
5.4-1 20 5.4-2 20 5.4-3 20 5.4-4 20 0- 5.4-5 20
- Amendment number not noted on page.
Am. 23 l
(
() -SECTION NO. b: SECTION NO. 7 - FIGURES:
Page No. Am. No. Figure No. Am. No.
6-1 21 7-1 4 6-2 21 7-2 4 6-3 21 7-3 4 6-4 21 7-4 4 6-5 21 6-6 21 6-7 21 SECTION NO. 7 - APPENDIXES:
b-8 21 6-9 21 Am. No.
6-10 21 6-11 21 Appendix 7A 17 (25 pgs.)
6-12 21 Appendix 7B 11 (6 pgs.)
6-13 21 6-14 21 6-15 21 SECTION NO. 8:
6 21 6-17 21 Page No. Am. No.
8-1 22 SECTION NO. 7: 8-2 22 8-3 11 i
j Page No. Am. No. 3-4 11
{/
s- 8-5 11 i 7-1 4 8-6 6 7-2 4 8-7 16 7-3 7 8-8 6 7-4 18 8-9 16 7-5 18 8-10 16 7-6 18 8-11 7 7-7 18 8-12 22 7-7a 18 8-13 22 7-8 4 8-14 22 7-9 23 8-15 22 7-9a 21 8-16 22 7-9b 21 8-17 22 7-10 5 8-18 22 7-11 4 8-19 22 7-11a 18 8-19a 23
- 7-11b 8 8-20 16 l 7-12 13 8-21 6 7-13 4 7-14 23 7-15 16 I
l SECTION NO. 7 - TABLES:
/'s
- (,_) Tabic No. _
Am. No.
7.3-1 5 (3 pgs.)
- Amendment number not noted on page.
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- . , _ . _ _ _ . _ . _ . _ . _ _ _ . ~ - - _ _ _ _ . _ . _ . _ . . , _ . . _ . , _ . . _ . _ _ _ _ _ _ _ _ . _ _ _ . _ . _
SECTION NO. 8 - TABLES: SECTION NO. 8_y APPENDIX 8A- FIGURES:
(.)
Table No Am. No. Figure No. Am. No.
8-1 11 8A-1 12 8-2 6 8A-la 13 8A-2 6 SECTION NO. 8 - APPENDIXES: 8A-2a 11 8A-3 Am. No. 8A-4 22 8A-5 6 Appendix 8A (Title Sheet) 6 8A-6 6 Appendix 8A-1 10 8A-7 6 Appendix SA-2 10 8A-8 10 8A-9 10 SECTION NO. 8 - APPENDIX 8A - TABLES: 8A-10 10 8A-11 10 Table No. Am. No. 8A-12 12 8A-13 (Missing) 8A-1 23 8A-14 12 8A-2 23 8A-15 12 8A-3 23 8A-16 15 8A-4 23 8A-17 18 8A-5 23 8A-18 22 8A-6 23 8A-19 22 8A-7 23 8A-20 22 s,j 8A-8 23 8A-21 22 8A-9 23 8A-22 22 SA-10 23 8A-23 22 i 8A-11 23 8A-24 22 8A-12 23 8A-13 23 SECTION NO. 9,:
8A-14 23 8A-15 23 Page No. Am. No.
8A-16 23 8A-17 23 i 21 8A-18 23 ii 21 8A-19 23 iii 21 8A-20 23 iv 21 8A-21 23 y 21 8A-22 23 9-1 4 8A-23 23 l 8A-24 23 SECTION NO. 9 - TABLES:
( 8A-25 23 8A-26 23 Table No. Am. No.
8A-27 23 l 8A-78 23 9.2-1 11 l 8A-29 23 8A-30 23 SECTION NO. 10:
8A-31 23 Page No. Am. No.
- O 10-1 10 l
10-2 10 i.
(
- Amendment number nc . noted on page Am. 23
~ ,.
m SECTION NO. 10: SECTION NO. 11.2.1:
Page No. Am. No. Page No. Am. No.
10-3 18 (5 pgs.)
- 10-4 3 10-5 10 SECTION NO. 11.2.2:
10-6 22 10-7 15 Page No. Am. No.
~10-8 15 10-9 18 (9 pgs.) 21 10-10 18 10-11 18 SECTION NO. 11.2.3:
SECTION NO. 10 - APPENDIXES: Page No. Am. No.
Am. No. (6 pgs.) 21 Appendix No. 10A: SECTION NO. 11.2.4:
- 1) Reactor Trip No. 11
- Page No. Am. No.
- 2) Reactor Trip No. 12 *
(4 pgs.)
- SECTION NO. 11:
SECTION NO. 11.2.5:
() Page No. Am. No.
Page No. Am. No.
11-1 21 11-2 21 (4 pgs.) 21 11-3 21 11-4 21 SECTION NO. 11.2.6:
11-5 21 11-6 21 Page No. Am. No.
11-7 21 11-8 21 (5 pgs.) 21 11-9 21 11-10 21 SECTION NO. 11.2.7:
11-11 21 11-12 21 Page No. Am. No.
11-13 21 11-14 21 (15 pgs.) 21 11-15 21 11-16 21 SECTION NO. 11.2.8:
i 11-17 21 d
11-18 21 Page No. Am. No.
11-19 21 11-20 21 (5 pgs.) 16 11-21 21 11-22 21 11-23 21 11-24 21
\- 11-25 21 11-26 21
- Amendment number not noted on page.
Am. 23
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f i
t l0 1 SECTION NO. 11.2.9:
l Page No. Am. No.
(3 pgs.) 21 i i SECTIO.': ':0. 11.2.10:
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l Page No. Am. No.
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- i
- SECTION NO. 11.2.11
Page No. Am. No.
(2 pgs.)
- SECTION NO. 11.2.12:
Page No. Am. No.
(5 pgs.)
- r.
I SECTION NO. 11.2.13:
Page No. An. No.
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Am. 23
--_-_____.,.___,_._.._.-..._,,_.._____.-_,__._._.____,_.a___,,___ _ _ _ _ _ , . , - - - . -
SUPPLEMENT NO. I', PART NO. 1:
Page No. Am. No.
i 16 11 16 111 16 iv 16 v 16 Question No. Page No. Am. No.
1 1 5 i 2 1 22 3 1 6 2 6 3 6 4 6 4 1 5 5 1 5 6 1 5
- 7. 1 5 8 1 13 9 1 5 10 1 16 10 - App. A la 1 13 s 2 13 lb 1 13 2 13 2 1 13 2 13 3 1 13 i 10a 1 5 10b 1 5 10c 1 5 10d 1 5
- 10e 1 5 10f 1 5 4 10g 1 5 10h 1 5 101 1 18 2 18 10j 1 18 10k 1 22 11 1 5 12 1 -5 13 1 5 14 1 5 O 15 16 1
1 5
5 i
- Amendment number not noted on page.
Am. 23
'l SUPPLEMENT NO. 1, PART NO. 1: (Cont'd.)
Question No. Page No. Am. No.
, 17 1 6 l 2 6 i 18 1 5 19 1 5 i
Table 20-1 20 .h 1 1 12 5
l l 2 12 21 1 18 l
j 22 1 5 23 1 5 24 1 5 25 1 5 26 1 5
, 27a 1 10 j 27b 1 12 i 27c 1. 12 27d 1 10 27e 1 5 j 28 1 5 j 29 1 18
, 30 1 5 iO 32 32 1
1 s
5 2 33 1 5 34 1 5 i
35 1 5 36 1 '5 36a 1 5 36b 1 5 Figure 1 to Q. 36 1 9 36c 1 5
, 37 1 5 l 38 1 5 39 1 5 39a 1 5
[ 39b 1 5 l 40 1 5 l 41 1 5 42a 1 18 j 2 18 i 3 18 4 18 42b 1 5 2 5 i 42c 1 5 43 1 5 44 1 5
- Amendment number not noted on page.
Am. 23
i iO phPPLEMENT NO. 1, PART NO. 1: (Cont'd.)
$ Question No. Page No. 'Am. No.
1 45 1 15 Attachment to Q. 45 4
- 1) Document No. 86-1 1105508-00 30
- 2) Document No. 69-1106001 43 46 1 5 47- 1 5
! 48 1 5 49 1 5 s 50 1 5 51 1 5 j 52 1 5 i 53 1 5 54 1 5
! 55 1 23 4 2 23 Attachment to Q. 55 13 23 56 1 5 1 57 1 5 O s8 59 i
1 s
5 60 1 5 61 1 5 62 1 9 63 1 9
- 04 1 9 65 1 5
., 66 1 22 67 1 9 i
4 J
l l
r O
- Amendment number not noted on page.
Am. 23
i O - SUPPLEMENT NO.1, PART NO. 2:
i Page No. Am. No.
i 16 ii '16 iii 16
- iv 16
- v. 16 vi 16 i Question No. Page No. Am. No.
i la/b 1 Ic 1 22 2a/b 1 13 2c/d 1 9 2 9 3 12 4 9 2e 1 9
- Attachment to Q. 2e
- . 1) Document No. 32- *
- 1106898-00 49 3 1 13 2 18 3 3 18
- Attachment to Q. 3 3
- 1) An Evaluation of an 9 10 Emergency Feedwater
! Flowrate of 500 gpm for 25^8 MhT BfiW i Plants
, 2) Document No. 86- 11 6 11-2587-00
]
i 4 1 22 5 1 7 2 7 3 7 i Attachment-to Q. 5-
- 1) Transducer / Computer 2 Speci fication 6 1 18 Attachment to Q. 6 1 18 2 18 s
3 18 i
1 7 1 6 4
4
- Amendment number not noted on page. Am.-23
_- . = _ ~ - -, .
4
'V SUPPLEMENT NO. 1, PART NO. 2: (Cont ' d. )
Question No. Page No. Am. No.
8 1 6 9 1 6 10 1 6 11 1 6 12 1 12 2 12 13 1 22 14 1 18 2 18 Attachment to Q. 14
- 1) Figure 1- 1 18
- 2) Figure 2 1 18
- 3) Letter / Report 26
- 15 1 12 2 12 3 7 Table 15-1 1 7 Figure 15-1 1 7 Document 86-1102525-00 29 7 lO l 16 1 21 17 1 6 18 1 6 19 1 6 20 1 6 21 1 6 22 1 6 i 2 6 23 1 6 24 1 6 25 1 9 2 18 Attachment to Q. 25 5 26 1 6 27 1 6 28 1 6 29 1 6 4
30 1 6 31 1 7 I 32 1 10 33 1 6 34 1 6 35 1 9 l 36 1 6 l 37 1 6 i
- Amendment number not noted on page.
-Am. 23
, . ~ - . - - , - . - . . . . . - . . . . . - . . . . . . . . , . . - . , - . . - . . . - . - . , . . , , . - - - . - . . - . . . - , ~ . . . . - . . . . - - . . . - . , . . . - - _ .
__ _. . _ _ _ _ m . . _ _ .. . . . .._ . . . .
1 15-SUPPLEMENT -NO. 1, PART NO. ' 2 (Cont 'd.) i l
Question No. Page No. Am. No.
f i 38 1 6
, 39 1 6 2
40 6
, 41 6
[ _42 6
- j. 43 6 44 1s 6
! 45 2 6 46 3 6
, 47 6 48 6 j 49 6 50 6
- 51 1 6 l 52 1 22
! 2 22
- 3 21 4 21 5 21 Attachment to Q. 52 3 21 l .
53 1 18
' 2 22 ,
i 54 1 6
! 55 1 7 Attachment to Q. 55 8 6 3
1 56 1 6 Attachment to Q. 56 1 6 l
i 57 1 6'
- i. 58 1 6 59 1 6
- 60 1 6 61 1 6 l 62 1 6 l 63 1 9 l 64 1 6 2 6
' 65 1 6
, 66 1 6 l 2 6
- i. 3 6 l 4 6 67 1 22 l 68 1 22 l 69 1 6
- 70 1 6 71 1 6 I 2 6 1 3 6 j 4 6
! 5 6 6 6 7 6 Am. 23 i
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4 I
SUPPLEMENT NO. 1, PART NO.'2 (Cont'd.)
Question No. Page No. Am. No.
4 l72. 1 6 i< 73 1 6 74 1 6 1
75 1 6 76 1 6
, 77 1 6 78 1 6 i 79
- ! 1 6 2 6 l
3 6 j 80 1 6-
! 81 1 6 82 1 6 2 6 I 3
- 83 . 1 6 84 1 6
.85- 1 6 86 1 -6 87 1 6 i
88 1 6 i s 2 6 3 6 89- 1 6
{
90 1 6
) 91 1 7 I- 2 7 Attachment to Q. 91 19 7 fi 92 1 6 i
93 1 6 i
2 6 l
94 1 6
{ Table 20-1 1 6
' 2 6
- 95 1 21 Attachment to Q. 95 l
Document No. 86-1120838-00 21
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i Am. 23
SUPPLEMENT NO. 1, PART NO. 3 Page No. / .d . No.
i 16 Questien No. Page No. Am. No.
I 1 10 2 13 3 13 Attachment to Q. 1 -
, 1) SDD-211A i 18 11 18 111 18 4 1 18 2 18 3 22 4 18 5 18 6 18 7 18 6 -
8 18 9 22 10 18 7- 18 gg 11 Figure 1 1 18 Figure 2 . 18 Figure 3 1 18 Table 1 1 18 Tabic 2- 1 18 Table 3 -
1 18 Table 4 1 18 Table 5 1 18 Tabic 6 1 18 Letter TMI-80-070 7 Letter TMI-80-036 6 Letter TMI-79-208 5 2 1 10 3 L 10 Attachment to Q. 3 5 4 1 12 2 12 Attachment to Q. 4 Figure 1 1 12 Figure 2 1 12 Figure 3 1 12 Figure 4 1 12 Figure 5 1 12 5 1 12 6 1 12 7 1 12
- knendment number not noted on page.
Am. 23
I i
l' j' i O SUPPLEMENT NO. 1, PART 3 (Cont'd.)
- 3. -
- Question No. Page No. Am. No.
k
, 8 1 12 1
4 -9 1 12 10 1 12 11 1 18
! 2 21 l 12 1 22
[ 2 22
! 3 22 13 1 13
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14 1 12 I
t i-i SUPPLEMEN. J. - 2 Operational Quality. Assurance 118 18 1 Plan for TMI-1 i
4
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Am. 23 l - -.. _ _ _. _ .
-I contact. isolators to preclude the propogation of faults into the Reactor Protection System (RPS). ' Die reactor will be tripped
-f~i through the existing RPS logic upon coincident signals from any M two of the four channels.
A bypass arrangement will be provided in order to allow for power escalation, starting the main turbine and normal shutdown of the main turbine. The main turbine trip bypass will be automatically placed in effect when reactor power is less than 20%.. The bypass will be automatically removed when the reactor power is increased above 20%. Bypass of the feedwater pump trip signal ~ is automa-tically placed in effect when reactor power is less than 10%. It will be . removed automatically when reactor power is raised above
- 10%. Tht bypass fnction will be accomplished individually in ~
each of the four channels by means of bistables which monitor the power range nuclear instrumentation.
The additional moduli,s required in the Reactor Protection System
'will be the same safoty grade equipment type used in the original system. Wiring for redundant channels will be separated and run in Se'ismic I, safety grade raceways except in the turbine build-ing. Since the turbine building is not Seismic I, the equipment and wiring.therein cannot be classified as Seismic I. However, all wiring in the turbine building for this system will be run in conduit and redundant channgis will be routed separately to minimize the probability of disabling more than one channel due to damage to the turbiae building. The system will be designed with normally closed contacta so that an open wire will represent
() a tripped conditien. The sigtals from the turbine building will
- o through contact isolators in the Reactor Protection System to preclude the propogation of faults in the system. -
2.1.1. 1.4 Design Evaluation The Safety Grade Reactor Trip scheme provides an anticipatory trip to the reactor, reducing the number of reactor trips on high pressure and the number of challenges to the Pressurizer PORV and safety valves. (Also see Supplement 1, Part 2, Question 15.)
2.1.1.1.5 Safety Evaluation The system is safety grade and meets the requirements of IEEE-279 including those for testability and single failure criterion.
The modifications which will be required to the Reactor Protec-tion System will not degrade the ability of that system to perform its design function. The design will result in an enhancement of nuclear safety.
- 2.1.1.1.6 Start-Up Testing
< This system will be tested during installation to verify its operation prior to start-up.
1O I 2.1-2 l
i 1 Am. 23 4
.c eer,,w,-.m.,...m.,ve,-.., vi*,--- - - , - . --+me, . w , w y.y y.w w w , -.,.%-,,,.-e--c-,pvww.wm~,,.,,-,y,-wy*n,-,,.-,----.ev-wy-w w w,we.- ,.--y..c---,vv,,+v.w ,-.,ew-, r, w-,.e vie.e--
conditions. Calculations have benn mada, using conservstiva narumptions, to demonstrate that a satisftctory eignal will be generated when any of the valves open. Calculations have been made for saturated, liquid and two phase flow. A summary of g these calculations is provided in Appendix 2A. The calculations
(") demonstrate that a satisfactory signal is developed for flows as low as normal makeup flow for all plant conditions of interest.
Tests run by B&W on the electromatic relief valve under reduced flow condiitons have confirmed the validity of this approach.
Because of the straight-forward and well known relationships that exist between flow conditions and differential pressure across the elbow, the signal from one dif fereatial pressure transmitter can be confidently predicted for any flow conditions.
For this reason it has been concluded that operating tests, which would be dif ficult since they involve opening the PORV and relief valves, will not be required.
Acoustic monitoring of the e_ ::tromatic relief valve makes use of well proven equipment and techniques which have been ued in the B&W Loose Parts Monitoring System. Test:. run on this valve at the B&W Alliance facility demonstrated that the acoustic monitor-ing system gave satisfactory results.
2.1.1.2.5 Safety Evaluation Instrument taps will be installed on elbows in the discharge piping of pressurizer code safety valves RC-RVIA and RC-RV1B and electromatic relief valve RC-RV2. This piping is classified as N2, Seismic I. Analysis has been performed to demonstrate that this modification will not degrade the integrity of the existing p pipe. The pipe classification has been maintained up to and V including the instrument root valves. The mounting of new equipment which will be located in the vicinity of safety related systems has been analyzed to ensure that no hazardous missiles will be generated in a seismic event. It has been concluded that this modification will not degrade any safety related systems.
Shock lor. dings were considered in the original design of the PORV and safety valve discharge piping. It was also e m idered for the desitn of the elbow tap instrument since water loop seals are maintained on the safety valves to prevent contact of steam during normal operation and the PORV discharge lines j
may see slug flow under certain conditions. When the safety /
l relief valve opens a slug of water is discharged to the tailpipes.
The small size of the elbow tap instrument lines compared to the 4 inch (PORV) tailpipe results in only a small portion of the pressure wave (due to slug flow) from effecting the elbow tap l
instrument. Any ef fect is further dampened by condensing pots
! in the inst 1rument lines close to the tailpipe elbow and due to the relatively long length of instrument sensing lines.
The differential pressure cell is also designed to withstand full
! system pressure of 2500 psi across the diaphram compared to an operational requirement uf 400 inches of water with no loss of accuracy or damage.
All of the equipment inside containment for Pressurizer PORV and safety valve detection will be seismically and environmentally (g) qualified. Work is underway to upgrade the portion of the system U outside containment. This involves specifying and procuring of .
additional equipment. l l
t
! 2.1-5 l Am. 23
2.1.2 Long Term Modification 2.1.2.1 Post Accident Monitoring O/
2.1.2.1.1. System Description Certain post accident monitoring capability will be provided in compliance with Reg. Guide 1.97, Rev.2 as discussed below. Pending the availability of appropriately qualified instrumentation and equipment , the following modifications will therefore be completed as soon as possible. The final design will be provided for NRC review.
Containment Pressure - Continuous containment pressure indication will be provided in the control room using a range from -5 psig to three times the design pressure of the containment. The pressure indication will be safety grade and will meet the design and qualification requirements of Reg. Guide 1.97. Redundant indication of pressure will be provided.
Containment Water Level - Continuous containment water level indication shall be provided in the control room. A safety grade wide range indicator from the bottom of containment to a level of 90 inches will be installed in accordance with the requirements of Reg. Guide 1.97. In addition, a narrow range indicator from the bottom to the top of the sump with continuous indication in the control room shall be installed which meets the requirements of Reg. Guide 1.89 and is capable of being periodically tested.
[)'
Containment Hydrogen Indication -Safety grade continuous indica-tion of containment hydrogen will be provided in the control room. The range of indication will be 0-10% concentratie.a assuming commercial availability over this range.
High Range Containment Radiation Monitor - Two safety grade containment radiation monitors that are physically separated shall be provided with recording display and continuous indicator presentation in the control room. The range of this monitor shall be 107 R/hr and shall detect photon radiation down to 60 Kev. The design of the radiation monitors shall be provided in accordance with Reg. Cuide 1.97 Rev. 2.
High Range Effluent Monitor - High range effluent monitors intended as the Long-Term modification are planned for each i normal gas release point.
The range of these monitors shall be as follows:
- Undiluted Containment Exhaust . . . . . . . . . 105 uci/cc
- Main Steam Lines . . .. . . . . . . . . . . . 102u ci/cc
- Auxiliary & Fuel Handling Building Exhaust . . . 103uci/cc
- Condenser 0FF GAS Exhaust . ....... . . . 103 uci/cc 2.1-30 Am. 23
R:guletory Guide 1.97 R;v. 2 (Dec. 1979) will be fcilsw2d for the design of range for the high range ef fluent monitors. Vital bue power shall be employed for each system's mooular assembly with the normal power eupplying the monitor pumps with diesel gene -
7s ations as back ups.
(\_ ;
High Range Ef fluent Radio lodine & Particulate Sampling Analysis -
The existing sampling system will be expanded and will include the addition of silver zeolite cartridges. The system design und operation will both decrease the activity on the cartridges so they can be handled and will decrease the xenon to iodine ratio. Counting of the cartridges will be by use of NaI crystal connected to a single or dual channel analyzer with appropriate window and discrimination settings for th 364 Kev gamma of I-131, or by use of a GELI/MCA system. The expanded portion of the sampling system would be placed in service follow-ing an accident and will be located in an applicable area exhibit- g ing low background. I Prior to incorporation of the expanded sampling system, procedures have been developed for the use of silver zeolite cartridges and l normal particulated filters for sampling with a Na1 detector and a single or dual channel analyzer for iodine and gross particulate release rate determination. Specific details to insure exposures are maintained as low as reasonably ~ achievable are incorporated into the procedures.
2.1.2.2 RCS Venting rS 2.1.2.2.1 System Cescription V
Power operated vents will be provided for the reactor coolant system in order to enhance natural circulation and adequate core cooling following an accident. The vents will be from the top of the pressurizer, the top of both hot legs using existing connections on the reactor coolant piping and f rom the Reactor Vessel Head. The discharge from the reactor vessel and hot leg vents will be directed to the containment atmosphere. The system is shown schematically in Figure 2.1.-11.
The hot leg vents will tie into existing hot leg vent piping in-side the secondary shield wall. As part of this modiiication, re-mote operation of the vent valves in the existing vent line from the pressurizer to the reactor coolant drain tank will be provided and the system will retain the existing venting capability. Con-trol and position indication for the power operated vent valves will be provided in the control room.
2.1-31 A
V Am. 23
2.1.2.4 Po*t Accid:nt Sempling Cepibility
,, 2.1.2.4.1 Reactor Coolant Sampling
! t Reactor coolant post accident samples will be obtained and analyzed using the existing system and analysis equipment. The sampling system is shown on flow diagram C-302-671 and is located in the nuclear sampling room at Elevation 306'-0" of the control room tower as shown on general arrangement drawing E-015-015. The pre-paration of sanples and chemical analysis will be performed in the radio-chemistry lab which is adjacent to the sampling room as shown on drawing E-015-015. The use of the existing equipment and facil-ities will be augmented with emergency plan procedures, use of long handled tools, lead shields and a shielded sample transport cart.
Reactor coolant following an accident will be sampled from the reactor coolant loop "B" cold leg letdown line via valves CA-V13 and CA-V2. Upon opening of these valves the reactor coolant is routed to sample cooler CA-C1 through the sample hood and sampled from valve CA-V16. From the sample hood the reactor coolant is then routed to a point upstream of the makeup purification system filters (MU-FIA/B) which discharge to the make up tank (MU-TI).
The following table lists the valves which are operated to line up the sampling system as previously described:
Reactor Coolant Sampling System Valves Reference Valve Tag No. Position Operator Location Flow Diagram 7w b Motor Containment C-302-6??
CA-V13 Open Building CA-V2 Open Motor Auxiliary C-302-671 Building CA-V25A Open Manual Nuclear C-302-671 Sampling Room Sample Cooler CA-V26A Closed Manual C-302-671 CA-V25B Closed Manual C-302-671 CA-V26B Open Manual C-302-671 CA-V26C Closed Manual C-302-671 CA-V25C Closed Manual Nuclear C-302-671 Sampling Room Sample Hood f) x-2.1-39 Am. 23
Reactor Coolant Sampling System Valves
,y Reference
(_,) Valve Tag No. Position Operator Location Flow Diagram CA-V33 open Manual C-302-671 CA-V110 Open Manual C-302-671 CA-V34 Open Manual ,C-302-671 CA-V35 Open Manual C-302-671 CA-V16 Open Manual C-302-671 CA-V29 Open Manual C-302-671 Valves CA-V13 and CA-V2 may be operated from a local control panel in the nuclear sampling room or the ES (Engineered Safeguards) panel in the control room.
When the sempling system is used under post-accident conditions the nuclear sampling room and the areas through which the sample line runs will become high radiation areas. These areas will be admin-istratively controlled during sampling evolutions to restrict person-nel access to limit personnel radiation exposure.
Figures 2.1-14 and 2.1-15
(~} show the affected areas and the calculated dose rate based on the
'/ source strengths for T=0 (Reactor Trip) tabulated below.
The taking and handling of a post-accident reactor coolant sample will be done in accordance with emergency procedure. The procedure uses a team of three technicians in order to limit the exposure to any one individual. For the first sample, to be taken at T=45 min.
the first technician will enter the nuclear sampling room to set up the equipment, perform a valve line-up and establish flow through the sampling system. Eouipment set-ur,and valve line-up are expected to take 4 minutes with negligible exposure ( 3 MREM based on a 50 MREM /hr field). When flow is established the general area radiation level wil. increase to 180 REM /hr and result in a exposure of 3 REM to the technician based on a stay time of 1 minute. The second technician will then obtain the sample by opening valve CA-V16,
}
l filling the sample bottle (30 m1 sample) from the CA-V16 drain line and then placing the sample bottle in the shielded cart, he will then close CA-V16. (CA-V2 and CA-V13 are closed from the control room). These operations are performed in the sample hood using long handled tools and with a stay time of 1.5 minutes results in an exposure of 4.5 REM (based on a dose rate of 180 ROH/hr). The third technician will move the shielded cart to the chemistry lab.
Moving the cart requires a stay of 1 minute resulting in an exposure of 3 REM (based on a dose rate of 180 REM /hr). The same procedure will be followed for a second sample assumed to be taken at T=8 hrs. The sampling evolution exposures are summarized below:
()
2.1+4 0 Am. 23
Racetor Coolant Sampling Dose Rates 45 Min. 8 Hr.
/~N Stay Time Dose Rate Dose Dose Rate Dose b 4.6 REM Technician 1 4 min. e 55 mr/hr 3 REM 55 REM /hr 1 min. 180 REM /hr 55 REM /hr Technician 2 1.5 min. 180 REM /hr 4.5 REM 55 REM /hr 1.4 REM Technician 3 1 min. 180 REM /hr 3 REM 55 REM /hr .9 REM
- general area reading of the TMI-1 nuclear sampling room normally.
Note: The 45 minute sample is based on the decision to take a sample at T=0 (reactor trip). 45 minutes is to be the earliest time a sample can be taken.
Sample preparation and analysis will be donc in accordance with emer-gency procedure. The procedure provides guidance to obtain analytical data on boron concentration, isotopic identification and chlorides.
Upon transport of the shielded sample cart to the chemistry lab, the 30 m1 sample will be transferred from the cart to a lead pig located in the chemistry hood. The hardware used for sample preparation and analysis will have been laid out per procedure prior to removing the
- sample bottle from the shielded cart.
Once a sample bottle is in the chemistry hood, 0.1 m1, 1.0 ml and 2.0 ml portions of the sample are pipeted into various containers for, (1) dilution in order to perform isotopic identification, (2) boron analy-sis and (3) chloride analysis, respectively. The estimated time to
()
perform the dilution and analyses (boron) is 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> resulting in an exposure of 2.98 REM for first sample and .86 REM for the second sample to the technician. The dose takes into account the use of lead shields long handled tools and the handling of small or diluted samples. Samples prepare for isotopic analysis are transported to the count room located in the turbine building for counting on the Geli detector. There is no exposure associated with transporting and counting since the samples must read < 1 MR/hr prior to leaving the chemistry lab. The chloride sample will be obtained and analysis will be done off-site within 4 days.
The overall time to obtain and analyze the sample is tabulated below.
I RCS Post Accident Sampling Sequence Time, hr Cumulative Time, hr
- 1. Reactor trip & decision to 0 0
- take sample i
- 2. Complete personnel and equip- 0.75 0.75 ment preparation for obtaining sample
- 3. Obtain sample 0.50 1.25
> 4. Analyze sample 1.75 3.0 i
2.1-41 t
Am. 23
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-- - ---.m --. , , , . . - - . . . - , - - - - , , ,~ -, , . -- .-
2.1.2.4.2 Cantainmsnt Atmospharo S:mpling Containment atmosphere, post accident, aill be analyzed by taking grab samples of the containment air and identifying the various
(] isotopes and measuring hydrogen concentration. The sample will be obtained from a sample station located abave the instrument air compressor (shown on drawing D-001-016) wtich is located in the southeast corner of the intermediate building. The sample system will use the existing penetrations for RM-A2 (Containment Air Monitoring System) as shown on drawing C-302-721 (Figure 2.1-16),
to sample air inside containment. Using these penetrations, a sample can be drawn from two areas inside containment, the contain-ment ventilation duct or the discharge of the containment. cooling fans.
The sampling system is shown on Figure 2.1-16. The system ties into the sample lines for RM-A2 downstream of containment iso-lation valves.CM-V3&4 and CM-V1 & 2 via the three way solenoid operated valves CM-V8 and CM-V7. Upon opening the containment isolation valves and positioning the three valves to divert flow to the sampling system, a jet pump is used to circulate the air from containment, through a sampler and back to containment. By proper positioning four way valve CM-V9 on the jet pump discharge, air can be drawn from the ventilation duct and returned to the discharge of the cooling fans or vice versa. The following table lists the valves which are operated to line up the sampling system as described:
'A Containment Atmosphere Sampling System Valves V
Valve Tag No. Position Operator Location CM-V1 Open Pneumatic Intermediate Building CM-V2 Open CM-V3 Open CM-V4 Open CM-V7 (New) Open to Sample i
System CM-V8 (New)
- CM-V9 (New) Open Motor CM-V10 (New) Open
- Valves are located inside the penetration cubicle next to and above the instrument air compressor cubicle in the southeast Ik corner of the intermediate building.
l 2.1-42 Am. 23
Tha v lves lict d cbova will be op; rat 2d f rom e new lec21 centrol pinal locetad in ths corridor at Eltvrtion 305'00" of the inter-mediate building. Once flow has been established from the control panel the technician proceeds to the sample station and withdraws
(~N a sample either through a septum on a 25cc sample bulb using a
\_/ syringe or removes the entire sample bulb f rom the system. After obtaining the sample the system is then transported in a shield to the chemistry lab for preparation for analysis. The system can also be used to sample containment air using filtration or absorption devices.
Operation of the sampling system, under post accident conditions will result in a generally low area radiation level at the samp-ling station.
Of the evolutions described previously, the only significant exposure will be when the technician handles the sample prior to transport.
The time to accomplish this evolution is 3 minutes resulting in an exposure of < 2.5 REM. All other sample system evolutions are either at the control panel which is located in an unrestricted area or af ter the system has been purged which reduces the general area radiation to a negligible level.
Af ter the sample has been transported to the chemistry lab, a sample is either prepared for gamma spectrometry on the GELI detec-tot or for hydrogen analysis on the gas chromatograph. If the 25cc sample bulb was used, a sample for gamma spectrometry is pre-pared by extracting Scc using a syringe and injecting it into a Sec glass vial. The glass vial is then transported to the count room.
Hydrogen analysis is done by withdrawing 0.5cc and injecting into
',_,) the gas partitioner which is located in the chemistry lab. Negli-
gible exposure is associated with these evolutions because of the sample size, use of shields and the time to perform the above.
2.1.2.4.3 Reactor and Containment Atmosphere Post Accident Source Terms The post accident source terms from which the dose rates are calcu-lated were taken from the Midland Final Safety Analysis Report, Table 11.1-2, Total Core Fission Product Acti rity Versus Time in Equi-librium Cycle. The table in the Midla ad FSAR is based on an iso-topic core inventory for 310 ef fectiv< full power days in an equilibrium cycle at a power level o' 2552 MWt. Source term in-formation was taken f rom the Midland FSAR in the absence of comparable information for THI-1. This information is slightly conservative since the THI-1 power level is 2535 MWt.
I The source strengths for reactor coolant and containment air following an accident are listed below. The reactor coolant source is based on the activity f rom 100% of the noble gases, 50% of the halogens and 1% of all other isotopes being diluted in the reactor coolant liquid volume. The containment air
- source is based on the activity f rom 100% of the noble gases and 25% of the halogens being dispersed within the air contained l
in the containment building f ree volume. The source strengths listed below are for times T=0 (reactor shutdown) and T=8 hrs.
(~; Personnel exposures for the various sampling evolutions are L'
2.1-43 Am. 23
bez:d on the T=45 min. sourca str:ngtha. Szmpling at T=45 nin.
allows for a decay period of 4' min. which takes into account a time to fall fuel and to disperse the fission products since
\ these events are not instantaneous processes.
(V Dose Rate Source Strengthe, I Reactor Coolant i
Time = 0 Time = 8 hrs Energy (MEV) 7/cc-sec Energy (MEV) 7/cc-sec
.13 1.23 x 1010 .094 6.59 x 109
.36 2.27 x 1010 .32 8.19 x 109
.74 4.59 x 1010 .61 8.02 x 109 1.23 1.19 x-1010 1.23 3.04 x 109 1.74 6.51 x 109 1.73 1.19 x 109 2.20 8.09 x 109 2.28 7.05 x 108 2.57 1.64 x 109 2.56 6.85 x 107 3.52 1.44 x 108 3.19 9.72 x 106 4.11 2.33 x 106 -4.68 1.34 x 106 II Containment Atmosphere Time = 0 Time = 8 hrs Energy (MEV) 7/cc-sec Energy (MEV) 7/cc-sec
() .13
.36 6.61 x 107 1.09 x 108
.095
.31 3.56 x 107 2.75 x 107
.74 1.38 x 108 .61 2.25 x 107 1.24 3.79 x 107 1.25 9.02 x 106 1.75 2.64 x 107 1.72 2.31 x 106 2.19 4.42 x 107 2.28 3.75 x 106 2.57 9.29 x 106 2.57 1.39 x 105 3.52 4.08 x 105 3.52 1.14 x 101 4.09 6.42 x 103 4.09 1.80 x 10-1 i
2 f -2.1.2.5 Reactor Coolant Pump Trip on HPI 2.1.2.5.1 System Description i The purpose of this proposed modification is to provide automatic i trip of the Reactor Coolant Pumps when degraded primary system conditiona associated with a LOCA have been detected. This will be accomplished by requiring that RCP trip be initiated when the Engineered Safeguards System has actuated Safety Injection and saturation margin has been lost.
The proposed logic will preclude RC pump trip during those events such as severe overcooling or very small breaks where maintenance s_-
2.1-44 I
Am. 23 [
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, -< - ,e-,-- , . - , - - - - , - - , , , - - , . , , . - , , , - - . - ,-. .,,- .,,, , , = - - . . . .,. -_n_.
- _- _ _ - . _ _ . ___- _ _ m . . _ .
of forced cooling is very desir:ble. The conceptual design das-cribed in this section is being submitted for NRC review and comment and will be implemented subject to concurrence of the NRC staff.
'O. 2.1.2.5.2 Design Bases
- . See response to Question 11 of Supplement 1, Part 3.
'2.1.2.5.3 System Design - To be provided later.
- 2.1.2.5.4 Design Evaluation - To be provided later 4
2.1.2.6 Auxiliary Feedwater System 7
Auto start of the emergency-feedwater (EFW) System is being imple-mented in two phases: 1. . Control Grade Auto Start .This is a non-safety related initiation as described in paragraph 2.1.1.7
- and it is a short-term approach, 2. Safety Grade Auto Start -
This will be a long-term modification where the initiation will amet the requirements for Class 1E system and the system is func-l tionally described below.
t- 1. The safety grade EFW auto start when implemented will automatic-ally initiate the system on presence of the following condi-tions with or without the availability of the off-site power:
- Loss of both normal feedwater pumps, or
- Loss of all four reactor coolant pumps, or
- Low differential pressure between-the normal feedwater ph t and main steam lines at either steam generator, The system initiation on low steam generator level will even-
! tually-be added. This will be done after the necessary analysis and engineering has been completed to insure that this signal
! will give a satisfactory actuation and will not interact with l
other plant functions. Loss of normal feedwater pumps is de-l tected by differential pressure switches across each pump (two
! switches per pump, i.e. , one switch per train).
! The model of differential pressure switches used for this L application has been seismically tasted. These switches have j temperature limits of -60 to 200*F. Since they will be located l
in Turbine Building which is a non-seismic building, the switches will be tied into their respective EFW intiating cir-cuits (Train A&B) through buffer devices and thus the switches will be treated as safety grade items to the extent possible.
The buffer devices are relays. similar to those described in the TMI-2 FSAR Section 7.3.2. This application is similar to the approved application described in that section.
! 2. All cables associated with the initiating logic will be quali-
! fled for Class IE apS' n. ion and the initiations will be de-
' signed to meet single failure criteria. All circuits will meet the regulatory criteria for-separation of Class IE circuits.
O i
2.1-45 l
Am. 23
- 3. The initiating logic will include hardwsra for th2 following purposes:
- Latching mechanism to seal-in the actuation h")
- Manual Bypass Capability
- Testability of the initiating circuit
- 4. Indication will be provided in the control room to identify the source of the initiation.
- 5. Annunciation will be provided in the control room to alarm:
- Auto start of the EFW system. This will be a common alarm for both the trains.
- Initiating conditions being bypassed. This will be a common alarm for all initiating conditions associated with with the same train.
2.1.2.7 Increased Range of Radiation Monitors (2.1.8.b)
The existing Radiation Monitoring System provides in-line monitor-2.1.2.7.1 ing capability for effluents from:
a) Auxiliary and Fuel Handling Building (RM-A8) b) Reactor Building Purge (RM-A9) c) Condenser Off-Gas (RM-AS)
Discharge from Waste Gas Decay Tanks is monitored by RM-A7 prior l (p) l to combination with other exhaust and after dilution by RM-A8. The I Reactor Building Hydrogen Purge System discharge is monitored by the I normal purge system monitor RM-A9.
The monitors, RM-A8 and RM-A9 are manufactured by Victoreen, Inc. and I
consist of:
I a) A fixed filter particulate monitor -Beta scintillation detector; sensitivity approximately 1.5 x 10 0 / min l
based on SR-90; full range 1 x 106 cpm.
b) A Fixed Charcoal Filter Iodine Monitor; NaI detector with fixed in window; Sensitivity approximately 1.3 x 10 9 fill range 1 x 106 cpm.
c) A gross gaseous monitor; Beta Scintillation detector; Sensitivity approximately 4 x 107 cpm full range Ci/ce; 1 x 106 cpm.
I d) Air sampling pump with normal sample flew of approximately 1
- cubic foot per minute.
Radiation monitor RM-A5 has only - gross gaseous monitor (c above) l situated on the discharge of the condenser vacuum pumps, exhausting i
i to the suction of the vacuum pumps. Flow through the monitor is 1 - regulated to maintain approximately 500 cc/ min. All monitors have Control xoom readout and recording.
l l >
2.1-46 i
l Am. 23
2.1.2.7.2 Long Term Modifications Increased range capabilities will be furnished for each of the
?~N effluent monitors described above (RM-A8, RM-A9, RM-AS) and the (m-) Main Steam lines. For the Long Term Modification additional monitoring ranges will be provided utilizing ionization chambers for the Reactor Building Purge Exhaust, the Condenser OFF-GAS Exhaust, the Main Steam Lines. The Auxiliary and Fuel Handling Building Exhaust will have extended monitoring ranges incorpor-ating a C.M. device. The sensitivity of the individual units will be determined by standard volume source calculations.
The sensitivity will assure that release rate of:
5,600,000 Ci/sec from Auxiliary & Fuel Handling Building 2,300,000 Ci/sec from Reactor Building Purge.
1,400 Ci/sec from Condenser Of f-cas based on minimum flow rates from each release path.
2,500 Ci/see from a single steam generator can be detected.
The installation of each monitor will include evaluation of the position of the monitor relative to other potential radiation sources and shielding necessary to minimize the effect of sources other than sample lines on the response of the monitor and recording.
() For each of the monitors described, the following applies:
Each will be powered from vital power, thereby providing redundancy in power supply.
Establishing sensitivities will be correlated to solid source calibrations. Procedures defining calibration method and frequency will be written to assure proper response of the instruments.
- Emergency procedures will be written to the use of the i radiation instrumentation in conjunction with flow information to determine release rate.
Emergency Plan implementing procedures describe the dis-semination of information obtained from monitors.
Procedures and evaluations will be available for NRC review prior to restart.
2.1.2.7.3 Short Term Modifications Increased range capabilities will be furnished for the Reactor Building Purge Exhaust, the Condenser Of f-Cas Exhaust, and the Main Steam Lines as a short term modification. This Short
("% Term Modification will consist of G.M. Tubes or ionization
( ,)_
chambers affixed to each of the effluent release paths described r
2.1-47 Am. 23
- _ _ - . - - .- - -. .- . - - _. ~-
~
in 2.1.2.1.1 (o' nly one detection' ayaten will be provided for each OTSG).' Remote readout will be provided to areas which are habit--
able during an accident. The Long Term Modification for the
'- ~ Auxiliary and Fuel Handling Building is projected to be complete
'before restart. If a Long Term Modification is not available by I,
start-up, due to equipment delivery problems, a Short Term Modi-
, fication utilizing a G.H. tube or ionization chamber will be in-corpora ted. All devices will have necessary shielding if back-ground effects are considered excessive.
The installation of each monitor will include evaluation of the position of the monitor relative to other potential radiai. ion sources and shielding necessary to minimize the effect of sources other than sample lines on the response of the monitor.
i i The sensitivity will assure that release rates of:
l 5,600,000 Ci/sec from Auxiliary & Fuel Handling Bldg.
2,300,000- Ci/sec from Reactor Building Purge.
1,400 Ci/see from Reactor Building Purge.
1 i 2,500 Ci/see from a' single steam generator can be detected.
l The range of these monitors is identical to the range capability of the long term modification.
4 For each of the' monitors described, the following applies:
Each will be powered from normal power with battery backup. -
Established sensitivities will be correlated to solid source calibration. Procedures defining calibration methods and frequency will be written to assure proper response of the instruments.
Emergency procedures will be written to the use of radiation instrumentation in conjunction with flow f nformation to de-termine release rate.
Emergency Plan implementing procedures describe the dissem-ination of information obtained from the monitors.
Procedures and evaluations for interim methods will be available e for NRC review prior to criticality if the long-term modifications are not completed by that time.
l i
- O 2.1-48 Am. 23 o . - , _ , . . . . _ .
_/ y , , _ .
TilREE MILE ISLAND UNIT NO. 1 TABLE 2.1-1 1.ist of Isolation Signal Override Capability Isolation Signal Penetration Reactor liigh 4 psig 30 psig 1600 psig Line No. Trip Radiation Building Building (SFAS) Break Conteinment Air Sample 108 N/A N/A C N/A N/A C_
R.B. Sump 353 C IB C N/A N/A N/A RCDT 330,331 C IB C N/A N/A N/A RCS Sample 328 C IB C N/A N/A N/A R.B. Purge 336,423 K K K N/A N/A N/A RCS Hakeup 323 N/A N/A C N/A C, N/A RCS Letdown 309 (HU-V2A/B) N/A IB C N/A C N/A (HU-V3) A N7A C N/A N/A N/A Demin Water 307 C N/A C N/A N/A N/A GTSG Sample 213, 214 C IB C N/A N/A N/A NSCCW 346, 347 N/A N/A N/A NO N/A NO ICCW 302, 333, N/A N/A N/A HO N/A NO 334 R.B. Air Coolers 421, 422 N/A N/A C N/A C N/A R.C. Pump Seal Return 329 N/A N/A NA NO N/A N/A Core Flood TK 348, 349 C N/A C N/A N/A N/A Legend C = Common Signal Override; initiating isolation condition may still exist.
, I = Individual isolation signal override capability; procedures governing override to be developed.
~t IB = Individual isolation signal bypass capability A = Automatic isolation signal override.
K = Common signal override wi th key interlock permissive.
NO = No override or bypass capability; initiating condition must clear to allow reopening of valve.
N/A = Not applicable.
No te : For combinations of initiating signals that are allowable, refer to Table 2.1-2.
Am. 23
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THREE MILE IStAND L' NIT NO. 1 TAELE 2.I-2 LIST OF cnNTAINMFNT ISOLATION VALVES REQUIRINC MCCIFICATIONS Valve Line Method Normal Post Actual Valve Penetration Valve Valve Sire, of Valve Accident Position Position Actes ten Signal Source No. S Service J stem Tag No. Tg In. Ac t ua t ion Position Faisting Modified Indication Es n.sg Modified Notes 108 Containment Ai r RM Of-V I Ball I Air Open Closed Closed Yes Sample 1.10 1.L6.10 CM-V2 Sall I Ai r Open Closed Closed Yes Of-V 3 Ba!! I Ai r Open Closed Closed Yes CM-V4 Sall I Air Open Closed Closed Ye s 213 Steam Generator CA CA-V4A Clohe 3/8 EMO Closed Closed Closed Yes 1.10 1.4.5.6,10 No B&W reccamendation Sample CA-VSA Clebe 3/8 Air Closed Closed Closed Yes 214 Steam G7nerator CA CA-V48 Clobe 3/8 EMO Closed Closed Closed Yes 1.10 1.4.5.6.10 Sample CA-V58 Clebe 3/8 Ai r Closed Closed Closed Ye s 302 Intermediate IC IC-V2 Cate 6 EMO open Class" open/ Closed Yes
- Cooling IC-V3 Cate I .Jt10 3_.7.8.9.10 6 Air Open Closed Open/ Closed Yes
! Water Outlet Line 307 Demin. Water to CA CA-Vl89 Cate 2 Air Open Closed Closed Yes Reactor Butiding 1.10 1.5.40 309 Intdoun Line to MU MU-V2A Clebe 2-1/2 EMO Open Yes Purification MU-v28 Clobe Closed Open/ Closed 1.10 1.2.4.6.10 2-l/2 EMO Open Closed Open/ Closed Yes 3.10 1.2.4.6.10 Demineralizers MU-V 3 Cate 2-l/2 Air Open Closed Open/C1-=ed Yes 1.10 1.5.6.10 323 aC Makeup MU Mu-vi8 Cate 2-1/2 Ai r Open Closed Closed Yes I.10 1.2.10 1 328 Pressurizer and CA CA-VI Clebe 3/8 EMO Closed Closed closed Reactor Cociant Ye s 3.10 1.4.5.6.10 1 CA-V2 Care 3/8 Ai r Closed Closed Closed Yes
- Sample Lines CA-V3 Globe 3/8 EMO Closed Closed Closed Yes CA-Vl3 Clobe 3/8 EMO Closed Closed Closed Yes 329 Reactor Coolant MU MU-V25 Clobe & EMO Open Closed open/ Closed Yes I.7.10
- Pump Seal Return 1.7.8.10 MU-V26 Cate 4 Air Open Closed Open/ Closed Yes 3 30 acactor Coolant WDC WDC-V3 Clebe 2 EMO open Closed closed Yes 3.10 1.4.5.10 Dratn Tank WDG-V4 Cate 2 Air Open Closed Closed Yes Vent 331 Reactor Coolant WDL WDL-V 303 Cate 4 EMO Closed Closed Closed Yes 1.10 1.4.5.I0 Drain Tank Fm p WDL-V304 Cate 4 Air Closed Closed Closed Yes
,1 Di sc ha rge J
333 Intermediate IC IC-V4 Cate 6 Ai r Open Closed Open/ Closed Yes 3,7,8.9.10
( Cooling Water f.L10 Supply Line i( 334 In t e rmed ia t e IC IC-V6 Cate 3 Air open Closed Open/ Closed Yes Cooling to 1.LIO 3.7.82 9.10 l[ CROM Cooling Co11e 4
}
i i
Pg. 2 of 2 THRf'E ftILE ISI.AND t' Elf NO.1 TABLE _2.5-2 (CONT'D) llEEF,pp,T,4tyN 'T ISOLATION
- V AL.VI.% BffFIRINC tmDtFICATIONS Valve Line *1c t hod Nornal Post Actual Valee Penetration Valee Valve $ s re , of Valve Accident Posteten Position Actuitten Stanal Source No. Service _ _, Systen Ta* he. h_ In. Ag r uat ion Position Futstina th*dlited Indleation Entseina teodified Notes 336 ' eactor B.itidinit AN AM-VIA Butter- 48 Air Closed Closed Closed Yes 1,10 1,4,5,10 Outlet Purge fly Line AM-Vl3 But t e r- 48 L!?n Closed Closed closed Yes fly 346 Reactor Coolant NS MS-V IS Cate 8 cm open Closed open/ Closed Yes 1,10 2,8,9,50, Punp flotar Cooling Water )
Supply !
t 347 Reactor Coolant NS NS-V4 Cate 8 tm Open Closed open/ Closed Yes 1,10 7,8,9,10 i Pump Motor -V 3' Cate 9 Im Open Closed Open/ Closed Yes 12 7,8,9,10, l Cooling Watar Return
- 35) Reactor Rutidtrg WDL WDL-V S 34 Cate 6 Air Closed Closed closed Tes 1,10 I,4,5,80 7 Sump. Drain WDL-VS)$ Cate 4 Air Closed Closed Closed Yes 421 Reactor tutIdtag RB RB-v2A Cate 4 tm om closed open Yes I 10 1,2,I0; Adu auto intttation of Foerg. R.B. cooling _on 4 psig Normal Air Coolers Supply R.B. and 1600 pets R2 C.
Line p_ressure isolation signa h 422 Reactor Butidica RB RS-V7 Cate 6 Air Open Closed Open Yes 1,10 1,7,10; Normal Air Coolers Return Line
- 42) Reactor Buildina AH AM-VIC Butter- 48 DU Closed Closed Closed Yes 1,4,10 1,4,5,10 Inlet Purge fly Line AM-VID Butter- 48 Air Closed Closed Closed Yes fly 348.149 Ceye Flood TK. CF CF-V2A48 Clobe i tnt Closed Closed Closed Yes t a lo 1.5.10 Sample and Ny f ill -Vl9A4B Cate ! Air Li nes -v20A4B _Ca_.t_e _i A_i r Valve Actuatten 3tgnal Sou p i
- 1) 4 psts reactor building pre w ce- Isolation 7) Classify lie to c etente Category I f
- 2) 1600 psig (SFAS) isolation 8) 30 putg rentar twet tdtng pressure isolation
- 3) Radiation alare, operator ac t ion required 9) Line break 4 *ilat tan stRnal
- 4) Migli radiation (non-safety) isolet tosi 10) Renate man.nl cont rol y 5) Reactor trip toelation
- 6) Override capability on individual valves U
....,___.._m. _ . .___.m . . _ . . . . . _ _ . _ . - ___ _. -~ ._ . . . . - - - . . . _ . . . . .=m.. ~. . ._ m . _mm_ . . _ _ . _ _ , , _ <
o- :
THREE MILE ISLAND UNIT NO. 1 '
TABLE 2.1-3 I
LIST OF CONTAIN* TENT PENETRATIONS REQUIRING ISOLATION ON HI-RADIATION '
Pene t ra tion leolation Radiation Valve Detector Type of No. Service System Tag No. Location Monitor 213 Steam Generator CA CA-V4A and Sample locate the monitors outt!de the R.B. Area
-V5A near the sampling line downstreas of Cam.a t
214 -V4B the contatrument isolation valve and Detectors
-VSB upstream of connection for Turb. t
'(New) f Plant sampling 309 Letdown Line to NU HU-V2A Utilize existing Rad.' Mc71 tor RM/L-1 Inline !
Purifkation -V2B located outside R.B.
Demineralizers (Existing) i 328 Pressurizer and CA CA-VI locate the monitor outside the R.B. Area Camma Reactor Coolant -V 2 - Detween the isolation valve and the {
Sample Lines Detector
-V1 samp?e cooler. (New)
-VI) 329 Reactor Coolant HU MU-V33A Pumps Seal Locate the online radiation monitor Area Camma
-338 downstream of the contatruent isola- Detector Peturn -33C tion valves outside of the R. ' B. for (New)
-3 3D Alarm Operator action is required to Or MU-V25 close valves.
-V26 330 Reactor Coolant WDC WDC-Y3
. Drain Tank locate the er nitor on the outside of Area Ge m as.d . Vent the tank.R.B. Strap monitor onto vent Detector
-V4 and drain lines are near each other 331 Reactor Coolant WDL (New)
WDL-V303 .
Drain Tank [
l Pmp Discharge -V304 336 Reactor Building AH AH-VIA Utiltre the existing purge outlet Inline Outlet and -V!B line Rad. Monitor RM/A-9 located ,
and Inlet Purge (Existing) '
-VIC outside of R.B.
423 Lines -Vi3 353 Reactor Building WDL WDL-V534 Sump Drain Locate an area radiation monitor Sump Area '
-V535 in the R.B. Sump mounted inside Monitor a seismically supported pipe. (New) 302 Intermediate Cooling IC IC-V2.3 locate the radiation monitor on the Incline 333 Supply & Return f and
-V4,6 6* IC return line between valve (Existing) 334 IC-V3 and the 2" pump rectre. line !
i Am. 23 l
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i l lO ! i l i i APPENDIX 2A i l i
- 1. BAW-1603, " Pilot Operated Relief Valve (PORV) Monitor Test," '
(Proprietary Report - forwarding letter only attached). ! l i 2. TDR-240, " Analysis of Two-Phase Flow Induced Differential Pres-l sure Acro:s TMI-1 Pressurizer Safety Valve Line Elbow Taps," l ! Rev. O. { 8 i O \ i I 8 l f t I i i i f r [s I e I 10 Am. 23 l i i
5.2.2 Operations and Maintenance Director, TMI-l g
- a. Function g)
's, The Director reports directly to the Vice President TMI-l and l assists him in the overail operation of TMI-1.
- b. Responsibility This position has direct responsibility for operating the unit in a safe, reliable and ef ficient manner; is responsible for of f site radioactive discharges and bears the respon-sibility for compliance with the operating licenses and the rules and regulations of the Commonwealth of Pennsylvania; supervises the Operations Group and Maintenance Group and the Radioactive Waste Processing and Shipment Group.
- c. Authority The authority of the Director, to act on behalf of the Vice President, TMI-1, is inherent in the position and commensurate l with the assigned responsibilities. It includes the authority to order the shutdown and cooldown of TMI-l whenever the health and safety of the public is endangered or when in his judgement a chutdown is warranted. It also includes the authority to 1 issue procedures, orders, and other directives required in the execution of the assigned responsibilities. Necessarily included in the responsibility for plant operation, compliance with Technical Specifications, is the authority to assign and prioritize requirements to the Plant Engineering, Training and
(')
\- Administration and Services Groups. Similarly, the authority of the Director includes the initiation and prioritization of corrective raintenance and preventative maintenance in the execution of his responsibilities. The Director may delegate his authority to the Ibnager, Plant Operation, IMI-l or Shif t Supervisor during absences. This delegation of authority extends to the issuance of standing orders and directives in support of the responsibilities assigned. In the absence or incapacitation of' the Vice President THI-1, this Director l is delegated the authority of that of fice for the centralized ,
control, supervision, coordination and planning of all aspects of TMI-l Operations.
- d. Minimum Qualifications The Operations and Maintenance Director, THI-1 shall possess l a Bachelors c egree in Science or Engineering and ten years of responsible power plant experience of which at least three years will be in nuclear power plant design, construction, startup, operation, maintenance, or technic.a1 services. A maximum of four years of the remaining seven may be fulfilled by academic training. This Director shall have acquired the l experience and training normally required for examination by the NRC for a SRO license whether or not the examination is taken.
( 5.2-6 Am.23
- f. Interfaces
- 1. Offsite 73
()
The Director interf aces with company, corporate, local Commonwealth of Pennsylvania, and Federal Government organizations in fulfillment of responsibilities assigned, State and Federal regulations, and directives received.
5.2.3 Manager, Plant Operations TMI-l
- a. Function The Manager, Plant Operations TMI-l has the responsibility l for directing the actual day-to-day operation of the unit.
He reports directly to the Operations and Maintenance Director, TMI-1. The Manager, Plant Operations TMI-l coordinates opera- l tions and related maintenance activities with the Manager, l Plant Maintenance TMI-1.
- b. Responsibility This position is respcasible for the day-to-day direction of the Operations personnel to ensure compliance with the conditions of the plant operating license and the Technical Specifications. He is also responsible for the supervision of the TMI-l Radioactive Waste Processing and Shipment
^ Group.
- c. Authority The Manager, Plant Operations 2MI-l has the authority l to shutdown and cooldown of TMI-I whenever the health and safety of the public is endangered or when, in his judgement, a shutdown is warranted.
- d. Minimum Qualifications The Manager, Plant Operations TNI-l will have a minimum of eight years of responsible power plant experience of which at least three years will be in nuclear power plant design, construc-tion, startup, operations, maintenance, or technical services.
A maximum of two years academic or related training may be included as part of the remaining five years of power plant experience. The Manager, Plant Operations TMI-l shall hold l a Senior Reactor Operators License. I
- e. Incumbent Qualifications Education: High School Graduate 1960 Military Service: U.S. Navy - 1960-1968 O 5.2-8 Am. 23
- ,,- - p,_ m i
~
/ \ .' w F1tiUPZ 5.3-1 GPU NOCLEAR CORPORATION l ~~l OFFICE OF TiiE PRESIDENT I I I I g President l 1 1 I I I
i l l Executive l
. ! Vice President i I L________ ________I Dairman - C.eneral Office Review Board.
i I I I I l vice President vice President Vice President Vice President _ Vice President Vice President Oyster Creek Tt4 I-1 WI-2 Technical Nuclear Administration Functions Assurance Operation Operation Operation Systems Qual. Assur.
- Fiscal Mgmt
- Maintenance Maintenance Maintenance
- Eng. & Design Training Human Resources Plant Engineeting . Plant Engineering
- Plant Engineering
- Licensing Nucle <sr Saf ety Assessment
- Security &
Decontamination Proj. Engineerin9 Facilities & Maintenance ' . Emergency Industrial Startup & Test Plar.ning Safety Construction __ Oyster Creek - Engineering Services System Lab. Iegal Services IOSRG - Forked River Vice President Vice President ~ Radiological & - Vice President Communications Environmental Maintenance & Controls Construction
- External Communication Hqtra Radeon
- Maint. & Constr. - Oyster Cred
- Internal Communication . Hqtra Environment Maint. & Constr. - TMI-l
- Site Fadeon .
Maint. & Constr. - m I-2
- Site Environment Am. 23
Further contributing to the availability and security of the liquid waste system is the fact that all of the above equipment is located within Seismic Class I structures that have been hardened to with-() stand an aircraf t impacc. Within these structures, all equipment that is anticipated to become a significant radiation source is housed within 2 to 3 foot thick shield walls for the protection of plant per-sonnel from radiation. The atmosphere of each of these shielded cubi-cles is maintained at a slightly lower pressure than that in surrounding areas to ensure that any radioactive gas leakage is away from plant per-s onnel. Based on the above indicated systems and equipment, the design basis waste liquid quantities generated annually are 49,000 Ci of mixed l is-sion products (excluding tritium) and 5.02 x 10 8 Ci oY tritium. With 17,500 gpm of the cooling tower effluent allocateE to Unit 1,~ Unit 1 annual discharge volume for 'which dilution credit may be taken is 3.48 x 1010 liters. This results in an annual average concentration of mixed fission products (excluding tritium) and tritium in the plant effluent of 1.4 x 10-6 Ci/ liter and 1.45 x 10-2u ci/ liter respectively. This annual average concentration of mixed fission products (including tri-tium) is within Appendix I to 10CFR 50 guidelines. 7 . 3 .1.1. 3 Epicor I Liquid Radwaste Treatment System 7.3.1.1.3.1 System Function and Design Objectives Existing plant equipment was not designed to process the quantity or radioactivity of the waste generated subsequent to the Three Mile Is-
'~1 land Unit 2 incident. A temporary custom-built externally located liquid radwaste treatment system, designated Epicor I, was installed to supplement the station's existing system.
The temporary system is designated to remove suspended and dissolved radioactive contaminants from liquid waste. Treatment is achieved through filtration and demineralization. Environmental protection is maintained by the use of features that provide leak and/or overflow protection. The discharge of radioactive gases is minimized. The system facilitates assembly and is flexible enough to conform to plant requirements and layout. 1 I ) O-7-9 Am. 23 i 1 t
.. - - - - _ . , - _ _ , - , _ _ _ _ , , , _ , , , , ~ , . . , , , . _ , _ - . . , . , , _ , - - . _ . . . -, _ , . - . . . . . . _ _ _ _ . . . _ _ _ , . _ _ _ _ _ , . . . , , _ . . . . _
- d. Af ter each complete or partial replacement of charcoal adsorber bank by verifying that the charcoal adsorbers
( remove > 99% of a halogenated hydrocarbon refrigerant (3 / test gas when they are tested in-place in accordance with AKS1 NS10-1975. 7.3.3.3 Implementation Schedule The testing schedule as previously described will be in place prior to the restart of Unit-1. 7.3.4 Nuclear Sampling 7.3.5 Nuclear Sampling Capabilities 7.3.5.1 Post-Accident Sampling N. See section 2.1.2.4.1 7.3,5.2 Improved In-Plant Radioiodine Monitoring Instrumentation See -section 2.1.2.1.1. .- 7.4 Affect of TMI-2 Recovery on TMI-l Operation Activities in TMI-l related to radwaste processing and activities in the comTn3n fuel' handling building will not be affected by n the THI-2 recovery program. As demonstrated in Section 7.2.3, C TMI-l does not have to rely on any T)tI-2 facilities for the processing of radwaste, Section 7.2.1 describes specifics to be taken to isolate 'the radwaste piping . systems of the two units. Through the isolation of the piping system, interface between the two unit's radwaste systems will be eliminated. Waste processing activities related to TMI-2 will be performed in the fuel handling building during the recovery program. These activities will not af fect activities in the auxiliary building because the areas will be separated by an environmental barrier (Section 7.2.2). Coutmuni-cation of the air spaces (Thi-1/TMI-2) of the fuel handling building will be minimized with appropriate modifications of the ventilation equipment in the building (See Supplement 1, Part 2, Question 52). Continuous access to the fuel handling building is not required for the safe operation of TMI-1 (with environmental barrier installed). 7-14 Am. 23
~
to Offsite dos:s are within the limits of 10CFR100. iii. Radially averaged enthalpy should not be greater than 280 , cal /gm at any axial location in any rod. !')
- 3. Mitigatio_n
- 1. The power excursion is limited by the Doppler coefficient.
ii. The power excursion is terminated by reactor trip on high pressure or high flux.
- 4. Conciusions The lower high pressure trip setpoint results in increased safety margins over the FSAR analysis. Improvements to the containment isolation signal (radiation +Rx trip) make release of fluid from the containment building less likely.
8.3.13 Feedwater Line Break Accident (TMI-2 FSAR, Section 15.1.8 S3-21.49, S 2-21.43, Reference 2, Q.3 of Supplement 1, Part 2)
- 1. Description This event has not been analyzed for TMI-1. The following description is based on FSAR analyses for TMI-2. . A loss of feedwater flow results in a loss of heat sink, primary system hcatup, increased pressurizer level and pressure, and reactor s trip on high RCS pressure. The TMI-2 analysis assumes a
) complete loss of feedwater due to a break upstream of the first feedwater line check valves. No analysis of loss of feedwater due to pump trip or valve closures were analyzed.
The loss of feedwater flow due to the postula.ed break is analyzed as an immediate ions of flow, which results in a bounding analysis for loss of feedwater events. The reactor is initially at 2772 Mw(t). Assumptions were made to provide two worst case scenarior-one for containment, and one for primary system conditio'.s. A double ended rupture (with a blowdown area limited by the feedwater header area) was analyzed; steam generators are l assumed to have a fouled inventory of e2,500 lbs. , and emergency feedwater is assumed to be 3: full flow within 40 seconds. The loss of feedwater is not directly calculated but taken as a conservative loss of heat demand (100-0% in 5 seconds for the affected generator and 100-0% in 20 seconds for the unaf fected generator). Reference 2 and Question 3, Supplement 1, Part 2 provide results for a loss of normal feedwater event. Table 8-2 compares the analysis assumptions to the plant design. 8-12 Am. 22
- 2. Acceptance Criteria e's i. Core thermal power shall not exceed 112% of rated power.
,b
- 11. Reactor cooJant system pressure shall not exceed code
, allowable limitr of 2750 psig. iii. Pressurizer does not become water solid during a loss i of feedwater transient.
- 3. Mitigation
- i. Reactor coolant system trip on hig pressure.
ii. The secondary system heat sink is restored by initiation of emergency feedwater to full flow within 40 seconds. Heat removal is through the turbine bypass valves or main steam relief valves.
- 4. Conclusions Results' of the TMI-2 f eedwater line break accident will become bounding for Unit I with the addition of a feedwater line break initiating signal. The addition of reactor trip on loss of feedwater increases the safety margin over the TMI-2 analysis. Lowering of the high pressure trip setpoint also increases safety margins since reactor trip will be initiated sooner. The RCS heatup is thus reduced.
))
PORV operation was not assumed in the fe line break analysis, so that the increase in the valve set, ; does not affect analysis results. The PORV would a .uate for the worst case feeCline break accident analyzed in the TMI-2 FSAR. Analyses are being performed by B&W to determine if a low level initiating signal would be suitable for auto EFW initiation during a feed line break accident. The B&W 4 analysis will take into account the lower thermal power level of TMI-1, as well as modeling of the OTSG in a more complete manner so that the heat demand will not be estimated so conaervatively. Furthermore, the back area will be based on the cross-sectional area of the feedwater injection nozzles rather than the cross-sectional area of the feedwater heater. This reduced blowdown area, along with the more detailed OTSG modeling is expected to reduce the severity of the feedwater line break substantially (refer to Appendix 8A, Figures 8A-21 and 22 for an example of expected results). As demonstrated by Table 8-2 and Q3 Supp 1, Part 2, TMI-1 acets the acceptance criteria for a loss of Feedwater transient and the analysis bounds the TMI-1 plant design. O 8-13 Am. 22
8.3.14 Waste Gas Decay Tank Rupture (FSAR Section 14.2.2.5) 4
- 1. Description G(-'g The rupture of a waste gas decay tank would result in radio-logical releases via the plant ventilation system. The tank contents as calculated assuming the activity evolved from degasing the primary coolant system af ter operation with 1%
failed fuel. .
- 2. Acceptance Criteria Doses shall not exceed the limits of 10CFR100.
- 3. Mitigation Elevated release of activity from the unit vent.
- 4. Conclusions This analysis has not been changed as a result of any plant modifications.
I 8.3.15 Small Break Loss of Coolant Accidents (LOCA)
- 1. Description
, vs Small break LOCA's are piping ruptures whose break areas (,,) range f rom as small as 0.005 it.Z to as large as 0.5 f t.2, These LOCA's may or may not involve depressurization of the Reactor Coolant System (RCS).
- 2. Acceptance Criteria
- 1. Local fuel cladding oxidation (metal water reaction)
, shall not exceed 0.17 times the total cladding thickness, or .05 the overall cladding mass. ii. Peak Cladding Temperature (PCT) shall not exceed 2200*F. l iii. A coolable geometry shall be maintained. iv. Long term cooling shall be assured.
- 3. Mitigation
- 1. Inventory will be maintained by the high pressure injec-tion system.
ii. Emergency Feedwater flow within 20 minutes of very small break LOCA's allows depressurization of the RCS and allows sufficient inventory addition by the HPI system to maintain core cooling. O 8-14 Am. 22
.. . . . - . . . _ - - . _ . - . - - - _ _ . - - . - . . _. .= - -
4 4
- 4. Conclusion i
j Pursuant to NRC regulations (10CFR50.46) and 10CFR50 Appendix () K) B&W performed generic LOCA analyses of their 177 f uel assembly lowered loop plants. Initially this sork was performed to meet the Interim Acceptance Critetia (IAC) and documented in ' BAW-10052. Later, the analyses were revised to ' the Final Acceptance Criteria (FAC) using the approved
- g. Appendix K model (BAW-10104). The FAC analysis results were documented in BAW-10103.
i The work performed for BAW-10052.was used as the basis for the small break LOCA Icoation and size sensitivity study and therefore no new work was performed for BAW-10103 other than analysis of three specific bEeak sizes and locations (0.04 l ft.2, 0.44 ft.2 and 0.5 ft.2 break sizes). 1 In April 1978, B&W identified an error in their ECCS model. l The error was also evident in the model used for the BAW-10052 sensitivity studies and therefore the basis for the accepta- ' bility of the small break analysis was eliminated. B&W
- . performed additional small break studies using the corrected '
model. The revised analyses are documented in a letter from J. H. Taylor, B&W to S. A. Varga, NRC dated July 18, 1978. These analyses cover break sizes 0.04 0.15, 0.2, 0.3, 0.13, and 0.17 ft.2, , .055, .07, .085, 0.1,
' Key assumptions for the small break LOCA analyses versus the
} TMI-1 plant design are given below:
BAW-10103 ' Ceneric TMI-1 i Reactor Power (MWt)- 2772 2335 Reactor Trip (psig) 1900 1900 , RC Pumps (LOOP) Coastdown Coastdown 1 1 i AFW Available** Yes-40 sec. Yes**** ESEAS HPI (psig) 1600 1600 , Operator Action Yes-c ross-connect none*** l
; EPI Distribution 70% to Core 70% to core j within 10 min. from time zero***
4 HPI Flow (gpm) 450 at 600 psig 500 at 600 psig***** i i I
** Amount assumed for generic analyses 550 gpm. The resonse to Supplement 1, Part 2 Question 4 demonstrates that 500 gpm is t
8-15 Am. 22
the minimum EFW required for TMI-1. TMI-1 is capable of delivering this minimum under the worst case single failure. ,~ Results of Reference 2 demonstrate that EFW is not required () before 20 minutes.
*** Prior to startup TMI-I will install HPI injection leg cross connects and flow control devices to eliminate operator action to cross connect HPI and equalize flow in all four injection legs.
**** For worst case LOCA in which offsite power is lost, EFW is initiated by the loss of feedwater or by the loss of 4 reactor coolant pump signals.
Also refer to the response to supplement 1, Part 3 Questions 1, 2 and 3. In all cases, TMI-1 plant specific information is as conservative or more conservative than the generic assumption. Since the TMI-2 accident, greater focus has been placed on small break LOCA's and the capability of the ECCS to mitigate them. Problems such as those discussed in Reference 21 (where the pressurizer stays full due to the loop seal arrangement despite loss of RCS inventory) have been addressed. These studies are documented in B&W's " Evaluation of TransienIt Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" May 7, 1979 (Reference 2). Breaks of 0.01, 0.02, and 0.07 ft.2 are analyzed utilizing varying assumptions on the availability and timing of AFW and HPI. These analyses use the same initial ,_, assumptions as used in BAW-10103 except that ESEAS is assumed to / \'s') occur at 1350 psig. Therefore, they are also bounding assumptions for TMI-l except for the distribution of HPI flow as discussed below. The analysis in Reference 2 also established that EFW flow is not required less than 20 minutes before any steam line break accident. In Reference 2, credit is taken for operator action to initiate HP I o r EFW. No mention is made as to whether whether operator action includes the time necessary to cross connect HPI as required in B&W's other small break accident analyses. TMI-l will com-plete the installation of permanent cross connection of the HFI prior to startup, therefore, operator action will not be neces-sa ry. All of the B&W small break LOCA analyses assume essen-tially equal backpressure for all four HPI injection points. This assumption is the basis for the 70%/30% flow split of HPI (assuming a single failure of one HPI train) between the core and the break respectively, after cross connection is accomplished. Such an equa? backpressure would not exist given an HPI line rupture. The back pressure on the broken HPI leg would be essentially zero and therefore the HPI loss out the break could be high resulting in inadequate injection to the core. The criterion established by B&W for the small break analysis requires that 70% of the total flow for one HPI pump be injected s us 8-16 Am. 22
into the broken legs of the reactor coolant system. This cri-teria applies to a 2772 MW thermal 177 fuel-assembly plant. 3 For TMI-1 with a licensed core power of 2535 MWt, the 70% - 30% {AJ criterion can be relaxed in direct proportion to the power reduction. This is justified based on the fact that the decay heat load following a small break LOCA is proportional to power and therefore cooling requirements will be directly proportional to the power at which the plant has operated. Therefore, for TMI-1, the acceptable flow split can be relaxed to 64% 6%. The 64%/36% flow split would not be obtained for an HPI lin e break as explained above. Therefore, operator action would be required to isolate the ruptured HPI line. The need to isolate could be determined by observing the individual flow indicators for the HPI legs. The high flow leg would then be isolated. This action would be contrary to the operators instinct and would require considerable judgment since the initial flow imbalance may not be dramatic. Since too great a chance for operator error (error of omission) exists, cavitating venturis will be added to the injection legs to limit flow in the broken leg. The venturis have been sized to limit flow in each leg to 137.5 gpm when only one high pressure injection pump is operating and Reactor Coolant System is at atmospheric pressure. The venturi design ensures that for the worst case HPI line break condition, the 64%/36% flow split can be achieved when Reactor System Pres-sure is less than 1500 psig. At RCS pressure conditions greater than 1500 psig, a flow split beyond the 64%/36% acceptance y-s criteria will occur. B&W has reviewed this situation and judged (_) the cavitating venturi performance is acceptable. This conclu-sion is based on the fact that under HPI line break conditions, the Reactor Coolant System will not expend significant time above 1500 psig and that during the time the RCS is above 1500 psig the cavitating venturi ensures that there is significant flow of high pressure injection into the RC system. B&W also notes that a much larger small break than a HPI line break sets the generic flow split criteria and therefore for a HPI line break the flow split criteria can be relaxed. l In addition to the benefits discussed above, the venturis provide two added benefits. First, they balance flow of the injection legs under all other small break conditions such that TMI-1 flow split will be within the bounds of the generic
- analysis (i.e. , 70%/30% flow split). SecondI7, the cavitating l can be relaxed.
1 ( S-17 Am. 22
8.3.16 Large Break Loss of Coolant Accidents (Reference FSAR Section 14.2.2.3) (} 1. Description Break sizes in the reactor coolant system (RCS) greater than 0.5 f t.2 are classified as large break loss of coolant accidents (LOCA's). These breaks involve rapid depressuri-zation of the RCS and are accompanied by rapid increases in containment pressure. Offsite doses are calculated from the design 5 asis radioactivity release to containment, and the design basis containment leak rate.
- 2. Acceptance Criteria
- 1. Peak fuel clad temperature does not exceed 2200'F.
ii. The corc is maintained in a coolable geometry. iii. Local fuel cladding oxidation (metal water reaction) shall not exceed 0.17 times the total cladding thick-ness of .05 times the total cladding mass. iv. Of fsite doses are within the limits specified by 10 CFR 100.
- 3. Mitigation
- 1. Core flood tank actuation at 600 psig to establish water 7')'N
( invento ry. ii. Low pressure injection system flow below 200 psig to , establish core cooling for the remainder of the accident. 111. Building spray addition to put iodine in solution with the containment water volume thus preventing release to the environment. iv. Containment leak tightness to limit radioactivity releases. l
- v. Switchover of the decay heat removal system suction source to the containment building sump on low-low BWST level.
- 4. Conclusion The calculated offsite dose resulting from the design basis l
LOCA will increase as a result of the deletion of sodium thiosulfate from the building pray system. Doses will still i be within the limits of 10 CFR 100. Dose calculations per-l formed for TMI-2 (see TMI-2 FSAR, Section 15 and Reference 5) demonstrate that design basis LOCA doses are within the limits of 10 CFR 100. The TMI-2 dose calculations were performed taking no credit for sodium thiosulfate. Since Unit 2 has a O 8-18 Am. 22
slightly large thermal pouer level and allowable containuent leak rate, then Unit 2 dose calculations conservatively bound the worst case LOCA dose for TMI-1. (v) Automated switchover of the BWST to the recirculation mode provides additional assurance that switchover will occur within the correct level band. Correct operator action had always been assumed in previous LOCA analyses. The automated switchover achieves the same function requirement by means of a safety grade control system. b.4
SUMMARY
AND CONCLUSIONS Plant modifications to TMI-1 allow the plant analyses to bound the expected plant behavior (see below). In some cases, analysis for TMI-2 have been referenced because they either analyze events that are not in the TMI-l FSAR (feedline break) or provide additional assurances of safety margins (steam line break).
- 1. Raising the PORV setpoint and lowering the high pressure trip setpoint affects all of the pressurization transients in the FSAR. Safety margins are improved since the high pressure trip setpoint has been lowered. No credit was taken for operation of the PORV, so that raising the valve setpoint has no effect on the FSAR analysis results.
The combined effect of the PORV and RPS setpoint changes vS are to decrease the probability of PORV operation. The (,,/ integrity of the primary coolant system will be challeng-ed less f requently, so that this change is in the conserva-tive direction. It should be noted that this codification could result in more f requent plant trips.
- 2. Reactor trip resulting from loss of feedwater results in Laproved safety margins for loss of feedwater events and does not degrade plant response for any accidents / transients.
- 3. Reactor trip as a result of turbine trip increases safety margins for the loss of feedwater or feed line break analy-ses. The effect of retaining or deleting plant features that permitted this event to occur without a reactor trip is being analyzed.
- 4. The addition of emergency feedwater initiating signals for the feedline break accident makes the TMI-2 f eedwater line break accident analysis bounding and conservative for TMI-1.
This event has additional safety margins beyond the TMI-2 analysis since both turbine and feedwater trips result in a reactor trip. This earlier reactor trip will result in a smaller heatup of the primary system. M (G 8-19 Am. 22
reactor trip. This earlier reactor trip will result in a smaller heatup of the primary system.
/~~T
(_) 5. Modifications to the high pressure injection system will al-low adequate HP1 flow for the spectrum of LOCA's. System per-formance is not degraded for any other accidents / transients in which HPI flow is initiated.
- 6. Upgrading of instrumentation inside containment assures that instrumentation will be functional in the postulated accident '
environments.
- 7. Automated switchover of the BWST to the recirculation mode provides additional assurance that switchover will occur within the correct level band. Correct operator action had always been assumed in previous LOCA analyses. The automated switchover achieves the same function requirement by means of a safety grade control system.
- 8. Dose calculation performed for TMI-2 demonstrate that the requirements of 10CFR100 3re met even af ter sodium thiosul-fate is deleted.
- 9. The transition to natural circulation following a complete loss of feedwater will be demonstrated by a startup test.
Reference 2 documents natural circulation tests and natural circulation events at B&W designed reactors. These tests and events demonstrate that natural circulation is a reli-
) able and effective means of core cooling.
w
- 10. An analysis of , loss of all AC will be performed as part of the B&W Owners Group ATOG program to determine what specific actions would be required to bring the plant to a safe shutdown condition.
- 11. A PORV setpoint of 2450 psig does not result in unacceptable interactions between the PORV and the pressurizer safety valves, whose setpoint is 2500 psig.
O 8-19a Am. 23
(\.,,) ((~) v ((v~) TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF 'INI-1 I. Loss of Offsite Power (LOOP) Purpose Results Case 1: 1840 gpm EFW lowered OTSG Show plant response to LOOP and Figure 8A-20 setpoi~t transition to natural circulation. Case 2: 1840 gpm EFW and 1% decay Examine overcooling potential Figure 8A-3 heat OTSG 1evel control at with 200*. EFW and minimum decay 12.5 ft. (max. rooldown case) heat. Case 3: Stuck open OTSG safety valves Examine plant performance for first Figure 8A-16 (17% of design flow) 10 minutes of secondary side depres-surization using one OTSG model. Case 3a: Same as Case 3, but with no Evaluate plant response to an inter- Figure 8A-2a EFW ruption of natural circulation caused by a loss of heat sink. Case 4: 500 gpm EFW using flow Examine long-term plant response with Figure 8A-8 limitation EFW flow limitation and LOOP from 100% power. Case 5: 1% decay heat with 1145 gpm . Examine effect of flow limiters in Figure 8A-4 EFW maximum and OTSG Ievel EFW system on max, cooldown case. control at 10 ft. Case S: EFW in superheat region Evaluate effect on transition to Figure 8A-14 natural circulation when model puts EFW in superheat region rather than downcomer. Shows more realistic plant response. L Case 10: No EFW Evaluate plant response to loss of Figure 8A-11 heat sink under natural circulation conditions. Case 15: 1% decay heat with 1840 gpm Show plant response with existing Figure 8A-13 EFW and OTSG 1evel control level setpoint and full EFW flow. at 20 ft. Am. 23
TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF BtI-1 (Continued) II. Station Blackout Purpose Results Case 1: Base Look for long-term plant response Figure 8A-7 to event, including voiding in the RCS LOOP. Case 2: 1 gpm pressurizer leakage Effect on plant response due ;o Figure 8A-12 and letdown isolation at cooldown of pressurizer steam 20 minutes space. Maximizes expected cooldawn. III. Loss of Feedwater Case 1: 460 gpm EFW flow and 1.2 Examine plant response to operation Figure SA-9 ANS decay heat of only one motor-driven EFW pump. Case la: 460 gpm EFW flow and .85 Figure SA-10 ANS decay heat Case Ib: 460 gpm EFW flow and 1.0 Figure 8A-24 ANS decay heat, no pres-surizer spray, reactor. trip on high pressure. Case 2: EFW flow limitation Look at plant transition to stable Figure SA-6 shutdown with EFW flow limited. Case 3: Figure SA-5
~
Base case 1840 gpm EFW and Show plant response to LOFW with no no equipment failures equipment failures and no operator actions. Case 4: Failure of EFW 1evel control Evaluate effect of continuous EFW Figure 8A-1 and l'6 ANS decay heat flow causing overcooling of OTSG. Am. 23
0 (O (O TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF TMI-1 (Continued) 111. Loss of Feedwater (Cont'd.) Purpose Results Case 8: No EFW Look at time available before llPI Figure 8A-19 must be initiated. Case 11: Loss of feedwater with trip Quantify benefit of anticipatory Figure 8A-15 on high pressure reactor trip. Case 14: 460 gpm EFW and 1.0 ANS de- Figure 8A-17 cay heat. EFW trip on low OTSG 1evel, spray on to de-lay llP trip. IV. Feed Line Break Accident Case 1: EFW initiation on feed / Demonstrate plant can tolerate Figure 8A-21 steam AP, 500 gpm the design basis.feedline break accident. Case 2: Delayed EFW Demonstrate that delay of EFW for Figure 8A-22 greater than 40 seconds still al-lows an acceptable plant response. Bounds analysis in which EFW is initiated on low OTSG 1evel signal. Am., 23
TABLE 8A-2 /~ ; LOSS OF OFFSITE POWER CASE 1 \_) MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 1840 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE
' SMALL SAFETIES : 1060/986 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA f ') PRESS HEATERS : 1 BANK v
PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 1 OFFSITE PONER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F PRESS MODEL : NON-EQUILIBRIUM Am. 23
TABLE 8A-3
/~N LOSS OF OFFSITE POWER CASE 1B U
- t k
MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE i EFW STEAM : AVAILABLE EFW CAPACITY : 1840 GPM i ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE l SMALL SAFETIES : 1060/086 PSIA. BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA j BANK 3 : 1072/997 PSIA (} PRESS HEATERS : 1 BANK PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP
'RPS TRIP DEFEAT : NONE f SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE l
DIESEL GENERATORS : 1 OFFSITE POWER : UNAVAILABLL DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F f) PRESS MODEL : NON-EQUILIBRIUM 1 Am. 23
m . .-. _ _ __ . _ _ _ __ ._____ _ _ ____ _ _ _ _ . . . . . . . . . .. _. i TABLE 8A-4 7 LOSS OF OFFSITE POWER CASE 2 l 4 MAKEUP : AVAILABLE ? LETDOWN : ISOLATED @ T = 5 SEC 1 1 EFW MECH : AVAILABLE ! EFW STEAM : AVAILABLE EFW CAPACITY 1840 GPM ATMOS DUMP : AVAILABLE 9 1025 PSIA 1 TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/1028 PSIA j BANK 1 : 1065/1033 PSIA 4 PRESS HEATERS : 2 BANKS (126 KW) . PRESS SPRAY : UNAVAILABLE !. I
- ps RC PUMPS
- COAST DOWN BEGINNING 9 T=O SEC i i \-)
-; RPS l' RIP : LOOP l RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 i
- OFFSITE POWER : UNAVAILABLE i
[ DECAY HEAT : 0.01 X ANS 5.1, 1971
- 1 : 2450/2400 PSIG i
PRESS SAFETY : 2500/2475 PSIG i
- DOPPLER COEF
- -8.4047 E-4 $/F i
i MODERATOR COEF : -4.4027 E-3 $/F l PRESS MODEL : NON-EQUILIBRIUM '(:). i Am. 23 L__________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
---___ m. . _ _- - _ - - _ - - _ _ _ _ _ --__-- _._ ____ __ -_ . _ . . . _ _ __
i 4 TABLE 8A-5 LOSS OF OFFSITE POWER CASE 3 O l MAKEUP : AVAILABLE LETDOWN : ISOLATED 0 T = 5 SEC i l EFW MECH : AVAILABLE EFW STEAM : AVAILABLE 1 EFW CAPACITY : 1840 GPM ATMOS DUMP : AVAILABLE O 1025 PSIA TURBINE BYPASS : AVAILABLE 1 SMALL SAFETIES : 1060/590 PSIA BANK 1 : 1065/590 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA PRESS HEATERS : 2 BANKS ( PRESS SPRAY : UNAVAILABLE { RC PUMPS : COAST DOWN BEGINNING 0 T=0 SEC i j RPS TRIP : LOOP ! RPS TRIP DEFEAT : NONE I l SPAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE ! DIESEL GENERATORF : 2 l OFFS 7TE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG i DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F .O Am. 23
TABLE 8A-6
/~T LOSS OF OFFSITE POWER CASE 3A 'v' i
MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 0 GPM ATMOS DUMP : hVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE SMALL SAFEPIES : 1060/14.7 PSIA BANK 1 : 1060/14.7 PSIA BANK.2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA
/~h PRESS HEATERS : 2 BANKS \_)
PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 0.85 X ANS 5.1, 1971
- PORV : 2450/2400 PSIG t
PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-7 (~} LOSS OF OFFSITE POWER CASE 4 V MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 490 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1060/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA l() PRESS HEATERS : 2 BANKS (126 KW) PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE j SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG t PRESS SAFETY : 2500/2475 PSIG i i DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-B LOSS OF OFFSITE POWER CASE 5 { MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : A/AILABLE EFW STEAM : AboILABLE EFW CAPACITY : 1140 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILAQLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1060/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA (} PRESS HEATERS : 3 BANKS PRESS-SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE FOWER : UNAVAILABLE DECAY HEAT : 0.01 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E~4 $/F i MODERATOR COEF : -4.4027 E-3 $/F Am. 23 _ _ _ _ _ _ . . _ .. . _ _ . _ , . _ , , . _ , _ ~ _ . , . _ . _ . _ , _ . _ _ . . . _ . _ _ , _ . _ . - _ . _ . , . . _ . , . _ _ _ - . _ -
s TABLE 8A-9 i LOSS OF OFFSITE POWER CASE 5A MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC A EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 1140 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/986 PSIA
- BANK 1 : 1060/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA
(} PRESS HEATERS : 1 BANK PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=O SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 i-i OFFSITE POWER : UNAVAILABLE DECAY HEAT : 0.01 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F , ( Am. 23
TABLE 8A-10 LOSS OF OFFSITE POWER CASE 8 MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE i EFW CAPACITY : 1840 GPM ATMOS DUMP : AVAILABLE @ 1025-PSIA TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA (} PRESS HEATERS : 3 BANKS PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 i PORV : 2450/2400 PSIG ., PRESS SAFETY : 2500/2475 PSIG i f DOPPLER COEF : -8.4047 E-4 $/F l 1 i MODERATOR COEF : -4.4027 E-3 $/F ' O 4 Am. 23 y -w,--.s- a g- -
-,-y,,,.,, e,n-,w--,, -- --,,,qw..,-n, w ,.--,-cr ,
7,,,-,-we , -v e ,,y--w,, ,,,-w,,,w-.,,-,v,-~.>p ,~m,y,nwe,m.,s-,w,,,,,r--.a.,,,,w ,-e.e ,,,,-w.e, , - ,
- _ __ . ~ _ _ _ _ . _ .. - _ . . - _ _ _ . . _ _ . m ._ . _
i
; TABLE 8A-11 LOSS OF OFFSITE POWER CASE 9
)
f MAKEUP : AVAILABLE J LETDOWN : ISOLATED @ T = 5 SEC 1 EFW MECH : UNAVAILABLE j EFW STEAM : AVAILABLE EFW CAPACITY : 490 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA c TURBINE BYPASS : AVAILABLE SMALL SAFETIES : ~1060/1028 PSIA A BANK 1 : 1065/1033 PSIA s { BANK 2
- 1070/1038 PSIA l BANK 3 : 1072/1040 PSIA
- j. PRESS HEATERS l
- 2 BANKS
+ PRESS SPRAY : UNAVAILABLE 4 RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC i RPS TRIP : LOOP RPS TRIP DEFEAT : NONE i 4 SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE 3 DIESEL GENERATORS : 2 4 1 OFFSITE POWER : UNAVAILABLE 1 DECAY HEAT : 1.0 X ANS 5.1, 1971 r. PORV- : 2450/2400 PSIG , PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F i MODERATOR COEF : -4.4027 E-3 $/F !o . Am'. 23 i,-..---..-_.--,,4-,.-----.-,---,
~ _ , , . - - , . - - - , . _ , , , - , . . . . . - , _ . . . . .- . . - . _ , , , - - , .-r--- ,_., -..-,
TABLE 8A-12 LOSS OF OFFSITE POWER CASE 10 MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : UNAVAILABLE
- EFW STEAM
- UNAVAILABLE EFW CAPACITY : 0 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1060/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS : 2 BANKS PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP 3
RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG i DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F 4 Am. 23
TABLE 8A-13 /~% LOSS OF OFFSITE POWER CASE 11 (/ MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T w 5 SEC EFW MECH : UNAVAILABLE EFW STEAM : UNAVAILABLE EFW CAPACITY : 0 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : UNAVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA (} PRESS HEATERS PRESS SPRAY 1 BANK UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=O SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : OPEN @ T=0 SEC, CLOSE @ 600 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F l( fl NODERATOR COEF : -4.4027 E-3 $/F Am. 23
I TABLE 8A-14 gg LOSS OF OFFSITE POWER CASE llA l (/ ! MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : UNAVAILABLE EFW STEAM : UNAVAILABLE EFW CAPACITY : 0 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : UNAVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA {} PRESS HEATERS PRESS SPRAY 1 BANK UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=O SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 1 l OFFSITE POWER : UNAVAILABLE LECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : OPEN @ T=0 SEC, CLOSE @ 600 PSIG PRESS SAFETY : 2500/2475 PSIG l D Ov P'_ _"D COEF : -8.4047 E-4 $/F () MODERATOR COEF '
-4.4027 E-3 $/F Am. 23
TABLE 8A-15 LOSS OF OFFSITE POWER CASE 15 MAKEUP : AVAILABLE LETDOWN : ISOLATED-@ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 1840 GPM ATMOS DUMP : AVAILABLE @ 1025 PSIA TURBINE BYPASS : UNAVAILABLE SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA (~T PRESS HEATERS : 1 BANK V PRESS SPRAY : UNAVAILABLE RC PUMPS : COAST DOWN BEGINNING @ T=0 SEC RPS TRIP : LOOP RPS TRIP DEFEAT : NONE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 1 OFFSITE POWER : UNAVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F O Am. 23
._ _ _ -. _. ..- ,.. _ . __. ,_ .__ ..- .... ~ _-.-_.._ _ . _._... _ _ _ , . _ _ - . _ . . _ , , _ - . . . _ . . . . _ - - . - - , _ - _ . _ . _ . . .
TABLE 8A-16 FEEDWATER LINE BREAK CASE 1 MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : UNAVAILABI.E LOOP A AVAILABLE LOOP B EFW STEAM : UNAVAILABLE LOOP A AVAILABLE LOOP B J EFW CAPACITY : 0 GPM LOOP A 500 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA
\'O/ BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA)
RC PUMPS : AVAILABLE RPS TRIP : HIGH PRESSURE RPS TRIP DEFEAT : VARIABLE LOW PRESSURE & TURBINE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 1.2 X ANS 5.1, 1971 PORV : 2450/2400 PSIG l PRESS SAFETY : 2500/2475 PSIG l DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-17, FEEDWATER LINE BREAK CASE 1A l MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5'SEC EFW MECH : UNAVAILABLE LOOP A AVAILABLE LOOP B EFW STEAM : UNAVAILABLE LOOP A AVAILABLE LOOP B EFW CAPACITY : 0 GPM LOOP A 500 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILADLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA 4 (} BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA) RC PUMPS : AVAILABLE l RPS TRIP : TURBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 i OFFSITE POWER : AVAILABLE DECAY HEAT : 1.2 X ANS 5.1, 1971 t
/'T l (_j PORV : 2450/2400 PSIG I
f PRESS SAFETY : 2500/2475 PSIG i l DOPPLER COEF : -8.4047 E-4 $/F 1 MODERATOR COEF : , - 4. 4 0 27 E-3 $/,F __ Am. 23
TABLE 8A-18 FEEDWATER LINE BREAK CASE 1B d(~N MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : UNAVAILABLE LOOP A AVAILABLE LOOP B EFW STEAM : UNAVAILABLE LOOP A AVAILABLE LOOP B l EFW CAPACITY : 0 GPM LOOP A 550 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 990 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 1065/1033 PSIA _( ) : BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS. : 5 BANKS PRESS SPRAY : AVAILABLE (z220/2170 PSIA) RC PUMPS : AVAILABLE RPS TRIP : TURBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1600 PSIG RCS PRESSURE , SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F
. 23 L MODERATOR _COEF : -84@fDLJR-X_ft&
E TABLE 8A j FEEDWATER LINE BREAK CASE 1C O MAKEUP- : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC i
'EFW MECH : UNAVAILABLE LOOP A AVAILABLE LOOP B )~
EFW STEAM : UNAVAILABLE LOOP A AVAILABLE LOOP B EFW CAPACITY : 0 GPM LOOP A 550 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA {~ 1 BANK 3 :- 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY ': AVAILABLE (2220/2170 PSIA) i RC PUMPS : AVAILABLE i RPS TRIP : HIGH PRESSURE RPS TRIP DEFEAT : VARIABLE LOW PRESSURE & TURBINE
-SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE 1
DIESEL GENERATORS : 2 i OFFSITE POWER : AVAILABLE DECAY HEAT : 1.2 X ANS 5.1, 1971 PORV : 2450/2400 PSIG l() PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF. : -8.4047 E-4 $/F . - - . ' MODERATOR COEF : -4.4027 E-3 $,F Am. 23
- - , . , - . . . . . _ - - _ . _ _ , . . . . _ . - . _ - . . . . , _ . . . _ . - . _ . _ . . . _ . _ _ . _ _ . . . , . . - . _ . , - ~ . . _ , . _ . _ . . - . . . _ . -
I TABLE 8A-20 1 f- FEEDWATER LINE BREAK CASE 1D ()g MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 720 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1000 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA 1 BANK 2 : 1070/1038 PSIA , BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS pd PRESS SPRAY : AVAILABLE (2220/2170 PSIA) l RC PUMPS : AVAILABLE , RPS TRIP : TURBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE j DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG ! DOPPLER COEF : -8.4047 E-4 $/F
'() MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-23 FEEDWATER LINE BREAK CASE 1F { }- MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC
- EFW MECH : UNAVAILABLE LOOP A l
AVAILABLE LOOP B EFW STEAM : UNAVAILABLE LOOP A AVAILABLE LOOP B EFW CAPACITY : 0 GPM LOOP A l 550 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA () BANv 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA)
- RC PUMPS
- AVAILABLE RPS TRIP : TURBINE RPS TRIP DEFEAT : NONE SEAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE i DIESEL GENE AATORS : 2
( OFFSITE POWER : AVAILABLE DECisY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF. : -8.4047 E-4 $/F Am. 23 i MODERATOR COEF _----. : _-4.4027 E-3 $/_F. - - - . . _ . _ . . . . - . , _ _ _ - , _ __..._.. __ , , . _ . . . _ . . _ ~ -
TABLE 8A-22 FEEDWATER LINE BREAK CASE 2A MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : UNAVAILABLE EFW STEAM : UNAVAILABLE LOOP A
- UNAVAILABLE LOOP B EFW CAPACITY : 250 GPM LOOP A 0 GPM LOOP B ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA
{} BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESF HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE, OFF AT REACTOR TRIP RC PUMPS : AVAILABLE RPS TRIP : TURBINE l RPS TRIP DEFEAT : NONE j SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILAl ' DECAY HEAT : 1.2 X ANS 5.1, 1971 PORV : 2450/2400 PSIG () PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F I MODERATOR COEF : -4.4027 E-3 $/F Am. 23
i TABLE 8A-23 1 LOSS OF NORMAL FEEDWATER CASE 1 MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : 1 MDP AVAILABLE EFW STEAM : 'AVAILABLE EFW CAPACITY : 460 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA l BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA
/~ PRESS HEATERS : 5 BANKS V}
PRESS SPRAY : AVAILABLE (2220/2170 PSIA) RC PUMPS : AVAILABLE RPS TRIP : TURBINE 1 RPS TRIP DEFEAT : NONE SFAS TRIP : 1500 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE
- DECAY HEAT
- 1.2 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F
; h) u MODERATOR COEF : -4.4027 E-3 $/F Am. 23 l
1 TABLE 8A-24 LOSS OF NORMAL FEEDWATER CASE lA MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : 1 MDP AVAILABLE EFW STEAM : UNAVAILABLE EFW CAPACITY : 460 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SHALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA (} PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA) RC PUMPS : AVAILABLE RPS TRIP : TURBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 0.85 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F () MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-25 LOSS OF NORMAL FEEDWATER CASE 1C [vD J MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE EFW STEAM : .AVAILABLE EFW CAPACITY : 460 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA-SMALL SAFETIES : 1060/986 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA gs PRESS HEATERS : 5 BANKS ^ L.) PRESS SPRAY : AVAILABLE @ T = 600 SEC RC PUMPS : AVAILABLE RPS TRIP : HIGH PRESSURE (2405 PSIA) RPS TRIP DEFEAT : TURBINE TRIP SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 3.9 X ANS 5.1, 1971 PORV : 2900/400 PSIG t PRESS SAFETY : 2500/2475 PSIG , DOPPLER COEF : -8.4047 E-4 $/F ()- MODERATOR COEF : -4.4027 E-3 $/F i Am. 23
, - . . ,.-,-r.,.- - m., - . . _ - ., -.
, , . , , ~ . _ . . . . . - - - . , , . - , ,,.,-,% , --, ,, ,y e,... ., , , _ - , ,.---e-
TABLE 8A-26
-- LOSS OF NORMAL FEEDWATER CASE 2 MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : AVAILABLE EFW STEAM : UNAVAILABLE EFW CAPACITY : 460 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA)
RC PUMPS : AVAILABLE RPS TRIP : TRUBINE RPS TRIP DEFEAT :- NONE SFAS TRIP : 1500 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F () MODERATOR COEF : -4.4027 E-3 $/F Am. 23
TABLE 8A-27 LOSS OF NORMAL FEEDWATER CASE 3 O v MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 0 SEC EFW MECH : AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 460 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025-PSIA SMALL SAFETIES : 1060/1028 PSIA BANK 1 : 1065/1033 PSIA BANK 2 : 1070/1038 PSIA BANK 3 : 1072/1040 PSIA PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA) RC PUMPS : AVAILABLE RPS TRIP : TRUBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1500 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFESITE POWER : AVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F [a'T - MODERATOR COEF : -4.4027 E-3 $/F Am. 23
4 TABLE 8A-28
/~T LOSS OF NORMAL FEEDWATER CASE 4 V
MAKEUP : AVAILABLE LETDOWN- : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE I-EFW STEAM : AVAILABLE EFW CAPACITY : 1145 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/986 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA
- i
() PRESS HEATERS : 5 BANKS PRESS SPRAY : AVAILABLE (2220/2170 PSIA) i RC PUMPS : AVAILABLE I
- RPS TRIP
- TRUBINE RPS TRIP DEFEAT : NONE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE
( DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 0.01 X ANS 5.1, 1971 PORV : 2450/2400 PSIG i PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F f (~x l \_) MODERATOR COEF : -4.4027 E-3 $/F Am. 23-
TABLE 8A-29 s LOSS OF NORMAL FEEDWATER CASE 8
%-) ,
MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 50 SEC EFW MECH : UNAVAILABLE EFW STEAM : UNAVAILABLE EFW CAPACITY : 0 GPM
~
ATMOS DUMP : 1065/1035 PSIA TURBINE BYPASS : 1065/1035 PSIA SMALL SAFETIES : 1065/1012 PSIA BANK 1 : 1080/1026 PSIA BANK 2 : 1090/1036 PSIA BANK 3 : 1117/1061 PSIA fs PRESS HEATER 3 : 5 BANKS N] PRESS SPRAY : AVAILABLE RC PUMPS : AVAILABLE RPS TRIP : TRUBINE RPS TRIP DEFEAT : NONE l l SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 l OFFSITE POWER : AVAILABLE l l DECAY HEAT : 0.65-X ANS 5.1, 1971 PORV : 2470/2245 PSIG PRESS SAFETY : 2450/2325 PSIG l DOPPLER COEF : -7.100 E-4 $/F MODERATOR COEF (_~ ) : -3.040 E-3 $/F ha. 23
TABLE 8A-30 v"N~ LOSS OF NORMAL FEEDWATER CASE 11 b MAKEUP : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC EFW MECH : AVAILABLE-EFW STEAM : AVAILABLE EFW CAPACITY : 1840 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE @ 1025 PSIA SMALL SAFETIES : 1060/986 PSIA BANK 1 : 1065/990 PSIA BANK 2 : 1070/995 PSIA BANK 3 : 1072/997 PSIA PRESS HEATERS : 5 BANKS
)
PRESS SPRAY : AVAILABLE (2220/2170 PSIA) RC PUMPS : AVAILABLE RPS TRIP : HIGH PRESSURE RPS TRIP DEFEAT : TURBINE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 1.0 X ANS 5.1, 1971 PORV : 2450/2500 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.404' E-4 $/F k MODERATOR ' COEF : -3.4027 E-3 $/F Am. 23 I
TABLE 8A-31 LOSS OF NORMAL FEEDWATER CASE 14 ( )4 - MAKEUP. : AVAILABLE LETDOWN : ISOLATED @ T = 5 SEC' EFW MECH : 'AVAILABLE EFW STEAM : AVAILABLE EFW CAPACITY : 500 GPM ATMOS DUMP : AVAILABLE TURBINE BYPASS : AVAILABLE SMALL SAFETIES : 1060/986 PSIA BANK 1 : 1065/990 PSIA BANK 2 : '970/995 PSIA BANK 3 : 1072/997 PSIA ( ))^ PRESS HEATERP : 5 BANKS PRESS SPRAY : AVAILABLE, OFF ON REACTOR TRIP RC PUMPS : AVAILABLE RPS TRIP : HIGH PRESSURE APS TRIP DEFEAT : TURBINE SFAS TRIP : 1600 PSIG RCS PRESSURE SFAS TRIP DEFEAT : 41 , CONTAINMENT PRESSURE DIESEL GENERATORS : 2 OFFSITE POWER : AVAILABLE DECAY HEAT : 1.2 X ANS 5.1, 1971 PORV. : 2900/400 PSIG PRESS SAFETY : 2500/2475 PSIG DOPPLER COEF : -8.4047 E-4 $/F MODERATOR COEF : -3.4027 E-3 $/F Am. 23-- .______________:___________________1______________________________
SUPPLEMENT NO.1, PART N0,1 O V OUESTION:
- 55. Bulletin 05B Item 1 Your procedure 1102-16, Natural Circulation, includes anticipatory filling of the OTSG prior to securing the reactor coolant pumps. Submit the analy-sis performed to provide guidance as to the expected system response.
RESPONSE
The attached analysis provides guidance regarding system response to anticipa-tory steam generator fill transients. The analysis provides:
- 1. Calculated OTSG fill rate as a function of main feedwater or EFW flow. Feedwater temperature is accounted for as a func-tion of equivalent primary system heat load.
- 2. The response of various B&W 177 facilities to OTSG fill tran-sients.
The analysis feedwater temperatureshows a (as strong dependence would be expectedof. eq)uivalent The attachedprimary analysessystem provide heat gui-load on Q b dance as to the expanded systems response to anticipatory filling of the OTSG. The following TMI-l operator guidelines were developed using these analyses: OPERATOR GUIDELINES FOR FILLING THE OTSG'S TO 50% ON THE OPERATE RANGE LEVEL WITH THE MAIN FEEDWATER SYSTEM
- 1. When main feedwater is used to fill tile steam generator to 50% with forced RCS flow, feed only through the startup valve. Control the rate of ficw to maintain OTSG pressure within 100 psi of the control setpoint.
- 2. If feedwater is below 100*F, fill one OTSG at a time.
- 3. If forced RCS flow is lost before reaching 50% in both steam generators, feed with emergency feedwater using the EFW guidelines.
OPERATOR GUIDELINES FOR FILLING THE OTSG'S TO 50% ON THE OPERATE RANGE LEVEL WITH THE EMERGENCY FEEDWATER SYSTEM
- 1. When emergency feedwater is used to fill the steam generators to 50% with forced RCS flow, then the rate of fill must be controlled in order to main-tain the OTSG pressure within 100 psi of the control setpoint. Throttle EFW flow as required to control OTSG pressure.
- 2. Use either two (2) motor driven pumps or one (1) turbine driven pump.
Am. 23
. SUPPLEMENT NO. 1. PART NO. 1 O PAGE NO. 2 TO QUESTION NO. 55
- 3. If forced RCS flow is lost before reaching 50% level in the OTSG's, observe
; the following restrictions:
A. Steam generator level should always be increasing. B. EFW flow should be continuous until the setpoint is l reached. C. EFW flow should be turned on full if natural circula-tion stops. 4
. O .
I } i 1 I I i lO i Am. 23 i
~
ATTACHMENT TO SUPPLEMENT NO. 1, PART NO. 1, , QUESTION NO. 55 O ana'vsrs 0F sTexa ceaea^T0a aartcre4T0av era Taaasitars , The analysis of the anticipatory fill transient was divided into two sections. The first evaluated the influence of fast fill rates on the RC system using main feedwater, and the second section examined the effect of cooling the primary loop when filling steam generators with the emergency feedwater system. i In the first section, changes in primary loop conditions were obtained for main feedwater with various temperatures and flow rates. Predictions were , substantiated by several instances of recorded 177 fuel assembly plant responses to steam generator fills using the main feedwater system. Figure i shows the relaticnship of MFW flow rate per steam generator i for observed fill rates using the start-up level indication. If the main feedwater system start-up valves are used, assuming a maximum flow rate of 1.5 million lbs/ hour, then fill rates greater than 60 inches per minute could occur in each steam generator. The effect of MFW temperature O wes inci ded eed covered e reese of te Pereture between 200 eed 400 F. Direct heating of the main feedwater by aspirated steam was assumed and in some instances the flow rate of steam condensed was approximately 35% of the MFW flow rate. If the heat transferred from the primary loop , to the secondary loop is sufficient, then an identical amount of steam will be vaporized from the volume of water to restore the condensed steam and maintain a nearly constant steam pressure. The heat required to exactly replace the condensed steam has been calculated and is shown in Figure 1 as a percentage of the full rated power. As a comparison, a typical upper limit to decay heat plus RC pump power is shown as a heavy line at approximately 5%. For MFW temperatures of 300 to 400*F, a fill rate of 60 inches per minute (as indicated by the start-up level) would lead to a negligible change in steam pressure if sufficient decay heat is available. As MFW temperature decreases toward 200 F, more heat is required from the steam and the primary loop and smaller fill rates are necessary in order to maintain proper steam pressure. For very low MFW temperaturet, it would be preferable 0 to riii o"ir o"e stee= 9emeretor et e time-Am. 23
86-1120427-02
~
O Fi9eres 2 end 3 disnier neeriy ideucicei nieat resnease et two simiier B&W plants for large rates of steam generator filling with the main feedwater system. In each case, main feedu ater flow was not terminated following the reactor trip and both steam generators were filled above the 20 foot level before MFW flow was manually tenninated. The flow rate of MFW is quite large, being either 1.5 or 3.0 million 1bs/ hour, and each steam generator was filled at rates of approximately 120 inches per minute. The important fact to notice is that steam pressure dropped from roughly 1000 psig to 950 psig or higher during the fill operation and the effect on primary loop cooldown was very small. If the MFW temperature had been lower than 435 + 20 F, a larger decrease in steam pressure would have been anticipated. (Actual plant data for low MFW temperature is not available.) Figure 4 exhibits the transient behavior of the reactor coolant system, namely RC pressure and pressurizer level, for the two events that occurred at the Oconee Nuclear Station. .At approximately one minute of filling the two steam generators with main feedwater, minimum values of PC pressure and pressurizer level were reached. This corresponds to the . time that the desired level was achieved and main feedwater flow rates were reduced to zero. The calculated values shown in Figure 1 appear to be conservative in comparison with real plant response. Since the calculations did not account for the heat stored in the metal of the steam generators, larger fill rates than predicted should be possible before a decrease in steam pressure would occur which would lead to cooling in the RC system. In the second section of the analysis, a similar evaluation of steam generator performance during an emergency feedwater fill transient was performed. Both calculations and plant data reveal that steam pressure responds immediately to any imbalance between the direct heating of the near ambient temperature EFW by the steam and the heat supplied from the primary loop. q V Am. 23
86-1120427-02 O Fioure 5 preseats e comperison of ectuei chenses ia steem senerator pressere due to selected fill rates using either main or emergency feedwater systems. Though all the plant data for EFW was recorded with all RC pumps shutdown, the effect of an additional 1/2 to 1% heat load due to running RC pumps would change the slope of the curve. However, the situation of RC pumps running and the operator selecting EFW to fill the steam generator has occurred very infrequently compared to either no RC pumps (and MFW pumps) [ and EFW system operation or RC pumps running and the MFW system used to fill ' both steam generators. , Figure 6 shows the severe impact that excessive EFW flow into two steam generators simultaneously has on the reactor coolant system. This was an unexpected loss of station power while the plant was operating at : l 407, power level. The EFW system w.i initiated and filled each steam generator at approximately 1200 gpm to a 100 to 120 inch level. In four i minutes, the pressurizer level had decreased below a zero indication and the operators terminated the wide open filling of the second steam generator prior to reaching a 120 inch level. i Figure 7 shows a similar loss of power event that was conducted at TMI-2 on April 22, 1978 as a test. The EFW system was used to fill both steam generators appt' imately equally with a flow rate of about 500 gpm I to each generator. TL ugh the fill operation was not complete at eight minutes into the test, minimum values cT P1 pressure and pressurizer ! level were reached at that time. The severity of the impact on the RC ! system was quite small but may have been aided by reduced letdown flow and increased makeup system flow, actions usually taken by the operator after a reactor trip.
! Figure 8 shows the EFW flew rate to one steam generator corresponding to observed fill rates on the start-up level. For the assumption of steam pressure at 900 psig and steam directly heating the EFW to a saturation temperature liquid, the relationship of fill rate and EFW flow ate is presented and compared to the 500 gpm flow rate limited by the cavitating venturis. As calculated before for MFW, the primary loop heat required i to replace all condensed steam and maintain constant steam pressure is also Am. 23
86-1120427-02 O shown in Figure 8. The dotted lines show that a total heat load of 3% full rated power is required to maintain steady conditions within both steam generators when the total EFW flow is 1000 gpm. The effect of variable EFW temperature is small since the expected range of EFW temperature is only about 50*F. If the operating history of the reactor is low at the time of the reactor trip, then the decay heat will be low and it will be insufficient to match the " wide open" flow rate of EFW of 500 gpm. Thus, each decay heat level establishes a band of allowable EFW flow rates that can be used for filling the steam generators to the desired level. Since these are not known at the time of reactor trip, the operator should manually control EFW flow to maintain steam pressure greater than 100 psi below the control setpoint for steam generator oressure. Then satisfactory conditions will be maintained in the RC system. If there is a choice between u< ' gi MFW or EFW to achieve 20 feet of water in each steam generator before tripping the last RC pumps operating, the operator should use main feedwater. There are fewer restrictions and less manual control required than for the emergency feedwater system. m Am. 23
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