Text

Amend 15 to Advanced BWR Ssar
	ML20065L313
Person / Time
Site:	05000605
Issue date:	11/30/1990
From:	; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20065L307	List: ML20065L305; ML20065L313;
References
	NUDOCS 9012060054
	Download: ML20065L313 (629)
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l p ABWR SSAR Amendment 15. Page change instruction ne following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), Bold page numbers represent a page that has been changed by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. !

1 CilAPTER 1 Add 2A 1,2 Add 2A 3,4 1.2 5,5a 1.2 5,5a Add 2A 5,6 1.26 1.2-6 Add 2A 7,8 Add 2A 9,10 i 13-2,3 13-2,3 Add 2A 11,12 13-4,5 13-4,5 Add 2A-13,14 13-6,7 13-6,7 Add 2A 15,16 13-8,9 13-8,9 Add 2A 17,18 Add 2A 19,20 1.6-2,3 1.6-2,3 Add 2A 21,22 1.6-4 1.6-4 Add 2A 23,24 Add 2A 25,26 j 1.8-42,43 1.8-42,43 Add 2A 27,28 1.8-44,45 1.8-44,45 Add 2A 29,30 Add 2A 31,32 1.92,3 1.9-2,3 Add 2A 33,34

( 1.93.1,4 1.9-4,5 Add 2A 35,36 Add 1.96 Add 2A 37J8 Add 2A 39,.m 1A.211,12 1A.2-11,12 Add 2A-41,42 l 1A.214,14a 1A.214,14a Add 2A 43,44 1A.2-17,18 1A.217,18 Add 2A-45,46 1AS-1 1A31 CilAPTER 3 CHAITER 2 3.2-1,2 3.21,2 2 il 2 11 3.28,8,1 3.28,8.1 3.29 3.29 2.0-2 2.0-2 3.2 10,11 3.2 10,11 3.2-21,21a 3.2-21,21 1 2 . 2 11 2 .2 -11 3.2-21b,22 3.2 21.2,22 2.21 2.21 3.2 29,30 3.2-29,30 Add 2.2 2 3.2 33,34 3.2-33,34 Add 3.2 34.1 Add 2 3 -11 Add 23-1 3.5 li,ili 3 .5 -11,111 3.5-8 3.5-8 Add 2A Cover Add 2A li,Ill 3.9 31,31.1 3.9-31,31.1 3.9-33,34 3.9 33,34 b

a 1-90120/.0054 *01130 POR ADOCK 05000605 K PDC

1 ABWR SSAR I 1

Amendment ;5 Page change instruction

'The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as I

page pairs (front & back). Bold page numbers represent a page that has been changet by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No.

PAGE No. PAGE No.

CHAPITR 4 4A Cover 4A Cover 4.0-li,iil 4 -11,111 4A li,iii 4A II,iu 4A 1 4A 1 4.1il,ili 4.1 li,Ill 4.11,2 4.1 1,1.1 Add 4B Cover Add 4.12 Add 4B 11 4.15,6 4.15,6 Add 4B 1 7 4.2.i1 4.211 Add 4C Cover 4.21 4.2 1 Add 4C Il 4.2 2 4.2 2 Add 4C 1 43 i1,i11 4 3 -11,111 CHAPTER 5 43 1,2 43-1,2 43-3 4.3-3 5.0-il,iii 5 .0 -i1,111 43 4 43-4,5 5.1 il 5 .1 11 4.4 li,ili 4 .4 11,111 5.11,2 5.11,2 4.4iv 4.4-Iv 5.13,4 5.13,4 i 4.4-1,2 4.4-1,2 5.15,6 5.15,6 Fig' 4.4 3,4 4.4-3,4 5.17,8 5.17,8 Fig

4.4-5,6 4.4-5,6 5.19 5.19,9.1 Fig

4.4-7,8 4.47,8 Add 5.19.2,93 Fig' 4.4-9,10 4.4-9,10 Add 5.19.4,9.5 Fig

4.4-11 4.4-11 Add 5.19.6 Fig' 5.1 10,11 5.1 10,11 4.5il 4 .5 11 4.5-4a 4.5-4.1 5.2 li,ili 5 .2 11,111 5.2-iv,v 5.2Iv,v 4.6-ii,ili 4 .6 -11,111 5.2 vii,5iii 5.2 vil, vill 4.6-iv,v 4.6Iv,v 5.2ix 5.2lx 4.6- 1,la 4.6-1,la 5.21,2 5.21,2 4.6-9,10 4.6-9,10 5.2-3,4 5.23,31 4.6-11,12 4.6 11,12 Add 5.24 4.6-13,14 4.6-13,14 5.2-5,6 5.2-5,6 4.6 15,16 4.6-15,16 5.27,8 5.27,8 5.210 5.2 10

These figures,11x17 foldouts, are being sent under separate cover 2

l

ABWR SSAR Amendment 15 Page change instruction

'Ihe following pages have been changed, please make the specified changes in your SSAR. Pages are listert as page pairs (front & back). Hold page numbers represent a page that has been changed by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No. }.'.AfiE No. PAGE No.

5.2 12 5.212 5.4 3,4 5.4-3,4 q 5.2 13,14 5.2-13,14 5.45,5.1 5.45,5.1 5.2 15,15.1 5.2 15,15.1 5.4-6 5.4-6 5.2 16,17 5.2 16,17 5.4-7,7a 5.4-7,7.1 :

5.2 17.1,17.2 5.2 17.1,17.2 5.4-8 5.48 5.2 18 5.2 18 5.4 9,10 5.4 9,10 5.2 19,20 5.2 19,20 5.410a 5.4 10.1 5.2 21,22 5.2 21,22 5.4-11,12 5.4-11,12 ,;

5.2 23,24 5.2 23,24 5.4 13,14 5.4-13,14 5.2 25,26 5.2 25,26 5.414a 5.4-14.1 5.2-27,28 5.2-27,28 5.4-15,15a 5.4-15,15.1 5.228a 5.2 28.1 5.416,16a 5.4 16,16.1 5.2-30 5.2 30 5.4 17,18 5.4 17,18 ,

5.2 31,32 5.2 31,32 5.4-18a 5.4 18.1 Add 5.2 32.1 5.4 19,20 5.4 19,20 5.2 33,34 5.2-33,34 5.4 21,21a 5.4 21,21.1 0- 5.2 35,36 5.2 37,38 5.2 35,36 5.2 37,38 5.4 22,22a 5.4 23,24 5.4 22,22.1 5.4-23,24 5.2 5.2 41 5.4 25,26 5.4 25,26 5.2-42,43 5.2-42,43 5.4-27,28 5.4-27,28 5.2-44,45 5.2-44,45 Fig' 5.4 29,30 5.4 29,30 5.2 46,47 5.2 46,47 Fig' 5.4 31,31a 5.4-31,31.1 5.2-48,49 5.2-48,49 Fig' 5.4-32 5.4 32 5.2 50,51 5.2 50,51 Fig' 5.4-33,M 5.4 33,34 Add 5.2 52,53 Fig

5.4-35,36 5.4-35,36 5.4-37,38 5.4-37,38 53 1,2 53 1,2 5.4-39,40 5.4-39,40 53 2.1 53-2.1 5.4-41,42 5.4-41,42 53-2a - 5.4-43 5.4-43 53-3,4 53-3,4 5.4-47,48 5.4-47,47.1 Fig'

$3 5,6 53 5,6 Add 5.4-48 Fig' 53-6a 5.36.1 5.4-49,50 5.4-49,50 53-7,8 53 7,8 5.4-51,52 5.4 51,52 Fig' 53 9,10 53 9,10 Add 5.4 52,1 Fig' 53-11,11a 5 3-11,12 5.4-53,54 5.4 53,54 Fig' '

53-12 -- 5.4 55,56 5.4 55,56 Fig' 53 13,14 5 3-13,14 5.4-57,58 5.4 57,58 Fig' Add 5.4-58.1,58 2 Fig

5.4-il,ili 5.4 il,ili Add 5.4 583 Fig' 5.4-vi,di 5.4-vi,vil 5.4-59,60 5.4 59,60 Fig

5.4-1,2 5.41,2 5.4-61,62 5.4-61,62 Fig

5.4-63,64 5.4 63,64 Fig' O

m

ABWR SSAR Amendment 15 Page change instruction The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back). Iloid page numbers represent a page that has been changed by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No.

CilAPTER 6 CilAITER 13 6.0-li,ili 6.0-II,ili 13.6-2 13.6 2 13.6 3 13.6-3 63 iv,v 63-Iv,v 6 3-9,10 63-9,9.1 CHAPTER 15 Add 6 3-10,10.1 6 3-11,11.1 63 11,11.1 15 il,ili 15 il,ili 6 3-12 6 3-12,12.1 15 iv,v 15 iv,v 6312a -

6 3-13,14 6 3-13,14 15.0-li,ili 1 5 .0 -11,111 6 3-15 63 15 15.0-1,2 15.0-1,1.1 63 18,19 6 3-18,19 Add 15.0-2,2.1 6 3-21 63-21 15.0-3 15.0-3 63 24 6 3-24 15.0 4,5 15.0-4,5 15.0-6,7 15.0-6,7 CilAITER 7 15.0-8,9 15.0-8,9 15.0-10,11 15.0-10,11 7A.2-1 22 7A.21 22 15.0-12,13 14.0-12,13 15.0-14 15.0 14 CilAITER 9 15.1 vi,sii 15.1 vi,vil 9.1 vi,vii 9.1 si,vil 15.11,2 15.1 1,2 9.12d,2e 9,1 2d,2c 15.1 3,4 15.13,3.1 9.1 12,13 9.11213 Add 15.1-4 15.1 5,6 15.1 5,5.1 93-il,ili 9 3 -11,i11 Add 15.1 6 15.1 7,8 15.1-7,8 9.4 il,ili 9.4-II,iii 15.19 15.1 9 15.1 10,11 15.1 10,11 9A.6 182 9A.61 182 15.1 12,13 15.1 12,13 15.1 15,16 15.1-15,16 CilAITER 10 15.1-17,18 15.1 17,18 15.1 19 15.1 19 103 1,2 10 3-1,2 15.2-si,sii 15.2-vi,sii CilAITER 11 15.2 viii,ix 15.2-vill,lx l 15.2-x,xi 15.2-x,xt 11.2 1 11.2 1 15.2 xii 15.2-xil 15.2 1,2 15.2 1,1.1 11 3-1 113-1 24 Add 15.2-2,2.1 0

=

nwa ssia O Amendment 15 Page change instruction

'Ihe following pages have been changed, please make the specified changes in your SSAR Pages are listed as page pairs (front & back). Bold page numbers represent a page that has been changed by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. --

15.2 3,4 15.2 3,3.1 Add 15C 11 Add 15.2-4 Add 15C.2 Il 15.2-5,6 15.2 5,6 Add 15C.21 15.2 7,8 15.2 7,8 Add 15C.2 2 5 15.2 9,10 15.2 9,10 Add 15C3-1 15.2 10.1 . Add 15C 41 15.2 11,12 15.2 11,12 Add 15C.5 il 15.2 13,14 15.2 13,14 Add 15C 51 15.2 15,16 15.2 15,16 Add 15C.5 2,3 :

15.2 17 15.2 17 Add 15C.5 4,5 15.2 18,19 15.2 18,19 Add 15C.5-6,7 15.2 20,21 15.2 20,21 Add 15C.5-8 15.2 22,23 15.2 22,23 15.2 24,25 15.2 24,25 Add ISD Cover 15.2 26,27 15.2-26,27 Add 15D Il 15.2 28,29 15.2 28,29 Add 15D.1 1 15.2-30,31 15.2 30,31 Add 15D.2 il 15.2 32,33 15.2 32,33 Add 15D.2-1 15.2 34,35 15.2 34,35 Add 15D.2 2 15.2 36,37 15.2 36,37 Add 15D3-il 15.2-38,39 15.2 38,39 Add 15D3 1 :

15.2-40,41 15 2-40,41 Add 15D3 2 '

Add 15D3-3 4 15 3-1,2 153 1,2 Add 15D.4-il 15 3-3,4 15 3-3,4 Add 15D 4-1 15 3-8,9 15 3-8,9 Add 15D.42 -

15 3 10,11 15 3-10,11 153-12,13 15 3-12,13 Add 15E Cover 15 3-14 153 14 Add 15EIl Add 15E.1 1 15.4-3,4 15.4-3,4 Add 15E.2-il 15.4-5,6 15.4 5,6 Add 15E.21 15,4-15,16 15.4-15,16 Add 15E.2 2 15.4 22,23 15.4-22,23 Add 15E3 il -

15.4-24,25 15.4-24,25 Add 15E31 Add 15E3 2,3 15.8 1 15.8 1 Add 15E3 4 Add 15E.4-1 Add 15C Cover Add 15E 5 il,ill Add 15C-Il O 5

ABWR SSAR Amendment 15. Page change instruction The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back) Bold page numbers represent a page that has been changed by Amendment 15.

REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No.

Add 15E.51,2 Add 19L511 Add 15E.5-3,4 Add 19LS 1 9 Add 15E.5-5,6 Add 19L.6-il,ill Add 15E57,8 Add 19L6-1 18 Add 15E.5 9,10 Add 19L7 Il Add 15E.511,12 Add 19L71 3 Add 15E513,14 Add 19L.8-il Add 15E515,16 Add 19L8-1 4 Add 15E517,18 Add 19L.9 il Add 15E.61 Add 19L9-1 2 Add 15E.71 Add 19L10-Il Add 19L10-1 2 CHAPTER 18 Add 19L.111 ISB il 18B-11 CilAl'TER 20 CilAFTER 19 20-il,iii 2 0 i1,111 19B ii,ili 19B il,ill 20.1 1.16 20.1 1.16 19B.21 19B.2-1 20.2 19,19.1 20.2 19,19.1 Add 20 3-19.2,19 3 1983-ii 19B3-11 19B3-1 1983-1 203-li,ila 203-il,ila 20 3 106,107 20 3-106,107 Add 19C-Il 20 3-108,109 20 3-108,109 19C.1 1 19C.1-1 20 3-109.1,110 20 3-109.1,110 Add 19C.2-Il 20 3-110.1,111 20 3-110 1,110.2 Add 19C.21 2 20 3 111.1 20 3 111,111.1 Add 19C3 il,lil 20 3 112,113 20 3 112,112.1 Add 19C3-1 35 Add 20 3-113 20 3 114,114.1 20 3-114,114.1 19E 21 19E.21 20 3-115,115.1 20 3 115,115.1 20 3 128,128.1 20 3-128,128.1 Add 19L Cover 20 3-255 290 20 3-255 290.1 Add 19L il,111 20 3-320 - 349 203-320 319 Add 19L11 Add 20 3-352,353 Add ' 19L21 Add 20 3-354 Add 19L3-1 Add 19L4-il 20.4 1 20.4 1 Add 19L4-1 2 20B 46,4/ 20B-46,47 o

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ABWR Standard Plant

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t CHAPTER 1 i PAGE STATUS PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 1 COVER 1.2 17 14 1,2-46 14 1.7-11 13 1 11 8 1,2 18 6 1.7 1 13 1 111 11 1.2-18a 6 1.2-48 14 1.7-2 13 1.2-49 14 1-tv 11 l'?'3 l3 1 2 19 4 1.1-11 8 1.2 194 6 1.7-4 13 1.7-5 13 1.1-1 9 1.2-19b 6 1.2 52 14 1,2-20 1.7 5.1 13 1.1-2 1 4 1.2-53 14 1.7-5.2 8 1.1-3 2 1.2-21 4 1.3-11 1 1.7 5.3 8 1.2-11 7 1.2 22 4 g,3,3 3,7,$,4 3 34 1.2 111 14 1*2-23 4 1,3 2 15 1.7 5.5 10 1,2 tv 14 1,2 24 4 1,3-3 15 1.7-5.6 10 1.2-v 1 1.2-25 4 1.3-4 15 1.7 5.7 13 1.2-vi 14 1.2-26 4 1.3 5 15 1.7 5.8 '13 1.2-vta 7 1.2 27 4 1,3-6 15 1.7 5.9 13 1.2-v11 6 1.2-28 4 1.3 7 15 1.7-5.10 13 1.2 vita 6 1.2 29 4 1.3-8 15 1.7-5.11 13 1,2 vilt 7 1,2-29a 4 l 1,3-9 1 1.7 5.12 13

(, ) 1,2 1 7 1.2-29b 4 1.3 10 1 1.7 5.13 13 v 1.2-2 1.2-29c 4 1

1,3-11 1 1.7-5.14 13 j 1.2-3 1' 1.2 29d 6 1.3 12 1 1.7-5.15 13 1.2 4 1 1.2 29e 6 1.3 13 6 1.7-5.16 13 1.2 5 7 1.2-29f 6 1,7-6 1.3 14 1 14 1.2 5a 15 1.2-29g 6 g,3,33 3 3,7,7 34 1.2 6 15 1.2-29h 6 1.3-16 1 1.7 8 6 ,

1,2-7 14 1.3 17 1 1.7-9 6 ;

1.2 8 14 1.2-29j 6 1.3 18 7 1.7 10 6 1.2 9 7 1.2 30 6 1,3-19 1 1.7-11 6 1.2 10 2 1.2 31 6 1.4-11 1 1.7-12 6 l- 1.2 11 14 1.2-32 6 '

~

1.4-1 8 1.2-12 7 1.2-33 6 1.7-14 6 1.4-2 1 1.2-13 14 1.2-34 6-

,,3 1.7-15 6-7 1.2-14 14 1.2-35 6 1.7-16 6 I 1.5-1 14 1,2-15 7 1.2-36 6 1.7-17 6 1.2 13 1.2-37 6 1.6-1 6 1.7-18 6 i 1.2-16a 7 1.2-38 6 1.6-2 8 1.8-11. 12 1= 1,2-16b 7 1.2-39 45 14 1.6 3 15 1.8-111 12 1.6-4 15

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November 30,1990 1I l

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_.____A

ABWR Standard Plant !

CHAPTER 1 1 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 1.8-1 1 1.8 36 8 1.8 73 12 1A.2-3 8 a

1.8 2 6 1,8 3 6 1.8 38 14 1.8-75 12 1A.2 5 4 1.8 4 6 1.8-5 6 1.8-40 8 1.8 77 12 1A.2 7 13

^*' II 1.8 6 6 1.8 7 6 1.8-42 14 1.8 79 12 1A.2 da 11

-1.8-8 6 l.8 9 8 1.8 44 15 1.8-81 12 1A.2 10 6

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^'

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10 1.9 4 15 1.8 23 10 . 1.8 59 13 1.9 5 15 1A.2 20 8 q 1.8-24 3 1.9-6 15 l.8-25 8 1.8-61 12 * '"

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1A COVER 1.8 26 2 1A31 15 1.8-27 7 1.8-63 12 1.8-64

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183 1.8-67 12

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Nmmber 30,1990 12 1

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.ABWR Standard Plant II l

.O CHAPTER 2

(_/ PAGE STATUS PAGE AMEND PAGE AMEND 2 COVfR 2A-17 15 2A 18 15 2- H 15 i

2.0-11 6 '

2A-20 15 2.0-1 6 !

2A-21 15 2.0-2 15 2A-22 15 2.1 11 6 2A-23 15 2.1-1 6 2.1 2 6

^'

2.1 3 9 2A 27 15 2.1 4 9 ' I 2.1 5 6

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2A-ti 15 2A 37 15 2A iii 15 2A 38 15 2A 1 15 2A-39 15 2A 2 15 2A-40 15 2A-3 15 2A-41 15 2A 4 15 2A-42 15 2A 5 15 . 2A 43 15 2A-6 15 2A-44 15 2A 7 15 2A-45 15 2A-8 15 2A-46 15 2A-9 15 2A-10 15

2A Il 15 2A-12 15 2A-13 15 2A 14 15 2A 15 15 2A-16 15 D

O 21 Nowmber 30,1990 l

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ABWR Standard Plant i f]-

( l CHAPTER 3 PAGE STATUS !

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 3 COVCR 3.1 21 9 3.2 11e 3 3.3 3 1 3.1 21.1 9 3.2 12 10 3.3-4 1

3. 0- H 7 3.0-111 7 3.1 22 8 3.2 13 12 3.4 11 6 3.0-iv. 1 3.1 23 1 ;

3.0-v 1 3.1 24 7 3.2-144 11 3.4 1 6

. H . 2 6 3.0-vt 1 3.1 25 1 3.1 26 7 3.2-16 6 3.4-3 6 3,gg 3 3.1-111 1 3.1 27 1

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_7 \

November 30,1990 31

h ABWR Standard Plant CHAPTER 3 PAGE _1"LO PAGE PAGE STATUS (Continued)

AMEND PAGE AMEND PAGE AMEND g

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3.7-15 1 3.8-15 1 3.6-30 1 3.7-16 1 3.7-53 1 3.8-16 7 -

3.6-30.1 10 3.7 54 1 3 g,g7 3,7,g7 3 7 3.6-31 7 3.7-17a 3 3.7 55 1 3.8 18 1 3.6-32 1 3.7-18 1 3.8 19' 7 3.6 33 1 3,7,39 g.

W7 1 M M 3.7 58 1 3.6-33,1 10 3.7-20 1 3.8-21 6 3.6-33.2 10 ' '

3.7-21 1 l

3.6-34 1 3.7 1 3.8 22- 1 3.6 35 1 3.7 1 3.8-23 8 3.7-24 11 3.7 62 1 3.6 36 1 3.7-24 1 3.8 .'3a 6 3.6-37 1 3.7 63 1 3.841 7 3.7-25 7 3.7 64 1 3.6-38 1 3.7-26 9 3.8-25 6 3.6 39 1 '

~

3.7-27 7

~

3.6-40 1 3.7-28 1 3.8-264 7 3.7 67 1 O

October 1,1990 32 h

ABWR Standard Plant

/~'N#

CHAPTER 3 N..J PAGE STATUS (Continued)

PAGE AMEND PAG! AMEND PAGE AMEND PAGE AMEND 3.8 27 6 3.9 vitt 1 3.9 29 7 3.9-58.4 14 3.8 27a 6 3.9-lx 1 3.9 30- 3 3.9-58.5 14 3.8 28 7 3.9-x 14 3.9-31 15 3.9 58.6 14 3.8 29 3 3.8-30 1 3.9 xit 14 3.9-32 1 3.9 58.8 14 3.8 31 1 3.9 1 1 3.9-33 7 3.8 32 7 3.9 2 2 3.9-34 15 3.9 58.10 14 3.8 33 2 3.9 . 11 3.9 34.1 11 3.8-34 3 3.9-3.1 11 3.9 8H 14 3.9 35 11 3.9-4 3.9 36 3.8 35 3 11 3 3.8 36 1 '"

3.9-5 3 3.9 36a 10

I 3.8-37 1 3.9 6 11 3.8 38 1 3.9 SP.16 14 3.9-7 3 3.9 38 1 g, g 7

3.8 39- 1 3.9-8 3 9,39 g 3.8 40 1 g 3.9-58.18 14 3,g,9 g 9,g

.9 58,19 14

-3.8-41 1 3.9-10 1 3.9-11 12 7 3.8-43 1 3.9-Ita 3 3.9 43 7

{j _ 3.8 44 1 3.9 12 7 3.9-44 14 3.8-45 1 3'9'l3 7 3'9'44'l l4 3.9-58.24 14 B-4 1 3.9-*5 3.9-14 1 14 3.9-58.25 14 3.9-58.26 14 3.8-49 1

" 3.9-58.28 14 3.8-50 1 3.9-48 1 3.9-58.29 14 1 3.9-58.30 14 8 2 g

3 3.8-53 1 3.8 54 1 3.9 58.32 14 3.9-20 13 9-52 3.8-55 1 3.9 53 1

3.9-21 13 1 3.8 56 1

3.9-22 7

3.9 61 3.9-11 1 1

~

3.9-62 1 3 C-111 11 3.9-24 7 3.9-56 3 3.9 Iv 1 3.9-57 3-3.9 25 7

' " 3.9-26 1 3.9 64 11 5 3.9-vi 13 3.9-58.1 3.10-11 3.9 27 8 14 1 3 9,g _ 7 3.9-28 1

3.9-58.2 14 3.9-58.3 14 3.10-tv 9 r

/

(

November 30,1990 3-3

ABWR !

i' Standard Plant CHAMER 3 PAGE STATUS (Continued) ffGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 3.10-1 8 3A.4 1 1 3A.8 13 1 3B.3-1 6 3.10 2 1 ^'

3A.4 2 1 3B.4 11 6 3.10-3 1. 3A.4-3 1 3A.8-15 1 38.4 111 8 3.10 4 7 3g,4,4 3A.8 16 1 g 38.4 tv 8 3.10-5 7 3A.4 5 1 3A.8 17 1 3B.4 1 6 3.10-6 1 3A.4 6 1 3.10-7 1 3A 4-7 1 3A.8 19 1 3.10-8 7

3A 4-8 1 3.10 9 1 3A.4 9 1 3A.8 21 1 38.6 1 6 3.10 10 1 '

3A.4-10 1 38.7-1 6 3.11-11 14 3A.8-23 1 3g,$,gg 3 3C C0'.ER 3A.8 24 1 3M M 3.11-1.1 14 3A.8 25 1 3A.5-2 1 3C 1 1 3 . I '. - 2 14 .. 3A.9 1 1

?.11-3 14 3 54 1 3A.10 1 1 3A COVER 3A.5-5 1 3AA.1i 1 3A.it 1 3A.5-6 1 3AA 1 1 3A.til 1 30 11 1 3A.1 1 9 30 ti1 'l 3A.1 2 1 3A.6-1 1 g ,4 30.1-1 1 3A.2-it 1 3A.6-2 1 30.2-11 1

.3A.2 1 1 3A.7-li 1 30.2-1 1 3A.2-2 1 3A.7-1 30.3-1 1 3A.2-3 1 3,' . 7-2 1 3AA 8 1 30.4 11 1

^ '

3AA 9 1 30,4 1 1 3A.3 1 1 3A.6 iti 1 3AA 10 1 30.4-2 1 3A.3 2 1 3A.8-' 7

-3A,3-3 1 -3A.8-2 7 3AA-12 1 30.5-1 1 3A.3-4 1 3A.8-3 1 3B COVER l 3A.3 5 1 3A.8-4 1 30.6-11 1 3A.3 6 7 3A.8-5 1 30.6-1 1 3A 3 7 1 3A.8-6 1 3B.1 11 6-3A.3 1 3A.8-7 1 39 3A.3 9 1 3A.8-8 1 38.1-1 6 3

3A.3 10 1 3A.8 9 1 38.2-11 7 3E COVER 3A.3-11 1 3A.8 10 1 38.2-1 6 3A.3-12 1 3A.8-11 1 3A.8-12 1 3E.1-1 1 October 1,1990 3-4 O

I ABWR l l

Standard Plant '

. f7 CHAITER 4 V PAGE STATUS PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 4 COVER 4.5-11 15 4.6 26 0 4A 29 0 4.0 11 - 15 4.5-1 12 4A COVER 4.0-111 15 4.5 2 8 4g,gg g$

4A 31 0 4.1-11 0 4.5 3 2 4A tit 0 4A-32 0 4.1 111 15 4.5-4 2 4A-iv 0 4A-33 0 4,1 1 15 4.5-4.1 15 4A-34 0 4.1 1.1 15 "I 4.6 11 0 4A-35 0 4.1-2 15 4.6 111 15 4A 1 15 4A 36 0 4.1 3 0 4.6-tv 15 4A 2 0 4A 37 0 4.1 4 0 4.6 v 15 4A-38 0 4A-3 0 4.1 5 15 4.6 1 15 4A 4 0

! 4.1 6 15 4.6 le 3 4A 40 0 4A-5 0 4.1-7 0 4.6 2 12 4A-41 0 4.2-11 15 4.6 3 12 4A-42 0 4.6 4 12 4A 7 0 4.2 1 15 4A-43 0 4.6 5 10 4A 8 0 4'2 2 15 u-44 0 4.6 5.1 10 4A 9 0 4*3 11 15 4A-45 0 g 4.6-6 12 4A-10 0

't 4.6 6.1 10 43,33 -0 4.3-1 15 4.6 7 0 4A 12 0 48 COVER 4.3 2 15 4'6-8 12

0. 48 11 15 4A-13 4.3 3 15 4'6-9 15 0 4B1-7 15 4A 14 4.3 4 15 4.6-10 15 4A-15 0 4'3.$ 15' 4C COVER 4.6-11 15 4A 16 0 4.4 11 - 15 4.6-12 15 4C-ti 15 4.4 111 15 4C-1 15 4.6-13 15 4A-18 0 4.4 tv 15 4 . 6 1 ". 0 4,4 1 15 .4.6-15 12 4.4 2 15 4.6 16 15 4A-20 0

-4.4-3' 15 4.6 17 10 4.4-4 15 4.6-18 0 4A 22 0 4.4-5 15 4.6-19 0 4A-23 0 4.4-6 15 4,6-20 0 4A-24 0 4.4 7 15 4.6-21 0 4A-25 0 4.4 8 15 4.6 22 12 4A 26 0 4,4 9 15 4.6-23 10 4A 27 0 4.4 10 15 4A-28 0 4.6-24 0 4.4-11 15 4.6-25 0 f%

i \

\,_/

4*I Nowmber 30,1990

ABWR ,

Standard Plant

['\ . CHAPTER 5 PAGE STATUS

'\s' PACE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 5 COVIR 5.2-10 15 5.2 36.3 13 5.2 52 15 5.2 36.4 13 5.2 53 15 5.0 11 13 5.2 11 3 5.0-111 15 5.2-11a 11 5.2 36.5 13 5.3 11 11 5.2 36.6 13 5.3-111 11 5.1 11 15 5.2 12 15 5.2-36.7 13 5.3dv 0 5.1 1 2 5.2-13 15 5.2 36.8 13 5.3 v 9 5.1-2 15 5.2-14 15 5.2 36.9 13 5.3 1 15 5.1 3 15 5.2-15 15 5.2 36.10 13 5.3 2 15 5.1-4 15 5.2 15.1 15 5.2 36.11 13 5.3 2.1 15 5.1 5 -15 5.2 16 13 5.2 36.12 13 5.1-6 15 5.2 17 15 5.3-3 15

.2 36.13 13 5.3-4 15 l 5.1 7 15 5.2 17.1 15 5.2-36.14 13 5.1-8 15 5.2-17.2 13 5.3 5 15 5.2 36.15 13 5.3 6 15 5.1 9 15 5.2-18 15 5.1 9.1 5.3 6.1 15 15 SN9 n. '3 'II I3 5.1 9.2 15 5.2 20 15 5,3 7 15 5.2 36.18 13 5.1-9.3 15 5*3-8 15 I 5.2-21 15

~ 'I0 I3 5.1-9.4 15 5.2 22 15 5.3-9 15 5.2-36.20 13 l 5.1-9.5 15 5.3-10 15 5.2 23 15 rx 5.2-36.21 13 5.3 11 15 I j 5.2-11 0 5.2 24 15

'# 5.2 111 15 5.2-36.22 13 5.3 12 15 5.2 25 15 5.2-tv 15 5.2-26 15 5.2 37 0 5.3 13 15 5.2-v 13 .-8 15 5.3-14 15 5.2-27 15 5.2 vt- 13 5.2 28 15 5.2 39 0 5.4 15 5.2 vl.1 13 . 0 0 5.4-Hi 0 5.2-26,1 15 1 5.2-vit 15 5.2-41 15 5.4 tv -10 U -29 13

. 0

~ 5,2-42 15 5.2-29.1 13 5.2-tx 15 5.2 43 15 5.4-vi 15 g, g

. H 15 5.2-1 15 5.2 29.3 13 5.2-43.1 13 5.2-2 15 . 3.2 13 5.4-1 15-5.2 30 15 5.2 3 15 .- 15 5.2-31 15 5.4-3 5.2 3.1 15 5.2-45 15 15 5.2 32 15

. 15 5.2-4 15 5.2-46 15

5. 2-32 J 15 5.4-5 15 5.2 47 15 ;

5.2-5 15

' ' .4 5.1 15 5.2-6 15 5.2 48 15 .

5.2-34 15 5.4 6 5.2 49 15 15 l 5.2-7 15 5.2 8 3 5.2-50 15 6 1 5.2-51 15 5.2-9 11

~

5.2 9.1 11 5.2-36.2 13 f

U}.

Ncumber 30,1990 33 ,

1 ABWR ;

Standard Plant CHAPTER 5 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND l 5.4 1 15 5.4-35 15 5A.1 1 13 5AB 17 13 5.4 7.1 15 5.4 36 15 '

5A.2 li - 13 5,4 8 15 5.4 37 15 5AB 19 13 5A*2 1 13 5.4 38 15 5AB 20 13 5.4 9 15 5A 2 2 V 5.4 10 15 5.4-39 15 5A*2 3 13 5.4-40 15 SAB 22 13 5.4 10.1 15 5A.5-1 13 544 15 5.4 11 15 _

5A.2-4 13 5.4 12 15 54,3 3 13 5.4 43 15 ,

5.4-13 15 5A 4 1 13 5.4 14 15 5.4 44 0 SA.4 2 13 5.4 45 0 5.4 14,1 15 SM-COVER 5.4 46 0 5.4 15 15 5M 1 13 5.4 15.1 15 5.4 47 15 5AA 2 13 5.4 47.1 15 5.4 16 15 5M 3 13 5.4-16.1 15 5.4-48 15 5AA 4 13 5.4-17. 15 5.4 49 15 5M 5 13 5.4 18 15 5.4-50 8 5AA 6 13 5.4 18,1 15 5.4-51 15 5M 7 13 6' 6 0 5.4 19 15 SAA-8 13 5.4 20 15 5.4 52.1 15 5AA-9 13 5,4 21 15 5.4 53 15 5AA 10 13 5.4 21.1 15 5.4-54 15 5AB-C0VER 5.4 22 .15 5.4-55 15 5AB-1 13

,5.4 22.1 15 5,4 56 15 SAB 2 13 5.4 23- 15 5.4 57 15

$g3,3 33 5.4 24 15 5.4 58 15 5AB 4 13 5.4 25- 15 5.4 58.1 15 SAB 5 13 4

5.4 26 15 - 5.4-58.2 15 SAB-6 13 5.4 27 15 5.4 58.3 15 5AB-7 13 5.4 28 15 5,4 59 15 5AB 8 13 5.4 29 - 7 5.4 60 15 5AB 9 13 5.4 30' 15 5.4-61 15 SAB 10 13 5.4-31. 15 5.4 62 15 5AB ll 13 5.4-31.1 15 5.4-63 15 5AB-12 13 5.4-32 15 5.4 64 15 5AB-13 13 5,4 33 15 5A COVER 'SAB 14 13 5,4 34 15 5A 11 13 5AB 15 13 5AB-16 13 0.

Nmtmber 30,1990 52

ABWR Standard Plant CHAPTER 6 PAGE STATUS PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 6 COVER 6.2 13 0 6.2 44 11 6.2 50.23 14

~~

6.2 45 6.0 11 15 2 6.0-1ii 13 6,2-15 10 6.2-46 3 6.2 50.25 14

< 6.0-tv 13 6.2 46a 3 6.2-16a 3 6.2-46b 3 6.2 50.27 14 6.0-1 2 6.2 50.28 14 6.2 17 11 6.2-46c 7 6.1-11 14 6.2 46d 6.2 50.29 14 6.2 18 9 7

^

6.1 1 0 6.2 46e 6.2 19 0 7 6.1-2 14 6.2 46f 6.2 50.31 14 6.2-20 7 7 6.2 50.32

~

6.1-3 14

- 14 Wn 7 6.2 46g 7

? 6.1-4 0 6.2-22 9 6.2 46h 7 6.2 50.33 14 6.1 5 0 6.2 47 9 6.2-22e 9 a 1-6 6.2 48 9 6.2-50.35 14 h_ 2 6.2 50.36 24 6'1 7 0 6.2 j 6.2-i4 9 2 6.2 11 0 6.2-50 0 6.2 50.37 14 6.2 1 H 0 6.2 50.1

6. -26 0 14 6.2 tv 0 6.2 50.2 14 6.2 50.39 14 6.2-v 0 6.2-50.3 14 7' .2 8 6.2-vt 2 6.2-50.3a 14 6.2-50.41 14 1 E * *8 6.2 50.42 14 6'2 v11 11 6.2 50.3b 6.2 30 2 14 6.2-vii1 10 6.2 50.4 14 6.2 50.43 14 6.2-tx 10

.2 32 11 6.2 50.5 14

~

6.2 txa 10 6.2 50.6 14 6.2-50.45 14 6.2-33 11 6.2-50.46 14 6.2 x 0 6.2 33.1 11 6.2 50.7 14 i 6.2 xt 11 6.2 50.8 14 6.2 50.47 14

6. W 11 f -

6.2 xit 13

- 6.2 35 11 6.2 50.48 9 6.2-50.10 14 6.2 50.49 14 6.2-1 0 6.2-36 11 6.2 2 0 6.2-50.11 14 6.2 50.50 14 6.2-36.1 11 6.2-50.12 14 6.2 50.51 14 623 0 6.2 4 3 6.2 50.13 14 6.2-50.52 14 6.2-37.1 10 6.2 50.53 '14 6.2-50.14 14 6.2-44 3 6.2-38 5 6.2 50.15 14' 6.2-50.54 14 625 6.2 39 0 6.2-50.16 14

, 6.2-50.55 14 6.2-40 3 6.2 50.56 6.2 50.17 14 14

6. 0 6.2 50.18 6.2 41 3 14 6.2 50.57 9 6.2-42 9 6.2-50.58 6.2 50.19 14 9 6.2 9 0 6.2 50.20 6.2-43 7 14 6.2 50.59 9 6.2-43a 7 6.2 50.21 14 6.2-11 0 6.2-50.22 14 6.2-12 0 6.2-52 0

=

r November 30,1990 61

ABWR Standard Plant CHAPTER 6 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 6.2 53 0 6.2 90 11 6.3 21 15 6.3 50- 2 6.2-54 0 6.2-91 11

6.3-22 2 6.2-55 0 6.2 92 11 '

6.3 23 10 6.2 56 10

6.2 93 11 6.2 57 10 6.3 54 2 6.2-58 0 6.3 24a 6 6.3-55 2 6.3 W 0 6.3 24b 6 6.2-59 0 '

6.3-tv 15 6.2 60 0

,,3,v g3 6.3 24c 6 6.3 57 2 6.2 61 0 ' '

6.3-vi 13 6.2-62 0 ' ' '

6.3 vil 7 6.2 63 0 6.3-24f 6 6.3 60 2 6Mt 7 6.2-64 0 6,3 tx *' ' ' '

?

6.2-65 0 .- 6 6.3-62 2 ;

6.3-x 7 6.2-66 0

6.3 x1 7 6.2 67 0

6.3 1 2 6.2 68 0 6.3-28 13 6.3-65 2 6.2 69 0 ' ' '

~

6.3 3 2 6.2 70 0

6.3 4 14 6.3 30 2 6.2-71 0 6.3 5 7 6.2 72 0 *

6.3-6 13 6.3 32 0 6.2 73 0 6.3 70 2 W7 7 6.3 M 6 6.2 74 0 ' '

6.3 8 13 6.3-34 6 6.2-75 0 .- 10 6.3-9 15 6.3-35 6 6.3 9.1 15 6.3 36 13 6.2-77 0 * '

6.3 10 15 6.3 37 13

' ' ~ ' '

6.3-10.1 15 6.3-38 13 6.2-79 0 '

6.3 11- 15 6.3-39 13

' ' -1 H 0 6.3 11.1 15 6.3 40 13 6.2-81 0

6.3 12 15 6.3 41 13 6.2 0 -

0 6.3-12.1 15 6.2 83 0 6.4 3 7 6.2-84 0- 6.4 4 7 6.3-14 7 6.3-15 15 . 5' 2 0

6.3-16 0- 6.3-46 2 6.2-87 9

6.3 17 2 6.3 47 2 6.2-87.1 9 6.4 7 10 6.2-88 6 6.4-7a 6 6.2-89 0 -

~

6.3-20 13 O

October 1,1990 62 I

l l

_ _-. -__ - _ _ _ _1

.-ABWR Standard Plant CHAPTER 7 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND EA,GE AMEND PAGE AMEND 7.6-21 2 7.6-58 2 7.7 11 2 7.7 38.1 11 7.6 22 2 7.6 59 2 7.7 12 2 7.7 38.2 11 7.6 23 2 7.6-60 2 7.7-13 2 7.7-39 11 7.6 23a 5 7.6 61 2 7.7 14 2 7,74g 7 7.6 24 2 7.6-62 2 7.7 15 2 7,7,y 7.6-25 2 7.6-63 2 7.7 16 2

. 2 7.6 26 2 7.6 64 2 7.7 17 3 7.6 27 2 7.6 65 2- 7.7 18 3 2

7.6-28 2 7.6 66 2 7.7.lBa 3

.- 2 '

7.6 29 2 7.6 67 2

'I'I' E 1,1 4s t 7.6 30 2 7.6 68 2 7.7 20 2 7,p ,

7.6-31 2 7.6 69 2 7,7 ,

7.6 70 7.7 22 7.6-32 2 t 2

. 9 2 7.6 33 2 7.6 71 2 l' 'E E y,1.so y 7.6 34 2. 7.6 72 2 7.7 24 2 7.6 35 2 7.6-73 2 I'I'II E 1,1.sg g 7.6-36 2 7.6 74 2 7.7 26 2

I 7.6-37 2 7.6-75 2 .

I*I'EI E

- ['% y,1.sg g ;

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2 7.6 76 7.6 77 2

2 7.7-28 7.7 29 2

2 7

7,g

~

7 i

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.- 2 7.6-41 2 7.6-79 5 7.7 31 2 7.7 58 2 7.6-42 2 7.6 80 14 7.7-32 2

.9 2 .;

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. 60 2 q 7.6 44 2 '7.7 11 11 7.7 34 11

7. - 1 2 l 7.7-111 11 f

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'

2 7.6-49 2 7.7 34.5 11

.- . -3 . 11 .7 66 2 ;

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' .- 2 7.6 57 2 7.7-38 11 1

i .

I

%_J October 1,1990 73 I

o l

ABWR Standard Plant CHAPTER 7 PAGE STATUS (Continued)

PAGE Ag3Q 7.7 74 2 7.7 15 2 1

7.7 76 2 'l 7.7 17 t l 7.7 78 2 7.7 79 2 7.7 80 2 7.7 81 2 1.7 82 2 ;

7.7 83 2 7.7 84 2 7.1 85 2 7.7 86 2 l 7.7 87 14 7.7 87.1 14 7.7 88 11 l _7.7 89 11 l 7.7 90 11 7.7 91 11 1 7.7 92 11 7.7-93 11 l 7.7 94- 11 7.7 95 11 7.7 96 11 L .7.8 11 11 7.8 1 11 7A COVER 7A li 6 7A-lil 6 7A.1 1 6 7A.2 22 15 7A.3 1 6 7A.4 1 6 7A.5 1. 6 1A.f-l 6 7A.7-1 8 O

November 30,1990 7-4

ABWR Standard Plant (n'

CHAPTER 9 PAGE STATUS PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 9 COVER 9.1 7 7 9.2-11 6 9.2-22 14

9. - 9. -lH 1 9.2 23 14 :

9 il 6 9 ill 11 9.1 9 7 9.2-iv 11 9.2-24 14 9.1-10 7 9.2-tva 11 9.2-25 14 g,gy 34

9. H 9. w H 9.2 25a 7 g,. .i 6 9.1 !!a 7 9.2-vi 14 9.2 25b 7 9.1 111 6

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9. I 9.2 la 8 9.1-v 6 9.2 26 6

' ' ' ' 2 H 9.2 27 6 9.1 vi 6 9.1 13b !!

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'I

9.2 31 6 g,g.1 6 9.2-4.1 14 9.1 14 6 9.1-la 6 9.2 32 6 9.1 1 6 9. 33 6 9.1-lb 7 9.2 6 14 9.1 16 6 9.2 34 6 9.1-2 11 9.2 6.1 14

9. l 12 9.2 34a 6 9.1 ta 6 9.2-7 9.1-18 6 14 9.2 3 2 O- 9.1 2b 9.1 2c 7

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'I "

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9. - a .2 -

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r 9.1 26 6 9.2 44 7 9.1-6b 6 g ,p r 9.1 6c - 6

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3II '

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- 9. - 1 2 9.3-la 9 9.1-61 6 9.1 37 6 9.3-2 6 9.3-3 12

[)

9.t November 30.1990 1

ABWR

' Standard Plant CHAPTER 9 PAGE STATUS (Continued) .l PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND ,

9.3-4 2 9.4-2c 2ta 10 9.5 10.1 10 9A.4 xvl 14 9.3-5 11 9. 0.2 10 %. bull 14 9.4-2j 6 9.3 6 2 9.4-2k 6 9.5-10.3 10 9A.4 1 - 271 14 9A.5 il 9.3-7 8 9.4 3 6 6 9.3-8 8 9.4-4 s 9.5 10a 6 9A.5 1 - 8 14

, ,g 4 9.3 10 4 9.5 11 2 9A.6-ti 14-9.3-11 7 9.5-12 6 9A.6 182 15 2

9.!-12 11 9.5-13 10 9.4-7a 6

" I 9.4 7b 6 9.5-14 14 9 3-13.1 11 9.4 7c 6 f.3-14 3 9.4 7d 6 J.3-15 6 9.5 17 14 9 6 9.3 154 6 ^

9.4 7f 6 9.3 16 2 9.4-7g 6 9A-ll 14

^ "

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9.4 8 6 9.3 23 12 '

9.d 9 6 ,

9.3-24 11 9A.2 8 14 9.5-11 14 9.4-11 15 9.5-111 6 9A.3-11 6

9.5-tv 6 9A.3-1 6 9.4 tv 6 9.5-v 10 gg,g g 9.5-va 10 9A.4-til 6

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l November 30,1990 92 e

ABWR-Standard Plant f^% CHAPTER 10 V. PAGE STATUS PAGT AMEND PAGE AMEND PAGE AW ND 10 CovtR 10,4.,, 3 ,,,,

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v

)

November 30.1990 10 l

i 1

Standard Plant i es

( I- CHAPTER 11 l PAGE STATUS !

PAGE AWEND PAGE ANEND

-11 COVER 11.5 7 4 )

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i N<wember 30,1990 t t .1 I

j

-ABWR Standard Plant CHAPTER 13

{ -

PAGE STATUS PAGE AMEND 13 C0VER 13-11 8 13.1-1 7 13.2 1 7 13.3-1 11 13.3 2 11 13.3 3 11 l 13.3 4 11 l l

13.3-5 11 13.4 1 -7 13.5-1 7 13.6-11 8 13.6-111 8 13.6 tv 8 13.6 1 8 13.6 2 15 13.6 3 15 l.

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t t

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( .i O

N<mmber 30,1990 13 1

7-ABWR Standard Plant CHAPTER 15

' ( ]-

PAGE STATUS AMEND fyf. AMEND PAGE AMEND PACE AMEND f't

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15.4-18 2 E i U-November 30, im 15 . I

ABWR Standard Plant CilAPTER 15 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 15.4 19 2 15.6 17 0 15.7 19 2 15A.4 3 0 15.4 20 2 15.6 18 10 15.7 20 2 15A.4 4 0 15.4 !! 2 15.6 19 8 15.7 21 2 15A.4-5 0

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15.7-18 2 1;A.4 2 15.6 16 2 0 R6N 0 15A.6 25 0 0

Nowmber 30,1990 15 2

ABWR Standard Plant

-r O CHAPTER 15 PAGE STATUS (Continued) gycz, Aufn3 FAtt AMEND PAGt AMEN 3 PAGE AMEND 15, ' 26 0 15A.6 65 0 15B.3 ti 6 ISE COVER IM.6 !? ? 15B.3*) 7 15t ti 15 15A.6 tB 2 15A.6-67 0 38.4 11 6 St.1 1 3 15A.6-68 0 IM.6 29 0 38.4 1 6 St.2 it n 15A.6 30 0 15A.6-69 0 15t.t 1 15A 6 70 0 15C C0vtR 15 1 M.6 31 0 g$g,g.g 35 15A.6 32 0 15A.6 71 0 15C il !$

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'O' 15A.6 39 0 15C.4 1 15 nt.5 H H 15A.6 40 0 1M.6 79 0 15C.5-tt 15 15E.5 tit 15 15A.6 80 0 15A.6 41 0 15[.5 1 15 15C.5 1 15 15A.6 42 0 15A.6-81 0 g$g,$,, 35 15A.6 82 2 15C.5 2 15 IM.6 43 0 15C.5 3 15t.5 3 15 O

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158.2 111 6 15A.6-63 0 15B.2 1 6 15A 6 64 0 15D.4 2 15 Ncatmber 30,1990 15 3

ABWR Standard Plant n) i

'v CHAPTER IN PAGE STATUS PAGE ANfWD PAGf Amno 18 COVfR iga,jl.1 jg 18 11 13 IBA.It 1 14 18-111 14 18A.13-1 14 18.1 1 13 188 C0VER 18.2-1 13 188 11 !$

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180-COVER IBA.4 1 14 180 11 14 f

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l

%d htmber .10,1990 18 1

ABWR Standard Plant CHAPTER 19 PAGE STATUS PAGE AMEND PAGE AMEhD PAGE AuthD PAGE AMEND 19 COVER 198 11 11 19D.5 1 10 19F.3 1 10 I'0' ' U Ib 190.6 11 8 19F.4 1 10 19 11 8 19 1it 8 19B tv 11 190.6-111 4 19G COVER 19 iv 14 19B.1 11 14 190.6-tv 4 19H COVER 19.1 11 4 198.1 1 14 190.6 1 10 g,g, g g 19.1 1 14 198.2 H 7 190.7 11 4 19H.1 1 10 I'""' '

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1 i

ABWR Standard Plant CilAPTER 19 py, ,3 19J.6-1 10 19K COVER 19K-ti 14 19K.1 1 14 19K.2 1 14 19K.3 1 14 19K.4 1 14 19K.$ ti 14 19K.5 1 14 19K.6 1 14 19L C0VIR 19L il 15 19L til 15 19L.1 1 15 19L.2 1 15 .

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Ncwembet 30,1990 19 2

ABWR Standard Plant

()

CilAPTER 20 PAGE STATUS FAGf AufhD PAGE ANEND PAGE AMfhD PAGf AuthD 20 COVER 20.2 4f 3 20.! Itb 3 20.2 14.8 10 20.2-49 3 20.2 Itc 3 20.2 14.9 10 g,g ,

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3

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) t0.2 14.7 10 20.2 !!c 3 O

November M,1990 20 1

ABWR Standard Plant CHAPTER 20 PAGE STATUS (Continued)

PAGE AMEND PAGE AMEND PAGE AMEND PAGE AMEND 20.2 !!d 3 20.3 13 9 20.3 48 12 20.3 72c 3 20.! 14 11 20.3 46.1 9 20.3 72d 3 9

I 20.3 15 13 20.3 49 7 20.3 12e 3 g g ,,3 20.3 16 11 20.3 50 3 0'I' '

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O l Ncntmber 30,1990 20 2 l

ABWR Standard Plant

- CilAITER 20

( PAGE STATUS (Continued)

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s)4 November 30,1990 20 3

ABWR Standard Plant CHAPTER 20 PAGE STATUS (Continued)

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_ 20.3 320 *,49

. 15

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November 30, IWO 20 - 4

ABWR 1

)

Standard Plant 1

CHAPTER 20 l PAGE STATUS (Continued)

PAGE AMEND 1

20B 20 11 20B-21 11 l 208-!! 11 ,

208-23 11 208 24 11 208-25 11 20B 26 11 20B 27 11 20B-28 11 208 29 11 20B 30 11 208 31 11 208 32 11 208 33 11 208-34 11 20B-35 11 20B 36 11 20B 37 11 208 3B 11 20B-39 11 208-40 11 20P-41 11 ,

206 42 11 200-43 11 208 44 11 208 45 11 70B 46 15 20B 47 15 l

i O

Nmrmber 30,170 20 5

i ABWR mac nry c Standard Plant (6) M onitoring of e ssential ge ne r at or s, 1.2.2.1.2 Description of Plant Environs O transformers, and circuits is provided in the main contrel room. 1.2J lJ.1 Meteorotagy 1.2.1.2.5.3 power Conversion Systems Process The safety related structures and equipment Control Criteria are designed to retain required functions for the loads resulting from any tornado with (1) Control equipment is provided to control the characteristics not exceeding the values reactor pressure throughout its operating provided in Table 2.01.

range.

Tornado missiles are discussed in Section (2) The turbine is able to respond automatically 3.5.

to minor changes in load.

1.2.2.1.2.2 Hydrobgy (3) Control equipment in the feedwater system maintains the water level in the reactor The safety design basis of the plant provides vessel at the optimum level required by steam that structures of safety significance will be separators. unaffected by the hydrologic parameter envelope defined in Chapter 2.

(4) Control of the power conversion equipment is possible from a centrallocation. 1.2.2.1.2.3 Geology and Seismology 1.2.1.2.6 Power conversion Systems Criteria The structures of safety significance for the plant are designed to withstand a safe shutdown Components of the power conversion systems earthquake (SSE) which results in a freefield shall be designed to perform the following basic peak acceleration of 0.3g.

objectives:

O (1) produce electrical power from the steam coming from the reactor, condense the steam 1.2.2.2 General Arrangement of Structures and Equipment into water, and return the water to the The principal structures located in the reactor as heated feedwater with a major plant are the following:

portion of its gases and particulate impurities removed; and (1) Reactor Building includes the containment, drywell, and major portions of the nuclear (2) assure that any fission products or steam supply system, steam tunnel, refueling radioactivity associated with the steam and area, diesel generators, essential power, condensate during normal operation are safely non essential power, emergency core cooling contained inside the system or are released systems, ilVAC and supporting systems; under controlled conditions in accordance with waste disposal procedures. (2) Service Building personnel facilities, and portions of the non essential llVAC; 1.2.2 Plant Description (3) Control building includes the control 1.2.2.1 Site Characteristics room, the computer facility, the cable tunnels, some of the plant essential 1.2.2.1.1 Site Location switchgear, some of the essential power, reactor building water system and the The plant is located on a site adjacent to or essential HVAC system, close to a body of water with sufficient capacity for either once through or recirculated cooling (4) Turbine Building - bouses all equipment or a combination of both methods. associated with the main turbine generator.

O t.2 5 Amendment 7 l

1 l

ABWR awime RM C !

Standard Plant Other auxiliary equipment is also located in this building. !

(5) Radwaste Building . houses all equipment associated with the collection and processing of solid and liquid radioactive waste generated by the plant.

The arrangement of these structures on t.Le plant site is shown in Figure 1.21. F;gures 1.2 2 through 1.2 30 show the equipment arrangement in the principal buildings.

1.2.2.3 Nuclear Systems The nuclear system includes a direct. cycle forced circulation boiling water reactor that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power conditions is shown in Figure 1.12. i 1.2.2.3.1 Reactor Core and Control Rods Fuel for the reactor core and the control rods

are described in Section 4.2.

O O

Amendment 15 1.2 Sa

ABM DA61mAC Sipitdard Plant RIV C y result in a leak flow that is less than the ECCS

( capacity allowing full core coverage. The

's internal pumps are a wet motor design with no shaf t seals, thereby providing increased reliability, reduced maintenance requirements 1.2.23.2 Reactor Vessel and int (rnals and decreased operational radiation exposure.

The RIP has a low rotating inertia. Coupled The reactor vessel contains the core and with the solid state adjustable speed drives supporting structures; the steam separators and (ASD) the RIP can respond quickly to load dryers; the reactor internal pumps; the control transients and operator demands.

rod guide tubes; the distribution lines for the feedwater, core flooders, and standby liquid 1.2.2J.4 Residual licat Removal System 1 control; the in core instrumentation; and other components. The main concections to the vessel The residual heat removal (RIIR) system is a include steamlines, feedwater lines, control rod system of pumps, heat exchangers, and piping drive and in core nuclear instrument housings, that fulfills the following functions:

core flooder lines, residual heat removal lines, ;

head spray and vent lines, standby liquid control (1) removes decay and sensibic heat during and l line, core differential pretsure lines, internal after plant shutdown. I pump and pump deck differectial pressure sensing ,

lines, and water level instrumen.tation. (2) injects water into the reactor vesse! I following a loss of. coolant accident ic The reactor vesselin designed and fabricated reflood the core in conjunction with other in accordance with applicable codes for a core cooling systems (Subsection 1.2.2.4.8);

pressure ol 1250 psig. The nomina' operating pressure in the steam space above the separators (3) removes heat from the containment following is 1040 psia. The vessel is fabricated of low a loss of coolant accident to limit t he

, alloy steel and is clad internally with stainless increase in containment pressure. This is v steel or Ni Cr Fe Alloy (except for the top head, accomplished by cooling and recirculating l nozzles, and nozzle weld rancs which are unclad). the suppression pool water.

I l The reactor core is cooled by demineralized 1.2.2.3.5 Reactor Water Cleanup System water that enters the lower portion of the core and boils as it flows upward around the fuel T.ne reador water cleanup system (CUW) rods. The steam leaving the core is dried by recirculates a portion of reactor coolant steam separators and dryers located in the upper through a filter demineralizer to remove portion of the reactor vessel. The steam is then particulate and dissolved impurities from the

directed to the turbine through the main reactor coolant. It also removes excess coolant steamlines. Each steamline is provided with two from the reactor system under controlled isolation valves in series; one on each side of conditions and provides clean water for the l the containment barrier, reactor head spray nozzle.

1.2.233 Reactor Internal Recirculation Pumps 1.2.23.6 Leak Detection and Isolation System t The reactor internal pumps (RIPS) are internal The leak detection and isolation system

! pumps which provide a continuous internal circu- consists of temperature, pressure, flow, and lation path for core coolant flow, The RIPS fission. product sensors with associated are located at the bottom of the vessel. The instrumentation, alarms, and isolation l

pump motors are enclosed in pressure housings functions. This system detects and annunciates under the vessel. A break in the housing will leakage (and closes isolation valves, as l

l O

V Amendment 15 1.26

ABWR mime nry c Standard Plant TABLE 1.31 d(m COMPARISON Oa' NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are related to rated power out for a single plant unless otherwise noted) l This Plant

GESSAR NMP2 Grand Gulf AHWR BWR/6 BWR/5 BWR/6 Design 278 872 238 748 251 764 251 800 ,

hrmal and Hydraulle (See Section 4.4) r Rated power 3,926 3,579 3,323 3,833 (MWI)

Design power 4,005 3,729 3,463 4,025 (MWt) (ECCS design basis)

Steam flow 16.843 15.40 14.263 16.491 rate, Mlb/hr o

at 420 F (FW Temp)

Core coolant 115.1 104.0 108.5 112.5 g

i flow rate

\ (Mlb/hr)

Feedwater flow 15.807 15.372 14.564 16.455 rate (Mlb/hr)

System pressure, 1,040 1,040 1,020 1,040 nominalin steam dome (psla) ,

Average power 50.6 54.1 49.15 54.1 l density (kW/()

Maximum linear 13.4 13.4 13.4 13.4 heat generation rate (kW/ft) l Average linear 6.0 5.9 5.40 5.93 heat generation rate (kW/ft) l Maximum heat 361/00 361,600 354,255 361,600 flux 2

(Btu /hr/ft )

Parameter @r the core loading in Figue 4.31 used in the sensitisin' anab. sis.

~

l V

Amendment 15 13-2 1

ABWR uume i Standard Plant nrv q TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

This Plant

GESSAR NMP2 Grand Gulf ,

ABWR BWR/6 BWR/5 BWR/6 Design 278-872 238 748 251 764 251 800 Wr==I and Hydraulle (Continud Average Heat (later) 159,300 144,032 160,300 flux 2

(Btu /hr ft )

Maximum UO2 (later) 3,435 3,325 3,435 temperature (OF)

Average (later) 2,185 2,130 2,185 volumeterie fuel temperature (DF)

Average 566 565 566 565 cladding surface temperature (OF)

Minimum critical 1.16 1.20 1.24 1.20 power ratio (MCPR)

' Coolant $27,7 527.6 527.5 527,9 enthalpy at core inlet (Btu /lb)

Core maximum 75 79 76.2 76 voids within assemblics Core average 14.5 14.7 13.1 14.6 crit quality

(% steam)

Feedwater 420 420 420 420 temperature (OF)

Parametersfor the core loading in Figure 4.3-1 used in the sensitisity analysis.

O Amendment 15 l}3

I ABWR mau Standard Plant RN C ,

i

n. TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Contin.ied) i nls Plant

GESSAR NMP2 Grand Gulf l ABWR BWR/6 BWR/5 BWR/6 i Design 278 872 238 748 251 764 251 800

]

nertnal and Hsdraulle (Continued)

Design power l peaking factor Maximum relative 1.40 1.40 1.40 1.40 ast.emble power Local peaking 1.25 1.13 1.24 1.13 factor Axial peaking 1.40 1.40 1.40 1.40 factor

(' Total peaking factor 2.43 2.26 2.43 2.26 Nuclear (first corr)

(Section 4 3) l Watcr/UO2 2.95 2.70 2.55 2.70 [

volume ratio (cold)

Reactivity with <0.99 <0.99 <0.99 < 0.99 strongest control rod out (ke rd l Dynamic void 5.20e @ 7.16 -8,57 7.14 coefficient 102% rated (c/%) at core output average voids ' 39.2 40.9$ 40.54 4131

(%)(EOC rated output)

Fuel temperature -0360 0.112 0.419 0396 doppler coeffi-cient (c/0C)

(EOC ratea output) l 'Pararnetersfor the core loading in Figure 4.31 used in the sensitivity anob' sis.

Amendment t5 1.3-4

i ABWR mame '

Standard Plant RfV C TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) !

This Plant

GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/$ BWR/6 j

{}niga 278 472 238 748 251 764 2!1400 i Nuclear inrst core) (Continued)

I Initial average 2.22 1.90 1.90 1.70 I U 235 enrichment (%)

Initial cycle 9,950 9,138 9,200 7,500 exposure l (mwd /short ton)

Fuel assembly (Section 4.2)

Number of fuel 872 748 764 800 l assemblics Fuel rod array 8x8 8x8 8x8 8x8 Overalllength 176 176 176 176 (inches) ;

l Weight of UO2 435 456 466 458 l per assembly (Ib)(pellet type)

Weight of fuel 675 697 698 697 assembly (lb)

(includes che mel)

Fuel Rods (Section 4.2) !

l Number of fuel 62 62 63 62 rods per assembly Outside diameter 0.483 0,483 0.493 0.483 (in.)

Cladding " 0.032 0.032 0.032 l

thickness (in.)

Parametersfor the core loading in figure 4.3-1 used in the sensitivity analysis.

" Proprietary information, subrnitted under separate cover.

O Amendment 13 1.35

ABWR m umic Standard Plant my e D TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN L'HARACTERISTICS (Continued)

This Plant

GE3SAR NMP2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 Design 278 872 238 748 251 764 251 800 Fuel Rods (Continued)

Diameteral gap, 0.009 0.009 0.009 0.009 pellet to-cladding (in.)

Length of gas 13.23 9.48 14 9.48 plenum (in.)

Cladding Zircaloy 2 Zircaloy 1 Zircaloy 2 Zircaloy 2 material" Fuel Pelicia (Section 4.2)

Material UO2 U02 U02 UO2

[

Density (% of %.5 95 95 95

%. theoretical)

'" 0.410 0.416 0.410 Diameter (in.)

Length (in.) '" 0.410 0.420 0.410 j l

Fuel Channel 1 (Section 4.2) l l

l Thickness 0.100 0,120 0.100 0.120 (in.) !

Cross section 5.48 x 5.48 5.45 x 5.45 5.48 x 5.48 5.45 x 5.45 l diuensions (in.)

l Material Zircaloy 4 Zircaloy 4 Zircaloy 4 Zircaloy 4 Gre assembly (Section 4.2)

Fuel weight 379,221 341,640 265,551 365,693 l

l: as UO2 (Ib)

,n *Parametersfor the core loading in Figure 4.31 used in the sensitivity analysis.

1 ( " Free standingloaded tubes

"

Proprietary infonnation, submitted under separate cowr, Amendment 15 1.3-6

F 9

ABWR uwmac i

Standard Plant RK C TABLE 1.31 l

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) l This Plant

GESSAR NMP2 Grand Gulf ;

ABWR BWR/6 BWR/$ BWR/6 Desiga 278472 238 748 251 764 2!! 4 00 Core Assembly (Continued)

Core diameter 203.3 185.2 160.2 191.5 (equivalent)

(in.)

Core height 146 150 146 150 (active fuel)

(in.)

Reactor Control System (Chapters 4 and 7)

Method of Movable Movable Movable Movable variation of controf controlrods control control .

reactor power rods and and variable rods and rods and variable forced forced coolant variable forced variable forced W

coolant now coolant coolant Dow Dow Dow Number of 205 177 185 193 movable con-trof rods Shape of Cruciform Cruciform Cruciform Cruciform movab!c control rods Pitch of 12.2 12.0 12.0 12.0 movable control rods Control BC4 BC4 BC4 BC4 I materialin granules gianules granules granules l movable rods compacted compacted compacted compacted j in SS tubes in SS in SS in SS I tubes tubes tubes l

l

Parametersfor the core loading in Figure 4.3-1 used in the sensitivin' analysis.

Amendment 13 1.37

ABWR mouc Standard Plant Riv c TABLE 1.31 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

'1%Is Plant

GESSAR NMP2 Grand Gulf AHWR RWR/6 BWR/5 BWR/6 Design 278-872 238 748 251 764 251 800 Reactor Control System (Continued)

(Chapters 4 and 7) 6 Type of Bottom Botto:a Bottom Bottom control rod entry entry entry entry drives electric locking locking locking hydraulic piston piston piston fine motion Type of Burnable Burnable Burnable Burnable temporary poison; po: son; poison; poison; Reactivity gadolinia- gadolinia- gadolinia- gadolinia-control for urania urania urania urania inital core fuel rods fuel rods fuel rods fuel rods In core neutron Instrumentation (Chapters 4 and 7)

Total number 208 164 172 176 of LPRM dectce.

tors Number ofin- 52 41 43 44 core LPRM penetrations Number of LPRM 4 4 4 4 detectors per penetration Number of SRM " 4 4 6 penetrations l Number ofIRM 10 " 8 8 8 penetrations Total nuclear 62 53 43 58 instrument l penetrations l l

Tarameters for the core loading in Figure 4.3-1 used in the sensitidty analysis.

" For the ABil'R, the SRM and IRM are combined as one and called SRNM.

knendment 15 1.38 l

- . ._~ .. - . . . _ _ . . ,

i ABM MA6100AC RIT. A Standard Plant l TABLE 1.31 !

COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued)

This Plant GESSAR NMP2 Grand Gulf ABWR HWR/6 HWR/$ HWR/6 Design 278-872 238 748 251 764 251 800 l Core Mechanical (Continued)

In core neutron j instrumentation (Continued)

*

j Source range N/A monitor range Intermediate N/A l range monitor 1 range j Starteo range N/ *. N/A N/A I neutron monitor Power range Approximately1% power to 125% power monitors range Local powe' 208 164 172 176 range monitors

' 1 l

l

tage power 4 4 6 8 age monitors l

Number and type 5 Sb Be 7 Sb Be 7 Sb Be 7 Sb Be I ofin core i neutron sources Reactor Vessel (Section 5.3) l Material Low alloy Low alloy Low alloy Low alloy l

stec!/ steel / steel / steel /

stainless stainless stainless stainless and Ni Cr Fe clad clad clad Alloy l- clad .

Shutdowrs through criticality

" Prior to criticality to lowpower

" Shutdowrs through lowpower O

Amendment i 1.3-9

ABWR nume Standard Plant nev e p- Table 1.61 REFERENCED REPORTS AHWR SSAR Report No. Title Section No.

22A7007 GESSAR 11, 238 Nuclear Island, BH'R/6 Standard Plant, 3.7 !

General Electric Cornpany, March 1982, & Amendments 19.2 l

1 21. 19.3 19D.3 19D.7 19E.2 19E.3 APED.5750 Design and Perfortnance of General Electric Boiling H'ater 5.4 Reactor blain Steam Line isolation l'alves, General l Electric Co., Atomic Power Equipment Department, March 1%9.

NEDO.10029 An Analytic Study on Brittle Fracture of GE.BH'R Vessel 5.3 Subject to the Design Basis Accident.

NEDO 10527 CJ. Paone and J.A. Woolley, Rod Drop Accident Analysis 15.4 for Large Bolling H'ater Reactors, Licensing Topical ,

i Report, Marcb 1972.

l NEDO.10585 F.G. Brutchscy, et al., Beharlot of lodine in Reactor 15.2 Y H'ater During Plant Shutdown and Startup, August 1972.

NEDO.10871 J.M. Skarpelos and R.S. Gilbert, Technical Derivation of 12.2 BH'R 1971 Design Basis Radioacthe biaterial Source Terms, March 1985.

NEDO.11209 04A Nuclear En*rgy Business Operations Quality Assurance 17.1 Prograin Description, Rev.7, May 1987.

NEDE 20566 General Electric Company, Analyticci h!odelfor Loss-of-Coolant Analysis in Accordance with 10CFR$0, Appendix K Proprietary Document, November 1975.

NEDO.20953A J.A. Woolley, Three Dimensional BH'R Core Simulator, 4A.4 January 1977.

NEDO 20533 W.J. Bilanin, The G.E. Afark til Pressure Suppression 6.2.7 Containment Anabsical blodel, Jvue 1974.

NEDO.205331 W.J. Bilanin, The G.E. hfark Ill Pressure Suppression 6.2.7 Containment Analytical Afodel, Supplement 1, September 1975.

i' f

Amendment 8 14 2

i ABWR zwime Standard Plant an.c Table 1.61 REFERENCED REPORTS (Continued)

Report No. Title ABWR SSAR Section No.

NEDE 20566-P A General Electric Company Analyllical hf0delfor 6.3.7 Loss of Coolant Analysis in Accordance with 10CFR50, Appendir K September 1986.

NEDE 21354 P BWR Fuel Channel Afechanical Design and Deflection, 3.9.7 September 1976.

NEDE 21175 P BWR/6 Fuel Assembly Evaluation of Combined Safe 3.9.7 Shutdown Eanhquake (SSE) and Loss of Coolant Accident l

1 (LOCA) Loadings, November 19?6.

NEDO 21985 Functional Capability Criteria for Essential hiark 11 3.9.7 Piping, September 1978, prepared by Battelle Columbus 1 aboratories for General Electric Company.

l NEDO-21778-A Transient Pressure Rises Affecting Fracture Toughness 5.3.4 Requirementsfor Boiling Water Reactors, January 1979.

I NEDO 21052 FJ. Moody, blaximum Discharge Rate of Liquid l'apor 6.2.7 bilrtures from l'essels, General Electric Company, September,1975.

NEDO 21159 2 Airbome Releasesfrom BifRsfoe Environmentalimpact 12.2.3 NEDO 21159 Evaluations, NEDO 211431 H. Careway, l'. Nguyen, and P. Stancavege, Radiological 15.210 Accident The CONACO3 CODE, December 1981 15.6.7 NEDE 21514 BWR Scram System Reliability Analysis, December 1976, 19D.6.6

, General Electric Company, 1 NEDO 22155 GE Report, Generation and hiitigation of Cornbustible Gas 6.2.7 blittures in inened B1IR blark 1 Containments, June 1982.

O 1,6-3 Amendment 15 l

. - .1

ABWR mame Standard Plant nev c Table 1.61 REFERENCED REPORTS (Continued)

ABWR SSAR Report No. Title Section No.

NEDE 22056 Failure Rate Data Manualfor GE BifR Components, Rev.2 193.5 January 17,1986, Class 111, General Electric Company. 19D3.6 NEDE 220$6 Reliability Analysis Data Manual, General Electric 19E.2.5 Company, January 1986.

NEDO 23909A II.A. Careway, V.D. Nugyen, and D.G. Weiss, Control Roorn 15.6.7 Accident Errosure Evaluation CRDOS Program, February 1981.

NEDO 24057 P Assessrnent of Reactor Internals Vibration in Bif'R/4 and 3.9.7 NEDO 24057 Bil'R/S Plants, November 1977. Also NEDO 24057 P, O Amendment 1, December 1978, and NEDE 2 P24057 Amendment 2, June 1979.

NEDE 24326-1 P General Electric Environmental Qualification Prograrn, 3.9.7 Proprietary Document, January 1983. 3.11.7 I

i NEDE 30637 B.M. Gordon, Conosion and C(wosion Controlin BilRs. P.S.2 28a NEDC 30259 H.A. Careway, D.B. Townsend, B.W. Shaffer, A Technique 15.6.7 <

for Evaluation of Bif'R MSIV Leakage Contribution to

, RadiologicalDose Rate Calculations, September 1985.

NEDE 31152 P GE Bundle Designs, December 1988 4.23 Amendment 15 1.64

ABWR zwioorc Standard Plant Riv c TABLE 1.8 20 RGs Applicable to ABWR (Continued)

ABWR Appl. Issued Appil.

RG No. Regulatory Guide Title EgL Dait Cahld Cornments 1.60 Design Response Spectra for Seismic Design 1 12/73 Yes of Nuclear Power Plants.

1.61 Damping Values for Seismic Design of Nu- 0 10/73 Yes clear Power Plants.

.: 1.62 ManualInitiation of Protective Actions. 0 10/73 Yes 1.63 Electric Penetration Assemblics in Contain- 3 2/87 Yes ment Structures of Nuclear Power Plants.

1.64 Quality Assurance Requirements for the De- Superceded See Table sign of Nuclear Power Plants. 17.0-1 1.65 Materials and Inspections for Reactor Ves. 0 10/73 Yes sel Closure Studs.

i 1.68 laitial Test Programs for Water Cooled 2 8/78 Yes Reactor Power Plants.

1.68.1 Preoperational and inithi Startup Testing 1 1/77 Yes of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants.

1.68.2 laitial Stattup Test Program to Demonstrate 1 7/78 Yes Remote Shutdown Capability for Water Cooled Nuclear Power Plants.

1.68.3 Preoperational Testing of Instrument and 1 7/78 Yes Control Air Systems.

1.69 Concrete Radiation Shicids for Nuclear Po- 0 12/73 Yes wer Plants.

1.70 Standard Format and Content of Safety Ana- 3 11/78 Yes lysis Reports for Nuclear Power Plants.

1.71 Welder Qualifications for Areas of Limited 0 12/73 -- Interface Accessibility.

1.72 Spray Pond Piping Made From Fiberglass- 2 11/78 Yes Reinforced Thermosetting Resin.

O Amendment 14 1.lL 12

I ABWR ummac nry c Standard Plant TABLE 1.8 20 RGs Applicable to ABWR (Continued)

O l ABWR Appl. Issued Appil.

RG No. Renulatorv Guide Title EtL Dals cubic? comments 1.73 Qualification Tests of Electric Valve Ope- 0 1/74 Yes ,

rators Installed traide the Containment of I Nuclear Power Plants. !

1.74 Quality Assurance Terms and Definitions. Superceded See Table l 17.0 1 1 1.75 Physical Independence of Electric Systems. 2 9/78 Yes l 1.76 Design Basis Tornado for Nuclear Power 0 4/74 Yes j Plants 1.77 Assumptions Used for Evaluating a Control 0 5/74 No PWR onif Rod Ejection Accident for Pressurized Water Reactors. ,

1.78 Assumptions for Evaluating the Habitability 0 6/74 Yes of a Nuclear Power Plant Control Room Dur.

ing a Postulated flazardous Chetnical Re-lease.

1.79 Preoperational Testing of Emergency Core 1 9/75 No PWR only Cooling Systems for Pressurized Water Reac-tors.

1,81 Shared Emergency and Shutdown Electric Sys- 1 1/75 Yes tems for Multi Unit Power Plants.

1.82 Water Sources for Long Term Recirculation 1 11/85 Yes Cooling Following less of Coolant Accident.

1.83 in Service Inspection of Pressurized Water 1 7/75 No PWR only Reactor Steam Generator Tubes.

1.84 Design and Fabrication Code Case Acceptabi- 24 6/86 Yes lity, ASME Section !!!, Disition 1.

1.85 Materials Code Case Acceptability, ASME 24 6/86 Yes Section !!!, Disision 1.

1.S6 Termination of Operating Licenses for Nu- 0 6/74 - Interface l clear Reactors.

O Amendraent 15 1.8 13

ABWR m amic Standard Plant nry c

/^T TABLE 1.8 20 RGs Applicable to ABWR (Continued)

AllWR Appl. Issued Appil-RG No. Renulatorv Guide Title EtL llatt enble? Comments 1.87 Guidance for Construction of Class 1 Compo- 1 6/75 No l nents in Elevated Temperature Reactors (Supplement to AShtE Section 111 Code Cases 1592,1593,1594,1595, and 15%).

1.88 Collection, Storage, and hiaintenance of Superceded See Table Nuclear Power Plant Quality Assurance Re- 17.0-1 cords 1.89 Eiwironmental Qualification of Certain 1 7/84 Yes Electric Equipment important to Safety for Nuclear Power Plants.

1.90 Insersice inspection of Prestressed Con. 1 8/77 - Interface crete Containment Structures with Grouted Tendons.

,m

, 1.91 Evaluations of Explosions Postulated to 2 2/78 Yes

! I Occur on Transportation Routes Near Nuclear V Power Plants.

1.92 Combining blodel Responses and Spatial Com. 1 2/76 Yes ponents in Seismic Response Analysis.

33 Availability of Electric Power Sourecs. 0 12/74 Yes 1.94 Quality Assurance Requirements for Instal- - - - See Table lation, inspection, and Testing of Structu. 17.0 1 ral Steel During the Construction Phase of Nuclear Power Plants.

1.95 Protection of Nuclear Power Plant Control 1 1/77 Yes Room Operators Against an Accidental Chlo-rine Release.

1.96 Design of hiain Steam Isolation Valve leak- 1 6/76 Yes age Control Systems for Boiling Water Reac-tor Nuclear Power Plants.

1.97 Instrumentation for Light Water Cooled 3 $/83 Yes Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.

p) i

%J Amendment 15 1,841

I ABM maime o..a..a m... mc TABLE 1.8 20 RGs Applicable to ABWR (Continued)

ABW1t Appl. Issued Appil.

Efs.ha, Rapdatary Guide 11tle Egy., Dalg galdgI Comments 1.98 Assumptions for Evaluating the Potential 0 3/76 Yes 1 Radiological Consequences of a Radioac-tive Offgas System Failure in a Boihng .

Water Reactor.

1.99 Radiation Embrittlement of Reactor 2 $/88 Yes Vessel Materials, fl 1.100 Selsmic Qualification of Electric Equip.

ment for Nuclear Power Plants.

2 6/88 Yes 1.101 Emergency Planning and Preparedness for 2 10/81 Yes Nudear Power Reactors.

1.102 Flood Protection for Nudear Power Plants. 1 9/76 Yes 1.105 lastrument Setpoints for Safety Related 2 2/86 Yes Systems. ;

1.106 Therinal Overload Protection for Electric 1 3/77 Yes Motors on Motor Operated Valves.

1.107 Qualifications for Cement Grouting for Pre- 1 2/77 Yes stressing Tendons in Containment Struc-tures.

1.108 Periodic Testing of Diesel Generator Units 1 8/77 Yes Used as Onsite Electric Power Systems at Nudear Power Plants.

1.109 Calculation of Annual Doses to Man from 1 10/77 Yes

(

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CPR Part 50, Appendix 1.

1.110 Cost Benefit Analysis for Radwaste Systems 0 3/76 Yes for Light Water Cooled Nudear Power Plants.

1.11.1 Methods for Estimating Atmospheric Trans- 1 7/77 Yes port and Dispersion of Gaseous Effluents in Routine Releases from Ught. Water Cooled Reactors.

O 1 3 -15 Amendment 10 l

l

ABWR 23umac Standard Plant nev c n Table 1.91 b

SUMMARY

OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT ITEM INTERFACE NO. SUBJECT TYPE SUBSECTION

-1.1 Standard review plan sections for remainder of CorJirmatory 1.8.4 plant identified as ' Interface" in Table 1,84; i 1.2 Applicabil'ty of regulatory guides for remainder Confirmatory 1.8.4 of plant in jentified as " Interface" in Table 13 20 13 Applicability of Experience Information Confirmatory / 1.8.4 for remainder of plant identified as Procedural

' interface"in Table 1.8-22 1.4 Emcrgency procedwes and emergency procedures Procedural 1A3.1 training prog.am 1.5 Procedures for removing safety-related systems Procedural 1A3.2 from service h

Q.

1.6 Inplant radiatiom moniton 1g Procedural 1A33 1.7- Reporting of Failures of Reactor System Procedural 1A3.5 Relief Valves 14 Report on ECCS Outage ' Procedural 1A3.6 2.1 - Envelope of ABWR Standard Plant Site Design Design & 2.2.1 Parameters Confirmatory 2.2 - StacLrd Review Plan Site Characteristics Confirmatory 2.2.2 l 23 CRAC 2 Computer Code Calculations Confirmatory 2.23 3.1 Site Specific Design Basis Wind Confirmatory 333.1 :

3.2 Site Specific Design Basis Tornado Confirmatory 333.2 33 Effect of remainder of plant st.uctures, Confirmatory 3333 systems and components not designed to tornado loads 3.4 Flood Elevation Design 3.43.1 3.5 Ground Water Elevation Design 3.43.2 3.6 Protection of ultimate heat sink Confirrr.atory 3.5.4.1 Amendment 15 - 1.9-2

ABWR nA6ta^c i Standard Plant Rev C Table 1.91

SUMMARY

OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT (Continued)

ITEM INTERFACE NO. SURJECT 'IYPE SUBSECTION Missles generated by natural phenomena from Confirmatory 3.5.4.2

{ 3.7 remainder of plant 3.8 Site proximity missiles and aircraft hazards Confirmatory 3.5.43 3.9 Protection against secondary missiles inside Confirmatory 3.5.4.4 containment

. 3.9a - Impact of Failure of Non Safety Related Confirmatory 3.5.4.5 Items Due to Design Basis Tornado 3.10 Details of pipe break analysis results Confirmatory 3.6.4.1 and protection methods 3.11 Leak before break analysis results Confirmatory 3.6.4.2 l 3.11a Seismic Parameters Confirmatory 3.7.5.1 -

3.12 Foundation Waterproofing Confirmatory 3.8.6.1 3.13 Site Specific Physical Properties and Confirmatory 3.8.6.2 Foundation Settlement

'3.14 Reactor Internals Vibration Analysis, Confirmatory 3.9.7.1 Measurement and Inspection Programs 3.15 ASME Class 2 or 3 Quality Group Confirmatory 3.9.7.2

. Components with 60 Year Design Life l 3.15a Pump and Valve Inservice Testing Program Confirmatory 3.9.73 3.16_ Equipment qualification report Confirmatory 3.10.5.1 3.17 Dynamic qualification report Confirmatory 3.10.5.2 3.18 Environmental Qualification Document Confirmatory 3.11.6.1 3.19 Enviromental Qualification Records Confirmatory 3.11.6.2 4.1 CRD inspection Program Procedural 4.5.3.1 5,1 Water Chemistry Design 5.2.6.1 l 5.2 Conversion ofIndicators Procedural 5.2.6.2 Amendment 15 1.93 mis-mmmmmmmmmmme ----um--mmemii-sm- - m

ABWR nuiwac Standard Plant nev c-Table 1.91 lr3 d

SUMMARY

OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT (Continued)

-ITEh! INTERFACE NO. SUBJECT TYPE SUBSECTION 53 Fracture Toughness Data Confirmatory 53.4.1 5.4 hiaterials and Surveillance Capsule Confirmatory 53.4.2 6.1 Protection Coatings and Organic hiaterials Confirmatory 6.13.1 6.2 External Temperature Confirmatory 6.4.7.1 63 hieterology(X/Qs) Confirmatory 6.4.7.2 ,

6.4 Toxic Gases Confirmatory 6.4.73 7.1 Effects of Sation Blackout on HVAC Confirmatory 7.8.1 7.2 Electrostatie Discharge on Exposed Confirmatory - 7.8.2 Equipment Components 73 Localized liigh Heat Spots in Semiconductor Confirmatory 7.83 hiaterial for Computing Desices 8.1 Stability of offsite power system Confirmatory 8.1.4.1 8.2 Picsel Generator Reliability Procedural 8.1.4.2 83 - Class IE Feeder Circuits Design 8.23.1 l

l 8.4 Non classIE Feeders Desi p 8.23.2 l

8.233 8.5 Specific ABWR Standard Plant / remainder of plant Design l power sysytem interfaces 8.6 Interupting Capability of Electrical Confirmatory 83.4.1 Distribution Equipment 8.7 Diesel Generator Design Details Confirmatory 83.4.2 8.8 Certified Proof Tests on Cable Samples Confirmatory 83.43 l 8.9 ElectricalPenetration Assemblies Confirmatory 83.4.4 1

8.10 Analysis Testing for Spatial Seperation Confirmatory 83.4.5 per IEEE 304 Amendment 15 1.94

, , . . - , , . . . - - - - - - - e ABM 2346teoac Standard Plant nev.c Table 1.91

SUMMARY

OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT (Continued)

ITEM INTERFACE NO. SUBJECT Th'PE SUBSECTION 8.11 DC Voltage Analysis Confirmatory 83.4.6 8.12 Seismic Qualification of Eyewash Equipment Confirmatory 83.4.7 8.13 Diesel Generator lead Table Changes Confirmatory 83.4.8 8.14 Offsite Power Supply Arrangements Procedural 83.4.9 8.15 Diesel Generator Qualification Tests Confirmatory 8 3.4.10 8.16 Defective Refurbished Circuit Breakers Confirmatory 8 3.4.11

- 8.17 Minimum Starting Voltages for Class' Confirmatory 8 3.4.12 1E Motors 9.1 New Fuel Storage Racks Criticality Analysis Confirmatory 9.1.6.1 9.2 New Fuel Storage Racks Dynamic and Impact Confirmatory 9.1.6.2 Analysis 93 Spent Fuel Storage Racks Criticality Analysis Confirmatory 9.1.63 9.4 Spent Fuct Storage Rack lead Drop Analysis Confirmatory 9.1.6.4 9.5 Ultimate heat sink capability Design 9.2.17.1 9.6 Makeup water system capability Design 9.2.17.2 9.7 Potable and Sanitary Water System Design 9.2.17 3 9.8 Radioactive Drain Transfer System Collection Design 9 3.12.1 Piping 9.9 Contamination of DG Combustion Air intake Confirmatory - 9.5.13.1

' 9.10 Use of Communication System in Emergencies Procedural 9.5.13.2 9.11 Maintenance and Testing Procedures for Procedural 9.5.13 3 Communication Equipment 9.12 Fire Hazard Analysis Database Confirmatory 9A.63 Amendment 1$ 1.9-5

ABWR nuime Standard Plant nev c eg - Table 1.91

-( ;-

SUMMARY

OF ABWR STANDARD PLANT IN'I'ERFACES i

WITH REMAINDER OF PLANT (Continued)

ITEM INTERFACE NO. SUBJECT TYPE SUBSECTION 12.1 Regulatory Guide 8.10 Confirmatory 12.1.4.1 12.2 Regulatory Guide 1.8 Confirmatory 12.1.4.2 123 Occupational radiation exposure Procedural 12.1.43 13.1 Physical Security Interfaces Design & 13.6 3 Procedural i 18.1 Main Control Room Design & 18.5 Confirmatory 19.1 Long term training upgrade Procedural 19A3.1 19.2 - 1.ong term program of upgrading of procedures Procedural 19A3.2

,G

/O 193 Purge system reliability Procedural 19A33 !

19.4 Licensing emergency support facility Procedural 19A3.4 19.5 In plant radiation monitoring Procedural 19A3.5 19.6 - Feedback of operating, design and construction Procedural 19A3.6 experience 19.7 Organization and staffing to oversee design and Procedural 19A3.7 construction 19.8 . Quality Assurance Program Design 19B ',.1 19.9 Prevention of Core Damage Procedural 19B3.2 19.10 Protection from ExternalThreats Design 19B33 19.11 Ultimate Heat Sink Models Design ' 19B3.4 19.12 Ultimate Heat Sink Reliability Design - 19B3.5 19.12a Main Transformer Design Design 19B3.6 l 19.13 ' Plant Siting Procedural 19B3.7 l 19.14 Interdesciplinary Design Reviews Procedural 19B3.8 Amendment 15 1.96 l'

l

A.BWR msimac Standard Plant nrv c

, 1A.2.17 Instruments for Monitoring 1A.2.19 Review and Modify Procedures (di Accident Conditions [II.F.3]

NRC Position for Removing Safety Related Systems From Service [lI.K.1(10)]

NRC Position Proside instrumentation adequate for monitor-ing plant conditions following an accident that in- Review and modify (as required) procedures for cludes core damage. removing safety related systems from service (and restoring to service) to assure operability status is Response known.

The ABWR Standard Plant is designed in accor- Response dance with Regulatory Guide 1.97, Revision 3. A de-tailed assessment of the Regulatory Guide, including See Subsection 1A.3.2 for interface requirement. {

the list of instruments, is found in Section 7.5.

1A.2.20 Describe Automatic and Manual 1A.2.18 Safety Related Valve Position Actions for Proper Functioning of Indication [lI.K.1(5)] Auxillary Heat Removal Systems When FW System Not Operable (ll.K.1(22)]

NRC Position NRC Position (1) Review all valve positions and positionir.g re.

quirements and positive controls and all related For boiling water reactors, describe the auto-test and maintenance procedures to assure matic and manual actions necessary for proper func-proper ESF functioning,if required, tioning of the auxiliary heat removal systems that are used when the main feedwater system is not operable (2) Verify that AFW valves are in open position. (see Bulletin 79 08, item 3).

/]

v Response Response (1) The ABWR Standard Plant is equipped with if the main feedwater system is not operable, a status monitoring that satistics the requirements reactor scram wdl be automatically initiated when of Regulatory Guide 1A7. See Subsection 7.1.2 reactor e level fai to Level 3. The operator can for detailed information on the status monitor- then mard.yici, ce the RCIC system from the ing equipment and e pability provided in the main control room, or the system will be automati-ABWR Standard Plant' design, cally initiated as hereinafter described. Reactor water level will continue to decrease due to boil-off in addition to the status monitoring, plant spe- until the low low level setpoint, Level 2, is reached.

cific procedures (see Subsection 1A.3.2) will At this point, the reactor core isolation cooling assure that independent verification of system (RCIC) system will be automatically initiated to sup-line ups is applied to valve and electrical ply makeup water to the RPV. This system will line ups for all safety related equipment, to continue automatic injection until the reactor water surveillance procedures, and to restoration level reaches Level 8, at which time the RCIC steam following maintenance. Through these proce- supply valve is closed.

dures, approval will be required for the performance of surveillance tests and mainte- In the nonaccident case, the RCIC system is not.

nance, including equipment removal from mally the only makeup system utilized to furnish sub-senice and return to senice, sequent makeup water to the RPV. When level reaches Level 2 again due to loss of inventory (2) This requirement is not applicable to the through the main steam relief valves or to the main ABWR. It applies only to Babcock & Wilcox condenser, the RCIC system automatically restarts

,q designed reactors, as described in Subsection 1A.2.22. This system then iJ Amendment 8 1A.2 11

l l

ABWR u^am^c H Standard Plant arv c maintains the coolant makeup supply. RPV pressure cral is fully described in NEDO 24708A, Additional <

Infonnation Requiredfor NRC Staff Generic Report l is theregulated by thevalves main turbine bypass automatic or manual which discharge to onoperation Bolling it'aterofReactors. An outline of this ;

J the condenser, description as applicable to the ABWR Standard Plant is provided in the following paragraphs.

To reniove decay heat during a planned isola-tion event, assuming that the main condenser is not Figure 7.71 illustrates the reactor vessel eleva-available, the safety relief valves are utilized to dump tions covered by each water level range. Additional the residual steam to the suppression pool. The sup- details may be found in Figure 5.13 (nuclear boiler pression pool will then be cooled by manual align- system P&lD). The instruments that sense the water ment of the RHR system into the suppression pool level are differential pressure devices calibrated to be cooling mode, which routes the pool water through accurate at a specific vessel pressure and liquid tem-the RFIR heat exchangers, cools it, and returns it to perature condition. The following is a description of the suppression poolin a closed cycle. Makeup each water level range.

water to the RPV is still supplied by the RCIC system. (1) Shutdown water lent range: This range is used to monitor the reactor water level during the For the accident case with the RPV at high pres- shutdown condition when the reactor system is sure, the llPCF systems can also be utilized to auto- flooded for maintenance and head removal. The matically provide the required makeup flow when the water level measurement design is the conden-water level drops below Level 1.5 setpoint. No sate reference chamber leg type that is not com-manual operations are required. if the IIPCF sys- pensated for changes in density. The vessel tem-tems are postulated to fail at these same conditions perature and pressure conditions that are used and the RCIC capacity is insufficient, the automatic for the calibration are 0 psig and 120 F water in

. depressurization system (ADS) will automatically the vessel. The two vesselinstrument penetra-lions elevations used for this water level mea-initiate depressurization of the RPV to permit the low pressure ECCS/LPFL mode of RHR systems to surement are located at the top of the RPV head provide makeup coolant, and the instrument tap just below the bottom of the dryer skirt.

Therefore,it can be seen that although manual actions can be taken to mitigate the consequences of (2) Narrow water level range: This range uses for its a loss of feedwater, there are no short term manual RPV taps the elevation above the main steam actions which must be taken. Sufficient systems exist outlet nozzle and the tap at an elevation near the

, to automatically mitigate these consequences, bottom of the dryer skirt. The instruments are calibrated to be accurate at the normal operating 1 A.2.21 Describe Uses and Types of RV points. The water level measurement design is p Level Indication for Automatic and the condensate reference chamber type,is not Manual Initiation of Safety Systems density compensated, and uses differential pres-

[II.K.1(23)] sure devices as its primary elements. The feedwater control system uses this range for its NRC Position water level control and indication inputs.

l For boiling water reactors, describe all uses and (3) Wide water level range: This range uses for its j' types of reactor vessellevelindication for both RPV taps the elevation above the main steam automatic and manualinitiation of safety systems. outlet nozzle and the taps at an elevation near Describe other instrumentation that might give the the top of the active fuel. The instruments are operator the same information on plant status. See calibrated to be accurate at the normal power op-Bulletin 79-08, item 4. erating point. The water. level measurement de-sign is the condensate reference type, is not den.

Response sity compensated, and uses differential pressure devices as its primary elements. These instru-

. The water level measurement for BWRs in gen- ments provide inputs to various safety systems O

Amendment 15 t A.212

1 4

ABWR u r6icoa c l Standard Plant REV C l l

differential pressure signals which isolate the RCIC (5) Earlier initiation of ECC systems, j f' turbine are processed through the leak detection and

(_ isolation system (LDS). Spurious trips are avoided because the RCIC has a bypass start system con-(6) Heat removal through emergency condensers, j l

trolled by valves F037 and F045 (,ee Figure 5.4-8, (7) Offset valve setpoints to open fewer valves per i l

RCIC P&ID). challenge, On receipt r / RCIC start signals, bypass valve (8) Installation of additional relief valves with a block F045 opens to r :essurize the Poe downstream and or isolation-valve feature to eliminate opening of accelerate the t arbine. The bypass line via F045 is the safety / relief valves (SRV's), consistent with small (1 inch) and natural'y limits t' initial flow the AShiE Code, surge such tha a differen'.ial pressure spike in the upstream pipe vill not owur. (9) Increasing the high steam line flow setpoint for main steam line isolat'on valve (htSIV) closure, After a p cduermined delay (approximately 510 seconds), steam supply valve F037 opens to ad- (10) Lowering the pres ure setpoint for htSIV mit full steam flow to the turbine. At this stage, the Closure, line downstream is already pressurized. Thus,it is highly unlikely that a differential pressure spike (11) Reducing the testing frequency of the could occur during any phase of the normal start up htSIV's, process.

1A.2.24 Reduction of Challenges and Failures of Relief Valves - Feasibility (13) Early removal ofleaking valves.

Study and System Modification

[ll.K.3(16)] An investigation of the feasibility and constraints of reducing challenges to the relief valves by use of

/7 NRC Position the aforementioned methods should be conducted, d Other methods should also be included in the feasi.

The record of rel ef valve failures to close for all bility study. Those changes which are shown to boiling water reactora (BWRs) in the past 3 years of reduce relief valve challenges without compromising plant operation is ap] proximately 30 in 73 reactor- the performance of the relief valves or other systems years (0.41 failures pe r reactor year). This has dem- should be implemented. Challenges to the relief onstrated that the failure of a relief valve to close valves should be reduced substantially (by an order would be the most fixely cause of a small break loss- of magnitude).

of coolant sccident (LOCA). The high failure rate is l

the result of a high relief-valve challenge rate and a Response I relatively high failure rate per challenge (0.16 fail-ures per challenge). Typically, five valves are chal- Response lenged in each event. This results in an equivalent failure rate per challenge of 0.03. The challenge and Gneral Electric and the BWR Owners' Group I failure rates can be reduced in the following ways: reponded to this requirement in Reference 6. This l response, which was based on a review of existing (1) Additional anticipatory scram on loss of feedwa- operating information on the challange rate of relief ter, valves, concluded that the BWR/6 product line had already achieved the " order of magnatude" level of (2) Revised relief valve actuation setpoints, reduction in SRV challange rate. The ABWR relief l

valve system also has similar design features which (3) Increased emergency core cooling (ECC) flow, also reduce the SRV challenge rate. With regard to inadvertently opened relief valves (IORV), the (4) Lower operating pressures, BWR/6 pint design evaluated for the Owners' l

Group report reflected a reduced level if IORC l compared with previous design because of h

O Amendment 15 t A.2 14

MM 23A6100AC Standard Plant climination of the pilot operated relief vane type of may c h

design. The ASWR design bas also eliminated the pilot operated relief valve type of design.

For the ABWR which has a solid state logic desigh with redundancy, the likelihood of an IORY is the sarne oc less than the BWR/6 design evaluated la connection with the Owners' Group report. The redundant solid state design has beco selected la order that the frequency of IORV with solid state logic becomes low enough so as to achieve the order of magnitude reduction in total SRV challenge rate required by NUREG 0737.

The redundant solid state design for SRV operation in the pressure relief mode consists of two duplicated reicroprocessor channels. Each microprocessor channel activates a separate load driver and both load drivers must be activated to cause operation of the SRV's in the relief mode.

Operation of the SRV's in the ADS mode also requires activation of two microproccuor channels i with separate load drivers to prevent unwanted SRV operation; however, two separate dual channel ,

systems are used to assure reliable operation in the '

ADS mode. Reliable operation in the pressure relief mode is assured by direct opening of the SRV against spring force, i

O

.\nwno ns ni t. LA 2144

MM . 23A6100Ac Standard Plant artv c switchover is implemented, licensees should verify

!] : that clear and cogent procedures exist for the manual i

.[v/- switchover of the RCIC system suction from the con-densate storage tank to the suppression pool.

I

Response

1 The RCIC system provided in the ABWR Stan-dard Plant includes an automatic switchover feature which will change the pump suction source from the l l condensate storage pool to the suppression pool.

The safety grade switchover will automatically occur upon receipt of a low level signal from the condensate storage pool or a high level signal from the suppression pool.

See Subsection 73.1.1.1.3 for additional infor-l ~- mation.

1A.2.29. ' Confirm Adequacy of Space Cooling for High Pressure Coolant injection and Reactor Core Isolation ;

Cooling Systems (II.K.3(24)]

NRC Position Long term operation of the reactor core isola.

FY tion cooling (RCIC) and high pressure coolant injec.

.V tion (IIPCI) systems may require space cooling to maintain the pump-room temperatures within allow-able limits. Licensees should verify the acceptability

- of the consequences of a complete loss of alternat-ing current power. The RCIC and HPCI systems should be designed to withstand a complete loss of '

offsite alternating current' power to their support sys-tems, including coolers, for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Response

The ABWR high pressure core flooder (HPCF)-

and the reactor core isolation cooling (RCIC) systems are provided space cooling via individual

' room safety grade air coditioning systems (See L Subsection 9.4.5). If all offsite power is lost, space cooling for the HPCF and RCIC system equipment

.would not be lost because the motor power supply for each system is from seperate essential power supplies.

(..

N. s Amendment 15 1A.2 17

)

ABWR numc

. Standard Plant nov c 1 A.2.30 Effect of less of Alternating. piping within primary coritainment. Two redundant Current Power on Pump Seals [II.K.3(25)] 7 day supplies of bottled air are available to compensate for leakage during long te;m usage, with NRC Positten replacement capability being provided for the remainder of the postulated accident to assure The licensees should determine, on a plant system functionaloperability. A maximum of three specific basis, by analysis or experiment, the of eight ADS valves need function to meet consequences of a loss of cooling water to the short term demands (see Subsection 19.3.1.3.1) and reactor recirculation pump seal coolers. The pump the functional operability of only one ADS valve will seals should be designed to withstand a complete loss fulfilllonger term needs. Less of pneumatic supply of alternating current (ac) power for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . pressure to the ADS SRV accumulator is alarmed to Adequacy of the seal design should be demonstrated, provide the reactor operator with indication of the failure of any of the redundant systems under hostle Response environmental condition.

The ABWR design features internal tecircula. The BWR Owners' Group sponsored an tion pumps (RIP) which do not require shaft seals. evaluation of the adequacy of the ADS During a LOPA, the RIPS shutdown automatically configurations. Evaluation results are summarized in but there are no shaft seals which require cooling the following paragraph, water restoration.

The accumulators are designed to provide two I A.2.31 Study and Verify Qualllication of ADS actuations at 70% of drywell design pressure, Accumulators on Automatic Depressurl. which is equivalent to 4 to 5 actuations at zation System Valves [II.K.3(28)] atmospheric pressure. The ADS valves are designed to operate at 70% of drywell design pressure because NRC Position that is the maximum pressure for which reactor depressurization through the ADS valves is required.

Safety analysis reports claim that air or nitrogen The greater drywell design pressures are associated accumulators for the automatic depressurization only with the short duration primary system system (ADS) valves are provided with sufficient blowdown in the drywell immediately following a capacity to cycle the valves open five times at design large pipe rupture for which ADS operation is not pressures, GE has also stated that the emergency required. For large breaks which result in higher core cooling (ECC) systems are designed to with. drywell pressure, sufficient reactor depressurization stand a hostile environment and still perform their occurs due to the break to preclude the need for function for 100 days following an accident. Licensee ADS. One ADS actuation at 70% of drywell design should verify that the accumulators on the ADS pressure is sufficient to depressurize the reactor and

. valves meet these requirements, even considering allow inventory makeup by the low pressure ECC normallcakage if this cannot be demonstrated, the systems. However, for conservatism, the Licensee must show that the accumulator design is accumulators are sized to allow 2 actuators at 70% of still acceptable. drywell pressure. See Subsection 6.8.1 for a description of the ADS N pneumatic supply.

2 The accumulators for the ADS valves are sized to provide two operating cycles at 70% of drywell design pressure. This cyclic capability is validated

. during preoperational testing at the station. The accumulators are safety grade components.

The 100 day, post. accident functional operability requirement is met through conservative design and redundancy; eight ADS valves are provided with code. qualified accumulators and seismic Category I Amenameni 6 1A.2-18

i MkN 23A6100AC ne v. c

. Standard Plant ,

,a

(., 1A.3 INTERFACES- appropiate, to improve the availability of the emer. l gency core cooling equipment.

1A.3.1 Emergency Procedures and Emergency Procedures Training Program Emergency procedures, developed from the

' emergency procedures guidelines, shall be provided and implemented prior to fuel loading. (See Subsec-tion 1A.2.1).

1A.3.2 Review and Modify Procedures for

- Removing Safety Related Systems From !

Senice Procedures shall br reviewed and modified (as required) for removing safety related systems from senice (and restoring to senice) to assure operabil-ity status is known. (See Subsections 1A.2.18 and 19) 1A.3.3 In Plant Radiation Monitoring -

. Equipment and training and procedures shall be prosided for accurately determining the airborne io-1 dine concentration in areas within the facility where .

._ . plant personnel may be present during the accident.

(See Subsection 1A.2.18) 1A.3.4 Reporting Failures of Reactor :

System ReliefValves Failures of reactor system relief valves shall be -f reported in the annual report to the NRC- (See Sub-1 seetion 1A.2.3.21.1).

- [1A.3.5 Report on ECCS Outages Starting from the date of commercial opera-tions, an annual report should be submitted which in-cludes instance of emergency core cooling system un-availability because of component failure, mainte-nance outage (both forced or planned), or testing, the following information shall be collected: i

_ (1) Outage date (2) Duration of outage (3) Cause of outage (4) Emergency core cooling system or component involved '

(5) Corrective action taken l p The above information shall be assembled into a !

.t report, which will also include a discussion of any changes, proposed or implemented, deemed Amendment 15 - 1A 31 l

l'

ABWR mamo I

- Standard Plant - REV.D l r CHAPTER 2 TABLE OF CONTENTS Sectlan Tiue Ease >

l 2 SITE CHARACTERISTICS -,

2.0

SUMMARY

2.0-1 2.1 LIMITS IMPOSED ON SRP SECTION 11 t ACCEITANCE CRITTRIA HY ABWR DESIGN 2.1 1 2.2 REOUIREMENTS FOR DETERMINATION OF 2.2-1 AHWR SITE ACCEPTABILITY 2.2.1 Design Bases Events 2.2 1 2.2.2 Severe Accident 2.2 1 23 INTERFACES 23 t E 23.1 Envelope of ABWR Standard Plant 23-1 Site Design Parameters 1

23.2 Standard Review Plan Site Characteristics 23-1 233 CRAC 2 Computer Code Calculations 23-1

' APPENDIX 2A INPtrr TO CRAC 2 COMPUTER CODE FOR

. DETERMINATION OF ABWR SITE ACCEITABILITY l

l 1

I :'

l'

(:

2 ..in

. Amendment 15 ,

i r , ,x

ABWR mime Standard Plant nrv.n g TABLE 2.01

, t-

'd ENVELOPE OF ABWR STANDARD PIANT SITE DESIGN PARAMETERS hlaximum Ground Water Ixvel: Extreine Wlod: Basic Wind Speed:

2 feet below grade 110 mph (1)/130 mph (2) blaximum Flood (or Tsunami) level:(3) Tornado:(4)

I foot below grade hiaximum tornado wind speed: 260 mph

- Translational velocity: 57 mph

- Radius: 453 ft Precipitation (for Roof Design): hiaximum atm AP: 1.46 psid Maximum rainfall rate: 19.4 in/hr(8) hiissile Spectra: Per ANSI /ANS 2.3 Maximum snowload: 50 lb/sq. ft.

Design Temperatures: Soll Properties:

Ambient Minimum Bearing Capacity (demand): 15ksf 1% Exceedance Values Minimum Shear Wave Velocity: 1000 fps Maximum: 1000F dry bulb /770F coincident wet - Liquification Potential:

bulb None at plant site resulting 3

-Minimum: 100F from OBE and SSE(7) 3 0% Exceedance Values (Historical limit)

- Maximum: 1150F dry bulb /820F coincident wet Seismology:

r bulb OBE Peak Ground Acceleration (PGA):

i l

b]- - Minimum: 400F '

Emergency Cooling Water Inlet: 950F O.10g(5) (6)

SSE PGA : 030g(5)

- Condenser Cooling Water Inlet : 11000 F SSE Response Spectra: per Reg. Guide 1.60 SSE Time History: Envelope SSE Response Spectra U) 50-year recurrence interval; value to be utilized for design of non safety related structures only.

s (2) 100-year recurrence interval; value to be utilized for design for safety related structures only.

U) Probable marimum flood level (PbfF), as defined in ANSI /ANS 2.8, " Determining Design Basis Flooding at Power Reactor Sites."

[#) 1,000,000-year tomado recurrence interval, with associatedparameters based on ANSI /ANS 2.3.

0)' Free field, atplant grade elevation.

(6) For conservatism, a value of 0.15g is employed to evaluate structural and component response

in Chapter 3.

3

0) See item 3 in Section 3A.]for additionalinfonnation. R (3 hiaximum value for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 sq. mile PhfP with ratio of 5 minutes to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months PhfP as found in Q) @)

National Weather Source Publication HbfR No. 52. Afaximum short term rate; 6.2in/5 min.

Amendment 15 2.02

1 ABWR 2mimo Standard Plant _

nrw. n l fN SECTION 2.2

\l CONTENTS l

section IltLe Eass 2.2.1 Desien Basis Events 2.21 2.2.2 Severe Accident 2.21 TAHLES Iable Iitic Ease E

2.21 Dose Related Goals 2.22 2.2-2 CRAC 2 Data input Listing 2.23 i

f '.

-%)

h 2.2il l -'

t Amendment 15

ABWR = *^n Standard Plant _.

nev. n 2.2 REOUIREMENTS FOR UTILITY are to be supplied by the licensing utility DETER 111 NATION OF AllWR SITE as specified in the CR AC 2 manual

- (VT ACCEPTAlllLITY (NUREG/CR 2326) and are site specific.

This section provides the requirements for the The basic reference case assumes no evacuation determination of ABWR site acceptability, or radiation shielding (Subgroup Evacuation) for risk Acceptability is required from the standpoint of both and dose calculations. However,if the results of design bases and severe accident. such an evaluation for a specific site are unacceptable, site specific evacuation and shielding 2.2.1 Design Bases Events parameters may be substituted in lieu of the reference values in Subgroup Evacuation.

For design bases events, the site is acceptable if all of the site characteristics fall within the envelope Analysis: The analysis for evaluation of a specific site of ABWR Standard Plant site design parameters will be accomplished with the CRAC 2 computer given in Table 2.0-1. For cases where a characteristic code as modified through Sandia National Lab mod exceeds its envelope it will be necessary for the 46. Basic input and code characteristics are applicant re?crencing the ABWR design to submit deseribed in NUREG/CR 2326 and analyses to demonstrate that the overall set of site NUREG/CR 2552, characteristic; do not exceed the capability of the design.

2.2.2 Severe Accidents The ABWR PRA results were calculated for an average or typical site, as outlined in Subsection 19E.3. Although these usuhs for . a good basis for 73 assessing the general ABWR capability to satisfy offsite dose related goals, they do not form a basis (V) for concluding that the ABWR would meet dose related goals at a specific site whose characteristics cannot be defined at the point of ABWR certification. Consistent with the certification concept that all key technical issues be resolved before certification, it is appropriate to define the process for determining future site acceptability.

This process is defined below in terms of (1) acceptance criteria, (2) data input, and (3) analysis.

Accentance Criteria: Site acceptability for severe accidents will be based upon a calculation using the CRAC 2 computer code. The results of such a calculation will be compared to the goals of Table 19E.3 7 as shown in Table 2.21. The site will be deemed acceptable if the results fall within the given goals.

Data input: The input to the CRAC 2 computer code will be a combination of ABWR and site parameters.

The CRAC 2 code input is divided into specific areas. The areas defined in Table 2.2 2 as ABWR will be used as input with their specific data listed in Appendix 2A. The areas defined as GENERAL are

/\ also provided in Appendix 2A. The areas defined as Arnendrnent t$ 2.21

ABMi Standard Plant mamo nev. n n

Q Table 2.2 1 DOSE RELATED GOALS Individual Risk <3.9 x 10 (0.1% of normal risk) 4 !

Societal Risk < 1.7 x 10 (0.1% of normal risk)

Probability of 25 Rem Whole Body Dose at ,

4 1/2 mile <10 per year Table 2.2 2 !

CRAC 2 DATA INPUT LISTING CRAC Parameter Groun Defined by Purpose

1. Spatial ABWR Site Radial Mesh

2. Site .ABWR Meteorological selection

3. Economic General Not used but required to run code

4. Population Utility Population description

5. Topography Utility Topography description

6. Isotope ABWR Reactor Core Inventory O 7. Leakage ABWR Release parameters C/ 8.' Dispersion - ABWR Building parameters

9. Evacuation- ABWR Evacuation modeling

10. Acute General Health physics

11. Latent General IIcalth physics

12. Chronic General Health physics

13. File 20 - Not bacd Same data as 4 and 5

14. File 21_ General Health physics j 15. File 27 Utility Meteorology data f

L

('

i i

. IN l' !( ;

Amendment 15 2.22

'ABWR meimo Standard Plant an'. n SECTION 2.3 CONTENTS Sectlon T1112 East 23,1 Envelope of ABWR Standard Plant Site Design Parameters 23 1 23.2 Standard Review Plan Site Characteristics 23 1 233 CRAC 2 Computer Code Calculations 23 1

, .' .y l

L i

! 23 il L'

Amendment 15

mame

- ABWR-Standard Plant nev n

.t <mt 2.3 INTERFACES i ~ ,/

23.1 Envelope of Standard Plant Design Parameters Compliance with the envelope of ABWR Standard Plant site design parameters of Table 2.0-1 shall be demonstrated for design bases events. (See Subseelion 2.2.1) l 23.2 Standard Review Plan Characteristics identification and description of all differences from SRP Section 11 Acceptance Criteria for site characteristics (as augmented by Table 2.1 1) shall be provided. Where such differences exist, the evaluation shall discuss how the alternate site characteristic is acceptable 2JJ CRAC 2 Computer Code Calculations Compliance with acceptance criteria, data input and analysis of Sulisection 2.2.2 for the ;

determination of ABWR iite acceptability for severe accidents shall be demons . rated.

[

I

?

'w) u.)

Amendment 15 2.11

i . i O

APPENDIX 2A-INPUT TO CRAC 2 COMPUTER CODE

k. FOR DETERMINATION OF ABWR SITE ACCEPTABILITY O

1 l

MM 23A6100AD Standard Plant nev.n 7

APPENDIX-2A V TABLE OF CONTENTS TABLES Table 11tle Eage 2A-1 Spatial Subgroup (NUREG/CR-2326, pg 211) 2A 1 2A 2 Site Subgroup (NU~ REG /CR 2326, pg 213) 2A 1 2A 3 Economic Subgroup (NUREG/CR 2326, pg 2 22) 2A 1 2A 4 Population Subgroup (NUREG/CR-2326, pg 2-26 2A 2 2A 5 Topography Subgroup (NUREG/CR 2326, pg 2-33) 2A 2 2A-6 Isotopic Subgroup (NUREG/CR 2326, pg 2 37) 2A 3 l

2A 7 Leakage Subgroup E (- (NUREG/CR 2326, pg 2-41) 2A-4 l

2A 8 Dispersion Subgroup.

(NUREG/CR 2326, pg 2 47) 2A 4 2A 9 Evacuation Subgroup (NUREG/CR 2326, pg 2-47) 2A 4 2A 10 A

' cute Subgroup (NUREG/CR 2326, pg 2 53) 2A 5 b' 2A 11 Latent Subgroup (NUREG/CR 2326, pg 2 57) 2A 5.

2A 12 Chronic Subgroup (NUREG/CR 2326, pg 2-62) 2A 5 2A 13 Scale Subgroup (NUREG/CR 2326, pg 2 72) 2A 10 2Ali Amendment 15

'M 33A6100AD

'. Standard Plant nev n APPENDIX 2A TABLE OF CONTENTS g-TABLES Table 11 tic Eagt 2A 14 Results Subgroup (NUREG/CR 2326,[7,2 74) 2A 10 2A 15 Individual Accident Event Groups ,

1.cakage Subgroup ;

(NUREG/CR-2326, pg 2-41) 2A 11

' 2A 16 . File Dose Conversion File (NUREG/CR 2326, pg 4 2) 2A 16 '

2A 17 File 27 Meteorological Data (NUREG/CR 2326, pg 4-6) 2A 46 l

4 4

b

\:

2A iii L

f w ww w - - mv

MM -

23A6100AD

- mv ^

Standard Plant -

Table 2A 1 l !* / SPATIAL SUBGROUP 2 (NUREG/CR 2326, pg 211 )

b SPATIAL 20 NO

.25 0.75 1.25 1.75 2.25 2.75 3.75 4.25 l

e 5.75 6.25 7.75 8.25 9.75 10.25 13.75 16.25 18.75 21.25 23.75 26.25 1

This reference specifies the location for the following CRAC input parameters, their definitions and formatting instructions.

Table 2A 2 SITE SUBGROUP (NUREG/CR 2326, pg 213)

[

SITE 1 GENTRIC SITE - 50001 29 06 Table 2A 3 l ECONOMIC SUBGROUP

()

3 u

ECONOMIC 54 NO (NUREG/CR 2326, pg 2 22) 499.0 3349.0 - 0.2 31527.0 4344.0 135.0 685.0 MAINE 5 9 0.077 0.182 250 0 485.0-N.H. 5 9 0.097 0.444 150.0 802.0 VT 5 9 0.283 0.791 177.0 657.0 MASS ' 5 9 0.123 0.283 372.0 1366.0

- R.I. .5- 9 0.081 0.220 476.0 2133.0 t

-CONN 5 .9 0.140 - 0313 500.0 2158.0 N.Y. 5 9 0315 0.579 188.0 612.0 NJ. -5 9 0.197 0.162 376.0 2222.0 PA 5 9 0307 0.413 239.0 669.0 -

OHIO 5 9 0.618 0.153 183.0 1516.0 IND 5 9 . 0.728 0.067 206.0 1498.0 ILL 5 9 0.795 0.041 213.0 1786.0 MICH 5 9 0.285 0.238 197.0 955.0 WIS 5 9 0.520 0.598 194.0 807.0 ;

M1NN 5 9 0.563 0.185 160.0 854.0

' 10WA' 5 9 0.944 0.050 242.0- 1458.0 MO- 5- 9 0.724 0.079 111.0 674.0 N.D. 5 9' O.922 0.047 45.0 306.0 S.D. 5 9 0.922 0.074 46.0 257.0 NEBR 5 9 0.%7 0.027 99.0 - 470.0 KANS. 5 9 0.915 0.034 92.0 437.0 p

-d DEL 4 10 0.471 0.046 .508.0 1725.0 Amendment 15 2A.1

, , ...-...a.i_,,-....e.

23A6100AD Standard Plant Rev A Table 2A.3 (Cont'd)

ECONOMIC SUBGROUP (NUREG/CR.2326, pg 2 22)

MD 4 10 0.414 0.227 273.0 1799.0 VA 4 10 0371 0.171 126.0 864.0 W.VA 4 10- 0.270 0.203 44.0 472.0

'N.C. 4 10 0368- 0.056 261.0 819.0 S.C. 4 10 0327 0.063 148.0 635.0 OA 4 10 0.417 0.058 164.0 609.0 FLA 4 10 0368 0.077 233.0 930.0 KY 4 10 0.557 0.117 141.0 792.0 TENN 4 10 0.507 C.140 118.0 669.0 ALA 4 10 0.400 0.041 144.0 515.0 MISS 4 10 0.475 0.047 135.0 520.0 ARK 4 10 0.494 0.030 158.0 - 691.0 LA 4 10 0332 0.087 137.0 763.0 OKLA 4 10 0.782 0.051 68.0 442.0 TEXAS 4 10 0.811 0.053 54.0 354.0 MONTAL 5 9 0.658 - 0.026 -20.0 186.0 1DAHO 5 9 0.894 - 0.114 93.0 485.0 WYOMING S 9 0.560 - 0.024 15.0 119.0 COLORADO 4 10 0.570 0.039 69.0 332.0 N. MEXICO 4 10 0.600 0.056 21.0 100.0-ARIZONA '4- 10 0.556 0.069 36.0 134.0 UTAH 4 10. 0.236 0.215 36.0 265.0 NEVADA 4 10 0.127 0.117 19.0 IN.0 WASH ' 5 9 0369 0.138 132.0 586.0 0REGON 5 9 0300 0.093 68.0 330h CALIF 4 10 0318 0.119 316.0 936.0 NOVA SCO . 5 9- 0.0 0.0 ' O.0 0.0 QUEBEC _ 5 9 0.0 0.0 0.0 0.0 ONTARIO - f. 9 0.0 0.0 0.0 0.0 BAJA CAL 5 9 0.0 0.0 0.0 0.0 SONORA 5 9 0.0 0.0 0.0 0.0 CHillUAHU S 9 0.0 0.0 0.0 0.0 Table 2A 4 POPULATION SUBGROUP (NUREG/CR 2326, pg 2 26)

To be supplied by utility for specified spatial mesh above. -

Table 2A 5 TOPOGRAPHY SUBGROUP (NUREG/CR 2326, pg 2 33)

To be supplied by utility for specified spatial mesh above. -

Anwndment 15 2A.2

23A6100AD Standard Plant an ^ .

Table 2A 6 ISOTOPIC SUBGROUP (NUREG/CR 2326, pg 2 37)

ISOTOPE 54 NO EVALUATED FOR 3926M%T CINDR SOURCE CO 58 7 3.730E + 05 7.130E+01 1.000E-02 1.000E 04 0.5 OF GESSAR CO-60 7 2.247E + 03 1.921E + 03 1.000E-02 1.000E 04 0.5 OF GESSAR KR 85 1 1.184E + 06 3.919E+ 03 0. O.

KR 85M 1 2.644E + 07 1.867E-01 0. O.

KR 87 1 5.070E + 07 5.278E 02 0. O.

KR 88 1 7.185E + 07 1.167E-01 0. O.

RB-86 4 1.844E + 05 1.865E+01 1.000E 02 1.000E-04

. SR-89 6 9.700E+ 07 5.200E+ 01 1.000E-02 1.000E-04 SR 90 6 1.014E + 07 1.026E + 04 1.000E 02 1.000E@

SR 91 6 1.242E + 08 3.950E-01 1.000E-02 1.000E-04

- Y 90 8 SR 90 1.094E+ 07 2.670E + 00 1.000E-02 1.000E-04 Y 91 - 8 SR 91 1.263E + 08 5.880E+01 1.000E-02 1.000E-04 ZR 95 8. 1.735E+ 08 6.550E+01 1.000E-02 1.000E@

ZR 97 8 1.781E + 08 7.000E 01 1.000E-02 1.000E-04 NB 95 8 ZR 95 1,734E+ 08 3.510E + 01 1.000E 02 1.000E-04 MO-99 7 1.966E+ 08 2.751E+ 00 1.000E 02 .1.000E @

TC-99M 7 MO 99 ' 1.6%E + 08 2.508E-01 1.000E-02 1.000E 04 RU 103 7 1.664E + 08 3.959E+01 1.000E-02 1.000E 04 RU 105 7 - 1.174E+ 08 1.850E-01 1.000E 02 1.0COE@

RU 106 7 5.909E+ 07 3.690E+02 1.000E 02 1.000E-04

~ RH 105 7 RU 105 9.907E+ 07 1.479E+ 00 1.000E 02 1.000E-04

'G SB 127 5 8.969E+ 06 3.800E + 00 1.000E-02 1.000E@

SB 129 5 3.172E + 07 1.808E-01 1.000E-02 1.000E@

TE 127 5 SB.127 8.853E+ 06 3.896E 01 1.000E-02 1.000E-04 TE 127M 5 1339E+ 06 1.090E+ 02 1.000E-02 1.000E-N TE-129 5 SB-129 2.983E+ 0714.861E 02 1.000E 02 - 1.000E 04 TE-129M 5 8.090E + 06 3340E+ 01 1.000E-02 1.000E-04 TE 131M 5 .1.464E + 07 1.250E+00 1.000E 02 1.000E-04

' TE 132 5 1.488E + 08 3.250E+00 1.000E 02 1.000E 04 1131 3 TE-131M 1.033E + 08 8.040E + 00 1.000E-02 1.000E 04 1 132 3 TE 132 1.510E+08 ' 9.521E-02 1.000E-02 1.000E-04 1 133 3 2.160E+08 .8.667E-01' 1.000E-02 L 1.000E-04 1 134 : 3 2378E + 08 3.653E-02 .1.000E 02 1.000E-04 1135 3 2.039E+ 08 2.744E 01 1.000E-02 1.000E-04 XE 133 11133 2.170E + 08 5.290E + 00 0. O.

XE 135 11135 3.806E + 07 3.821E-01 0. - 0.

CS 134 4 2.103E+ 07 7.524E+ 02 1.000E-02 1.000E 04 CS 136 4 4.630E + 06 1300E+01 1.000E-02 1.000E-04 w .CS 137 4 1305E+ 07 1.099E+04 1.000E 02 1.000E-04 BA 140 - 6 1.863E + 08 1.279E+01 1.000E 02 1.000E@

LA 140 8 B 'v140 1.974E + 08 1.676E+ 00 1.000E 02 1.000E-N CE 141 8 1.727E + 08 3.253E+01 1.000E-02 1.000E-04 CE-143 8 1.629E+08 1375E+00 1.000E 1.000E-04 CE 144 8 1387E+ 08 2.844E + 02 1.000E 02 1.000E@

PR 14*, 8 CE 143 1.612E + 08 1358E+01 1.000E-02 1.000E-N O Amendment 15 2A 3

ABWR awwan l Standard Plant nev 4 l Table 2A 6 (Cont'd) g' ISOTOPIC SUBGROUP (NUREG/CR 2326, pg 2 37) )

ND 147 8 7.103E + 07 1.099E+ 01 1.000E 02 1.000E 04 NP 239 8 2.401E + 09 2350E + 00 1.000E 02 1.000E-04 PU 238 8 CM 242 6.224E + 05 3.251E + 04 1.000E-02 1.000E-04 PU 239 8 NP 239 5364E + 04 8.912E + 06 1.000E 02 1.000E 04 PU 240 8 CM 244 8.826E + 04 2.469E+06 1.000E 02 1.000E-04 PU 241 8 2.121E + 07 5333E + 03 1.000E-02 1.000E 04 AM 241 8 PU 241 1.726E + 04 1.581E+05 1.000E 02 1.000E44 1 CM 242 8 1.260E+ 07 1.630E + 02 1.000E 02 1.000E 04 CM 244 8 2.885E + 05 6.611E+03 1.000E 02 1.000E-04 Table 2A 7 ;

LEAKAGE SUBGROUP (NUREG/CR 2326, pg 2 41)

(This group input as a dummy in reference deck and overwritten by individual cases.)

(See INDIVIDUAL CASES)

LEAKAGE 1 - NO DUMMY INPUT OVERLAYED IN ACTUAL RUN SEE TABLE 2A 15 ABWR CS1 1.0 31.9 2.78 1.5 4.0E + 07 10.

1.0E + 00 0.8 0.8 1.0E-03 1.1E-03 2.6E-04 1.5E 07 .

Table 2A 8 DISPERSION SUBGROUP (NUREG/CR 2326, pg 2-45)

DISPERSION 54.0 37.7 4 0 0 Table 2A 9 EVACUATION SUBGROUP (NUREG/CR 2326, pg 2 47)

EVACUATE 1 NO YES 1.0 0. C. O. 24135. O. 2. 1.0 1.0 1. 1.0 1.0 1.0 1.0 1.0 1.0 2.66E-4 2.66E-4 133E-4 2.66E-4 8045. 90. 95. 3. 0 Amendment 15 2A4 0

M\ -

23AM00AD Standard Plant wA Table 2A 10 7) i u ACUTE SUHGROUP (NUREG/CR ' 2326, pg 2 53)

ACUTE 7

- T MARROW 320. 400. 510. 615. .03 .5 L LLI WALL 2000. 5000, 5000, 5000. 1. 1, 1.

LUNG 5000. 14800, 22400. 24000. .24 .73 1.

W BODY 55. 150. 280. 370, 30 0 0.

LUNG 3000, 3000.1 6000. 6000. .05 1.0 0.

LLI WALL 1000. 1000.1 2500. 2500. .05 1.0 0.

THYROID 1.E10 - 1.E10 1.E10 - !.E10 1.01 1.0 0.0 i Table 2A 11 LATENT SUBGROUP (NUREG/CR 2326, pg 2 57)

LATENT 8 ,

10 CENT EST 30. 5. 300. 2.5 TMARROW LEUKEMIA 2.836E-05 2.720E-051.872E-05 1382E-05 9.720E-06 6.770E 06 4.040E-06 1.700E 06 4.900E 07 0.0 1.0 LUNG LUNG' 2.749E-05 2.749E-05 2.749E 05 1.587E-05 8.130E 06 3.990E 06 1.500E 06 - 2.200E-07 0.0 0.0 0.5

' /7 OTHER BREAST 3.172E-05 3.172E-05 3.172E-05 i.831E-05 9380E 06 4.600E-06 l V 1.730E-06 2.500E 07 0.0 0.0 1.000E + 09 1.107E-05 1.064E 05 6.990E-06 3.020E-06 1.670E-06 9.100E 07 SKELETON - BONE .

4.200E 1.200E-07 1.000E-08 0.0 1.0 LLI WALL G1TRK 1.688E 05 1.688E-051.688E-05 9.740E-06 4.990E-06 2.450E 06 '

9.200E 07 1300E 07 0.0 0.0 1.0 OTHER OTHER 4.235E-05 3.557E 05 2.539E-05 1.466E 05 7.520E-06 3.690E 06 1390E 06 2.000E-07 0.0 0.0 1.0 '

W BODY WBODY 1.579E 04 1.533E-041.274E-04 7.542E-05 4.141E 05 2.241E 05 L000E 05 2.620E 06 5.000E-07 0.0 1.0 THYROID THYROID 334E45 1.00E 09 Table 2A 12 CHRONIC SUBGROUP -

(NUREG/CR 2326, pg 2 62)

CHRONIC EXPOSURE 6 10 1' 1.000 - -365. 25550. 3.0 15.0 SR 90 0.0525 0.0718 RU 106 0.0397 0.0533 CS-137 0.0525 0.105 f PU 238 0.0529 0.107 k

Amendment 15 2A-5 l

, ABWR 2346im40

- Standard Plant Rev A Table 2A.12 (Cont'd)

CHRONIC SUBGROUP (NUREG/CR 2326, pg 2 62)

PU 239 0.0530 0.108

' PU 240 0.0530 0.108 PU 241 0.0520 0.101 AM 241 0.0530 - 0.108 CM 242 0.0292 0.0327

~ CM 244 0.0522 0.102 3 11 1.0 365. 365. 14.0 2.0 33 CS 134 8.44 4.22 LUNO 6.47E + 4 731E + 4

^;: T MARROW 6.50E+4 734E+4 SKELETON 6.41E + 4 7.24E + 4 TECL 6.41E + 4 7.24E + 4 ST WALL 7.40E + 4 834E + 4 SI+ CONT 8.05E + 4 9.09E+ 4 ULI WALL 7.95E+ 4 8.%E + 4 LLI WALL - 8.28E + 4 933E + 4 THYROID 6.49E + 4 733E + 4 ;

OTHER 6.27E + 7.08E+ 4 W BODY 632E + 4 7.14E + 4 TESTES 7.57E+ 4 8.55E + 4 '

- OVARIES 6.68E + 4. 7.55E + 4 '

CS-136 -- 2.84 1.42 LUNG 8.82E+ 3 8.82E + 3 T MARROW 9.29E+3 9.29E+3 SKELETON 9.10E + 3 9.10E + 3 -

TECL 9.10E + 3 9.10E + 3 t

- ST WALL - - 1.15E + 4 1.15E + 4 SI+ CONT ' 1.19E + 4 1.19E + 4

' ULI WALL 1.20E + 4 1.20E + 4 LLI WALL 135E+ 4 135E+ 4 THYROID 9.23E + 3 9.23E+ 3 OTHER' . 8.88E + 3 8.88E +3 W BODY 8.%E + 3 8.%E + 3 ,

TESTES. 1.03E+ 4 1.03E+ 4 OVARIES 9.48E+ 3 9.48E + 3 i CS 137 8.44 4.22 1 LUNO 4.71E + 4 5.59E + 4 T MARROW 4.73E+4 5.61E+4 SKELETON ' 4.68E+4 5.56E+4 TECL 4.68E+ 4 5.56E + 4 ST WALL ' 5.18E + 4 6.13E + 4 St + CONT 539E+4 639E+4 ULI WALL 5.40E + 4 639E + 4

- LLI WALL 5.64E+ 4 6.64E + 4 i

THYRO!D 4.68E + 4 5.55E+ 4 OTHER- 4.60E + 4 5.45E+ 4 Amendment 15 2A.6

23A6100AD Standard Plaint nn. 4 n Table 2A 12 (Cont'd) l b CHROhlCSUBGROUP (NUREG/CR 2326, pg 2 62) )

1 W BODY 4.62E + 4 5.49E + 4 TESTES 5.18E + 4 6.15E + 4 OVARIES 4.81E+ 4 5.70E+ 4 1 2 2 1.0 365. 365. 14.0 2.0 33 SR.89 .397 0.402 LUNG 2.91E + 3 5.f1E+ 2 '

T MARROW 2.63E+ 4 5.26E+ 3 SKELETON 5.95E + 4 1.19E + 4 TECL 6.00E + 4 1.20E + 4 ST WALL 1.56E + 4 3.12E + 3 Si + CONT 2.73E+ 4 5.45E + 3 UL1 WALL 1.46E+ 5 2.91E+ 4 LLL WALL 4.27E + 5 8.53E + 4 THYROID 2.91E+ 3 5.81E+ 2 OTHER 2.91E+ 3 5.SIE + 2 W BODY 9.55E + 3 1.91E + 3 TESTES 2.91E + 3 5.81E + 2 OVARIES 2.91E+ 3 5.81E + 2 SR 90 .505 .588 LUNG 1.59E + 4 3.18E+ 3 5.50E + 2 1.80E+ 1

( T MARROW 1.04E+ 6 2.08E+ 5 5.25E+ 4 1.29E + 4 1.00E + 4 3.10E + 3 y SKELETON 3.08E + 6 6.15E + 5 2.57E + 5 9.81E + 4 1.09E + 5 4.30E + 4 TECL 2.64E + 6 5.27E + 5 1.93E + 5 6.77E+ 4 7.20E + 4 2.76E + 4 ,

ST WALL 2.03E + 4 4.05E+ 3 5.50E+ 1 1.80E + 1 Si + CONT 2.64E 4 4 5.28E + 3 5.50E + 1 1.80E + 1 ULI WALL 1.06E+ 5 2.11E+ 4 5.00E+ 1 2.00E + 1 LLL WALL 4.06E + 5 8.12E + 4 5.00E + 1 THYROID 1.59E + 4 3.18E + 3 5.50E + 1 1.80E + 1 OTHER 1.59E + 4 3.18E + 3 5.50F + 1 130E+1 WBODY 2.76E+ 5 5.52E+4 2.03E+ 4 7.44E + 3 8.08E + 3 3.13E + 3 TESTES 1.59E + 4 3.18E + 3 5.50E + 1 1J10E+ 1 OVARIES 1.59E+ 4 3.18E+ 3 5.50E+ 1 1.80E+ 1 2 9 1.0 0.0 365. 14.0 0.0 10.0 1133 1.03E 8 0.00486 LUNG 8.53E + 2 1.58E + 2 T MARROW 7.99E+ 2 1.48E+ 2 SKELETON 7.8SE + 2 1.46E + 2 TECL 7.88E + 2 1.46E + 2 ST WALL 1.11E+ 4 2.06E+ 3 Sl + CONT 2J0E+ 3 4.25E + 2 ULIWALL 6.16E + 3 1.14E + 3 LLI WALL 9.83E + ? 1.82E + 3 THYROID 1.73E+ 6 3.21E+ 5 OTHER 9.07E 4 2 IIeE + 2 W BODY 1.46E + 3 2.70E + 2 TESTES 7.13E + 2 132E + 2

)

b Amendment 15 2A.7

l ABWR urumro Standard Plant Rev A [

I s

Table 2A.12 (Cont'd)

CHRONIC SUBGROUP '

l (NUREG/CR 2326, pg 2 62)

OVARIES 9.99E + 2 1.85E + 2 1131 1.00E 8 0.692 LiJNO 1.92E + 3 3.56E + 2 l

T MARROW 1.55E+ 3 2.87E+2 l SKELETON 1.67E + 3 3.10E + 2 TECL 1.67E + 3 3.10E + 2 ST WALL 6.16E + 3 1.14E + 3 SI+ CONT 1.79E+ 3 332E + 2 ULI WALL 4.49E + 3 832E + 2 LLI WALL 1.03E + 4 1.91E + 3 THYROID - 9.07E + 6 1.68E + 6 OTHER 2.20E + 3 4.07E + 2 W BODY 4.75E + 3 8.79E + 2 TESTES 7.51E+ 3 139E + 2 OVARIES 1.15E + 3 2.21E + 2 4 2 1.000 3650, 3650, 2400. 5.0 f.0 CS 134 .164 .0547 LUNO 731E+4 T MARROW 734E+ 4 SKELETON 7.24E + 4 TECL 7.24E + 4 ST WALL 834E+4 SI+ CONT 9.09E + 4 ULI WALL 8.96E + 4 LLI WALL 933E+4 THYROID 733E+4 OTHER 7.08E + 4 W BODY 7.14E + 4 TESTES 8.55E + 4 OVARIES 7.55E + 4 CS.137 .250 .0835 LUNO 5.59E + 4 T MARROW 5.61E+ 4 SKELETON 5.55E + 4 TECL $.56E + 4 ST WALL - 6.13E + 4 l

Sl+ CONT 639E + 4 ULI WALL 639E+4 LLI WALL 6.64E + 4

- THYROID 5.55E + 4

'OTHER 5.45E+4 W BODY 5.49E + 4 TESTES 6.15E + 4 OVARIES 5.70E + 4 ;

SR-89 .0136 .0068 i I

LUNG 5.81E+2 Amendment 15 2A 8 0

l L __ __ _ . . .

_ _ _ . _ _.j

4 ABWR 2.ummo Standard Ptsnt %A f] Table 2A.12 (Cont'd) v CHRONIC SUBGROUP (NUREG/CR.2326, pg 2 62)

T MARROW 5.26E+3 SKELETON 1.19E + 4 TECL 1.20E + 4 ST WALL 3.12E + 3 51+ CONT 145E+ 3 ULI WALL 2.91E + 4 LLL WALL 8.53E + 4 THYROID 5.81E + 2 OTHER 5.81E+2 W BODY 1.91E + 3 TESTES 5.81E4 2 OVARIES 5.81E + 2 SR 90 1.340 0.669 LUNO 3.18E + 3 5.50E + 2 1.80E + 1 T MARROW 2.08E+ 5 5.25E+ 4 1.29E+ 4 1.00E + 4 3.10E + 3 3.103 + 3 SKELETON 6.15E + 5 2.57E + 5 9.81E + 4 1.09E + 5 430E+4 4.30E + 4 TECL 5.27E + 5 1.93E + 5 6.77E + 4 7.20E + 4 2.76E + 4 2.76E + 4 ST WALL 4.0$E + 3 5.50E + 1 1.80E + 1 SI + CONT 5.28E + 3 5.50E + 1 1.80E+ 1 UL1 WALL 2.11E + 4 5.00E + 1 2.00E + 1

, LLL WALL 8.12E 4 4 5.00E + 1 THYRO!D 3.18E + 3 5.50E + 1 1.80E + 1 OTHER 3.18E + 3 5.50E + 1 130E + 1 W BODY 5.52E+ 4 2.03E + 4 7.44E+ 3 8.08E + 3 3.13E+ 3 3.13E+ 3 TESTES 3.18E + 3 5.50E + 1 1.80E + 1 OVARIES 3.18E + 3 5,50E + 1 1.80E + 1 10 11 0333 365. 10950. 3285. 5.0 25.0 CO 58 CO 60 NB 95 ZR 95 RU 103 RU 106 1131 CS 134 CS 136 CS 137 Amendment 15 2A 9 1

ABWR mwon Standard Plant n,< x l

Table 2A 13 l

SCALE SUBGROUP (NUREG/CR.2326, pg 2 72) i l

SCALE 36 NO j

1. E02. E03. E05. E07. E01. E12. E13. El

5. E17. E11. E22. E23. E25. E27. E21. E3

2. E33. E35. E37. E31. E42. E43. E45. E4

7. E41. E52. E53. E55. E57. E51. E62- E6

3. E65. E67. E61. E7 Table 2A 14 RESULTS SUHGROUP (NUREG/CR 2326, pg 2 74) l RESULT 28 NO 1

ACUTE FATALITIES ACUTE INJURIES F0F W/BMR DS >200 RSK OF FAT INT 2 1.0E-06 RSK OF FAT INT 4 1.0E-06 &

RSK OF FAT INT 10 1.0E 06 W RSK OF FAT IKT14 1.0E-06 FATAL RADIUS (MI)

RSK OFINJ INT 2 1.0E 06 RSK OFINJ INT 14 1.0E-06 RSK OFINJ INT 18 1.0E 06 INJUR RADIUS (MI)

ACU BMR DS INT 2 1.0E 02 ACU BMR DS INT 10 1.0E-02 ACU BMR DS INT 14 1.0E-02 ACU BMR DS INT 18 1.0E-02 ACU TIIY DS !NT 2 1.0E 02 ACU TIIY DS INT 10 1.0E 02 ACU TIIY DS INT 14 1.0E-02 ACU TIIY DS INT 18 1.0E 02 TOT LAT/ INITIAL TOT LAT/ TOTAL CANCER RSK lbrr 2 1.0E 06 CANCER RSK INT 14 1.0E 06 CANCER RSK INT 18 1.0E-06 DECON AREA DECON DIST TOT WBODY MANREM 1.0E 02 O

Amendment 15 2A 10

ABM momio Standard Plant n n.,

Table 2A.15 l

INDIVIDUAL ACCIDENT EVENT GROUPS LEAKAGE SUBGROUP (NUREG/CR.2326, pg 2-41) 1 l

l GE PROPRIETARY . PROVIDED UNDER SEPARATE COVER O .

I l

l l-O Amendment 15 2A Il

ABM 2W1ho Standard Plant Rev.A Table 2A 15 (Cont'd)

I INDIVIDUAL ACCIDENT EVENT CROUPS LEAKAGE SUBGROUP (NUREG/CR 2326, pg 2 41) i i

d I GE PROPRIETARY . PROVIDED UNDER SEPARATE COVER O

I l

Amendment 15 2A 12 9

f

__ _- . . , _ . .,_ ._ __ . .. ._ _,y, . . . . _ . . . , ,,

.ABWR m aman Standard Plant Rev.A

Table 2A 15 (Cont'd)

INDIVIDUAL ACCIDENT EVENT GROUPS LEAKAGE SUBGROUP (NUREG/CR.2326, pg 2 41)

GE PROPRIETARY PROVIDED UNDER SEPARATE COVER i

i

.4 l

l

{-

l.

i O 2A 13 Amendment 15 l

y .. _ _ _ . . . __ _ . _ _ . . . . _

ABWR mamn Standard Plant Rev.A Table 2A 15 (Cont'd)

INDIVIDUAL ACCIDENT EVENT GROUPS '

LEAKAGE SUBGROUP (NUREG/CR.2326, pg 2 41)

GE PROPRIETARY . PROVIDED UNDER SEPARATE COVER O

I Amenom.m ts h

' ~ ^ . . . - . . . . . .

..n.,_ , , , .

l MN 23AWOAD Standard Plant n,, 3 Table 2A 15 (Cont'd)

INDIVIDUAL ACCIDENT EVENT GROUPS LEAKAGE SUBGROUP (NUREG/CR.2326, pg 2-41)

I 1

GE PROPRIETARY . PROVIDED UNDER SEPARATE COVER l

! s t

l O Amendment 15 2A 15

r ABWR ur6tman Standard Plant Rev A Table 2A.16 FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326, pg 4 2) co.58 co-60 kt 85 kt 85m kt 87 kt-88 tb 86 sr 89 . 90 sr 91 y 90 y 91 rr 95 zr 97 ntw95 mo 99 te 99m ru 103 105 ru 106

, th 105 te 127 te 127m te 129 te 129m te 131m te.132 stw127 sb 129 l131 t I

l132 1133 1134 l135 xe 133 xe135 cs 134 cs 136 cs 137 ba 140 la.140 cc 141 cc 143 cc 144 pr 143 nd 147 op 239 pu 238 pu 239 pu.240 pu 241 am 241 cm 242 cm.244 lung

.5900e + 05 .5900e + 05 .2000e + 040. O. O. O.

.1030c + 03 .2100e+ N .1130c + 06 .2010e + 00 4600e + 06 .4600e + 06 .7400c + 060 . .1000e + Ofo. O.

.2660c + 03 .5580c + 04 .2920e + 06 .5670c + 00

.1800e + 00 .1800e + 000 . O. O. O. O.

.2300c + 00 4820e + 01 .2510e + 03 ,4470c-03

.2100e + 00 .2100e + 000 . O. O. O. O.

.9220e + 01 .1300c + O2 .1760c + 05 3220e 01

.9600e + 00 .9600e + 000 . O. O. O. O.

.1730e + O2 .1750c + O2 .8400c + 05 .1720e + 00

.2000e + 01 .2000e + 010 . O. O. O. O.

.1110e + 03 .1300e + 03 .2090c + 06 .4470c + 00

.1400e + 05 .1400e + 050 . O. O. O. O.

.9270c + 01 .1730c + 03 ,1020c + 05 .1940c 01

.7800e + 04 .7800c + 040 . O. O. O. O.

O. O. O. O.

.1600c + 05 .1600e + 05 .2000e + NO , O. O. O.

O. O. O. O.

.4300s + 04 .4300e+ NO 0. O. O. O.

.8250c+02 .1930c + 03 .8580e + 05 .1600e + 00 3300c + 05 3300c + 050 0. O. O. O.

O. O. O. O.

.2000c + 06 .2000e + 060 0. O. O. O.

.2810e + 00 .5660c + 01 3080c + 03 .5940e 03

.1300c + 06 .1300e + 060 0. O. O. O.

l .7720e + 02 .1670c + 04 .8440c + 05 .1520e + 00

.1500c*05 .1500c + 050 0, 0, 0. O.

.1310c + 03 .5100c + 03 .2120e + 05 4000c 01 3100c + 05 3100c + 050 0. O. O. O.

l .7820e + 02 ,1540c + 04 .8590c + 05 .1.%0e + 00

.1600c + 05 .1600c + 050 0, 0, 0. O.

.2180c + O2 .2910e + 03 .2080e + 05 3420c 01

.8900e + 02 .8900e + 020 0. O. O. O.

.8120e + 01 .1350c + 02 .1360e + 05 .2540c-01 L .5400e + 05 .5400e + 050 0. O. O. O.

.5520e + 02 .1090e + 04 .6050e + 05 .1050c + 00

.2200c + 04 .220J' e + NO 0. O. O. O.

.4850c + 02 .T.,70c + O2 .9200e + 05 .1670c + 00

.2500e + 07 /4500c + 07 .1400c + 070 0. O. O.

.2060e + 02 .4300e+ 03 .2250e+05 .40(Oc01 1

1 l

Amendment 1, 2A.16 l

I ABWR uAmorn Rev. A Standard Plant T Table 2A.16 (Cont'd)

G(^ FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2) 360tk + 04 3600e + 040 0. O. O. O.

l

.7530c + 01.5010e + O2 .8940c + 04 .1610c-01

.1600e + 04 .1600e + 040 0. O. O. O.

3430c + 00 .7670c + 00 .4980c + 03 .8780c-03

.1200e + 06 .1200e + 060 0. O. O. O.

.1340e + 013370c + 02 .1350c + 04 .5610c 03

.5600c + 03 .5600c + 030 0. O. O. O.

.1810e + 01.1830c + 01 .9550e + 04 .1350c 01

.1500e + 06 .1500e + 060 0. O. O. O.

.1020e + 02 .2220e + 03 .6480c + 04 .6970c-02

.1100c+ 05 .1100e + 050 0. O. O. O.

.1390c + 03 .oM + 03 .1590e + 06 .2940c + 00

.3000e + 05 3000c + OM 0. O. O. O.

i

.1690c + 03.2880c 4 04 .2230c + 05 .4190c-01

.2500c + 05 .2500e + 0$0 0. O. O. O. l

.7050c + O2 .8650c + 03 .7950c + 05 .1430e + 00 3200e + 043200e + 040 0. O. O. O.

.6990c + O2 .9780c + O2 .1360c + 06 .2530c + 00

.2400e + 04 .2400e + 040 0. O. O. O.

.4150c + O2 .6630c + 03 .4610e + 05 .8220c-01

.1000e + 04 .1000e + 040 0. O. O. O.

g-

' .9230c + O2 .1010e + 03 .2700c + 06 4830c + 00 3100e + 04 3100e + 040 0. O. O. O.

.6500c + 02 .2910e + 03 .8090c + 05 .1460c + 00

.5Wie + 03 .5600c + 030 0, 0, 0. O.

3870c + 02 3880c + 02 .2690c + 06 .5000c + 00

.2500c + 04 .2500c + 040 0. O. O. O.

.1400e + 03 .2690e 4 03 .2040c + 06 .4000c + 00

.4100c + 00 .4100c + 000 0. O. O. O.

.5750c + 01.8080c + 02 .6440c + 04 .6970c 02

.9400e + 00 .9400e + 000 0. O. O. O.

.1890c + 02 .4160e + 02 .2750c + 05 .5060c.01

.4500c + 05 .4500e + 05 .6000c + 040 0. O. O.

.1660c + 03 3470c 4 04 .1810e + 06 3280e + 00

.8200c + 04 .8200c + 040 0. O. O. O.

.2160c + 03 3820e + 04 .2380c + 06 .4440e + 00 3400e + 05 3400c + 05 .6000e + 040 0. O. O.

.5840e + O2 .1240c + 04 .6450c + 05 .1150e + 00

.6300c + 04 .6300c + 040 0. O. O. O.

.3970c + O2 .3460c + 04 .2580c+05 .4140c 01

.1600e + 05 .1600e + 050 0. O. O. O.

.2330c + 03 .1710e + 04 .2730c + 06 .5390c + 00

.6200c + 05 .6200e + 050 0. O. O. O.

.7640c + 01.1500e + 03 .8400e + 04 .1500c 01

.1300c + 05 .1300c + 050 0. O, 0. O.

3140c + 02.2000e + 03 3720e + 05 .6080c 01

.2100e + 07 .2100e + 07 .8000c + 060 0, O. O.

2A 17 f Amendment 15 l

l

.,,.e..

MM 33A61ooAn Standard Plant %A Table 2A.16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326 pg 4 2)

.4980c + 01 .1070e + 03 .2250c + 04 3440c-02

.4900e + 05 .4900e + 050 0. O. O. O.

O. O. O. O.

3700e+05 3700c+05 .1000c+040 0. O. O.

.1570c + 02 .2700c + 03 .1740c + 05 .2780e 01

.9200e + 04 .9200c + 04 .1000e + 030 0. O. O.

.1870e + O2 .1740c + 03 .2150e + 05 .2650c-01

.1200c + 09 .1200e + 09 .1900e + 090 0. O. O.

.5410e-01 .1140c + 01 .5920c + 02 .9580c 05

.1200e + 09 .1200e + 09 .1700e + 090 0. O. O.

.2960e.01 .6220c+ 00 3240c+02 .5420e.05

.1200e + 09 .1200c + 09 .1700e + 09 .1000e + 080 0. O.

.5150c 01 .1080c + 01 .5640e+02 .9170c-05

.6400e + 05 .6400e + 05 4660c + 06 3000c + 05 .1000e + 05 .1000c + 05 .1000e + 05 3530e-05 .1560e-02 .7740c 02 .2940c 09

.1300e + 09 .1300e + 09 .1800e + 090 0. O. .1000c + 08 i

.4840c + 01 .1020e + 03 .5300e + 04 3220c-02

.7600c + 08 .7600c + 08 .1100c + 080 0. O. 0-A370c-01 .9050e + 00 .4790c + 02 .8310c.05

.1300e + 09 .1300e + 09 .1800e + 090 0. O. O.

.1240c + 01 .2610e + 02 .1360c + 04 .1070e 02 t marrow

.7950c + 03 .3000e + 04 .1000e + 030 0. O. O.

.1230c + 03 .2500e + 04 .1350c + 06 .2400c + 00

.2000c + 04 .2100c + 05 .3500c + 05 .1000e + 04 .1000c + 040 0,

.2960c + 03 .6220e + 04 3250e + 06 .6310e + 00

.6100e + 00 .6100e + 000 0, 0, 0. O.

.2970c + 00 .6220e + 01 3250e + 03 .5780c-03 j 3900e+00 3900c+000 0. O. O. O, i

.1570c + 02 .2210e + 02 3000c+05 .5500c 01 l

.1300e + 01 .1300e + 010 0. O. O. O.

.1930e + 02 .1960e + O2 ,9370e+05 .1920e+ 00 3100c +01 3100c + 010 0. O. O. O.

.1190c + 03 .1390c + 03 .2250c + 06 .4830e + 00 '

3250e + 04 .6500e + 040 0. O. O. O.

.1090c + 02 .2020c + 03 .1200e + 05 .2270c 01 3350c + 04 .1300e + 050 0, 0. O. O.

O. O. O. O.

.6100c + 04 .1100e + 06 .4200e + 06 .1300e + )6 3000e+ 05 3000e + 05 .1000e + 05

0. O. O, 0.

.2150e + 033100e + 03 .1000e + 020 0. O. O.

.1020e + 03 .2390c + 03 .1030c + 06 .193Ge + 00

.4700c + 03 .5100c + 030 0. J. O. O.

O. O. O. O.

.1430e + 04 .9200e + 04 .1000e + 030 0. O. O.

3010e + 00 .6060c + 01 3300e+03 .6390c-03

.6700e+ 03 3500c + 04 .1000c + 030 0. O. O.

Amendment 15 2A 1B

</

ABWR 2.wimso Standard Plant n,v 4 (q') Tabk 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.9460c + 02 .2040e + 04 .1030e + 06 .1870c + 00

.1900c + 03 .1900c + 030 0. O. O. O.

.1670c + 03 .6520e + 03 .2510e + 05 .4720c-01

.57W + 03 .1400e + 040 0. O. O. O.

.9170c + 02 .1800e + 04 .1010e + 06 .1830e + 00

.1250c + 03 .1300c + 030 0. O. O. O.

.3130c + 02 .4650c + 03 .2700e+05 .4440c 01

.1100c + 02 .1100c + 020 0. O. O. O.

.1740c + 02 .2880c + 02 .2910c + 05 .5420c-01 4050c + 03 .1100c + 040 0. O. O. O.

.7160e + 02 .1420e + 04 .7830c + 05 .1360e + 00

.2400c + 02 .2400e + 020 0. O. O. O.

6440c + O2 .9990c + 02 .1220e + 06 .2210e + 00

.4400e + 03 3600e + 04 .2600c + 040 0. O. O.

.2650c + O2 .5540c + 03 .2910e + 05 .5220e 01

.2300e + 02 .2300e + 020 0. O. O. O.

.1280c + 02 .8550c + 02 .1520e+05 .2740c 01 3900e + 01 3900c + 010 0. O. O. O.

.4540c4 00 .1010e+01 .6590c + 03 .1160e.02

.1820e + 03 .7500e + 03 .5000e + 020 0. O. O.

.4080c + 01 .9200c + 02 .4310c + 04 ,1790c-02 I .1100e+01 1100e+010 0. O. O. O.

i .2420e + 01 .2440c + 01 .1270c+05 .1810e-01 3750c + 03 .8300e + 03 .1000c + 020 0. O. O. ,

.1420e + 02 3070c + 03 .9200e + 04 .9920c-02 3000c + 03 3100e + 030 0. O. O. O.

.1710e + 03 .1100e + 04 .1920e + 06 3560e + 00

.9400e + 03 .1000e + 040 0, 0. O. O.

.2170c + 03 3630c + 04 3880e+05 .7310c-01 3100c + 033300e + 030 0. O. O. O.

l .9070e+02 .1110e+04 .1020e + 06 .1840e + 00 l .4600c + 02 ,4600e + 020 0. O. O. O.

.8250c + 02 .1160e + 03 .1600e + 06 .2970e + 00

.1500e + 03 .1900e + 030 0. O. O. O.

r

.5450c + O2 .8730c + 03 .6060c + 05 .1080c +(A)

.5000e + 02 .5000e + 020 0. O. O. O.

.1130c + 03 .1230e + 03 3290c + 06 .5890c + 00

.9350e + 02 .9400e + 020 0. O. O. O.

.8130c + 02 3750e + 03 .1010e + 06 .1830c + 00

.2000e + 02 .2000e + 020 0. O. O. O.

.4550e + 02 .4560c + 02 3160c +06 .5890e + 00

.9100e + 02 .9100e + 020 0. O. O. O.

.1600c+03 3180c+03 .2260c+ 06 4420c + 00

0. O. O. O,

.1600c + 01 .1600e + 010

.1310e + 02 .1840e + 03 .1470e+05 .1590c-01

.2100e + 01 .2100e + 010 0. O. O. O.

p 3160c + 02 .6980c + O2 ,4620e + 05 .8470c 01 NJ 2A 19 Amendment 15

23A61COAD Standard Plant nev A Table 2A 16 (Cont'd)

, FILE 20 DOSE CONVERSION FILE g

(NUREG/CR.2326 pg 4 2)

.4950e + 04 .4300e + 05 .5000e + 040 0. O. O.

.2030e + 03 .4260e + 04 .2230e + 06 .4030e + 00 3550c + 04 .6000e + 040 0. O. O. O.

.2640c + 03 .4680c + 04 .2930c + 06 .5420e + 00 3250c + 04 .3100e+ 05 .6000e +040 0. O. O.

.7560e + 02 .1600e + 04 .8340c + 05 .1490c 4 00

.2100e+04 3400c+040 0. O. O. O.

l .5000e + O2 3980c + 04 3500e+05 .5610e-01 l .6700e + 03 .6800e + 030 0. O. O. O.

.2610e + 03 .1920e + 04 3073c + 06 .6060e + 00

.1130e + 03 .2700e + 030 0. O. O. O. ,

' .1650e + 02 3240e + 03 .1810c + 05 .3220e 01

.9550e + 02 .1100e + 030 0. O. O. O.

4830c + 02 .3080c + 03 .5740c+05 .9360c 01

.2350c + 03 3600e + 04 .5600e + 040 0. O. O.

.7840c + 01 .1670c + 03 .4980c v 04 .7610c-02

.1780c + 02 3400c + 020 0. O. O. O.

O. O. O. O.

.1400e + 03 .1900c + 03 .1000e + 020 0. O. O.

.2480e + 02 .4260e + 03 .2740e+05 .4390e-01

.6200e + 02 .6400e + 020 0. O. O. O.

3490e+02 3260c+03 .4020e+05 .4970c 01

.1710e + 03 .6000e + 04 .2240c + 06 3400e + 06 3000e + 06 .2300e + 06 .2000e + 06

.2390c + 00 .5020e + 01 .2620e+03 ,4250e-04

.1590c + 03 .5600e + 04 .2240e + 06 3500e + 06 3400e + 06 .2800e + 06 3000e + 06

.1190c + 00 .2490c + 01 .1300e+03 .2170e-04

.1640c + PJ .5600e + 04 .2240e + 06 3500e +063500e + 06 .2700c + 06 3000e + 06

! .2180cs 00 .4570e+01 .2380e+03 3890c-04

.4200c-01 .6100e + 01 .1790c + 04 .5700e + 04 .7500e + 04 .8000c + 04 .9000c + 04

.%30e 04 .4520c-02 .2240c-01 .8530c.09

.2650c + 03 .7M0c + 04 .2430e + Go .3800e + 06 3600e + 06 3100e + 06 3000e + 06

.1400c+02 .2950e+03 .1530c+05 .9330c-02

.2030c + 03 3200e + 04 .2900e + 04 .1200e + 04 .1200e + 04 .1000e + 04 .5000e + 03

.2050e + 00 .4250c + 01 .2240c + 03 3890c-M

.2010e + 03 .6600c + 04 .2030e + 06 .2200c + 06 .1500e + 06 .1000e + 06 fiX)0e + 05 3240e+01 te10c+02 3560e + 04 .2810c-02 skeleton 3300e + 03 .2500e + 04 .1000e + 030 0. O. O.

.1230c + 03 .2500e + N .1350e + 06 .2410e + 00

.8000e + 03 .1800e + 05 3000e + 05 .2000e + 040 0. O.

3010e 4 03 .6270c + 04 3290e + 06 .6390e + 00

.1500e + 00 .1500e + 000 0. O. O. O.

.2960c + 00 .6240e + 01 3250e + 03 .5780e 03

.1900c + 00 .1900e + 000 0. O. O. O.

.1580c + 02 .2220e + 02 3000c +05 .5500c-01

.8300e + 00 .8300c + 000 0. O. O. O.

.1980c + O2 .2000e + 02 .9640e + 05 .1970e + 00 Amendment 15 2A.20 O

i

MM 23A6100AD Standard Plant Rn A Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.1800c + 01 .1800e + 010 0. O. O. O.

.1230c + 03 .1440c 4 03 .2310e + 06 .4940c + 00

.1700c + 04 .6500e + 040 0. O. O. O.

.1080c + O2 .2020e + 03 .1190c + 05 .2270c-01

.3000c + 04 3000c + 050 0. O. O. O.

O. O. O. O.

3800c + 04 .2600e + 06 .1340c + 07 .6000c + 06 .2000c + 06 3000c + 06 .1000e + 06

0. O. O. O.

.2000c + 03 3400c + 030 0. O. O. O.

.1020e + 03 .2390c + 03 .1040c + 06 .1940c + 00

.8600c + 03 .1000c + 040 0. O. O. O.

O. O. O. O.

.1200e + 04 .1900c + 050 0. O. O. O.

3080e + 00 .6240c + 01 3380c + 03 .6530c 03 ,

.2800c + 03 .3300e 4 04 .1000c + 030 0. O. O.

.9440c + 02 .2030c + 04 .1040c + 06 .1880e + 00

.1300c + 03 .1300e + 030 0 0. O. O.

.1680e+ 03 .6(10e+03 .2530e+05 .4750c 01

.2600e + 03 .1200c + 040 0. O. O. O.

.9150c + O2 .1800e + 04 .1010e + 06 .1830e + 00 i .1000c + 03 .1100e + 030 0. O. O. O.

! O 3130c + 02 .4650e + 03 .2690c 605 .4440e 01 b .1000c + 02 .1000e + 020 0. O. O. O.

.1740c + O2 .2880c + 02 .2910e + (A .5420e-01

.1800c + 03 .8800c + 030 0. O. O. O.

.7160c + 02 .1420e + 04 .7810e + 05 .tyoc + 00

.1700c + O2 .1700e + 020 0. 6. O. O.

.6450e + O2 .9960c + 02 .1220e + 06 .2220e + 00

.1900c + 03 3400c+ 04 .2500e + 040 0. O. O.

.2660c + 02 .5550c + 03 .2900c + 05 .5220c-01

.1600c + 02 .1600e + 020 0. O. O. O.

.1280e + 02 .8500e + 02 .1520e + 05 .2740e.01

.5200c + 01 .5200c + 010 0. O. O. O.

i .4530e + 00 .1010e + 01 .6600e+03 .1160e-02

.1300c + 03 '000c + 04 .NX)0c + 030

_ 0. O. O.

.3810e + 0'. 8680c + 02 .402 + 04 .1670c-02

.1200c + .11 .1200c 4010 0.. 0 0. O.

.2410e e 01 .2440e + 01 - .1270e + 05 .1810e-01

.260rc + 03 .1400c + 040 0. O. O. O. ,

.14".0c + 02 3050c + 03 .9150c + 04 .9830c.02

.2.40c + 03 .2500e + 030 0. O. O. O.

.1720e + 03 .1100c + 04 .1940c + 06 3580e + 00 D .8000c + 03 .9100e + 030 0. O. O. O.

.2170c + 03 3640e + 04 3870c + 05 .7280c-01 L .2100c + 03 .2600e + 030 0. O. O. O.

.9100c + 02 .1110e + 04 .1020c + 06 .1840c + 00 I

p 3800e + 023900e + 020 0. O. O. O.

Amendment 15 2A-21

ABWR wmo Sigpdard Plant bA Table 2A.16 (Cont'd)

FILE 20 DOSE CONVERSION FILE <

(NUREG/CR.2326. pg 4 2)

.8X0e + 02 .1160e + 03 .1600e + 06 .2990c + 00

.1200c + 03 .2100e + 030 0. O. O. O.

.5450c + O2 .8750c + 03 .6030c + 05 .1080c + 00

.4700e + 02 .4700c + 020 0. O. O. O.

.1130c + 03 .1230e + 03 3290e + 06 .5890c + 00

.9200c + 02 .9200c + 020 0. O. O. O.

.8130c + O2 3750c + 03 .1010e + 06 .1830c + 00 '

1900e + 02 .1900c + 020 0. O. O. O.

.43,2 + O2 .4570c + O2 3160c + 06 .58 % ,(0

.8700c + u. 5t700e + 020 0. O. O. O.

.1620e + 03 3220e + 03 .2Xoc + 06 4500e+OC

.3600c + 00 3600c + 000 0. O. O. O,

.1300e + O2 .1820e + 03 .1450c + 05 .1570c 01

.7200c + 00 .7200c + 000 0. O. O. O.

.3170c + O2 .6990c + O2 .4620e+ 05 .8470c 01

.2000c + 04 .4200c + 05 .5000c + 040 0. O. O.

.2030c + 03 .4260e + 04 .2230c + 06 .4030c+(0

.2000c + 04 .5900c + 040 0. O. O. O.

t

.2650c + 03 .4690c + 04 .2930c + 06 .5440e + 00 l- .1300e + 04 .3100c + 05 .5000c + 040 0. O. O.

.7590c + O2 .1600c + 04 .8370c + 05 .1490c + 00

.2000c + 04 .5200c + 040 0. O. O. O.

, .5030c + 02 .4040c + 04 .3500e + 05 .5610c-01 l .6700c + 03 .7000e + 030 0. O. O. O.

.2660c + 03 .1950c + 04 3120e + 06 .6140c + 00

.6100e + O2 3200c + 030 0. O. O. O.

.1650c + 02 3220e + 03 .1810e + 05 3220c-01

.7100e + 02 .1100c + 030 0, 0 O. O.

.4790c + O2 .3060c + 03 .5700e + 05 .9310e-01

.2000c + 03 .7200c + 04 .1180c + 050 0. O. O.

.7820e + 01 .1670c + 03 .4950c + 04 .7550e.02

.2400e + 02 .8600e + 020 0. O. O. O.

O. O. O. O.

.8500e + 02 .1900e + 03 .4000e + 02 .1000e 4 020 0. O.

2460c + 02 .4230c + 03 .2720e + 05 .4340c 01 4900c + 02 .5400c + O2 .2600c+02 .4000e + 02 4000c + 02 3000c + 02 .4000e + 02 3490c + O2 3250c + 03 4010e + 05 .4950c.01 3500c + 05 3400e + 07 .1270c + 09 .1900e + 09 .1700e + 09 .1400e + 09 .1300e + 09

.2260c + 00 .4740e + 01 .2480c + 03 .4020c-04 3300e + 05 3100e + 07 .1270c + 09 .2000e + 09 .1000e + 09 .1800e+(N .1700c + 09

.1110e + 00 .2340e + 01 .1220e + 03 .2030c 04 3300c + 05 .3200c + 07 .1270e + 0) .2000e + 09 .1900e+09 .1800e + 09 .1700c + 09

.2050c + 00 .4320e + 01 .2250c + 03 3660c 04

.1200c + 01 .2700c + 04 .9770e + 06 3120e + 07 .4200c + 07 .4700e + 07 4000e + 07

.9460c 05 .4170c 02 .2070c 01 .7860c 09

.3900e + 05 3600e + 07 .1360e + 09 .2100e+(9 .2000e + 09 .18(Oc + 09 .1700e+(D

,1300e + 02 .2720e + 03 .1420c 4 05 .8620e-02 O

Amendment 15 2A 22

I 23A6100AD Standard Plant Rev A

/~ Table 2A 16 (Cont'd)

( FILE 20 DOSE CONVERSION FILE !'

(NUREG/CR 2326. pg 4 2) 4300e + 05 .1800e + 07 .16We + 07 .7000e + 06 .7000e + 06 .(000e + 06 .4000e + 06 ,

.1930c + 00 .4000e + 01 .2120e+03 3680c44

.4100e + 05 3700e + 07 .1160e + 09 .1300e + 09 .8000e + 08 .5000e + 08 .4000e + 08

.3070e + 01 .6440c + 02 3360c + 04 .26700-02 tee 1 3300e + 03 .2500e + 04 .1000e + 030 0. O. O.

.1300c + 03 .2640c + 04 .14%e + 06 .2530c + 00

.8000e + 03 .1800e + 05 .3000e + 05 .2000e + 040 0. O.

.3120e + 03 .6540e + 04 3420e + 06 .6640e + 00 3700e + 00 3700e + 000 0. O. O. O.

3110e + 00 .6540e + 01 3410e + 03 .6080e.03

.2800e + 00 .2800c + 000 0. O. O. O.

.1670c + 02 .2360e + 02 3190c + 05 .5860e-01

.1000e + 01 .1000e + 010 0. O. O. O.

.20$0c + 02 .2080c + O2 .9990e + 05 .2040c + 00

.2400e + 01 .2400e + 010 0. O. O. O.

.1270e + 03 .1480c + 03 .2400e + 06 .5140e + 00

.1700e + 04 .6500e + 040 - 0. O. O. O.

.1150c+02 .2140c+03 .1270e+ 05 .2400c 01 !

3200c + 043000e + 050 0. O. O. O.

0. O. O. O.

3800e + 04 .2400e + 06 .1060e + 07 .5000e + 06 .2000e + 06 .2000e + 0(4.

(' O. 0. O. O.

.2200e + 03 3700e + 030 0. O. O. O.

.1070e + 03 .2510e + 03 .1100e + 06 .2030c + 00

.8000e + 03 .9700e + 030 0. O. O. O.

O. O. O. O.

.1200e + 04 .2000e + 050 0. O. O. O.

.3140c + 00 .6330e + 01 .3440e + 03 .6670e-03

.2800e + 03 3300e + 040 0. O. O. O.

l

.9980e + 02 .2150e + 04 .1100e + 06 .1970c + 00

.1300e + 03 .1300e + 030 0. O. O. O.

.1750c + 03 .68%e + 03 .2640c + 05 .4970c-01

.2600e + 03 .1200e + 040 0, 0. O. O. ,

3710e + O2 .1910e + 04 .1070c + 06 .1940e + 00

.1@c + 03 .1100e + 030 0. O. O. O.

3320a02 .4960e+03 .2850c+05 .4690e-01

.1000e + 01 .1000e + 020 0. O. O. O.

.1870e + O2 3110e + 02 3140c + 05 .5830e 01

.1800e + 03 .8800e + 030 0. O. O. O.

.7530e + 02 .1490e + 04 .8230c + 05 .1430c + 00

.1700e+ O2 .1700c.+ 020 0. O. O. O.

.6780e + 02 .1050e + 03 .1290e + 06 .2330e + 00

.1900e + 03 3400e + 04 .2500e + 040 0. O. O.

.2790c+O2 .5820e+03 3060e+ 05 .5470c.01

.1600e + 02 .1600e + 020 0. O. O. O.

f .1370c + 02 .9090e + 02 .1610e+05 .2920e-01 Amendment 15 2A.23

ABWR momro Standard Plant wA Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE I (NUREG/CR 2326, pg 4 2) !

.5400e + 01 .5400e + 010 0. O. O. O.

.4770c + 00 .1070c + 01 .6930e + 03 .1220c 02

.1%0c + 03 .1900c + 04 .1000e + 030 0. O. O.

.4670e + 01 .1050e + 03 .4950e + 04 .2060c 02

.1200e + 01 .1200e + 010 0. O. O. O.

.2560c + 01 .2580c + 01 .1350c + 05 .1910c 01

.2600e + 03 .1400e + 040 0. O. O. O.

.1510e +02 3260c + 03 .9810c + 04 .1060e-01

.2300e + 03 .2500e + 030 0, 0. O. O.

.1810e + 03 .1160c + 04 .2030c + 06 3780e + 00

.8000e + 03 .9100e + 030 0. O. O. O.

.2290e + 03 3830c + 04 .4130c + 05 .7780c 01

.2100e + 03 .2600e + 030 0. O. O. O.

l .9540c + 02 .1170e + 04 .1080c + 06 .1930c + 00 3800e + O23900e + 020 0. O. O. O.

.8710e + O2 .1220e + 03 .16% + 06 3140c + 00

.1200e + 03 .2100e + 030 0. O. O. O.

.5730e + 02 .9170e + 03 .6370e + 05 .1140e + 00

.4700e + 02 .4700c + 020 0. O. O. O.

.1190e + 03 .1300e + 03 .3460e + 06 .6190c + 00

.9200e + O2 .920Ge + 020 0. O. O. O.

.8540c + 02 3950c + 03 .1060e + 06 .1920e + 00 l .1900e + O2 .1900e + 020 0. O. O. O.

.4810e + 02 .4820e + 02 3340e + 06 .6190e + 00

.8700e + 02 .8700c + 020 0. O. O. O.

.1680c + 03 3340e + 03 .2370c + 06 .4640e + 00

^

.9200e + 00 .9200e + 000 0. O. O. O.

.1430e + 02 .2010e + 03 .1600e + 05 .1740c-01

.1300e + 01 .1300c + 010 0, 0, 0, 0.

3370e + 02 .7420e + 02 .4910c+05 .9030e 01

.2000e + 04 .4200c + 05 .5000e + 040 0. O. O.

.2140c + 03 .4490c + 04 .2350e + 06 .4250e + 00

.2000c + 04 .5900e + 040 0. O. O. O.

.2800e +03 .4950c+04 .3090c + 06 .5750e + 00

.1300e + 04 3100e + 05 .5000e + NO 0. O. O.

.7940c + 02 .1680e + 04 .8760e + 05 .1560c + 00

.2100e + 04 .5500e + NO 0, 0. O. O.

.5260c + O2 .4180c + 04 3690c + 05 .5920c-01

.6800e + 03 .7100e + 030 0. O. O. O.

.2740c + 03 .2010e + N 3210c + 06 .6330c + 00 l .6100e + 02 3200c + 030 0. 0. O. O.

.1790c + 02 .3500e + 03 .1960e + 05 .3500e 01

.7200e + 02 .1100c + 030 0. O. O. O.

.5140e + 02 3280e + 03 .6110e + 05 .9940c 01

.1900c + 03 .6700c + 04 .1130e + 050 0. O. O.

.8410e + 01 .1790e + 03 .5400e + N .8250c 02

.2600e + 02 .9300e + 020 0. O. O. O.

Amendment 15 O

2A 24

ABWR m om^o Standard Plant Rev A Table 2A.16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326 pg 4 2)

O. O. S. O.

.8600e + O2 .1900e + 03 .4000e + 020 0. O. O.

.2640c + O2 .4540c + t'3 .2930e + 05 .va70e 01

.4800e + 02 .5200e + 020 0. O. O. O.

3740c + O2 3490c + 03 .4300c + 05 .5310c-01

.6300e + O2 .6000c + 04 .2240c + 06 3400e+ 06 3000e + 06 .2300e + 06 .3000e + 06

.2790c + 00 .5870e + 01 3060c+03 .4940e-04

.5900e + O2 .5600e + 04 .2240e + 06 3500e + 06 3400c + 06 .2800e + 06 3000e + 06

.1370e + 00 .2880c + 01 .1500c+03 .2500c-04

.5900e + O2 .5700e + 04 .2240c + 06 3(00e + 06 3400e + 06~ 3700e + 06 .%00e + 06

.2530c + 00 .5320e + 01 .2780c+03 .4500c 04

.5900c.01 .1100c + 02 .1990c + 04 .6000e + 04 .8000e + 04 .8000e + 04 .9000e + 04

.1110c 04 .4880e 02 .2430c-01 .9220c-09

.9900e + O2 .7300e + 04 .2530c + 06 3900e + 06 3500e + 06 .4000e + 06 _V00e + 06

.1520e + O2 3190c + 03 .1660c+05 .1010c-01

.7600e + O2 .3200c + 04 .2900e + 04 .1300e + 04 .1100e + 04 .1000e + 04 .'000e + 03

.2410c + 00 .4990c + 01 .2640c + 03 .4580c.04

.7400e + O2 .6600e + 04 .2030c + 06 .2300e + 06 .1500e + 06 .9000e + 05 .6000e + 05 3500c + 01 .7350c + O2 3830e + 04 3030c 02 st wall

.9600e + 03 .5100c + 04 .1000e + 030 0. O. O.

.9530c + 02 .1940c + 04 .1040e + 06 .1860c + 00

.2300e + 04 3700e+ 05 6000e+ 05 3000c + 040 0. O.

2510e + 03 .5260e + 04 .2750c + 06 .5330c + 00

.1800c + 00 .1800e + 000 0. O. O. O.

.2430c + 00 .5100e + 01 JI/Oc + 03 .4750c 03

.2200e + 00 .2200e + 000 0. O. O. O.

.6800c 4 01 .9580e+ 01 .1300e+05 .2389c-01

.1000e + 01 .1000e + 010 0. O. O. O.

.1690c + O2 .1710e + O2 .8200e + 05 .1680c + 00

.2100c + 01 .2100c + 010 0. O. O. O.

.1060c +03 .1230c+03 ,1990c+06 .4250e + 00

.2200e + 04 .7000e + 040 0. O. O. O.

.8310e + 01 .1550c + 03 .9200c + 04 .1740c-01

.8700c + 03 .1800e + 040 C. O. O. O.

O. O. O. O.

.8900c + 03 .6100e + 04 .2100c + 04 .2000e + 030 0. O.

O. O. O. O.

.6200e + 03 .6800e + 030 0. O. O. O.

.8240e+ 02 .1940c + 03 .8380e + 05 .1560e + 00

.1400c + 04 ,1400e + 040 0. O. O. O.

O. O. O. O.

0. O. O. O,

.1100e + 04 .1600e + 040

.2740c + 00 .5520e + 01 3000e + 03 .5810c-03

.9300e + 03 .5700e + 04 .1000e + 030 0. O. O.

.7480c + 02 .1610e + 04 .8170e + 05 .1480c + 00

.1800e + 04 .1800e + 040 0. O. O. O.

Amendment 15 2A.25

23A6100AD Standard Plant Rev A Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326. pg 4 2) l .1370e + 03 .5340c + 03 .2060e+05 3890e 01

.7600c + 03 .2300e + 040 0. O. O. O.

.7010e + 02 .1380c + 04 .7700c + 05 .1400e + 00

.8500e + 03 .8600c + 030 0. O. O. O.

.2090c + 02 .2660c + 03 .2070e + 05 3420e-01

.2700c + 02 .2700c + 020 0. O. O. O.

.6250e + 01 .1040c + 02 .1050c+05 .1960c-01

.6600e + 03 .1900e + 040 0. O. O. O.

.5830c + 02 .1160e + 04 .6400e + 05 .1110e + 00 3200e + 033200e + 030 0. O. O. O.

A920e + O2 .7320e + O2 .9370e + 05 .1700e + 00 3200e + 04 .7600e + 04 3400e + 040 0. O. O.

.2150c + O2 .4500e + 03 .2360e + 05 .4250e-01

.2700e + 03 .2700c + 030 0, 0. O. O.

.5560e + 01 3700e + O2 .6580c + 04 .1190c-01

.1600e + 03 .1600e + 030 0. O. O. O.

3570e + 00 .7980e + 00 .5180e+03 .9140c 03

.5200c + 03 .1500c + 040 0. O. O. 0. -

.1070e + 01 .2850e + O2 .1050c + 04 4390c 03

.5700e + 02 .5700e + 020 0. O. O. O.

.1820e + 01 .1840c + 01 .9550c + 04 .1360c-01

. .1200c + 04 .1800e + 040 0. O. O. O.

.1040c + 02 .2260c + 03 .6680c + 04 .7190c 02 .

.1000c + 04 .1000c + 040 0. O. O. O.

l- .1260e + 03 .8170c + 03 .1440e + 06 .2660c + 00 1- .2200e + 04 .2400e + 040 0, 0. O. O.

.1610e + 03 .2780c + 04 .1660c+05 3110c.01

.1100e + 04 .1200e + 040 0. O. O. O.

l '

.7150c + 02 .8780c + 03 .8060c + 05 .1450c + 00

.4300c + 03 .4300c + 030 0. O. O. O.

.6530c + 02 .9150e + 02 .1270e + 06 .2360e + 00

.2400e + 03 .2800e + 030 0, 0, 0. O.

.4330c + O2 .6920e + 03 .4800e+05 .8580e-01 3700e + 033700e + 030 0. O. O. O.

.9060c + 02 .9940c + 02 .2650e + 06 .4750c + 00 3900c+03 3900e+030 0. O. O. O.

.6680e + 02 .2960e + 03 .8310e + 05 .1510e + 00

.2700e + 03 .2700e + 030 0. O. O. O.

l 3580e + 023580c + 02 .2480e + 06 .4610e + 00 l .4200e + 03 .4200e + 030 0. O. O. O.

L .1340c + 03 .2520e + 03 .1940e + 06 3810e + 00 4000e + 00 .4000e + 000 0. O. O. O.

4440c + 01 .6240e + 02 .4970e + 04 .5390c 02

.1000e + 01 .1000e + 010 0, 0. O. O.

l' .1440c + 02 .3170e + 02 .2090c + 05 38@c01

.2500e + 04 .4800e + 05 .6000e+ NO 0. O. O.

.1620e + 03 3390e + M .1770c + 06 3220e + 00 1 Amendment 13 2A.26

l Mhh 21A610MD Standard Plant Rev A i

l Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326. pg 4 2)

.2500c + 04 .6900e + 040 0. O. O. O.

.1920e 403 3400e + 04 .2120e + 06 .3940c + 00

.1600c + 04 3300e + 05 .7000e + 040 0. O. O.

.6180c + 02 .1310e + 04 .6820e + 05 .1220e + 00

.9600c + 03 .1500e + 040 0. O. O. O.

39$0e+023360c 404 .2630c + 05 .4190c-01

.1700e + 04 .1700e + 040 0. O. O. O.

.2250c + 03 .1650c + 04 .2640c + 06 .5190e + 00

.4200e + 03 .5900e + 030 0. O. O. O.

.5890c + 01 .1150c + 03 .6470c + 04 .1150c 01

.7800e + 03 .7900e + 030 0. O. O. O.

.2700e + 02 .1720e + 03 3210e + 05 .5220e 01

.2600e + 04 .4000c + 04 .8000e + 030 0. O. O.

.4500e + 01 .9690c + 02 .1740c + 04 .2650c.02

.6100e + 03 .6200e + 030 0. O. O. O.

O. O. O. O.

.6500e + 03 .7300e + 030 0. O. O. O.

.1520e + 02 .2620e + 03 .1690e+05 .2690c 01

.5200e + 03 .5300c + 030 0. O. O. O.

.1400c + 02 .1310e + 03 .1610e + 05 .1990c-01

.1200e + 04 .7200e + 04 .2230c + 06 3400e + 06 3000e + 06 .2300e + 06 .2000e + 06 3520c 01 .7380e + 00 3850e + 02 .6220c 05 v .1100c + 04 .6700e + 04 .2230e + 06 3500e + 06 34COc + 06 .2800e + 06 3000e + 06

.2110e-01 .4420c + 00 .2300e + 02 3860c.05

.1100e + 04 .6800e + 04 .2230c + 06 3500e + 06 3500e + 06 .2700e + 06 .3000e + 06 3440c-01 .7230c + 00 3780e + 02 .6140c-05

.1100e + O2 .1700e + O2 .1780e + 04 .5600e + 04 .7600e + 04 .8000e + 04 .8000e + 04 30$0c 05 .1350c02 .6680e-02 .2540c-09

.1300e + 04 .8200e + 04 .2420e + 06 3700e + 06 3600e + 06 3200e + 06 .3000e + 06

.4180c + 01 .8780c + 02 .4570e + 04 .2780c 02

.1300c + 04 .4500e + 04 .2900e + 04 .1300e + 04 .1100e + 04 .1200e + 04 ,1000e + 04

.2760c-01 .5710e + 00 3020e + 02 .5250e 05

.1300e +04 .7900e+04 .2020e + 06 .2300e + 06 .1400e + 06 .1000e + 06 .(OO0e + 05

.1030e + 01 .2160c + 02 .1130c + 04 .8920c-03 si+ cont

.1900c + 04 .2800e + 040 0. O. O. O.

.8430e + 02 .1710e + 04 .9290c + 05 .1640e + 00

.4700e + 04 .1200c + 05 .1300e + 05 .1000e + 040 0. O.

.2170e + 03 .4560c + 04 .2380e + 06 .4640c + 00

.1800e + 00 .1800e + 000 -0. O. O. O.

.1760e + 00 3680e + 01 .1920e + 03 3420e-03

.2300e + 00 .2300e + 000 0. O. O. O.

.7330e + 01 .1030c + 02 .1400e + 05 .2570c.01

.1000c + 01 .1000c + 010 0. O. O. O.

.1410e + 02 .1420e + 02 .6820e + 05 .1400e + 00

.2200e + 01 .2200e + 010 0. O. O. O.

( .9180e + 02 .1070c + 03 .1720e + 06 3670e + 00 Amendment 15 2A 27

ABM as46imso Standard Plant Rev A Table 2A.16 (Cont'd) ,

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2) l

.1800c + 04 .6600e + 040 0. O. O. O.

.7670c + 01 .1430e + 03 .8450c + 04 .1600c-01

.1200e + 04 .2200e + 040 0. O. O. O.

O. O. O. O.

.1200e + 04 .6400e + 04 .2100c + 04 .1000e + 03 .1000e + 030 0.

O. O. O. O.

.9600e + 03 .1000e + 040 0. O. O. O.

.6510e + O2 .1520e + 03 .6860e + 05 .1270c + 00 3200e+04 3300e+040 0. O. O. O.

O. O. O. O.

.2700e + 04 .3300e + 04 .1000e + 030 0. O. O.

.22ide + 00 .4580c + 01 .2490c + 03 .4810c-03

.2000c + 04 3200c + 040 0. O. O. O.

.6170c + 02 .1330c + 04 .6750c + 05 .1220e + 00 3900e + 043900e + 040 0. O. O. O.

.1010c + 03 .3930e + 03 .1700e + 05 .3190c-01

.1600e + 04 .1900e + 040 0. O. O. O.

I470c + 02 .1270c + 04 .7110e + 05 .1290e + 00

.1900c + 04 .1900e + 040 0. O. O. O.

.1700c + 02 .2260e + 03 .1620e + 05 .2670c-01

.1100e + 02 .1100e + 020 0. O. O. O.

.6250c + 01 .1040e + 02 .1050c+05 .1950c 01

.1400e + 04 .1700e + 040 0. O. O. O.

.4220e + 02 .8360e + 03 .4630c + 05 .8030c-01

.4900e + 03 .4900e + 030 0, 0. O. O.

3750e + O2 .5710e + 02 .7120e + 05 .1290c + 00

.7400c + 04 .1000c + 05 3000e + 040 0. O. O.

.1580e + 02 3300e + 03 .1730c + 05 3110c 01

.5900e 4 03 .5900e + 030 0. O. O. O.

.5990c + 01 3990c +02 .7090e + 04 .1280e 01

.2400c + 03 .2400e + 030 - 0, 0. O. O.

.2630c + 00 .5870c + 00 3810c+03 .6720c-03

.9500e + 03 .2400e + 040 0. O. O. O.

.5160c + 00 .1530c + 02 .4740c + 03 .1970e 03 3800e + 023800e + 020 0. O. O. O.

.1390c + 01 .1410c + 01 ,7320c + 04 .1040c 01

.2200e+04 .2700e +040 0. O. O. O.

.7700c + 01 .1670c e 03 .4780e + 04 .5140c-02

.1700e + 04 .170Ce + 040 0. O. O. O.

.1140e + 03 .723)c + 03 .1300e + 06 -.2400c + 00

.1500e + 04 .150N + 040 0. O. O. O.

.1340c + 03 .2290e + 04 .1760e+05 3310e 01

.2400e + 04 .2500e + M 0. O. O. O.

.5480c + 02 .6730c + 03 .6180c + 05 .1110e + 00

.6500e +03 6500e + 030 0. O. O. O.

.5690e + O2 .7960e + 02 .1100e + 06 .2060e + 00

.9600e + 02 .1300e + 030 0. O. O. O.

Amendment 15 2A.28 O

ABWR zuum^o Standard Plant Rn ^

Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326, pg 4 2) 3170c + O2 .5080c + 03 3530e+05 .6280e-01

.8700e+ O2 .8700c+ 020 0. O. O. O.

.7340c + O2 .8040c + O2 .2140e + 06 3830c + 00

.1200c + 03 .1200c + 030 0. O. O. O.

.%I60c + O2 .2250c + 03 .6300c + 05 .1140c + 00

.4600e + 02 A600e + 020 0. O. O. O.

3170c + 02 3170c + 02 .2200c + 06 A080e + 00

.1300e + 03 .1300c + 030 0. O. O. O.

.1140c + 03 .2170c + 03 .1660c + 06 3250e + 00 '

.4200e + 00 .4200c + 000 0. O. O. O.

A090e + 01 .5750c + 02 A580e + 04 .4970c 02

.1000e + 01 .1000e + 010 0, 0. O. O.

.1500c + O2 3300c + 02 .2180c + 05 A000c-01

.2500e + 04 .5200e + 05 .7000c + 040 0. O. O.

.1320e + 03 .2770c + 04 .1450c + 06 .2620e + 00

.2600e + 04 .7500c + 040 0. O. O. 0

.1770c + 03 3140c + 04 .1960e + 06 3640e + 00

.1500c + 04 3500e + 05 .7000e + 040 0. O. O.

.4470e + 02 .9440c + 03 A920c + 05 .8780e 01

.1600e + 04 .2X10e 4 tr.v 0. O. O. O.

3100e + O2 .2770c+ 04 .1980e+05 3170c 01 O

3600e + 043600e + 040 0. O.

.1880c + 03 .1380c + 04 .2210e 4 06 AXoc + 00

0. O.

O.

.1000c + 04 .1100c + 040

.5800c 4 01 .1140c + 03 .6380c + 04 .1140c 01

.1800c + 04 .1800e + 040 0. O. O. O.

.2450c + 02 .1560c + 03 .2910e + 05 A750e-01

.6700e + 04 .7400c + 04 A000e + 030 0. O. O.

3870c + 01 .8310e + 02 .1670c + 04 .2550c 02

.1500e + 04 .1500c + 040 0. O. O. O.

O. O. O. O.

.1600c + 04 .1600e + 040 0. O. O. O.

.1180c + 02 .2030c + 03 .1310e + 05 .2090c-01

.1200e + 04 .1200c + 040 0. O. O. O.

.1460c + 02 .1360e + 03 .1680c + 05 .2080c 01

.2900e + 04 .9000c + 04 .2310e + 06 3300c + 06 _%00c + 06 .2.%0e + 06 .2000e + 06

.1540c-01 3230e + 00 .1680c + 02 .2730e-05

.2700e + 04 .8400c + 04 .2220e + 06 3500e + 063400c + 06 .2800e + % 3000e + 06

.9990c-02 .2100e + 00 .1090c + 02 .18.%e 05

.2700c + 04 .8500e + 04 .2210e + 06 3600e+ 06 3400e + 06 .2700e + 06 3000e + 06

.1680e-01 3520c+ 00 .1840c + 02 3000c-05

.2700e + O2 3500e + 02 .1760c + N .5600e + 04 .7600e + 04 .8000e + 04 .8000e + 04

.2110c 05 .9280e 03 4620c.02 .1750c 09 3100e + 04 .9700e + 04 .2400e + 06 3700e + 06 3600e + 06 3200e + 06 3000e + 06

.2880e + 01 .6060c + O2 3150c + 04 .1920e 02 3200e + 04 .6500c + N .2800e + 04 .1700e + 04 .1000e + 04 .1000e + 04 .1000e + 04

.9920e-02 .2050e + 00 .1090c + 02 .1880e 05 Amendment 15 2^.29 l

1

ABWR mamo Standard Plant nu A Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326 pg 4 2)

.3100e + N .9800e + 04 .2000e + 06 .2300e + 06 .1500e + 06 .9000e + 05 .6000e + 05

.8240e + 00 .1730e + 02 .9020c + 03 .7140e 03 uli wall 3400e + 04 .4500e + 040 0. O. O. O.

.9730c + 02 .1980e + 04 .1070c + 06 .1900c + 00

.8500e + 04 .1800e + 05 .1700c + 05 .1000e + 040 0. O.

.2730c + 03 .5730c + 04 3000e+ 06 .5830c + 00

.1800e + 00 .1800e + 000 0. O. O. O.

.1890c + 00 3960e + 01 .2070e+ 03 3690e 03

.2300e + 00 .2300c + 000 0. O. O. O.

.7540c + 01 .1060e + 02 .1440c + 05 .2640c.01

.1000c + 01 .1000c + 010 0. O. O. O.

.1900e + O2 .1930c + O2 .9200c + 05 .1890e + 00

.2200e + 01 .2200- + 010 0. O. O. O.

.1330c + 03 .1550c + 90 .2510e + 06 .5360c + 00

.2100e + 04 .6900e + 040 C. O. O. O.

.8950c + 01 .1670c + 03 .9900c + 04 .1870c 01

.4900c 4 04 .5800e + 040 0. O. O. O.

O. O. O. O.

.4200c + 04 .9400c + 04 .2600c + 040 0. O. O.

O. O. O. O.

.2400e + 04 .2400e + 040 0. O. O. O.

.7380e + 02 .1730c + 03 .7920e + 05 .1470c + 00

.1700e + 05 .1700e + 050 0. O. O. O.

O. O. O. O.

.1600e + 05 .1800e + 050 0. O. O. O.

.3010c + 00 .6060e + 01 3300c + 03 .6390c-03

.5500e + 04 .7400e + 040 0. O. O. O.

.6960e + O2 .1510e + 04 .7600e + 05 .1370e + 00

.1300e + 05 .1300e + 050 0. O. O. O.

.1100e + 03 .4290c + 03 .2050e + 05 3860c 01 3200e >04 3600c+040 0. O. O. O.

.7550e + 02 .1490e + 04 .8300c + 05 .1510c + 00

.8600e + 04 .8600e + 040 0, 0. O. O.

.1830c + 02 .2380c + 03 .1780e + 05 .2920e 01

.1300e + 02 .1300e + 020 0, 0. O. O.

.6120e + 01 .1020e + 02 .1020e + 05 .1910e-01

.4100e + 04 .4500e + NO 0. 0, 0. O.

.4540c + 02 .9000c + 03 .4980e + 05 .840c 01

.1100c + M .1100e + 040 0. O. O. O.

.4050c + 02 .6140e + 02 .7690e + 05 .1390e + 00

.4200e + 05 .4700e + 05 3000e + N0 0. O. O.

.1710e + O2 3580e + 03 .1880e + 05 3390e-01

.2500e + 04 .2500e + NO 0. O. O. O. l

.6160e + 01 .4100c + 02 ct290e+ M .1320c.01

.7300e + 03 .7300c + 030 0. O. O. O.

.2820e + 00 .6300e + 00 .4090c + 03 .7220e 03 Amendment 15 2A.30 l

1 ABWR := =^o Standard Plant um A Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.5500e + 04 .9900e + 040 0. O. O. O.

.6880e + 00 .1920e + 02 .6560c + 03 .2730e-03

.2600e + O2 .2600e + 020 0. O. O. O.

.1530c + 01 .1540e + 01 .8020c + 04 .1140c-01

.1200c + 05 .1E0e + 050 0. O. O. O.

.8410e + 01 .1830e + 03 .5220e + 04 .5610c-02

.4900e + 04 .5000e + 040 0. O. O. O.

.1350e 4 03 .8530c + 03 .1550c + 06 .2860c + 00

.3100e + 04 3200c + 040 0. O. O. O.

.1530c + 03 .2620e + 04 .1810e+05 3420e-01

.1000c + 05 .1000e + 050 0. O. O. O.

.M100e + O2 .7360c + 03 .6760e + 05 .1210e4 00

.1300c + 04 .1300e 4 040 0. O. O. O.

.6700e 4 02 .9370c + 02 .1310e + 06 .2420e + 00

.1700e + 03 .2000e + 030 0. O. O. O.

3410e + 02 .5450c + 03 .3780c+05 .6750c-01

.9800e + 02 .9800e + 020 0. O. O. O.

.8450c + 02 .9260c + 02 .2470c + 06 4420e + 00

.2300c + 03 .2300c + 030 0. O. O. O.

.5670c + O2 .2520e + 03 .7060c + 05 .1280c + 00 3700e+O2 3700e,020 0. O. O. O.

h V

3720e + O23730c + O2 .2580c + 06 .4810e + 00

.2100e + 03 .2100e + 030 0, 0, 0, 0,

.1420e + 03 .2680c + 03 .2110e + 06 .4140c + 00

.4200c + 00 .4200e + 000 0. O. O. O.

.4120e + 01 .5790e + O2 .4620e + 04 .5000e 02

.1000e + 01 .1000e + 010 - 0. O. O. O.

.1540e + 02 3410c + 02 .2250c + 05 .4140e-01

.2500e + 04 .5200e + 05 .6000e + 040 0. O. O.

.1500e + 03 3130c + 04 .1640e+06 .2970c + 00

.2500c + 04 .7300e + 040 0. O. O. O.

.2080c + 03 3670c + 04 .2300e + 06 .4250e + 00

.1600c + 04 3500e + 05 .6000e + 040 0. O. O.

.4810e + O2 .1020e + 04 .5300e+05 .9440c 01

.4900e + 04 .5500e + 040 0. O. O. O.

.3600e + O2 3490e + 04 .2120e+05 3390c 01

.1100e + 05 .1100e + 050 0. O. O. O.

.2410e + 03 .1770e + 04 .2820e + 06 .5580e + 00

.5200e+ 04 .5300e + 040 ' 0. O. O. O.

.5710e + 01 .1120c + 03 .6270c + 04 .1120e.01

.7600e + 04 .7600e + 040 0, 0. O. O.

.2650e + 02 .1690e + 03 3140c+05 .5110c 01 3900e + 05 .4200e + 05 .1000e + NO 0. O. O.

.4680e + 01 .1010e + 03 .1660c + 04 .2530e.02

.8700c + 04 .8800e + 040 0. O. O. O.

0. - 0. O. O.

q .7800e + 04 . 7800e + 040 0. O. O. O.

G 2A-31 Amendment 15

ABWR awan Standard Plant Rev A Table 2A.16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326, pg 4 2)

.1250c + 02 .2150c + 03 .1380e+05 .2210e 01

.5600c + 04 .5600e + NO 0. O. O. O.

.1480e + O2 .1380e + 03 .1700e + 05 .2100e.01

.1700e + 05 .2400e + 05 .2260e + 06 3400e + 06 .2900e + 06 .2200e + 06 3000e + 06

.1630c-01 3430c + 00 .1790e+02 3890e-05

.1600e + 05 .2200e + 05 .2280c + 06 3500e + 06 .3300e + 06 3700e + 06 .2000e + 06

.1150e-01 .2410c + 00 .1250e + 02 .2100e-05

.1600c + 05 .2300e + 05 .2270e + 06 3500e + 06 3400e+06 36(X)e + 06 .3000e + 06

.1770c 01 3730c + 00 .1940e + 02 3170e 05

.1(/X)e + 03 .1800e + 03 .1820e + 04 .5600e + 04 .7400e + 04 .8000e + 04 .8000e + 04

.2340e05 .1030c-02 .5120c-02 .1940e-09

.1800e + 05 2600e + 05 .2440c + 06 3700e + 06 36(X)e + 06 30(X)e + 06 3(XX)e + 06 3200e + 01 .6730e + 02 3500e + 04 .2130e 02

.1900e + 05 .2300c + 05 .?D00e + 04 .2000e + 04 .1000e + 04 .1000e + 04 .1000e + 04

! .1080c 01 .2230e + 00 .1180c + O2 .?N0e 05

.1800e + 05 .2600e + 05 .?o40c + 06 .2300e + 06 .1400e + 06 .1000e + 06 .6000e + 05

.8670c + 00 .1820e + 02 .9460c + 03 .75(X)e-03

!!i wall

.4280c + 04 .4650c + 040 0. O. O. O.

.7270e + 02 .1480c + 04 .7970e + 05 .1420e + 00

.1010e + 05 .1230e + 05 3920e + NO 0. O. O.

. .2170c + 03 .4540c + 04 .2370c + 06 .4610e + 00 i .6120c.01 .6120c-010 0. O. O. O.

.1760e + 00 3680c + 01 .1920e+03 .3420c 03

.9460c 01 .9460c 010 0. O. O. O.

.5760c + 01 .8110e + 01 .1100e+05 .2020e-01

.8100e + 00 .8100c + 000 0, 0. O. O.

.1440c + 02 .1460c + 02 .6990c + 05 .1430c + 00

.2070c + 01 .2070c + 010 0. O. O. O.

l .9500c + 02 .1110e + 03 .1790e + 06 3830c + 00

.1400e + 04 3900e + 040 0. O. O. O.

.6390e + 01 .1190c + 03 .7N0c + 04 .1330e-01

.6860c + 04 .6860c + NO 0. O. O. O.

O. O. O. O.

.9100c + 04 .1300c + 05 .1300e + 040 0. O. O.

O. O. O. O.

.1430c + 04 .1480c + NO 0. O. O. O.

.6240c + 02 .1470e + 03 .6470c + 05 .1210e + 00

.2600e + 05 .2600e + 050 0. O. O. O.

O. O. O. O.

.2300e + 05 .2600e + 050 0. O. O. O.

.2540e + 00 .5120e + 01 .2790e + 03 .5390c.03 3920e + 04 .4480e + 040 0. O. O. O.

.5580c + 02 .1200c + 04 .6110e + 05 .1100e + 00

.1100e + 05 .1100e + 050 0. 0, 0, 0,

.1010c + 03 3920e + 03 .1680e + 05 3170c 01 3110c + 04 3240c + 040 0. O. O. O.

O Amendment 15 2A 32

ABWR zwa=^o Standard Plant Rev A Table 2A.16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326. pg 4 2)

.5390e + 02 .1060e + 04 .5920e + 05 .1070c + 00

.7400c + 04 .7770c + 040 0. O. O. O.

.1560e + 02 .2020e + 03 .1520e + 05 .2520e 01

.6930e + 01 .6930c + 010 0. O. O. O.

.5060c + 01 .8400c + 01 .8470c + 04 .1580e 01

.2000e + 04 .2200e + 040 0. O. O. O.

.4220e + 02 .8360c + 03 .4630c + 05 .8030c-01

.4920c + 03 .4920c + 030 0. O. O. O.

[~.

3620e + O2 .54Mc + 02 .6890c + 05 .1250e + 00

.8880c + 05 .9620e + 05 .7400e 4 040 0. O. O.

.1570c+O2 3280c+03 .1720c+05 3080c.01

0. O. O.

4080c + 03 .4080e + 030 O.

r .4710c + 01 3130c+ 02 .5570c+04 .1010e-01

.1790c + 03 .1790c + 030 0. O. O. O.

g

.2590e + 00 .5800e + 00 3770c + 03 .6640c-03

' 0. O. O. O.

3570c + 04 .5040e + 040 3860c + 00 .1260c + O2 3330e + 03 .1380e 03

.2560e + 01 .2560e + 010 0, O. O. O.

.1330c + 01 .1340c + 01 .6980c + 04 . .9920c-02

0. O, O. O.

.1500e + 05 .1670c + 050

.7460c + 01 .1620e + 03 .4700e + 04 .5060c-02 O .5350c + 04 .5350c + 040 0.

.1010e+ 03 .6490e+ 03 .1160e + 06 .2140e + 00 0.

O.

3300e + 04 3350e + 040

.12Mc + 03 .2140c + 04 .1380e + 05 .2610e 01

.9100c + 04 .9100e + 040 0. O. O. O.

.5260c + 02 .6460c + 03 .5930c + 05 .1060e + 00 4860c + 03 .4860e + 030 0. O. O. O.

[

.5120e + 02 .7170e + 02 .9990e + 05 .1850c + 00

0. O. O. O.

.7260e + 02 .7920e + 020 3140e + 02 .5030c + 03 3490c + 05 .6220c 01 4440c + 02 .4440c + 020 0. O. O. O.

.6930c + 02 .7600e + 02 .2020e + 06 3640e + 00

.1520e + 03 .1520e + 030 0. O. O. O.

.5030e + 02 .2220e + 03 .6250e + 05 .n30c + 00

.1720e + 02 .1720e + 020 0. O. O. O.

.2800e + O2 .2800e + 02 .1940c + 06 3610e + 00

.1080c + 03 .1080c + 0% 0. O. O. O.

.1160c + 03 .2170e + 03 .1710e + 06 3330c + 00

.1300c + 00 .1300e + 000 0. O. O. O.

3280c + 01 .4610e + 02 3670e+04 3970c-02 4850c + 00 .4850c+ 000 0. O. O. O.

.1200e + 02 .2640c + 02 .1740e+05 3190c-01

.2100e + 04 3970e + 05 .5250e + 040 0. O. O.

.1220e + 03 .2550c + 04 .1330c + 06 .2410e + 00

.2320e + 04 .6390c + 040 0. O. O. O.

9 .1530e + 03 .2710e + N .1690e + 06 3140e + 00 Amendment 13 2^*33

ABWR wmn Standard Plant Rev A Table 2A.16 (Cont'd) g FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.9120e + 03 .1730c + 05 .2880e + 040 0. O. O.

.4460c + 02 .9440e + 03 .4920e + 05 .8780c 01

.7840c + 04 .7840c + 040 0. O. O. O.

.3130e + O2 .2900c + 04 .1920e + 05 .XLSI0c-01

.1070c + 05 .1070c + 0$0 0. O. O. O.

.1980c + 03 .1460c + N .2320e + 06 .4580c + 00

.1200c + 04 .1200e + NO 0. O. O. O.

.4680e + 01 .9170c + 02 .5140c + 04 .9170e-02

.5760e + 04 .5760c + 040 0. O. O. O.

.2@0c + 02 .1330c + 03 .2480e+05 .4060c 01

.8520e + 05 .8520e + 05 .7100e + 040 0. O. O.

3690c + 01 .7960c + O2 .1350c + 04 .2060c-02

.6750c + 04 .6750c + 040 0. O. O, 0.

l 0. O. O, O.

3990c + 04 .4180c + 040 0. O. O. O.

.1110e + 02 .1910e + 03 .1240e + 05 .1970c-01

.1040e + 04 .1040c + 040 0. O. O. O.

.1170c + 02 .1090c + 03 .1340c + 05 .16(Oc-01

.5000e + 04 .6100e + 04 .2290c + 05 3400e + 05 3000e + 05 .2700e + 05 .2000e + 05

.7960e 01 .1670c + 01 .8710e + 02 .1410e 04 4600c + 04 .5600c + 04 .2240c + 05 3500e + 05 3400e + 05 3300e + 05 3(00e + 05 3290e-01 .6910e + 00 3600e + 02 .ro30c-05

.4700e + 04 .5700e + 04 .2330c + 05 3500e + 05 3400e + 05 3200e + 05 .3000e + 05

.7110e 01 .1490c + 01 .'n90c + 02 .1270c 04 0, 0, 0. O. O. 0. O.

.1980c-05 .8720e 03 .4330c 02 .1640c-09

.5200e + 04 .6400e + 04 .2460c + 05 3700c + 05 3200e + 05 .4000e + 05 3000e + 05

.2710e + 01 .5690c + O2 .2960c + 04 .1800e 02

.5500c + 04 .6100e + 04 3000e + 03 .2000e + 03 .1000e + 03 .1000e + 03 .1000e + 03

,6920c.01 .1430c + 01 .7590c + 02 .1310c04

.5200e + 04 .6400e + N .2060c + 05 .2300e + 05 .1500e + 05 .9000e + N .6000e + N

.7270c + 00 .1530c + O2 .7950e + 03 .6280e 03 thyroid

.1100e + 03 3200e + N .1000e + 030 0. O. .

.8700e + 02 .1780e + 04 .9640c + 05 .1710e + 00

.2400e + 03 .2100e + 05 3700e + 05 .1000e + 04 ,1000e + No 0.

.2510e + 03 .5280c + 04 .2730e + 06 .5330e + 00

.1800c + 00 .1800c + 000 0, 0. O. O.

.2550e + 00 .5380e + 01 .2800e + 03 .5000c03

.2000e + 00 .2000e + 000 0. O. O. O.

.1000c + 02 .1400c + 02 .1890e+05 3500c 01

.9700e + 00 .9700e + 000 0. O, 0, 0,

.1770c + 02 .1790c + 02 .8580c + 05 .1740c + 00

.2000e + 01 .2000e + 010 0. 0, 0, 0,

.1080c + 03 .1260c + 03 .2020e + 06 .4310e + 00

.5000c + 03 .6500e + N0 0. O. O. O.

,7300c + 01 .1370c + 03 .8060c + 04 .1540c 01 Amendment 13 2A 34

2M6100AD Standard Plant an. x

g Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326, pg 4 2)

.2600c + 03 .1500e + 040 0. O. O. O.

O. O. O. O.

7,00e + 03 .5900e + 04 .2100e + 04 .2000e + 030 0. O.

O. O. O. O.

.1300e + 03 .2000e + 030 0. O. O. O.

.8300e + 02 .1950e + 03 .8320e + 05 .1540e + 00 .,

.8200e + 01 .2100e + 020 0. O. O. O.

O. O. O. O.

.7100e + 01 3900c + 030 0. O. O. O.

.2930e + 00 .59t0c + 01 ' 3220e + 03 .6220e 03

.7900e + 02 3500e + 04 .1000e + 030 0. O. O.

.7240c + 02 .1540e + 04 .7880c + 05 .1430e + 00

.,7700e + 02 .8600c 4 020 0. O. O. O.

.1440e + 03 .5580e + 03 .2110e + 05 . 3940e-01 1

.8100e + 02 1300c + 040 - 0, O. O. O.

'.6230:+ 02 .1220e + 04 .6830e + 05 .1220e + 00

.9400e+02 .1500e+030 0. O. O. 0. .

.2420e + 02 .3460c + 03 .2150c + 05 .3560c-01 4600e + 02 .4600e + 020 0. O. O. O.

.1180c + 02 .1970e + O2 .1990c+05. 3720c 01'

.5200e + O2 .9600e + 030 0, 0. O. O.

'G. .61400 ; 02 :.1220e + 04 .6710e + 05 ',1170e + 00 L d ~

1400e + 02 .1500e + 020 0. O. O. O.

l '. 5260e + 02 .8010e + O2 .1010e + 06 .1820e + 00

.4800e + 02 3600e + 04 .2700e + 040 0, 0. O.

.2270c + 02 .4760c + 03 .2480e+05 .4470e-01

.6400e + 01 .9900e + 0N 0.- 0. O. O.

.8200e + 01 .5450c + 02 .9640c + 04 .1730c 01

.2900e +01 3000e + 010 0, 0, 0, 0, 3

3830c+60 .8600e+00 .5590c+03 .9810c-03

.1600e+02 .2100e+030 0. O. O. O.

.2660e + 01 .6200e + 02 .2790e + 04 .1150e-02

.8100c + 00 .8100e + 000 0. O. O. O.

.1960c + 01 .1980c + 01 .1020e+ 05 .1470c 01 ,

f p .4300e+02 3900e+030 0. 0. O. O.

.1160e +(,2 .2500e + 03 .7620e + 04 .8170e-02 ,

.4500e + 04 ' .9500e + 050 0, 0. O. . O.

.1240c 403 .8100e+03 .1410e+06 .2610e + 00 4800e + 05 .9700e + 050 0. 0. - . O. O.

.1680e + 03 .2850c + 04 .2440e+05 .4580e 01 l .1000e + 03 .2200e + 030 - l 0. O. - 0. O.

.7430e + O2 .9200e + 03 .8320e + 05 . 1500e + 00 ,

3700e+02 3800e+020 0. O. O. O.

.6150e+02 .8700e+02 .1200e + 06 .2250c + 00

.1300e + 06 .1100e + 070 0, 0. O. O.

.4650c+ 02 .7370e+03 .5150e+05 .9170c-01 l ..~ p .6600c + 04 .6600e + 040 0. O. O. O.

U 2A-35 Amendment 15

ABWR 23A6100AD Standard Plant nev.A Table 2A 16 (Cont'd)

~

FILE 20 DOSE CONVERSION FILE .

(NUREG/CR 2326 pg4 2) l .9000e + 02 .9900e + 02 .2640c + 06 .4720e + 00

!. ,1200e + 06 .1800e + 060 0, O. O. 0,

.6970e + O2 3150e + 03 .8670e + 05 .1560e + 00

.1100e + 04 .1100e + 040 0, 0, 0. O.

3330e + 02 3340e +02 .2300c + 06 .4280e + 00

.4300e + 05 .4400e + 050 0, 0, 0. 0,

.1380c+03 .26600+ 03 .1980c+06 3890e + 00 3900e +00 .4000e + 000 0, O. O. O.

.8900e + 01 .1250c + 03 .9900c + 04 .1070c 01

.9100:+ 00 .9100e + 000 0. O. O. O.

.2050c + 02 ' .4520e + 02 .2980c+05 .5470e 01

.5800e 4 03 .4300e + 05 .5000e + 040 0. 0, 0,

.1590c + 03 3310e + 04 .1730e + 06 3140e +00

.6900e + 03 .6000e + 040 - 0, 0. O. 0,

.1860c + 03 3280e + 04 .2050c + 06 3810e + 00 l 3600e+03 3100e+05 .5000e + 040 0. O. O.

.6540c + 02 .1380e + 04 .7170e + 05 .1280e + 00

.2200e + 03 .1200e + 040 0. O, L O. O.

.4270e + 02 .3550c + 04 .2880c+05 .4610e 01

.1500e + 03 .2300e + 030 0, 0. O. O,

.2370e + 03 .1740e + 04 .2770c + 06. 5470e + 00

.6000c + 01 .9200e + 020 0, 0, O. O. .

.1130c + 02 .2200e + 03 ,1240c + 05 .2220e-01

.1800e+02 .2500e+020 0. O. 0, 0.

3290e + 02 .2100e + 03 ' 3900e + 05 .6360c.01

.5100e + 01 3500e + 03 .2600e + 030 0. O. O.

.6060c + 01 .1300e +03 3370c + 04 .5140e.02

,9300c-01 .1000c + 010 0. O, O. O, 0, - 0. O. O.

.1200e+02 .6900e+02 .1000e + 010 0, 0, 0,

.1910e + 02 3280e + 03 .2110e + 05 3360c 01 .

.8200e + 01 .1500e + 020 0, O. O. O,

,22/0e+02 .2110e+03 . 2600c +05 3220e-01

~ 1800e + 02 ' .5900c + 04 .2240c + 06 3400e + 06 .2900c + 06 .2400c + 06 .2000c + 06

.4400e.01 ' .9200e + 00 .4820e + 02 .7780e 05

,1600e + 02 .5600e + 04 .2240c + 06 3500c + 06 .3400e + 06 .2800e + 06 3000e + 06 3200c-01 .6900e + 00 3590c + 02 . 6000e-05

.1600c + 02 5600e + N .2240c + 063500e + 06 3400e +06 .2800e + 06 3000e+06.

.4500c.01 .9500c + 00 .4910e + 02 .778')e 05 ;

~

.1700e 02 .5300e + 01 .1690e + 04 .5700e + N .7600e + 04 .8000e + 04 .8000e + N

.4800c05 .2120c-02 ,1060e.01 .4000c-09

.2000e + O2 .6500e + 04 ' .2430e + 06 3700e +06 3500e + 06. 3300e + 06 3000e + 06.

,6600c + 01 .1380e + 03 .7270c + 04 .4390c-02

.2100e + 02 3200c + 04 ~ .2800c + 04 .1300c + 04 .1200e + 04 .1000c + 04 .5000e + 03 3200c.01 .6600c + 00 3590c +02 .6110c-05

.2000e + O2 .6600e + 04 .2030e + 06 .2200e + 06 .1500e + 06 .1000c + 06 6000e + 05

.1740c + 01 3590c +02 .1890e + 04 .1500e-02 2A-36 O

Amendment 15

~l l

23A6100AD Standard Plant Rev.A 1

ll Table 2A.16 (Cont'd)-

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326, pg 4 2)-

other

.2400e + 03 3400e + 040 0. O. O. O.

'.1400e + 03 .2860c + 04 .1540e + 06 .2750e + 00

.5800e + 03 .2400c + 05 .4000e + 05 .2000e + N .1000e + N0 0, 3310e + 03 .6960c + 04 3630e+06 .7060c + 00

.3100c + 00 3100e + 000 0, 0. O. O.

.2960c + 00 .6230e + 01 - 3250e + 03 .5810e-03

.2600e + 00 .2600e + 000 - O. O. O. O.

.1260e + 02 .1780e + O2 .2410e + 05 .4430e.01

.1000e + 01 .1000e + 010 0. O. O. O.

.2200e + 02 .2220e + 02 1070c + 06 .2180e + 00

.2300e + 01 .2300e + 010 0. O. O. O.

.1410e + 03 .1640e + 03 .2640e + 06 .5640e + 00

.5000e + 03 .6500e + 040 0. O. O.- 0,

.1280e + 02 .2380c + 03 .1410e + 05 .2670c-01

.2600e + 03 .1500c + 040 0. O. 0.. O.

O. O. O. O.

0, 0, l: .1800e + 03 .5900e + 04 .2100c + 04 .2000c + 030 3 0, 0. O. O.

.1500e+03 .2200e+030 0. O. O. 9. '

I

-- .1080e + 03 .2520e + 03 .1120e + 06 .2100e + 00 - !

.2000e+02 5200e+020 0. O.

'(' O. O.

O.

.1800e + 02 .9700e + 03 .1000e + 020 0. O. O.

3210e + 00 .6480c + 01 .3520e + 03 .6810e-03

.1900e + 03 3700e + 04 .1000e + 030 0. O. O.

.1030c + 03 .2240e + 04 .1130e + 06 2040c + 00

.1700e + 03 .1800e + 030 0, ..

O. O. O.

.1680c + 03 .6580c + 03 .2690e+05 .5060e-01

.1900e + 03 .1500e + 040 . O. O. O. O.

.1080e+03 .2120e+04 .1190c + 06 .2150c + 00 '

.6200e+02 .1100e+030 0. O. O. O.

.2930e+02 3950e+03 . 2780c+05 .4570e-01

.8200c + 01 .8200e+ 010 0. O. O. O.

.1130e+02 .1880e+02 .1890c+05 3530e.01 !

'.1200e+03 .1100e+040 0.- 0, O. O.

.7280e + 02 .1450c + 04 .7810e + 05 .1360e + 00

.2400e + 02 .2500c + 020 0. O. O. O.

.6330c + 02 .%70c + 02 ' .1200e + 06 .2170c + 00

.7700c + 02 3700e + 04 .2800e + 040 - 0. O. O.

.2660e + 02 .5560e + 03 .2920e+05 5250c-01

.1500e + 02 .2100e + 020 ' O. O. O. O.

.1030c+02 .6850e+02 .1210e+05. 2190c-01

.4900e + 01 .5100e + 010 0. O. O. O.

.4480c+00 .1000c+01 .6480c + 03 .1140e.02 3100e + 02 .4700e+ 030 0. O. O. O.

.5750c + 01 .1260e + 03 .6160c + 04 .2560e 02 2A.37 Amendment 15

l ABWR mamm ,

Standard Plant . Rn A l Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326, pg 4 2) l l

.1300e + 01 .1300e + 010 0. O. O. O.

.2520c + 01 .2540c + 01 .1320c+05 .1870c-01

.7900e + 02 .7400e + 030 0. O. O. O.

.1500e+02 3240c + 03 .9850c + 04 .1060c-01

.2200e + 03 3300e + 030 0. O. O. O. ;

.1870c + 03 .11% + 04 .2120e + 06 3950c + 00

.5000c + 03 .1000c + 040 0. O. O. O.

.2240c + 03 3810e + 04 3070c + 05 .5770c.01

.1700e + 03 3300e + 030 0. O. O. O.

.9240c + 02 .1130c + 04 .1050c + 06 .1880c + 00

.4900e + 02 .5000e + 020 0. O. O. O.

.9280c + 02 .1300e + 03 .1800e + 06 3370c + 00

.8500e + 02 .2700e + 030 0. O. O. O.

.53% + 02 .8570e + 03 .5970e + 05 .1070c + 00

.5400e + 02 .5400e + 020 0. O. O. O. 1

.1220e + 03 .1330c + 03 3560c + 06 .6360c + 00

.9700c + O2 .1000e + 030 0. O. O. O.

.8370c + 02 3800c + 03 .1040e + 06 .1870c + 00

.2200e + 02 .2200c + 020 0. O. O. O.

.5200e + 02 .5210e + 02 - 3600e + 06 .6720e + 00

,1000e + 03 .1000e + 030 0. O. O, 0.

.1720e + 03 3340c + 03 .24% + 06 .48% + 00 l

.W)0e + 00 .6900e + 000 0. O. O. O.

.1040c + 02 .1460c + 03 .1170e + 05 .1260e.01

.1200e + 01 .1200c + 010 0. O. O. O.

.2570c + 02 .5680c + 02 3770e + 05 .6920c.01

.5700c + 03 .4100c + 05 .5000c + 040 0, 0. O.

.2200e + 03 .4610e + 04 .2420e + 06 ,4370e + 00

.6700c + 03 .5800c + 040 0. O. O. O.

l

.2920e + 03 .5160c + 04 3220e + 06 .6000e + 00 I

3600e +03 3000c + 05 .6000e + 040 0. O. O.

l

.7540c + 02 .1610c + 04 .83% + 05 .14% + 00

.3200c + 03 .1400c + 040 0. O. O. O.

.5060c + 02 ' .4210e + 04 3430c + 05 .5500e.01 3900e + 03 .5800e + 030 0. O. O. O.

.2800c + 03 .2050e + 04 32W + 06 .6470e + 00

.2200e + O2 .1800e + 030 0. O. O. O.

.1110e+02 .2180c+03 .1230e +05 .21 % .01

.5400c + 02 .7800e + 020 0. O. O. O.

.4470c + 02 .2850c + 03 .5310c + 05 .8640e.01

.2700e + O2 .1500e + 04 .19uve + 040 0. O. O.

.6910c + 01 .1480c + 03 3570e +04 .5440c.02

.1200e + 01 - ,1300e + 020 - 0.

(' O. O. O.

L 0. O. O. O.

I' 3900c + 02 .1400e + 03 .1000c + 020 0. O. O. )

l .2250e + 02 3870c + 03 .24% + 05 3980c.01 l e)

.1000c + 010. I

.2900e+02 .4700e +02 .1000e + 01 .1000e + 010.

! Amendment 15 M.38 L ,

l

y ;

23A6100AD Standard Plant nev ^

O Table 2A 16 (Cont'd)

N.)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326 pg 4 2)

.2710c +02 .2530c+03 3120e+05 3840c.01

.1900e + 03 .1900e + 06 .4610e + 07 .4100e + 07 3100e + 07 3000e + 07 .2000e + 07

.2380c + 01 .5000e + 02 .2610e + 04 4220e .1800e + 03 .1800e + 06 .4520e + 07 .4300e + 07 .4000e + 07 3000e + 07 3000e + 07

.9040c + 00 .1900e + 02 .9890c + 03 .1650e-03

.1800c + 03 .1800c + 06 .4520e + 07 .4400c + 07 3900e + 07 3000e + 07 3000e + 07

.2040c + 01 .4290e + 02 .2230c + 04 3620c-03

.1800e 01 .1600e + 03 - 3080c + 05 .6300e + 05 .7600e + 05 .7000e + 05 .6000e + 05

.8010c05 3530c-02 .1760e-01 .6660c.09

.2200e + 03 .2000e + 06 .4800e + 07 .4600e + 07 3400e + 07 .4000e + 07 .2000e + 07

.1100e + 02 .2300c + 03 .1200e + 05 .7280c.02

.2300e + 03 ' .9200c + 05 .7800e + 05 .2000e + 05 .1000e + 05 .1000e + 05 .1000e + 05

.2170e + 01 .4490c + 02 .2380c + 04 .4120c-03

.2200e + 03 .2100e + 06 . 4190e + 07 .2800e + 07 .1600e + 07 .9000e + 06 3000e + 06

.2330e + 01 .4890e + 02 .2550e+04 .2020e 02 w body

.2800c + 03 .4100e + 04 .1000c + 030 0, 0, 0,

.1100e + 03 .2240c + 04 .1200e + 06 .2160e + 00

.6800e + 033000e + 05 .4900e + 05 .2000e + 04 .1000e + 040 0.

.2820e + 03 .5880c + 04 3070e+ 06 .6000e + 00

/ - 3100e + 00 .3100c + 000 0. O. O. O.

I

( .2420e + 00 .5050e + 01 3660e+ 03 .4750e 03

.2600e + 00 .2600e + 000 0. O. O. O.

.1040c+02 1470c+02 .1980c+05 3640c 01

.1000e + 01 .1000c + 010 0. O. O. O.

.1810e + 02 .1830c + 02 .8760e + 05 .1810e + 00

.2300e + 01 .2300c + 010 0. O. O. O.

.1170c + 03 - .1350e + 03 .2180c + 06 .4670e + 00

.6000e+03 .6600e+040 0. O. O. O.

.9890c + 01 .1850c + 03 1090e + 05 .2070c-01

.4200e + 03 .4100e + 040 0. O. O. O.

O. O. O. O.

3100c + 03 2800e + 05 .1120e + 06 .5000e + 05 .2000e + 05 .2000e + 05 .1000e + 05 0' O. O. O.

.2200e + 03 ' 3100e + 030 0. - 0. O. O.

.8730c + 02 .2050e+ 03 .9110e + 05 .1690e + 00

.4100e + 03 .7800e + 030 ~ 0. O. O. O.

O. O. O. - 0, 3400c+03 .5600e+040 0. O. 0, 0,

.2940e + 00 ~.5910e + 01 3220e+03 .6250e 03

.2700e+03 .5500e+04 .1000e + 030 0. O. O.

.8180e + 02 .1770e + 04 .9020e + 05 .1620e + 00

.4600e + 03 .5200e + 030 0. O. O. O.

.1390c + 03 .5380e + 03 .2240e+05 .4220e-01

.2300e+03 .1900e+040 0. O. O. O.

.8340e + 02 .1640e + 04 .9200e + 05 .16600 + 00 O

,/

.2400c + 03 .4200e + 030 0. O. O. O.

2A 39 Amendment 15

l +

l ABWR usaman Standard Plant nev a l

l l

\

Table 2A 16 (Cont'd) 1 FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.2380e + 02 3250e + 03 .2230e+05 3640c-01

.9800e + 01 .9800e + 010 0. O. O. O.

.9700e + 01 .1620e + O2 .1640c+05 3060e-01

.1900e + 03 .1900e + 040 0. O. O. O.

.5860c + 02 .1160c + 04 .6450c + 05 .1110e + 00

.6300e + 02 .6600e + 020 0, 0. O. O.

l l .5190e + 02 .7940c + 02 .9900e + 05 .1790e + 00 l .8300e + 03 .4000e + 05 .2200e + 050 0. O. O.

.2180c + 02 .4560e + 03 .2390c + 05 .4310c-01

.6900e + 02 .9600e + 020 0. O. O. O,

.8500e + 01 .5670c + 02 .9990c+ M .1820c.01 3300c + 023400e + 020 0. O. O. O.

3660e + 00 .8130c + 00 .5280e + 03 .9360c 03

.1400e + 03 .2300e + 04 .1000e + 030 0. O. O.

i .2530c + 01 .5840c + 02 .2640c + 04 .1100c-02

.9800e + 01 .9800c + 010 0. O. O. 0

.1970c + 01 .1980e + 01 .1020e + 05 .1470c 01

.3400c + 03 3000e + 040 0. O. O. O.

.1140e + 02 .2460c + 03 .7310c + M .7830e 02 3500c + 03 .5500e +030 0, 0. O. O.

.1500e + 03 .9600c + 03 .1700e + 06 3140c + 00 l .7000e + 03 .1500c + 040 0. O. O. O.

.1810e + 03 3080c + 04 .2520e + 05 .4750c.01 3600e +03 .7900c + 030 0. 0, 0. O.

.7510e + 02 .9200e + 03 .8500e + 05 .1510e + 00

. .1000c + 03 .1100c + 030_ 0. O. O. O.

l .7460e + 02 .1040c + 03 .1450c + 06 .2680e + 00

.1500e + 03 6000e + 030 0. O. O. O.

.4410e + 02 .7060c + 03 .4910e + 05 .8720c-01

.7000c + 02 .7000e + 020 0. O. O. O.

.9800c + 02 .1070c + 03 .2860e + 06 .5110e + 00

.1800e + 03 .2000c + 030 0. O. O. O.

.6870e +02 3110e+03 .8500e + 05 _ .1540c + 00 3000e + 02 = 3000e + 020 0. O. O. O.

.4120e + 02 .4140c + 02 .2860e + 06 .5330e + 00 l .1500e + 03 .1500e + 030 0, 0. O. O.

.1490e + 03 .2850c + 03 .2150e + 06 .4190e + 00

.6600c+00 .7000e+000 0. O. O. O.

l .7500e + 01 .1050c + 03 .8410e + M .9060e 02

.1200c + 01 .1200e + 010 0. O. O. O.

.2120e + 02 ',4620e + 02 3090e +05 .5670e-01

.6100e + 03 .4100c + 05 .6000c+ N0 0. O. O, l .1760e + 033690c + N .1930e + 06 3500e+00

'~

.7100c + 03 .5900c + NO 0. O. O. O.

.2320c + 03 .4100e + 04 .2570c + 06 .4780e + 00

.4000c + 03 3000e + 05 .6000c + 040 0. O. O.

.6170c + 02 .1310e + 04 .6830e + 05 .1220e + 00 Amendment 15 2A-40 O

l.

ABWR . 2mioaro Standard Plant Rev A i

]u Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR.2326, pg 4 2)

.4400c + 03 .1900c + 040 0. O. O. O.

.4230c + 02 3650c + 04 .2770e + 05 .4440c-01

.6200e + 03 .9200e + 030 0. O. O. O.

.2450c+03 .1800e + 04 .2870c + 06 .5670c + 00 .

.1200e + 03 .1100e + 040 0. O. O. O.

.9300c + 01 .1820c + 03 .1020c + 05 .1830c 01

.2100e + 03 3400e + 030 0, 0. O. O.

3510c + 02 .2240e + 03 .4160c+05 .6810c-01

.7200e + 03 3200e + 05 .1500e + 050 0. O. O.

. 5630e + 01 .1200e + 03 .2810e + 04 .4310c-02

.1600e + 03 .8200e + 030 0. O. O. O.

O. O. O. O.

.1800e + 03 . 7600e + 03 3000e + 020 0. O. O.

.1770c + 02 3050c + 03 .1960c + 05 .3140e 01

.1400e + 03 .2400e + 030 0. .1000e + 020 0,

.2170e + 02 .2020c + 03 .24W + 05 3080c-01

.2000e + 05 .2200e + 07 .1580c + 08 .1700e + 08 .1500e + 08 .1200e + OS .1100e + 08

.2970c + 00 .6200e + 01 3250e+03 .5250e 04

.1800c + 05 .2000c + 07 .1500c + 08 .1800c + 08 .1700c + 08 .1600e + 08 .1400c + 08

.1260c + 00. 2630c + 01 .1370e + 03 .2300c.04

.1800e + 05 .2100e + 07 .1590e + 08 .1800c + 08 .1600e + 08 .1600e + 08 .1500e + 08

ig)

L

.2600c + 00 .5470e + 01 .2860c + 03 4640e-04

.2900e + 01 .1200c + 04 .9880c + 05 .28@c + 06 3700e + 06 3500e + 06 .4000e + 06

.4960c-05 .2210c 02 .1090c-01 .4170e-09

.2000e + 05 .2200e + 07 .1680e + 08 .1900e + 08 .1700c + 08 .1600e + 08 .1500e + 08

.6900e + 01 .1430e + 03 .7530e + 04 .4560c 02

.2200e + 05 1300e+ 07 3000e + 06 .1000c + 06 .1000e + 060. .1000e + 06

.2640e + 00 .5460e + 01 .2890c+03 .5000e-04

.2100e + 05 .2300e + 07 .1470e + 08 .1100e + 08 .8000e + 07 .4000e + 07 . 2000e + 07

.1650e+01. 3460c+O2 ,1800e+04 ~ .1420e 02 o testes

.9100e + 02 3900e + 03 .1000e + 020 0. O. O.

.1030c + 03 .2090e + 04 .1130c + 06 .2010e + 00

. 2000e + 03 .2100e + 04 .4000e + N . 2000e + 03 .1000e + 030 0.

.2300e + 03 .4800e + 04 .2510e + 06 ~ .4890c + 00

.1800e + 00 .1800e + 000 0. O. O. O.

.2690e + 00 .5660c + 01 2960c + 03 .5280e 03

.2000e + 00 .2000e + 000 0. O. O. O.

.1350c + 02 .1910e + 02 .2590c + 05 .4780c-01 e .9900e + 00 .9900e + 000 0. O. O. O.

.1320e+02 .1340e+ 02 .6450c + 05- .1320e + 00

.2100e + 01 .2100e + 010 0. O. O. O.

.7000e + 02 .8100e + 02 .1320e + 06 .2830e + 00

.5000e + 03 .6500e + N0 0. O. O. O.

.8990e + 01 .1660e + 03 .9810e + N .1870e 01

- .2600e + 03 .1500e + NO 0. O. O. O.

O. O. O. O.

Amendment 15 2A-11

4 ABWR 234612 40 Standard Plant RM. A Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326, pg 4 2)

.1800e + 03 .5900e + 04 .2100e + 04 .2000e + 030 0. O.

~0. O. O. O.

.1300e + 03 .2000e + 030 0. O. O. O. !

.8790c+02 .2070c+03 .8840e + 05 .1640e + 00

.8200c + 01 .2100c + 020 0. O. O. O.

O. 0. O. O.

.7100e + 01 3800e + 03 .1000e + 020 0. O. 0,

.2210c + 00 .4440c + 01 .2420c + 03 .4670c-03

.6900c+02 3300e+030 0, 0. O. O.

.8130e + 02 .1760e + 04 .8960e + 05 .1610e + 00

.5500e + 02 .6400c + 020 0. O. O. O.

.1500e + 03 .58% + 03 .2N0e+05 3860e-01

.7100c + 02 .2500c + 030 0. O. O. O,

.75% + 02 .14% + 04 .8330e + 05 .1520e + 00

.2500e + 02 .4200e + 020 0. O. O. O.

.24% + 023320e + 03 .23%+05 3920e-01

.5300e + 01 .5300e + 010 -0. O. O. O.

. 9110e + 01 .1510c + 02 .1520e+05 .2840c-01 _.

.5100e + 02 .2400e + 030 0. O. O. O.

.6530c + 02 .1300e + 04 .7100e + 05 .1240e + 00

.5000e + 01 .5500e + 010 0. O. O. O.

.5720e + 02 .8910e +02 .1090e + 06 .1970e + 00 '

.4700e + 02 .2300e + 04 .1900e + 040 0, O. O.

.2380e+02 .4970e+03 .2610c+05 .46%-01 l .6300e + 01 .9800c + 010 - 0. O. O. O.

L .1110e + 02 .7370e + 02 .1320e + 05 .2360c-01

.2800e + 01 .2900e +010 0. O. O. O. !

,4100e+00 .9140e+00 .5950c+03 .1050c 02

.1600e + 02 .2000e + 030 0. O. O. O.

L 3420e + Ol' .7770c + 02 3600e+N .1500e-02

.5800e + 00 = .5800e + 000 0. O. O. O.

.2150e+01 2170e+01 .1130e + 05 .1610c-01

'.4100e+02 3300e+030 ' O. O. O. O.

.1260c + 02 .2740e + 03 .8270c+04 .8860c-02 j

.8300e 62 .1200e+030 0. O. O. O.

.1360c+ J3 .8860e+03 .1530c+06 .2850e + 00 3100e + 03 .5900c + 030 - 0, O. O. O.

.1840c + 03 ' 3080e + 04 3250c +05 .6140c-01

.8900e+02 .1700e+030 0, 0, 0. O, L .8070c + 02 .9950e + 03 .9090e + 05 .1630e + 00

.2200e + 02 .2200e + 020 0. O. O. O.

.6760c + 02 .9570e + 02 ' .1320e + 06 .2470e + 00

- .5000e + 02 .8600e + 020 - 0, 0. O. O.

.4940e + 02 .7850c + 03 .54% + 05 .9750c-01 '

3700e+02 3700e+020 0. O. O. O.

tram +02 .1050e+03 .2800e + 06 .5030e + 00

.7000e+02 .7200e+020 0. O. O. O.

Amendment 15 2A.42 O

ABM 2346i004o Standard Plant nev x Table 2A 16 (Cont'd)

Q' &

j.

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326 pg 4 2)

L

.7150c + O2 3220e + 03 .8880c + 05 .1620e + 00

.1400e + O2 .1400e + 020 0. O. O. O.

3750c + 023760c + 02 .2600e + 06 .4830e + 00

.6600c + O2 ,6600c + 020 0. O. O. O.

.1250c + 03 .2520e + 03 .1730c + 06 3390c + 00 3800e + 00 3800e + 000 0. O. O. O.

.7620e + 01 .1080c + 03 .8520e + 04 .9280c 02

.9000e + 00 .9000e + 000 0. O. O. O.

.2740c + 02 .6040c + 02 .4000e+05 .7360e-01

.6100e + 03 .5000e + 05 .6000e + 040 0. O. O.

.1760e + 03. 3670e + 04 .1930c + 06 3500e + 00

.6700e + 03 .6600e + 040 0. O. O. O.

.2160e + 03 3820e + 04 .2380e + 06 .4440e + 00 3700e + 03 .3400e + 05 .6000e + 040 0. O. O.

.6850c + 02 .1460c + 04 .7630c + 05 .1350c + 00

.2400e + 03 .1200e + 040 0. O. O. O.

.4250c + O2 3110e + 04 3150e+05 .5060e-01

.1200e + 03 .1900e + 030 0, O. O. O.

.1990e + 03 .1460c + 04 .2330c + 06 .4580c + 00

.6500e + 01 .1700c + 020 0, 0. O. O.

.8860c + 01 .1730e + 03 .9670c + 04 .1740c-01

( .1600e + 02 .2300e + 020 0. O. O. O.

U .4070c + 02 .2590c i 03 .4840c+05 .7890c.01

.5500e + 01 .1100e + 03 .1600e + 030 0. O. O.

,4790e + 01 .1030e + 03 .2730c + N .4170c 02

.9300c-01' .1000c + 010 0. O. O. O.

O. O. O. O.

.1200e + O2 .2600c + 020 0. O. 0. ' O.

.1940e+02 3340e+03 .2150e+05 3420c 01

.8500e + 01 .1400e + 020 _. O. .1000e + 010 0.

.2520e+O2 2350e+03 .2900e+05 3580c-01 : 4

~.6300e + 02 .2100e + 05 - .8090e + 06 1170c + 07 .1100c + 07 .9000e + 06 . 8000e + 06

'.2360c + 00 .4970c + 01 .2600e + 03 .4190c-04

.5900e + 02 .2000: + 05 .7900e + 06 .1290c + 07 .1200e + 07 : .1100e + 07 .1100e+ 07 '

.1060e + 00 ~ .2230c + 01 .1160e + 03 .1940e 04

.5900e + 02 .2000e + 05 ,8000e+06 .1280e+07 .1200e + 07 .1200e + 07 .1000e + 07 4

~'

.2080c + 00 ' .4380e + 01 .2290c + 03 3700e 04

.6200c 02 1900e + 02 - .6280e + 04 .1970e + 05 .2700e + 05 .2900e + 05 .2800e + 05 -

.4380e-05 .1940c-02 .%10e 02 3650e-09

.7000e + 02 .2300e + 05 .837C e 06 .1320e + 07 .13( Oc + 07 .1200e + 07 .1000e + 07

.6000e + 01 .1260c + 03 .6610e+04 .3veuc 02

.7500c + 02 ,1200c + 05 .1000e + 05 .4000c + 04 .40C )e + N .4000e + 04 3000e + 04

.2100c+00 .4350c+01 .2300c + 03 3990e-04 i

.7100e + 02 .2400e + 05 .7260e + 06 .8500e + 06 .50C Oc + 06 3000e +06 3000e + 06

.1470c + 01 3080e + 02 .1610c + 04 .1270c 02 e

b Amendment 15 2A 43

r

.M 3M6100AD Standard Plant ne a Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

OYafieS

.9700e + 03 .2300e + 040 0. O. O. O.

.6370c + O2 .1290e + 04 .7010e + 05 .1220e + 00 i L-l2

.2100e + 04 .8500e + 04 .8500e + 04 .1000e + N0 0. O.

.2240c + 03 .4700e + 04 .2440c + 06 .4780e + 00 l: .1800c + 00 ,1800e + 000 0. O. O. O.

i

.8900e .1860e + 01 .9720e + 02 .1720c-03

.2300e + 00 .2300e + 000 0, 0. O. 0.'

.5700e + 01 .8020c + 01 .1080e + 05 .1980c-01

.1000c + 01 .1000e + 010 0. O. O. O.

.14% + 02 .1510e + 02 .7220e + 05 .14% + 00

.2300e +01 .2300e + 010 0. 0, 0, 0,

.1140e + 03 .1340c + 03 .2160c + 06 4640c +00

.5000e + 03 .6500c + 040 0. O. O. O.

.6090c + Ol' .1130e + 03 .6660e + 04 .1260e.01

.2600e + 03 .1500e + 040 0. O. O. O.

! 0. 0, 0. O.

ll .1800e + 03 .5900e + 04 .2100e + 04 .2000e + 030 0. O.

!: 0.. O. O. O.

.2300e + 033000e + 0.10 0. O. O. O.

4460e + 02 .1020c + 03 .5110e + 05 .9440c 01 i

.8200e + 01 .2100e + 020 0. O. O. O L

0. 0; 0. O.

.9500e + 01 3900e + 03 .1000e + 020 0. O. O.

.2740c + 00 .5480e + 01 3000e + 03 .5830e-03

.7800c + 03 .2300e + 040 0. - 0. O. O.

.4120e + 02 .9000e + 03 .4530e+05 .8170e 01 l ' 6100e+03 .6900e+030

. 0.- 0. O. O.

.5600c+02 . 2180e +03 .14W+05 .2780e-01 11 .7800e + 03 .1600e + NO : 0. O. - 0. O.

.5090c + 02' ,1010e + 04 .5620e + 05 .1020e + 00 L .2100e + 03 3400e + 030- 0. O. O. O.

1

.1030c + O2 .1410e + 03 .9720e + N .1580c-01 l- .8600e + 01' .8600c + 010 0. O. O. 0. -

L .4150c + 01 .6800e + 01 .6920c + 04 .12 h 01

.5300e + 03 .1200e + 040 0, 0. O. O.

.2150c + 02. 4200c + 03 .2350e + 05 .4080e-01

.5600e + 02 .5900e + 020 0.- 0. O. O.

'.2100e + 02 3280e + 02 3940e + 05 .71 % 01

,.2400e+ 03 .2800e + 04 .2000e+ NO 0. O. O.

- 8500c + 01 .1770e + 03 .9290e + N .1670e-01

.5900e+02 .7900e+020 0. O. O. O.

2.4640e + 01 3080e + 02 .5510e + N .9860c-02 3400c+01 3500c + 010 0. O. O. O.

.1380c + 00 3120e + 00 .2010e + 03 3530e 03 SJ00e +02 .2500c +030 0. O. O. O.

.9100c+00 .2100e+02 .9460c +03 3920e 03 O

Amendment 15 2A44

l 1

MM 23A6100AD Standard Plant Rev A Table 2A.16 (Cont'd) ;

O) q i FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326 pg 4 2) l O. l

.7900c + 00 .7900e + 000 0. O. O.

.8400e + 00 .8400e + 00 .4380e + 04 .61 % -02

.1000e + 03 .4600e + 030 0, 0. O. O.

.4500c + 01 .9800c + 02 .2700e + 04 .2940e 02

.5800e + 03 .7600e + 030 0. O. O. O.

.9800e + 02 .6030e + 03 .1120e + 06 .2090c + 00

.5400e + 03 .9700e + 030 0. O. O. O.

.9900e + O2 '.1660e + 04 .1380e + 05 .2580e.01

.5300e + 03 .8800e + 030 0. O. O. O.

3150e + 02 3920c + 03 35% + 05 .6440e-01

.7800e + 02 .7900e + 020 0. O. O. O.

.4570c + 02 .6450e + 02 .8940c + 05 .1670e + 00

.6300e + O2 .1000e + 030 0. O. O. O.

.1640c + 02 .2640c + 03 .1840e + 05 3250e 01

.4300c + 02 .4300e + 020 0, 0, 0. O.

.5340e + 02 .5860e + 02 .1550c + 06 .2780c + 00

.7/00e + O2 .8000e + 020 0. O. O. O.

3230e + O2 .1440c + 03 '.4040c + 05 .7250c-01

~.1700e + O2 .1700e + 020 0. . O. O. O.

.2560c+02 .2570e +02 .1780c + 06 3310e + 00

.8300e + 02 .8300e + 020 0, 0, 0, 0.

p$

.1160c + 03 .2170e + 03 .1770e + 06 3470c+ 00 0.

.4200c +00 ' 4300e+000 O. O. O.

3080c + 01 .4360c +02 3450e+ 04 3720c 02

.1000c + 01 .1000e + 010 0. O. O. O.

l.

.1130e + 022480c + 02 .1650e + 05 3000c 01 l! .5800e + 03 .4400e + 05 .5000e + 040 0. O. O.

j

.8900c + 02 .1870e + 04 .9810e+05 .1750e+00

.6500c + 03 .5900e + 040 ' O. O. O. O. ;

.14% + 03 .2640e + 04 .1650e + 06 3060e + 00 l1 3600c +03 ' 3100c + 05 6000e + 040 0.~ 0. O.

.2270c + 02 .4800e + 03 .2470e + 05 .4440c-01

.6000c+03 .1900c +040 - 0.- 0. O. O.

.2400e + 02 .2830e + 04 .1080e + 05 .1750c-01

.1100e + 04 .1600c + 040 0. O. O. O.

l .2030c + 03 .14% + 04 . 2370c + 06 .46% + 00

.1100e+03 .2000e+030 0. O. O. 0..

3920e + 01 .7700e + 02 .4300e + 04 .7670e.02

.2300e+03 3000e+030 0. O. O. O.

.1800c + 02 ' .1150e + 03 .2140c+05 3500c-01

.6000c+02 2400c+03 .1800c+030 ' O. O. O.

3540e+01 7610e+02 .1170c + 04 .1790c.02

.9300c.01 .1000e + 010 0. O. O. O.

O. 0' O. 0. '

.1800c + 03. 3100c + 030 0. O. O. O.

.7100e + 01 .1200e + 03 .7710e + 04 .1250e-01

.1300e + 03 .1900e + 030 - 0. O. O. O.

Amendment 15 2A 45

1

~

ABWR uts m o Standard Plant hA Table 2A 16 (Cont'd)

FILE 20 DOSE CONVERSION FILE (NUREG/CR 2326. pg 4 2)

.1080c + 02 .1000e + 03 .1240c + 05 .1540c-01

.2200e + 02 6800e + 04 .2630c + 06 3800e + 06 3400e +06 3100e + 06 .2000e + 06 3320e 01 .7020e + 00 3640c + O2 .5920e-05

.1900c + 02 .6400e + 04 .2540c + 06 .4000e + 06 3400e + 06 4000e + 06 3000e + 06

.2020c.01 .4250e + 00 .2220e + 02 3690e-05

.2100c + 02 .6400e + 04 .2540e + 06 .4100c + 06 .4300c + 06 3000e + 06 .4000c + 06 3220c 01 .6770e + 00 3510e + 02 .5720e-05

.2200c-02 i(100e + 01 .1990c + 04 .6400c + 04 .8600e + 04 .9000e + 04 .9000e+ N

.2970c 05 .1300c-02 .6460c .2460c 09

.7300e + 02 .7400e + 04 .2730c + 06 4300e + 06 3900e + 06 .4000e + 06 3000e + 06

.4060c + 01 .8520e + O2 .4450c + 04 .2690c-02

.2600e+02 3700c+04 3200e+ 04 .1500e+ N .1300e + 04 .1300e + 04 .1000e + 04

.2690e 01 .5590e + 00 .2950e + 02 .5110c-05

.2500e + 02 .7500e + 04 .2320e + 06 .2600e + 06 .1700e + 06 .1100e + 06 .6000e + 05

.8900e + 00 .1870c + 02 .9720e + 03 .7720c-03 Table 2A 17 FILE 27 METEOROLGICAL DATA (NUREG/CR 2326, pg 4-6)

To be supplied by utility for specific site.

-O_,

O Amendment 15 2A.46

ABWR mwn Slandard Plant av n b

U 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS (as discussed in Section 3.7) in combination with other appropriate loads.

ABWR Standard Plant structures, systems and The seismic classifications indicated b components are categorized as nuclear safety- Table 3.21 rneet the requiremrats of Regulatory related or non nuclear safety related (see Table Guide 1.29 except as otherwise noted in the 3.21). The safety related structures, systems table.

and components, perform nuclear safety related functions as defined here, and are classified in 3.2.2 Quality Group Classifications accordance with Subsection 3.2.3. In addition, specific design requirements are identified for Quality group classifications as defined in the safety related equipment commensurate with NRC Regulatory Guide 1.26 are shown in Table '

their safety classification (see Table 3.2 2 and 3.21 for all components under the heading, 3.23). ' Quality Group Classification *, Although not within the scope of Regulatory Guide 1.26 A safety related function is a direct or definitions, component supports, core support support function that is necessary to assure: structures and primary containment boundary that are within the scope of ASME Code,Section III,

1. the integrity of the reactor coolant are assigned per Tables 3.2 2 and 3.2 3, a pressure boundary; or quality group classification as identified in l Table 3.21,

2. the capability to shut down the reactor and maintain it in a safe condition; or Qualit:' group classifications and design and fabrication requirements defined in Regulatory-

3. the capability to prevent or mitigate the Guide 1.16 are indicated in Tables 3.21 and (

jm consequences of accidents which could result 3.2 3, re spectively. Figure 6.2 38 depicts I g

) in potential offsite exposures comparable to quality group classifications of the components

" in major systems.

the guidelines exposures of 10CFR100, l 3.2.1 Scismic Classification 3.2.3 SLfety Classifications l l l ABWR Standard Plant structures, systems, and Safety related structures, systems, and

- components, including their foundations and components of the ABWR Standard Plant are supports, that must remain functional in the classified for design requirements as Safety

! event of a safe shutdown earthquake (SSE) are Class 1, Safe,ty Class 2, or Safety Class 3 in designated as Seismic Category 1. accordance with their nuclear safety importance, These safety classificatiou are All safety related ABWR Standard Piant identified on Table 3.21 for poncipal l structures, components, equipment, and systems structures, systems, and components. Components are classified as Seismic Category I, except within a system are assigned different safety f those (e.g., pipe whip restraints), as noted on classes depending upon their differing safety L

l Table 3.21, which need not function during but importance; a system may thus have componerit:, in l n shall remain functic nal after the event of an uore than one safety class. Safety SSE. Also some nomafety related structures, classification for supports within the scope of N

l systems, and compo' ents are classified as Seismic ASME Code,Section III, depends ',pon that of the Category I as noted en Table 3.21, supoorted component. j The Seismic Category I structures, systems and Tle definitions of the safety classes in this components are designed to withstand,.wlthout section are based on Section 3.3 of ANS Standard loss of function, the appropriate seismic loads 52.1, and examples of their broad applica- l l

l } %

i; l 1 M l

Arnendment 15 W

1 ABM 2346ioaan Standard Plant REV. B tion are given. Decause of specific design con. (4) Ensure emergency core cooling where the siderations, these general definitions are equipment provides coolant directly to the subject to interpretation and exceptions. Table core (e.g., emergency core cooling systems);

3.21 identifies component classifications on a i component by component basis. (5) Provide or maintain sufficient reactor l

coolant inventory for emergency core cooling Minimum design requirements for various (e.g., suppression pool),

safety related classes are delineated in Tables 3.2 2 and 3.2 3. Where possible, reference is Safety Class 2 includes the pressure-l made to accepted industry codes and standards retaining partions of the following:

I which define design requirements commensurate l with the safety related function (s) to be (1) Those components of the control rod system performed. In cases where industry codes and which are necessary for emergency negative standards have no specific design requirements, reactivity insertion; l ti.e sections that summarize the requirements to be implemented in the design are indicated. (2) Emergency core cooling systems; 3.23.1 Safety Class 1 (3) Primary containment vessel; l Safety Class 1 (SC 1) applies to all compo- (4) Post accident containment heat removal nents of the reactor coolant pressure boundary systems; (as defined in 10CFR50.2), and their supports, whose failure could cause a loss of reactor (5) Pipes having a nominal pipe size of one inch coolant at a rate in excess of the normal makeup. or smaller that are part of t'ac reactor system, and which are within the scope of the coolant pressure boundary, ASME Code, Section 111.

Safety Class 2 structures, systems, and compo-l Safety Class 1 components are identified in nents are identified in Table 3.21.

l Table 3.21, 3.233 Safety Class 3 3.23.2 Safety Class 2 Safety Class 3, (SC-3) applies to those struc-

. Safety Class 2 (SC 2) applies to pressure- tures, systems, and components, not included in-retaining portions, and their supports, of pri- SC 1 or 2, that are designed and relied upon to mary containment and to other mechanical equip. accomplish the following nuclear safety-related ment, requirements for which are within the scope functions:

of the ASME Code,Section III, that are not included in SC 1 and are designed and relied upon (1) Provide for functions defined in SC 1 or 2 to accomplish the following nuclear safety- by means of equipment, or portions thereof, related functions: that is not within the scope of the ASME Code,Section III; (1) Provide primary containment radio ctive material holdup or isolation; (2) Provide secondaiv conRinmera radioactive material holdup, isolation, or heat removal; (2) Provide emergency heat removal for the primary containment atmosphere to an (3) Except foi primaq containment boundary ex-intermediate heat sink, or emergency removal M.aion functions, ensure hydrogen concentra-

.of radioactive material from the primary tion control of the primary containment atmo-containment atmosphere; sphere to acceptable limits; (b Introduce emergency negative react vity i to (4) Rernove radioactive material from the atmos-make the reactor suberitical; phere of confined spaces outside primary Amendment 3 3.22 O

- ABWR 23^am^u RPV. D Standard Plan!

TABLE 3 21 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca- Classi- Assurance Seismic Princloal Componenta ggg b gc Deationd Recuiremente Catecorvf b'.olts BI Reactor Pressure Vessel System /

Fuel Assemblies 8-

1. Reactor vessel 1 C A B 1 g

2. Reactor vessel support skirt 1 C A B 1 and stabilizer

3. Reactor vessel appurtenances 1 C A B 1 (g) pressure retaining portions

4. Supports for CRD housing, 1 C A B 1 in. core housing and recircu.

lation internal pump ,

5. Reactor internal structures - 2 C B B 1 feedwater, RHR/ECCS high pressure core Gooder spargers

6. Reactor internal structures- 3 C -- B 1 safety related components including core support structures (See Subsection 3.9.5)

~ 7. Reactor internal structures - N C -- - --

non safety related components (See Subsection 3.9.5)

8. Control rods 3 C - B 1

9. - Power range detector hardware 3 C - - - P I including startup range detector

10. Fuel assemblies 3 C -- B 1

11. Reactor Internal Pump 1 C A B 1 Motor Casing B2 Nuclear Boller System

1. Vessels levelinstrumenta- 2 C B B I tion' condensing chambers O

Amendment 15 3.28

1

'ABWR ma.

Standard Plant mn TABLE 3.21 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca. Classl. Assurance Seismic Princinal Comnonent# fjgssb ge ggd - Reaulrement' Catenorvf b'atts

2. Vessel air accumulators 3 C C B I (for ADS and SRVs)

3. Pipingincluding supports - 2/3 C B/C B 1 (b) g safety / relief valve discharge g i

r l

O e

L s

I O

Amendment 11 3gg

- ABWR 2w-c Standard Plant nity n p)

TABLE 3.21 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca. Classi- Assurance Seismic Princloal Comnonenta cin33b ge [1cationd Reautrement' Catecorv fNotes B2 Nuclear Holler System (Continued)

4. Pipingincluding supports 1 C,SC A B 1 main steamline (MSL) and feed-water (FW)line up to and in-cluding the outermost isolation valve

5. Piping includir.g supports- 2 SC B B 1 h MSL from outermost M isolation valve to and including seismic interface restraint and FW from outermost isolation valve to the seismic interface restraint including the shutoff valve and the restraint O-V

6. Pipingincluding supports MSL from the scismic interface N SC,T B - --

(r) gg AA restraint to the turbine stop vahe

7. Deleted I !

j 8. Piping FW beyond the seismic N SC,T D ,- - -!

J interface restraint l 9. Deleted

10. Pipe whip restraints MSL/FW 3 SC,C - B --

[

l- g 11. Pipingincluding supports other M within outermost isolation valves .

a. RPV head vent 1 C A B I (g)

b. RPV head spray . 1 C A B I (g)

c. Main steam drains 1 C,SC A B i (g) !

g 12. Piping including supports-other g beyond outermost isolation valves

a. RPV head vent N C D -- -

b. RPV head spray N SC D -- --

c. Main steam drains N SC D

(%

Amendment 15 3.29 1

LABM 2346ioore REY.B

- Standard Plant p).

w TABLE 3 21 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety loca. Classl. Assurance - Seirmic Principal Comnonenta gggg b ge ficationd Requirement' CatenoryI Ep.its B2 Nuclear Boller System (Continued)

13. Pipingincluding supports - 2/N C.SC B/D B/-- 1/- (g) h instrumentation beyond

" outermost isolation valves

14. Safety / relief valves 1 C A B I l

l 15. Valves MSL and FW 1 C,SC A B -1 isolation valves, and other

+ FW valves within containment

16. Valves FW,other beyond 2 SC B B 1 l

l outermost isolation vlaves up to

, and including shutoff valves 17.- Valves within outermost isolation valves l

l.

1- a. RPV head vent .1 C A- B I (g)

b. RPV head spray 1 C,SC A B I (g)-

c. - Main steam drains 1 C,SC A B I (g) l l 18. Valves,other

a. RPV head vent 3 C C B I

b. RPV head spray 2 SC B B I

c. Main Steam Drain N SC D -- -

19. Valves instrumentation beyond 2/N SC- B/D B/ - 1/- (g) outermost isolation valves l i

20. Mechanicalmodules.instrumen- 3 C,SC - B I tation with safety-related function

21. Electrical modules with safety- 3 C,SC,X --- B I (i) related function ,

22. Cable with safety related 3 C,SC,X -- B 1 function O

Amendment 3 32 10

ABWR meime -

Standard Plant nev.n

- TABLE 3.2 1 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca. Classi. Assurance Seismic Princloal Comnonenta flaub ge ficationd f Reautremente Catenorv b'o,ts,3 B3 Reactor Recirculatlou System 1,- Piping Primary side, motor 3 C C B 1 (s) @ h' l coolmg system "N

2. Pipe Supports 3 C C B 1

3. Pump motor cover 2 C B B I

4. Pump non pressure retain- N C - - --

ing parts including motor, instruments, electrical cables and seals

5. Valves 3 C C B I (g)

C1 CRD System

1. Valves with no safety related 2 SC B B 1 (g) function (not part of HCU)

2. Piping including supports. 2 C,SC B B I (j)

$ insert line n

3. Piping other(pumpsuction, N -SC D --- -- (g) pump discharge, drive header)

4. Hydraulic control unit 2 SC - B 1 (k)

5. Fine motion drive motor N C -- --- --

6. CRD Drive water pumps N SC D --- --

g 7. Control Rod Drive 1/3 C A /--- B I

8. Electrical modules with 3 C,SC -- B 1 safety function

9. Cable with safety related 3 C,SC,X B 1 function 9

Amendment 15 3.2 11

l l

ABM 2346ioort REV B-1 Standard Plant

/j TABLE 3 21 U- 1 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca- Classi- Assurance Selsmic Prineinal Comnonenta ci,33b ge ficationd Reaulttment' CatenonI b'q1n i

HI Main Control Room Panel 1

1. Panels 3/N X ~

B/ -- 1/ - (aa)

2. Electrical Modules with 3 X - B 1 safety related function

3. Cable with safety related 3 X -- B I function

.4 Other mechanical and N X ~ ~ ~

electrical modules H2 Local Control Panels i -/ O u

C/ 1. Panels or Racks 3/N C,SC,X --

B/~ I/~ (aa) ,

i

2. Electrical modules witi- 3 C,SC,X - B I safety related funcOn

3. Cable with safety related 3 C,SC,Y, -- B I function

4. Other mechanical and- N _C,SC,X - - -

electrical modules El Radioactive Drain Transfer System L Drain pipingincluding supports N ALL D- - --

l and valves radioactive (except)

RZ,X) g 2. Drain piping including supports N ALL D - -

g and valves nonradioactive

3. Piping and valves contain. 2 C,SC B B I

, ment isolation

4. Other mechanical and N ALL - - -

electrical modules Amendment 7 3.2 21 l

~

1 l

23A6100AE Standard Plant uv n TABLE 3.21 '

L - CIASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Loca. Classi. Assurance Seismic Princloal Comnonent" Safet4 lign* fication Reautreme'nt' Catenorv I

Qasa Notes ,

K2 Radweste System

1. Piping including supports N C,SC B. B I !

and valves forming part of containment boundary

2. Pressure vessels including N W - - - (p)

3. Atmospheric tanks including N C,SC,H, - -- - (p) _3 supports T,W l

' 4. 15 PSIG Tanks and supports N- W -- -- -- (p)

,. 5. Heat exchangers and supports N C,SC,W - - -- (p) i

\

L ' 6, Pipingincluding supports 'N C,SC,H - -- -- (p) and valves T,W

. 7. Pumps including support. N- C,SC,H -- - -- (p) -

T,W 7

. N1 Power Conversion System 1

1. (Deleted)

2.' Branch line of MSL including - N SC,T B B -

(r) :

supports between the second ,

isolation valve and the turbine stop valve from branch point at ..

MSL to and including the first valve in the branch line (

3. Main feedwater line i.MFL) N SC B B. I including supports fram second

' isolation valve branci hnes and components beytod up to outboard shutoff valvea ;

, -l O

Amendment 13 3.2 21.1 l l 1

.l MM. 23A6100AE Standard Plant REY. R O TABLE 3.21 d

CIASSIFICATION

SUMMARY

(Continued)-

Quality Group Quality Loca. Classi- Assurance Seismic Safet4 Principal Component" QAIA UQn' fication Reautrement' Catenorv b' gigs

4. Turbine bypass piping - N T D - -

including supports

5. Turbine stop valve, turbine N T D - --

(1)(n)(o) bypass valves, and the main steam leads from the turbine controlvalve to the turbine casing -

6. - Feedwater system components N T D -- -

beyond outboard shutoff valve

7. Turbine generator N T - - -

l L 8. Condenser ' N T - - --

9. ' Aih ejector equipment N T- - - -

' 10. Turbine gland sealing N T D - --

l system components j h

11. Circulating water system N T D -- -

N2 Offgas System

' 1. Pressure vessels including N T. - - - - -- (p)(q) supports

2. . Atmospheric tanks including N T - - - (p)(q)-

supports

3. 0-15 psig tanks including . N .T - -- -- (p)(q) supports

4. . Heat exchangers including N - - --

(P)(q)

L supports t

5. Piping including supports N '.' ~ - -- (P)(q) and valves

6. Pumps including supports N T -- - -- (p)(q) fL)\.

Amendment 15 3.2-21.2

ABWR m 6i m e-Standard Plant anv n TABLE 3.21 g CIASSIFICATION

SUMMARY

(Continued) -

Quality Group Quality 1xca. Classi. Assurance Seismic Princloal Component" Safet4 Reautrement' Category Qan ll2n fication haln P1 Makeup Water System (Purified) i 1

1. Piping locluding supports and 2 C B B 1 valves forming part of the con-tainment boundary

2. Demineralizer water storage N O D -- --

tank including supports -

3. Demineralizer water header - 2 SC B B 1 piping including supports and valves -

4 . Piping including supports and N O D -- --

lm

! valves

5. Other components N O D - -

P2 Makeup Water System (Condensate)

1. Condensate storage tank N O D- -- --

(w) including supports 2.' ' Con'densate hea' der piping 2 SC~ B B 1

.. g including supports and vakes n

3. Piping including supports and N O D. - -

valves ;

4 Other components N O D -. - -

- l ; P3 Reactor Building Cooling Water System 1.' Piping and valves forming part '2 SC,C B B 1 (g) 3 of primary containment boundary g.

.g 2. Other safety related piping, 3 SC,C C B I including supports pumps and valves l'

t e

Amendment 12 3.2 22 j

ABM ux61oort l Standard Plant RTV H 1

TABLE 3.21 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safety Loca- Classi. Assurance Seismic Princinal Comnonerita fJau b ge ficationd Reautremente CatenorvI b'gtn -

U2 Heating, Ventilating,and Air Conditioning Systems' (Continued)

h. Other safety related 3 SC,RZ, - B I valves and dampers H,X

i. Electrical modules with 3 SC,RZ, - B 1 safety-related function H,X

j. Cable with safety related 3 SC,RZ, -- B 1 function H,X

2. Non safety related equipment' l a. ' HVAC mechanical or N SC,RZ,H, - - - - -

[ . electrical components X,SC,

\ ,

with non safety related W,T

!. . functions L:

(' U3' Fire Protection System

1. Piping including supports and 2 C B B I i valves forming part of the

. primary containment boundary

2. Other pipingincluding supports N . SC,C,X, D - --

(t) (u) and valves RZ,H <

3. Pumps - N F D - ---

(t) (u)

4. Pump motors N F .- -- --

(t) (u)

5. Electrical modules N C,SC,X, -- - -

(t) ~ (u) l' RZ,H, T,w Includes thermal and radiological erwironmental controlfunctions within the ABH'R Standard Plant scope.

" Controls erwironment in rooms or areas containing non. safety related equipment within the

. ABHR Standard Plant.

3 Amendment 7 3.2-29

ABM 2346ioore erv. n Standard Plant TABLE 3.21 CLASSIFICATION

SUMMARY

(Continued) G-Quality Group Quality Loca. Assurance Seismic Safet4 Classi. d I Principal Component" Casa llan* ncation Reautrement' Catenorv b'Olta U3 Fire Protection System (Continued)

6. CO actuation modules N RZ,T - --- -

(t) (u) 2

7. Cables - N SC,C,X, -- --- --

(u)

RZ,T,W

8. Sprinklers N SC,X D -- -

(u)

U4 CivilStructures

1. T Reactor Building (Secondary 3 SC,RZ - B I Containment and Clean Zone)

2. Control Building 3 X - B 1

3. Service Building N H --- -- --

G,.

4. Radwaste Building substructure 3 W - B I_

5. Turbine Building N T -- <- -

(v) l O

Amendment 15 3.2 30

l l

l MkN 23A6100AE Rn n Standard Plani

-(J l NOTES (Continued)

The design and construction specifications for the HCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but of the remaining parts and details. For example: (1) all welds are LP inspected; (2) all socket welds are inspected j r

for gap between pipe and socket bottom; (3) all welding is performed by qualified welders; and (4) ]

all work is done per written procedures. Quality Group D is generally applicable because the codes j and standards invoked by that group contain clauses which permit the use of manufacturer standards and proven design techniques which are not explicitly defined within the codes for Quality Groups A, ;

B, or C. This is supplemented by the OC technique described.

1. The turbine stop vahr is designed to withstand the SSE and maintain its integrity.

m. The RCIC turbine is not included in the scope of standard codes. The assure that the turbine is fabricated to the standards commensurate with safety and performance requirements, General Electric has established specific design requirements for this component which are as follows:

1. All welding shall be qualified in accordance with Section IX, AShiE Boiler and Iressure Vessel ,

Code.

2. All pressure containing castings and fabrications shall be hydrotested at 1.5 h the design ;

pressure, I

f j l

3. All high pressure castings shall be radiographed according to:

g ASThi E 94 E 141 U;

s E 142 maximum feasible volume -

i E-71,186 or 280 Severity level 3 l 4; As cast surfaces shall be magnetic-particle or liquid penetrant tested according to AShiE Code,

Section III, Paragraphs NB 2575, NC 2576, or NB 2576, and NC 2576.

5. Wheel and shaft forgings shall be ultrasonically tested according t; ASThi A 388.

6. Butt welds shall be radiographed and magnetic particle or liquid penetrant tested according to s the AS' ... Eoiler and Pressure Vessel Code Acceptance standards shall be in accordance with AShfE f Boiler and Pressure Vessel Code Section Ill, Paragraph NB 5340, NC.5340, NB 5350, or NC-5350, a L

respectively. l

7. . Notification shall be made on major repairs and records maintained thereof.

[ 8. Record system and traceability shall be according to AShfE Section III, NCA 4000.

i

9. Control and identification shall be according to AShf E Section 111, NCA 4000.

l; >

_10. Procedures shall conform to ASME Section III, NB 5100 and NC-5100.

L

11. Inspection personnel shall be qualified according to AShiE Section 111, NB 5500 and NC 5500.

l C\

Amendment 10 3.2 33 l

l 1

ABWR 2mman -

Standard Plant nrv. n - !

NOTES (Continued)

n. All cast pressure retaining parts of a size and configuration for which volumetric methods are O effective are examined by radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards is used as an alternate to radiographic methods. Examination procedures and acceptance standards are at least equivalent to those defined in Paragraph 136.4, Nonboiler External Piping, ANSI B31.1. -

o. The following qualifications are met with respect to the certification requirements:

1. The manufacturer of the turbine stop valves, turbine control valves, turbine bypass valves, and main steam leads from turbine control valve to turbine casing utilizes quality control procedures equivalent to those defined in GE Publication GEZ-4982A, General Electric Large Steam Turbine Generator Quality Control Program.

i ertification obtained from the manufacturer of these valves and steam loads demonstrates that ,

the quality control program as defined has been accomplished. !

The followy requirements shall be met in addition to the Quality Group D requirements:

1 Alllongitudinal and circumferential butt weld joints shall be radiograph:d (or ultrasonically tested to equivalent standards). Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrate examination may be substituted.

Examination procedures and acceptance standards shall be at least equivalent to those specified -

as supplementary 'ypes of examinations, Paragraph 136,4 in ANSI B31.1.

2 All fillet and socket welds shall be examined by either magnetic particle or liquid penetrate

- methods. All structural n'ttachment welds to pressure retaining materials shall be examined by 4 -

cither magnetic particle or liquid penetrate methods. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1, 3 All inspection records shall be maintained for the life of the plant. These records shall include data pertaining to qualification of inspection personnel, examination procedures, and examination results,

p. Regulatory Guide 1.143 furnishes complete design guidance relating to seismic and quality group classification and quality assurance provisions for radioactive waste management systems, structures and components,

q. Detailed scismic design criteria for the offgas system are provided in Section 11.3.

r. The main steam lines from the containment outboard isolation valves and all branch lines 21/2 inches t in diameter and larger,= up to and including the first valve (including lines and valve supports) are ~

designed by the use of an appropriate dynamic scismic system analysis to withstand the operating bases e'arthquake (OBE) and safe shutdown earthquake (SSE) design loads in combination with other appropriate loads, within the limits specified for Class 2 pipe in the ASME Section Ill. The mathematical model for the dyn'amic seismic analyses of the main steam lines and branch line piping ,

includes the turbine stop valves and piping to the turbine casing. The dynamic input loads for design of the main steam lines are derived from a time history model analysis or an equivalent method-as described in Section 3.7.

O Amendment 13 3.2 34 l

1

ABWR usamru l~

Standard Plant REV.B EDIDi (Continued)

!qI

s. The recirculation motor control system (RMCS) is classified Quality Group C and Safety Class 3 which is in accordance with the requirements of 10CFR50.55a. The XMCS, which is part of the reactor coolant pressure boundary (RCPB), meets 10CFR50.55a(c)(2). Postulated failure of the RMCS piping j cannot cause a loss of reactor coolant in excess of normal makeup (CRD return or RCIC flow), and the RMCS is not an engineered safety feature. Thus,in the event of a postulated failure of the RMCS piping during normal operation, the reactor can be shutdown and cooled down in an orderly manner, and reactor coolant makeup can be provided by a normal make up system (e.g., CRD return or RCIC system). .

Thus, per 10CFR$o.55a(c)(2), the RMCS need not be classified Quality Group A or Safety Class 1.

Since the RMCS is not an engineered safety feature (e.g., it does not provide emergency reactivit) ,

control, emergency core coolant, or primary reactor containment), the system need not be classified Quality Group B.or Safety Class 2. The RMCS is classified Quality Group C and Safety Class 3, however, the system is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code,Section III, Class 1 criteria as specified in Subsection 3.9.3.1.4 and Figure 5.414 ,

t. There is a limited quality assurance program for the Fire Protection System.

u. Special seismic qualification. ar.d quality assurance requirements are applied.

v. See Subsection 113.4,6 for the offgas vault seismic requittements.

l'

w. The condensate storage tank will be designed, fabricated, and tested to meet the intent of API ,

Standard API 650. In addition, the specification for this tank will require: (1) 100% surface !

examination of the side wall to bottom joint and (2) 100% volumetric examination of the side wall )

seld joints.

(G'-) x. The crancs are designed to hold up their loads under conditions of OBE and to maintain their -

positions over the units under conditions of SSE.

y. All off engine components are constructed to the extant possible to the ASME Code, Section Ill, Class p 3.

z. Components associated with a safety related function (e.g., isolation) are safety related.- .;

i aa. Structures which support or house safety related mechanical or electrical components are safety related, i bb.' A quality assurance requi ements shall be applied to ensure that the design, construction and testing j( ; requirements are met.

i i

f

r V

Amendment 15 3.2441

i gg 21M10GAE Standard.Elant

/~'s SECTION 3.5 V :

CONTENTS Section M .7ESE 5 3 $.1 Missile Selection and Descrintion 3.51 ,

3.5.1.1 Internally Generated Missiles (Outside Containment) 3.53 '

3.5.1.1.1 Rointing Equipment 3.5-3 l

3.5.1.1.1.1 Missile Characteritation 3.53 33.1.1.1.2 RClC Steam Turbine 15 3 3.5.1.1.13 Main Steam Turbine 3.53 3.5.1.1.1.4 Other Missile Analysis 3.53 )

i 3.5.1.1.2 Pressurized Components 3.53 l l

3.5.1.1.2.1 Missile Characterization 3.53 i p

3.5.1.1.2.2 Missile Analyses 3.53 3.5.1.13 Missile Barriers and leadings 3.55 3.5.1.2 Internally Generated Missiles (Inside Containment) 3.56 3.5.1.2.1 Rotating Equipment 3.56 3.5.1.2.2 Pressurized Componems 3.56 3.5.1.23 Missile Barriers and leadings 3.5-6 3.5.1.2.4 Evaluation of Potential Gravitational Missiles inside Containment 3.5-6 3.5.13 Turbine Missiles 3.5-6 1

3.5.1.4 Missiles Generated by Natural Phenomena 3.5-6 3.5.15 Site Proximity Missiles (Except Aircraft) 3.57 3.5.1.6 Aircraft llazards 3,57

./#%

1 3.5 li Amendment 7

ABWR.

$1godard Phnt *$37n SECTION 3.5 g

CONTENTS (Continued) l Section Ig[g ggg 3J.2 Structures. Systems, and Comnonents to be I Protected from Externally Generated Mlasites 3.57 I 3.53 Barrier Deslan Procedures 3.57 j 1

3.53.1 LocalDamage Prediction 3.57 3.53.1.1 Concrete Structcres and Barriers 3.57 3.53.1.2 Steel Structures and Barriers 3.57 l 3.53.2 Overall Damage Prediction 3.57 3.5.4 Interfaces 3.57 3.5.4.1 Protection of Ultimate Heat Sink 3.58 ,

3.5.4.2 hiissile Generated by Natural Phenomena from Remainder of Plant Structures, Systems and Components 3.58 3.5.43 Site Preximity hiissiles and Aircraft llazards 3.58 3 3.5.4.4 Secondary hiissiles inside Containment 3.58 l

3.5.4.5 Impact of Failure Non Safety Related Structures, 3.58 l Systems, and Components Due to a Design Basis Tornado 3.5J References 3.58 i

ILLUSTRATIONS Elgure Ihic bge i 3.51 hiissile Velocity and Displacement Characteristics l

Resulting from Saturated Steam and Water Blowdowns (1050 psia Stagnation Pressure) 3.59 l

3.5 iii Amendment l$

ABM zw-u RIV H Standardflant l (9 valent static load concentrated at the impact impact the safety function of a safety related V area is determined. The structural response to systems and components will be provided to the this load,in conjunction with other appropriate NRC by the app!! cant referencing the ABWR design loads, is evaluated using an analysis design. (See Subsection 3.5.1.4),

procedure similar to that in Reference 6 for '

rigid m*ssiles, and the procedure in Reference 7 3.5.5 References for deformeble missiles. ,

1. C. V. Moore, The Design of Barricades for 3.$A Interfaces flasardous Pressure Systems, Nucleat Engineering and Design, Vol. $,1967.

3.5.4.1 Protection of Ultimate Heat Sink

, 2. F. J. Moody, Prediction of Blowdown Thrust j Compliance with Regulatory Guide 1.27 as and /rt forces, ASME Publication 69.HT 31,

} related to the ultimate heat sink and connecting August 1%9.

1 conduits being capable of withstanding the A effects af externally generated missiles shall be 3. A. Amirikan, Design of Protective Struc-demonstrated (See Subscetion 3.5.2). turcs, Bureau of Yards and Docks, Publica-tion No. NAVDOCKS P 51, Department of the 1 3.5.4.2 ltl6siles Generated by Natural Phenomena Navy, Washington, D.C., August 1060. ;

from Res 'alader of Plant Structurws, Systems and i Compone6 ts 4. A. E. Stephenson, Full Scale Tornado. Mis.

sile impact Tests, EPRI NP 440, July 1977, The tem sinder of plant structures, systems, prepared for Electric Power Research and compon snts shall be analytically checked to Institute by Sandia Laboratories, ensure that s.uring a site specific tornado they will not generate missiles exceeding the missiles 5. W. B. Cottrell and A. W. Savolainen, U. S.

considered under Subsection 3.5.1.4. Reactor Containment Technology, ORNL-

. C.V) t NSIC 5, Vol.1, chapter 6, Oak Ridge Na-l 3.5.4.3 Site Proximity Missiles and Aircraft tional Laboratory.

l Haanrds.

! 6. R. A. Williamson and R. R. Alvy, /rnpact

! Analyses shall be provided that demonstrate Effect of Fragments Striking Structural ,

that the probability of site proximity missiles Elcments, Holmes and Narver, Inc., Revised (including aircraft) impacting the ABWR Standard November 1973.

g Plant and causing consequences greater than 10CFR l

Part 100 exposure guidelines is 3.10'7 per year 7. J. D. Riera, On the Stress Analysis of l (See Subsection 3.5.1.6) Structures Subjected to Aircraft impact L

Forces, Nuclear Engineering and Design, l

3.5.4.4 Secondary Missiles Ins!de Containment North Holland Publishing Co., Vol. 8,1968.

Protection against the secondary missiles 8. American National Standard For Estimating

$ inside containment described in Subsection Tornado and Other Extreme Wind Characteris-3.5.1.2.3 shall be demonstrated. tics at Nuclear Power Sites, ANSI /ANS 2.3.

9. River Bend Station Updated Safety Analysis g L l 3.5.4.5 Impact of Fallure of Non Safety Related Rcport, Docket No. 50 458, Volume 6, pgs. s Structures, Systems,and Components Due to a 3.5 4 and 3.5 5, August 1987.

Design Basis Tornado An evaluation of all non safet related structures, systems, and components (not housed g in a tornado structure) whose failure due to a j design basis tornado missile that could adversely

(

Amendment 15 35-8

~ _ _ _ _ _ _ _ _ .

ABWR mai:

Standard Plant niv n

'O 3.9.3.4 Component Supporti correspond to those used for design of the sup-() ported pipe. The component loading The design of bolts for component supports combinations are discussed in Subsection is specified in the ASME Code Section 111, 3.9.3.1. The stress limits are per AShtE 111, Subsection NF. Stress limitt, for bolts are given Subsection NF and Appendix F. Supports are in NF 3225. The rules and stress limits which generally designed either by load rating must be satisfied are those given in NF 3324.6 method per paragraph NP 3260 or by the stress multit tied by the appropriate stress limit factor limits for linear supports per paragraph for the particular service loading level and NF 3231. The critical buckling loads for the stress category specified in Table NF 3225.21. Class 1 piping supports subjected to faulted loads that are more severe than normal, upset Moreover, on equipment which is to be, or and emergency loads, are determined by using may be, mounted on a concrete support, sufficient the methods discussed in Appendicer F and XVil holes for anchor bolts are provided to limit the of the Code. To avoid buckling in the piping anchor bolt stress to less than 10,000 psi on the supports, the allowable loads are limited to nominal bolt area in shear or tension. two thirds of the determined critical buckling loads.

Concrete anchor bolts which are used for pipe support base plates will be designed to the The design of all supports for non nuclear applicable factors of safety which are defined in piping satisfies the requirements of ANSI I&E Bulletin 79 02, ' Pipe Support Base Plate B31.1, Paragraphs 120 and 121.

Designs Using Concrete Expansion Anchor Bolts,'

Revision 1 dated June 21,1979. For the major active valves identified in Subsection 3.9.3.2.4, the valve operators are 3.9.3.4.1 Piping not used as attachment points for piping supports.

N Supports and their attachments for casential

[b ASME Code Section 111, Clars 1,2, and 3 piping The design criteria and dynamic testing re-are designed in accordance with Subsection NF' up quirements for the AShiE 111 piping supports g to the interface of the building structure. The are as follows:

g building structure component supports are de-signed in accordance with the AISC specification (1) Piping Supports All piping supports are for the Dcnign, Fabrication, and Erection of designed, fabricated, and assembled so Structural Steel for buildings. The loading com- that they cannot become disengaged by the l binations for the various operating conditions movement of the supported pipe of i equipment after they have been installed.

All piping supports are designed in accordance with the rules of Subsection NF of the ASME Code up to the building structure interface as defined in the

Augmented by the following: (1) application of project design specifications.

Code Case N 476, Supplement 89.1 which governs l the design of single angle members of ASME Class (2) Spring 11 angers The operating load on 1,2,3 and MC linear component supports; and (2) spring hangers is the load caused by dead when eccentric loads or other torsional loads are weight. The hangers are calibrated to en-not accommodated by designing the load to act sure that they support the operating load through the shear center or meet ' Standard for at both their hot and cold load settings.

Steel Support Design *, analyses will be performed Spring hangers provide a specified down in accordance with torsional analysis methods travel and up travel in excess of the such as: ' Torsional Analysis of Steel Members, specified thermal movement.

USS Steel Manual *, Publication T114 2/83.

O

)

Amenht t$ 3A31

ABWR = = =

Standard Plant RIN B (3) Snubbers The operating loads on snubbers are the loads caused by dynamic events (e.g., seismic, RBV due to LOCA and SRV discharge, discharge through a relief valve line or valve closure) during various operating conditions. Snubbers restrain piping against response to the vi.

bratory excitation and to the associated differential movement of the piping system support anchor points. The criteria for locating snubbers and ensuring adequate load capacity, the structural and me-chanical performance parameters used for snubbers and the installation and inspec-tion consider ations for the snubbers are es follows:

(a) Required Load Capacity and Snubber Lo-cation The entire piping system including valves and support system between an-chor points is mathematically modeled for complete piping structural analysis. In the dynamic analysis, the snubbers are modeled as a spring with a given spring stiffness depending on the snubber size. The analysis determines the forces and moments acting on each piping components and the forces acting on the snubbers due to all dynamic loading and operating conditions defined in the piping design specification. The forces on snub-bets are operating loads for various operating conditions. Those loads are assumed no to exceed the snubber design load capacity for various operating condit.lons, i.e., design, normal, upset, emergency and faulted.

O Amendment 11 3.9 31.1

ABM :sreaoorn RIV B

$1&ndard Plant O significant modes of the piping system; (i) There are no visible signs of damage or impaired operability as a result of o Displacements are measured to storage, handling, or determine the perf ormance installation.

characteristics specified; (ii) The snubber location, o Tests are conducted at various orientation, position temperatures to ensure operability setting, and configuration over the specified range; (attachments, extensions, etc.) are according to design o Peak test loads in both tension and drawings and specifications, compression are required to be equal to or higher than the rated load (iii) Snubbers are not seized, requirements; and frozen or jammed, o The snubbers are tested for various (iv) Adequate swing clearance is abnormal environmental conditions, provided to allow snubber Upon completion of the abnormal movements.

environmental translent test, the snubber is tested dynamically at a (v) If applicable, fluid is to be frequeney within a specificd recommended level and not be frequency range. The snubber must leaking from the snubber operate normally during the dyncmic system, test.

(vi)- Structural connections such O (d) Snubtxt Installation Requirements An installation instruction manual is as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter required by the pipe support design pins are installed correctly, specification. This manual is required to contain instructions for storage, if the period between the handling, erection, and adjustments (if initiai_ pre service necessary) of snubbers. Each snubber examination and initial has an installation location drawing system pre operational tests which contains the installation location exceeds 6 months because of -

of the snubber on the pipe and unexpected situations, structure, the hot and cold settings, reexamination of items 1,4, and additional information needed to and 5 will be performed, install the particular snubber. Snubbers which are installed incorrectly or otherwise fall (c) Snubber Prc r.cryice Examination to meet the aboyc requirements will be repaired The pre service examination plan of all or replace:I and re examined snubbers covered by the Chapter 16 tech- in accordance with the above nical specifications will be prepared, criteria.

This examination will be made after snubber installation but not more than 6 (4) Struts The design load on struts months prior to initial system pre oper- includes those loads caused by dead ational testing. The pre service weight, thermal expansion, seismic forces examination will verify the following: (i.e., OBE and SSE), other RBV loads, O

Amendment 7 3.9 33 I

ABWR m6im Standard Plant RPV D system anchor displacements, and reaction (P/Perit) + (4/9 erit) + (f /' crit) forces caused by relief valve discharge or valve closure, etc. < (1/S.F.)

Struts are designed in accordance with ASME where:

Code Section III, Subsection NF.3000 to be capable of carrying the design loads for q = longitudinalloa '

various operating conditions. As in case of P = external pressure snubbers, the forces on struts are obtained r = transverse shear stress from an analysis, which are assured not to S.F. = safety factor exceed the design loads for various = 3.0 for design, testing, service operating conditions. levels A & B

= 2.0 for Service Level C 3.9.3.4.2 Reactor Pressure Vessel Support Skirt = 1.5 for Ser ice level D.

The ABWR RPV support skirt is designed as an 3.9.3.4.3 Reactor Pressure Vessel Stabillier ASME Code Class I component per the requirements !

of ASME Code Section 111, Subsection NF'. The The RPV stabilizer is designed as a Safety loading conditions and stress criteria are given Class i linear type component support in in Tables 3.91 and 3.9 2, and the calculated accordance with the requirements of ASME stresses meet the Code allowable stresses in the Boiler and Pressure Vessel Code Section ill, critical support areas for various plant Subsection NP. The stabilizer provides a operating conditions. The stress level margins reaction point near the upper end of the RPV assure the adequacy of the RPV support skirt. An to resist horizontal loads due to effects such analysis for buckling shows that the support as carthquake, pipe rupture and RBV. The skirt complies with Subparagraph F.1332.5 of ASME design loading conditions, and stress criteria 111, Appendix F, and the. loads do not exceed two are given in Tables 3.91 and 3.9 2, and the thirds of the critical buckling strength of the calculated stresses meet the Code allowable skirt. The permissible skirt loads at any stresses in the critical support areas for elevation, when simultaneously applied, are various plant operating conditions, limited by the following interaction equation:

3.9.3.4.4 Floor. Mounted MQor Equipment (Pumps, Heat Exchangers, and RCIC Turbine)

- Since the major active valves are supported by piping and not tied to building structures, valve " supports

do not exist (See Subsection 3.9.3.4.1).

The IIPCF, RilR, RCIC, St.C, FPCCU,

' Augmented by the following: (1) application of SPCU, and CUW pumps; RMC, RHR, Code Case N 476, Supplement 89.1 which governs RWCU, and FPCCU heat exchangers; and RCIC the design of single angle members of ASME Class turbine are all analyzed to yerify the ,

1,2,3 and MC linear component supports; and (2) adequacy of their support structure under l when eccentric loads or other torsionalloads are vahs plant operating conditions. In all i not accommodated by designing the load to act cases, the load stresses in the critical through the shear center or meet ' Standard for support areas are within ASME Code allowables.

Steel Support Design *, analyses will be performed in accordance with torsional analysis methods Selsmic Category I active pump supports are such as: " Torsional Analysis of Steel Members, qualified for dynamic (seismic and other RBV)

USS Steel Manual *, Publication T114 2/83. loads by testing when the pump supports 9

Amendment t3 3.9.M

ABWR Standard Plant 2aamn myc ;

CHAPTER 4 p

TABLE OF CONTENTS Section lille Ease 4 REACTOR 4.1 SUMM ARY DESCKIFrlON 4.11 4.1.1 Reactor Pressure Vessel 4.11 j l

4.1.2 Reactor internal Components 4.11 l

4.13 Reactivity Control Systems 4.12 ,

i 4.1.4 Analysis Techniques 4.13 l l 4.2 FUEL SYFTEM DESIGN 4.21 l l

4.2.1 Discussion 4.2-1 p 4.2.2 Interfaces 4.21 4.23 References 4.21 43 NUCLEAR DESIGN 43.1 l

43.1 Discussion 43 1 l

l 43.2 Interfaces 43 1 433 References 43-1 4.4 H[L)tMAl, HYDRAULIC DESIGN 4,4 1 1

1 4.4.1 Discussion 4.41 l

l Amendment 15 1-1

l' ABWR man Standard Plant RIN. C CHAPTER 4 g

TABLE OF CONTENTS (Continued)

Section Tult East 4.4.2 Deseeiption of the Thermal and liydraulie Design of the Reactor Coolant System 4.4-2 4.43 Interfaces 4.4-4 4.5 REACTOR MATERIALS ,

4.5.1 Control Rods System Structural Materials 4.51 4.5.2 Reactor Internal Materials 4.53 i

4.53 Interfaces 4.5 4a 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6.1 Information for Control Rod Drive '

System 4.6 1 4.6.2 Evaluations of the CRDs 4.6 10 l 1

4.63 Testing and Evaluation of the CRDs 4.6 13 1 4.6.4 Information for Combined Performance of l Reactivity Conti01 Systems 4.6-16 i 4.6.5 Evaluation of Combined Performance 4.6 16 i

APPENDIX B CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR TYPICAL AHWR 4A 1 AP.fLNDIX 4B FUEL LICENSING ACCEL'TANCE CRITERIA APPENDIX 4C CONTROL ROD LICENSING ACCEPTANCE CRITERIA 4.iii

- Amendment t$

ABM namosu Standard Plant UN. A SECTION 4.1 CONTENTS

! Sectlon ]l[jg b 4.1.1 Reactor Pressure Vessel 4.1 1 4.1.2 Reactor Internal Comnonents 4.11 4.1.2.1 Reactor Core 4.11 4.1.2.1.1 Fuel Assembly Description 4.12 4.1.2.1.2 Assembly Support and Control Rod Location 4.12 4.1.2.2 Shroud 4.12 4.1.23 Shroud Head and Steam Separators 4.1 2 4.1.2.4 Steam Dryer Assembly 4.12 4,13 Etasilyltv control svstems 412 4.13.1 Operation 4.12 4.13.2 Description of Control Rods 4.12 4.133 Supplementary Reactivity Control 4.12 4.1.4 Analvals Techniques 4.13 4.1.4.1 Reactor Internal Components 4.13 4.1.4.1.1 NASTRO4V 4.13 4.1.4.1.2 SAP 4GO7 4.13 4.1.4.13 HEATER 4.13 4.1.4.1.4 USAGOB01 4.13 4.1.4.1.5 ANSYS 4.13 4.1.4.1.6 CLAPS 4.14 4.1.4.1.7 ASIST 4.14 0

4.1 ii ,

-um .- .

MM 2M6100A11 Standard Plant arrv. c SECTION 4.1 CONTENTS (Continued) i i

Section Ililt East

, 4.1.4.1.8 SEISMO3 4.1 4 l

4.1.4.1.9 SASS 1015 4.15 4.1.4.2 Fuel Design Analysis 4.15 4.1.4.3 Reactor Systems Dynamics 4.15 l 4.1.4.4 Nuclear Analysis 4.15 4.1.4.5 Neutron Fluence Calculations 4.15 4.1.4.6 Thermal Hydraulic Calculations 4.15 4.1J (Deleted) 4.16 l

ILLUSTRATIONS Figure 1111e East 4.11 Core Configuration with Location l ofInstrumentation 4.1 7 !

l 4.1.iii Amendment 15 e~

l ABWR muun Standard Plant RIV C

4. REACTOR q

O 1 4.1

SUMMARY

DESCRIPTION core spray and core flooding spargers. Except for the Zircaloy in the reactor core, these The reactor assembly consists of the reactor reactor internals are stainless steel or other pressure vessel, pressure cos.taining corrosion resistant alloys. The fuel assemblies ,

appurtenances including CRD housings, in core (including fuel rods and channel), control )

instrumentation housing and the head vent and blades, shroud head and steam separator j spray assembly plus the reactor internal assembly, and steam dryers and in core components described in Subsection 4.1.2. Figure instrumentation dry tubes are removable when the 3.91 (Reactor Vessel and Internal Assembly) reactor vessel is opened for refueling or l shows the arrangement of the reactor assembly maintenance, components. A summary of the important design and performance characteristics is given in 4.1.2.1 Reactor Core S u b s e c t io n 1.3.1.1. Loading conditions for I l

reactor assembly components are specified in important features of the reactor core are:

Subsection 3.9.5.2.

(1) T1.c bottom.cntry cruciform control rods, i For the purpose of this SSAR, a typical fuel Rods of this design were first introduced in and control rod design and core loading pattern the Dresden 1 reactor in April 1961 and have was used as the basis for the system response accumulated thousands of hours of service, studies in Section 6.3 and Chapter 15. The actual fuel and control rod designs and core (2) Fixed in core fission chambers (LPRMs) loading pattern to be used at a plant will either provide continuous local power range neutron have been approved or will meet criteria approved flux monitoring. A guide tube in each by the USNRC, and will be provided to the USNRC in core assembly provides for a traversing

(' for information. The fuel and control blade ion chamber (TIP) for calibration and axial design and core loading pattern used for the detail. Start up range neutron monitors system response studies are documented in this (SRNMs) are located at fixed locations chapter; information to be provided by the between the (LPRMs) as shown on Figure utility referencing the ABWR design is contained 4.11. The in core location of the start up in the interface subsection of each of these and source range instruments provides sections, coverage of the large reactor core and provides an acceptable signal to noise ratio

' 4.1.1 Reactor Pressure Vessel and neutron to gamma ratio. Allin core instrument leads enter from the bottom and The reactor pressure vessel includes the the instruments are in service during ,

reactor internal pump (RIP) casing and flow refueling. In core instrumentation is restrictors in each of the steam outlet nozzles presented in Subsection 7.6.1.

and the shroud support and pump deck which form L the partition between the RIP suction and (3) As shown by experience obtained at Dresden 1

) discharge. The reactor pressure vessel design and all other BWR plants, utilizing the in-l and description are covered in Section 5.3. core flux monitor system, the desired power distribution can be maintained within a 4.1.2 Reactor Internal Components large core by proper control rod scheduling.

The major reactor internal components are the (4) The fuel channels provide a fixed flow path core (fuel, channels, control blades and for the boiling coolant, serve as a guiding

( instrumentation), the core support structure surface for the control rods and protect the i (including the shroud, top guide and core plate), fuel during handling operations.

the shroud head and steam separator assembly, the l (5) The mechanical reactivity control permits O steam dryer assembly, the feedwater spargers, the Q

Amendment t5 4.11

ABWR m ima Standard Plant RLY C criticality checks during refueling and provides maximum plant safety. The core is designed to be suberitical at any time in it's operating history with any one control rod fully withdrawn and the other control rods fullinserted.

(6) The selected control rod pitch represenu a practical value of individual control rod reactivity worth, and allows adequate clearance be!'w the pressure vessel between CRD mechan.sms for case of maintenance and removal.

i i

O' Amendment 13 O

4.11.1

.. < . - .y - _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ . ,

ABWR m6ima myc l

Standard Plant O

(") (7) The reactor core is arranged as an upright circular cylinder containing a large number 4.1.2.4 Steam Drpr Assembly of fuel cells and is located within the core The information on the steam dryer assembly shroud inside the reactor vessel, is presented in Subsection 3.9.5.1.2.3.

4.1.2.1.1 Fuel Asserably Description 4.1.3 Reactivity Control Systems The fuel assembly description is discussed in 4.13.1 Operation Section 4.2.

The control rods perform dual functions of 4.1.2.1.2 Assembly Support and Control Rod power distribution shaping and reactivity Location control. Power distribution in the core is controlled during operation of the reactor by A few peripheral fuel assemblies are manipulation of selected patterns of rods. The supported by the core plate. Otherwise, rods, which enter from the bottom of the near individual fuel assemblies in the core rest on cylindrical reactor core, are positioned to fuel support pieces mounted on top of the control counterbalance steam volds in the top of the rod guide tubes. Each guide tube, with its core and effect significant power flattening orificed fuel support, bears the weight of four assemblics and is supported by a control rod These groups of control elements, u,ed for drive penetration nozzle in the bottom head of power flattening, experience a somewhat higher the reactor vessel. The core plate provides duty cycle and neutron exposure than the other lateral support and guidance at the top of each rods in the control system.

control rod guide tube and directs the reactor recirculation into the orificed fuel support and The reactivity control function requires that p through the fuel assemblies. The top guide, all rods be available for either reactor ' scram' mounted on top of the shroud, provides lateral (prompt shutdown) or reactivity regulation, V support and guidance for the top of each fuel Because of this, the control elements are assembly. The reactivity of the core is con- mechanically designed to withstand the dynamic trolled by cruciform control rods and their forces resulting from a scram. They are associated mechanical hydraulle drive system. connectedto bot tom m ou nt ed, The control rods occupy alternate spaces between electro. hydraulically actuated drive mechanisms fuel assemblics Each independent drive enters which allow either electric motor controlled the core from the bottom, and accurately axial positioning for reactivity regulation or positions its associated control rod during hydraulle rap.id scram insertion. The design of lL normal operation with an electric motor driven the rod.to drive connection permits each blade

! ball screw. For scram hydraulic pressure acts on to be attached or detached from its drive the hollow cylinder to exert several times the without disturbing the remainder of the control force of gravity to insert the control rod during system. The bottom. mounted drives permit the the scram mode of operation. Bottom entry allows entire control system to be left intact and optimum power shaping in the core, case of remain operable for tests with the reactor L refueling and convenient drive maintenance, vessel open.

i

( 4.1.2.2 Shroud 4.1J.2 Description of Control Rods

.The information on the shroud is presented in A description of the control rods is Subsection 3.9.5.1.1.1, presented in Section 4.6.

4.1.2J Shroud Head and Steam Separators 4.1JJ Supplementary Reacthity Control The information on the shroud head and sep. The core control requirements are met by use arators is presented in Subsection 3.9.5.1.2.1. of the combined effects of the movable control Amendment 13 (12

ABWR m an i

Standard Plant uvc element technique. Although it is strictly a linear program, approximate non linear analysis (2) SEPST can be performed by an iterative scheme called This ECP is the SEl'M post processor. SEPST

Equivalent Linear Method.*

conderaes the SF'.5M output data into a form which is morv practical to interpret. It 4.1.4.2 Fuel Design Analysis determines and prints the initial values, the maximum and minimum values for all The fuel design analysis models are approved components, and the times of their or developed using criteria approved by the ocr.urrence, in addition, it generates the USNRC.

rerponse time history plots of selected cota pone nt s. 4.1.4.3 Reactor Systtms Dynamics (3) CRTil The analysis techniques and computer codes used in reactor systems dynamics are approved or The CRTF1 program uses, as input, the scaled developed using criteria approved by the USNRC, or composite horizontal acceleration time histories at the mid fuel and end fuel 4.1.4.4 Nuclear Analysis positions to determine the clamping forces to be applied to the analysis model friction The analysis techniques are approved or elements; to determine the scram uplift developed using criteria approved by the USNRC, forces on a bundle; to determine inertial forces of the fuel in order to obtain 4.1.4.8 Neutron Fluence Calculations reaction forces on both ends of the fuel; and to determine fuel center deflection and Neutron vessel fluence calculations were uplift forces due to scram, carried out using a two dimensional, discrete ordinates, Sn transport code with general O'- 4.1.4.1.9 SASS 101 anisotropic scattering.

SASS 101S can be used to perform dynamic This code is the most widely used two soil. structure interaction analysis in two or dimensional, discrete ordinated code which 3 three dimensions. The seismic environment solves a wide variety of radiation transport t consists of an arbitrary three dine.nsional (3D) problems. Tha program will solve both fixed superposition of inclined body waves and surface source and multiplication problems. Rectangular waves. The site consists of semi infinite (X, Y, or RZ) and polar (R, G) geometry are clastic of viscoclastic halfspace. The structure allowed with various boundary conditions. The l and the soil can be modeled using a combination fluence calcula&.a. incorporate, as an initial of 3D solid element,3D beam element, four node starting point, neutron fision distributions quadrilateral plate /shell element,2D, four node prepared from core physics data as a distributed !

plane strain element,3D spring element, and source. Anisotropic scattering was considered stiffness / mass matrix clement. Seismic load in for all regions. The cross sections were pre-the form of an acceleration time history can be pared with 1/E flux weighting Polynominal expan-applied to one of the three global directions at slon matrices for anistropic scattering but did a control point on a soll layer interface. not include the resonance self shleiding External forces or moments such as impact loads, factors, wave forces, or loads from the rotating machinery can be introduced directly to nodes in the 4.1.4.6 Thermal llydraulic Calculations soil structure system.

The thermal. hydraulic models are approved by 3 ASS 101S is formulated in the frequency domain or developed using criteria approved by the using the complex response method and the finite USNRC.

O Amendment 13 4.15

ABM 23461=rn Standard Plant niv. c 1

4.1.5 (Deleted) 9 O;

i l

Amendment 15 4.14 O

ABWR numn Standard Plant nrv. c

( SECTION 4.2 CONTENTS Section M '4age 42.1 Discussion 4.21 4.2.2 Interfaces 4.21 4.2.3 Etftrences 4.21 ILLUSTRATIONS Ugurs 11tle East l

4.21 Control Rod Assembly Used in 4.22 Response Analyses O

l l

~

Aaendment 15

ABWR maman nry c Standard Plant 4.2 FUEL SYSTEM DESIGN O 4.2.1 Discussion The fuel to be loaded in an ABWR is any f9el i

L design approved by the USNRC or that meets the criteria documented in Appendia 4B. Using these designs will assure that all fuel system design requirements are met.

To demonstrate ABWR system response in this SSAR, a reference core of BP8x8R fuelis used.

This core is shown in Section 4.3; information for this fuel design is provided in Reference 1.

Each utility referencing the ABWR design may have different fuel and core designs which will be 4.2.3 References provided by the utility to the USNRC for information. S e e Subsection 4.2.2.1 for 1. GE Fur / Bundic Des /pe r, NEDE 31152P.

interf ace requirements.

2. GE ControlRod Desig s, (To be issued).

The control rods perform the dual function of power shaping and reactivity control. A discussion of the rod control system components is presented in Section d.6. .

The control rod design to be used in an ABWR h any design approved or that meets the criteria documented in Appendix 4C, To demonstrate the ABWR system response in this report, a control rod design of sheathed cruciform array of stainless steel tubes filled with boron. carbide was used. This design is documented in Reference 2 and shown in Figure 4.21. The control blade design to be used at the plant will be provided by the utility referencing the ABWR design to the

= USNRC for information. See Subsection 4.2.2.2.

-- for interface requirement.

4.2.2 Interfaces 4.2.2.1 Fuel Design The fuel bundle name and a reference to documentation of the fuel design will be provided by the utility referencing the ABWR design to the USNRC foi information.

g 4.2.2.2 Control Blade Design The control blade model and reference to documentation of the control blade design will be provided by the utility referencing the ABWR design to the USNRC for information.

f 9 Amendment 15 4.21 1

ABWR 23y m ui Standard Plant nrv c l

l O HANDLE O ,

, 0 0 0 0 0 l

t 0 O! o

'O NEUTRON D 0 ABSORBER g

e RODS e

SHEATH C

0 g 0 g BLADE O

, l O

O g U

/

/Nn, ,

l COUPLING SOCKET 8722642 L

l- Figure 4.2 1 CONTROL ROD ASSEMBLY USED IN RESPONSE ANALYSES Amendment l' 4.22 i

ABWR n^aa^n Standard Plant RIV C l SECTION 4.3 CONTENTS Section lhlt EA82 43.1 Discussion 43 1 43.2 Initdtsta 431 43.2.1 Core leading Pattern 431 43.2.2 Core Effecthe Multiplication Values 43 1 O

O Amendment 15 l

ABWR m omin Standard Plant RIV C SECTION 4.3 TABLES g

Table 31 tit East 43 1 Calculated Core Effective Multiplication and Control System Worth . No Voids, 200C (for the core loading in Figure 431) 43 3 ILLUSTRATIONS !

Figure lult East 43 1 Core leading Map Used for Response Analyses 43-4 43-2 Typicalleading Patterns 43 5 (Using Blank Fuel Bundles in Initial Cycle)

O Amendment 15 ;

d

ABWR 2wan Standard Plant Riv c

/~~T 4.3 NUCLEAR DESIGN 4.3.2.2 Cort Effective htuttiplication Values

\ )

4.3.1 Discussion The beginning of each cycle Kert for uncontrolled, fully controlled, and strongest The nuclear design of the fuel has either rod withdrawn conditions or similar cases and been approved by the USNRC or meets the nuclear the increase in cold core reactivity throughout criteria approved by the USNRC as documented in the cycle will be provided by the utility Appendix 4B. For the purpose of this SSAR, a referencing the ABWR design to the USNRC for reference core loading of 872 fuel bundles was information, developed to use as the basis for the system dynamic response analyses given in Section 6.3 and Chapter 15. This reference core loading is given in Figure 4.3.1. Individual plant core loadings are expected to include different fuel designs and enrichments. In addition the first core loading in a plant may use a varying number of blank fuel bundles rather than enriched bundles in parts of the core. Some or all of I these blank fuel bundles could be replaced with enriched bundles in later cycles. An example of this type cf a loading pattern is shown in Figure ,

I 4.3 2. The core loading for each plant will be provided to the USNRC for information as documented in Subsection 4.3.2.1. The core effective multiplication factor (Kett) at cold reactor conditions is calculated each cycle to

/3 assure the adequacy of the plant shutdown h margin. For the core given in Figure 4.31, the Kett for the cases of all rods withdrawn (uncontrolled), all rods inserted (controlled),

and the strongest rod withdrawn are provided in Table 4.31. Kett values for these or similar cases will be provided by each plant the to USNRC for information. See Subsection 4.3.2 for interface requirements.

A variety of control rod patterns can be used with each core loading to meet the plant Technical Specification limits. An example of these control rod patterns and their impact on the power distribution within the core is given in Appendix 4A.

4.3.2 Interfaces 4.3.2.1 Cort Loading pattern The fuel design, number of bundles of each enrichment, and location of each bundle to be loaded will be provide by the utility referencing the ABWR design to the USNRC for information.

p

( )

LJ Amendment 15 43-1

A L* W R m uonra ,.

Sandard?lant Rf?V. c !

O l

i

+

l i

l I i

+

0 t l

O Amendment 15 4.3-2

-- , + + , . . __,-..,.--,..w--

- - - ,- ,- ..~- -. ..ew, ,,,,,,_-.-t-g * - - - -

ABWR m ,,,

nyc l

l Standard Plant Table 4.31 j CALCULATED CORE EFFECTIVE MULTIPLICATION ,

AND CONTROL SYSTEM WORTH . NO YOIDS,200C (for the core loading in Figure 4.31)

Begin.aing of Cycle, K effective Uncontrolled 1.0876 Fully Controlled 0.9145 Strongest Controf Rod Out 0.9591 R. Maxirnuna increase in Cold Core Reactisity with Exposure Cycle, Ak 0.0000 t

O l

i I

O g3 Amendment 15

- - - ~ - . _ _ . _ _ . . . _

l ABWR .wamn nrv c S_ tandard Plant 1

( l

( ) 1 U/ 4 4 4 )

NOTE: LOADING PATTERN 15 SHOWN FOR -

OUARTER CORE ONLY. ROTATIONAL SYMMETRY APPLIES. 4 4 4 2 2 1 2 -

-, J 4 2 2 1 1 2 2 3 4 4 4 1 2 1 1 2 1 2 4 4 2 2 1 2 1 2 3 1 2 1 5 4 2 1 1 2 1 2 2 1 2 1 3 2 6 -

4 2 2 1 2 1 2 3 1 2 3 1 3 7 4 2 1 1 2 1 1 2 1 2 3 1 2 1 8 2 3 2 3 9 4 2 1l 2 1 3 3 1 1 1 4 1 2 1 1 3 3 1 3 1 2 2 1 3 10 4 1 2 1 2 2 1 1 2 1 2 3 1 2 1 11 4 2 3 2 3 2 2 3 2 12

(~] 2 2 1 1 1 1 1

~

/ 3 2 3 13 4 2 1 2 2 1 2 3l 1 2 1 3 1 L

4 1 1 3 1 2 3 1 2 3 1 3 3 1 2 1 14 4 2 1 2 1 2 3 1 2 2 1 2 2 1 2 3 2 15 4 1 2 1 2 3 .1 2 1 1 2 3 1 2 3 1 2 16 4 2 1 2 3 1 2 2 3 3 2 1 2 3 1 2 3 17

- l mm .e l J+1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 BUNDLE TYPE ENRICHMENT NUMBER OF BUNOLES 1 HIGH ENRICHMENT,3.18 wt % 308 2 MEDIUM ENRICHMENT,2.19 wt % 324 l

l 3 LOW ENRICHMENT,1.23 wt % 148 4 NATUR AL URANIUM,0.71 wt % 92 e 87 2LM1 f x I )

V Figure 4.3-1 CORE LOADING MAP USED F0'A RESPONSE ANALYSES Amendment 15 01

_v

l ABWR mamn Standard Plant an c e'

i GE PROPRIETARY provided under separate cover (Figure 4.3 2) f,ggg Amendment 4.35 15 i

e Amendment 13 4}5

t.

MNE 2M6mW ;

Standard Plant RTV C O sectio"4 4 -

CONTENTS Section M EAgt 4.4.1 Diguuj.QD 441 4.4.2 Descrintion of the Thermal and Htdraulle Dtijgn of the React (Coolant Sntem 4,4.2 y

g a

4.4.2.1 Plant Conry;uration Data 4.4 2 -

4.4.2.1.1 Reactor Coolant System Configuration 4.4 2 mumm u

E O _

Amendment 13

'M

=

ABM 234sioo48 i

Standard Plant nrV.A APPENDIX 4A h ILLUSTRATIONS (Continued) l El811m Blic P_ast lI j 4A3 2a Summary of Haling Condition i 111% Rated Core Fiow 4A 9

! 4AS-2b Relative Axial Power at 9.0 GWd/MT l l- Cyde Exposure (Haling;) 4A 10 l

l

! 4A3 2c Relative Axial Exposure at 9.0 GWd/MT Cycle Exposure (Haling) 4A10 4A3 2d lategrated Power per Bundle (Haling) at 9.0 GWd/MT Cyde Exposure 4A 11 4A.) 2e Average Bundle Exposure (Haling) at 9.0 GWd/MT Cyde Exposure 4A 11 ;

l 4AS 3a Summary of 0.22 GWd/MT Condition 4A 12 4AJ-35 Relative Axial Power at 0.2 GWd/MT Cycle Exposure 4A 13 1 4A3 3e Relative Axial Exposure at 0.2 GWd/MT Cyde Exposure 4A 13 ,

4AS 3d lategrated Power per Bundle at 0.2 GWd/MT Cyde Exposure 4A 14 : l 4AS 3e Average Bundle Exposure at 0.2 GWd/MT Cycle Exposure 4A 14 L

4A34a Summary of1.10 GWd/MT Condition 4A 15 4A3-4b Relative Axial Power at 1.1 GWd/MT Cycle Exposure 4A 16 - l 4A3-4c Relative Axial Exposure at 1.1 GWd/MT Cyde Exposure 4A 16 4A3-4d Integrated Power per Bundle at l

l 1.1 GWd/MT Cycle Exposure 4A 17 4A3-4e Average Bundle Exposure at 1.1 GWd/MT Cycle Exposure 4A 17 4AJ Sa Summaryof 2.20GWd/MTCondition 4A 18 4A lii E

4 ABWR- mnen nry e Standard Plant SECTION 4.4 !

l TABLES i

IRhlt Dlt EAST j 4.4-1 Typical Thermal and Hydraulic Design Characteristics of the Reactor Core 4.45 4.42 Void Distribution for Analped Core 4.4-7 4.4-3 Flow Quality Distribution for Analvzed Core 4.4-8 4.4-4 Axial Power Distribution Used to Generate Void and Quality Distributions 4.49 l for Analyzed Core

'4.45 Reactor Coolant System Geometric Data 4.4-10 ILLUSTRATIONS Figure Iuls Ease

, 4.4-1 Power Flow Operating Map Used for 4.4 11 System Response Study 1

i e1 4.4-iv Amendment t$

3p' O x

3%

l 1

l ABWR maan i Standard Plant nrv. c l r~

t 4.4 THERMAI HYDRAULIC DESIGN N_

4.4.1 Discussion The thermal hydraulic design has been approved by the USNRC or meets the criteria approved by the USNRC as documentad in Appendix 4B. An evaluation of plant performance is dependent upon the fuel design. Typical thermal hydraulic paratneters for the ABWR are compared to those for a typical BWR/6 plant. The values for the ABWR are based on tbc core loadirig given in Figure 4.3 1.

For information tbc axial distribution of core void fractions for the average radial channel and the maximum radial channel (end of node value) for the core shown in Figure 4.31 are given in Table 4.4 2. Tbc core average and maximum exit value is also provided. Similar distributions for steam quality are provided in Table 4.4 3. Tbc core average axial power distribution used to produce these tables is given in Table 4.4 4 e v)

,q Amendment 15 4.4-1

. . = . - .. .. - -- - . .

I l

.AB M - 2 m ioorn nv c i

Standard Plat.t 4A.2.3.3 Regions of the Power Flow Map 4A.2.3.5 Flow Control Region ! This reg!on defines the system The normal plant startup procedure requires operational capability with the the startup of all RIPS first and maintain at reactor internal pumps running at their minimum pump speed (30% of rated), at their minimum speed (30%). Power which point reactor heatup and pressurization changes, during normal startup and can commence. When operating pressure has been shutdown, will be in this region. The established, reactor power can be increased, normal operating procedure is to start This power flow increase will follow a line up along curve 1. .vithin Region I of the flow control map shown in Figure 4A 1. The system is then brought to the Region il This is the low power area of the desired power. flow level within the normal oper. '

operating map where the carryover ating area of the map (Region IV) by increasing through steam separators is expected the RIP speeds and by withdrawing control rods. ,

to exceed the acceptable value.

Operation within this region is Control rod withdrawal with constant pump precluded by system interlocks. speed will result in power / flow changes along lines of constant pump speed (Curves 1 through Regionll! This is the high power / low flow area 8). Change of pump speeds with constant control of the operating map which the system rod position will result in power / flow changes is the least damped. Operation within along, or nearly parallel to, the rated flow this region is precluded by SCRRI control line (curves A through F).

(Selected Control Rods Run.In). ,

RegionIV This represents the normal operating zone of tho map where power changes

. v[- f can be made, by either control rod

. movement or by core flow changes, through the change of the pump speeds.

4A.2.3A Design Features for Power Flow Control -

4A.2A Thermal and Hydraulle Characterlatics The following limits and design features are Summary Table -

>' employed to maintain power. flow conditions shown in Figure 4.41: The thermal. hydraulic characteristics are provided in Table 4A.1 for the core and tables (1) Minimum Power Limits at intermediate and of Section SA for other portions of the reactor High Core Flows: To prevent unacceptable coolani system, separator performance, the recirculation systern is provided with an interlock to reduce the RIP speed.

(2)' Pump Minimum Speed Limit:. The Reactor Internal Pumps (RIPS) are equipped with Anti Rotation Devices (ARD) which prevent a tripped RIP from rotating backwards. The ARD begins operating at 300 rpm decreasing .,

speed. In order to prevent mechanical wear in the ARD, minimum speed is specified at 300 rpm. However, to provide a stable .

operation, the minimum pump speed is set at :l 450 rpm (30% of required).

' Amtadment 15 44-3 H

l

s ,

ABWR . m om^n Standard Plant Rev. c

(/ Table 4,41 L TYPICAL THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE B%R/6 ;

General 3neratine Conditions 238 748 ABWR*

Reference design thermal output (Mwt) 3579 3926 ;

Fower level for enginected safety features (Mwt) 3730 4005 Steam Dow rate, at 4200F final feedwater 15.40 16.84 temperature (millions Ib/hr)

Core coolant now rate (millions Ib/hr) 104.0 115.1 L

Feedwater flow rate (millions Ib/hr) 15.367 16.807 L

System pressure, nominalin steam dome (psia) 1040 1040 System pressure, nominal core design (psia) 1055 1055 Coolant saturation temperature at core design 551 551 pressure (OF)

/\

50.6 G. Average power density (kW/ liter) 54.1 2 73,303 83,176 Core total heat transfer area (ft )

Design operating minimum critical power ratio See Table 15.0-1 (MCPR)

Core inlet enthalpy at 4200F FFWT (Btu /lb) 527.7 527.6 Core inlet temperature at 4200F FFWT (DF) 533 533 Core maximum exit voids within assemblies (%) 79.0 75.1 Core average void fraction, active coolant 0.414 0.408

Based on the core loading in Figure 4.3-J usedfor sensitivity studies.

O Amendment 15 44-5

ABWR 2346 oorn uv. c Standard Plant

,.a

-/ Table 4.4 2 VOID DISTRIBUTION FOR ANALYZED CORE l l

Core Average Value 0.408 )

Maximum Exit Value 0.751 Active Fuel length .146 inches l Core Average Maximum Channel Ngig (Averane Node Valsg) End of Node Value)

Bottom of Core 1 0 0 2 0 0.011 3 0.010 0.069 )

4 0.045 0.162 5- 0.104 0.260 6 0.174 0346 7 0.247 0.421 8 0.315 0.481 <

9 0372 0.529 10 0.419 0.567 I 11 0.458 0.597

.-- 12 0.489 0.623 b 13 0.515 0.643 .;

14 0.536 0.660 15 0.554 0.675

~'

'(

16 0.570 0.688 -

17 0.585 0.700 18 0.599 0.712 ;

19 0.611 0.723 20 0.623 - 0.733 21 0.633 0.740 22 - 0.641 0.746 1 >

-23 0.646- 0.750

. Top of Core 24 - 0.649 0.751 F

D L-Il 1:

l -' -

Amendment 15 447 1.

ABWR mamn Standard Plant arv. c .

Table 4.4-4 AX1AL POWER DISTRIBUTION USED TO GENTRATE VOID AND QUALin' DISTRIBUTIONS FOR ANALYZED CORE M Avtal Power Factor i

Bottom of Core 1 0.38 2 0.69 3 0.93 4 1.10 5 1.21

6 1.30 L 7 1.47 8 1.51 9 1,49 10 1.44 11 1.36 12 1.28-13 1.16 14 1.06 i 15 1.01

pI -

f 16 17 0.97 0.94 18 0.97 19 0%

l- 20 0.91 21 0.77 22 0.59 i D OM

. Top of Core 24 0.12 ,

l 1 :'

l t

Amendment 15 4.49 l

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ABWR n^6im^n nov c Standard Plant I I I I I I I I I I I I I

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R R S 8 8 8 R S 8 ? R R S USMOd 1N33W3d d'4'II Amendment 15

ABWR uuima ;

Standard Plant nry e SECTION 4.5 -

CONTENTS ,

Section Tule Eage 4.5.1 Control Rod System Structural Materfah 4.5-1 4.5.1.1 Materials Specifications 4.51 4.5.1.2 Austenitic Stalnless Steel Components 4.5-2 4.5.13 Other Materials 4.52 4.5.1.4 Cleaning and Cleanliness Control 4.52 43.2 Reactor Internal Materials 4.53 1

4.5.2.1 Material Specifications 4.5-3 4.5.2.2 Controls on Welding 4.53 4.5.23 Nondestructive Examination of Wrought Seamless Tubular Products 4.53

- I O 4.5.2.4 Fabrication and Processing of Austenitic dL Stainless Stect Regulatory Guide Conformanes 4.5-4' 4.5.2.5 Other Materials 4.5-4 4.53 Interfaces 4.5-4.1

-t 4.53.1 CRD Inspection Program 4.5-4.1 t

.) .

Amendment 15

a ABWR mamn Standard Plant arv c SECTION 4.4 g i CONTENTS (Continued)

Section Ihle P. age 4.4.2.1.2 Reactor Coolant System Thermal ifydraulic 4.42 Data 4.4.2.13 Reactor Coolant System Geometric Data 4.4.2 4.4.2.2 Operating Restrictions on Pumps 4.4-2 4.4.23 Power Flow Operating Map 4.42 4.4.23.1 Limits for NormalOperations 4.42 4.4.23.2 Other Performance Characteristics 4.4-2 4.4.233 Regions of the Power Flow Map 4.43 4.4.2.3.4 Design Features for Power Flow Map 4.4-3 4.4.23.5 Flow Control 4.43 O

4.4.2.4- Thermal and 11ydraulic Characteristics 4.4 Summary Table i 4.43 Interfaces 4.4-4 -

)

4.43.1 Power Flow Operating Map _ 4.4-4 4.4.3.2 - Thermal Limits 4.4-4 i

h 4.4 iii O

A:nendmcot t3

ABWR m6imo RTV C Standard Plant years in BWR applications. Extensive laboratory tests have demonstrated that XM 19 is a suitable material and that it is resistant to stress corrosion in a BWR environment.

4.53 Interfaces 4.53.1 CRD inspection Program The CRD inspection progrem shallinclude provisions to detect incipient defects before they could become serious enough to cause operating problems. [See Subsection 4.5.1.2(2))

1 O

Amendment 15 4541

l ABWR 2246moin Standard Plant m -

j. SECTION 4.6 -

CONTENTS ;

Section 31th Eagt

- 4.6.1 Information for Control Rod Drive System 4.6 1 4.6.1.1 Design Bases 4.61 4.6.1.1.1 Safety Design Bases 4.61 ,

4.6.1.1.2 Power Generation Design Basis 4.6 1 i 4.6.1.2 Description 4.6-1 4.6.1.2.1 Fine Motion Control Rod Drive Mechanism 4.G2

(

4.6.1.2.2 FMCRD Components 4.6-3 4.6.1.2.2.1. Components for Fine Motion Control 4.63 4.6.1.2.2.2' Components for Scram 4.6-4 1

4.6.1.2.23 L FMCRD Pressure Boundary 4.6-4 4.6.1.2.2.4 Rod Position Indication 4.65-L 4.6.1.2.2.5 Scram Position Indication 4.65 1.

4.6.1.2.2.6 Control Rod Separation Detection 4.6-5 4.6.1.2.2.7 Bayonet Coupling 4.6-5 4.6.1.2.2.8 FMCRD Brake 4.6-6 4.6.1.2.2.9 _ Integral Internal Blowout Support . 466 4.6.1.2.2.10 Materials of Construction 4.6-6'

~ 4.6.1.23 Control Rod Drive Hydraulic System 4.6-6 l :E

l 4.6.1.23.1- Hydraulic Requirements 467 ?

4.6.1.23.2 System Description 4.67 4.6.1.23.2.1 CRD Supply Pump 4.67 4.6 il ,

I l 1

ABM 2mioara arv. c Standard Plant overall plant control characteristics, core thermal power limits, etc. The power flow map 4.4.2 Description of the Thermal and for the power range of operation shown in Figure Hydraulle Design of the Reactor Coolant 4.41 was used in the system response analyses System documented in Section 6.3 and Chapter 15. The specific power flow operating map for each plant will be provided to the USNRC for information.

I See Subsection 4.4.3 for interf ace 4A.2.1 Plant Configurstloe Data requirements. The nuclear system equipment, nuclear instrumentation, and the reactor 4A.2.1.1 Reactor Coolant System Configuration protection system, in conjunction with operating l-procedures, maintain operations within the area

! The reactor coolant system is described in of the operating map for normal operating l Seuion 5.4. conditions. The boundarie . on this map are as l 4A.2.1.2 Reactor Coolant Sy%m Thermal Hydraulk De:a Natural Circulation Line,0: The operating state of the reactor moves along this line The steady state distribution of temperature, for the normal control rod withdrawal pressure and flow rate for each flow path in the sequence in the absence of recirculation -

reactor coolant system is shown in Figure 5.11. pump operation.

l 4A.2.1.3 Reactor Coolant System Geometric Data 102% Power Rod Line ,or Rated Power (Whichever Is Less): The 102% power rod Volumes of regions and components within the line passes through 102% power at 100%

reactor vessel are shown'in Figure 5.12. flow. Some plaots may establish an operating map in which 102% power can be achieved at lower flow. The operating state -

Table 4.4 5 provides the flow path length, height, liquid level, minimum elevations, and for the reactor follows this rod-line (or minimum flow areas for each major flow path similar ones) during recirculation flow volume within the reactor vessel and changes with a fixed control rod pattern; recirculation loops of th i rea'ctor coolant however, rated power may not be exceeded.

- systems.

Steam Separator Limit Line: This line l 4A.2.2 Operating Restrictions on Pumps results from the' requirements to have acceptable moisture carryover fraction from Expected recirculation' pump performance steam separator.-

curves are shown in Figure 5.4 3 These curves L are valid for all conditions with a normal 4A.23.2 Other Performance Characteristice l L operation range varying from approximately 20 percent to 115 percent of rated pump flow. Other performance characteristics shown on the power flow operating map are: .

Subsection 4.4.2.3 gives the operating limits imposed on the recirculation pumps by cavitation, Constant Rod Lines A, B, C, D, E, F: These pump load, bearing design flow starvation, pump lines show the change in flow associated speed,'and steam separator performance, with power change while maintaining constant control rod position.

4A.23 Power Flow Operating Map Constant Pump Speed Lines 1,2, ?,4,5,6, 4A.23.1 Limits for Normal Operation 7, 8: These lines show the change in flow associated with power changt s w hile A BWR must operate with certain restrictions maintaining RIP speeds at a constaat speed.

because of pump net positive suction head (NPSH),

(4-2 O

Amendment t$

. ~ -

> ABWR- m-u Standard Plant nry c I SECTION 4.6 i

CONTENTS (Continued) 7 Ses119.n - Etic P. age

'4.6.23.4 Evaluation of Scram Time - 4.6-12 4.6.23.5 Analysis of Malfunction Relating to Rod Withdrawal 4.6-12 4.6.23.5.1 Drive Housing Failure 4.6-12 4.6.23.5.2 Rupture of Hydraulic Line(s) to Drive Housing Flange 4.6 12 4.6.23.53 Total Failure of All Drive Flange Bolts 4.6 12 4.6.23.6 Scram Reliability '4.6 13 4.63 Testine and Verification of the CRDs 4.613 4.63.1 ' Testing and Inspection - 4.6 13 4.63.1.1 Development Tests 4.6 13 -

, .[Q Factory Quality ControlTests 4.6 13 l 4.63.1.2-4.63.13 . OperationalTests 4.6 14 4.63.1.4' Acceptance Tests 4.6 14 ,

4.63.1.5 Surveillance Tests 4.6 15 4.63.1.6 FunctionalTests 4.6 15 s

=4.6.4 Information for Combined Performance of Reactivity Control Systems 4.6 16' 4.6.4.1 Vulnerability to Common Mode Failures 4.6-16 4.6.4.2 Accidents Taking Credit for Multiple

, Reactivity Systems 4.6 16 4.6.5 - Evaluation of Combined Performance. 4.6 16' 4.6 iv Amendment 15

ABWR mamn Standard Plant nrv c 4.4.3 Interfaces 4.4.3.1 Power Flow Operating Map The specific power flow operating map to be used at the plant will be provided by the utility to the USNRC for information.

4.4.3.2 Thermal Limits The thermal limits for the core loading at the plant will be provided by the utility to the USNRC for information.

O Amendment 15 4.44 9

l l

ABWR MA6100AD Standard Plant REV C (j 4.6 FUNCTIONAL DESIGN OF (5) Each positioning device shall provide a REACTIVITY CONTROL SYSTEMS means to prevent or limit the rate of control rod ejection from the core due to a The reactivity control systems consists of break in the drive mechanism pressure boun-control rods and controf rod drives, dary. This is to prevent fuel damage re-supplementary reactivity controlir. the form of sulting from rapid insertion of reactivity.

l burnable poison in the fuel, and the standby liquid control system (described in Subsection 4.6.1.1.2 Power Generation Design Basis 9.3.5) .

The control rod drive system (CRDS) design Evaluations of the reactivity control systems shall meet the following power generation design against the applicable General Design Criteria (GDC) are contained in the follov ing subsections; bases:

QDE Subseedon (1) The design shall provide for controlling changes in core reactivity by positioning 23 3.1.2.3.4 neutron absorbing control rods within the 5" 25 3.1.2/1.6 corc.

26 3.1.2.3.7 27 3.1.2.3.8 (2) The design shall provide for movement and 28 3.1.2.3.9 positioning of control rods in increments to 29 3.1.2.3.10 enable optimized power control and core 4.6.1 Information for Co'.itrol Rod Drive i System 4.6.1.2 Description L o.

l -- V 1-4.6.1.1 Design Bases The CRDS consists of fm' e motion control rod drive (FMCRD) mechanisms, and the CRD hydraulic 3: 4.6.1.1.1 Safety Design Bases system (including pumps, filters, hydraulic control. units, interconnecting piping, instru-The controi rod drive CRD mechanical system mentation and electrical controls). The CRDS, shall meet the following safety design bases: in conjunction with the rod control and infor-mation system (RC&lS) and reactor protection (1) . The design shall provide for rapid control system (RPS) performs the following functions:

rod insertion (scram) so that no fuel damage results from any moderately frequent event (1) Controls changes in core reactivity by

. (see Chapter 15). positioning neutron absorbing control rods within the cors in response to control (2) The design shall include positioning signals from the RC&IS.

devices, each of which individually supports and positions a control rod. (2) Provides movement and positioning of control rods in increments to enable optimized power (3) Each positioning device shall be capable of control and core power shape in response to holding the control rod in position and control signals from the RC&lS.

preventing it from inadvertently withdrawing outward during any non. accident, accident, (3) Provides the ability to position large post. accident and seismic condition, groups of rods simultaneously in response to control signals from the RC&lS.

(4) Each positioning device shall be capable of detecting the separatior. of the control rod (4) Provides rapid control rod insertion (scram) from the drive mechanism to prevent a rod in response to manual or automatic signals drop accident. from the RPS so that no fuel damage results from any plant transient.

.d Amendment 15 4.61

ABM 23461oorn arv. c Standard Plant Table 4.41 (Continued) g TYPICAL THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE BWR/6 General Oneratina Co editlans Lglg M*

Active coolant flow area gr a.ssembly (in.2) 15.164 15.678 Core average inlet wlo.ity (ft/sec) 6.98 6.43 Maximum inlet velor.ity (ft/sec) 8.54 7.45 Total core pressure drop (psi) 26.4 24.4 Dre support plate pressure drop (psi) 22.0 20.0 Average orifice pressure drop Central region (psi) 5.71 8.75 Average orifice preuure drop Peripheral region 18.68 17.69

- (psi)

Maximum channel pressure loading (psi) 15.40 10.9 Average Power assembly channel pressure loading 14.1 9.5 (bottom) (psi)-- l Shroud support ring and lower shroud pressure 25.7 23.9 loading Upper shroud pressure loading (psi) 3.7 3.5

Based on the core loading in Figure 4.3-J usedfor sensitivity studies.

44-6 O

Amendment 15 -

ABM DA6100AD Standard Plant RN C t

accumulator charging line prevents loss of water water header is indicated in the control room.

5 pressure in the event supply pressure is lost. An alarm is provided to indicate excessive differential pressure which will lif t the During normal plant operation, the accumulator control rods if not limited, piston is seated at the bottom of its cylinder.

Loss of nitrogen decreases the nitrogen pressure, The pump purge water is furnished to the RIPS which actuates a pressure switch and sounds an and CUW pumps at the required flow rates as alarm in the control room. specified on the CRD process flow diagram (Figure 4.6 9).

To ensure that the accumulato* is always able a produce a scram, it is continuously monitored To assure the continuous ability to trip, the for water leakage. A float. type level switch charging water header maintains the HCU actuates an alarm if water leaks past the piston accumulators at a high pressure. The scram barrier and collects in the accumulator valves remain closed except during and after instrumentation block. trip operation, so that no flow passes through the charging water header, f ressure in the 4.6.1.2A Control Rod Drive System Operation charging water header is monitored. A significant degradation in the charging header The operating modes of the control rod drive pressure will result in a low pressure warning system (CRD) are normal operation, rod insert. alarm and rod withdrawal block, and, eventually, tion / withdrawal, and scram. These operational reactor sc:am if further degradation occurs.

modes are described in the following sections.

4.6.1.2A.2 Control Rod lasertion/ Withdrawal 4.6.1.2A.1 Normal Operation Normal insertion and withdrawal of the l During normal operation the CRD provides the control rods is provided by the electric stepper

.f l motor on the FMCRDs. The motors receive their l proper amount purgingof flow plusrequired forA drive purging

'(d and for pump (RIPS CUW pumps). insert / withdraw commands via signals from the multi stage centrifugal pump pressurizes the rod control and information system (RC&lS). The system with water from the treated makeup water objective of the RC&lS is to provide the system (de oxygenated) and/or condensate storage operator with the means to make changes in tank, CRD pump minimum bypass flow to the . nuclear reactivity so that reactor power level condensate storage tank is utilized to prevent and distribution can be controlled. The system pump overheating if the pump discharge is allows the operator to manipulate control rods.

blocked. The total pump flow is the sum of this The design bases and further discussion of the '

bypass flow, the CRD purge flow through the flow .RC&lS are covered in Chapter 7.

control valve, and the recirculation and reactor water cleanup pump purge flows. A full capacity 4.6.1.2A3 Scram standby pump is available. Condensate water is 1: processed by filters (drive watc filters) in the Upon loss of electric power to both scram l

pump st.ction and discharge. A redundant set of ~ pilot valve solenoids, the associated HCU drive water filters are provided and are normally applies the . drive insert forces to its.

on standby. Differential pressure monitoring respective drives using a precharged accumulator :

devices and control room alarms are used to contained within the HCU (The N 2 /H2 O l monitor the filter elements to detect plugging of accumulator having previously been charged with L the filters. charging water from the CRD). The drives l l insert the control rod blades rapidly. The The purge water for each drive is provided water displaced from the drives is discharged through the purge water header. The purge water into the reactor vessel. Also, on receipt of a control valve automatically controls purge water. scram signal, each FMCRD will automatically flow to the drive mechanisms. Differential start and the ball out on each drive will be pressure between the reactor vessel and the purge driven upward by the screw shaft until it is I r V) Amendment 15 4.69 l.

l l

i i

-ABM zwioorn Standard Plant REV C j Table 4.4 3 FLOW QUALITY DISTRIBUTIONFOR ANALYZED CORE

! Core Average Value 0.145 Maximum Exit Value . 0.258 Active Fuel Insth .146 laches Core Average Maximum Channel Egdg (Averman Node Value) (End of Node Valne) i l

Bottom of Core 1 0 0 ;

2 0 0

.i 3 0 0.002 4 0 0.008 5 0 0.019 6 0.005 0.033 7 0.016 0.051 8 0.028 0.069 9 0.040 0.087 10 0.052 0.105 11 0.064 0.122 12 0.074 0.137 13 0.085 0.152 14 0.094 - 0.165 15 0.103 0.177 -

16 0.110 0.189 17 0.118 0.200 18 0.126 0.212

. 19 0.134 0.224 20 0.142 0.235 21 0.148 0.245 22- 0.154 0.252 .

s 23 0.158 0.257 '

Top of Core 24 0.159 . 0.258 bU Amendment 15 4.4-8 l

Y ABM nwoorn Standard Plant uvc f.

C.

l 4.6.2.3.2 Functional Fallurv of Critical operation. The control rod cannot rotate Components since it is always constrained between four fuel assemblies and the hollow piston has The consequences of a functional failure of rollers which operate in a track within the ,

CRD system critical components have been FMCRD. Only structural failure would permit !

evaluated and the results are covered in or result in control rod to FMCRD uncoupling Subsection 4.6.2.3.2.2. which in turn could only result in rod drop

,. If the redundant switches failed to sense l 4.6.2.3.3 Precluding Excessive Rate of separation. In that case the rate of rod Reactivity Addition - drop could exceed acceptable reactivity addition rates, but the combined failures of Excessive rates of reactivity addition are the structure and the redundant switches is precluded in the design of the FMCRD, Prevention incredible because the components are of rod ejection due to FMCRD pressure boundary designed and tested to prevent structural failure is described in Subsection 4.6.2.3.2.2. failure through mechanical damage as Rod drop is prevented through: described in the following Subsection.

(1) Two redundant Class 1E switches in the FMCRD sense separation of the hollow piston, which positions the control rod, from the ball nut. These switches sense either separation of the piston from the nut or separation of the control rod from the piston, and block further lowering of the nut thereby p preventing drop of either the control rod or the control rod and hollow piston as an G assembly (Refer to 4.6.1.2.2.6 Lor further details).

.(2) Two redundant spring. loaded latches on the I hollow piston open to engage in openings in the guide tube within the FMCRD to catch the hollow piston if separation from the nut were to occur. .These latches open to support the hollow piston (and control rod) following every scram until the ball nut is run in to provide the normal support for the hollow piston (and control rod).- The rate at which the hollow piston (and control rod)

I would drop is less than the allowable rate p as evaluated in Chapter 15.

l (3) The control rod to hollow piston coupling is a bayonet type coupling. Coupling is verified by pull test for the control rod H upon initial coupling at refueling and again each time an attempt is made to drive beyond the " full out' position during reactor operation. The control rod can only be uncoupled from the FMCRD by relative rotation which is not possible during 1

Amendment t$ 4.6-11

l ABWR mama l Standard Plant nev. e Table 4.4 5 g REACTOR COOLANT SYSTEM GEOMETRIC DATA Height Elevation Flow and of Bottom Average Path. Uguld of Each Flow 14ngth level Volume

Areas

. find Ilad _ find ifL81.

A. Lower Plenum 182.9 182.9 0.0 209.8 182.9 B. Core 171.5 171.5 182.9 172.8 "

171.5 includes bypass C. Upper Plenum and 143.2 143.2 354.4 177.9 Separators 143.2 D. Dome (Above Normal 307.1 30'.1 520.0 325.4 Water Level) 0 E. Downcomer Area 497.8 497.8 72.5 174.3 497.8 9

i

}

Reference Point is vessel bottorn zero,

" For the core loading given in Figure 4.3-1 used for sensitivity studies. ,

Amendment IS ( 4 10 9

ABM ur6ioorn REV C Standard Plant O device which engages when the lower housing taneous with the initiation of hydraulic (spool piece) is removed for maintenance. This fast scram. This provides a diverse means device prevents rotation of the spindle and hence to assure control rod insertion.

control rod motion when the spool piece is removed. The two components of the anti rotation 4.6.3 Testing and Verification of the CRDs device are (1) the upper half of the coupling between the lower housing drive shaft and ball 4.6.3.1 Testing and Inspection spindle, and (2) the back seat of the middle flange (see Figure 4.61). The coupling of the 4.63.1.1 Development Tests lower housing drive shaft to the ball spindle is splined to permit removal of the lower housing. The initial development of the FMCRD invoked The underside of the upper coupling piece has a testing of a prototype based on a European drive circumferentially splined surface which engages design. Testing of this prototype included more with a mating surface on the middle flange back than 600 scrams and 67,000 motor driven cycles, seat when the spindle is lowered during spool A subsequent prototype was developed for piece removal. When engaged, spindle rotation is installation in an operating BWR for the purpose prevented. In addition to preventing rotation, of demonstrating FMCRD performance under actual this device also provides scaling of leakage from BWR operating conditions. This in. plant FMCRD the drive while the spool piece is removed, in prototype has been tested extensively prior to the unlikely event of the total failure of all installation at the operating plant, including the drive flange bolts, the anti. rotation device over 500 scrams and 63,000 step cycles. The will engage when the spool piece falls, thus inplant FMCRD is currently being installed at La preventing rod ejection. Salle Unit 2 where it will be tested for one complete operating cycle.

4.6.2.3.6 Scram Reliability 9 A reference FMCRD prototype design, based on High suam reliability is the result of a refinements of initial development prototypes numbr. of features of the CRD system. For described above, has been developed and tested, e x *..n ple : To date, testing of this reference prototype has included over 1,000 serams and 150,000 step (1) Each accumulator provides sufficient stored cycles. These tests have demonstrated the energy to scram two CRDs at any reactr following:

pressure.

(1) The drive casily withstands the forces,

-(2) Each pair of drive mechanisms has its own pressures and temperatures imposed, scram valve and dual solenoid scram pilot valve; therefore, only a single scram valve (2) No abnormal distortion or deformation was needs to open for scram to be initiated, found. Wear, abrasion and corrosion was Both pilot valve solenoids must be negligible.

-de energized to initiate a scram.

(3) The basic scram speed of the drive has a (3) ' The RPS and the HCUs are designed so that satisfactory margin above minimum plant the scram signal and mode of operation requirements at any reactor vessel pressure, override all others.

4.63.1.2 Factory Quality Control Tests (4) The hollow piston and guide tube are designed so they will not restrain or The detailed quality control specifications prevent control rod insertion during scram, and procedures have not as yet been prepared but will follow the general pattern established for

($) Each CRD mechanism initiates electric motor such specifications and procedures in BWRs driven insertion of its control rod simul- presently under construction and in operation.

Amendment 15 4.6 13

ABM 23rraoorn Standard Plant Rf'V C nuclear instrumentation; and ;

The af ter installation, pr sstartup tests (Chapter 14) include normal and se am motion and (b) when the rod is fully withdrawn the are primarily intended to verile that piping, first time, observe that the drive will valves, electrical components and iratrumentation not go to the overtravel position.

are properly installed. The test specifications Observation of the separation switches include criteria and acceptable ranges for drive provides direct indication that the speed, scram valve response times, and control control rod is following the drive pressures. These are tests intended more to during withdrawal, but does not provide i document system condition rather than tests of a direct check on coupling integrity, performance. Additionally , observation of a response from the nuclear instrumentation during

[

- As fuel is placed in the reactor, the startup an attempt to withdraw a control rod test procedure (Chapter 14) is followed. The provides another indirect indication tests in this procedure are intended to that the rod and drive are coupled. The demonstrate that the initial operational overtravel position feature provides a characteristics meet the limits of the positive check on the coupling specifications over the range of primary coolant integrity, for only an uncoupled drive temperatures and pressures from ambient to can reach the overtravel position, operating. The detailed specifications and procedures are similar to those in BWRs presently (4) During operation, accumulator pressure and under construction and in operation. level at the normal operating value are verificd.

t 4.63.1.5 Surveillance Tests Experience with CRD systems of the same type ,

i The surveillance requirements (SR) for the CRD indicates that weekly verification of system are described below. While these accumulator pressure and level is sufficient requirements have not yet been formalized the to assure operability of the accumulator intent is to follow the general pattern portion of the CRD system.

_ catablished for surveillance testing in BWRs presently under construction and in operation. (5) At the time of each major refueling outage, each operable control rod is subjected to (1) Sufficient control rods shall be withdrawn, scram time tests from the fully withdrawn following a refueling outage when core position.

_ alterations are performed, to demonstrate with adequate shutdown margin that the core Experience indicates that the scram times of can be made suberitical at any time in the the control rods do not significantly change subsequent fuel cycle with the maximum worth over the time interval between refueling 7 control rod pair having the same HCU or the outages. A test of the scram times at each single rod attached to the unpaired HCU,if refueling outage is sufficient to identify of greater worth withdrawn and all other any significant lengthening of the scram operable rods fully inserted. times.

(2) Each partially or fully withdrawn control 4.63.1.6 Functional Tests rod is exercised one or two steps at least once each week, The detailed requirements for functional testing have not as yet been formally (3) The coupling integrity shall be verified for established, but the intent is to follow the each withdrawn control rod as follows: general pattern established for such testing in BWRs currently under construction and in

~

(a) when the rod is first withdrawn, observe operation. As such, it is anticipated that the the control rod separation switch functional testing program of the CRDS will response and discernible response of the consist of a five year maintenance life test and Amendment 12 4 15

l 1.

i-L - APPENDIX 4A -

L i TYPICAL CONTROL ROD PATI' ERNS i AND l ,

' ASSOCIATED POWER DISTRIBUTION .

FOR ABWR-p

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Standard Plant nrv. c Ih APPENDIX 4A CONTENTS .

1 Section 11111 East L

4A TYPICAL CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR ABWR L' 4A.1 Introduction 4A.1 L

4A.2 Power Distribution Strategy 4A.3 4A.3 Results of Core Simulation Studies 4A 4 L ~ 4A.4 References 4A.46 L

TABLES Table ~11tle East t L

4A.11 Basic Control Strategy for Typical ABWR 4A 2 4A.3-1 Incremental Exposure Steps and Related LM Figure Numbers 4^.5 i

ILLUSTRATIONS Figure 11tle Ear:

4A 3-la Summary of Haling Condition 100% Rated -

Core Flow 4A.6

4A.3.lb - Relative Axial Power at 8.5 GWd/MT Cycle Exposure (Haling) 4A.7 4A.31c - Relative Axial Exposure at 8.5 GWd/MT Cycle Exposure (Haling) 4A 7 ;

4A.3-1d - Integrated Power per Bundle (Haling) at 8.5 GWd/MT Cycle Exposure 4A.8 4A.3-le Average Bundle Exposure (Haling) at 8.5 GWd/MT Cycle Exposure 4A.8

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Amendment 15

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l ABWR mama l Standard Plant REV.C SECTION 4.6 g'

CONTENTS (Continued) l Due East Section 4.6.1.23.2.2 Accumulator Charging Pressure 4.67 4.6.1.23.23 Purge Water Header 4.6-8 4.6.1.23.2.4 Hydraulic Control Units 4.6-8 4.6.1.2.4 Control Rod Drive System Operation 4.6-9 4.6.1.2.4.1 Normal Operation 4.69 4.6.1.2.4.2 Control Rod Insertion / Withdrawal 4.6-9

'4.6.1.2.43 Scram 4.6-9 4.6,1.2.5 - Instrumentation 4.6-10 4.6.2 - Evaluation of the CRDs 4.6 10 ,

L4.6.2.1- Failure Mode and Effects Analysis 4.6-10 4.6.2.2 Protection from Common Mode Failures 4.6 10 i 4.6.23 Safety Evaluation 4.6 10 <

4.6.23.1 Forces for Expulsion 4.6-10

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f u 4.6-11 =t 4.6.23.2 Functional Failure of Critical Components

4.6.233 - Precluding Excessive Rates of Reactivity Addition 4.6-11

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4 . 6 . i11 Amendment t5

ABWR msimr, RM C Standard Plant (3

. (j . 4A.1 INTRODUCTION This appendix contains a typical simulation of l an equilibrium cycle. The control rod patterns used are just one example of a set of control rod l patterns which radial and axial couldneeded power shaping be used to provide the to meet tbc Technical Specifications.

The basic control rod strategy for this case is provided in Table 4A.11.

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Amendment t$ 4A 1

ABWR m6me REY.C Standard Plant SECTION 4.6 $

ILLUSTRATIONS Eignot Ihle East 4.61 Control Rod Drive Schematic 4.6 17 4.6-2 Control Rod Drive Unit (Cutaway) 4.6 18 4.63 Full in Indication 4.6-19 4.6 4 CRD Separation Detection 4.6-20 4.6-5 Control Rod to Control Rod Drive Coupling 4.6 21 4.6-6 FMCRD Centrifugal Brake 4.6-22 4.6-7. InternalBlowout Support Schematic 4.6-23 4.6-8 CRD System P&ID 4.6 24 4.69 ' CRD System Process Diagram 4.6-26 9

4.6-v Amendment 13

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APPENDIX 4B FUEL LICENSING ACCEPTANCE CRITERIA ;

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i (5) Gathers rod status and rod position data for rod pattern control, performance monitoring,

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operator display and scram time testing. j i

(6) Prevents undesirable rod pattern or rod i motions by imposisg rod motion blocks in order to protect the fuel. ;

(7) Psevents and miticstes the consequences of a rod drop accident by detecting rod ,

separation and controlling rod pattern. j I

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Amendment 3 4g O

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ABWR mamn

$1RDdard Plant ._

nu e SECTION 4B C

CONTENTS

[ Sesitan Ihle East 4B FUEL LICENSING ACCEPTANCE CRITERIA 48.1 tetroduction 4B 1

General Criteria 4B 1 48.2 4B.3 Derinal Mechanical 4B 1 h

48.4 Nuclear 4B 3 48.5 Hydraulle 4B 4 48.6 Safety unit MCPR 4B 4 j 48.7 Operating unit MCPR 4B 5 48.8 Critical Power Comlation 4B 5 48.9 Stability 4B 6 48.10 Overpmsure Protection Analysis 4B-6 48,11 Ims Of Coolant Accident Analple Me'tbods 4B 7 48.12 Refueling Accident Analpla 4B 7 48.13 Usamptable Results For lefrequent incidents (Unexpected Operstloa Occumaces) 4B 7 48.14 Referrom 4B 7 4B il Amendment 13

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ABM usuoarn nrv e Standard Plant back in contact with the hollow piston. This completes the rod full.in Insertion and places the CRDS in a condition ready for restarting the reactor.

After reactor scram, Indication that the scram has gone to completion (rods in full in position) is displayed to the operator.

Following scram completion (each ball out re engaged with its hollow piston), the scram signals are reset and each accumulator is rechstged with water from CRDS.

4.6.1.2.8 lettrumentation The Instrumentation for the CRDS is defined on the system P&lD, Figure 4.6 8. Supervisory instrumentation and alarms such as accumulator trouble, low charging header pressure, purge water / reactor vessel high differential pressure are adequate and permit surveillance of the CRD System's readiness.

The design bases and further discussion are covered in Chapter 7.

4.6.2 Evaluations of the CRDS 4.6.2.1 Failure Mode and Effects Analysis This subject is covered in Appendix 15B.

4.6.2.2 Protection from Common Mode Fallum The position on this subject is covered in Appendix 15B.

44.2J.1 Forces for Expulsion 4.6.23 Safety Evaluation An analysis has been performed which The safety evaluation of the control rod evaluates the maximum pressure forces which drives is given below, could tend to eject a control rod from the core. The results of this analysis are given in Subsection 4.6.2.3.2.2.2. In summary, if the check valve in the drive flange were to fall to close, which is unlikely, calculations and tests indicate that the maximum ejection velocity would exceed allowable rates. However, a centrifugal brake in the FMCRD stops the rod ejection before this allowable ejection velocity is exceeded. These forces are similar to scram acceleration forces.

Amendment is u.to

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ABM 2mioun Standard Plani uvc O

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! GE PROPRIETARY provided under separate cover

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4B I through 4B . 7 1$

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ABM 234suoxn nry c Standard Plant 4.6.2JJ.2 Rupture of Hydrsulle une to l Drise Housing Flange For the case of a scram insert line break, a partial or complete circumferential opening is postulated at or near the point where the line enters the housing flange. This failure, if not mitigated by special design features, could result in rod ejection at speeds exceeding maxi.

mua allowable limi.s of 4 in/sec (assuming rod pattern control) or 6 inches maximum travel distance before full stop. Failure of the scram 4.6.2J.4 Evaluation of Scram Time insert line would cause loss of pressure to the underside of the hollow piston. The force The rod scram function of the CRD system resulting from full reactor pressure acting on provides the negative reactivity insertion the cross sectional area of the hollow piston, required by safety design basis 4.6.1.1.1(1), plus the weights of the control rod and hollow The scram time shown in the description is piston, is imposed on the ball nut. Tbc ball reflected in plant transient analyses (Chapter nut in turn translates this resultant force into 15).

a torque acting on the spindle. When this torque exceeds the motor residual torque and l 4.6.23J Analysis of Malfunction Relatlas to seal friction, reverse rotation of the spindle Rod Withdrawal will occur permitting rod withdrawal Analyses show that the forces generate'd during this post.

e There are no known single malfunctions that ulated event can result in rod ejection speeds I cause the unplanned withdrawal of even a single which exceed the maximum allowable limits.

control rod. However,if multiple malfunctions are postulated, studies show that an unplanned The FMCRD design provides two diverse means rod withdrawal can occur .t withdrawal speeds of protection against the results of a that vary with the combination of malfunctions postulated scram insert line failure. The first postulated. means of protection is a ball check valve located in the middle flange of the drive at the 4.6.2.3J.1 Drive Housing Failure scram port. Reverse flor during a line break l

will cause the ball to move to the closed The bottom head of the reactor vessel has a position. This will prevent loss of pressure to penetration for each CRD location. A drive the underside of the hollow pis:on, which in housing is raised into position inside each turn sill prevent the generation of loads on the penetration and fastened by welding. The drive drive which could cause rod ejection.

is raised into the drive housing and bolted to a flange at the bottom of the housing. The second means of protection is the FMCRD brake described in Subsection 4.6.1.2.2.8. In in the unlikely event of a failure of the the event of the failure of the check valve, the drive housing to vessel attachment welo passive brake will prevent the ball spindle (including a failure through the housing or along rctation and rod ejection, the fusion line of the housing to stub tube weld) or the flange bolting attaching the drive to the .4.6.23JJ Total Fallure of All Drhe l housing, ejection of the CRD and attached control Flange Bolts rod is prevented by the integral internal blowout support. The details of the this internal The FMCRD design provides an anti. rotation blowout support structure are contained in Seetion 4.6.1.2.2.9.

4612 O

Amendmcat 13

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APPENDIX 4C CONTROL ROD LICENSING ACCEPTANCE CRITERIA O .

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ABWR w Slagdard Plant MV A Quality control of welding, best treatment, 4.6.3.1.3 operstionalTests ,

dimensional tolerances, material verification and similar factors is maintained throughout the Af ter installstion, all rods and drive manuf acturing process to assure reliable mechanisms can be tested through their full !

performance of the mechanical reactivity control stroke for operability.

components. Some of the quality control tests performed on the CkD mechanisms and HCUs are The switches which detect separation will l listed below- .

provide indication and automatic tod withdrawal j block should a control rod separate from the (1) CRD mechanism tests:

drive mechanism during rod withdrawal.

Additionally, the operator can observe the (a) Pressure welds on the drives are in. core monitor indications to verify that the .

i hydrostatically tested in accordance with control rod is following the drive mechanism, ASME codes. All control rods that are partially withdrawn from the core can be tested for rod following by (b) Electrical components are checked for inserting or withdrawing the rod one or two electrical continuity and resistance to steps and retusning it to its original position, ground, while the operator observes the in core monitor indications.

(c) Drive parts that cannot be visually inspected for dirt are flushed with filtered To make a positive test of control rod to CRD water at high velocity. No significant coupling integrity, the operator can withdraw a foreign material is permitted in effluent control rod to the end of its travel and then water, attempt to withdraw the drive to the overtravel position. Failure of the drive to overtravel (d) Drive shaft seals are tested for leakage to demonstrates the lategrity of the rod.to. drive demonstrate correct seal operation. coupling.

(c) Each drive are tested for shim motion and Hydraulle supply subsystem pressures can be control rod position indication, observed from instrumentation in the control room. Scram accumulator pressures can be (t) Each drive is subjected to cold scram tests observed on tbc nitrogen pressure gages, at various reactor pressures to verify correct scram performance. 4.6.3.1.4 Acceptance Testa (2) HCU tests: Criteria for acceptance of the individual CRD mechanisms and the associated control and (a) Hydraulic systems are hydrostatically tested protection systems are incorporated in '

in accordance with the applicable code. specifications and test procedures covering three distinct phases: (1) pre.installationt (2)

(b) Electrical components and systems are tested after installation prior to startup; and (3) l for electrical continuity and resistance to during startup testing, ground.

The pre. installation specification defines (c) Correct operation of the accumulator criteria and acceptable ranges of such pressure and level switches are verified. characteristics as seal leakage, friction and scram performance under fixed test conditions (d) The unit's ability to perform its part of a whleh must be met before the component can Se scram is demonstrated, shipped.

4 614 9

l-M\N 2.M6100AD Standard Plant an. e O SEcTiON 4C CONTENTS ,

Section Illig East 4C CONTROL ROD LICENSING ACCEPTANCE CRITERIA 4C.1 latroduction 4C 1 4C.2 General Criteria 4C 1 4C3 Basis For Acceptahce 4C 1 4C3.1 Streu, Strain and Fatigue 4C 1 4C3.2 Control Rod Insertion 4C 1 l

l 4C33 Control Rod Material 4C 1 4C3.4 ReactMty 4C-1 4C.4 References 4C 1

(

l I- 4C Il Amendment 15 l -

. 1 ABWR m6ms avc Standard Plant a 1.5 times design life test program. In addition to maintenance and design life tests, I the program covers crud / contamination testing. I seismic misalignment, channel bulge, failed buffer, rod drop (to test hollow piston latch functionality), and rod ejection (to test FMCRD brake functionality).

4.6.4 Information for Combined Performance of Reactivity Control Systems 4.6.4.1 Yulnerability to Common Mode Failures The reactivity control system is located such that it is protected from common mode failures due to missiles, fallures of moderate and high energy piping, and fire. Sections 3.4,3.5 and 3.6, and Subsection 9.5.1 discuss protection of essential systems against missiles, pipe breaks and fire.

4.6.4.2 Acc! dents Taking Credit for Multiple ReactMty Systems There are no postulated accidents documeued l in Chapter 15 that take credit for two or more reactivity control systems preventing or mitigating each accident.

4.6.5 Evaluation of Combined Perforraance As indicated in Subsection 4.6.4.2, credit is not taken for multiple reactivity control systems for any postulated accidents documented in Chapter 15.

l 4616 9

Amendment 13

ABM 2mimsn mye Standard Plant O

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4C 1 Amendment 15 I

i ABM nwoorn Standard Plant arv c CHAPTER 5 TABLE OF CONTENTS Section M Pagg 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIITION 5.11 ,

5.1.1 Schematic Flow Diagrams 5.12 5.1.2 Piping and lastrumentation Diagrams 5.12 5.13 Elevation Drawings 5.12 5J th"IIGRITY OF REACTOR COOLAhT PRESSURE BOUNDARY 5.21 5.2.1 Compliance with Codes and Code Cases 5.21 5.2.2 Overpreuure Protection 5.21 5.23 Reactor Coolant Preuure

't e Boundary Materials 5.27 5.2.4 Presenice and insenice Inspection and Testing of Reactor Coolant Preuure Boundary 5.2 15.1 5.2.5 Reactor Coolant Preuure Boundary and Core Cooling Systems Leakage Detection 5.2 18 5.2.6 Interfaces 5.2 28 5.2.7 References 5.2 28

$3 REACTOR VESSEL 53 1 l

p 53.1 Reactor Vessel Materials 53 1 53.2 Pressure / Temperature Limits 53 6 l 533 Reactor VesselIntegrity 53-7 l 53.4 Interfaces $3 12 l:

53.5 References 53 12 5.0 il Amendment 13 1

MM Standard Plant 23A6tooAB nev. e l

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CHAPTER 5 g TABLE OF CONTENTS (Continued) i Section Iitle East 5.4 COMPONENT AND SUBSYSTEM DESIGN $.41 5.4.1 Reactor Recirculation System 5.41 5.4.2 Steam Generators (PWR) 5.45 l l

5.4.3 Reactor Coolant Piping 5.45 l 5.4.4 Main Steamline Flow Restrictors 5.45 l

' 5.4.5 Main Steamline Isolation System 5.4-6 5.4.6 Reactor Core Isolation Cooling 5.4 10 l System 5.4.7 Residual Heat Retr. oval System 5.4 16 5.4.8 Reactor Ws.er Cleanup System 5.4-24 ;

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1 5.4.9 Main Steamline.: and Feedwater 5.4 27 Piping l l

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5.4.10 Pressurizer 5.428 5.4.11 Pressurizer Relief Discharge 5.4 28 System f 4.12 Valves 5.4 28 1 4.13 Safety / Relief Valves 5.4-29 5.4.14 Component Supports 5.4 30 5.4.15 References 5.4 30 A?'ENDIX 5A DETECTION AND SIZING CAPABILI'IY TEST FOR REGULATORY GUIDE 1.150 5.0-lii Amendment 15 O

ABM as6>oorn

-- Standard Plant ..BI:y C SECTION 5.1 CONTENTS Sectlon HLle P.agt 5.1.1 Schematic Flow Diagrams 5.12 512 Piping and instrumentation Dianrams 5.12 5.1.3 ElesatIon Drawinns $.12 ILLUSTRATIONS Hgurt Elle East 5.11 Rated Operating Conditions of the ABWR 5.13 ,

5.12- Coolant Volumes of the ABWR 5.14 5.13 Nuclear Boiler System P&lD 5.15 ,

5.14 (Deleted) 3.1 10 i

5.1 ii O Amendment 15 l

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ABWR momn Standard Plant RPV H

5. REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTION the suppression pool. The pressure relief ,

system also acts to automatically depressurire the nuclear system in the event of a The reactor coolant system (RCS) includes loss of coolant accident (LOCA) in which the those systems and companents which contain or feedwater, reactor core isolation cooling (RCIC) transport fluids coming from, or going to the and high pressure core flooder (HPCF) systems l scactor core. These systems form a major portion fall to maintain reactor vessel water level.

of the reactor cof, ant pressure boundary (RCPB). Depressurization of the nuclear system allows This chapter provides information regarding the the low pressure flooder systems to supply reactor coolant system and pressure containing enough cooling water to adequately cool the appendages out to and including isolation fuel.

valving. This grouping of components is defined as the RCPB. Subsection 5.2.5 establishes the limits on nuclear sy % k.*..ge inside the drywell so The RCPB includes all pressure containing th . appropriate act'on can be taken before the components such as pressure vessels, piping, integrity of the nuclear system process barrier pumps, and valves, which are: is impaired.

(1) part of the RCS, or The reactor vessel and appurtenances are described in Section 5.3. The major safety (2) connected to the RCS up to and including any wasideration for the reactor vessel is the

/q and all of the following: - ability of the vessel to function as a

_ D) radioactive material barrier. Various (a) ihe outermost containment isolation combinations of loading are considered in the va've in piping which penetrates primary vessel design. The vessel meets the reactor containment, requirements of applicable codes and criteria.

The possibility of brittle facture was (b) the second of the two valves normally considered, and suitable design, material closed during normal reactor operation selection, material surveillance activity, and in system piping which doet not operational lin4its were established that avoid penetrate primary reactor containment, conditions where brittle fracture was possible.

and The reactor recirculation system (RRS)

(c) the RCS safety / relief valve (SRV) provides coolant flow through the core, piping. Adjustment of the core coolant flow rate changes reactor power output thus providing a means of

~ This chapter also deals with various following plant load demand without adjusting subsystems to the RCPD which are closely allied control rods. The RRS is designed to provide a to it. Specifically, Section 5.4 presents these slow coastdown of flow so that fuel thermal subsystems, limits cannot be exceeded as a result of recirculation system malfunctions. The reactor The nuclear synem rressure relief system recirculation pumps are located inside the protects the RCFB frori damage due to over- reactor vessel, thus eliminating large piping pressure. To protect ; gainst overpressure, connections to the reactor vessel below the core pressure-operated relief valves are provided that and also eliminating the reactor recirculation can discharge steam from the nuclear system to- system piping.

O V

Amendment 2 $.1 1

ABWR mn=a Standard Plant uvc The main steamline flow restrictors of the a portion of reactor coolant through a filter-venturi type are lastalled in each main steam dcminerallrer to remove particulate and dis-nozzle on the reactor vessel inside the primary solved impurities with their associated corro- ::

containment. The restrictors are designed to sion and fission products from the reactor cool- ,M limit the loss of coolant resulting from a main ant. It also removes excess coolant from the steamline break inside or outside the primary reactor system under controlled conditions, containment. The coolant loss is limited so that reactor vessel water level remains above the top 5.1.1 Schematic Flow Diagrams ,

of the core during the time required for the main steamline isolation valves to close. This action Schematic flow diagrams (Figures 5.11 and protects the fuel barrier. 5.12) of the RCS show major components, principal pressures, temperatures, flow rates, Two isolation valves are installed on each and coolant volumes for normal steady state main steamline. One is located inside, and the operating conditions at rated power.

other is located outside the primary containment. If a main steamline brcsk occurs 5.1.2 Piping and Instrumentation Diagrams inside the containment, closure of the isolation valve outside the primary containment seals the Piping and instrumentation diagrams covering prirnary containment itself. The main stenmline the systems included within RCS and connected isolation valves automatically isolate the RCPB systems are presented as follows:

when a pipe break occurs outside containment.

This action limits the loss of coolant and the (1) the nuclear boiler system (Figurc 5.13);

release of radioactive materials from the nuclear i syrtem. (2) main stcam (Figure 5.13b & c); I !

The RCIC system provides makeup water to the (3) feedwater (Figure 5.13d);

core during a reactor shutdown in which fcedwater ,

flow is not available. The system is started (4) recirculation system (Figure $.4-4); i automatically upon receipt of a low reactor water level signal or manually by the operator. Water '(5) reactor core isolation cooling system is pumped to the core by a turbine pump driven by (Figurc 5.4 8);

reactor &tCam.

(6) residual heat removal sysicm (Figure The residual heat removal (RHR) system 5.410)1 and includes a number of pumps and heat exchangers that can be used to cool the nuclear system under (7) reactor water cleanup system (Figure 5.4-12) a variety of situations. During normal shutdown and reactor servicing, the RHR system removes 5.1.3 Elevation Drawings residual and decay heat. The RHR system allows decay heat to be removed whenever the main heat The elevation drawing showing the principal l sink (main condenser) is not available (l.c., hot dimensions of the reactor and connecting systems standby). One mode of RHR operation allows the in relation to the containment are provided in removal of heat from tt %ary containmert Figures 1.2 2 and 1.2 3.

following a LOCA. Another operational mode of the RHR system is low pressure flooder (LPFL).

The LPFL is an enginected safety feature fos use during a postulated LOCA. Operation of the LPFL is presented in Section 6.3.

The reactor water cleanup system recirculates O

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2 CORE OUTLET 73.8 52 2x10 288 358.3 3 SEPARATOR OUTLET (STEAM DOME) 73.1 7.64x108 287 661.6 STEAMLINE (2ND ISOLATlON

4 70.3 7.64 x10' 285 661.6 VALVE)

FEEDWATER INLET (INCLUDES 6 0 74.9 7.78x10 216 221.1 CLEANUP RETURN FLOW)

, Figure 5.1 1 RATED OPERATING CONDITIONS OF THE ABWR ,

Amendment 15 5.13

ABWR naim4s Standard Plant arv.c 3

VOLUME OF FLUID (M )

A LOWER PLENUM 102 O l l'

B CORE 70 C UPPER PLENUM AND SEPARATORS 60 D DOME (ABOVE NORMAL WATER LEVEL) 236 ;

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MM 21A6100All Remndard Plant REV. A SECTION 5.2 CONTENTS

-. m e.

8.2.1 Comnllance with Codes and Coden Cases 5.21 5.2.1.1 Compliance with 10CFR50,Seetion 50.55a 5.21 5.2.1.2 Applicable Code Cases 5.21 8.2.2 Overnressure Protection 5.21 5.2.2.1 Design Basis 5.21 512.1.1 Safety Design Bases 5.21 512.1.2 Power Generation Design Bases 5.21 5.2.2.13 Discussion 5.21 5.2.2.1.4 Safety / Relief Yalve Capacity 5.22 5.2.2.2 Design Evalualion 5.22 5.212.1 Method of Analysis 5.22 5.2.2.2.2 System Design 5.22 5.2.2.211 Opcisting Conditions 5.22 5.2.2.212 Transients 5.22 512.2.23 Safety / Relief Valve Transient Analysis Specification 5.23 512.2.2.4 Safety / Relief Valve Capacity 5.23 5.2.2.23 Evaluation of Results 5.23 5.2.2.23.1 Safety / Relief Valve Capacity 5.23 5.2123.2 Pressure Drop in inlet and Discharge 5.23 O

5 . 2 11

ABWR Etandard Plant mumin REY C SECTION 5.2 CONTENTS (Continued)

Section lula P.ast 5.2.23 Piping and Instrument Diagrams 5.23 5.2.2.4 ' Equipment and Component Description 5.24 5.2.2.4.1 Description 5.24 ,

5.2.2.4.2 Design Parameters 5.2-6 5.2.2.43 Safety / Relief Valve 5.26 i 5.2.2.5 Mounting of Safety Relief l Valves 5.26 ,

1 5.2.2.6 Applicable Codes and Classification 5.26 5.2.2.7 Material Specifications 5.2-6 5.2.2.8 Process Instrumentation 5.27 1

5.2.2.9 Sptem Reliability 5.27 5.2.2.10 Inspection and Testing 5.27 )

1 5.23 Reactor Coolant Pressurr ,

Boundary Materials 5.27 l 5.23.1 Material Specifications 5.27 1 5.23.2 Compatibility with Reactor Coolant 5.27 5.23.2.1 PWR Chemistry of Reactor Coolant 5.27 1

5.23.2.2 BWR Chemistry of Reactor l Coolant 5.27 5.23.2.2.1 Fuel Performance Considerations 5.28 5.29 I 5.23.2.2.2 Radiation Field Buildup 5.2 iii Amendment 13

ABM zwioorn Standard Plant nrv c l

7" SECTION 5.2

( CONTENTS (Continued)

Sectlon Ette East l

5.23.2.23 Sources ofimpurities 5.29 5.23.2.2.4 IASCC Considerations 5.2 12 l 5.23.23 Compatibility of Construction Materials with Reactor Coolant 5.2 12 l

5.2.3.2.4 Compatibility of Construction Materials with External Insulation 5212 l 5.233 Fabrication and Processing of Ferritic Materials 5.2 12 i 5.233.1 Fracture Toughness 5.2 12 5.233.2 Control of Welding 5.213

[ 5.233.2.1 Regulatory Guide 1.50: Contrel of Preheat Temperature Employed y) for Welding ofI.ow Alloy Steel 5.213 5.233.2.2 Regulatory Guide 134: Control Electroslag Weld Properties 5.2 13 5.233.23 Regulatory Guide 1.71: Welder ,

Qualification for Areas of -l 1imited Accessibility 5.2 13 5.2333 Regulatory Guide 1.66:

Nondestructive Examination of Tubular Products 5.2 13 5.233.4 Moisture Control for Low Hydrogen, Covered Arc Welding Electrodes 5.2 14 5.23.4 Fabrication and Processing of Austenitic Stainless Steels 5.2 14 !

5.23.4.1 Avoidance of Stress / Corrosion Cracking 5.2 14 5.23.4.1.1 Avoidance of Significant Sensitiration 5.2 14 1 C\'

U 5.2 iv Amendment 15 -

ABMW 22^^ma RW C Standard Plant SECTION 5.2 CONTENTS (Continued) $

Sectlon Title East 1

5.23.4.1.2 Process Controls to Minimize Exposure to Contaminants 5.2 14 5.23.4.13 Cold Worked Austenitic Stainless Steels 5.2 15 5.23.4.2 Control of Welding 5.2 15 5.23.4.2.1 Avoidance ofliot Cracking- 5.2 15 5.23.4.2.2 Regulatory Guide 134:

Electroslag Welds 5.215 l

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5.23.4.23 Regulatory Guide 1.71: Welder Qualification for Areas of Limited Accessibility 5.215 1

5.23.43 Regulatory Guide 1.66: Non destructive i

Examination of Tubular Products 5.2 15.1 5.2.4 Preservice and Inservice Insnecting ;

and Testina.nLEtattfir C.nolant !

Pressure Roundarv 5.2 15.1 l 1

5.2.4.1 Class 1 System Boundary 5.2 15.1 5.2.4.1.1 Definition 5.2 15.1 5.2.4.1.2 Exclusions 5.2 15.1 5.2.4.2 Accessibility 5.2 16 5.2.4.2.1 Reactor Pressure Vessel Access 3.216 5.2.4.2.2 Piping, Pumps, Valves and Supports 5.2 16 5.2.43 Examination Categories and Methods 5.2 17 5.2.43.1 Examination Categories 5.2 17 5.2.43.2 Examination Methods 5.2 17 5.2.43.2.1 Ultrasonic Examination of the Reactor Vessel 5.217 5.2.43.2.2 Visual Examination 5.2 17 Amendmen 13

ABWR msimo standard Plant RTV. C y SECTION 5.2

(}. CONTENTS (Continued) una m na 5.2.5.4.1 Totallealage Ratc 5.2 26 5.2.5.4.2 Identified Leakage inside Drywell 5.226 5.2.5.5 Unidentified Leakage Inside Drywell 5.226 5.2.5.5.1 Unidentified Leakage Rate 5.2 26 I 5.2.5.5.2 Matgins of Safety 5.2 27 5.2.5.5.3 Criteria to Evaluate the Adequacy and Margin of the Leak Detection System 5.2 27 j 5.2.56 Differentiation between Identified and Unidentified Leaks 5.2 27 l

5.2.5.7 Sensitivity and Operability Tests 5.2 27 i O- i

' \M 5.2.5.8 Testing and Calibration 5.2 27 5.2.5.9 Regulatory Guide 1.45: Compliance 5.227 5.2.6 Interfaces 5.2 28 5.2.6.1 Water Chemistry 5.228 5.2.6.2 Conversion ofIndicators 5.2 28 l

l 5.2.7 Refmam 5.228 5.2.nl

-(

Amendment 15

ABM 234aooxa Standard Plant REV C ;

SECTION 5.2 TABLES S' 1 Eisut HLit East 5.21 Applicable Code Cases 5.229 5.22 Systems Which May Initiate During Overpressure Event 5.2-30 >

5.23 Nuclear System Safety / Relief Valve Setpoints 5.2 31 5.2-4 Reactor Coolant Pressure Boundary Materials 5.2 32 5.25 BWR Water Chemistry 5.2 34 5.26 LDS Control & Isolation Functions vs Monitored Variables 5.2 35 5.27 Leakage Sources vs Monitored Trip Alarms 5.2 36 528 Examinations Category 5.2 36.1 5.29 Ultrasonic Examination of RPV:

Reg. Guide 1.150 Complit ace 5.2 36.22 ,

ILLUSTRATIONS Figure Bilt East r-5.21 Safety Action Valw Lift Characteristics 5.2 37 5.22 MSIV Closure with Flux Scram and Installed i Safety / Relief Valve Capacity 5.238 5.23 Safety / Relief Valve Schematic Elevation 5.2 39 5.24. Safety / Relief Valve and Steamline Schematic ' 5.2 40 5.25 (Deleted) 5.2 41 5.2-6 (Deleted) 5.2 42 5.2 7a RPV Examination Areas 5.2 43 5.2 7b Typical Piping System Isometric 5.2 43.1 i

5.2 viii l Amendment t$

ABWR ursioorn arv c Etandard Plant O i'tusrairiosscce ti=#ea) l Figure Illic East 5.2 7c Scan Coverage Limitation 5.2 43.2 5.28 leak Detection Sptem . IED 5.2-44 I

i l

I O

I 5.2.ix O-Amendment i$

ABWR mma l Standard Plant um e '

$.2 INTEGRlW OF REACI'OR COOLANT formance with 10CFR50, Appendix A, General Dr.

(~ PRESSURE BOUNDARY sign Criterion 15. Preoperational and startup

( instructions are given in Chapter 14.

This section disrdsses measures employed to provide and maintesn the integrity of the reactor $.2.2.1.1 Safety Design Bases i coolant pressure Soundary (RCPB) for the plant 1 design lifetime. The nuclear pressure relief system has been ;

designed to-5.2.1 Compliacce witF. Codes and Code Cases (1) prevent overpressurization of the nuclear system that could lead to the failure of the 5.2.1.1 Compilance Wth 10CFR50, Sc-tion 30.$$a RCPB; Table 3.2 4 shows the Code applied to (2) provide automatic depressurization for small a components. Code edition, applicable addenda, breaks tu the nuclear system occurring with

$ and component dates will be in accordance with maloperation of both the reactor core 10CFR50.55a. Isolation cooling (RCIC) system and the high pressure core flooder (HPCF) system so that 5.2.1.2 Applicable Code Cases the low pressure flooder (LPFL) ruode of the residua: heat removal (RHR) system can ,

The reactor pressure vessel and appurtenances operate to protect the fuel barrict; ,

and the RCPB piping, pumps, and valves will be designed, fabricated, and tested in accordance (3) permit verification ofits operability; and with the applicable edition of the ASME Code,in. .

cluding addenda that were mandatory at the order (4) withstand adverse combinations oiloadings date for the' epplicable cornponents. Section and forces resulting from normal, upset,

- 50.55a of 10CFR50 requires code case approval for emergency, or faulted conditions.

('

\

N Class 1,2, and 3 components. These code cases l contain requirements or special rules which may $.2.2.lJ Power Generation Design Bases

be used for the construction of pressure retain.

ing components of Quality Group Classification A, The nuclear pressure relief system safety / _

B, and C. The various ASME code cases that may relief valves (SRVs) have b en designed to meet g be applied to components are listed in Table the following power generation bases:

5.21.

t (1) discharge to the containment suppression l Regulatory Guides 1.84 and 1.85 provide a list pool, and l of ASME Design and Fabrication code cases that have been generically approved by the Regulatory (2) correctly reclose following operation so Staff. Code Cases on this list may, for design that maximum operational continuity is

. purposes, be used until apptr priately annulled, obtained.

Annulled cases are considered active for equipment that has been contractually committed 5.2.2.1.3 Discusalon

. to fabrication prior to the annulment.

$ The ASME Boller and Pressure Vessel Code 5.2.2 overpressure Protection requires that each vessel designed to meet Section 111 be protected from overpressure under This subsection evaluates systems that protect upset conditions, the RCPB from overpressurization. -

5.2.2.1 Design Basis Overpressure protection is provided in con.

A Amendment 15 5.21

__ _ ____ ___ - _. _ _ __ _ _ _ _ _ _ ____- ___ .____ - - -- ~ _ - . _ _ _ - _ - - . - - .

MN Standard Plant uAsimAs arv c The SRV setpoints are listed in Table 5.2 3 dump valve of the turbine control valve and satisfy the ASME code specifications for hydraulle actuation system. The rosition i safety valves because all valves open at less switches are actuated when the respeed.c valves than the nuclear system design pressure of 87.9 are closing and following 15% travel of full kg/cm g. stroke. The pressure switches are actuated when l a fast closure of the turbine control valves is The automatic depressurization capability of initiated. Credit is not taken for the the nuclear system pressure relief system is powet. operated mode. Credit is only taken for evaluated in Section 6.3 and Section 7.3, the safety / relief valve capacity which opens by the spring mode of operation direct from inlet The following eriteria are used in selection pressure.

of SRVs:

The rated capacity of the pressure. relieving (1) must meet requirements of ASME Code, Section devices shall be sufficieat to prevent a rise in j

!!!; pressure within the protected vessel of more than 110p of the design pregiure (1.10 x 87.9 (2) must qualify for 100% of nameplate capacity kg/cm g = 96.7 kg/cm g) for eyents credit for the overpressure protection defined in Section 15.2. I function; and Full account is taken of the pressure drop on (3) must meet other performance requirements both the inlet and discharge sides of the such as response time, etc., as necessary to valves. All combination safety / relief valves provide relief functions, discharge into the suppression pool through a discharge pipe from each valve which is designed The SRV discharge piping is designed, to achieve sonic flow conditions through the installed, and tested in accordance with ASME valve, thus providing flow independence to Code, Section 111. dischstge piping losses, r

5.2.2.1.4 Safety / Relief Valve Capacity SRV capacity of this plant is adequate to 5.2.2.2 Design Evaluation limit the primary system pressure, including transients, to the requirements of ASME Boller 5.2.2.2.1 Method of Analple and Pressure Vessel Code Section III, Nuclear Power Plant Componchts, up to and including The method of analysis is approved by the NRC applicable addenda. The essential ASME require. or developed using criteria approved by the NRC.

ments which are met by this analysis follow. ~

5J 2.2.2 Sptem Dealen

L lt is recognized that the protection of vessels in a nuclear power plant is dependent A parametric study was conducted to upon many protective systems to relieve or determine the required steam flow capacity of l' terminate pressure transients. Installation of the , 'Vs based on the following assumptions, pressure relieving devices may not independently provide complete protection. The safety valve 5.2.2.2.23 Jperating Conditions

( sizing evaluation gives credit for operation of the scram protective system which may be tripped (1) operating power = 4005 MWt (102% of nuclear by either one of two sources: a direct or a flux boiler rated power);

l i trip signal. The direct scram trip signal is derived from position switches mounted on the (2) vessel dome pressure 173.1 kg/cm# g ; and l main steamline isolation valves, the turbine stop valves, or from pressure switches mounted on the 5.22 Amendment is

1

(

ABWR m-prv c l

Standard Plant j i

nominal setpoints. This is to account f ar ini. l (3) steamflow = 7844 T/h (102.7% of nuclear tial setpoint errors and any instrument setpoint )

ph'~_ l drift that might occur during operation. Typic. i boiler rated steamflow).

ally the assumed setpoints in the analysis are These conditions are the most severe because at least 1% above the actual nominal setpoints, l maximum stored energy exists at these Conservative SRV response characteristics are conditions. At lower power conditions the also assumed; therefore, the analysis conserv-transients would be less severe, atively bounds all SRV operating conditions.

5.2.2.2.2.2 Transients 5.2.2.2.2.4 Safety / Relief Yalve Capacity The overpressure protection system is capable Sizing of the SRV capacity is based on of accomodating the most severe pressurization establishing an adequate margin from the peak i transient. The evaluation of transient behavior vessel gressure to the vessel code limit (96.7 based on the core loadiag shown in Figure 4.31 kg/cm g) in response to the reference demonstrates that MSIV closure with failure of transients.

direct scram (i.e., scram occurs on high flux) is the most severe pressurization transient. Other The method used to determine total valve fuel designs and core loading patte.ns, including capacity is as follows.

loading patterns similar to Figure 4.3 2, will not affect the conclusions of this evaluation. Whenever the system pressure increases to the Analysis of this event will be performed each valve spring set pretsure of a group of valves, cycle and the results provided for information to these valves are assumc3 to begin opening and to the USNP.C. Table 5.2 2 lists the systems which reach full open at 103% of the valve spring set could initiate during the MSiV closure flux scram pressure. The lift characteristics assumed is event. shown in Figure 5.21.

O V

5.2.2.2.2.3 Safety / Relief Valve Transient Analysis Specification 5.2.2.2.3 Evaluation of Results 5.2.2.2.3.1 Safety /Rellef Oe Capacity (1) Simulated valve groups:

The required SRV capacity is determined by spring action s afety mode . 5 groups analyzing the pressure rise from an MSIV closure with flux scram transient. Results of this l (2) opening pressure setpoint (maximum safety analysis is given in Fi;ure t 5.2 2. The peak Umit): vessel, bottom pressure calculated is 89.6 kg/cm g, which is 3 ell below the acceptance spring. action safety mode: limit c. 96.7 kg/cm g.

group 1 82.8 kg/cm',g 5.2.2.2.3.2 Pressure Drop in Inlet and group 2 83.5 kg/cmg Discharge group 3 84.2 kg/cmg group 4 84,9 kg/ cmp Pressure drop in the piping from the reactor- l group 5 85.6 kg/cm g vessel to the valves is taken into account in calculating the maxituum vessel pressures.

(3) reclosure pressure setpoint (% of opening Pressure drop in the discharge piping to the setpoirit) both modes: suppression poolis limited by proper discharge line sizing to prevent back. pressure on each SRV ma.ximum safety limit (ud in analysis) % from exceeding 40 percent of the val.e inlet

" pressure, thus assuring choked flow in the valve

$ minimum operationallimit 90 orifice and no reduction of valve capacity dre to the discharge piping. Each SRV has its own p The opening and reclosure setpoints are separate discharge line, assumed at a conservatively high level above the Amendment 15 $.2 3 w

h MM 23A6100AB Standard Plant nev. c 5.2.2.3 Piping and lastamment Diagrams Figures 5,13 and 5.2 3 show the schematic location of the follewloj pressure relieving devlees for:

_ l (1) the reactor coolant system; 9

i Amendment 15 5.23.1 O

l l

,-j

ABWR. ursioorn REV C Standard Plant (2) the primary side of the auxiliary or reclose as generated heat decays. The pressure emergency systems interconnected with the increase and relief cycle will continue with O. primary system; and lower frequency and shorter relief discharges as the decay beat drops off.

(3) any blowdown or heat dissipation system connected to the discharge side of the Remote manual actuation of the valves from pressure relieving devices, the control room is recommended to minimize the total number of these discharges with the intent The schernatic arrangements of the SRVs are of achieving extended valve seat life, shown in Figures 5.2 3 and 5.2 a.

The SRV is opened by either of the following 5.2.2.4 Equipment and Component Description two modes of operation:

5.2.2.4.1 Description (1) The safety (steam pressure) mode of l operation is initiated when the direct and The nuclear pressure relief systern consists increasing static inlet steam pressure of SRVs located on the main steamlines between overcomes the restraining spring and the the reactor vessel and the first isolation vahe frictional forces acting against the inlet within the drywell. These valves protect against steam pressure at the main disc or pilot overpressure of the nuclear system, disc and the main disc moves in the opening direction at a f aster rate than The SRVs provide three main protection corresponding disc movements at higher or functions: lower inlet steam pressures. The condition at which this action is initiated is termed (1) overpressure relief operation (the valves the ' popping pressure

and corresponds to are opened using a pnu:matic actuator upon the set pressure value stamped on the receipt of an automatic or manually- nameplate af the SRV.

initiated signal to reduce pressure or to limit o' pressure rise); (2) The relief (power) mode of operation is initiated when an electrical signal is (2) overpressure safety . neration (the valves received at any of the solenoid valves function as safety valves and open to located on the pneumatic actuator assembly, prevent nuclear system overpressurization - The solenoid valve (s) will open, allowing they are self actuated by inlet steam pressurized air to enter the lower side of pressure if not already signaled open for the pneumatic cylinder piston which pushes relief operation); the pistcn and the rod upwards. This action pulls the lifting mechanism of the mdn or (3) depressurization operation (the ADS valves pilot disc thereby opening the valve to open automatically as part of the emergency allow inlet steam to discharge through the core cooling system (ECCS) for events SRV until the inlet pressure is near or l involving stuall breaks in the nuclear system equal to zero.

process barrier. The location and number of the ADS valves can be determined from Figure The pneumatic operator is so arranged that if 5.25). it malfunctions it will not prevent the valve from opening when steam inlet pressure reaches Chapter 15 discusses the events which are the spring lift set pressure, expected to activate the primary system SRVs.

The section also summarizes the number of valves For overpressure SRV operation (self. actuated expected to operate in safety (steam pressure) or spring lif t mode), the spring load

, mode of operation during the initial blowdown of establishes the safety valve opening setpoint the valves and the expected duration of this pressure and is set to open at setpoint first blowdown. For several of the events it is expected that the lowest set SRV will reopen and Amendment 15 5.24

ABM 23xsiooxii

' Standard Plant an c The ADS utilizes selected SRVs for (q

V) designated in Table 5.2 3. In accordance with the ASME code, the full lift of this mode of depressurization of the reactor as described in operation is attained at a pressure no greater Section 6.3 Each of the SRVs utilized for than 3% above the setpoint, automatic 4 pressurization is equipped with an air accumuth.or and check valve arrangement.

The spring loaded valves are designed and These accumulators assure that the valves can be constructed in accordance with ASME 111, NB 7640, held open following failure of the air supply to as safety valves with auxiliary actuating the accumulators. The accumulator capacity is devices, sufficient for one actuation at drywell design pressure, or 5 actuations at normal drywell For overpressure relief valve operation pressure.

(power actuated mode), valves are provided with pressure sensing devices which operate at the Each SRV discharges steam through a discharge setpoints designated in Table 5.2 3 When the line to a point below minimum water levelin the set pressure is reached, a solenoid air valve is suppression pool. The SRV discharge lines in operated which in turn actuates the pneumatic the drywell are classified as quality group C piston / cylinder and linkage assembly to open the and Seismic Category I. The SRV discharge lines valve. in the wetwell air space are classified as Quality Group B and Seismic Category 1. SRV The maximum opening delay from when the discharge piping from the SRV to the suppression pool consists of two parts. The first is l pressure start of disk just exceeds motion the relief is 0.5 seconds setpoint of which to at one end to the SRV and at its other the attached time to energize the SRV solenoid shall not end to the diaphragm floor penetration which l exceed 0.4 seconds, When the piston is actuated, acts as a pipe anchor. The second part of the the delay time (maximum elapsed time between SRV discharge piping extends from the diaphragm receiving the overpressure signal at the valve floor penetration to the SRV quencher in the actuator and the actual start of valve motion) suppression pool. Because the diaphragm floor p:

Q will not exceed 0.1 second. The maximum clapsed acts as an-anchor on this part of the line,it time between signal to actuator and full open is physically decoupled from the main steam header. l

[ position With the SRV of inlet valve will not pressure exceed

>70.3 0.25 seconds.

kg/cm 2g and initial SRV pressure <4% of inlet pressure. As a part of the preoperational and startup testing of the main steamlines, movement of the The'SRVs can be operated individually in the SRV discharge lines will be monitored, power actuated mode by remote manual controls from the main control room. The SRV discharge piping is designed to limit valve outlet pressure to.approximately 40 There is one solenoid provided on each SRV for percent of maximum valve inlet pressure with the non ADS power actuated operation. The logic for valve wide open. Water in the line more than a the SRV power actuated relief function requires few feet above suppression pool water level two trip signals to open the SRVs. The failure would cause excessive pressure at the valve of one pressure transmitter will not cause the discharge when the valve is again opened. For SRVs to open. Each SRV is provided with its own this reason, two vacuum relief valves are pneumatic accumulator and inlet check valve. The provided ~on each SRV discharge line to prevent accumulator capacity is sufficient to provide one drawing an excessive amount of water into the SRV actuation. Subsequent actuations for an line as a result of steam condensation following overpressure event can be spring actuations to termination of relief operation. The SRVs are limit reactor pressure to acceptable levels. located on the main steamline piping rather than on the reactor vessel top head, primarily to simplify the discharge piping to the pool and to avoid the necessity of having to remove sections Amendment 15 5.2-5 l

i

]

~ $$ 2WtooAs -

REV.C Standard Plant u of this piping when the reactor head is removed i

l. for refueling. In addition,. valves located on L 'the steamlines are more accessible during a .

I shutdown for valve maintenance.

The ADS automatically depressurires the nuclear system sufficiently to permit LPFL mode ,

1:

of the RHR system to operate as a backup for the i Lx HPCF. Further descriptions of the operation of the automatic depressurization feature are presented in Section 6.3 and Subsection 7.3.1.

5.2J.4.2 Design Parameters !

l-l The specified op-rating transients for components within the RCPB are presented in Subsection 3.9.1 Subsection 3.7.1 provides a

l. discussion of the input criteria for design of g

Seismic Category I structures, systems, and components, The design requirements established to protect the principal components of the ;

reactor coolant system against environmental

. effects are presented in Section 3.1L The design criteria and analysis methods for

. 5.2.2.4.3 Safety / Relief Valve considering Imds due to the SRV discharge is 1,

contained in Subsection 3.9.3.3.

5.2.2.6 Applicable Codes and Classiflestion .

The vessel overpressure protection system is i The design pressure and temperature of the desigaed to satisfy the requirements of Section !

valve inlet is %.7 kg/cm 2g at 3080C. 111 of the ASME Boiler and Pressure Vessel Code. The' general requirements for protection The valves have been designed to achieve the against overpressure of Section III of the Code ;

maximum practical number of actuations consistent recognize that reactor vessel overpressure pro-

'with state-of.the art technology, tection is one function of the reactor protec- <

tive systems and allows the integrati_on of .j pressure relief devices with the protective j systems of the nuclear reactor. Hence, credit is taken.for the scram protective system as a a complementary pressure protection device.L The L NRC has also adopted the ASME Codes as part of j their. requirements'in the Code of Federal l 5.2.23 Mounting of Safety Relief Valves Regulations (10CFR50.55a).

The safety relief valves are located on the 5.2.2.7 MaterialSpecifications main steam piping.

Material ~ specifications-for pressure-retaining components of SRVs are in Table 5.2 4.

.5.24 O

Amendment 15

ABWR man Standard Plant RM C q 5.2.2.8 Process lastrumentation certification from the valve manufacturer that design and performance requirements have been

-("7 Overpressure proteetion process met. This includes capacity and blowdown instrumentation is listed in Table 4 of Figure requirements. The set points are adjusted, 5.13. verified, and indicated on the valves by the vendor. Specified manual and automatic i 5.2.2.9 System Reliability initiated signal for power actuation (relief mode) of each SRV is verified during the The system is designed to satisfy the re- preoperational test program, quirements of Section Ill of the ASME Boiler and Pressure Vessel Code. The consequences of fail. It is not feasible to test the SRV setpoints l ure are discussed in Subsections 15.1.4 and while the valves are in place. The valves are 15.6.1. mounted on 1500 pound primary service rating flanges. They can be removed for maintenance or 5.2.2.10 Inspection and Testing bench checks and reinstalled during normal plant shutdowns. The valves will be tested to check The inspection and testing applicable SRVs set pressure in accordance with the requirements utilizes a quality assurance program which of the plant technical specifications. The complies with Appendix B of 10CFR50. external surface and seating of all SRVs are 100% visually inspected when the valves are The non radioactive SRVs are tested at a removed for maintenance or bench checks. Valve suitable test facility in accordance with quality operability is verified during the preopera-control procedures to detect defects and to prove tional test program as discussed in Chapter 14.

operability prior to installation. The following tests are conducted: 5.2.3 Reactor Coolant Pressure Boundary-Materials (1) hydrostatic test at speciiled test con-

- p]

1 ditions (ASME code requirement based on design pressure and temperature);

5.23.1 Material Specifications l

j Table 5.2 4 lists th( rincipal pressurc (2) thermally stabilize the SRV to perform retaining materials and the appropriate material l

quantative steam lackage testing at 10.5 specifications for the RCPB components.

kg/cm2g below the SRV nameplate valve with an acceptance criteria not to exceed 5.2.3.2 Compatibility with Reactor Coolant 0.45 kg/h leakage; 5.23.2.1 PWR Chemistry of Reactor Coolant (3) full flow SRV test for set pressures and blowdown where the valve is pressurized with Not applicable to BWRs.

saturated steam, with the pressure rising to the valve set pressure. (The SRV must be 5.23.2.2 BWR Chemistry of Reactor Coolant adjusted to open at the nameplate setpressure 11%, unless a greater tolerance A brief review of the relationships between I is established as permissible in the water chemistry variables and RCS materials overpressure protection report in t' e valve performance, fuel performance, and plant design specification); and radiation fields is presented in this section t and further information may be obtained from l g -(4) response time test where each SRV is tested Reference 10a.

l to demonstrate acceptable response time j- based on system requirements. The major environment related m.iterials l performance problem encountered to date in the l

The valves are installed as received from the RCS of BWRs has been intergranular stress factory. The GE equipment specification requires corrosion cracking (IGSCC) of sensitized G austenitic stainless steel. IGSCC in sensitized U

smendment 15 5.27 u

ABWR man rTy; Sandard Plant material adjacent to welds in Type 304 and Type manipulate the corrosion potential in laboratory 316 stainless steel piping systems has occurred tests) (Reference 10).

in the past. Substantial research and develop-ment programs have been undertaken to understand As the corrosion potentialis reduced below Q M

the IGSCC phenomenon and develop remedial men. the range typical of normal BWR power operation sures. For the ABWR,IGSCC resistance has been (+ 50 to 50 mVSHF)' * EI" I I""""IIY I achieved through the use of IGSCC resistant mate- IGSCC appears at ~230 mVSHE. It is apparem D rials such as Type 316 Nuclear Grade stainless that a combination of corrosion potential (which A steel and stabilized nickel base Alloy 600M and can be achieved in a BWR by injecting usually <

1 ppm hydrogen into the feedwater) plus tight E 182M. N conductivity control (0.2 pS/cm) should permit Much of the early remedy development work fo- BWRs to operate in a regime where sensitized cused on alternative materials or local stress stainless steels are immune to IGSCC. In the reduction, but recently the effects of water che- reactor vessel, the excess hydrogen reacts with E mistry parameters on the IGSCC process have rece- the radioloytic oxygen and reduces the electro- (

ived increasing attention. Many important fea- chemical corrosion potential (Reference 10a and E tures of the relationship between BWR water chem- 10b). The reactor water cleanup sys'em, which #

istry and IGSCC of sensitized stainless steels processes reactor water at a rate of 2% of rated feedwater flow, removes both dissolved and

(

bave been identified. E undissolved impurities that enter the reactor y A

W Laboratory studies (References 3 and 4) have water. The removal of dissolved impurities A shown that although IGSCC can occur in simulated reduces the conductivity into the region of BWR startup em'ironments, most IGSCC damage pro- immunity to IGSCC.

bably occurs during power operation. The normal BWR emironment during power operation is -280 Since the ABWR has no sensitized stainless C water containing dissolved oxygen, hydro- steel, IGSCC control by bydrogen injection is gen and small concentrations of ionic and non- not required. However, irradiation assisted y lonic impuritiesfconductivity generally below stress corrosion cracking (IASCC) can occur in A 0.3 pS/cm at 25 C). It has been well docu- highly irradiated annealed stainless steel and mented that some ionic impurities (notably sul- nickel base alloys. Preliminary in reactor and fate and chloride) aggravate IGSCC, and a number laboratory studies (Reference 11) have indicated of studies have been made of the effects of indi- that HWC will be useful in mitigating IASCC.

vidualimpurity species on IGSCC initiation and growth rates (References 3 thru 7). This work In reactor and laboratory evidence also indi-clearly shows that IGSCC can occur in water at cates that carbon and low alloy steels also tend 280 C with 200 ppb of dissolved oxygen, even at to show improved resistance to envhonmentally low conductivity (Iow impurity levels), but the assisted cracking with both increasing water pu-rate of cracking decreases with decreasing impu- rity and decreasing corrosion potential r.

I rity content. Although B\"R water chemistry (Reference 12). N guidelines for reactor water cannot prevent IGSCC, maintaining the lowest practically achiev- 5.2.3.2.2.1 Fuel Performance Considerations y able impurity levels will minimize its rate of progression (References 5 and 9). Nuclear fuel is contained in Zircaloy tubes n

that constitute the first boundary or primary

!. Stress corrosion cracking of ductile materials containment for the highly radioactive species L

in aqueous environments often is restricted to generated by the fission process; therefore, the c

specific ranges of corrosion potential', so a integrity of the tubes must be ensured. Zirca-l number of studies of impurity effects on IGSCC loy interacts with the coolant water and some have been made as a function of either corrosion coolant impurities. This results in oxidation potential or dissolved oxygen content (the by the water, increased bydrogen content in the dissolved oxygen content is the major chemical Zircaloy (hydriding), and, often, buildup of a i

variable in BWR type water that can be used to layer of crud on the outside of the tube. Ex-

Also called electrochemical corrosion potential or ECP, see Reference 9.

518 Amendment 3 l

l

ABM nacimin RIV C Standard Plant (1) Conductivity oxygen control cannot be achiev_ed through traditional chemistry and operational prac.

- O- Increasing levels of many lonic impurities tices. Oxygen control to low, plant spec-adversely influence both the stress corri- ific levels can be obtained through hydro-sion cracking behavior of RCS materials, the gen injection. Control of reactor water rate of radiation field buildup and also can oxygen during startup/ hot standby may be affect fuel performance. Therefore, conduc- accomplished by utilizing the de aeration tivity levels in the reactor water should be capabilities of the condenser, Independent maintained at the lowest level practically control of control rod drive (CRD) cooling achievable, water oxygen concentration of <50 ppb during i

power operation is desirable to protect (2) Chloride against IGSCC of CRD materials. Carbon stects exhibit minimal general corrosion and Chlorides are among the most potent promo- release rates in water with a conductivity ters of IGSCC of sensitized stainless stects less than 0.1mS/cm if the concentration of and are also capable of inducing transgran- oxygen is in the range of 20 to 1000 ppb.

ular cracking of nonsenitized stainless Regulation of reactor feedwater dissolved 7 steels. Chlorides also promote pitting and oxygen to 20 to 50 ppb during power op-

crevice attack of most RCS materials. Chlo- eration will minimize corrosion of the rides normally are associated with cooling condensate and feedwater system and reduce water inleakage, but inputs via 'adwaste the possibility of locally increasing processing systems have also ocsorred, reactor wates oxygen concentrations, it is important to note that for oxygen concentra-Because chloride is implicated in several tions below 20 ppb, the data indicates an different corrosion phenomena, its level in increase in the corrosion and corrosion prod-reactor water should be kept ;s low as uct release for carbon stects.

practically achievable during power operation.

'(3) Sdak High iron inputs into the reactor have been associated with excessive fuel deposit Recently, sulfate has been found to be more buildup. Proper regulation of feedwater aggressive in promoting IGSCC of sensitized purity and dissolved oxygen levels will Type 304 stainless steel in EWR type water minimize iron transport to the reactor.

(in laboratory tests) than any other lon, This, in tten, should minimize fuel deposits including chloride. Sulfates have also been and may asat in controlling radiation-

- implicated in environment assisted cracking buildup.

of high nickel alloys and carbon and low alloy steels. Sulfate ingress can result (6) Fluoride from cooling water inleakage, regenerant chemical inleakage, or resin ingress. Fluoride promotes many of the same corrosion

- phenomena as chloride, including IGSCC of '

(4) QDgm sensitized austenitic stainless steels, and

- may also have the potential to cause corro-E Dissolved oxygen has been identified as a sion of Zircaloy core components. If floride major contributor to IGSCC of sensitized is present, it will be measured for

=

L stainless steels and reduction of oxygen diagnostic purposes.

~, content is known to reduce the tendency for

[ pitting and cracks of most plant materials. (7)Dranles

" During power operation, most of the oxygen Organic compounds can be introduced into the 3

content of reactor water is due to the radi-olysis of water in the core and, therefore, RCS via turbine or pump oil leakage, rad-waste, or makeup water systems. Of par-l_ ticular concern is the possibility that

- halogenated organic compounds (e.g.,

Amendment 15 5.2 10 r

1 23A6G.AB mandard Plant uv c g Water quality specifications for the ABWR re- (2) nickel based alloy (including 600 and X 750) e quire that the condenser is to be designed and and alloy steel; erected to minimize tube leakage and to facill-g tate aslotenance. Appropriate features ere in. (3) carbon steel and low alloy steel; ;

2, corporated to detect leakage and segregate the !

EA source. Tbc valves controlling the cooling water (4) some 400. series martensitic stainless steel !

l to the condenser sections are required to be (all tempered at a minimum of 5930C); l j operable from the control room so that a leaking section can be scaled off quickly. (5) Colmonoy and Stellite hardfacing material '

(or equivalent); and

$.2.3.2.2.4 1ASCC Conalderations !

(6) precipitation hardening stainless steels, Plant experience and laboratory tests indicate 17 4PH and XM.13 in the H1100 condition. ,

that irradiation assisted stress corrision l cracking (IASCC) can be initiated in solution All of these construction materials are annealed stainless steel above certain stress resistant to stress corrosion in the BWR cool-levels after exposure to radiation, ant. General corrosion on all materials except ;

carbon and low alloy steel, is negligible.

Extensive tests have also shown that IASCC has Conservative corrosion allowances are provided i

l not occurred at fluence levels below ~5x1020 for all exposed surfaces of carbon and low alloy

-n n/cm2 (E>1MeV) even at high stress levels. Ex. steels.

,. E periments lodicate that as fluence increases I

above this threshold of 5x1020 n/cm2 , there The requirements of GDC 4 relative to compat-is a decreasing threshold of sustained stress be. ability of components with environmental condi.

l low which IASCC has not occurred. (Examination tions are met by compliance with the applicable of top guides in two operating plants which have provisions of the ASME Code by compliance with l

creviced designs has not revealed any 1ASCC.) the recommendations of Regulatory Guide 1.44.

j. l.

1!

lL -

Contaminants in the reactor coolant are '

Reactor core structural components are design. controlled to very low limits. These controls l ed to be below these thresholds of exposure and/ are implemented by limiting contaminant levels or stress to avoid 1ASCC In addition, crevices- of elements (such as halogens, S, Pb) to as low g have been eliminated from the top guide design in as possible in miscelleancous materials used n order to preveat the synergistic interaction with .during fabrication and installation. These -

IASCC. materials (such as tapes, penentrants) are l 'l completely removed and cleanliness is assured.

In areas where the 5x1020 n/cm2 threshold No detrimental effects will occur on any of the j '

.of irradiation is not practically avoided, the materials from allowable contaminant levels in stress level is maintained below the stress the high purity reactor coolant. Expected L

threshold. High purity grades'of materials are radiolytic products in the BWR coolant have no .

l used in control rods to extend their life. Also adverse effects on the construction materials.

HWC introduced in the plant design to control .

i ICSCC may also be beneficial in avoiding 1ASCC. 5.2.3.2.4 Compatibility of Construction L

Materials with External Insulation l 5.23.2.3 Compatibility of Construction Materials with Reactor Coolant ' All non. metallic insulation applied to auste-nitic stainless steel meets Regulatory Guide The cinstruction materials exposed to the 1.36.

reactor coolant consist of the following:

5.2.33 Fabrication and Processing of Ferritic i

(1) solutio 0 annealed austenitic stainless Materials I steels (both wrought and cast), Types 304, 304L,316L.1,316L and XM.19; 5.2.3.3.1 Fracture Toughness 5.2 12 Amndmem t$

l 1

ABM Standard Plant

. 2 mioors RPV C i

. Compliance with Code requirements shall be in perature control requirements and welding proce.

accordance with the following: dure qualifications supplementing those in ASME !

Sections III and IX.

(1) The ferritic materials used for piping, pumps, and valves of the reactor coolant The use of low alloy steel is restricted to

l pressure boundary are usually 63.5 mm or the reactor pressure vessel, Other ferritic I- less in thickness. Impact testing is components in the reactor coolant pressure performed in accordance with NB 2332 for boundary are fabricated from carbon steel thicknesses of 63.5 mm or less. The materials, materials comply with Appendix G, Section G 3100 of ASME Code Section Ill. Preheat temperature employed for welding of low alloy steel meet or exceed the recommenda.

-(2) Materials for bolting with nominal diameters tions of ASME Code Section III, Subsection NA. )

exceeding 25.4 mm are required to meet both Components are either held for an extended time the 0.64 mm lateral expansion specified in at preheat temperature to assure removal of hy.

NB 2333 and the 6.2 kg m Charpy V value spe- drogen, or preheat is maintained until post weld cified in 10CFR50, Appendix G. The 6.2 kg m heat treatment. The minimum preheat and maximum requirement stems from the ASME Code where interpass temperatures are specified and

, l it applies to bolts over 100 mm in diame- monitored.

g ter, starting Summer 1973 Addenda. Prior to this, the Code referred to only 2 sizes of

- bolts (.s. 25.4 mm and > 25.4 mm). GE continued the two size categories, and added the 6.2 kg m as a more conservative requirement.

. j /]- (3) The reactor vessel complies with the requi- All welds were condestructively examined by V rements of NB 2331. The reference tempera- radiographic methods. In addition, a supple-ture (RTNDT) is established for all mental ultrasonic examination was performed, required pressure retaining materials used -

in the construction of Class 1 vessels. - 5.233.2.2 Regulatory Guide l.34: Control of This includes plates, forgings, weld Electroslag Weld Properties material, and heat affected zone. The:

RTNDT differs from the all ductility For electroslag welding applied to structural temperature (NDT) in that in addition to joints, the welding process variable specified-passing the drop test, three Charpy V Notch- in the procedure qualification shall be specimens (traverse) must exhibit 6.9 kg m monitored during the welding process, absorbed energy and 0.89 mtn lateral _

expansion at 330C above the RTNDT. The 5.23.3.2.3 Regulatory Guide 1.71: Welder core beltline material must meet 10.4 kg m Qualification for Arvas of United Accessibility absorbed upper shelf energy.

Welder qualification for areas of limited l (4) Calibration of instrutnent and equipment accessibility is discussed in Subsection g shall meet the requirements of the ASME 5.2.3.4.2.3.

Code, Section Ill, paragraph NB 2360.

-L 5.2.3.3.3 Regulatory Guide 1.66: Nondestrue.

5.2.33.2 Control of Welding tive Examination of Tubular Products 5.233.2.1 Regulatory Guide 1.50: Control of Regulatory Guide 1.66 describes a method of Preheat Temperature Employed for Welding of implementing requirements acceptable to NRC re.

I.ow Alloy Steel garding nondestructive examination requirements of tubular products used in R"PB This Regula-Regulatory Guide 1.50 delineates preheat tem. tory Guide was withdrawn on Se,.* ember 28,'1977, by the NRC because the additiord eequirements 5213 Amendment 15

1 I

MM 22A6100AB an c 1

Standard Plant 1

imposed by the guide were satisfied by the ASME with the guide lines of NUREG 0313, to avoid l l Code. significant sensitization.

l Wrought tubular products are supplied in Process controls are esercised during all stages l accordance .oth applicable ASTM /ASME material of component manufacturing and constructio spadrications. Additionally, the specification minimize contaminants. Cleanliness controls are nor the tubular products used for CRD housings applied prior to any elevated temperature treatment. l speelfied ultrasonic examination to paragraph For applications where stainless steel surfages are l NB 2550 of ASME Code Section 111. exposed to water at temperatures above 93 C low l

carbon (<0.03%) grade materials are used. Fu l These RCPB components meet 10CfR50 Appendix B critical applica. tions, nuclear grade materials ,

I requirements and the ASME Code requirements thus (carbon content .5. 0.02%) are used. All materials l assuring adequate control of quality for the are supplied in the solution heat treated condition. E l products. Special sensitization tests are applied to assure N that the material is in the annealed condition.

5.2.3.3.4 Moisturv Control for Im Hydrogen, Covered Arc Weidlag Electrodes During fabrication, gay heafing operation (except 982 C are avoided, unless l

welding) between 427 i Sultable identification, storage, and handling followed by solution heat treatment. During welding, i of electrodes, flux, and other welding material heat input is controlled. The interpass temperature will be maintained. Precautions shall be taken is also controlled. Where practical, shop welds are to minimize absorption of moisture by electrodes solution heat treated. In general, weld filler and flux, material used for austenitic stainless steel base metals is Type 308L/316L/309L with an average of 8%

(of Fn) ferrite content.

5.2.3.4.1.2 Process Controls to Minimlae Exposure to Contaminants Exposure to contaminants capable of causing stress / corrosion cracking of austenitic stainless l steel components all cleaning was and processing avoided materials whichby carefully con contact the stainless steel during manufacture, construction, and installation.

Special care was exercised to insure removal of L surf ace contaminants prior to any heating 5.23.4 Fabricatlem and Processing of operations. Water quality for cleaning, rinsing, Austenitic Stainless Steels flushing, and testing was controlled and monitored.

Sultable protective packaging was provided for 5.23.4.1 Avoidance of Stress / Corrosion components to maintain cleanliness during shipping Cracking and storage.

5.2.3.4.1.1 Avoidance of Significant The degree of surface cleanliness obtained by Sensitlantion these procedures meets the requirements of Regulatory Guides 1.37 and 1.44. '

l- When austenitic stainless the temperature range982 427, steels C, they are gre heated in u l2 g considered to become ' sensitized' or susceptible li to intergranular corrosion. The ABWR design

! complies with Regulatory Guide 1.44 and 5.2 14 Amendment 15 l'

ABM 21uioorn REV.C Standard Plant 5.2.3.4.13 Cold Worked Austenitic Stainless high. alloy steels or other materials such as Steels static and centrifugal castings and bimetallic joints should comply with fabrication require-Cold work controls are applied for components ments of Sections 111 and IX of the ASME Boiler madu of austenitic stainless steel. During and Pressure Vessel Code. It also requires fabrication cold work is controlled by applying additional performance qualifications for limits in hardness, bend radii and surface finish welding in areas of limited access, on ground surfaces.

All ASME Section ill welds are fabricated in i 5.23.4.2 Control of Welding accordance with the requirements of Sections ill and IX of the ASME Boiler and Pressure Vessel 8.23.4.2.1 Avoidance of Hot Cracking Code. There are few restrictive welds involved in the fabrication of BWR components. Welder Regulatory Guide 1.31 describes the acceptable qualification for welds with the most restric-method ofimplementing requirements with regard live access is accomplished by mockup welding, to the control of welding when fabricating and Mock.up is examined by sectioning and radiography 5 joining sustenitic stainless steel components and (or UT). A sysr is.

Tbc Acceptance Criterion II.3.b.(3) of SRP Written welding procedures which are approved Section 5.2.3 is based on Regulatory Guide by GE are required for all primary pressure boun- 1.71. The ABWR design meets the intent of this dary welds. These procedures comply with the regulatory guide by utilizing the alternate requirements of Sections Ill and IX of the ASME approach as follows:

Boiler Pressure Vessel Code and applicable NRC Regulatory Guides. When access to a non volumetrically examined ASME Section 111 production weld (1) is less p All austenitic stainless steel weld filler than 305 mm in any direction and (2) allows l g

materials were required by specification to have welding from one access direction only, such a minimum delta ferrite content of 8 FN (ferrite weld and repairs to welds in wrought and cast number) determined on undiluted weld pads by low alloy steels, austenitic stainless steels magnetic measuring instruments calibrated in and high nickel alloys and in any combination of accordance with AWS specification A4.2 74, these materials shall comply with the fabrica-tion requirements specified in ASME Boiler and Delta ferrite measurements are not made on Pressure Vessel Code Section 111 and with the qualification welds. Both the ASME Boiler and requirements of Section IX invoked by Section Pressure Vessel Code and Regulatory Guide 131 III, supplemented by the following requirements: ,,,

specify that ferrite measurements be performed on 2 A

undiluted weld filter material pads when magnetic (1) The welder performance qualification test instruments are used. There are no requirements assembly required by ASME Section IX shall for ferrite measurement on qualification welds. be welded under simulated access condi.

tions. An acceptable test assembly will 5.23.4.2.2 Regulatory Guide 134: Electrostag provide botb a Section IX weIder Welds performance qualification required by this Regulatory guide.

l See Subsection 5.233.2.2. If the test assembly weld is to be judged by bend tests, a test specimen shall be 5.2.3.4.23 Regulatory Guide l.71: Welder removed from the location least favorable Qualification or Areas of Umited Accessibility for tbe welder. If tbis test specimen cannot be removed from a location Regulatory Guide 1,71 requires that weld prescribed by Section IX, an additional fabrication and repair for wrought low. alloy and bend test specimen will be required. If

'O the test assembly weld is to be judged by Amendrnent 15 5.2 15

1 ABM ux61oors nrv. c l Standard Plant 1

radiography or UT, the length of the weld to of the ASME B&PV Code Section XI. -

be examined shall include the location least l favorable for the welder. 5.2A.1 Class 1 System 8oundary Records of the results obtained in welder 5.2A.I.1 Definition accessibility qualification shall be as ,

certified by the manufacturer or installer, The class 1 system boundary for both shall be maintained and shall be made preservice and inservice inspection programs and )

accessible to authorized personnel, the system pressure test program includes all those items within the Class 1 and Quality Group Socket weld with a 50A nominal pipe size and A boundary on the piping and instrumentation

[ under ate excluded from the above drawings (P&lDs). That boundary includes the 1 requirements, following:

(2) (a) For accessibility, when more restricted (1) Reactor pressure vessel aecess conditions than qualified will (2) Portions of the main steam system obscure the welder's line of sight to (3) Portions of the feedwater system the extent that production welding will (4) Portions of the standby liquid control

require the use of visual aids such as system N mirrors. The quellfication test as- (5) Portions of reactor water cleanup system l sembly shall be welded under the more (6) Portions of the residual heat removal system restricted access conditions using the (7) Portions of the reactor core isolation

,4 visual aid required for production cooling system welding. (8) Portions of the high pressure core flooder L, system I

(b) GE complies with ASME Section IX.

Those portions of the above systems within (3) Surveillance of accessibility qualification the Class 1 boundary are those items which are requirements will be performed along with part of the reactor coolant system up to and normal surveillance of ASME Section IX including any and all of the following:

performance qualification requirements.

- (1) the outermost containment isolation valve in 5.2.3A 3 Ragulatory Gulde l.66: the system pipig which penetrates primary Nondestructive 5'uamination of Tubular Producta reactor containment.

For discussion of compliance with Regulatory (2) the second of two valves normally closed Guide 1.66, see Subsection 5.2.3.3.3. during normal reactor operation in system piping which does not penetrate primary 5.2.4 Preservice and Inservice reactor containment.

Inspection and Testing of Reactor t Coolant Pressure Boundary (3) the reactor coolant system safety and relief valves, This subsection describes the preservice and l- inservice inspection and system pressure test (4) the main steam and feedwater system up to programs for NRC Ouality Group A, ASME Boiler and including the outermost containment Class 1, items.* It describes those programs isolation valve.

' implementing the requirements of Subsection IWB Pressure Vessel (B&PV) Code Section 111 and XI 5.7A.1J Exclusions q Portions of systems within the reactor

~

llems as used in this subsection are products coolant pressure boundary, as defined in constructed under a Certificate of Authorisation 5.2.4.1.1, that are excluded from the Class 1 (NCA 3120) and material (NCA 1220). See Section boundary are as follows:

111, NCA.1004 footnote 2. .

5.2 15.1 Amendment 13

ABWR zu6icorn Standard Plant new e (1) those components where,in the event of (2) RPV Welds Above Top of the Biological Shield i]f

/ postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly Wall Access to the reactor pressure vessel welds above manner, assuming makeup is provided by the top of the biological shield wallis prmided by the reactor coolant makeup system only; and removable insulation panels. This design provides reasonable access for both automated as well as (2) components which are or can be isolated manual ultrasonic examination.

from the reactor coolant system by two valves (both closed, both open, or one closed (3) Closure Head, RPV Studs, Nuts and Wasbers rnd one open). Each such open valve is capable of automatic actuation and if the The closure head is dry stored during refueling, other valve is open its closure time is such Removable insulation is designed to provide access that,in the event of postulated failure of the for manual uhrasonic examinations of closure head component during normal reactor operation, welds. RPV nuts and washers are dry stored and each valve remains operable and the reactor are accessible for surface and visual (VT 1) can be shut down and cooled dotvn in an examination. RPV studs may be volumetrically orderly manner assuming makcup is examined in place or when removed, provided by the reactor coolant mahup system only. (4) Bottom Head Welds 5.2.4.2 Accessibility Accer.s to the bottom head to shell weld and bottom head seam welds is prosided through openings in f All items within the Class 1 toundary are the RPV support pedestal and removable insulation designed, to the extent practicable, to provide panels around the cylindricallower portion of the L access for the examinations required by ASMC vessel. This design provides access for manual or Section XI,IWB 2500. Items for which the automated ultrasonic examinatis u equipment.

/G design is known to have inherent access Sufficient access is provided to partial penetration l

C restrictions arc described in Subsection 5.2.4.8. nozzle weIds, i.e., CRD penetrations, instrumentation nozzles and recirculation internal 5.2.4.2.1 Reactor Pressure Vessel Access pump penetration welds, for performance of the visual, VI' 2, examination during the system leakage Access for examinations of the reactor and system bydrostatic examinations.

pressure vessel (RPV) is incorporated into the L design of the vessel, biological shield wall and (5) Reactor Vessel Support Skirt vesselinsulation as follows:

The integral attachment weld from the number four l (1) RPV Welds Below the Top Biological Shield shell course forging to the RPV skirt will be l

Wall . examined ultrasonically. Sufficient access is "

provided for either manual or automated ultrasonic The shield wall and vesselinsulation behind examination. Access is provided to the balance of the shield wall are spaced away from the the support skirt for performance of visual, VT 3, RPV outside surface to provide access for examination.

remotely operated ultrasonic examination devices as described in Subsection 5.2.4.3.2.1 5.2.4.2.2 Piping, Pumps Valves and Supports Access for the insertion of automated devices is provided through removable Physical arrangement of piping pumps and valves insulation panels at the top of the shield wall provide personnel access to each weld location for and at access ports at reactor vessel aozzles. performance of ultrasonic and surface (rnagnetic Platforms are attached to the bioshield wall particle or liquid penetrant) examinations and sufficient to provide access for installation of remotely access to supports for performance of visual, VT 3, operated nozzle examination devices. examination. Working platforms are provided in some p

Amendment 13 5.2.t6

AB%R uxsioorn l Standard Plant a,, c .I areas to facilitate servicing of pumps and valves. IWA 2232 (a), and Section V, Article 4. In addition the Platforms and ladders are provided for access to ultrasonic examination system shall meet the piping welds including the pipe to-reactor vessel requirements of Regulatory Guide 1.150 as described in -

nozzle welds Removable thermal insulation is Table 5.2 9. RPV welds and nozzles subject to provided on welds and components which require examination are shown in Figure 5.2 7a.

frequent access for examination or are located in high radiation areas. Welds are located to permit The GE reactor vessel inspection system (GERIS) ultrasonic examination from at least one side, but meets the detection and sizing tequitements of where component geometries permit, access Regulatory Guide 1.150, as cited in Table 5.2 9. Inner from both sides is provided. radius examinations are performed from the outside of the nozzle using several compound angle transducer Restrictions: For piping systems and wedges to obtain complete coverage of the required i portions of piping systems subject to volumetric examination volume. E.'ectronic gating used in GERIS l

and surface examination, the following piping system reeords up t) 8 different reflectors designs are not used: simultaneously to assure taat all relevant indications are recorded. Appendix SA de nonstarted compliance with

1) Valve to Valve Regulatory Guide 1.150.

(2) Valve to Reducer l (3) Valve to Tec 5.2.4.3.2.2 Visual Examinatiot ;

(4) Elbow to Elbow (5) Elbow to Tec Visual examination methods, VT 1, VT 2 and VT 3, (6) Nozzle to elbow shall be conducted in accordance with ASME Section I (7) Reducer to elbow XI, IWA 2210. In addition, VT 2 examinations shall (8) Tee to tee meet the requirements of IWA 5240.

(9) Pump to valve Direct visual, VT 1, examinations shall be Straight sections of pipe and spool pieces conducted with sufficient lighting to resolve a 0.8mm l shall be added between fittings. The minimum black line on an 18% neutral grey card. Where direct (

length of the spool piece has been determined by visual, VT 1, examinations are conducted without the '

l using the formulate L = 2T + 152mm, where L use of mirrors or with other viewing aids, clearance (of equals the length of the spool piecc (not at least 610mm of clear space)is provided where feasible l i . including weld preparation) and T equals the pipe wall thickness. for arm'sthe head length and (508mm) shoulders of the surface to beof a man within a work examined.

5.2.4.3 Examination Categories and Methods At locations where leakages are normally expected i

and leakage collection systems are located, (e.g., valve 5.2.43.1 Examination Categories stems and pump seals), the visual, VT 2, examination shall verify that the leakage collection system is The examination category of each item is listed operative, in Table 5.2 8. The items are listed by system

j. and line number where applicable. Table 5.2 8 Piping runs shall be clearly identified and laid out i

also states the method of examination for each such that insulation damage, leaks and structural distress l

item. The preservice and inservice examination will be esident to a trained visual examiner.

' plans will be supplemented with detailed drawings showing the examination areas, such as 5.2.43.2.3 Surface Examination Figures 5.2 7a and 5.2 7b.

i Magnetic particle and liquid penetrant examination l 5.2.43.2 Examination Methods techniques shall be performed in accordance with i ASME Section XI, IWA 2221 and IWA 2222, 5.2.4.3.2.1 Ultrasonic Examination of the respectively. Direct examination access for magnetic Reactor Vessel particle (MT) and penetrant (PT) examination is the same as that required for direct visual (VT 1)

Ultrasonic examination of the RPV will be examination (Subsection 5.2.4.3.2.3), except that conducted in accordance with ASME Section XI, l

Amendment 15 5.2-17 I

ABWR 2 mima Standard Plant nev c !

.q additional access shall be provided as necessary to may result in improvements in examination reliability

Q enable physical contact with the item in order to perform the examination. Remote MT and PT and reductions in personnel exposure, generally are not appropriate as a standard 5.2.433 Data Recording examination process, however, boroscopes and mirrors can be used at close range to improve the Manual data recording will be performed where angle of vision. As a minimum, insulation manual ultrasonic examinations are performed, removal shau expose the area of each weld plus at Electronic data recording and comparision analysis are least 152mm from the toe of the weld on each to be employed with automated ultrasonic examination side, insulation wiu generally be removed 406mm equipment. Signals form each ultrasonic transducer on each side of the weld. will be fed into a data acquisition system in which the key parameters of any reflectors will be recorded, The 5.2.43.2.4 Volumetric Ultrasonic Dirret data to be recorded for manual and automated methods Examination are:

Volumetric ultrasonic dircet examination (1) Location shall be performed in accordance with ASME (2) Position Section XI,IWA 2232. In order to perform the (3) Depth below the scanning surface examination, visual eccess to place the head and (4) Length of the reflector l shoulders within 508mm of the area of interest (5) Transducer data including angle and frequency shall be provided where feasible. Nine inches (6) Calibration data between adjacent pipes is sufficient spacing if there is free access on each side of the pipes. The The data so recorded shall be compared with the transducer dimension has been considered: a results of subsequent examinations to determine the l 38mm diameter cylinder,76mm long placed with behavior of the reflector.

-access at a right angle to the surface to be p examined. The ultrasonic examination instrument 5.2.4.4 Inspection latervals l

Q has been considered as a rectangular box 305 x 305 x 508mm located within 12m from the The inservice inspection intervals for the ABWR i transducer. Space for a second examiner to will conform to Inspection Program B as described in I monitor the instrument shall be provided if Section XI,IWB 2412. Except where deferralis necessary, permitted by Table IWB 25001, the percentages of examinations completed within each period of the Insulation removal for inspection is to allow interval shall correspond to Table IWB.24121. Items sufficient room for the ultrasonic transducer to selected to be examined within the 10 year intervals are

, described in Table 5.2-8.

L l scan 152mmthe exarnination where area.

T is pipe thickness, Aminimum is the distance of 2T plus required on each side of the examination area.The 5.2.4.5 Evaluation of Examination Results l insulation design generally leaves 406mm on each side'of the weld, which exceeds minimum Examination results will be evaluated in accordance requirements. with ASME Section XI,IWB 3000 with repairs based l on the requirements of IWA 4000 and IWB 4000.

5.2.43.2.5 Alternative Examination Techniques Re examination shall be conducted in accordance with I the requirements of IWA 2200. The recorded results As provided by ASME Section XI, shall meet the acceptance standards specified in IWA 2240, alternative examination methods, a IWB.34001.

combination of methods, or newly developed l- techniques may be substituted for the methods 5.2.4.6 System bakage and Hydrostatic Pressure Tests 1 specified for a given item in this section, prosided that they are demonstrated to be equivalent or 5.2.4.6.1 System bakage Tests I

superior to the specified method. This prosision allows for the use of newly developed As tequired by Section XI,1WB 2500 for Category examination methods, techniques, etc., which B-P, a system leakage test shall be performed Amendment 15 5.2 17.1

ABWR 2 m ioox.

Standard Plant a, e in accordance with IWB 5221 on all Class 1 Section XI, Table IWB 25001 for category B D, will be components and piping within the pressure inaccessible for ultrasonic examination. The ultrasonic retaining boundary following each refueling examination !s conducted from the outside surface of outage. For the purposes of the system leakage the RPV and, as such, the outside radius of the nozzle test, the pressure retaining boundary is defined in forging limits the movement of the ultrasonic search Table IWB 25001, Category B P, Note 1. The unit. The typical scan limitation resulting from this system leakage test shallinclude a VT 2 geometry ls illustrated in Figure 5.2 7c. The extent of examination in accordance with IWA 5240. The the examination coverage limitation for each nozzle will system leakage test will be conducted be determined during the preservice examination of the approximately at the maximum operating RPV nozzle to vessel welds, pressure and temperature indicated in the applicable process flow diagram for the system as 5.2.412 Reactor Pressure Vessel Bottom Head Weld indicated in Table 1.71. The system hydrostatic test (Subsection 5.2.4.6.2), when performed is Access to the bottom head weld for ultrasonic acceptable in lieu of the system leakage test. examination is limited due to pecettations in the bottom head for the reactor internal penps, control rod 5.2.4.6.2 Hydrostatic Pressure Testa drives and incore monitors. Partial coverage will be possible. The extent of the examination coverage As required by Section XI,IWB 2500 for limitation for each nozzle will be determined during the Category B.P, the hydrostatic pressure test shall preservice examination the RPV bottom head weld.

be performed in accordance with ASME Section IWB 5222 on all Class 1 components and piping 5.2.4A3 Reactor Pressure Vessel Bottom Head to-Shell within the pressure retaining boundary once Weld during each 10 year inspection interval. For purposes of the hydrostatic pressure test the Access to the RPV bottom head to shell weld for pressure retaining boundary is defined in Table ultrasonic examination is limited due to the proximity of IWB 25001, Category B.P, Note 1. The system the RPV skirt pedestal on the shell side, and the curved hydrostatic test shallinclude a VI' 2 cxamination surface of the bottom head itself interferes with in accordance with IWA 5240. For the purposes examination from the bottom head side of the weld, of determining the test pressure for the system Only limited examination coverage can be achieved oc hydrostatic test in accordance with IWB 5222 (a), this weld, however, the actual extent of the examination the nominal operating pressure shall be the coverage will be determined during the preservice maximum operating pressure indicated in the examination of the RPV bottom head to shell weld.

process flow diagram for the nuclear boiler system, rigure 5.13.

l 5.2.4.7 Code Exemptions As provided in ASME Section XI, IWB 1220, certain portions of Class 1 systems are exempt from the volu.actric and surface- l l

examination requirements of IWB 2500. These portions of systems are specifically identified in Table 5.2 8.

5.2.4.8 Relief Requests 5J.4.8.1 Reactor Pressure Vessel Nozzles l~

Due to the inherent geometry of the RPV I nozzles, some portion of the nozzle to vessel weld required volume as specified in ASME Amendment 13 5.217J el

ABWR 2 mima g Standard Plant - REV.C r

of the leak detection and isolation system L (LDS). The LDS (Figure 5.2 8) consists of temperature, pressure, radiation and flow sensors with associated instrumentation, power supplies and logic used to detect, indicate, and alarm leakage from the reactor primary pressure boundary and,in certain cases (see Subsections 7.3.1.1.2, 7.6.1.3 a n d 7.7.1.7), t o initia t e closure of isolation valves to shut off leakage external to the containment. The system is designed to be in conformance with Regulatory Guide 1.45 (for leak detection funt.tions) and IEEE 279 (for isolation function).

Abnormal leakage from the following systems within the primary containment (drywell) and within selected areas of the plant outside the drywell (bott. Inside and outside the reactor building) is detected, indicated, alarmed, and, in certain cases, isolated:

(1) Main steamlines (2) Reactor core isolation cooling (RCIC) system (3) High pressure core flooder (HPCF)

(4) Residual heat removal (RHR) system

^

(5) Reactor water cleanup system (CUW)

(6) Feedwater system (7) Cuoiant systems within the drywell (8) Reactor pressure vessel (10) Miscellaneous systems Leak detection methods (in accordance with Regulatory Guide 1.45) differ for the plant areas inside the drywell as compared to those areas outside the drywell. These areas are considered separately as follows.

5.2.5.1.1 Detection of14akage Within Drywell 5.2.5 Reactor Coolant Pressure Boundary and Core Cooling Systems Leakage Detection The primary detection method for.small

- unidentified leaks within the drywell includes 5.2.5.1 leakage Detection Methods drywell floor drain sump pump activity and sump level increases, drywell cooler condensate flow RCPB leakage detection is a primary function rate increases, and airborne gaseous and particulate radioactivity increases. The Amendment 15 5.2 iB

ABM 2346ioorn Standard Plant RIV C e sensitivity of these primary detection methods appropriate isolation valves, l 3.785 for unidentified leakage liters / min within within one hour. the drywell These is The alarms, indication and isolation trip variables are continuously indicated and/or functions performed by the foregoing leak recorded in thc cootrol room. If the detection methods are summarized in Tables 5.2 6 unidentified leakage increases to the Technical and 5.2 7.

Specification limit, the detection instru-mentation channel (s) will trip and activate an Listed below are the variables monitored for alarm in the control room. No isolation trip detection of leakage from piping and equipment will occur, located within the drywell:

The secondary detection methods, pressure and (1) High drywell temperature temperature of the drywell atmosphere, are used to detect gross unidentified leakage. High (2) High temperature in the valve stem seal l drywell pressure will alarm and trip the (packing) drain lines isolation logic which will result in closure of the containment isolation valves. High drywell (3) High flow rate from ihm drywell floor and temperature is recorded and alarmed only, equipment drain suraps The detection of small identified leakage (4) High steamline flow rate (for leaks within the drywell is accomplished by monitoring downstream of flow elements in main steamline drywell equipment drain sump pump activity and an). RCIC steamline) sump levelincreases. The equipment drain sump levelinstruments and the fillup and/or pumpout (5) High drywell pressure l~ timers will activate an alstm in the control room j l when totalleak rate react es 95 liters / min. (6) High fission product radiation Equipment , drain sump pump activity and sump (7) Reactor vessellow water level IcVel increases will be caused primarily from naks from large process valves through valve (8) Reactor vessel head seal drain line high stem drain lines. pressure The determination of the source of other (9) SRV discharge piping high temperature.

identilhd leakage within the drywell is accomplishod by monitoring the reactor vessel 5.2.5.1.2 Detection orleakage External head seal drain line pressure, by monitoring to Drywell temperature in the valve stem seals drain line to the equipment drain sump, and by monitoring The areas outside the primary containment temperature in the SRV discharge lines to the (drywell) which are monitored for primary coolant j suppression pool to detect leakage through each leakage are: (1) the equipment areas in the of the SRVs.- All of these monitors continuously reactor building;(2) the main steam tunnel; and indicate and/or record in the control room and (3) the turbine building. The process piping, will trip and activate an alarm in the control for each system to be monitored for leakage, is l

room on detection of leakage from monitored located in compartments or rooms separate from

' components, other systems, where feasible, so that leakage may be detected by area temperature indications. Each Excessive leakage inside the drywell (e.g., leakage detection system will detect leak rates process line break or loss of coolant accident) that are less than the established limits for is detected by high drywell pressure, low reactor unidentified or identified leakage, water level or high steamline flow (for breaks downstream of the flow elements). The The monitored areas are monitored by single l instrumentation channels for these variables will element thermocouples for sensing high ambient p trip when the monitored variable exceeds temperature in each area. The temperature Q predetermined limits to activate an alarm and ele m e n t s are located or shielded trip the isolation logic which will close Amendment t5 5.2 19 I

k

ABM aa6tooan Standard Plant REV C so that they are sensitive to air temperature (1) Within reactor building:

only and not radiated heat from hot piping or equipment. Increases in ambient temperature or (a) Main steamline and RCIC steamline high in differential temperature will indicate Icakage flow of reactor coolant into the area. These monitors have sensitivities suitable for detection of (b) Reactor vessellow water level reactor coolant leakage into the monitored areas of 95 liters / min or less . The temperature trip (c) High flow rate from reactor building setpoint will be a function of the room size and sumps outside drywell the type of vect!Mion provided. These monitors provide alann and indication and recording in the (d) High ambient temperature or high I differential in equipment areas of RCIC, l control close theroom and will appropriate trip the isolation isolation valves, logic to e.g., the RHR, and the hot portions of the CUW l

main steam tunnel arra temperature rnonitors will close the main steamline .:9 MSL drab isolation (c) RCIC turbine exhaust line high diaphragm valves the CUW isolation valves, pressure Leakage detection will be provided in the (f) High differential mass flow rate in CUW turbine building. The turbine building monitors piping will also alarm and indicate in the control room and trip the isolation logic to close the main (g) High radiation in the RHR, CUW, and RIP, l steamline isolation valves and MSL drain and FPC reactor building cooling water isolation valves when leakage exceeds 95 heat exchanger discharge lines liters / min. (intersystem leakage)

Large leaks external to the drywell (e.g., (h) RCIC steamlinc low pressure process line breaks outside of the drywell) are detected by low reactor water level, high process (2) Within steam tunnel (between primary line flow, high ambient temperatures in the containment and turbine building):

l piping drain sumpor activi equipment areas,flow t y, high differential floor (CUWor equipment(a) High radiation in main steamlines (steam only), low steamline pressures or low main tunnel) condenser vacuum. These monitors provide alarm and indication in the control room and will trip (b) Main steam tunnel high ambient air the isolation logic to cause closure of temperature or high differential appropriate system isolation valves on the temperature l indication of excess leakage.

(3) Within turbine building (outside secondary Intersystem leakage detection is accomplished containment):

l by monitoring cooling radiation water (RCW) coolant of thelines return reactor building from the (a) Main steamline low pressure reactor internal pumps (RIP), residual heat removal (RHR), and reactor water cleanup system (b) Low main condenser vacuum (CUW) and fuel pool cooling heat exchangers.

l This monitoring is provided by the process (c) Turbine building ambient temperature in radiation monitoring system, areas traversed by main steam lines Listed below are the variables monitored for 5.2.5.2 leak Detection Instrumentation and detection of leakage from piping and equipment Monitoring located external to the primary containment (dryv ell):

Amendment t$ 5.2-20 9

l l

l

'M 2146100AI)

Standard Plant - uv c 6.23.2.1 tank Detection lastrumentation and - This flow is monitored by one channel of flow ON Monitoring inalde the Drywell instrumentation located to measure flow in i the common condensate cooler drain line which l (1) Drvwell Floor Drain Sumn Monitorine drains the condensate from all of the drywell coolers to the drywell floor drain sump. The I The drywell floor drain sump collects transmitter and its associated comparator unidentified leakage such as leakage from provide main control room flow readout and control rod drives, floor drains, valve trip and alarm on high flow conditicos flanges, closed cooling water for reactor approaching the unidentified discharge rate services (e.g., RIP motor cooling), limit. Location of the common header is such condensate from the drywell atmosphere that at least a 25% safety margin is coolers and any leakage not connected to the available for flow transmitter pressure head drywell equipment drain sump. The sump is requirements.

l . equipped with two pumps and specialinstru-mentation to measure sump fdlup and pumpaut (4) Drvwell Temocrature Monitorine times and provide continuous sump level rate of change monitoring with control room The ambient temperature within the drywell is indleation and alarm capabilities for monitored by four single element excessive fill rate or pumpout frequency of thermocouples located equally spaced in the the pumps. The drain sump instrumentation vertical direction within the drywell. An has a sensitivity of detecting reactor abnormal increase in drywell temperature l coolant leakage of 3.785 liters / min within a could indicate a leak within the drywell.

60 minute period. The alarm setpoint has an adjustable range up to five gpm for the drywell floor drain sump.

(2) Drvwell Eauinment Drain Sumn Monitorine Ambient temperatures within L

Oi

~

The drywell equipment drain sump collects the drywell are recorded and alarmed in the l main control room. Air temperature only identified leakage from identified monitoring sensors are located such that they leakage sources. This sump monitors leakage are sensitive to reactor coolant leakage and .

.l from valve stem packings, RPV head flange not to radiated heating from pipes and seal, and other known leakage sources which equipment, are piped directly into the drywell equipment drain sump. The number of sump - (5) Drvwell Fission Product Monitorina pumps and the types of. drain sump .

instrumentation is the same as that used for Primary coolant leaks within the drywell are the drywell floor drain sump.- The detected by radiation monitoring of monitoring channels measure sump level rate continuous drywell atmosphere samples. - The of change and sump fillup and pumpout times, fission product radiation monitors provide with main control room indication and alarm gross counting of radiation from radioactive capabilities. Collection in excess of particulates, and radioactive gases. The l background leakage would indicate an count levels are recorded in the control room increase in reactor coolant leakage from an and alarmed on abnormally high activity.

identiliable sourcc. leve1. ,

(3) Drvwell Air Cooler Condensate Flow (6) Drvwell Pressure Monitoring Monitoring Drywell pressure is monitored by pressure The condensate flow rates from the drywell transmitters which sense drywell pressure atmosphere coolers are monitored for high relative to reactor building (secondary drain flows, which indicate leaks from containment) pressure. Four channels of piping or equipment within the drywell.

O, Amecdment 15 5.2 21

-ABWR ==

Syndard Plant REV. C drywell monitoring are provided by the cleanup, reactor core isolation cooling, and nuclear boiler system. A pressure rise residual heat removal systems are fitted with above the normally indicated values will drain lines from the valve stems, from indicate a possible leak or loss of reactor between the two sets of valve steam packing.

coolant within the drywell. Pressure Leakage through the inner packing is carried exceeding preset values will be alarmed in to the drywell equipment drain sump. leakage the main control room and required safety during bydro-testing may be observed in drain action will be automatically initiated. -line sight glasses installed in each drain '

line. Also, each drainline is equipped with (7) Reactor Venel Head Finane Seal Monitating temperature sensors for detecting leakage. A remote operated solenoid valve on each line A single channel of pressure monitoring is may be closed to shut off the leakage flow provided for measurement and control room through the first seal in order to take indication of pressure between the inner and advantage of the second seal, and may be used outer reactor head flange seals. High during plant operation, in conjunction with pressure will indicate a leak in the inner the sump instrumentation, to identify the O ring seal. This high pressure is specific process valve which is leaking.

annunciated in the main control room (no isolation). A pressure tap for this (11) Main Steamline Hlah Flow Monitorina (for measurement is provided by the nuclear leaks downstream of flow elemenid boiler system. Leakage through both inner and outer seals will be detected by other High flow in each main steamline is monitored i drywellleak detection instrumentation. Any by four differential pressure transmitters '

leakage through the inner seal can be that sense the pressure difference across a directed to the drywell equipment drain flow restrictor in that line. The pressure sump, taps are included with the main steam system. High flow ' rate in the. main .

(8) Reactor Recirculation Pumn Motor Lenkane steamlines during plant operation could Monitorine indicate a break in one or more of the lines. High flow exceeding preset values in l Excess leakage from the RIP motor casing any of the four main steamlines will result will be detected by the drywell floor drain in trip of the MSIV isolation logic to close

. sump monitors described in (1) above, all the MSIVs and the MSL drain valves and 3 annunciate in the main control room. Each

. (9) Safetv/ Relief Valve l eakaoe Monitorina monitoring channel will also include control ,

room recording by the PMCS.

SRV leakage is detected by temperature . .

sensors located on each relief valve (12) Reactor Vaual low Water izvel Mnnitorina discharge line such as to detect any valve outlet port flow. Each of the temperature The nuclear boiler system provides reactor channels includes control room recording and water level monitoring for the LDS functions alarm capabilities. The temperature sensors and for safety f"nctions of other systems, are mounted using thermowells in the Sixteen channels of monitoring (four in each discharge piping several feet from the valve division to provide trip signals at four 1 body to prevent false indication. The different water levels, i.e., levels 3, 2,

monitoring of this leakage is provided by 1.5 and 1) are provided for the LDS the nuclear boiler system. functions, e.g., RHR, CUW, MSL and isolations of other portions of the plant. The safety

.(10) Valve Stem Packina Lenkase Monitorina related performance requirements of the level monitoring channels are a function of the Large (two inch or larger) remote nuclear boiler system.

power operated valves located in the drywell for the nuclear boiler, reactor water (13) RCIC Steamline Flow Monitorine (for leaks downstream of flow elements Amendment t$ $.2-22

ABM 2mioarn nrv. c Standard Plant The steam supply line for motive power for piping, process instrumentation piping or

,) operation of the RCIC turbine is monitored control rod drive hydraulic control unit >

v for abnormal flow. Four channels of flow piping, is collected in several reactor l building floor drain sumps. Background measurement are provided for detection of steamline breaks downstream of the flow leakage is identified during preoperational elements by LDS flow transmitters which tests. The number of pumps and the l sense differential pressure across elbow instrumentation used for monitoring both the taps in the RCIC turbine supply steamline. reactor building floor and equipment and High steamflow exceeding preset values will equipment drain sumps, are similar to those result in the closure of the RCIC steamline used for monitoring the drywell floor drain isolation valves, warmup bypass valve, and sump as described in Subsections 5.2.5.2.1(1) trip the turbine isolation valve, Isolation and 5.2.5.2.1(2). The reactor building floor trip signals from one division will tiose and equipment drain sump monitoring channels the outboard isolation valves w' ale trip measure sump levels and sump fillup and signals from a second divisica will close pumpout times and initiate alarms when the inboard RCIC steamlics isolation valve setpoints are exceeded, and warmup bypass velve Any isolation signal to the RCIC logic will also trip the (3) Reactor Water Cleanun System Differential RCIC turbine. LDS mea,urements are taken as Flow Monitorinn close to tbr reactor vessel es possible to maximin LDS coverage. The suction and discharge flows of the reactor water cleanup system are monitored Tables 5.2-6 and 5.2 7 summarize the actions for flow differences between that coming from taken by each leakage detection function. Table the reactor and that returning to the reactor 5.2 6 shows the systems which detect gross or to the main condenser. Temperature comp.

leakage and those in which immediate automatic ensated flow differences greater than prefet isolation is initiated. The systems which are values cause alarm and isolation. Bypas

' O tapable of detecting small leaks initiate an slarm-in the control room (Table 5.2 7). The time delay interlocks are provided fc-delaying the isolation signals and prevent isolation initiation during normal CUW surge operator may manually isolate the leakage sources or take other appropriate action, conditions. Flow in the CUW suction line from the reactor and in the CUW return lines 5.2.5.2.2 14ak Detsetion instrumentation and to the reactor and in the blowdown liac to Monitorlag Estemal to Drywell the radwaste system is monitored by twelve differential flow transmitters (four for each

. (1) Visual and Audible Insnection line). CUW flow measurements are taken as close to the reactor vessel as possible to Accessible areas are inspected periodically maximize the degree of coverage of the LDS and the temperature, pressure, sump level channels. The outputs of the flow and flow indicators discussed below are transmitters in the suction line are compared i monitored regularly. Any instrument with the outputs from the discharge lines and

l. Indication of abnormal leakege will be alarms in the control room and isolation l . investigated, signals are initiated when higher flow out of j the reactor vesselindicates that leaks equal L (2) Reactor Buildine Floor and Eaulement D,mjn to the established leak rate limits for alarm

!. Sumn Monitorine or isolation may exist. Net flow indication li readout is provided in the control room.

Reactor building equipment drain sumps collect the identified leakage from known (4) b4ain Steamline Area Temnerature Monitors sources from within onclosed equipment areas. Leakage from unknown or unidentified High temperature in the main steamline tunnel L

sources, such as from shutdown cooling

system piping, reactor water cleanup system Amendment 15 5.2 23

ABM 2346ioorn l Standard Plant arv. c l

area is detected by single element thermo- signal and a trip signal for that division's l

couples. Four thermocouples are used for isolation logic to close the respective measuring main steam tunnel ambient system isolation valves.

temperatures and are located in the area of

l. the main steam lines tunnel area. All temperature elements are located or shielded so as to be sensitive to air temperatures

.l and not to the radiated heat from hot

. equipment. High ambient temperatures will j alarm in the control room and provide i signals to close the main steamline and MSL l drain line isolation valves, and the CUW l isolation valves. . High ambient temperature l in the steam tunnel area can also indicate leakage from the reactor feedwater piping or equipment within the tunnel. Isolation of the feedwater lines, if necessary, may be accomplished by manual closure by the (6) Main Steamline R=4=tian Monitorina operator of valves located in the feedwater lines in the steam tunnel. Monitoring of Main steamline radiation is monitored by the main steamline area outside the steam gamma sensitive radiation monitors of the tunnel and before the inlet to the turbine process radiation monitoring system (PRRM) as is provided with sufficient ambient . a functional interface to the LDS (and RPS),

temperature sensors to cover the fulllength The PRRM provides four divisional channel of the steam lines in the turbine building. trip signals to the LDS to close all MSIVs ,

and the MSL drain valves upon detection of j high radiation in-the main steamline area, The channel signals are combined so as to downstream of the outboard MSIVs. A reactor provide the four divisional trip signals trip (scram) is also initiated by the same .

.used as inputs to the LDS isolation logic PRRM channel trip signals. The detectors are l for closure of the MSIVs and MSL drain geometrically arranged to detect significant

lines. High ambient turbine building increases in radiation level with any number

. temperatures (main steamline areas) will of main steamlines in operation. Control

also be indicated in the control room. The room readout and alarms are provided by the: '

turbine building temperature elements are PRRM system.

located.so as not to be sensitive to

. radiated heat from hot equipment. .(7) RCIC Stenmune Pressure Monitors (5)iTemnerature Monitors in Eauinment Areas Pressure in the RCIC steamline is monitored by LDS instruments to' provide RCIC turbine Dual element thermocouples are installed in. shutoff and closure of the RCIC isolation l the RCIC, RHR and CUW equipment rooms for valves on low steamline pressure as'a l sensing high ambient temperature in these . protection for the RCIC turbine. This steam-areas. These elements are located or line pressure is monitored by four pressure shielded so that they are sensitive to air transmitters, each connected to one taps of temperature only and not to radiated heat the two elbows used for RCIC steamflow from- hot equipment. Four ambient measurement, and upstream of the RCIC

. temperature channels are provided in each steamline isolation valves (see Subsection equipment area. ' Each of the four channels 5.2.5.2.1 (13)). Low pressure is alarmed in the control room and low pressure isolation

~

drive voting logic in two divisions (three divisions for RHR) which provides an alarm signals close the same RCIC valves as those- l closed by the RCIC steamflow monitoring instruments.

Amendment 15 3.2 24 1

l

ABM 2mioorn an c Standard Plant (8) RCICTurbine hhautt Line Diaohrmem Prenure reactor coolant leakage into the reactor

/' .

Monitors building cooling water (RCW) system which l.

supplies coolant water to the RHR beat exchangers, to the reactor internal pumps Pressure between the rupture dise diaphragms in the RCIC system turbine exhaust vent line (RIPS) beat exchangers, to the CUW non. regen. I is monitored by four channels of pressure crative heat exchangers, and also to the fuel i l lastrumentation. The instrumentation pool cooling heat exchangers. One process channel equipment and piping are provided by sensing channel is provided in each of the the RCIC system as an interface to the LDS. three RCU loops to monitor for radiation due The two logic channels of division I trip on to coolant leakage into the RSW. Each high pressure to close the inboard RCIC channel will alarm on high radiation ,

isolation valves and the channels of conditions indicating process leakage into l Division 11 trip to close the outboard the RCW system. The process radiation - l isolation valves. Either divisional logic monitoring system prosides the monitoring of channel will also trip the turbine. this variable No isolation trip functions are perfctmed by these monitors. J (9) Main Steamline & Pressure Monitorina (12) Larne 12aks External to the Drvwell Main Steamline low presure is monitored by four pressure t'ransnMers (one in each The main steamline high flow monitoring, the line) that sense the pressun downstream of reactor vessel low water level monitoring and the outboard MSIVs. The sensing points are the RCIC steamline flow monitoring, discussed '

located as close as possible to the turbine in Subsection 5.2.5.2.1, Paragraphs 11,12 stop valves.L Low steamline pressure at the and 13, can also indir. ate large leaks from points monitored can be an indication of an the reactor coolant piping external to the excessive steamline leak or a malfunction of drywell.

the reactor pressure control system. The O

V-transmitters are provided by the nuclear 5.2.5.2.3 Summary boiler system. The LDS will automatically !

initiate closure of all MSIVs and the MSL Tables 5.2 6 and 5.2 7 summarize the actions-

. drain valves if pressure at the turbine end - taken by each leakage detection function. Table of the main steamlines decreases below a - 5.2 6 shows that those systems which detect gross ,

preselected value when the reactor mode leakage initiate immediate automatic isolation

. switch is in the *RUN" position. action to terminate the gross leakage or minimize loss of reactor coolant. The systems which are (10) Main Condenser Low Vacuum Manitorinn capable of detecting smallleaks initiate an alarm ,

in the control room as shown in Table 5.2 7. In l Low main condenser vacuum could indicate addition, Table 5.2 6 shows that two or more

, that primary reactor coolant is being lost leakage detection methods are provided for each l l through the main condenser. Four channels system or area that is a potential-source of q

j. of mkin condenser pressure monitoring are leakage. Plant operating procedures will dictate L provided by the nuclear boiler system. The the action an operator is to take upon receipt of '

LDS utilizes the low vacuum signals to trip an alarm from any of these systems.- The operator the MSIV logie on low condenser vacuum and can manually isolate the violated system or take close all MSIVs and the MSL drain valves. other appropriate action.

The condenser sacuum trip signals can be bypassed by a manual keylocked bypass switch l

In the control room during startup and shutdown operations.

(11) Intersystem 12akane Monitorinn Radiation monitors are used to detect

a. 5.2 25 Amendment t$

l l

MM 23A6100AB RTV.C Standard Plant liters / min and 3.785 liters / min, respectively. l The total leakage rate limit is established low enough to prevent overflow of the sumps. The equipment drain sumps and the floor drain sumps, j which e

A time delay is provided for two 10 mcollect allleakage, are each pumped out by 3 /hr pumps.

CUW differential flow isolation signals to prevent system isolation during CUW surges, if either the total or unidentified leak rate limits are exceeded, an orderly shutdown shall be Tbc LDS is a four divisional channels which initiated and the reactor shall be placed in a are redundantly designed so that failure of any cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , single element within a channel will not interfere with a required detection of leakage or 5.23.4.2 Identined bakage Inside Drpell a required isolation. In the four division LDS, where inadvertent isolation could impa!r plant The valve stem packing of large power operated performance (e.g., closure of the MSIVs), any valves, the reactor vessel head flange seal and single channel or divisional component other seals in systems that are part of the malfunction will not cause a false indication of reactor coolant pressure boundary, and from which L

l l leakage and will not cause a false isolation normal design identified source leakage is l trip. Only one of the four channels will trip expected, are provided with leakoff drains. The and two or more channels are required to trip :n nuclear system valves inside the drywell and the-order to cause closure of the main steamline reactor vessel head flange are equipped with isolation valves. The LDS thus combines a very double seals The leakage from the inner valve l high probability of operating when needed with a stem packings and from the reactor vessel head very low probability of operating falsely, The flange inner seal, which discharge to the drywell system is testable during plant operation. equipment drain sump, are measured during plant l-l operation. Leakage from thr main steam SRVs, 5.2.5.3 Indication in the Control Room discharging to the suppression pool, is monitored l _

by temperature sensors mounted in thermowells in l Leak detection methods are discussed in the individual SRV exhaust lines. The f Subsection 5.2.5,1, Details of some of the LDS ttermowells are located several feet from the L alarms, recordings and other indleations in the ialve bodies so as to prevent false indication.

control room are discussed in Subsections These temperature sensors transmit signals to the 4 5.2.5.1,1, 5.2.5.1.2, 5.2.5.2.1 a n d 5.2.5.2.2, sontrol room for monitoring. Any temperature l l

Further details of.the LDS control room increase detected by these sensors, that is above indications are included in Subsections the ambient temperatures, indicates SRV leakage.

l 7.3.1.1.2. 5.23.5 Unidentined bakage inside the Drywell 5.23,4 Limits for Reactor Coolant hakage 5.233.* Unidentined bakage Rate -

5.23.4.1 Total bakage Rate The unidentified leakage rate is the portion The total reactor coolant leakage rate con- of the total leakage rate received in the drywell L sists of all leakage, identified and unidenti- ' sumps that is not identified as previously L fled,-that flows to the drywell floor drain and described. A threat of significant compromise to I equipment drain sumps. The total leakage rate the nuclear system process barrier exists if the limit is well within the makeup capability of the barrier contains a crack that is large enough to RCIC system'(182 m 3 /hr). The total reactor propagate rapidly (critical crack length); The coolant leakage rate limit is established at 95 unidentified leakage rate limit must be low be.

liters / min. The identified and unidentified cause of the possibility that most of the leakage rate limits are established at 95 5.2 26 O

Amendment 15

ABWR mams Standard Plant nrv. c unidentified leakage rate might be emitted from a The established limit is sufficiently low so d(3 single crack in the nuclear systern process that, even if the entire unidentified leakage barrier. rate were coming from a single crack in the nuclear system process barrier corrective action An allowance for leakage that does not could be taken before tb integrity of the

barrier would be threatened.

l identifiable compromise barrier integrity is established andplant for normal is not operation. The leak detection system will satisfactorily detect unidentified leakage of 3.785 liters / min l The unidentified leakage rate limit is within the drywell.

l corrective established actionat before 3.785the liters / minbarrier process to allow time 5.2.5.6 for Differentiation Between identified and could be significantly cornpromised. This Unidentified L4aks unidentified leakage rate is a small fraction of the calculated flow from a critical crack in a Subsection 5.2.5.1 describes the leak l primary system pipe (Appendix 3E). detection methods utilized by the leak detection system. The ability of the leak detection system to differentiate between identified and '

unidentified 5.2.5.4 and 5.2.5.5.leakage is discussed in Subsections '

5.2.5.7 Sensitivity and Operability Tests 5.23.5.2 Margins of Safety Sensitivity, including sensitivity tests and The margins of safety for a detectable flaw response time of the leak detection system, and to reach critical size are presented in the criteria for shutdown if leakage limits are

/N- Subsection 5.2.5.5.3. Figure 5.2 8 shows general exceeded are covered in Subsections 5.2.5.1.1, h relationships between crack length, leak rate, 5.2.5.1.2, 5.2.5.2,1(1) and 7.3.1.1.2.

stress, and linesize using mathematical models.

Testability of the LDS is contained in 5.215.3 Criteria to Evaluate the Adequacy and Subseetion 7.3.1.1. 2 ( 10) . l Margin of L4ak Detection System 5.2.5.8 Testing and Calibration For process lines that are normally open, there are at least two different methods of Provisions for testing and calibration of the detecting abnormalleakage from each system leak detection and isolation system are covered comprising the nuclear system process barrier, in Chapter 14.

located both inside the primary containment (drywell) and external to the drywell, in the 5.2.5.9 RegulatoryGulde 1A5: Compliance reactor building the steam tunnel and the turbi '

building (Tables 5.2 6.and 5.2 7). The These guidelines are prescribed to assure that instrumentation is designed so it can be set to leakage detection and collection systems provide provide alarms at established leakage rate limits maximum practical identification of leaks from and isolate the affected system if necessary, the RCPB.

The alarm points are determined analytically or based on measurements of appropriate parameters Leakage is separated into ider.tified and made during startup and preoperational tests, unidentified categories and each is independently monitored, thus meeting Position C.1 The unidentified leakage rate limit is based, requirements.

with an adequate margin for contingencies, on the crack size large enough to propagate rapidly. Leakage from unidentified sources fr.om inside the drywell is collected into the floor drain O sump and monitored with an accuracy better than D

Amendment LS 5.2 27 l

l l

.l l

l

ABM 23^61oarn REV.C Standard Plant 3,785 liters / min thus meeting Position C.2 and testingis provided.

i requirements, These satisfy Position C.8 requirements.

By monitoring (1) fic.or drain sump fillup and pumpout rate, (2) airborne particulates, and (3) Limiting unidentified leakage to the 3.785 air coolers condensate flow rate, Position C.3 is liters / min and identified to 95 liters / min satisfied, satisfies Position C.9.

o Monitoring of the reactor building cooling water heat exchanger coolant return lines for 5.2.6 Interfaces l CUW radiation and the due fuelto leaks pool within cooling theheat system RHR, RIP 5.2.6.1 andChemistry Water exchangers satisfies Position CA. For system detail, see Subsection' 7.6.1.2. The remainder of plant will meet the water chemistry requirements given in Table 5.2 5.

The floor drain sump monitoring, air particu-lates monitoring, and air cooler condensate moni- 5.2.6.2 Conversion of Indications l 3.785 toring are liters designed

/ min within oneto detect hour, leakage Procedures thus snecting rates of and graphs will be provided to Position C 5 requirements, operations for converting the various indicators into a common leakage equivalent (See Subsection The fission products monitoring subsystem is 5.6.5.9).

qualified for SSE. The containment floor drain sump monitor, air cooler, and condensate flow 5.2.7 References meter are qualified for OBE, thus meeting Position C.6 requirements. 1. (Deleted)

Leak detection indicators and alarms are provided in the main control room. This -

satisfies Position C.7 requirements. Procedures 2. (Deleted) .

and graphs will be provided by the applicant to plant operators for converting the various 3. - D.A. Hale, The Effect of BWR Startup En- ;

indicators to a common leakage equivalent, when' vironments on Crack Growth in Structural 0

-necessary, thus satisfying the remainder of Alloys, Trans, of ASME, vol 108, January Position C.7 (See interf ace requirements 1986.

Subsection 5.2.6.2). The leakage detection system is equipped with provisions to permit 4. F.P. Ford and M. J. Povich, The Effect of testing for operability and calibration during Oxygen / Temperature Combinations on the the plant operation using the following methods: Stress Corrosion Susceptibility of Sensi-tised'T 304 Stainless Steel in High Purity (1)' simulation of trip signal; Water, Paper 94 presented at Corrosion 79, Atlanta, GA, March 1979.

(2) ' comparing channel to channel of the same leak detect' . method (i.e., area tempera. 5. BWR Normal Water Chemistry Guidelines: 1986 ture monitormg); Revision, EPRI NP-4946-SR, July 1988.

(3) operability checked by comparing one method 6. B.M. Gordon, The Effect of Chloride and versus another (i.e., sump fillup rate ver. Oxygen on the Stress C.arrosion Cracking of sus pumpout rate and particulate monitoring Stainless Steels: Review of Literature, or air cooler condensate flow versus surnp Material Performance, NACE, Vol.19, No. 4, fillup rate); and April 1980.

(4) continuous monitoring of floor drain sump 7. W.J. Shack, et al, Environmentally Assist-l level, and a source of water for calibration ed Cracking in Light Water Reactors: Annual Report, October 1983 September 1984, Amendment 15 5.2-28

ABWR nuixa saudariinnt nrv c NUREG/CR-4287, ANL 85 33, June 19S5.

8. D.A. Hale, et al, Bil'R Coolant inapurities Program, EPRI, Palo Alto, CA, Pinal Report on RP2293 2, to be published.

9. K.S. Brown and G.M. Gordon, Effects of Bil'R Coolant Chemistry on the Propensiy ofIGSCC

< initiation and Growth in Creviced Reactor Internals Components, paret presented at g the Third International Symposium of Envi-d ronmental Degadation of Materials in Nucle-k at Power Systems, ANS NACE.TMS/AIME, Traverse City, Michigan, September 1987.

@I n~ 10. B.M. Gordon et al, EAC Resistance of Bit'R Materials in Hil'C, Fredeeding of Second International Symposium Environmental Degration of Materials in Nuclear Power

Systems, ANS, LaGrange Park, ILL 1986.

10a. BWR Hydrogen Water Chemistry Guidelines:

. 1987 Revision EPRI NP 4947.SR, December '

l 1988.

E 10b. Guldline for Permanent BWR Hydrogen Water Chemistry Installations: 1987 Revision, EPRI NP 5203 SR.A.

.g l 11. B.M. Gorden, Corrosion and Corrosion a Controlin Bil'R's, NEDE.30637, December

g. 1984.

lDE h 7, 12. B.M. Gordon et al, Hydrogen IVater Chemisty l"} for Bil'R's. Materlois Beharlor, EPRI NP.5080, Palo Alto, CA, March 1987.

E O ;

Amendmu U 5.2 %1 l

l ABWR mamn i Etandard Plant ntv c i

o. t Table 5.2 2 V SYSTEMS WHL"H MAY INITIATE DURING OVERPRESSIJF.E EVE.*T l

i

Sand laulaJna/Trin knal

Reactor Protection Resetor shutdown on high flux RCIC ON when reactor water levelis at L2 OFF when reactor water levelis at L8 1 Recirculation System Four pumps OFF when reactor water level is at L3 Remaining six pumps OFF when reactor water levelis at L2 Four pumps (the same four tripped at L3)

OFF when reactor pressure is at 79.1 kg/cm2 g CUW OFF when reactor water level'is at L3

(

I

Vesselleveltrip settings (Figure 3.3 2). ,

1.

l i

Amendment 15 $ 2.M

MWR zwi=^n "Y C Standard Plant Table 5.2 3 NUCLEAR SYSTEM SAFETY / RELIEF VALVE SETPOINTS Set Pituures and Capacities ASME Rated Capacity at 103%

Spring Spring Set Relicf Number

Set Preuure Preuure of Preuure (kg/hr Set Pressure Valves (kg/cm2g) each) (kg/cm2g) 1 80.8 395,000 76.6 1 80.8 395,000 77.3 4 81.$ 399,000 78.0 4 82.2 402,000 78.7 4 82.9 406,000 79.4 4 83.6 409,000 80.1

Eight of the SRl's serve in the automatic depressurhation function.

O Amendment 13 5.2 31

_ _ =----_ _ _ - _ - - - - - - - _ - _ - - _ _ - _ _ - _ _ _ - _ - . _ _ _ - . .

t ABM 2mixin Standard Plant av c !

Table 5.2 4 i

REACTOR COOLANT PRESSURE BOUNDARY MATEklALS Component [ grin Material Snecincation (ASTM /ASMD Main Steam l Valve Body Cast Carbon steel SA312 LCB Cover Forged Steel SA350LF2 Foppet Forged Steel SA350LF2 Valve stem Rod 17-4 pH SA SM 630 Body bolt Bolting Alloy steel SA 540 B23 CLS

, Hex nuts Bolting Nots Alloy steel SA 194 GR7 Main Steam Saferv/ Relief Valtg Body Forging Carbon steel ASME SA 105 or Casting Carbon steel ASME SA 352 LCB Bonnet (yoke) Forging Carbon steel ASME SA 105 l or Casting Carbon steel ASME SA 352 LCB NonJe (seat) Forging Stainless steel ASME SA 182 Gr F316 or Casting Carbon steel ASME SA 350 LF2 Body to Bar/ rod low allow steel ASME SA 193 Gr B7 bonnet stud Body to Bar/ rod Carbon steel ASME SA 194 Gr 2H bonnet out Dise Forging Alloy steel ASME SA 637 Gr 718 or Casting Stainless steel ASME SA 351 CF 3A Spring washer Forging Carbon steel ASME SA 105 Adjusting Screw Alloy steel ASME SA 193 Gr B6 or Set point adjust. Forgings Carbon and alloy Multiple specifications ment assembly steel parts Spindle (stem) Bar - Frecipitation. ASTM A564 Type 630 hardened steel Spring Wire or Steel ASTM A2 4 Gr 4161 N Belhille washers Alloy steel 45 Cr Mo V67 l

Main Steam Pininn l

Pipe Seamless Carbon steel SA 333 Gr. 6 Contour nonje Forging Carbon steel SA 350 LF 2 200A 1500# Forging Carbon steel SA 350 LF 2 large groove flange l Amendment 15 5.2 32 l

l 1

AB\M 22^62=^n RIV C Standard Plant Table 5.24 REACTOR COOLANT PRESSURE BOUNDARY MATERIALS (Continued)

Component EDIE MAltdA} Snecincation (ASTM /ASME) 50A special Forging Carbon s'.cel SA 350 LF 2 nozzle Elbow Seamless Carbon steel SA 420 Head fitting /pene- Torging Carbon steel SA 350 LF 2 tration piping O

O Amendment 13 5.2 32.1

ABWi ursioasa REY C Standard Plant

(~'

\

Table 5.24 REALTOR COOLANT PRESSURE BOUNDARY MATERIALS (Continued)

Component E9DE Material Enecification (ASTM /ASME)

Recirculation Purnp Motor Housing and Heat Exchancer Pump Motor Casing Forging Carbon steel ASME SA $08CL.I or til Stud (ease) Bolting Alloy steel ASME SA 540CL 3 Grp B24 i or SA 193, B7 i Nut (ease) Bolling Alloy steel ASME SA 194 Grp 7 Bottom flange (cover) Forging Carbon steel ASME SA $33 Gr B Class I or SA $08 CL 111 Motor heat exchanger Forging Carbon steel SA $33 B1, SA $08 ULIll :

or plate 1

Shell & Heads !

Motor coolant pipe Pipe Stainless steel ASME SA 312 TP 304L Thermowell Bar Stainless steel ASME SA 479 TP 304L i

O V m

)

Middle flange Forging Stainless steel SA 182, F304L or 316L Spool piece Forging Stainless steel SA 182, F304L or 316L Mounting bolts Bar Alloy steel SA 194, B7 Seal housing Forging Stainless steel SA 182, F304L or 316L ;

Seal shaft Bar Stainless steel SA 479, XM 19 ,

Shaft Nut Bar Stainless steel SA 564,17 4PH i

Reactor Pressure Vessel Plate Mn 1/2 Mo-1/2 Ni SA 533, Type B Class 1 Forging 3/4 Ni 1/2 Mo- SA 508, Class 3 1 Cr V Low Alloy Jorging C Si SA 508 Class 1 )

i Carbon steel i Forging Ni Cr Fe SB 166 Type 60 High nickelalloy

~

Forging 16 Cr 12 Ni 2 Mo SA 182 F316L .

High alloy Type 316 Stainless steel Amendment 15 5.2-33

MM 21 w 00AD RIN C Sandard Plant Table 5.24 REALTOR COOLANT PRESSURE BOUNDARY MATERIALJi (Contloped)

Component Entm Material EnNification (ASTM /ASMfd Recirculation Pumn Motor Housinc and Heat Exchancer Pump Motor Casing Forging Carbon steel ASME SA 508CL.I or til Stud (case) Bolling Alloy steel ASME SA $40CL.3 Grp B24 i or SA 193,B7 i Nut (ease) Bolting Alloy steet ASME SA 194 Grp 7 Bottom flange (cover) Forging Carbon steel ASME SA 533 Gr. B Class I or SA 508 CL Ill Motor heat exchanger Forging Carbon steel SA 533 B1, SA $08 CLIII or plate Shell & Heads Motor coolant pipe Pipe Stainless steel ASME SA 312 TP 304L Thermowell Bar Stain!eas steel ASME SA-479 TP 304L CED Middle flange Forging Stainless steel SA 182, F304L or 316L Spool piece Forging Stainless steel SA 182, F304L or 316L Mounting bolts Bar Alloy steel SA 194, B7 Seal housing Forging Stainless steel SA 182, F304L or 316L Seal shaft Bar Stainless steel SA 479,XM 19 ,

Shaft Nut Bar Stainless steel SA 564,17 4PH Reactor Pressure Vessel Plate Mn 1/2 Mo 1/2 Ni SA 533, Type B Class 1 Forging 3/4 Ni 1/2 Mo- SA 508, Class 3 Cr V Low Alloy Forging C Si SA 508 Class 1 Carbon steel Forging Ni Cr Fe SB 166 Type 60 High nickel alloy Forging 16 Cr 12 Ni 2 Mo SA 182 F316L High alloy Type 316 Stainless steel Amendment 13 5.2 33 l

l+

23A6100AB l Standard Plant RFY C Table 5.2 5 !

BWR WATER CHEMISTRY Electro.

Chemical Corrosion ,

Concentrations

ConductMtv Potential Parts Per Billion (ppb) pS/cm pHat iron Conner Chloride Sulfate Orvnen" at 250C 250C V at 250C Condenute < 20 <2 <4 <4 <10 - 0.075 -

l Condensate Treatment Effluent and Feedwater < 2.2 < 0.1 <032 <032 20 50 <0.059 -

l Jteactor Water ,

(a) Normal Operation < 20 <1 < 20 < 20 " <03 ~7 < 0.23 (b) Shutdown < 20 <1 < 20 < 20 - < 1.2 ~7 -

(c) Hot Standby < 20 <1 < 20 < 20 < 200 <03 ~7 -

(d) Depressurized < 20 <1 < 20 < 20 high (may <1.2 5.6-8.6 -

be 1000 to 8000)

Control Rod Drive l Coolina Water < 2,2 < 0.1 <032 <032 20 50 < 0.059 --

t i i

These limits should be met at least 90% of the time.

Some revision of anyen values may be established after hydrogen water ehemistry has been established.

O I

Amendment 13 5.2.M i

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. _ _ _ _ _ . . . - . - __ __._____.___..____.._._...._.._.-_..m___.__ . _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ _ . . - . . _ _ _ . . _ . . _ _ _ . . . . _ .

1 23A6100AB l- Riandard Plant arv. e LEAKAGE SOURCES VS MONITORED TRIP ALARMS Table 5.2 7 g h

Monitored b

}b}g }

Ea'rYable E l h a, i l le 3 b . ] l i

                                                                                                                                 .l 3                          1 J b },

Leaka ge h J j $ p l g s

                                                                         %                                jiP                                          13             l, ] 1                          ]

l li). , 3 l i n i i nv i ts i I X X X X Recite Pwnp Mon Casing I X X X X X RCIC Swamline RCic Wawr n RHR Water O x x Q I X X X X X X O X X I X X X X X X 0 X X X X X @ l X X X X X X 0 X X Recirc Panp Motor Casing O Reactor Wasel He=1 Seal O Vahe Stem Packing O X

MisceDanseus LAak3 0 X l = Inside DrywellL4akage O m Outside Dryweu Leakage h Reactor coolant leakage in cooling waier to RHR Ha, RIP Ha, CUW Non regen Ra's or to FP cooling Ha. Amendment 15 5.2 36 _ _ , . _ _ _ _ _ . _ . . . _ , . . _ . - - - . - . . . - - . _ . - . . . , . - - - - - - . . ~ . _ ,

B %'R 2xuaora C^ meandard Plant t 100 I SAFETY VALVE d l OPENING g CHARACTERISTICS B ! e I [ 60 - g

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                                    >                                      I O                           !.

g VALVE I i STROKE

                                                       . -     TIME     5
                                                         -C     0.3     3
                                                       '1 TIME bec) tg = TIME AT WHICH PRESSURE EXCEEDS THE VALVE SET PRESSURE 87 20345 Figure 5.21 SAFETY ACTION VALVE LIFT CHARACTERISTICS O

5.2 37

                     'l'             1 NEUTRON Fl.UX                                                             1 VESSEL PRESSURE RISE (pul g
 ;        150 2 PEAK FUEL CENTER TEMPERATURE 3 AVERAGE SURFACE HEAT FLUX                       300 2 STEAM LINE PRESSURE RISE (pul 3 TURBIME PRESSURE RISE (ps.)

gg y i 4 FEEDWATER FLOW 4 REllEF VALVE FLOW (pcti s 5 5 VESSEL STEAM FLOW 5 SAFETY VALVE FLOW (gt) 12. 3 6 TURBINE GTEAM FLOW (pct) N w { 2 tr 100 T 200 N u 4 E

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0 4 8 12 16 0 4 8 12 16 i TIME (sec) TIME (sec) 1 VOID REACTIVITY 1 LEVEL (INCH.REF-SEP SKIRTl 1 2 DOPPLER REACTIVITY 2" 2 W R SENSED LEVEL (en.) ; -b 3 SCHAM REACTIVITY 3 N R SENSED LEVEL (en.1 4 TOTAL REACTIVITY _ 4 CORE INLET FLOW (pct) Q

      ,                              5 PUMP F LOW 3 (pct)
                                                                                                                \       4 100                                                                             0
                                                                                                         >      3                          f o               5                                                              2                                        2 g          2.3                                                                $                                                      f 1

3__ 3 3 m 0 ' _ S\_ 5 w l.2 4 5 - l --100 -2 .... .... ,_ l 0 4 8 12 16 0 1 2 3 4 l TIME (sect TIME (sect ,, s. 8770306 ,{ is Figure 52-2 MSIV CLOSURE WITH FLUX SCRAM AND $$ INSTALLED SAFETY / RELIEF VALVE CAPACITY O O O

I MN 23A6100AD mndard Plant an,. c l O l i. l l O l 1 9 i i Figure 5.2-5 (Deleted) Amendment 13 5.241 s

 - ,   -,7--e           _.. ..%,,, ,... . , . ,                   .,    ,4 e v. , , , , . . - - . . - , . . -       . , - . . . , , , - . - - - , .

i l ABWR zuaman I Standard Plant arv c l O I l 1 i J l I l

t I Figure 5.2-6 (Deleted) j ' $.242 Amendment 15

l ! ABWR m aman : Standard Plant nrv c e

                                                                                     ^
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                  /                                                                                                                \

Y m/ WELD SEAM , E 8 N2 : MAIN STEA0 NS : SD OUTLET N7 : LPFL AND SD OUTLET I N3 : FEEDWATER N6 : HPCF AND SLC INLET NB : SD AND CVW OUTLET I Figure 5.2-7a RPV EXAMINATION AREAS j l 5.2 s) Amendnient 1) l E _ . . _ . __.__..___.._.i -

1 ABWR mma RrY C

    $1andard Plant 5 3 REACTOR VESSEL                                   these vessel components is in accordance with q                                                       procedures qualified per ASME Section lit and Q  5.3.1 Reactor Yessel Malerlais                       lx requirements. weld test samples are requir-ed for each procedure for major vessel full-5.3.1.1 Materials Specifications                     penetratloc welds. Tensile and impact tests are perfornied to determine the properties of The. materials used in the reactor pressure      the base me tal, heat affected zone, and weld vessel and appurtenances are shown in Table 5.2 4 metal.

together with the applicable specifications. Submerjed are and manual stick electrode The RPV materials shall comply with the provisions of the ASME Code Section 111, Appendix weldingis pnotocesses welding are applied for employed. structural welds.Electroslag l 1 and meet the specification requirements of Preheat and laterpass temperatures employed for 10CFR$0, Appendix G. welding of low alloy steel meet or exceed the values given in ASME, Section 111, Appendix D. 53.1.2 Special Procedures Used for Manufactur. Post weld heat treatment at 5930C minimum is Ing and Fabrication applied to all low alloy steel welds, l ! The reactor pressure vessel is primarily con- Radiographic examination is performed on all structed from low alloy, high. strength steel plate pressure containing welds in accordance with re-and forgings. Plates are ordered to ASME SA 533, quirements of ASME, Section 111, Subsection NB TYPE D, Class 1, and forgings to ASMt1 SA 508, 5320. In addition, all welds are given a Class 3. These materials are melted to fine grain supplemental ultrasonic examination. practice and are supplied in the quenched and tem. pered condition. Further restr!:tions include a The materials, fabrication procedures, and requirement for vacuum degassing to lower the hy. testing methods used in the construction of BWR O d drogen level and improve the cleanliness of the low alloy steels Materials used in the core reactor pressure vessels meet or exceed require. ments of ASME Section Ill, Class 1 vessels. beltline region als) spe.ify limits of 0.05% max-imum copper and 0.C!$% n.aximum phosphorous con- 5.3.1.3 Special Methods for Nondestructive tent in the base materials and a 0.08% maximum Examination copper and 0.020% maximum phosphorous content in weld materials. The materials and welds on the reactor pres-sure vessel are examined in accordance with Studs, nuts, and washers for the main closure methods prescribed and meet the acceptance re-

f. age are ordered to ASME SA 540, Grade B23 or quirements specified by ASME, Section 111. In G'.ade B24. Welding electrodes for low alloy steel addition, the pressure retaining welds are ul-are low hydrogen type ordered to ASME SFA 5.5. trasonically examined using manual techniques.

The ultrasonic examination method, including All plate, forgings, and bolting are 100% ultra- calibration, instrumentation, scanning sensitiv-sonically tested and surface examined by magnetic ity, and coverage, is based on the requirements particle methods or liquid penettant methods in ac. Imposed by ASME, Section XI, Appendix 1. Accep-cordance with ASMl! Section lit, Division 1. tance standards are equivalent or more restrie-tive than required by ASME, Section XI. Fracture toughness properties are also measured and controlled in accordance with Division 1. 5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels All fabrication of the reactor pressure vessel is performed in accordance with GE approved draw- 5.3.1.4.1 Regulatory Guide 1.31: Control of ings, f abrication procedures, and test proce- Stainless Steel Welding dures. The shells and vessel heads are made from Controls on stainless steel welding are dis-p formed plates or forgings, and the flanges and nozzles from forgings. Welding performed to join cussed in Subsection 5.2.3.4.2.1. Amendment 1$ 5M l

ABWR m un RFV C Statidard Plant 5.3.1.4.2 Regulatory Guide 134: Control of of low. hydrogen electrodes to prevent hydrogen Electrostag Weld Properties cracking (provided in Subsection 5.2.3.3.4) does not explicitly meet this requirements, the ABWR y control will assure that cracking of components g l See Subsection 5.233.2.2. made from low. alloy steels does not occur during ' fabrication. Further, the ABWR control rninimi. 5.3.1.43 Regulatory guide 1.43: Control of zes the possibility of subsequent cracking rc. Stainless Steel Weld Cladding of Im. Alloy Steel sulting from hydtogen being retained in the Components weldment. Reactor pressure vessel specifications requite All welds are nondestructively examined by that all low alloy steel be produced to fine radiographic methods. In addition, a supplemen. grain practice. The requirements of this regula. tal ultrasonic examination is performed.

tory guide are not applicable to BWR vessels.

53.1.4.6 Regulaton Guide 1.71: Welder 5.3.1.4.4 Regulatory Guide 1.44: Control of Qualification for Areas of Limited Accessibility the Use of Sensitized Stainless Steel Qualification for areas of limited accessibil. Sensitization of stainless steel is controlled ity is discussed in Subsection 5.2.3.4.2.3. by the use of service proven materials and by use of appropriate design and processing steps in. 53.1.4.7 Regulaton Gulde l.99: Effects of 0 cluding solution heat treatment, corrosion resis. Residual Elements on Predicted Radiation Damage A tant cladding, control of welding heat input, to Reactor Pressure Vessel Materials ( control of heat treatment during fabrication and control of stresses. Predictions for changes in transition tem. perature and upper shelf energy are made in accor. I 5.3.1.4.5 Regulatory Guide 1.50: Control of dance with the requirements of Regulatory Guide Prvheat Temperature For Welding im. Alloy Steel 1.99. Regulatory Guide 1.50 delineates preheat tem. 5.3.1.4.8 Regulatory Guide 1.37: Quality perature control requirements and welding proce. Assurance Requirements for Cleaning of Fluid dure qualifications supplementing those in ASME Systems and Associated Components of Water. Sections ill and IX. Cooled Nuclear Power Plants The use of low. alloy steel is restricted to The cleaning of systems and components on the i the reactor pressure vessel. Other ferritic com. site during and at the completion of construction ponents in the reactor coolant. pressure boundary is accomplished to written procedures which assure are fabricated from carbon steel materials. both cleanliness and that the components are not O exposed to materials or practices which will A Preheat temperature employed for welding of degrade their performance. For components low alloy steel meet or exceed the containing stainless steel the procedures will recommendations of ASME Code Section lit, comply with Regulatory Guide 137. The procedures Appendix D. Components are either held for an will prohibit contact with low melting point extended time at preheat temperature to assure compounds, substances which are known to cause removal of hydrogen, or preheat is maintained stress corrosion cracking or which can release in until post. weld heat treatment. The minimum any manner substances that can cause such preheat and maximum interpass temperatures are problems. In addition there are controls placed specified and monitored. on the use of grinding wheels and wire brushes that assure that they cannot introduce degrading Acceptance Criterion ll.3.b(1)(a) of SRP materials either through prior usage or through Section 5.2.3 for control of preheat temperature their materials of construction. In this context requires that minimum and maximum interpass degradation includes stress corrosion cracking. temperature be specified. While the ABWR control Controls also controlintroduction of unnecessary Amendment 15 5.32

c ABM awiooxi, nrv c Standard Plant dirt and require control of dirt producing In regard to 10CFR$0, Appendix H, the l' c surveillance test material is selected processes such as welding or grinding including (}x prompt cleaning, on the basis of the requirements of ASTM

'                                                                  E185 82 and Regulatory Guide 5.99 to
          $3.1.5 Fracture Toughness                                provide a conservative adjus'         (Jfer.

ence temperature for u ,1tli n e 5.3.1.5.1 Compilance with 10CFR50, Appendix G materials. The weld test p: t for the surveillance program specimens has the Appendix G of 10CFR50 is interpreted for principal working direction parallel to Class 1 primary coolant pressure boundary compo- the weId seam to assure that nent of the ABWR reactor design and complied with heat affected zone specimens are trans-as discussed in Subsections 5.3.1.5.2 and 5.3.2. verse to the principal working direc. The specific temperature limits operation of the tion. See Subsection 5.3.4.2 for reactor when the core is critical are based on materials and surveillance capsule 10CFR$0, Appendix G Paragraph IV, A.3. See interface require ments. Subsection 5.3.4.1 for fracture toughness data interface requirements. (3) Records and Procedures for Impact Testing (G lil.C)

          $3.1J.2 Methods of Compliance The following items are the interpretations and methods used to comply with 10CFR50, Appendix G.

(1) Material Test Coupons and Test Specimens l (Gill A)

~'\

(V Test coupons are removed from the loca-tion in each product form as specified in subarticle NB 2220 of the ASME Code, Section Ill. The heat treatment of the test coupons is performed in accordance with subarticle NB 2210. It is understood that separately pro. duced test coupons per Subparagraph NB 2223.3 inay be used for forgings. (2) Location and Circulation of Test Specimens (G lli A) The test specimens are located and ori. l ented per ASME, Section 111, Paragraph NB 2322. Transverse Charpy V impact specimens are used for the testing of plate and forged material other than bolting and bars. Longitudinal specimens are used for bolting and bars. Both longitudinal and transverse i specimens are used to determine the re-quired minimum upper shelf energy level

  ,                  of the core belt line materials.

53 2.1 Amendmem 15

_ ABWR 2mwam ut c m _ andard Plant !h Preparation of impact testing procc- safety to those required for shells and dures, calibration of test equipment, heads are demonstrated using a l/4 T postu-O- and the retention of the records of lated defect at all locations, with the ex-these functions and test data comply ception of the main closuis flange to the with the requirements of the ASME Code, head and shell discontinuity locations. Ad. Section Ill. Personnel conducting ditional instruction on operating limits is impatt testing are qualified by experi. r6 quired for outside surface flaw stres enca training or qualification testing greater than 6.0 mm at the outside surface l that demonstrates competence to perform of the flange to shell joint based on tests in accordance with the testing pro- analysis made for ABWR reactor vessels using cedure,

the It will be of these areas, that smaller defects can be calculations demonstrated, using methods a test mockupshown in WRCB i (4) Charpy V Curves for the RPV Beltline L (G Illa and G.IVA 1) detected by the ultrasonic inservice examina.

tions procedures required at the adjacent A full transverse Charpy V curve is de. weld joint. termined for all beats of base material and weld metal used in the core beltline region with a minimum of three (3) specimen tested at the actual TNDT-L Tbc minimum upper shelf energy level for base material and weld metal in the l beltline region is 10.4 kg im as re. quired by G IVA.1. (7) Fracture Toughness Margins in the Control of Reactivity (Appendix G.IV A). In regard to 0111 A,it is understood that separate, unitradiated baseline ASME Code, Section 111, Appendix G, was used specimens per ASTM E 185, Paragraph in determining pressure / temperature limita. 6.3.1 will be used to determine the tran- tions for all phases of plan operation, sition temperature curve of the core beltline base material, HAZ and weld $3.1.6 MaterialSurveillance metal. 83.1.6.1 Compilance with Reactor Vessel (5) Bolting Material Material Surveillance Program Requirements All bolting snaterial exceeding one inch The materials surveillance program monitors diameter has a minimum of 6.4 kg im changes in the fracture toughness properties of charpy V energy and 0.64 mm lateral ex- ferritic materials in the reactor vessel pansion at the minimum bolt preload tem- beltline region resulting from exposure to perature of 210C, neutron irradiation and thermal environment. l (6) Alternative Procedures for the Calcula- Reactor vessel materials surveillance

  -                                                           tion of Stress Intensity Factor (Appen-          specimens are provided in accordance with re-dix G IV A)                                      quirements of ASTM E 185 and 10CRF $0, Appendix

11. Materials for the program are selected to rep-Stress intensity factors are calculated resent materials used in the reactor beltline by the methods of ASME, Section III, region. Specimens are manufactured from a plate Appendix G. Discontinuity regions are or forging actually used in the beltline region evaluated as shell and head areas, as and a weld typical of those in the beltline part of the detailed thermal and stress region and thus represent base metal, weld mate-analyses in the vessel stress report, rial, and the weld heat affected zone material.

r Considerations are given to membrane and The plate and weld are heat treated in a manner bending stresses, as outlined in Para- which simulates the actual heat treatment per-graph G 2222. Equivalent margins of formed on the core region shell plates of the 1 Amendment 1$ $M

ABM n46iown RFV C Standard Plant U completed vessel. Each in. reactor surveillance temperature at end of life is less than 34 C, capsule contains 36 Charpy V. notch and 6 tensile and the end.of life upper shelf energy exceeds 69 specimens. The capsule loading consists of 12 kg.m. (See response to Question 251.5 for the y Charpy V Specirnens each of base metal, weld metal, calculation and analysis associ. ated with this g heat affected zone material, and 3 tensile estimate). I specimens each from base metal and weld metal. A set of out of reactor basellne Charpy V. notch 5.3.1.6.4 Positioning of Surveillance Capsules specimens, tensile specimens, and archive material and Methods of Attachment (Appendix H.ll B (2)) are provided with the surveillance test specimens. Neutron dosimeters and temperature Surveillance specimen capsules are located at monitors will be located within the capsules as re. three azimuths at a common elevation in the core quired by ASTM E 185. beltline region. The scaled capsules are not at. tached to the vessel but are in welded capsule Three capsule are provided in accordance with holders. The capsule holders are mechanically re. requirements of 10CFR50, Appendix H,since the pre tained by capsule holder brackets welded to the dicted end of the adjusted reference temperature vessel cladding. Since reactor vessel spe. cifi. of the reactor vessel steelis less than 38 C. cations require that all low. alloy steel pressure l vessel boundary materials be produced to The following proposed withdrawal schedule is fine grain practice, underclad cracking is of no in accordance with ASTM E 185. concern. The capsule holder brackets allow the removal and reinsertion of capsule holders. Al-First capsule: After 6 effective full power though not code parts, these brackets are de. years signed, fabricated, and analyzed to the require. Second capsule: After 15 effective full power ments of ASME Code Section 111. A positive years spring. loaded locking device is provided to re. Third capsule: Schedule determined -sed on tain the capsules in position throughout any an-g results of first two capsules per ASTM E 185, ticipated event during the lifetime of the Paragraph 7.6.2. vessel. See Subsection 5.3.4.2 for interface requirements pertaining to materials and Fracture toughness testing of irradiated cap. survellance capsules, sule specimens will be in accordance with require. ments of ASTM E 185 as called out for by 10CFR50, in areas where brackets (such as the survell. Appendix H. lance specimen holder brackets) are located, addi. tional mondestructive examinations are performed 8.3.1.6.2 Neutron Flux and Fluence Calculations on the vessel base metal and stainless steel weld deposited cladding or weld. buildup pads A description of the methods of analysis is during vessel manufacture. The base metal is ul-contained in Subsections 4.1.4.5 and 4.3.2.8. trasonically examined by straight beam techniques to a depth at least equal to the thickness of the

      $.3.1.6.3 Predicted irradiation Effects on            bracket being joined. The area examined is the Beltline Materials                                    area of width equal to at least half the thick-ness of the part joined. The required stainless Transition temperature changes and changes in steel weld deposited cladding is simi larly exam-upper. shelf energy shall be calculated in accor. Ined. The full penetration welds are liquid. pen.

dance with the rules of Regulatory Guide 1.99 etrant examined. Cladding thickness is required Reference temperatures shall be established in ac to be at least 3.2 mm. These requirements have cordance with 10CFR50, Appendix G, and NB.2330 of been successfully applied to a variety of bracket the ASME Code, designs which are attached to weld deposited stainless steel cladding or weld buildups in many l l Since weld material chemistry and fracture operating BWR reactor pressure vessels, toughness data are not available at this time, the limits in the purchase specification were Inservice inspection examinations of core used to estimate worst case irradiation effects, beltline pressure.rctaining welds are performed l These estimates show that the adjusted reference l l Amendment 15 $.3 4 i l

ABWR wmn Standard Plant nrv c vessel. If a bracket for mechanically retalning Code, Section lit, Class I, requirements. The (Vo) surveillance. specimen capsule holders were reactor pressure vessel clo,ure studs are SA.540 located at or adjacent to a vessel shell weld,it Grade B23 or 24 (AISI 4340). The maximum allow. would not interfere with the straight. beam or able ultimate tensile strength is 119.5 half. node, angle. beam inservice inspection ultra- kg/mm2 . Also, the Charpy impact test re. sonic examinations performed from the outside quirements of NB.2333 will be satisfied (the surface of the vessel. lowest Cy energy will be greater than the re. quirement of 6.4 kg.m at 21 OC; tbc lowest rc.

      $.3.1.6.$ Time and Number of Dostmetry Measure.       ported Cy expansion will exceed the 0.64 mm re.

ments quired). GE provides a separate neutron dosimeter so in relationship to regulatory position C.2.b, that fluence measurements may be made at the the bolting materials are ultrasonically exam. vessel ID during the first fuel cycle to verify ined in accordance with ASME Code Section 111, the predicted fluence at an early date in plant NB.2580, after final heat treatment and prior to operation. This measurement is made over this threading as specified. The requirements for ex, short period to avoid saturation of the dosim. amination according to ASME Code, Section 11, eters now available. Once the fluence to thermal SA488 and ASTM A614 were complied with. The power output is verified, no further dosimetry is procedures approved for use in practice are considered necessary because of the linear rela. Judged to insure comparable material quality and tionship between fluence and power output. It are considered adequate on the basis of compli. will be possible, however, to install a new dosim. ance with the applicable requirements of ASME eter,if required, during succeeding fuel cycles. Code, Subsubarticle NB.2580. 5.3.1.7 Reactor Vessel Fasteners The straight. beam examination is performed on 100 percent of cylindrical surfaces and from

/]

(, The reactor vessel closure head (flange) is both ends of each stud using a 19 mm maximurr. fastened to the reactor vessel shell flange by diameter transducer. The reference standard for multiple sets of threaded studs and nuts. The the radial scan contains a 12.7 mm diameter lower end of each stud is installed in a threaded flat bottom hole with a depth of 10 percent of hole in the vessel shell flange. A nut and Se thickness. The end scan standard is per washer are installed on the upper end of eac$. ASTM A614. Surface examinations are performed stud. The proper amount of preload c% be on the studs and nuts after final heat treatment applied to the studs by sequential tensioning and threaded as specified in the guide, in accor. using hydraulic tensioners, dance with ASTM A614. Any indication greater than the indication from the applicable calibra. liardness tests are performed on all main tion feature is unacceptable. The distanec/ampli. closure bolting to demonstrate that heat treat. tude correction turve for the straight beam end ment has been properly performed, scan of main closure studs, nuts, and washers are established as follows: 5.3.1J Regulatory Guide 1.65 For cylinder having a length (L) to 0.D. Regulatory Guide 1.65 defines acceptable mate. ratio of 7 of less, the distance / amplitude rials and testing procedures with regard to curve is established by a minimum of three reactor vessel closure stud bolting for test points along the test distance. For cyl. l light water cooled reactors, inders having length to O. D. ratios larger than 7, the minimum number of test points is The design and analysis of reactor vessel four. The test points are nearly equally l bolting materials is in full compliance with ASME spaced along the test distance. One calibra. l l tion hole is located at a test distance equal to L/2. 3 (O

                                                                                                        $.3 $

Amendment 15

ABWR memo Standard Plant RN C 5.3.2 Pressure / Temperature Limits non. nuclear bestup and cooldown following a nuclear shutdown. 5.3.2.1 Umit Curves Reactor Operation The pressure /teroperature limit curves in Figure 5.31 are based on the requirements of Curve C in Figure 5.31 specifies limits ap. 10CFR50, Appendix G. The pressure / temperature plicable for operation whenever the core is l limits look different than SRP Section 5.3.2 critical except for low level physics tests. because the ABWR temperature limits are based on a more recent revision of Regulatory Guide 1.99. 5J.2.1.4 Reactor Vessel Annealing All the vessel shell and head areas remote from inplace anneallag of the reactor vessel, discontinuities plus the feedwater nozzles were because of radiation embrittlement,is not an-evaluated, and the operating limit curves are ticipated to be necessary. based on the limiting location. The boltup limits for the flange and adjacent shell region are based $J.2.1J Predicted Shiftin RT and on,a minimum metal temperature of RT DT plus Drop in Upper Shelf Energy (hndix G.W B) 33 C. The maximum throughwall temperafure gradi-ent from continuous heating or cooling at 55.5 C For design purposes the adjusted reference per hour was considered. The safety factors nil ductility temperature and drop in the applied were as specified in ASME Code, Appendix upper shelf energy for BWR vessels is predicted G, and Reference 2. using the procedures in Regulatory Guide 1.99. The material for the vessel will be 5 rovided The calculations (see response to Question

     ,., with the following requirements of RT             as  251.5)   are based on the specified limits on determined in accordance with Branch 'NcSTnical Phosphorous (0.020%), Vanadium (0.05%), Copper             e-Position MTEB 5 2: shell and flanges 29 C; (0.08%) and Nickel (1.2%) in the weld material.                 5 l%      nozzles 29"C and welds 29 C.                          In plate material, the limits are Copper (0.05%)

and Nickel (0.73%). Forgings will have the same 53.2.1.1 Temperature Umits for Boltup chemistry as plate but the nickel limit is 1%. i l Minimum closure flange and fastener tem. The ABWR neutron fluences are low when l peratures of RT ND plus 33 C are required for compared with the past reactors because of the

   ' e-   tensioning at prcIoad condition and during fact that the incorporation of internal pumps 6   detensioning. Thus, the minimum limit b 29"C increased the annulus between the shourd and the l       + + 33 C = + 4 C.                                     vessel wall.

53.2.1.2 Temperaturv umits for iSl Hydro. A surveillance program in accordance with static and I2ak Pressurv Tests ASTM E 185 will be used. The surveillance ,, program willinclude samples of base metal, weld g Pressure (measured in the top head) versus tem. metal and heat affected zone material, perature (minimum vessel shell and head mett/. Rm- Subsection 5.3.1.6 provides added detail on the perature) limits to be observed for the test and surveillance program. operating conditions are specified in Figure 5.3 1. Temperature limits for preservice and $3 2.2 Operating Procedures inservice tests are shown in Curve A of Figure 5.31. A comparison of the pressure versus tem-perature limit in Subsection 5.3.2.1 with in. 5.3.2.1.3 Operating Umits During Heatup, tended normal operation procedures of the most Cooldon, and Corereration severe service level B transient shows that l those limits will not be exceeded during any l fleatup and Cooldown. foretecable upset condition. Reactor operating procedures have been established so that actual Curve B in Figure 5.31 specificc limits for transients will not be more severe than those ' Amendment 13 $S6 l

ABM 2mioaru arv c I Standard Plant j for which the vessel design adequacy has been dem. i onstrated. Of the design transients, the service ! (V-) l level B condition producing the most adverse tem. perature and pressure condition an3where in the vessel head and/ or she!! ascas yields a minimum i fluid temperature of 27ti oC and a maximurn peak j pressure of 85.4 hg/cm2 g. Scram auto. 1 matically occurs as a r sult of this event prior I to a possible reduction in fluid temperature to 1210C at a pratsure of 65.4 kg/cm g. Per 2 Figure 5.31, both the 85.4 kg/cm 2g vessej ] pressure at 276 OC (Curve C) and the 65.4 kg/cm2 g at 121 OC (Curve B) are within the calculated margin against nonductile failure. 5.3.3 ReactorVesselintegrity The reactor vessel material, equipment, and services associated with the reactor vessels and appurtenances would conform to the requirements of the subject purchase documents. Measures to ensure conformance included prosisjons for source evaluation and selection, objective evidence of ' quality furnished, inspection at the vendor source and examination of the completed reactor vessels. GE provides inspection surveillance of the G m, mons 3xi

I ABWR - mmo ; Enandard Plant uv : reactor vessel fabricator improcess manufactur- load, dead weight, and residual stresses; n i ( Ing, fabrication, and testing operations in accor-dance with the GE quality assurance program and (3) the radiation effect on material tough. l approved inspection procedures. The reactor ness (RT N shift and critical stress in. ; vessel fabricator is responsible for the first tensity); ankT level inspection of manufacturing, fabrication, and testing activities and GE is responsible for (4) methods for calculating crack tip stress the first level of audit and surveillance inspec- intensity associated with a monuniform tion, stress field following the design basis acci. ; dent. i Adequate documentary evidence that the reactor vessel material, manufacture, testing, and inspec. This analysis incorporated very conservative tion conforms to the specified quality assurance assumptions in all areas (particularly in the requirements contained in the procurement specifi- areas of heat transfer, stress analysis, effects cation is available at the fabricator plant site, of radiation on material toughness, and crack tip stress latensity). Therefore, the results re-Regulatory Guide 1,2, Thermal Shock to ported (Reference 1) provide an upper bound ap. Reactor Pressure Pessr/s, states that potential proach, it is concluded that catastrophic , reactor pressure. vessel brittle fracture, which failure of the pressure vessel due to DBA is may result from emergency core cooling system op- shown to be impossible from a fracture mechanics l eration, need not be reviewed in individual cases point of view. In the case studies, even if an I if no significant changes in presently approved acute flaw does form on the vessel inner wall, core and pressure vessel designs are proposed. it will not propagate as the result of the DBA. l If the margin of safety against reactor i i pressure. vessel brittle fracture due to emergency The criteria of 10CFR50, Appendix G, are in. I

   ' cooling system operation is considered unaccept-      terpreted   as establishing the requirements of an.            l L

able, an engineering solution, such as annealing, nealing. Paragraph IV B requires the vessels to l l j '

 \     could be~ applied to assure adequate recovery of be designed for annealing of the beltline only the fracture toughness properties of the vessel where the gredicted value of adjusted RT material. RG 1.2 requires that engineering solu. exceeds 93 C, as defined in Paragraph lNBb tions be outlined and requires demonstration that   of ASME Code Section !!!. This predicted value the design does not preclude use of the solu-       is not exceeded, therefore, design for annealing tions,                                              is not required, An investigation of the structural integrity        For further discussion of fracture toughness               i of boiling water reactor pressure vessels during of the reactor pressure vessel, refer to Subsec-                  !

I a design basis accident (DBA) has been conducted tion 5.3.1.5. l (Reference 1), it has been determined, based on methods of fracture mechanics, that no failure of 53.3.1 Design the vessel by brittle fracture as a result of DBA will occur. 5.3.3.1.1 Description - l The investigation included: 53.3.1.1.1 Reactor Vessel (1) a comprehensive thermal analysis consider. The reactor vessel (Figure 5.3 2) is a verti. Ing the effect of blowdown and the low. cal, cylindrical pressure vessel of welded con. .! l pressure coolant injection system reflooding; struction. The vessel is designed, fabricated, tested, inspected, and stamped in accordance (2) a stress analysis considering the effects with ASMF. Code, Section 111, Class 1 require-of pressure, temperature, seismic load, jet ments including the addenda in effect at the i date of order placement (Table 3.2 4), j O 5.37 Amendment 15 l l l l ,

ABWR == Standard Plant nry e Design of the reactor vessel and its support by vertical stilt legs from the bottom b:sd. system muts Selsmic Category I equipment require. This support is designed to carry the weight of ments. The materials used in the reactor pres- peripheral fuel elements, neutron sources, core sure vessel are listed in Table 5.2 4. plate, top guide and the steam separators and to laterally support the fuel assemblics and the The cylindrical shell and top and bottom heads pump diffusers. Design of the shrvud support of the reactor vessel are fabricated of low alloy also secounts for pressure differentials across steel, the interior of which is clad with stain- tbc shroud support plate, for the restraining less steel weld overlay except for the top bead effect of components attacheo to the support, and nozzle weld zones. The bottom bead is clad and for earthquake loadings, The shroud suppori l with Ni Cr Fe alloy. design is specified to meet appropriate ASME Code stress limits. Inplace annealing of the reactor vessel is not necessary because shifts in transition tempera- 533.1.1J Protection of Closure Studs ture caused by irradiation during the 60 year life can be accommodated by raising the minimum Tbc BWRs do not use bora'ed water for reactiv-pressurization temperature, and the predicted ity control during normal operation. This sub. section is therefore not applicable. l exceed value 930C.of adjusted Radiationreference embrittlementtemperature is not a does not problem outside of the vessel beltline region 53J.1.2 Safety Design itesis because the irradiation in those areas is less than 1 X 1018 avt with neutron energies in The design of the reactor vessel and appurte-excess of 1 McV. The use of existing methods of nances meets the following safety design bases. I predicting embrittlement and operating limits which are based on a 40 year life are considered (1) The reactor vessel and appurtenances will to be applicable to a 60 year life because the withstand adverse combinations of S age degrading mechanism is irradiation and fati- loading and forces resulting from op. A gue duty which are calculated for the 60 year cration under abnormal and accident condi-life. Time / temperature effects will either not tions, have any effect or will produce a small beneficial co anticaling. (2) To minimize the possibility of brittle fracture of the nuclear system process Quality control methods used during the fabri- barrier, the following are required: l cation and assembly of the reactor vessel and ap-purtenances assure that desien specifications are (a) impact properties at temperatures rela-met, ted to vessel operation have been specified for materials used in the reactor vesse!; Tbc vessel top head is secured to the reactor vessel by studs and nuts. These nuts are tight- (b) expected shifts in transition tempera-ened with a stud tensioner. The vessel flanges ture during design life as a result of envi-are scaled with two concentric metal scal rings roomental conditions, such as neutron flux, designed to permit no detectable leakage through are considered in the design and operational the inner or outer seal at any operating condi- limitations assure that NDT temperature tion, including heating to operating pressure and shifts are accounted for in reactor opera-l temperature at a maximum, tate of 550C in any Lion; and one bour period. To detect seal failure, a vent tap is located between the two nal rings. A mo. (c) operational margins to be observed with nitor line is attached to the tap to provide an regard to the transition temperature are indication of leakage from the inner scal ring specified for each mode of operation, seal. 533.13 Power Generation Design Bases 533.1.1.2 Shroud Support The design of the reactor vessel and appurte-The shroud support is a circular plate welded nances meets the following power generation to the vessel wall and to a cylinder supported design bases: Amendment 13 334

ABM Standard Plant m aiooxa nry c (1) Tbc reactor vessel has been designed for reactor vessel bottom head and are welded to Inconel stub tubes. Each housing transmits O a usefullife of 60 years. loads through the stub tubes to the bottom head (2) External and internal supports that are of the reactor. These loads include the weights

 -                   integral parts of the reactor vessel are of a control rod, a control rod drive, a control located and designed so that stresses in rod guide tube, a four lobed fuel support piece, i                     the vessel and supports that result from and the four fuel assemblies that rest on the reactions at these supports are within fuel support piece. The housings are fabricated ASME Code limits,                          of Type 304 austenitic stainless steel.

(3) Design of the reactor vessel and appurte- $33.1.43 In Core Neutros Flux Monitor nances allows for a suitable program of Housings inspection and surveillance. [ Each in core neutron flux monitor housing is L

          $33.1.4 Reactor Vessel Design Data                    inserted though the in core penetrations la the bottom head and welded to inconel stub tubes.

The reactor vessel design pressure is 87.9 kg

           /cm 2g and the design temperature is 3020C.               An in core flux monitor guide tube is welded The maximum installed test pressure is 109.9 kg to the top of each housing and a startup range
           /cm 2g.                                               neutron monitor (SRNM) or a local power. range monitor (LPRM) is bolted to the scal /rlog flange 533.1.4.1 Vessel Support                              J. 2he bottom of the housing outside the vessel y        '7.6).

The vessel support is constructed as an inte. gral part of the RPV pedestal. Steel anchor $33.1.4.* ; . tor Vessel lasulation bolts set in the conciete extend through the bearing plate and secure the flange of the The reactor pressure vessel insulation is re-O reactor vessel support skirt to the bearing pte and thus to the support pedestal. The design is in accordance with ASME Code Section Ill, Divi-flecthe metal type, constructed entirely of series 300 stainless steel and designed for a 60 year life. The insulation is made of prefabri-slon 1, NF The connection is a friction type cated units enginected to fit together and main-joint where the bolts are pretensioned to the tain the insulation effielency during tem-extent necessary to ensure that there will be no perature changes. The insulation is designed to relative movement between the RPV and its pedes- remain in place and resist damage during a safe tal. Shear forces are resisted by friction shutdown earthquake. Each unit is designed to between the skirt flangeplate and the pedestal permit free drainage of any moisture that may mounting plate or shear between the flange and accumulate in the unit and prevent internal pres- ; mounting bolts, sure buildup due to trapped gases. Loading conditions arc as follows: The insulation for the reactor pressure vessel is supported from the biological shield A & B = Power Range + OBE wall surrounding the vessel and not from the , vessel shell, lasulation for the upper head and C = Power Range + OBE + Scram flange is supported by a steel frame independent of the vessel and piping. During refueling the D = Power Range + SSE + Pipe Break support frame alone the top head insulation is removed. The support frame is designed as a Seismic Category I structure, insulation access

              $33.1.4.2 Control Rod Drive Housings                  panels and insulation around penetrations is de-signed in sections with quick release latches The control rod drive housing are inserted through the control rod drive penetrations in the 5.39 Amendmem 15

ABWR == Standard Plant krv.c which provide for case of installation a.id rerno. 5.3.3.2 Materials of Construction val for vessel inservice inspection and mainte. cance operation. Each insulation unit has lift. All material used in the construction of the ing fittings attached to facilitate removal. In. reactor pressure vessel conform to the require. sulation units attached to the shield wall are ments of ASME Code, Section 11 runterials. The not required to be readily removable except vessel heads, shells, flanges, and nozzles are around penetrations, fabricated from low alloy steel plate and forg. ings purchased in accordance with ASME Specifica. At operating conditions, the insulation on the tions SA.533 Type B, Class 1 and SA 508 Class shleid wall and around the refueling bellows has 3. Special requirements for the low. alloy steel an average maxirnum beat transfer rate of 176 plate employed on the interior surfaces of the kcal/m2h of outside insulation surface. The vessel consists of austenitic stainless steel maximum heat transfer rate for insulation on the weld overlay. The material in the beltline top head is 163 kcal/m2 h Operating condi. region is SA 508 Class 3 forged rings, tions are 2880C for the outside temperature of the reactor vessel and 570C for the drywell Tbest asterials of construition were selected air. The maximum air temperature is 660C, because they provide adequate strength, fracture except fer the head area above the bulkhesd and toughness, fabricability, and compatibility with refueling seal which has a maximum allowable tem. the BWR environment. Their suitability has been l perature of 930C. demonstrated by long term successful operating experience in reactor senice. 53.3.1.4.5 Reactor Vessel Nnules The npected peak neutron fluence at the 1/4 All piping connected to the reactor vessel t location used for evalution is less than 6 x , nozzles has been designed not to exceed the allow. 1017 avt for 60 years, the calculated shift able loads on any nozzle. The vessel top head in RTNDT is 15.50C for weld metal and A l nozzle is provided with flanges with small groove 4.40C for base metal and the drop in upper facings. The drain nozzle is of the full penetra. shelf energy is 1,38 kg.m for welds and 1.1 kg.m , tion weld design. The feedwater inlet nozzles, for base metal, core flooder inlet nozzles, and ECCS flooding nozzles have thermal sleeves. Nozzles connecting 5.333 Fabrication Methods to stainless steci piping have safe ends or exten. stons made of stainless steel. These safe ends The reactor pressure vessel is a vertical cy. or extensions were welded to the nozzles after lindrical pressure vessel of welded construction the pressure vessel was heat treated to avoid fabricated in accordance with ASME Code, Section furnace sensitiration of the stainless steel, 111, Class 1, requirements. All fabrication of The material used is compatible with the material the reactor pressure vessel was performed in ac. of the mating pipe, cordance with GE. approved drawings, fabrication procedures, and test procedures. The shell and

           $3.3.1.4.6 Materials and inspections                 vessel head were made from formed low. alloy steel plates or forgings and the flanges and The reactor vessel was designed and fabricated    nozzles from low. alloy steel forgings. Welding      l In accordance with the applicable ASME Boiler and    performed to join these vessel components was in Pressure Vessel Code as defined in Subsection accordance with procedures qualified to ASME, 5.2.1. Table 5.2 4 defines the materials and Section 111 and IX requirements, Weld test specifications. Subsection 5.3.1.6 defines the samples were required for each procedure for compliance with reactor vessel material surveil.      major vessel full. penetration welds.                 4 lance program requirements.

Submerged are and manual stick electrode 53.3.1.4.7 Reactor Vessel Schematie welding processes were employed. Electroslag welding was not applied. Preheat and interpass The reactor vessel schematic is shown in temperatures employed for welding of lomalloy Figure 5.3 2. steel met or exceeded tM requirements of ASME Section 111, Appendix D. Post weld heat treat-Amendment 15 5.3 M I i l I

ABM standard Plant imioorn nrv c J l ment of $93 C minimum was applied to all (2) if the coolant temperature difference be-low alloy steel welds, tween the dome (inferred from P (sat))

    -                                                                     and the bottom head drain exceeds 55 C,       l All previous BWR pressure vessels have                   neither reactor power level nor employed similar fr5;rication methods. These                   recirculation pump flow shall be vessels heve operated for an extensive number of               increased-years and their service history is rated excellent.                                                The limit regarding the normal rate of heatup and cooldown (item 1) assures that the vessel 533A laspection Requirementa                          closure, closure studs, vessel support sHrt, control rod drive bousing, and stub tube All plates, forgings, and bolting were 100% ul- stresses and usage remain within acceptable
 ,          trasonically tested and surface examined by mag- limits. Vessel temperature limit on recircu.

netic particle methods or liquid penetrant me- lating pump operation and power level increase thods in accordance with ASME Code, Section Ill, restriction (Item 2) augments the item 1 limit Welds on the reactor pressure vessel were exam- in further detail by assuring that the vessel ined in accordance with methods prescribed and bottom head region will not be warmed at ex. meet the acceptance requirements specified by cessive rate caused by rapid sweep out of cold l ASME Code, Section III. In addition, the pres. coolant in the vessel lower head .egion by sure retaining welds were ultrasonically examined recirculating pump operation or natural circula-using acceptance standards which are required by tion (cold coolant can accumulate as a result of ASME Code, Section XI. control drive inleakage and/or low recirculation flow rate during startup or hot standby). 53J 5 Shipment and lastallation These operational limits when maintained en. The completed reactor vesselis given a thor- sure that the stress limits within the reactor ough cleaning and examination prior to shipment, vessel and its components are wi: bin the thermal l ( The vessel is tightly scaled for shipment to pre- limits to which the vessel was designed for nor-vent entry of dirt or moisture. Preparations for mal operating conditions. To maintain the integ-shipment are in accordance with detailed written rity of the vessel in the event that these op-procedures. erational limits are exceeded, the reactor ves-sel has been designed to withstand a limited On arrival at the reactor site the reactor number of transients caused by operator error, vessel is examined for evidence of any contamina- Also, for abnormal operating conditions where tion as a result of damage to shipping covers, safety systems or controls provide an automatic Measures are taken during installation to assure temperature and pressure response in the reactor l that vessel integrity is maintained; for example, vessel, the reactor vessel integrity is main-access controls are applied to personnel entering tained since the severest anticipated transients the vessel, weather protection is provided, and have been included in the design conditions. periodic cleanings are performed. Therefore, it is concluded that the vessel integ-rity will be maintained during the most severe 533.6 Operating Conditions postulated transients since all such transients are evaluated in the design of the reactor Procedural controls on plant operation are vessel. Implemented to hold thermal stresses within ac. ceptable ranges and to meet the pressure /ternpe. 533.7 Inservice Suiveillance reture limits of Subsection 5.3.2. The restrictions on coolant temperature are as Inservice inspection of the reactor pressure follows: vessel will be in accordance with the require-ments of the ASME Boiler and Pressure Vessel (1) t, e average rate of change of reactor Code, Section XI. The vessel will be examined coolant temperature during normal heatup once prior to startup to satisfy the preopera. O and cooldown shall not exceed 55 C tional requirements of IWB 2000 of ASME Code, O during any one hour period; Section XI. Subsequent inservice inspection

                                                                                                               $.3-11 Arnendrnent 15

2 ABM 23461ocan - Standard Plant Bgy.J; = will be scheduled and performed in accordance a with the requirements of 10CFR50.55a, subpara. graph (g) as described in Subsection 5.2.4. .. The materials surveillance program will monitor changes in the fracture toughness proper-ties cf ferritic materials !n the reactor vessel g beltline region resulting from exposure to 1 neutron irradiation and thermal environment.

Specimens of actual reactor beltline material will be exposed in the reactor vessel and peri-odically withdrawn for impact testing. Operating procedures will be modified in accordance with

            . test results to assure adequate brittic. fracture                                --

control, Material surveillance' programs and inservice inspection programs are in accordance with appli. , l cable 10CFR50 Appendix H and ASME Code require-ments and provide assurance that brittle. fracture control and pressure vessel integrity will be maintained throughout the service lifetime of the reactor pressure vessel. 5.3.4 Interfaces n 5.3A.1 Fracture Toughness Data Fracture toughness data baseu on the limiting reactor vessel materials will be provided (See.

             . Subsection 5,3,1.5.1).

5.3A.2 Materials and Surveillance Capsule The following will be identified; the specific materials in each surveillance capsule; - the capsule lead factors; -the withdrawal schedule for each surveillance capsule;. the neutron fluence to be received by each capsule at

tW time'_ of its withdrawal; and, the vessel AQof life peak neutron fluence (See Subsection us.6.4).

              - 5.3.5 References.

1. An Analytical Study on Brittle Fracture of GL BWR VesselSubject to the Design Basis Ac-

                     - cident, (NEDO 10029).
                '2.   . Transient Pressure Rises Affecting Fracture Toughness Requirements for Boiling Water Reac-tors, January 197o, (NEDO 21778-A).

Amendm<nt 15 5.3 12 m ms - . . . .. . . . . .

 -_.-x-   .      _ _ . ,                _ . _ _         _ _ _ _ _ _ _ _ _ - . _             - . . . . _ .       . . - . . _      _ _ . . . _ _ _ _ . . _ . - _ . . . . _ _ _ . . .      _ - . _ . _ . . _ _ . . _ _
            'ABWR                                                                                                                                                                  n^om^n Standard Plant                                                                                                                                                           nov.c n                                                                                                                                                                                                             )

98.5 ' l A B C 84.4 _ A SYSTEM HYDROTEST LIMIT WITH FUEL IN VESSEL B NON. NUCLEAR HEATING

                -                                                                                                              LIMIT C NUCLEAR (CORE CRITICAL)

LIM!T

g. NOTE

LIMITS ARE BASED ON L e 10CFR50, APPENDIX G 56.3 _

, W B C ] 42.2 _ , u ' 1 T W ' 28.1 21.9 E x,,em a BOLTUP: 14.1 _ 0 1 l l l l 0 38 43 149 204 260 l h- MINIMUM REACTOR VESSEL METAL TEMPERATURE ('C) Figure 5.31 MINIMUM TEMPERATURE REQUIRED VERSUS REACTOR PRESSURE Amendment 15 5.3 13

1 ii 83A6100AD Standard Plant _ nov c- l N e.

                                                                                                 >EAD VENT AND SPRAY
                                                                                                                                             #            g CLOSURE HEAD                         ,                                                           ,

9 : REFUCLING SEAL SKIRT STEAM ORYERN 'l l b, B 9 i j 1-

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i SHROUD HEAD BOLT STEAM SEP ARATORQ_1 l

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CORE FLOODER T- SHROV" '<AD - Low PRESSURE

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SHUTOOWN l-C00VNG OUTLEi M HtCH PRESSURE

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                                                                                                                                   '                                   I RDV SUPPORT SKIRT                                                                                            l                       FUEL SUPDORT t                   .

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GU:0E TUBE , ! .1, SMROUD SUPPORT ' p RIP IMDELLER RPV ORAW LINE\

                                                                                                                !                                                                           6 RE ACTOR INTEPNAL
                                                                                                                " Y~ ll                                                                     'r-                     PUMP (R;P) WOTOR ll                                          \                                       CASING CR0 HOUSNG                         l                                                                    W-CORE ' HOUSWG
                                                                                                                      - Bi ii                         i ls                         'i e,                     ;;                            ;                          CR0 RESTRAINT BE AM Figure 5.3-2 REACTOR KEY FEATURES                                                                                                                                                g Amendment 15                                                                                                                                                                                      5.3 14 I

1

f

                'ABWR-                                                               2mma .

Standard Plant RTN. C ce SECTION 5.4 y}-

    \                                                CONTENTS                                       1 Section                         Ihlt                 E.agt 5.4.1.      Reactor Recirculation System             5.410.a 5.4.1.1     System Design Bases                      5.4-1 5.4.1.2     Power Generation Design Bases            5.41                  i 5.4.13      Description                              5.4-1 5.4.13.1    Recirculation Motce Cooling Subsystem    5.4-2 5.4.13.2    Recirculation Motcr Purge Subsystem      5.4-2 5.4.133     Recirculation Motor Inflatable Shaft -                           l Subsystem                                5.43                   !

5.4.1.4- Operation 5.44 5.4.1.5 Safety Evaluation - 5.4-4

      . . . -            5.4.1.6 '   Inspection and Testing                   5.45

[j Q 5.45 1, Q- 5.4.2 Steam Generators (PWR)

                                                                                                  ,1
                                                                                                     ]

l' 5.45 5.43 Eeactor Coolant Pininn l Main Steamilne Flow Restrictors 5.4-5 l 5.4.4 5.4.4.1 Safety Design Bases 5.4-5 l 5.4.4.2 - Description 5.45 5.4.43 Safety Evaluation - 5.4-6 l 5.4.4.4 . Inspection and Testing 5.46 I 5.4.5 Main Steamline Isolation System :i,4-6 J 5.4.5.1. Safety Design Bases 5.4-6 15,4.5.2 Description 5.4-7. 5.4.53 . Safety Evaluation 5.48 5.4.5.4 Inspection and Testing 5.4-9 5.4.6 Reactor Core Isolation Coollne fT l Q System 5.4 il 5.4-10 i l . Amendment 15 l l

AB M ux6ioorn Standard Plant REV A SECTION 5.4 p CONTENTS (Continued) $-- !

Sect lon Tidg Eagg 5.4.6.1_- Design Basis 5.4 10 5.4.6.1.1 Residual Heat and Isolation 5.4 11 5.4.6.1.1.1 Residual Heat 5.4 11 5.4.6.1.1.2 1 solation 5.4-11 5.4.6.1.2 Reliability, Operability, and Manual Operation 5.4-12 5.4.6.1.2.1 Reliability and Operability 5.4-12 I 5.4.6.1.2.2 Manual Operation 5.4 12 I S.4.6.13 las of Offsite Power 5.4-12 i

            ' 5.4.6.1.4 '    Physical Damage                        5.4-12                   j 5.4.6.1.5       Environment                           5.412 5.4.6.2         System Design                          5.4-12 5.4.6.2.1       General                                5.4-12                     4 5.4.6.2.1.1    Description                            5.4.13 5.4.6.2.1.2    Diagrams                               5.413 i

5.4.6.2.13 -Interiocks '5.413 5.4.6.2.2 Equipment and Component Description 5.4 13 0

 $           '5.4.6.2.2.1     Design Conditions                      5.4-13 5.4.6.2.2.2    Design Parameters                      5.4-14 5.4.6.2.3      Applicable Codes and Classifications    5.4-14 5.4.6.2.4       System Reliability Considerations      5.4-14 5.4.6.2.5      Eystem Operation                       5.415 9

5.4-lii

c ABWR ursioorn Standard Plant arv.c n

  ~v SECTION 5.4                              '

CONTENTS (Continued) Section M P.agt 5.4.9.2 Power Generation Design Bases 5.4-27 5.4.93 Description 5.4 28 5.4.9.4 Safety Evaluation 5.4-28 i 5.4.9.5 Inspection and Testing 5.4 28 i 5.4.10 Pressuriatt 5.4-28 l 5.4.11 Pressurtrer Rellef Discharne System 5.4 28 5.4.12 Valves 5.4 28 L' l-5.4.12.1 Safety Design Bases - 5.4-28 5.4.12.2 Description 5.4 29 5.4.123 - Safety Evaluation $.4 29 N 5.4.12.4_ Inspection and Testing 5.4 29 5.4.13 Safetv/ Relief Valves 5.429 5.4.13.1 Safety Design Bases 5,4-29 5.4.13.2 Description 5.4-30 , 5.4.13 3 - Safety Evaluation 5.4 30

              . _5.4.13.4       Inspection and Testing.                  5.4 30

.o, 5.4.14 Comnonent Sunnorts 5.4-30 3 5.4.14.1 Safety Design Bases 5.4 30 - i 5.4.14.2 ' Description 5.4-30 5.4.14 3 Safety Evaluation 5.4 30 5.4.14.4 (Deleted) 5.430' 5.4.15 References 5.4-30 5.4.vi Amendment 15

                    - ABWR                                                                       m am^s-Standard Plant                                                                 REV.C SECTION 5.4 TABLES
                                                                                                          -g Table                         Ikle                     East 5.41     Reactor Recirculation System Design           5.4-32 Characteristics 5.42      Design Parameters for RCIC System Components 5.4-33 5.4-3     RHR Pump / Valve Logic                       5.438 5.4-4     RHR Heat Exchanger Design and                5.4-40 Performance Data 5.45-     Component and Subsystem Relief Valves        5.4 41 5.4-6     Reactor Water Cleanup System Equipment Design Data                                   5.4-43 ILLUSTRATIONS
        ,                            Figure                        IRic                     East 5.41     Reactor Internal Pump Cross Section           5.4-44 5.42    - ABWR Recirculation Flow Path                 5.4-45 5.4-3     Reactor Internal Pump Reference Charactcristics                              5.4-46

-T 5.44 Reactor Recirculation System P&lD 5.4-47 5.45 Reactor Recirculation System PFD 5.4-48 5.4-6 Main Steamline Flow Restrictor 5.4 49 l 5.4-7 Main Steam Isolation Valve 5.4 50 5.4 8 Reactor Core Isolation Cooling System P&lD 5.4 51 5.49 RCIC System PFD 5.4-53 5.4-10 Residual Heat Removal P&lD 5.4-55 p: 5.4-11 RHR PFD 5.4 59 5.4-12 Reactor Water Cleanup System P&ID 5.4-61 5.4 13 Reactor Water Cleanup System PFD 5.4-63 5.4-vii Amendment 15

ABM 2 m ioorn RIV C Standard Plant 5.4 COMPONENT AND SUBSYSTEM 5.4 3. Regulation of reactor power output over

  &     DESIGN                                                an approximate power range (70% s reactor power V     5.4.1 Reactor Recirculation System output i100% rated output) without need for moving control rods, is thus made possible by varying recirculation flow rate over the flow 5.4.1.1 Safety Design Bases                           control range. The configuration of the RRS with 10 RIPS is shown on the RRS P&lD and Pro-The reactor recirculation system (RRS) has cess Diagram, Figures 5.4 4 and 5.4 5 respec-been designed to meet the following safety design tively. RRS design characteristics are present-bases:                                                 ed in Table 5.41, Control of the reactor power through the flow control region is provided by (1) An adequate fuel barrier thermal margin the recirculation flow control system (RFCS) as shall be assured during postulated tran- described in Section 7.7. The RFCS closely sients.                                         relates to the RRS in that it provides properly condi.tioned control and logic signals, which (2) The system shall maintain pressure integrity regulate the reactor coolant recirculation flow during adverse combinations of loadings and      rate produced by the RRS under various forces occurring during abnormal, accident, steady state, transient, upset, and emergency and special event conditions,                    modes of NSSS operation. The following three subsystems are designated as part of the RFCS:

i 5.4.1.2 Power Generation Design Bases ' (1) Adjustable speed drive (ASD) subsystem The RRS meets the following power generation design bases: (2) Recirculation pump trip (RFT) subsystem l (1) The system shall provide sufficient flow to (3) Core flow measurement (CFM) subsystem l- remove heat from the fuel. More detail on these latter three subsystems is l [-)-

  '     (2) The system shall provide an automatic load given in Section 7.7.

following capability over the range of 70 to 100% rated power. S addition to the RIPS, several subsystems are also iduded as part of the RRS to provide (3) System design shall minimize maintenance closely related, or closely supporting, fun < - situations that would require core disassem- tions to the RRS in composite or to the RIPS as

             . bly and fuel removal,                           individual components. These subsystems are as follows:

5.4.1.3 Description (1) Recirculation motor cooling (RMC) subsysterr. The RRS features an arrangement of ten reactor coolant recirculation pump units commonly (2) Recirculation motor purgc (RMP) subsystem referred to as reactor internal pumps (RIPS). A cross section of a RIP is shown in Figure 5.41. (3) Recirculation motor inflatable shaft seal Collectively these provide forced circulation of (RMISS) subsystem, the reactor coolant through the lower plenum of the reactor and up through the lower grid, the The RIPS, as well as each of these sub-reactor core, steam separators, and back down the systems, are further described in later para-domicomer annuius as shown on Figure 5.4 2. The graphs. recirculation flow rate is variable over a range termed the flow control range- from minimum The motor casing has a closure assembly, at

       - flow established by certain pump performance char- its bottommost end, termed a motor cover. The acteristics to above the maximutn flow required to motor cover, in addition to its reactor obtain rated reactor power as shown on Figure I

d Amendment 13 544

ABWR uwen Standard Plant PJN C i pressure boundary closure function, provides a Heat pickup by the RMC subsystem process foundation for the bearing assembly which holds coolant is rejected via the RMHX to the reactor i the non rotating bearing elements of the thrust building cooling water system as shown on Figure bearings. The motor cover is scaled to the motor 5.4 4. casing with a single, Flexitallic type gasket. The recite motor (RM) region surrounded by the The RMHX is a vertically oriented, inner surface of the motor casing and the inner shell and tube U tube heat exchanger with a surface of the motor cover,is termed the motor bottom water box, designed for 4 tube passes, cavity. 1 shell pass as shown schematically on Figure 5.4 4. Principal approximate sizing parameters The principal element of the stretch tube feature a carbon steel shell outside diameter of section is a thin walled Inconel tube configured approximately 400 mm and approximately 2700 mm as a hollow bolt fitting stound the pump shaft length,87.9 kg/cm 2g design pressure and and within the pump nozzle, it has an external 3020 design ternperature, Tubes are stain. lip (bolt head) at its epper end and an external less steel material designed for external pres-threaded section at its lower end. The stretch sure loading. Shell tube sheet and water box ma-tube function is to achieve tight clamping of the terialis carbon steel. The RMHX stands taller internal pump diffuser to the gesketed, than the RM motor casing, but the bottoms of internal mount end of the RPV pump notzle, at the each are located approximately at the same eleva-extremes of thermal transients and pu.np operating tion. RMC subsystem primary coolant from the l conditions. Clamping action is achieved by (1) RIP motor cavity flows outbound from a nozzle capturing, with the stretch .be upper lip, a near the top of the motor casing, and through 63 mating lip on the diffuser, (~.) a stretch tube mm stainless steel piping which courses across out threaded onto the stre'ch tube lower end and upward to the RMHX primary coolant inlet where it projects into the upper region of the nozzle located near the top of the RMHX shell, motor cavity. When the stretch tube is This RMC flow proceeds downward, under the com-hydraulically pretentioned, the prescribed bined action of driving pressure head developed preload is exerted on the diffuser. (when the RIP is running) by the RM auxiliary im- - peller and by buoyancy head developed by tem-l 5.4.13.1 Recirculation Motor Cooling Subsystem perature (density) differences existing over the vertical closed loop path lengths. In moving During RIP operation, heat is generated by downward through the shell, this primary coolant the RM internals (windings and conductor electri-cal losses; viscous beating) and is also con- sweeps guided byback and flow horizontal fourth across baffic / tubethe tube bundles ! support ducted from the vessel (RPV and primary coolant) plates. Flow exits from the shell through a ; to the motor cavity water and internals. There- nozzle located just above the tube sheet and fore cooling is required for the RM, crosses, via 63 mm piping, directly back to the l RIP motor casing on a piping run which is ar.

                           ~ These RM internals, including the water ranged primarily in a horizontal plane. Upon en-present in the motor cavity, are cooled by a cir- tering the RM casing, this primary coolant is culating water process which cycles the water in drawn into the suction region of the RM auxil-the motor cavity out through the RMC subsystem to iary impeller, where it is then driven upward a recirculation motor heat exchanger (RMHX) and through the RM to begin another circuit around through return piping connections back to the this RM RMHX RM flow loop.

RM. There is one RMHX per RIP located near the RM and within the reactor support pcdestal. 5.4.1.3.2 Recirculation Motor Purge Subsystem l While the RIP is operating, flow circulation is RIP maintenance radiation doses are l powered principally shown in Figure 5.41. Theby the RM RMHXs auxiliary impeller are positioned minimized by preventing the buildup of reactor vertkally such that should the RM stop during primary coolant impurities on RM components. ructor operation, natural circulation through Such prevention is provided by the recirculation the RMC Subsystem piping will occur at flow rates motor purge (RMP) subsystem, which supplies to l l sufficient to limit the RM temperature to accept- each RIP, a flow of clean water to an RM able values. Amendment 15 5.4-2

Ja4 . ABM 23moorn RN C Standard Plant shaft stretch tube annular region located just operative lower temperature limit depends on the O above the RM upper journal bearing, reactor operating state. These lower tem-(_) perature limits for RMSP water at the entrance to the RIP are: 300C whenever the reactor l primary coolant water temperature is above 1000C; and,100C whenever the reactor The rod control system (RCS) is the source for primary coolant water temperature is 1000C or l pure water supply to the RMP subsystem as shown below. These limits are set to preclude exces-on Figurc 5.4 4. RCS water supply pressure is ap- sive temperature cycling on the pump shaft in l proximately 156 kg/cm g,2 and will range in tbc region where the RMP water first encounters l temperature from just a few degrees above conden. reactor primary coolant, i.e., the region from sate storage tank temperature to a high tem- the top of the stretch tube to the joint with perature of about 700C. At the connection from the impeller at the top of the pump shaft. The the RCS, the RMP subsystem controls the 10 RIP RMP water supply from RCS normally originates purge flow to values shown for position 8 on from the main condenser botwell. The RCS tem-Figure 5.4 5. perature ordinarily will be in the 40 to 600C range at the point of delivery to the RMP sub-RMP flow then passes into a pipe header, system as shown oa Figure 5.4 5, Since the main outside the drywell wall, where the flow becomes run of RMP piping passes through the top of the distributed to an individual pipe to each RIP. lower dryweil equipment airlock, across the Between the header and the containment pipe pen- drywell, and up to the RIPS, and since the flow etration, on each line a manual flow control rate is so low, heat pickup from drywell atmo-valve is provided and an in line flow indicating sphere will ensure temperature at entrance to switch. This permits the plant operator to the RIPS will be above the required lower ! regulate the RMP flow to each RIP within the limit. Heaters for RMP subsystem flow will not range specified for position 7 on Figure 5.4 5. be required. This conclusion is consistent with European RIP experience, and is confirmed by de-p g The lower bound flow rate value assures that a positive upward moving flow, around the pump tailed engineering analyses, shaft and into the reactor, will always be it is expected that c once daily check by the maintained. This action thus precludes con- plant operator, to confirm that flow rate to taminated reactor water from entering the motor each RIP is within the required bounds, will be cavity and in turn, the RMC subsystem piping and the only attention needed for this subsystem, equipment. The upper bound flow rate value is Rarely willit be required for the operator to set to prevent conditions which might produce adjust the manual flow control valve. rapid temperature cycling (and thus produce high cycle fatigue) on the pump shaft, instrumentation is provided to monitor RMSP subsystem performance and proside warning alarms

     !      In addition to the above bounds on RMP sub- for individual RIP high or low flow conditions, system flow rate into each RIP, upper and lower temperature bounds also apply. An upper tem- 5.4.13 3 Recirculation Motor Inflatable Shaft perature limit to the RMP water, at the inlet to Seal Subsystem the RIP, of 700C has been established to pre-clude deterioration of the inflatable seal (resil-      An inflatable seal is designated as a second.

iency), which could occur under prolonged high ary seal. A primary seal preventing downflow of temperature operation. Since the maximum supply reactor water into the motor cavity is provided water temperature from the RCS to the RMP sub- by contact faces on the pump shaft and stretch system interface is 600C, and sir.cc fluid at tube. Ordinarily separated, this primary seal this high temperature would experience only heat becomes functional when the RM, and in turn the l losses along the pipe run to the RIP, the RMP sub- pump shaft,is lowered during the RIP disman-system design inherently assures that this upper tling sequence. temperature bound will not be exceeded. [h The inflatable seal made from cleastomeric ma-terial and housed inside the upper (neck) region Lower temperature bounds also apply, and the Amendment 15 5A-3

         .AB M                                                                                         ux61oorn Standard Plant -                                                                               REV C of the motor cavity (below the stretch tube lower       The requirements on the RIPS apply equally to end) is provided. When activated, this seal the RRS subsystems. For the conditions when the functions to prevent downflow of reactor water RIPS arc not required to operate, pressure integ-from the RPV into the motor cavity. This allows rity of the RCPB must be maintained.

the motor cavity to be drained and the RM to be removed from the motor casing for repair or main- The ranges of steady state conditions over l tenance work. The RMISS is the subsystem which which RIP operation is required, is indicated on enables manually activating the seal when the the process diagram for the reactor reactor is shutdown and the motor is stopped. recirculation system (Figure 5.4 5). Capa-l The RMISS applies pressurizing water to the side bilities for the system with one RIP out of - of the seal closest to the rnotor casing inside service are listed; this diagram states that the surface. Such pressurization causes the seal RRS shall provide rated core flow with one RIP l l member to displace, and press tightly against the out of service. With 7 or 8 RIPS operating, pump shaft, producing the scaling action. A pres- plant operation is possible at reduced power. sure equalizing line is connected on the line which activates the seal and down to the motor The RMS subsystem, including the RMHXs, is re-casing drain take off point. This pressure equal- quired to operate whenever the RIPS are operat-izing line is open for normal operation of the ing. Additionally, this subsystem must function RIP. The differential pressure that is produced in the period following trip of any RIPS until by RIP auxiliary impeller action when the RIP is such time as temperature of reactor primary operating ensures that a small outward pressure coolant has been brought below the value listed assisting seal retraction will be present to on the process diagram reactor recirculation j assure that contact does not take place between system (Figure 5.4 5) representing the normal the rotating pump shsft and the inflatable seal, exit temperature of RMC subsystem fluid leaving the motor cavity. 5.4.1.4 Operation Moreover, the RMC subsystem is required to - The RRS is required to operate during startup, function throughout all events in which electric normal operation, and hot standby. It is not re- power to the RIPS is lost. Loops B and C of the quired to operate during shutdown cooling. RCW, which are cooling water sources to the RMC During various moderately frequent transient and subsystem, are required to be immediately certain infrequent transients, various RIP operat. reconnected during this power event. ing modes will be required such as: bank of-five RIPS runback to 30% speed form loss of one 5.4.1.5 Safety Evaluation reactor feed pump, trip from current reactor protection conditions; or, runback to 30% speed Reactor recirculation system malfunctions and subsequent trip. These control actic,s are that pose threats of damage to the fuel barrier all produced through control actions of the RFCS, are described and evaluated in Chapter 15. It is shown in Chapter 15 that none of the malfunc-A description of system / component primary op- tions result in significant fuel damage. The crational requirements is given below, recirculation system has suIficieat flow coastdown characteristics to maintain fuel The RIPS are required to operate in the modes thermal margins during abnormal operational tro - directed by the RFCS, without sustaining damage sients. and without experiencing wear under normal op-erations over the time period remaining until Piping and pump design pressures for the their normal scheduled removal from the reactor reactor recirculation system are based on peak for refurbishment. The intended refurbishment in- steam pressure in the reactor dome, appropriate l terval is five years. An average of two of the pump head allowances, and the elevation head ten RIPS is scheduled for removal for refurbish- above the lowest point in the recirculation ment, with these operations to be performed pump. Piping and related equipment pressure during the scheduled refueling outage, parts are chosen in accordance with applicable codes. Use of the listed code design criteria assures that a system designed, built, and op. Amendment 15 $M

_ __ _ _ _ ~. . _ _ . _ _ _ . _ _ _ _ _ . MM 23A6100AB RM C Standard Plant

                . ersted within design limits has an extremely low                    shaft to the thrust bearing. ' The weak link in this path is the bearing to
     '/*       _ failure l probability mechanism.of failure   caused Purchase           by any known specifications            shaft bolt which is loaded to less than require that integrity of the pump motor case be                   its ultimate strength by the ejection maintained through all transients and that the                     event and hence would not be expected pump remain operable through all normal and upset                  to fail.

transients._ The design of the pump and motor bearings are required to be such that dynamic (c) If the weld fails, the stretch tube - load capability at rated operating conditions is fails and the bearing to shaft bolt not exceeded during the safe shutdown earthquake, fails then the vertical restaints come in play. These restraints are Pump overspeed will not occur during the stainless steel rods which run from , course of a loss of coolant accident (LOCA) due lugs on the vessel to lugs on the motor i to a anti rotation device (ARD) which is located cover. The restraints are designed at the bottom of the RIP motor and prevents a specifically.to preclude shootout and backward rotation of the RIP. The ARD also are designed to the same criteria as prevents backward rotation during normal plant used for pipe restraints, m operation when~ one RIP is stopped and the other l' RIPS are operating. A failure modes and effects analysis of RIP is presented in Appendix 15B. l _ Each RIP is contained in a pressure' boundary

                 - housing that is attached to the RPV by a weld to         5.4.1.6 Inspection and Testing

a stub tube located in the RPV head knuckle section (See' Figure 5.41)._ Mitigation of a Quality control methods are used during fab-

                 - hypothetical failure of the weld is assured by rication and assembly of the reactor

the following: recirculation system to assure that design *
specifications are met. Inspection and testing

[3/ i :(a) The weld is bridged by the stretch tube procedures are described in Chapter 3. The a

which is, in principle a long hollow reactor recirculation system is thoroughly s bolt. The normal f.inction of the cleaned and flushed before fuel is loaded ini-stretch tube is to b'old the pump . tially. ; diffuser in place, in the event of weld failure, the stretch tube is the first During the pre operational test program, the

                               - member'.to resist ejection of the          reactor recirculation system is hydrostatically housing. The stresses in the stretch        tested at 125% reactor vessel design pressure.

tube, resulting from a guillotine 'Preoperational tests on the reactor , failure of the weld, would be 85% of the minimum specified. ultimate strength, recirculation system erstion of the pumps and,also flowinclude checking control system - op- l ,t Thus, the stretch tube may be reasonable and is discussed in Chapter 14. considered to mitigate the event. During the startup test program, horizontal i l _~ (b) In the event'that the stretch tube also and vertical motions of the RIP motor casing is o breaks, the RIP assembly will move observed. -; downward a small amount until the - l impeller -ba ckse at s. The backseat Nuclear system responses to . feature is used during RIP motor recirculation pump trips at rated temperatures i servicing to prevent leakage of reactor and pressure are evaluated during the startup p coolant when the motor cover is tests and plant power response to recirculation remuved. In the event of weld and flow control is determined. stretch tube failure,' the backseating will result in the RIP shaft restraining 5.4.2 Steam Generators (PWR) the ejection load, with the load path y being from the backseat through the Not applicable to this BWR. Amendment 15 545 l-

ABWR nam ^s REV,C Standard Plant

     ' 5.4.3 Reactor Coolant Piping                                                      .

Since the RIPS are located'inside the RPV, there is no major external reactor coolant piping connected to the ABWR pressure vessel. 5.4.4 Main Steamilne Flow Restrictors j

     . 5AA.1 Safety Des!gn Bases The main steamline flow restrictors were de-signed to:                                                                           >
     '(1), Limit the loss .of coolant from the reactor vessel following a steam line rupture outside the containment to the extent that the                                        '

c . reactor vessel water level remains high enough to provide cooling within the time re-quired to close the main steam line isola-tion valves; (2) Withstand the maximum pressure difference ex-

          ' pected across the restrictor following com-plete'severence of a main steamline;
      -(3) Limit the amount.of radiological release i

outside of the drywell prior to MSIV closure-and , t (4)'. Provide trip signals for MSIV closure. 5AAl Description - 1 J A main'steamline. flow restrictor (Figure

       -5,4 6)'is provided for- each of the four main-steamlines by giving the inside bore of each RPV steam' outlet nozzle the shape of 'a flow restric '                                j'
      - ing venturi.-

The restletor limits the coolant blowdown rate from the reactor vessel in the event.a main steamline break occurs outside the containment to

      - a (choke) flow rate equal to or less than 200% of                                      ,

rated steam flow at 72.1 kg/cm g. 2 upstream pressure, The flow restrictor is designed a id fabricated in accordance with'ASME Code, Flaid i Meters. l w < The flow restrictor has no moving par'.s. Its L mechanical structure can withstand'the velocities i l and forces associated with a main l V e l Amendment 15 5.4-5.1 l 1

ABM ursioen arv. c Standard Plant stean:line break. The maximum differential to erosion / corrosion in a high velocity steam at. -A = mosphere. The excellent performance of stain-l Ispresstre between corservatively assumedinside and to be 96.7 outside kg/cm 2 g, lessof thehigh steelin vessel velocity steam appears to be the r: actor vessel ASME Code limit pressure, due to its resistance to corrosion. A protec-tive sui;e film forms on the stainless steel The ratio of venturi throat diameter to which prevents any surface attack and this film steamline inside diameter of approximately 0.5 is not removed by the steam, re'sults in a maximum pressure differential (unrecovered pressure) of about 0.7 kg/cm2g Hardness has no significant effect on at 100% of rated flow. This design limits the erosion / corrosion. For example hardened carbon steam flow ; a severed line to less than 200% steel or alloy steel will erode rapidly in appli-rated flow, yet it results in negligible increase cations where soft stainless steel is unaf-in steam moisture content during normal op- fccted. eration. The restrictor is also used to measure steam flow to initiate closure of the main Surface finish has a minor effect on steamline isolation valves when the steam flow crosion/ corrosion, if very rough surfaces are e xceeds preselected operational limits. The exposed, the protrading ridges or points will vessel dome pressure and the venturi throat pres- erode more rapidly than a smooth surface. Expe-

    -l sure are used as l'ie high and low pressure rience shows that a machined or a ground surface sensing locations.                                    is sufficiently smooth and that no detrimental crosion will occur.

5AA.3 Safety Evaluation SAAA laspection and Testing in the. event a main steamline should break outside the containment the critical flow phenom- Because the flow restrictor forms a perma-

enon would restrict the steam flow rate in the _ nent part of the RPV steam outlet nozzle and has

        .venturl throat to 200% of the rated value. Prior no moving components, no testing program beyond R          to isolation valve closure, the total coolant the RPV inservice inspection is planned. Very l-losses from the vessel are not sufficient to slow erosion which occurs with time, has been ac.

cause core uncovering and the core is thus ad- counted for in the ASME, Section 111 design equately cooled at all times, analysis; Stainless steel resistance to erosion has been substantiated by turbine inspections at Analysis of the steamline rupture accident ;the Dresden Unit 1 facility. These inspections (Subsection 15.6A) shows that'the core remains have revealed no noticeable effects from erosion _ covered with water 'and that the amount of radioac- on the stainless steel nozzle partitions. The tive materials released to the environs through :Dresden inlet velocities are about 100 m/sec and the main steamline break d_oes not exceed the the exit _ velocities are 200 to.300 m/sec. guideline values of published regulations. However, calculations'show that, even if the crosica rates are as high as 0.1 mm per year, after 40 years of operation, the increase in-restrictor choked flow rate would be no more than 5%. A 5% increase in the. radiological. dose calculated for the postulated main steamline break accident is' insignificant. 5.4.5 Main SteamlineIsolation System The steam flow restrictor is exposed to steam of about 1/10% moisture flowing at velocities of 5A.5.1 Safety Design Bases 45 m/sec (steam piping ID) to 180 m/sec (steam restrictor throat). The flow restrictor is Type The main steamline isolation valves, indi. 308 weld overlay clad. This is similar to the vidually or collectively, shall: Type 304 cast stainless steel used in previous flow restrictors. It has excellent resistance Amendment 15 5.4-6

ABM ux6ioorn - RIN C Standard Plant

     - (1) close the main steamlines within the time - pressure balancing hole in the poppet. When p           established by design basis accident           the hole is open, it acts as a pilot valve to        I
  ;           analysis to Ilmit the release of reactor       relieve differential pressure forces on the coolant;                                       poppet. Valve stem travel is sufficient to give flow areas past the wide open poppet (2) close the main steamlines slowly enough that greater than the seat port area. The poppet simultaneous closure of all steam lines will travels approximately 90% of the valve stem not induce transients that exceed the          travel to close the main steam port area; nuclear system design limits;                  approximately the last 10% of the valve stem travel closes the' pilot valve.         The air (3) close the main steamline when required            cylinder actuator can open the poppet with a despite single failure in either valve or in   maximum differential pressure of 14.1 the associated controls to provide a higa      kg/cm2 g across the isolation valve in a level of reliability for the safety funt-      direction that tends to hold the valve closed, tion; A Y pattern valve permits the inlet and (4)    use pneumatic (N2 or air) pressure and/or      outlet passages to be streamlined; this spring force as the motive force to close- minimizes pressure drop during normal steam flow the redundant isolation valves in tbe          and helps prevent debris blockage.

Individual steamlines. The valve stem penetrates the valve bonnet (5) use local stored eeergy (pneumatic pressure through a stuffing box that has two sets of and/or springs) to close at least one isola- replaceable packing. A lantern ring and tion valve in each steam pipeline without leak.off drain are located between the two sets relying on the continuity of any variety of of packing. . electrical power to furnish the motive force l to achieve closure; Attached to the upper end of the stem is an ! O. air cylinder that opens and closes the valve and V l (6) be able to close the steamlines, either a hydraulic dashpot that controls its speed. during or after seismic loadings, to assure The speed is adjusted by a valve in the isolation if the nuclear system is breached; hydraulic return line bypassing the dashpoi and piston.

      -(7) have the. capability for testing during               Valve quick closing speed is 3 4.5 seconds normal operating condit!ons to demonstrate     when N2 or air is admitted to the upper piston     ,

that the valves will function, compartment. The valve can be test closed with - l a 45 60 second slow closing speed by admitting 5.4.5.2 Description N2 or air to both the upper and lower piston compartments. L Two isolation valves are welded in a horizon- ! .tal run of each of the four main steam pipes; one The pneumatic cylinder is supported on the valve'is as close as possible to the inside of valve bonnet by actuator support and spring the dry,vell, and the other is just outside the guide shafts. Helical springs around the

      - containment.                                          spring guide shafts close the valve if gas pressure is not available. The motion of the l           Figure 5.4 7 shows a main steamline isolation      spring seat member actuates switches in the near valve. Each is a Y pattern, globe valve. Rated        open, near closed vavle positions.

steam flow through each valve is 1.918 x

       '106 kg/hr. The main disc or poppet is at-               The valve is operated by pneumatic pressure tached to the lower end of the stem. Normal and by the action of compressed springs. The steam flow tends to close the valve, and higher control unit is attached to the gas cylinder.

inlet pressure tends to hold the valve closed. This unit contains three types of control valves The bottom end of the valve stem closes a small that open and close the main valve and excercise Amendment 15 $.4 7

4 ABWR m si m a-Standard Plant 'REV.C it at slow speed: pnuematic, AC from Division I, , and AC from Division II. Remote manual switches ! In the control room enable the operator to , operate the valves, j Operating gas is supplied to the valves from l the' plant N2 or instrument air system. An pnuematic accumulator between the control valve < and a check valve prosides backup operating gas. 1 Each valve is designed to accomodate saturated steam at plant operating conditions l with a moisture content of approximately 0.3% an oxygen content of 30 ppm, and a hydrogen content

     .of 4 ppm. The valves we furnished in con-formance with a design pressure and temperature rating in excess of plant operating conditions to accommodate plant overpressure conditions.

In the worst case, if the main steamline should rupture downstream of the valve, steam flow would quickly increase to 200% of rated flow. Further increase is prevented by the venturl flow restrictor. O. I ( t i O Amendment t3 5.47.I' l

ABM 2346ioorn Standard Plant REY.C During approximately the first 75% of closing, line valve installations are designed as Seismic A the valve has little effect on flow reduction, Category I equipment. The valve assembly is ij because the flow is choked by the venturi manufactured to withstand the safe shutdown restrictor. After the valve is approximately 75% earthquake forces applied at the mass center of closed, flow is reduced as a function of the the valve with the valve located in a horizontal valve area versus travel characteristic, run of pipe. The stresses caused by horizontal and vertical seismic forces are assumed to act The design objective for the valve is a simultaneously. The stresses caused by seismic minimum of 60 years service at the specified oper- loads are combined with the stresses caused by sting conditions. Operating cycles are estimated other live and dead loads including the operM. l to be 1500 in 60 years and 3750 execise cycles in ing loads. The allowable stress ce this combina-60 years. tion of loads is based on a percentage of the al-lowable yield stress for the material. The in addition to minimum wall thickness required parts of the main steam isolation valves that by applicable codes, a corrosion allowance is constitute a process fluid pressure boundary are added to provide for 60 years service. designed, fabricated, inspected, and tested as required by the ASME Code Section Ill. Design specification ambient conditions for l normal plant operation are 57 C normal tem- 5.4.5.3 Safety Evaluation perature and 60% humidity in a radiation field of 202 rad /hr neutron plus gamma, continuous for In a direct cycle nuclear power plant the design life. The inside valves are not con- reactor steam goes to the turbine and to other tinuously exposed to maximum conditions, par- equipment outside the containment. Radioactive ticularly during reactor shutdown, and valves materials in the steam are released to the envi-outside the primary containment and shielding are rons through process openings in the steam in ambient conditions that are considerably less system or escape from accidental openings. A severe, t q large break in the steam system can drain the water from the reactor vessel faster than it is l The main steamline isolation valves are de. replaced by feedwater. l signed to close under accident environmental con-l ditions of 1710C for one hour at drywell design The analysis of a complete, sudden steamline pressure. In addition, they are designed to break outside the containment is described in remain closed under the following post accident Subsection 15.6.4. The analysis shows that the environment conditions: fuel barrier is protected against loss of cooling if main steam isolation valve closure is (1) 1710C for an additional 2 hours at drywell within specified limits, including instrumenta-drywell pressure of 3.16 kg/cm 2g, tion delay to initiate valve closure after the break. The calculated radiological effects of (2) 1600C for an additional 3 hours at drywell the radioactive material assumed to be released design pressure of 3.16 kg/cm 2g, with the steam are shown to be well within the guideline values for such an accident. (3) 1210C for an additional 18 hours at 1.8 L kg/cm 2g maximum, and The shortest closing time (approximately 3 seconds) of the main steam isolation valves is (4) 930C for an additional 99 days at 1.4 also shown to be satisfactory. The switches on kg/cm2g the valves initiate reactor scram when specific ( conditions (extent of valve closure, number of To resist sufficiently the response motion pipe lines included, and reactor power level) I from the safe shutdown earthqude, the main steam are exceeded (Subsection 7.2.1). The pressure !' rise in the system from stored and decay heat l' I b U Amendment 15 5A-8 l

ABWR u^a*^n Standard Plant REV C may cause the nuclear SRVs to open briefly, but system, each valve is tested as discussed in (n) v' the rise in fuel cladding temperature will be in. significant. No fuel damage results. Chapter 14. Two isolation valves provide redundancy in The ability of this Y pattern globe valve to each ste amline so either can perform the isola-close in a few seconds after a steamline break, tion fuiction and either can be tested for under conditions of high pressure differentials leakage after the other is closed. The inside and fluid flows with fluid mixtures ranging from valve, the outside valve, and the respective mostly steam to mostly water, has been demon- control systems are separated physically. strated in a series of dynamic tests. A full size,20 inch valve was tested in a range of The isolation valve is analyzed and tested steam water blowdown conditions simulating postu- for carthquake loading. The loading caused by lated accident conditions (Reference 1). the specified carthquake loading is required to be within allowable stress limits and with no The following specified hydrostwc, leakage, malfunctions that would prevent the valve from and stroking tests, as a minimum, are performed closing as required, by the valve manufacturer in shop tests: Electrical equipment that is associated with (1) To verify its capability to close at set- the isolation valves and operated in an accident tings between 3 and 4.5 sec (response time environment is limited to the wiring, solenoid for full closure is set prior to plant op- valves, and position switches on the isolation eration at 3,0 see minimum, 4.5 see valves. The expected pressure and temperature maximum), each yalve is tested at rated pres- transients following an accident are discussed l sure (71.1 kg/cm g) and no flow. in Chapter 15, i 5.4.5.4 laspection and Testing (% l l () (2) Leakage is measured with the valve seated. The main steam isolation valves can be func. The specified maximum seat leakage, using tionally tested for operability during plant op-cogd water at design pressure, is 2 eration and refueling outages. The test provi-cm /hr/25.4 mm. cf nominal valve size. In sions are listed below. During refueling outage l the main steam isolation valves can be function. addition, an air seat lealgage test is con. l ducted using 2.85 kg/cm g pressure up- ally tested, leak tested, and visually inspect-stream. Maximum permissible leakage is ed. l 0.74/sec/25.4 mm. of nominal valve size. 1 (3) Each valve is hydrostatically tested in ac- tested and exercised individually to the 90% l cordance with the requirements of tiie appli- open position in the slow closing mode, cable edition and addenda of the ASME Code. During valve f abrication, extensive i nondestructive tests and examinations are conducted. Tests include radiographic, liquid. penetrant, or magnetic-particle ex- Leakage from the valve stem packing is col-aminations of casting, forgings, welds, lected and measured by the drywell drain l hardfacings, and bolts, system. During shutdown while the nuclear L system is pressurized, the leak rate through the inner valve stem packing can be measured by col, lecting and timing the leakage. The leak through the pipeline valve seats After the valves are installed in the nuclear can be measured accurately during shutdown by g i i l Amendment 15 549

ABWR m aman Standard Plant RIN. C the following suggested procedure:

enough period to obtain meaningful data. An alternate means of leak testing the outer iso-l (1)- With the reactor at approximately $20C and lation valve is to' utilize the previously normal water level and decay heat being re- noted steamline plug and to determine leakage moved by the RHR system in the shutdown by pressure decay or by inflow of the test cooling mode, all main steam isolation valves medium to maintain the specific test pres. are closed utilizing both spring force and sure. air pressure on the operating cylinder. During pre startup tests following an exten-(2) Nitrogen is introduced into the reactor sive shutdown, the valves will receive the same

 '                vessel above normal water level and into the hydro tests that are imposed on the primary connecting main steamlines and pressure is system.

l raised to 1.4 2.1 kg/cm 2g. An alter-nate means of pressurizing the upstream side Such a test and leakage measurement program of the inside isolation valve is to utilize a ensures that the valves are operating correctly, steamline plug capable of accepting the 1.4 - 2 2.1 kg/cm g pressure acting in a direc. 5.4.6 ReactorCorelsolation CoolingSystem tion opposite the hydrostatic pressure of the ? fully flooded reactor vessel. Evaluations of the reactor core isolation ; cooling system against the General design g (3) A pressure gage and flow meter are connected Criteria (GDC) 5,29,33,34 and 54 are provided to the test tap between each set of main in Subsection 3.1.2. Evaluations against the L steam isolation valves. Pressure is held ECCS GDC 2,17,27,35,36 and 37 are provided l below 0.07 kg/cm2 g, and flow out of the below. space between each set of valves is measured to establish the leak rate of the inside iso. Compliance with GDC 2. The RCIC system is lation valve, housed within the reactor building which - provides protection against wind, floods, (4) To leak check the outer isolation valve, the missiles and other natural phenomena. Also, 1 . reactor and connecting steamlines are flooded RCIC system and its components are designed to to a water level that gives a hydrostatic withstand earthquake and remain functional head at the inlet to th- lnner isolation following a seismic event. valves slightly hight than the pneumatic - test pressure ,o be applied between the Compliance with GDC 17. The RCIC is a part of

                 . valves. .This assures essentially zero the ECCS network. It is powered from a Class leakage through *be inner valves. If neces- 1E source independent of the HPCF power sary to achieve the desired water pressure at sources.

'l the inlet to the inner isolation valves, gas system, it is redundant to the twv HPCF lo from 'a suitable pneumatic supply is intro- which comprise the high pressure ECCS (1 RCIC duced into the reactor vessel top head. Ni, and 2 HPCF). Since independent Class 1E power trogen pressure (1.4 2.1 kg/cm g) 2 is supplies are provided, redundancy and single then applied to the space between the isola- failure criteria are met, GDC 17 is satisfied.

                 . tion valves. The stem packing is checked for leak tightness. Once any detectable stem Compliance with GDC 27. As discussed in packing leakage to the drain system.has been Subsection 3.1.2.3.8.2, the design of the accounted for, the seat leakage test is con. reactivity under postulated accident conditions ducted by shutting off the pressurizing gas with appropriate margin for stuck rods. The and observing any pressure decay. The volume capability to cool the core is maintained under between the closed valves is accurately all postulated accident conditions by the RHR

_known. Correction for temperature variation system. Thus, GDC 27 is satisfied without RCIC during the test period are made,if necessary, system. to obtain the required accuracy. Pressure and temperature are recorded over a long Compliance with GDC 35. The RCIC in conjunction with HPCF, RHR and auto Amendment 15 5.4 10 i

MM 23A6100AD ' Standard Plant .imyJ , depressurization systems perform adequate core (1) a loss of coolant (LOCA) event;

       &                  cooling to. prevent excessive' fuel clad Q'                 temperature during LOCA event.         Detailed discussion of RCIC meeting this GDC is described (2) vessel isolated and maintained at hot standby; in Subsection 3.1.2.

(3) vessel isolated and accompanied by loss of Compilance with GDC 36. The RCIC system is coolant flow from the reactor feedwater designed such that in service inspection of the system; system'and its components is carried out in accordance with the intent of ASME Section XI. (4) complete plant shutdown with loss of normal The RCIC design specification requires layout and feedwater before the reactor is depressur- l arrangement of the containment penetrations, ized to a level where the shutdown cooling process piping, valves, and other critical system can be placed in operation; or equipment outside the reactor vessel, to the

                     ; maximum practical extent, permit access by (5) loss of AC power for 30 minutes.

personnel and/or appropriate equipment for testing and inspection of system integrity. Acceptance criteria 11.3 of SRP Section 5.4.6 Compliance with GDC 37. The RCIC system is states function that thethe without RCIC system availability must of any AC perform its l designed such that system and its components can power. Review Procedure 111.7 further requires be periodically tested to verify operability, that there be sufficient battery capability.for = Systems operability is demonstrated by two hours of operation. -While RCIC is designed , preoperational and periodic testings in for 30 minutes of operation during loss of ac accordance with RG'1.68. Preoperational test power, the battery capacity should allow over will ensure proper functioning of controls, four hours of operation,.which would meet this ' instrumentation, pumps and valves. Periodic requirement. testings confirm systems availability.and n().

                      -operability-through out the life of the plant.         During loss of AC power, RCIC when started at

,~V During normal plant operation, a full flow pump water level 2 is capable of preventing water l test is being-performed periodically to assure level from dropping below the level which ADS systems design flow and head requirements are mitigates (Level 1). This accounts for decay attained. All RCIC systems components are heat boil off and primary system leakages, t capable of individual functional testings during plant operation. This includes sensors, Following a reactor scram, steam generation instrumentation, control: logics, pump, valves, will continue at a reduced rate due to the core L and more. Should the need for RCIC operation fission product decay heat. At this time the ,

                     ; occur while the system is being tested, the RCIC- turbine bypass system will divert the steam to-                    7
                     . system and its _ components will automatically L           ,            re aligned to provide cooling water into the icactor. The above test requirements satisfy                                                                        ,

i GDC 37, 1: 5.4.6.1 Design Basis The reactor core isolation cooling (RCIC) system is a safety system which consists of a i turbine, pump, piping ' valves, accessories, and l instrumentation designed to assure that suffi-cient reactor water inventory is maintained in 7 l, the reactor vessel to permit adequate core cool-ing to take place This prevents reactor fuel- 1

                       ' overheating during the following conditions:                                                                        L i

Amendment 15 - 5.4-10.i l l 1

m a MM 23A6100AD

c Standard Plant ' nn e the main condenser, and the feedwater system will (3) replenishes reactor vesselinventory.

_.. ! supply the makeup water required to maintain reactor vessel inventory. Redundantly the HPCS system performs a similar function, hence providing single failure In the event the reactor vessel is isolated protection. Both systems use different reliable and the feedwater supply unavailable, relief electrical power sources which permit operation valves are provided to automatically (or remote with either onsite or offsite power. Addition-manually) maintain vessel pressure within desir- ally, the RHR system performs a residual heat able limits. The water level in the reactor removal function. vessel _will drop due to continued steam gen.

                                           . eration by decay heat. Upon reaching a predeter.         5.4.6.1.1.2 Isolation mined low level, the RCIC system will be initi-ated automatically. The turbine driven pump will            isolation valve arrangements include the fol-supply demineralized make up water from (1) the lowing:

condensate storage tank (CST) to the reactor

vessel and (2) the suppression pool. Seismically (1) Two RCIC lines penetrate the reactor coolant installed level instrumentation is provided for pressure boundary. The first is the RCIC automatic transfer of the water source with steamline which branches off one of the main manual override from CST to suppression pool on steamlines between the reactor vessel and receipt of either a low CST water level or high the main steam isolation nives. This line

                    't-suppression pool level signals (CST water is                 has two automatic motor operated isolation primery source). The turbine will be driven with             valves, one is located inside and the other la portion of the decay heat steam from the                     outside the drywell. An -automatic
                  -                             reactor vessel and will exhaust to the suppres-             motor operated inboard RCIC isolation bypass

sion pool. Suppression pool water is not usually valve is used. The isolation signals noted

                                            . demineralized and hence should only.be used in                earlier close these valves.
                                             ' the event all sources of demineralized water have
  -                                         : been exhausted;                                          (2) The RCIC pump discharge line is the other line that penetrates the reactor coolant 7                                   _._ During RCIC operation, the suppression pool
                                                                             ~

pressure boundary, which directs flow into a = J :shall act as the heat sink for steam generated by feedwater line just outboard of the primary _ l rea'ctor decay heat. This will result in a rise containment. This line has a testable ' check Lin pool water temperature. RHR heat exchangers

valve and an automatic motor operated valve c are used to maintain pool water temperature

located outside primary containment.

                                            > within acceptable limits by cooling the. pool 4                           water.                                                (3) The RCIC turbine exhaust line also pene.

trates the containment. Containment

                                             . 5.4.6.1.1 Residual Heat and Isolation                         penetration is located about a meter above e                                                       .

the suppression pool maximum water level A

                                    ,        . 5.4.6.1.1.1 Residual Heat                                     vacuum-breaking line'with two vacuum breakers in series runs in the suppression

_ , _ The RCIC system shallinitiate and discharge, pool air space and connects to the RCIC. within 30 seconds, a specified constant flow into t u r bi n e e x h a u's t line inside the the reactor' vessel over a specified pressure containment. Located outside the. range. The RCIC water discharge into the reactor containment in the turbine exhaust line is a

       =                                         vessel varies between a temperature of 4.50C up              remote manually controlled motor operated to and including a temperature of 770C The                   isolation valve.

_ .' mixture of the cool RCIC water and the hot steam

does the following: (1) quenches steam, (4) The RCIC pump suction line, minimum flow pump discharge lir.e, and turbine exhaust (2) removes reactor residual heat, and line penetrate the containment and are sub-Amendment 15 5.4-11 l . . . . . _

1 ABM aarsioorn Standard Plant RrW C I merged in the suppresslor pool. The isolation A design flow functional test of the RCIC valves for these lines:are outside the contain- system may be performed during normal plant op-ment and require automatic isolation operation, eration by drawing suction from the suppres ion except for the turbine exhaust line which has poel and discharging through a full 5 test remote manual operation. return line to the suppression pool. All compo-nents of the RCIC system are capable of indi-The RCIC system design includes interfaces with vidual functional testing during normal plant op- , redundnal leak detection devices, monitoring: eration. System control provides automatic j return from test to operating mode if system (1) a high pressure drop across a flow device in initiation is required, and the flow is auto-the steam supply line equivalent to 300 matically directed to'the vessel. percent of the steady state steam flow at l '83.8 kg/cm 2abs preasurc; Also, see Subsection 5.4.6.2.4. (2) a high area temperature utilizing tem- 5.4.6.1.2.2 Manual Operation l perature switches as described in the leak detection system (high area temperature shall in addition to the automatic operational fen- l be alarmed in the control room); tures, provisions are included for remde. manual I startup, operatiot, and shutdown of the RCIC l(3) a low reactor pressure of'3.5 kg/cm g 2sy tem provided initiation or shutdcwn signals - minimum; and do not exist. (4) a high pressure between the RCIC turbine 5.4.6.1.3 Loss of 0ffsite Power exhaust rupture diaphragms. The RCIC system power is derived from a reli-These devices, activated by the redundant power able source that is maintained by either onsite supplies, automatically isolate the steam supply or offsite power, to the RCIC turbine and trip the turbine. HPCF provides redundancy for RCIC should RCIC become 5.4.6.1.4 Physical Damage I isolated. ' The system is designed to the requirements l

   - 5.4.6.1.2 Reliability, Operability, and Manual      presented in Table 3.21 commensurate with the:

l Operation safety importance of the system and its equip- i ment. The RCIC is physically located in a dif. 5.4.6.1.2.1 Reliability and Operability ferent quadrant of the reactor building and uti- , lizes different divisional power and seperat- l The RCIC system (Table 3.21) is designed com- electrical routings than its redundant system as , mensurate with the safety importance of the system discussed in St.bsection 5.4.6.1.1.1 a n d l and its equipment. Each component is indi 5.4.6.2.4. I vidually tested to confirm compliance with system requirements. The system as a whole is tested 5.4.6.1.5 Environment

   ! during both the start up and pre operational phases of the plant to set a base mark for system      The system operates for the time intervals reliability. To confirm that the system main. and the environmental conditions specified in tains this mark, functional and operability Section 3.11.

testing is performed at predetermined intervals throughout the life of the plant. 5.4.6.2 System Design 5.4.6.2.1 General i

O Amendment 15 5.4 12

ABM 234sioorn Standard Plant REV.C 5.4.6.2.1.1 Description (7) High reactor water level (Level 8) closes F037, F012, F045 and subsequently F004 and The summary description of the reactor core F011. This level signal is sealed in and isolation cooling system is presented in Subsec- must be manually reset. It will tion 5 4.6.1 which defines the general system automatically clear if a low reactor water functions and components. The detailed descrip- level (Level 2) reoccurs. tion of the system,its components, and operation is presented in the following subsections. (8) High turbine exhaust pressure, low pump suction pressure,110% turbine electrical 5.4.6.2.1.2 Diagrams overspeed, or an isolation signal actuates the turbine trip logic and closes the Thc following diagrams are included for the turbine trip and throttle valve When the RCIC systems: signal is cleared, the trip and throttle valve must be reset from the control room. (1) Figure 5.4 8 is a schematic diagram showing components, piping, points where interface (9) Overspeed of 125 percent trips the me- l system and subsystems tie together, and in- chanical trip which is reset at the strumentation and controls associated with turbine, subsystem and component actuation. (10) An isolation signal closes F035, F036, (2) Figure 5.4 9 is a schematic showing tem- F048, and other valves as noted in Items perature, pressure and flows for RCIC op- (6) and (8), cration and system process data hydraulic re-quirements. (11) An initiation signal opens F001 and F004, F037, F012 and F045 when other permissives 5.4.6.2.1.3 Interlocks are satisfied; starts the gland seal system; and closes F008 and F009. The following defines the various electrical interlocks: (12) High and low inlet RCIC steamline drain 4 pot (1) Valves FO39 and F047 are two key locked open :58.levels respectively open and close !

   >$           valves with individual key locks.

(13) The combined signal of low flow plus pump S (2) The FOO1 limit switch activates when not discharge pressure opens and, with in-

     $          fully closed and closes F008 and F009.               creased flow closes, F011. Also see items (5), (6) and (7),

(3) The F039 limit switch activates when full open and :lears permissive so F037and F045 5.4.6.2.2 Equipment and Component Description to open. 5.4.6.2.2.1 Design Conditions (4)' The 1 ?,7 and turbine trip and throttle valve limit switchs activate when not fully closed Operating parameters for the components of to initiate the turbine governor valve the RCIC system are shown in Figure 5.4 9. The signal ramp' generator and to clear permis- RCIC components are: sives for F004 to open. (1) o. e 100% capacity turbine and accessories; (5) The F037 limit switch activates when fully closed and permits F031, F032, FN0 and FN1 (2) one 100% capacity pump assembly and accesso-to open and closes FON and F011. ries; and (6) The turbine trip throttle valve (part of (3) Fiping, valves, and instrumentation for:

 .              C002) limit switch activates when fully closed and closes F004 and F011.                      (a) steam supply to the turbine, l

Amendment 15 5.4 13

AB M 2mioota Standard Plant aw c (b) turbine exhaust to the suppression pool, Evaluation of reliability of the instrumentation for the RCIC shows that no failure of a single (c) make up supply from the condensate initiating sensor either prevents or falsely .'

storage tank to the pump suction, starts the system. (d) make up supply from the suppression pool In order to assure HPCF or RCIC availability I to the pump suction, and for the operational events noted previously, certain design considerations are utilized in (c) pump discharge to the feedwater line, a design of both systems, full flow test return line, a minimum , flow bypass line to the suppression (1) Physicalladependence pool, and a coolant water supply to ac-cessory equipment. The two systems are located in separate areas of the reactor building. Piping runs The basis for the design conditions is the are separated and the water delivered from

                - ASME B&PV Code Section III, Nuclear Power Plant                   each system enters the reactor vessel via Components,                                                       different nozzles.

(2) Prime Mover Diversity and Independence L l Independence is achieved by using a steam turbine to drive the RCIC pump and an elec-tric motor driven pump for the HPCF system. Analysl: of the net positive suction head The HPCF motor is supplied from either (NPSH) available to the RCIC pump in accordance normal AC power or a separate diesel gen- q with the recommendations of Regulatory Guide 1.1 erator, is provided in Table 5.4 la. (3) Controlindependence

           - 5.4.6.2J.2 Design Parameters Independence is secured by using different Design parameters for the RCIC system compo-                  battery systems to provide control power to nents are given in Table 5.4 2, See Figure 5.4 8                 each unit. Separate detection / initiation for cross reference of component numbers.                        logics are also used for each system,                                 i 5.4.6.2.3 Applicable Codes and Classifications              (4) EnvironmentalIndependence
           .         The RCIC system components within the' drywell-               Both systems are designed to meet Safety l                 including the outer isolatio_n valve are designed                 Class 1 requirements. Environment in the                              ;
             - in accordance with ASME Code Section III, Class                     equipment rooms is maintained by separate
                                                                                                                                                         ~

1, Nuclear Power Plant Components. The RCIC auxiliary systems.

          - System is also designed to S?ismic Category I.

(5) PeriodicTesting _ !. The RCIC system component-classifications and those for the condensate storage system are given A design flow functional test of the RCIC. J in Table 3.21. can'be performed during plant operation by; taking suction from the suppression pool and

         . 5.4.6.2.4 System Reliabil_ity Considerations discharging through the full flow test return line back to the suppression pool.

To assure that the RCIC will operate when The discharge valve to the feedwater line-

         . necessary and in time to prevent inadequate core                        remains closed during the test and reactor cooling, the power supply for the system is taken                  operation is undisturbed. All components of from reliable immediately available energy sour-                   the RCIC system are capable of individual l

as. Added i.tsurance is given by the capability for periodic testing'during station operation, functional eration. Controltesting during system designnormal providesplant op- ( . : Amendment 15 5.4 14

h MM- 2 W 100AD Standard Plant' Rev. c automatic return from test to operating mode if system initiation is required, and the flow is au-l- .. tomatically directed to the vessel, I l' /' N . sjs i l' 1 '. I .' ( . Il' 3 t t ' Y 1' I~ I

                   ' Amendment 15                                                                                                                     5.4 14.I'

ABM 23xaoorn Standard Plant RM C (6) General overspeed trips) opens to accelerate the turbine [ )_ to an initial peak speed of approximately 1500 V Periodie inspections and maintenance of the rpm; now under governor control, turbine speed turbine pump unit are conducted in accor- is returned to the low limit turbine speed dance with manufacturers instructions, demand of 700 to 1000 rpm. After a Valve position indication and instrumenta- predetermined delay (5 to 10 sec), the steam tion alarms are displayed in the control supply valve leaves the full closed position and room. the ramp generator is released. The low signal select feature selects and sends this increasing 5.4.6.2J System Operation ramp signal to the governor. The turbine increases in speed until the pump flow satisfies Manual actions required for the various modes the controller setpoint. Then the controller of RCIC are defined in the following subsections. leaves saturation, responds to the input error, and integrates the outptt signal to satisfy the 5.4.6.2.5.1 Standby Mode input demand. During normal plant operation, the RCIC The operator has the capability to select system is in a standby condition with the motor- manual control of the governor, and adjust speed operated valves in their normally open or nor- and flow (within hardware limitations) to match mally closed positions as shown in the piping and decay beat steam generation during the period of instrumentation diagrr.m (F&lD) included in Figure RCIC operation. 5.4 8. In this mode, the RCIC pump discharge line is kept filled. The relief valve in the - .? pump delivers the makeup water to pump suction line protects against overpressure the reactor vessel through the feedwater line, from backleakage through the pump discharge isola- which distributes it to obtain mixing with the tion valve and check valve, hot water or steam within the reactor vessel.

           '5.4.6.2.5.2 Emergency Mode (Transient Events                              The RCIC turbine will trip automatically upon         4 and LOCA Events)                                                       receipt of any signal indicating turbine                 ;

overspeed, low pump suction pressure, high Startup of the RCIC system occurs auto- turbine exhaust pressure, or an auto. isolation matically either upon receipt of a reactor vessel signal. Automatic isolation occurs upon receipt - low water level signal (Level 2) or.a high of any signal indicating: drywell pressure signal. During startup, the turbine. control system limits the turbine pump (1) A high pressure drop across a flow device in

           . speed to its maximum normal operating value,                                the steam supply line equivalent to 300% of -

L controls transient acceleration, and positions the steady-state steam flow at 83.8 the turbine ~ governor valve as required to kg/cm 2abs, maintain constant pump discharge flow over the pressure range of the system. Input to the- (2) A hlgh arca temperature. turbine governor is from the flow controller monitoring the pump discharge flow. During (3) A low reactor pressure of.3.5 kg/cm 2g standby conditions, the flow controller output is minimum, saturated at its maximum velue, (4) A high pressure between the turbine exhaust When the RCIC system is shut down, the low rupture diaphragms, signal select feature of the turbine control system selects the idle setting of a speed ramp The steam supply F037, steam supply bypass generator. The ramp generator output signal valve F045 and cooling water supply valve F012 during shutdown corresponds to the low limit step will close upon receipt of signal indicating and a turbine speed demand of 700 to 1000 rpm. high water level (Level 8) in the reactor vessel. These valves will reopen (auto-restart) l l 't -l On RCIC system start, the bypass valve F045 should an indication of low water level (Level (provided to reduce the frequency of turbine 2) in the reactor vessel occur. Water Level 2 l Amendment 15 5A 15 i I _____m__ m__.m..__.--_--- _-.___m___ ______.__ _ _ _ _ _ _

j 2184100AB _ Standard Plant m l aut'omatically reset the water level trip signal. The RCIC system can also be started, operated, ,

     .and shut down remote manually provided initiation                   -

or shutdown signals do not exist. 5A.6.2.53 Test Mode ~ A design functional test of the RCIC N e lM i 9 Amendment 15 5.4 15.1 f _ .----- __---.m__.----__ __-m__

ABM ursiman Standard Plant REV C system may be performed during normal plant op- vessel outside the core shroud (vir the feedwa-eration by drawing suction from the suppression ter line on Division A and via th iow pressure O. pool and discharging through a full flow test flooder subsystem discharge return lir:e on return line back to the suppression pool. The Divisions B and C). discharge valve to the vessel remains closed during test mode operation. The system will The RHR provides two independent containment automatically return from test to operating mode spray cooling systems (on loops B and C) each if system initiation is required and the flow having a common header in the wetwell and a com-will be automatically directed to the vessel, mon spray header in the drywell and sufficient capacity for containment depressurization. 5.4.6.2.5.4 Limiting Single Failurt Shutdown cooling uction is taken directly The most limiting single failure with the from the reactor via three shutdown cooling i RCIC system and its HPCF system backup is the suction nozzles on the vessel. Shutdown cooling failure of HPCF. With an HPCF failure, if the return flow is via the feedwater line on loop A capacity of RCIC system is adequate to maintain and via low pressure flooder subsystem discharge reactor water level, the operator shall follow return lines on loops B and C.

  ,                       Subsection 5.4.6.2.5.2: However, if the RCIC capacity is inadequate, Subsection 5.4.6.2.5.2           Connections are provided to the upper pools still applies, but additionally the operator may on t'wo loops to return shutdown cooling flow to also initiate the ADS described in Subsection the upper pools during normal refueling activi-l6.3.2.2.2.                                             ties if necessary. These connections also allow the RHR to provide additional fuel pool cooling 3                           5.4.6.3 Performance Evaluation                       capacity as required by the fuel pool cooling

[ system during the initial stages of the refuel-The analytical methods and assumptions in ing outage. evaluating the RCIC System are presented in Chap-O ter 15 and Appendix 15A. The RCIC system pro-vides the flows required from the analysis (Fig-The RHR provides an AC independent water addition subsystem which consists of piping and ute 5.4 9) within a 30 second interval based upon manual valves connecting the fire protection considerations noted in Subsection 5.4.6.2.4. system to the RHR pump discharge line on loop C downstream of the pump's discharge check valve. 5.4.6.4 Preoperational Testing This flow path allows for injection of water into the reactor vessel and the drywell spray The preoperational and initial startup test during severe accident conditions in which all program for the RCIC system is presented in AC power and all FG pumps are unavailable. Chapter 14. Additionally, an of bookup outside the reactor building for connection of a fire truck l 5.4.7 Residual Heat Removal System pump to an alternate water source is provided. Evaluations of residual heat removal (RHR) sys- As shown in Table 5.4 4, the RHR heat l g tem against the applicable General Desig.. Crite- exchanger primary (tube) side design pressure is r g ria (GDC) are provided in Subsection 3.1.2 and 35.0 kg/cm2 8 and the secondary (shell) side 5.4.7.1.4. design pressure is 14.0 kg/cm 2g. This pressure distribution is acceptable for the 5.4.7.1 Design Basis following teasons: The RHR is composed of three electrically and (1) Heat exchanger primary side leakage is mechanical independent divisions designated A, B, accommodated by the surge tank of the and C. Each division contains the necessary pump loop of the reactor building piping, pumps, valves and beat exchangers, in cooling water system. The inlet to the low pressure flooder mode, suction is taken the secondary side of the heat from the suppression pool and injected into the Amendment 15 SA-16 I l

ABWR tw=n Standard Plant nvc exchanger is sixays open to this (4) Shutdown cooling to remove decay and continvo':;iy cunning pump loop, sensible heat frorr the reactor. This , includes the safety.<eh ed requirements (2) The ABWR design bases against that the reactor mus 's b'ought to a cold interf acing LOCAs essentially climinates interfacing LOCA concerns by ; requiring that: (a) two or more malfunctions are necessary to expose piping systems to reactor operating pressure with > design pressures greater than or equal to ane third reactor operating pressure (e.g., RHR j heat exchanger primary side); and , (b) three or more tr.alfunctions are n:cessary to l expose piping systems to reactor eperating pressure with design pressures less than one third reactor operating pressure (e.g., R11R ! heat exchanger secondary side). l Further, the interfacing LOCAs design j bases requires the motor operated ECCS { injection valves to be tested wi th the reactor vessel at low pressure and ECCS ] I injection lines to have inboard testable I check valves with position indication in ' the control room. $.4.7.1.1 Functional Design Itasis The RilR provides the following four principal ' functions: (1) Core cooling water supply to the reactor to compensate for water loss beyond the normal control range from any cause up to and including the design basis (LOCA). (2) Suppression pool cooling to remove heat re-

     !cesed to the suppression pool (wetwell),as necessary, following heat inputs to the pool.

(3) Wetwell and drywell sprays to remove heat , and condense steam in both the drywell and wetwell air volumes following a LOCA. In addition, the drywell sprays are intended to provide removal of fission products released during a LOCA. O Amendment 13 $.416.1 l

ABWR m-n Sandard Plant my c shutdown condition using safety grade 5.4.7.1.1.2 Test Mode

/] equipment as well as the non safety V functions associated with refueling and Full flow functional tests of the RilR can be servicing operations, performed during normal plant operation or during plant shutdown by manuel operation of the Also, other secondary functions are provided, RHR from the control room. Fn , sting such as periodic testing, fuel pool cooling, pool during normal plant operall 'c .np is draining and AC independent water addition. sterted and th' return lira to m mression pool is opened. A reverse sequence i., ed to The RHR system has ten different operational terminate this test. Upon receipt of an configurations that are discussed separately to automatic initi. ation signal while in the flow provide clarity. testing mode, the system is retuined to automatic control.
$.4.7.1.1.1 Law Pressure riooder (LPFL) Mode
$.4.7.1.1.3 Minimum riow Mode Each loop in the low pressure flooder subsystem provides core cooling water supply to if the main discharge flow reaches a compensate for water loss beyond the normal predetermined low value, the minimum flow valve control range from any cause up to sr.d including in that loop will automatically open to provide the design basis (LOCA). This subsystem is some pump flow. During this mode, water is inillaicd automatically by a low water level in pumped from the suppreuion pool and returned to the reactor vessel or high pressure 19 the the suppression pool via the low flow bypass drywell. Each loop in the system can also be line. Sufficient main discharge flow will cause placed in operation by means of a manual the minimum flow valve to close automatically.

Initiation pushbutton switch. 5.4.7.1.1.4 Standby Mode l (q

  *g During the low pressure flooder mode, water is pumped from the suppression poolinitially and         During normal plant operation, the R11R loops diverted through the minimum flow lines until the   are in a atandby condition with the injection valve in the discharge line is motor operated valves in the normally open or signalled to open on low reactor pressure. As       normally closed position. The valves on the the injection valve opens on low reactor            suppression pool suction line are open and the pressure, flow to the RPV comes from the            minimum flow valves are open; the test valves suppress,n pool, through the RIIR heat exchanger,   and injection valves are clesed. The RilR pumps and the injection valve. This creates a flow are not running while the standby water leg, signal that closes the minimum flow line. The pumps (line fill pumps) are running to keep the RilR shall be capable of delivering flow into the pump discharge lines filled. The relief valves reactor vessel within 36 seconds after receipt of in the pump suction and pump discharge lines the low pressure permissive signal following protect the lines against overpressure, system initiation. This assumes a one second delay for the instrumentation to detect the low 5.4.7.1.1J Suppression Pool Cooling pressure permissive and generate an initiation signal to the injection valve. Consequently, the       The suppression pool cooling subsystem pro.

36.second RilR requirement is consistent with the vides means to remove heat released into the 37 second injection time assumed in LOCA suppression pool, as necessary, following heat analyses. Additionally, the time for the pumps additions to the pool. During this mode of to reach rated speed, from the receipt of at operation, water is pumped from the suppression (least one actuation signal, is 29 seconds, pool through the RHR heat exchangers, and back l to the suppression pool. This subsystem is The system remains in this mod until manually initiated manually, stopped Ly the operator. [N Amendment 13 5417

1 ABWR swwm Standard Plant uvc 5.4.7.1.1.6 Wetwil and Drywell Spray Cooling SRP 5.4.7, requires the RHR suction side isolation valves to have independent diverse Two of the RHR loops provide containment interlocks to prevent the valves from being spray cooling subsystems. Each subsystem opened unless the reactor coolant system (RCS) provides both wetwell and drywell spray cooling. pressure is below the RHR system design This subsystem provides steam condensation and pressure. While the ABWR RHR design does not primary containment atmospheric cooling following explicitly meet this requirement for diversity, a LOCA by pumping water from the suppression it does meet the intent of the requirement to pool, through the heat exchangers and into the provide high reliability against inadvertent wetwell and/or drywell spray spargers in the pri. opening of the valves. The pressure signal that mary containment. The drywell spray mode is provides the interlock function is supplied from initiated by operator action post LOCA in the 2 out of 4 logic, which has four independent presence of high drywell pressure.1he wetwell pressure sensor and transmitter inputs. Tbc spray mode is initiated by operator action. The independence is provided by cath being in a wetwell spray mode is terminated automatically by separate instrument division. Furthermore, the a LOCA signal. The drywell spray mode is termi- inboard and outboard valves of a common shutdown nated automatically as the injection valve starts cooling suction line are operated by different to open, which results from a LOCA and reactor electrical divisions, depresurization. Both wetwell and drywell spray modes can also be terminated by operator action. 5 . 4 . 7 . 1 . 1 .15 Fuel Pool Cooling

     $.4.7.1.1.7 Shutdown Cooling                           Two of the RHR loops provide supplemental fuel pool cooling during normal refueling The shutdown cooling subsystem is manually activities and any time the fuel pool heat load activated by the operator following insertion of exceeds the cooling capacity of the fuel pool the control reds and normal blowdown to the main   heat e4 changers. For normal refueling condenser, in this mode, the RHR removes           activities where the reactor well is flooded and residual heat (decay and sensible) from the        the fuel pool gates are open, water is drawn reactor vessel water at a rate sufficient to cool  from the reactor ?mtdown suction lines, pumped l it to 600C within 24 hours after the control         through the RHR 1. eat exchangers and discharged rods are inserted. The conditions are achieved 'hrough the reactor well distribution spargers.

for normal operation where all three RHR loops For 100% core removal, if necessary, water is are functioning together. The subsystem can drawn from the fuel pool cooling (FPC) system maintsin or reduce this temperature further so skimmer surge tanks, pumped through the RHR heat that the reactor can be refueled and serviced. exchangers and returned to the fuel via the FPC l l system cooling lines. These operations are l For emergency operation where one of the RHR initiated and shut down by operator action. loops has failed, the RHR system is capable of bringing the reactor to the cold shutdown 5.4.7.1.1.9 Reactor Well and Equipment Pool l condition of 1000C within 36 hours following Drain reactor shutdown. The RHR provides routing and connections for Reactor water is cooled by pumping it directly emptying the reactor well and dryer / separator from the reactor shutdown cooling nonics, pit equement oool to the suppression pool. through the RHR heat exchangers, and back to the Water is pep #d or drained by gravity through vessel via feedwater on one loop and via the low the FPC system return lines to the RHR shutdown l pressu(re flooder subsystem on the other suctiontwo lines and then to the radwaste or the loopst suppression pool. 1 l This subsystem is initiated and shut down by 5.4.7.1.1.10 AC Independent Water Addition operator action. l The AC independent water addition mode of The Branch Technical Position RSB 51, section the RHR provides a means for introducing water U 1.(b) and (c), of the RHR Standard Review Plan, from the fire protection system (FPS) directly Amendment 15 5 A-18

ABWR zwun ; Standard Plant nvc directly into the reactor pressure vessel to signal closes the RHR containment isolation (n) b the drywell spray header degraded plant conditions when AC power is not available from valves that are provided suction. Subsection 5.2.5 providesforan the shutdown cooling ]; explanation cither onsite or offsite sources. The RHR of the leak detection system and the isolation provides the piping and valves which connect the signals

see Subsection 5.2.5.2.1 (12) and Table FPS piping with the RHR loop C pump dischat 0. A6.

piping. The manual valves in this lir., permit , adding water from the FPS to the l'.rfR system if The RHR pumps are protected against damage the RHR is not operable. The primary means for from a closed discharge valve by means of supplying water through this connection is by use automatic minimum flow valves wh ch open on low of the diesel. driven pump in the FPS. A backup mainline flow and close on high maltline flow. to this pump is provided by a connection on the outside of the reactor building which allows 5.4.7.1J Design Basis for Pressure Reikf - hookup of the FPS to a fire truck pump. Capacity The vessel injection mode is intended to The relief valves in the RHR system are sired prevent core damage during statiot blackout after on one of three basis: P"IC has stopped operating, and to provide an in.vessui core melt prevention mechanism during a (I) thermalteliaf only, severe accident condition. If the AC. independent water addition mode is not actuated in time to (2) valve bypass icakage only, prevent core damage, core melting and vessel . failure, then it covers the corium in the lower (3) c, .rol valve failure and the subsequent

l. drywell when initiated and adds water to uncontrolled flow which results.

containment, thereby slowing the pressure rise. l l O ~ The drywell spray mode prevents high gas V temperatures in the upper drywel' and adds additional water to the containmeni, which increases the containment thermal mass and slows the pressurization rete. Additionally, the drywell spray provides fission product scrubbing to reduce fission product release in the event of failure of the drywell head. Operation of the AC. independent water addition mode is entirely manual. All of the valves which must be opened or closed during fire . vater L addition are located within the same ECCS valve room. The connection to add water using a fire i, truck pump is located outside the reactor l building at grade level. 1-5.4.7.1.2 Design Basis for Isolation of RHR System from Reactcr Cooiant Systene The low pressure portions of the RHR system o are isolated from full reactor pressure whenever the primary system pressure is above the RHR . l system design pressure. (See Subsection l 5.4.7.1.3 for further details.) In addition, automatic Isolation occurs for reasons of maintaining water inventory which are unrelated to line pressure rating. A low water level Amendment is 5+1s.1 l

ABWR m.mn Standard Plant uvc _' Redundant interlocks prevent opening valves discussed in Sections 3.5, 3.6, 3.7, a n d / to the low pressure suction piping when the Subsection 9.5.1 's reactor pressure is above the shutdown range. These same interlocks initiate valve closure on 5.4.7.2 Systems Design increaslag reactor pressure. 5.4.7.2.1 System Diagrams in addition, a high pressure check valve will close to prevent reverse flow if the pressure All of the components of the RHR syst:rn are should increase. Relief valves in the discharge shown in the P&lD (Figure 5.410). A piping are sized to account for leakage past the description of the controls and instrumentation check valve, is presented in Subsection 7.3.1.1.1 emergency core cooling systems control and instrumen-5.4.7.I.4 Design Basis With Respect to General tation. Design Criterion 5 Figurc 5.411 is the RHR process diagram and The RHR system for this unit does not share data. All of the r.izing modes of the system are equipment or structures with any other nuclear shown in the process data. The interlock block unit. diagram (IBD) for the RHR system is provided in Section 7.3. 5.4.7.1.5 Design Basis for Reliability and Operability Interlocks are provided to prevent: (1) drawing vessel water to the suppression pool, The design basis for the shutdown cooling (2) opening vessel suction valves above the mode of the RHR System is that this mode is suction lines or the discharge line design controlled by the operator from the control pressure, (3) inadvertent opening of drywell room. The only operations performed outs!de of spray valves during RHR operation where the l the control room for a normal shutdown are manual injection valve to the reactor is open and when i O)

 '       operation of local flushing water admission drywell pressure is not high enough to require valves, which are the means of providing cican the drywell spray for pressure reduction, and water to the shutdown portions of the RHR system.     (4) pump start when suction valve (s) are not open. A description of the RHR system logic Three separate shutdown cooling loops are (i.e., laterlocks, permissives) is presented in provided; and although the three loops are Table 5.4-3.

l required the reactorfor shutdown coolant under tonormal can be brought circumstances, 1000C in 5.4.7.2.2 Equipment and Component Description f.es than 36 hours with only two loops in operation. The RHR syste t is part of the ECCS (1) System Main pumps and therefore is required to be designed with redundancy, piping protection, power separation, The following are system performance require-etc., as required of such systems. (See Section ments the main pumps must satisfy. The pump 6.3 for an explanation of the design bases for equipment performance requirements include ECCS Systems.) additional margins so that the system perfor-r ance requirements can be achieved. These Shutdown suction and discharge valves are margins are standard GE equipment specifica-required to be powered from both offsite and tion practice and are included in procure-s'andby emergency power for purposes of isolation ment specifications for flow and pressure and shutdown following a loss of offsite power. measuring accuracy and foi power source fre-quency variation. 5.4.7.1.6 Design Basis for Protection from Physical Damage Number of Pumps 3 The design basis for protection frot.2 physical Pump type Centrifugal (g'} damage, such as internally generated missiles, pipe break, seismic effects, and fires, are Drive unit type M otor Amendmem 13 14 19

ABWR man Standard Plant ,_ RIV C 3 into the shutdown cooling mode below a Design flow rate 954 m /h nominal vessel pressure of 9.5 Total discharge head 125 m kg/cm2g, at design flow rate The RHR heat exchanger capacity is re-Maximum bypass flow 3 147.6 m /h quired to be sufficient to meet each of these functional requirements. The Minimum total 220 m max limiting function for the RHR heat ex-discharge head at 195 m min changer capacity is post LOCA contain-maximum bypass flow rate ment cooling. The heat exchanger capac-ity, K, is 88.45 kcal/0C.sec per Maximum runout flow 1130 m /h 3 bcat cxehanger. Maximum pump brake $50 kw The performance characteristics of the horsepower heat exchangers are shown in Table 5.4 4. Net positive suction 2.4 m head (NPSH) (3) Valves at 3.28 ft above the pump floor setting All of the directional vmives in the system are conventional gate, globe, and check Process fluid 10 to 182oC valves designed for nuclear service. The in-temperature range jection valves are high speed valves, as op-eration for RHR injection requires. Valve (2) Heat Exchangers pressure ratings are to provide the control or isolation funetion as necessary; i.e., The RHR heat exchangers have three major all vessel isolation valves are treated as functional requirements imposed upon them. Class 1 nuclear valves at the same pressure They are as follows: as the primary system. (a) Post LOCA Containment Cooling. The RHR (4) ECCS Portions of the RHR System limits the peak bulk suppression pool l temperature to less than 970C by The ECCS portions of the RHR system include direct pool cooling with two out of the those sections that inject water into the three divisions. reactor vessel. (b) Reactor Shutdown. The RHR removes The route includes suppression pool suction enough residt.al heat (decay and sen- strainers, suction piping, RHR pumps, dis-slble) from the reactor vessel water to charge piping, RHR heat exchangers, injec-l cool it to 600C within 24 hours after tion valves, and drywell piping into the the control rods are inserted. This vessel nozzles and core region of the mode shall be manually activated after a reactor vessel. blowdown to the main condenser reduces the reactor pressure to below 9.5 Pool cooling components include pool suction kg/cm2 g with all three divisions in strainers, piping, pumps, heat exchangers, operation, and pool return lines. (c) Safe Shutdown. The RHR brings the Containment spray components are the same as reactor to a cold shutdown condition of pool cooling components except that the l less than 1000C within 36 hours of spray headers replace the pool return lines, control rod insertion with two out of the three divisions in operation. The 5.4.7.2.3 Controls and instrumentation RHR is m a nu ally a cliva t e d Controls and instrumentation for the RHR system are described in Section 7.3. Amendment 15 5A20 i

ABWR m aman Standard Plant RTV C m The relief valves for the RHR system (Ell) are (7) In the absence of a valid LOCA signal with- [ ) listed in Table 5.4 5 and the operating character- out high drywell pressure and without the V istics of each valve (i.e., their relieving injection valve being fully closed, it is pressure) are tabulated. All of the E11 relief not possible to open the drywell spray valves in Table 5.4 5 are Quality Group B, Safety vnives in a loop when the corresponding Class 2, and Seismic Category I. All of the containment isolation valve in the same loop relief valves in Table 5.4 5 are standard is opent i.e., the two valves, in series, configurations meeting all applicable codes and are both not to be open during shutdown or standards. None of these valves are air operatad surveillance testing. nor can their setpoint be changed by the opcrators 5.4.7.2.3.2 Heat Exchanger leak Detection 5.4.7.2.3.1 Interlocks A radiation detector is provided in the main loop of the reactor building cooling water (RCW) (1) The valves requiring a keylock switch are system, which cools the secondary side of the F001 and F0148, C as indicated on the RHR RHR heat exchanger, if radioactive water from l P&lD, Figure 5.410. the primary side of the heat exchanger leaks to the secondary side, the radiation detector will (2) It is not possible to open the shut down signal a radiation increase soon after the RHR connection to the vessel in any given loop is started. Conformation is achieved through a whenever the pool suction, pool discharge sample port on the specific RHR pipe line of the l valve or wetwell spray valves are open in RCW system. the same loop to prevent draining the vessel to the pool. 5.4.7.2.4 Applic::ble Codes and Classifications (3) Redundant interlockt. prevent opening the (1) Piping, Pumps, and Valves p shutdown connections to and from the vessel V whenever the pressure is above the shutdown (a) Process side AShiEIII Class 1/2 range. Increasing pressure trip shall cause (b) Service water side closure of these valves. AShiE lll Class 3 l (4) A limer is provided in each pump minimum (2) Heat Exchangers flow valve control circuitry so that the pump has an opportunity to attain rated (a) Process side AShiE III Class 2 speed and flow before automatic control of TEhtA Class C the valve is activated. This time delay is l necessary to prevent a reactor water dump to (b) Service water the suppression pool during the shutdown side AShiElli Class 3 operation. TEhtA ClassC l (5) It is not possible to operate the RHR main (3) Electrical Portions I pumpr. without an open suction source because these pumps are used for core, vessel and (a) IEEE 279 containment cooling and their integrity must be preserved. (b) lEEE 308 (6) Redundant interlocks prevent opening and 5.4.7.2.5 Reliability Considerations automatically closes the shutdown suction ennnections to the vessel in any given loop The RHR system has included the redundancy w;enever a low reactor level signal is requirements of Subsection 5.4.7.1.5. Three present, completely redundant loops have been provided to p remove residual heat, each powered from a Amendmera 15 $441

ABWR warn Standard Plant RIN C separate emergency bus. All mechanical and electrical components are separate. Two out of three are capable of shutting down the reactor - within a reasonable length of time. ; 5.4.7.2.6 Manual Action (1) Emergency Mode [ Low pressure flooder (LPFL) mode] O O Amendment 15 5.4 21.1

ABWR mama . Standard Plant arv c Each loop in the subsystem is initiated (4) Wetwell and Drywell Spray Cooling (h automatically by a low water level in the V reactor vessel or high pressure in the Two of the RHR loops provide containment drywell. Each loop in the system can also spray cooling subsystems. Each subsystem be placed in operation by means of a Manual provides both wetwell and drywell spray Initiation pushbutton switch, cooling. This subsystem provides steam condensation and primary containment During the LPFL mode, water is initially atmospheric cooling following a LOCA by pumped from the suppression pool and pumping water from the suppression pool, diutted through the minimum flow lines through the heat exchangers and into the until the injection valve in the discharge wetwell and/or drywell spray spargers in the line is signalled to open on low reactor primary containment. The drywell spray mode , pressure. As the injection valve opens on is initiated by manual operator action low reactor pressure, flow to the RPV comes post LOCA in the presence of high drywell from the suppression pool, through the RHR pressure. The drywell spray mode is heat exchanger, and the injection valve. Initiated as required by manual operator This creates a flow signal that closes the asuon. The drywell spray mode is minimum flow line. terminated automatically following a LOCA signal as the injection valve opens, and the The system remains in the operating mode wetwell spray is terminated automatically by until manually stopped by the operator, a LOCA signal. Both drywell and wetwell spray can be terminated manually by operator (2) Test Mode action with no permissive interlocks to be satisfied, i Full flow functional testir.g of the RHR can be performed during normal plant operation (5) Shutdown Cooling e or during plan shutdow n by manual opcration i (' of the RHR from tb., controt room. For plant The shutdown cooling subsystem is manually testing during nr. mal plan'. operation, the activated by the operator following pump is started and the return line to the insertion of the control rods and normal suppresslor pool is opened. A reverse blowdown to the main condenser. In this sequence is used to terminate this test, mode, the RHR removes residual heat (decay Upon Nccipt of an automatic initiation and sensible) from the reactor vessel water signe' while in the Dow testing taode, the at a rate sufficient to cool it to 600C syr.cm is returned tu automatic control. within 24 hours after the control rods are inserted. The subsystem can maintain or (3) Suppression Pool Cooling reduce this temperature further so that the reactor can be refueled and serviced. The suppression pool cooling subsystem, which is initiated and stopped manually, Rector water is cooled by pumping it provides means to remove heat released into directt' from the reactor shutdown cooling the suppression pool, as necessary, nozzles, through the RHR heat exchangers, following heat additions to the pool, and back to the vessel (via feedwater on one During this mode of operation, water is loop and via the LPFL subsystem on the other pumped from the suppression pool through the two loops). RHR heat exchangers, and back to the suppression pool. This subsystern is This system is initiated and shut down by initiated manucily, manual operator action. l J(%. l Amendment l$ $ 4-22

ABWR msimn Standard Plant uvc (6) Fuel Pool Cooling . Two of the RilR loops provide so '!ccuental fuel pool cooling during norri.. meling activities and any time the fuel pool heat load exceeds the cooling capacity of the fuel pool heat exchangers. For normal refueling activities where the reactor well is flooded and the fuel pool gates are op:n, water is drawn from the reactor shutdown suction lines, pumped through the RHR heat exchangers and discharged through the reactor well distribution spargers. For 100% core removal, if necessary, water is drawn from the fuel pool cooling (FPC) system skimmer surge tanks, pumped through the RilR heat exchangers and returned to the l fuel pool via the FPC system cooling lines. Ther,c operations are initiated and shut down by operator action. O i i O Amendment 15 ' 5.4-22.1

ABWR zwan Standard Plant RIV C (7) Reactor Well and Equipment Pool Drain (5) system flushing time. Since the exchangers are designed for the fouled condition with O) ! The RHR provides routing and connections for emptying the reactor well and equipment pool relatively high service water temperature, the units have excess capability to cool when first to the suppression pool after servicing, cut in at high vessel temperature. Total flow Water is pumped or drained by gravity and mix tempergture must be controlled to avoid j through the FPC system return lines to the exceeding a $$ C/ hour cooldown rate. See l RiiR shutdown suction lines and then to the Subsection 5.4.7.1.1.7 for minimum shutdown time suppression pool, to reach 100 C. l (8) AC-Independent Water Addition 5.4.7J.2 Worst Case Transient The RIIR is provided with piping and valves Several limiting events were considered for which connect the RHR loop C pump discharge RilR heat exchanger sizing. Those events were: piping to the fire protection system (FPS) and to a reaetor building external fitc (1) fudwater line break (fvLB) truck pump hookup. These connections allow for addition of FPS water b the reactor (2) main steam line break pressere vessel, or the drywell s > ray header or wetwell spray header during events when (3) inadverient opening of a relief valve AC power is unavailable from both onsite and offsite sources. Operation of the RiiR in (4) normal shutdown the AC independent wser addition mode is entirely manual. All valves required to be (5) emergencyshutdown opened or closed for operation are located within the same loop C ECCS valve room to (6) ATWS provide case of operation. l It was determined for post LOCA suppression l \ 5.4.7.3 Perfor-nance Evaluation pool temperature control, that the FWLB is the l most limiting event. The worst case conditions RilR system performance depends on sizing its for the event assumes one R11R heat exchanger heat exchanger and pumping flow rate failure instead of one diesel generator characteristics with enough capacity to satisfy failure. When one heat exchanger fails, the the most limiting events. The worst case heat generated by the pump is still added to the transient established the heat excganger size, containment, and also one additional pump flow l given the pumping flow of 954 m /h for each carries the reactor decay heat more effectively RilR loop. The shutdown cooling mode requirements to the suppression pool. Therefore, a single were satisfied within the RilR characteristics failure of a RilR heat exchanger is the most established by the worst case transient, limiting single failure. 5.4.7.3.1 Shutdon With All Components The heat exchanger size was established to Available limit the suppression pool peak temperature to 97 C. This is acceptable to the ABWR for the l A typical curve is not included to show following reasons: vessel cooldown temperatures versus time because of the infinite variety of such curves that is (1) The ABWR wetwell pressure becomes high, possible due to: (1) clean steam systems that high enough to provide more than 11 C may allow the main condenser to be used as the subcooling with 97*C pool temperature heat sink when nuclear steam pressure is when the peak pool temperature occurs. Insufficient to maintain steam air ejector performance; (2) the condition of fouling of the (2) Because it takes 4 to 6 hours to reach the exchangers; (3) operator use of one or two peak pool temperature, shutdown cooling will J cooling loops; (4) coolant water temperature; and be initiated before the peak pool temp. 1 Amendment 15 5423 j 1 1 l

ABWR mamn SAndard Plant REV C erature. The energy release from the reactor will be controlled by the shutdown cooling system, and there is no need to release the reactor energy to the pool. Normal shutdown cooling is a nonsafety-

 $.4.7.3.3 Emeryncy Shutdown Cooling                related event and is therefore analyzed asmming that all three RHR loops are operational.

The design requirements for ABWR emergency shutdown cooling capability are specified in The design heat exchanger capacity is Regulatory Guide 1.139, as follows: sufficient to meet the normal shutdown cooling criteria. The reactor shutdown cooling system (SDCS) should be capable of bringing the reactor to 5.4.7.4 Pre operationel Testing a cold shutdown condition within 36 hours following reactor shutdown with only offsite The pre operational test program and startup power or onsite power available assuming the tests program discussed in Chapter 14 are used most limiting single failure, to generate data to verify the operational capabilities of each piece of equipment in the The limiting condition is for the case with system: each instrument, each set point, each loss of offsite power which would disable the logic element, each pump, each heat exchanger, forced circulation. The most limiting single each valve, and each limit switch. In addition, f ailure is the loss of one RiiR division these programs verify the capabilities of the (designated as N 1 case). Therefore, for the system to provide the flows, pressures, cooldown emergency shutdown cooling purpose, one of the rates, and reaction times required to perform ' bases of RHR heat exchanger sizing is to meet the all system functions as specified for the system following requirements: or component in the system data sheets and process data. The ABWR RilR in shutdown cooling mode should be capable of bringing the reactor to Logic elements are tested electrically; l cold shutdown conditions (1000C) within 36 valves, pumps, controllers, and relief valves hours following reactor shutdown for N 1 are tested mechanically. Finally, the system is case, witti only onsite power available, tested for total system performance against the design requirements using both the offsite power The ABWR selected design configuration meets and standby emergency power Preliminary heat all design requirements and is consistent with exchanger performance can be evaluated by the heat exchanger sire required for post LOCA operating in the pool cooling mode, but a vest,el pool temperature control, shutdown is required for the final check due to the small temperature differences available wD j 5.4.7.3.4 Normal Shutdown Cooling pool cooling. " After a normal blawdown to the main 5.4.8 Reacto AYater Cleanup Spam condenser, the shutdown cooling subsystem is activated, in this mode of operation the RilR The reactor water cleanup (CUW) system is l shall be capable of removing enough residual heat classified as a primary power generation system, decay and sensible) from the reactor vessel a part of which forms a portion of the reactor l w(he control rods are inserted, t ater to cool it to 600C within 24 hours after coolant prer,sure boundary (RCPB). The remaining portion of the system is not part of the RCPB because it can be isolated from the reactor. The CUW system may be operated at any time l durirg normal reactor operations. 8 Amendment 15 5421 1

ABM 2 m ioo w Standard Plant uv. c 5.4.5.1 line of the RHR loop *B" The cooled effluent p V l I The CUW system: of the NRHXs goes through the CUW pumps to the Iwo filter demineralizers for eleanup. CUW system discharge is split to feedwater lines 'A' (1) removes solid and dissolved impusities from and 'B' The system P&lD is provided in Figure the reactor coolant and measures the reactor 5.4 12 water conductivity in accordance w th Regulatory Guide 1.56,' Maintenance of Water The total capacity of the system, as shown on Purity in Boiling Water Reactors *; the process flow diagram in Figure 5.713 is equivalcut to 2% of rated feedwater flow. Each (2) provides containment isolation that places pump, NRHX, and fliter demineralizer is capable I the major portion of the CUW system outside of 50% system capacity operation, with the one l the RCPB, limiting the potential for RHX capable 100% system capacity operation, significant release of radioactivity from the primary system to the secondary The operating temperature of the containment; filter demineralizer units is limited by the ion exchange resins; therefore, the reactor coolant (3) discharge excess reactor water during must be cooled before being processed in the startup, shutdown, and hot standby filter demineralizer units. The regenerative conditions to the main condenser or radwaste heat exchanger transfers heat from the tubeside or suppression pool; (hot process inlet) to the shellside (cold

                                                                . process return). The shellside flow returns to (d) provides full system flow to the RPV head the reactor. The non regenerative heat spray as required for rapid RPV cooldown and exchanger cools the process further by rapid refueling; and                              transferring heat to the reactor building cooling water system, f~       (5) minimizes RPV temperature gradients by

( maintaining circulation in the bottom head The filter demineralizer units are pressure

  '              of the RPV during periods when the reactor        precoat type filters using powdered ion exchange internal pumps are unavailable,                   resins. Spent resins are not regerable and are sluiced from the filter demineralizer unit to a l     The CUW system is automatically removed from backwash receiving tack from which they are service upon SLCS actuation. This isolation transferred to the radwaste system for prevents the standby liqulJ reactivity control processing and disposal. To prevent resins from material from being removed from the reactor entering the reactor in tle event of failure of water by the cleanup system. The design of the a filter demineralizer resin support, a strainer CUW system is in accordance with Regulatory Guide        is installed on the filter demineralizer unit.

1.26 and Regulatory Guide 1.29 Each straiact and filter demineralizer vressel has a control room alarm that is energized by 5.4.8.2 System Description high differential pressure. Upon further increase in differentail pressure from the alarm l The CUW is a closed loop system of piping, point, the fitter demineralizer will l l circulation pumps, a regenerativeautomatically heat exchanger, isolate. nf n regenerative heat exchangers, reactor water pressure boundary isolation valves, a reactor The backwash and precoat cycle for a water sampling station, (part of the sampling filter demincializer unit is automatic to system) and two precoated filter demineralizers. minimize the need for operator intervention. , During blowdown of reactor water swell, the loop The filter demineralizer piping configuration is 1- l Is open to the radwaste or suppression pool. The complete and crud traps are eliminated. A single loop has two parallel pumps taking common bypass line is provided around the filter-suction through a rea.enerative heat exchanger demineralizer units. (RHX) and two parallel non regenerative heat exchangers (NRHX) from both the single bottom O'u/ l head drain line and the shutdown cooling suction j Amendment 15 5 4 2$

_. . - .-~ -_ -- . ABWR mma l Standard Plant RIV C ' In the event of low flow or loss of flow in demineralizer compartment is normally permitted the system, the precoat is maintained on the only after removal of the precoat. Penetrations septa by a holding pump. Sample points are through compartment walls shall be located so as provided in the common influent header and in not to compromise radiation shielding each effluent line of the filter demineralizer requirements. Primarily, this affects nozzle units for continuous indication and recording of locations on tanks so that wall penetrations do system conductivity. High conductivity is not 'see' the tanks. Generally, this means anneciated in the control room. The influent piping through compartment walls should be sample point is also used as the normal source of above, below, or to the side of reactor coolant grab samples. Sample analysis filter demineralizer units. The local control also indicates the effectiveness of the panel shall be outside the vessel compartment filter demineralizer units, and process valve cell, located convenient to the RWCU system. The tank which receives } l The suction line (RCPB portion) of the CUW backwash shall be locsted in a separate shielded system contains two motor operated isolation room belcw the filter demineralizer units, valves which automatically close in response to signals from the leak detection and isolation The filter demineralizer vents are piped to system, actuation of the standby liquid control the backwash receiving tank. Piping vents and l system, and high filter demineralizer inlet drains are directed to low conductivity temperature. Subsection 7.3.1.1.2 describes the collection in radwaste. System pressure relief l leak detection and isolation system setpoints valves are piped to radwaste. Refer to Figure l that are summarized in Tables 5.2 6 and 5.2 7. 5.412 for the exact configuration. This isolation prevents loss of reactor coolant I and release of radioactive saaterial from th: A remote, manually operated gate valve on the j reactor, prevents removal of liquid reactivity return line to the feedwater lines in the steam control material by the cleanup system should the tunnel provides long term leakage control. SLCS be in operation, and prevents exceeding the Instantaneous reverse flow isolation is provided l design temperature of the CUW and the by cher' valves in the CUW piping. l filter demineraliter resins. The RCPB isolation valves may be remote manually operated to isolate CUW system operation is controlled from the l the system equipment for maintenance or main control room. Filter demineralizing servicing. Discussion of the RCPB is provided in operations, which include backwashing and Section 5.2. precoating, are controlled automatically from a process controller or manually from a local Each filter demineralizer vessel shall be panei. Section 7.7 provides an interlock block , installed in an individual shielded compartment, diagram, l The compartments shall not require accessibility I during operation of the filter demineralizer 5A.8.3 System Evaluation unit. Shiciding is required due to the concentration of radioactive products in the The CUW system, in conjunction with the l filter.demineralizer process system. Service condensate treatment system and the fuel pool l space shall be provided the filter demineralizer cooling and cleanup system, maintains reacNr for septa removal. All inlet, outlet, vent, water quality during all reactor operating modes drain, and other process valves shall be located (normal, hot standby, startup, shutdown, and outside the filter demineralizer compartment in a refueling). separate shielded area together with the i necessary piping, strainers, holding pumps and The CUW system has process interfaces with I instrument elements. Process equipment and the RHR, control rod drive, nuclear boiler, , controls shall be arranged so that all normal radwaste, fuel pool cooling and cleanup (FPC), operations are conducted at the panel from reactor building cooling water systems, RP\ , and outside the vessel or valve and pump cc upartment suppression pool. The CUW suction is from the shiciding walls. Access to the filter. RHR *B' shutdown suction line and the RPV bottom head drain. The CUW system main process pump l Amendment 15 5.4-26 I 1

_ .~ _ _ _ _ _ . - _ _ _ _ _ _ ,_ __ MM 2wtoaAn Standard Plant arv c , motor cavities are purged by water from the the high pressure back valves is designed to [,) l control rod drive system. CUW system return flow Ouality Group D. V is directed to either the nuclear boiler system (feedwater lines), directly to the RPV through A tabulation of CUW system equipment data, l the RPV bead spray, suppressioe pool or radwaste includ ing temperature pressure and flow l through the CUW dump lii . . CUW filter. capacity is provided in Table 5.4 6. deminerallrer backwash is to the backwash receiving tank (BWRT) located in the FPC (BWRT 5.4.9 Main Steamilnes and Feedwater Piping accommodates backwash from the CUW the FPC, and the suppression pool cleanup system). The 5.4.9.1 Safety Design Bases non regenerative heat exchanger is cooled by the reactor building cooling vater system. Other in order to satisfy the safety design bases, utility or support interfaces exist with the the main steam and feedwater lines are designed instrument air system and the condensate and as follows: plant air systems for the filter deminerallrer backwash. (1) The main steam, feedwater, and associated dra lines are protected from potential The type of pressure precoat cleanup system d% due to fluid jets, missiles, reaction used in this system was first put into operation forces, pressures, .nd temperatures resulting in 1971 and has been in use in all BWR plants from pipe breaks. Operating plant l brought experience on has line shownsince that thethen. CUW system, de- (2) The main steam, feedwater, and drain lines signed in accordance with these criteria, are designed to accot. aodate stresses from provides the required BWR water quality. The internal pressures and earthquake loads l ABWR CUW system capacity has been increased to a without a failure that could lead to the nominal of 2% of rated feedwater from the release of radioactivity in excess of the O original 1% sire. This added capacity provides guideline values in published regulations. d additional margin agt. inst primary system intrusions and component availability. The (3) The main steam and feedwater lines are acces-nonregenerative heat exchanger is sired to sible for inservice testing and inspection. maintain the required process temperature for 100% system flow. During periods of water (4) The main steamlines are analyzed for dynamic rejection to the suppression pool or radwaste, loadings due to fast closure of the turbine , CUW system flow may be reduced slightly to stop valves. compensate for the loss of cooling flow through the RPV return side of the regenerative heat (5) The main steam and feedwater piping from the ex6 anger, reactor through the seismic interface

      .l       Th'c CUW system is classified as a nonsafety system. .The reactor isolation valves are          (6) The main steam and feedwater piping and t-         . classified as important to safety. System piping       smaller connected lines are designed in l           and components within the drywell, including the        accordance with the requnements of Table                     e l           suction piping up to and including the outboard         3.2 1.

suction isolation valve, and all containment isolation valve including interconnecting piping 5.4.9.2 Power Generation Design Bases assembly, are Seismic Category I, Quality Group L l A. The three flow elements that are used for CUW (1) The main stearnlines are designed to conduct system leak detection meet Seismic Category I and steam from the reactor vessel over the full Quality Group A requirements so as to maintain range of reactor power operation, structural integrity during a faulted condition. i All other non safety equipment is designed as Nonseismic, Quality Group C. Low pressure piping in the backwash and precoat area downstream of Amendment 13 5427

                                                                                                                                 )
                                                                                                                             -e

ABWR 21A6100All Standard Plant nry c (2) The feedwater lines are designed 9 conduct The materials used in the piping are in water to the reactor vessel over the full accordance with the anplicable design code and range of reactor power operation, supplementary requirements described in Section 3.2. The valve between the outboard isolation 5.4.9.3 Description valve and the shutoff valve upstream of the RilR entry to the feedwater line is to effect a The main steam piping is described in Section closed loop outside containment (CLOC) for 10.3. The main steam and feedwater piping from containment bypass leakage control (Subsections the reactor through the containment isolation 6.2.6 and 6.5.3). interfaces is diagrammed in Figure 5.13. The general requirements of the feedwater As discussed in Table 3.21 and shown in system are described in Subsections 7.1.1.7, Figurc $.13, the main steamlines are Quality 7.7.1.4, 7.7.2.4, and 10.4.7. l Group A from the reactor vessel out to and includ-ing the outboard MSIV and Quality Group B from 5.4.9.4 Safety Evaluation . the outboard MSIVs to the turbine stop valve. I They are also Seismic Category I only from the Differential pressure on reactor internals l reactor pressure vessel out to the scismic inter- under the atsumed accident condition of a rup- , face restraint, tured steamilne is limited by the use of flow l restrictors and by the use of four main steam-The feedwater piping consists of two $$0 A lines. All main steam and feedwater piping will diameter lines from the feedwater supply header be designed in accordance with the requirements to the reactor. Isolation of each line it, defined in Section 3.2. Design of the piping in accomplished by two containment isolation valves accordance with these requirements ensures consisting of one check valve inside the drywell meeting the safety design bases, and one positive closing check valve outside containmec4 (Figurc 5.13) Also included in 5.4.9.5 Inspection and Testing this portion of the line is a manual maintenance l valve (F005) between the inboard isolation valve Testing is carried out in accordance with and the reactor nozzle. The design temperature Subsection 3.9.6 and Chapter 14. Inservice and pressure of the feedwater line is the same as inspection is considered in the design of the that of the reactor inlet nozzle (i.e., 87.9 main steam and feedwater piping. This consider-kg/cm 2g and 3020C). ation assures adequate working space and access for the inspection of selected components. The feedwater piping upstream of the second isolation valve contains a remote, manual, 5.4.10 Pressurizer motor operated gate valve and upstream of the s gate valve, a seismic interface restraint. The Not Applicable to BWR Si " outboard isolation valve and the seismic inter-face restraint provide a quality group transi- 5.4.11 Pressurizer Relief Discharge System tional point in the feedwater lines. Not Applicable to BWR As discussed in Table 3.21 and shown in l Figure 5.13 the feedwater piping 16 Quality 5.4.12 Valves Group A from the reactor pressure vessel out to and including the outboard isolation valve, 5.4.12.1 Safety Design Bases g Quality Group B from the outboard isolation valve

to and including the seismic interface restraint, Line valves, such as gate, globe, and check and Quality Group D beyond the shutoff valve.

The feedwater piping and all conr, cered piping of l 65A or larger nominal site is Scismic Category I only from the reactor pressure vessel aut to and including the seismic inter face restraint. Amendent 15 5 4-28

ABWR 2WlooAn Saridard Plant RPV Q

  ,n   valves, are located in the fluid systems to       normal plant operation may be partially

(*) perform a mechanical function. Valves are components of the system pressure boundary and, exercised during this period to assure their operability at the time of an emergency or having moving parts, are designed to operate faulted condition. Other valves, serving as a efficiently to maintain the integrity of this system block or throttling valves, may be boundary. exercised when appropriate. The valve. e de under the internal Leakage from critical valves steam is pressure /tempuie loading as well as the monitored by use of double packed stuffing boxes external loading experienced during the various with an intermediate lantern leakoff connection system transient operating conditions. The for detection and measurement of leakage rates. design criteria, the design loading, and acceptability criteria are as specified in Motors used with valve actuators will be Subsection 3.9.3 for ASME Class 1,2, and 3 furnished in accordance with applicaHe industry valves. Compliance with ASME Code is discussed standards. Each motor actuator will be assem. In Subsection 5.2.1. bled, factory tested, and adjusted on the valve for proper operation, position, torque switch 5.4.12.2 Description setting, position transmitter function (where applicabic), and speed requirements. Valves Line valves are manufactured standard types will be to demonstrate ad:quate stem thrust (or designed and constructed in accordance with the torque) capability to open or close the valve requirements of ASME Code Section ill for Class within the specified time at specified differ-1,2, and 3 valves. All materials, exclusive of ential pressure. Tests will verify no seals, packing, and wearing components, shall mechanicci damage to valve components during endure the 60 yes.r plant life under the full stroking of the velve. Suppliers will be environmental conditions applicable to the required to furnish assurance of acceptability g) particular system when appropriate maintenance is (V periodically performed. of equipment for the intended service based on any combination of: Power operators will be sired to operate (1) test stand data, successfully under the maximum differeotial pressure determined in the design specification. (2) prior field performance, 5.4.*2.3 Safety Evaluation (3) prototype testing, and Line valves will be shop tested by the (4) engineering analysis, manufacturer for performability. Pressure - retaining parts are subject to the testing and Prc. operational and operational testing examination requirements of Section 111 of the performed on the installed valves consists of ASME Code. To minimize internal and external total circuit checkout and performance tests to leakage past seating surfaces, maximum allowable verify specd requirements at specificd leakage rates are stated in the design differential pressure. specifications for both back seat as well as the l main seat for gate and globe valves. 5A.13 Safety / Relief Valves Valve construction materials are compatible The reactor component and subsystem SRVs are with the maximum anticipated radiation dosage for listed in Table 5.4 5. The RHR relief valves the service life of the valves, ate discussed separately in Subsection 5.4.7.1.3. 5.4.12.4 Inspection and Testing 5.4.13.1 Safety Design liases (~} Valves serving as containment isolation V valves which must terr.ain closed or open during Overpressure protection is provided at Amendment 7 5429

ABWR muun i Standard Plant myc l Support types and materials used for lisolatable SLC, RilR, HPCF,portions of the and RCIC. The following systems: relief valves fabricated support elements are to conform with 1 will be selected in accordance with the rules set Sections NF 2000 and NF 3000 of ASME Code forth in the ASME Code Section III, Class 1,2, Section 111. Pipe support spacing guidelines of and 3 components. Other applicable sections of Table 121.1.4 of ANSI B31.1, Power Piping Code, the ASME Code, as well as ANSI, API, and ASTM are to be followed. Codes, will be followed. 5.4.14.2 Description 5.4.13 2 Description The use and the location of rigid type Pressure relief valves have been designed and supports, variable or constant spring type constructed in accordance with the same code supports, snubbers, and anchors or guides are to class as that of the line valves in the system, be determined by flexibility and seismic / dynamic

                                          .                  stress analyses. Component support elements are Table 3.21 lists the applicable code classes      manufacturer standard items. Direct weldment to for valves. The design criteria, design loading, thin wall pipe is to be avoided where possible, and design procedure are described in Subsection 3.9.3.                                                5.4.14J Safety Evaluation 5.4.13J Safety Evaluation                                The flexibility and seismic / dynamic analyses are to be performed for the design of adequate The use of pressure relieving devices will component support systems including all tran- l assure that over pressure will not exceed 10% sient loading conditions r:xpected by each above the design pressure of the system. The component. Provisions are to be made to prmide number of pressure relieving devices on a system spring type supports for the initial dead weight or portion of a system has been determined on loading due to hydrostatic testing of steam this basis,                                           systems to prevent damage to this type support.

5.4.13.4 (Deleted) 5.4.14.4 Inspection and Testing L After completion of the installation of a ' L support system, all hanger elements are to bc visually examined to assure that they are in correct adjustment to their cold setting position. Upon hot start up operations, thermal growth will be observed to confirm that spring type hangers will function properly 5.4.14 Component Supports between their hot and cold setting positions. Final adjustment capability is provided on all Support elements are provided for those hanger or support types. Weld inspections and components included in the RCPB and the connected standards are to be in accordance with ASME Code systems. Section III. Welder qualifications and welding procedures are in accordance with ASME Code 5.4.14.1 Safety Design Bases Section IX and NF 4300 of ASME Code Section Ill. Design loading combinations, design 5.4.15 References i procedures, and acceptability criteria are as described in Subsection 3.9.3. Flexibility 1. Design and rcrformance of Genera / Electric calculations and selsmic analysis for Class 1,2, Bolling Water Reactor Main Strom Line and 3 components are to be confirmed with the Isolation l'alves, General Electric Co., appropriate requirements of ASME Code Section Atomic Power Equipment Department, March III. 1%9 (APED 5750). O j Amendment 15 3.4 30 L l 1 l=

I ABWR maimo ; Standard Plant Rry c ( Table 5.41 d REACTOR RECIRCULATION SYSTEM DESIGN CHARACTERISTICS Number of Reactor Internal Pumps (RIP) and Heat Exchangers- 10 RIP Motor Housing and lleat Exchanger Shell Internal Design Pressure / Design Temperature - 87.9 kg/cm2gj302oC RIP Motor Heat Exchanger Tube Design Pressure / Temperature 87.9 kg/cm2 g/3020C external and 14 kg/cm2 g/7(PC laternal i Single RIP Parameters at Rated Reactor Power and Rated Core Flow given below: 10 RIPS 9 RIPS l' Ump Onerating Oneratine 3 flow (m /h) 6,912 8,291 Flow (106 kg/h) 5.22 6.26 Total Developed Head (m) 32.6 35.8 Suetion Pressure (kg/cm2abs) 73.9 73.9 [ ,) V Required NPSil (m) 5.6 7.0 Avallable NPSil (m) 134 134 Water Temperature (maxOC) 278 278 4 Pump Brake llorsepower (BilP) ~ 791 -1042 Mntor Motor Type Wet induction Rated Speed (rpm) -1350 -1500 Minimum Speed (rpm) 450 450 Motor input Power (kw) 681 893 Phase 3 3 Frequency (hz) variable 0 25 0 25 Rotationalinertia (kg m2 ) 19.5-24.0 19.5 24.0 . r Rated Voltage 3.3 kV 3.3kV Amendment 15 5A 31

ABWR m6mn Standard Plant Rev.c TABLE 5.4.la NET POSITIVE SUCTION HEAD (NPSH) AVAILABLE TO RCIC PUMPS A. Suppression pool is at its minimum depth, El. 3740mm. B. Centerline of pump suction is at El. 7200mm. C. Suppression pool water is at its maximum temperatue for the given operating m ode, 77'C. I D. Pressur is atmospheric above the suppression pool. l E. Maximum suction strainct losses are 1.46 m (50% plugged) l NPSil = IIATM + IIS *IIVAP 'IIF where:

                             =   atmospheric head liAThi ils        =   static head                                                       3 l                                                                                                   Y liVAp      =   sapor pressure head lip       =    Frictional head including strainer l                  Minimum Exnecled NPS11
   !              Maximum suppression pool temperature is 77'C
                            =    10.73m IIAThi lis       =   3.46m
                            =   4.22m l                  HVAP lip       =   1.82m Strainer head loss = 1.46m l

NPSil available = 10.73 + 3.46 4.22 1.82 - 8.15m NPSil required = 7.3m

NPSil Reference Point O

j Amendment 13 S.4 31.1 l

ABWR zwimin samadard Plant uv c Table 5.4 2 k DESIGN PARAME1TRS FOR RCIC hTSTEM COMPONEh"I3 (1) RCIC Pump Operation (CD01) Flow rate lajealon flow 182 m3/h 3 Cooling water flow 4 to 6 m gh Totalpumpdischarge 188 m /h (includes no margin for pump wear) Water temperature range 10 to 600C, continuous duty 60 to 770C, short duty NPSH 7.3m minimum Developed head ome -t *.0.S kg/cm 2abs reactor pressure # 186 m at 11.6 kg/cm2abs reactor pressure Maximum pump 675 kw at 900 m developed shaft power head ' f 125 kw at 186 m developed , s head > Design pressure 120 kg/cm2g (2) RCIC Turbine Operation (CX102) 7 liigh Pressure low Presure Condition Condition Reactor Pressure 83.5 kg/cm 2abs 11.6 kg/cm2abs (saturated temperature) Steam inlet pressure 82.8 kg/cm2abs, 10.5 kg/cm2 hs, minimum minimum Turbine exhaust 1.1 to 1.8 kg/cm 2abs, 1.1 to 1.8 kg/cm2abs, pressure maxin. 2m maximum Design inlet pressure 87.9 kg/cm2g at saturated temperature i. Amendment 15 5 4 .

l ABM 2:462=48 l Sinadard Plant BEY _.C

Table 5.4 2

-.)

f ;

    '\../                             DESIGN PARAMETERS FOR RCIC SYSTEM COMPOSThTS (COSTINUED)

Design exhaust 10kg/cm2g at pressure saturated temperature (3) RCIC leak off orifices Sized for 3.2 mm diameter minimum (D017, D018) to 4.8 mm diameter maximum Flow element (FE007) Flow at full meter 3 250 m /h differential pressure Normal temperature 10 to 770C System Design 87.9 kg/cm2g/3020C pressure /temperatute Maximum unrecoverable 032 kg/cm2 loss at normal flow Installed combined accuracy 12.5% at normal flow and normal (Flow element, Flow transmitter and Flow ' indicator) (4) Vabn Operation Requirements Steam supply valve (F037) Open and/or close against full differential pressure of 82.8 kg/cm 2within 15 seconds

   . f) .
   ' '\)

Amendment 15 5.4-33

ABM 2346ioarn Standard Plant RIY C Table 5.4 2 DESIGN PAIMMETERS FOR RCIC SYSTEM COMPONENTS (CONTINUED) Pump discharge valve Open and/or cloe gainst full (R04) differential prec. : of 98.4 kg/cm2 within 15 seconds. Pump minimum flow by- Open and/or close against full pass valve (F011) differential pressure of 98.4 kg/cm2 within 5 seconds l RCIC steam is>>lation Open and/or close against full l vaht (F035&0036) differential pressure of 82.8 kg/cm2 l within 30 seconds l Cooling water ;,ressure Self contained dowstream sensing I controlvaht (FU13) control valve capable of maintain- 1 ing constant dowstream pressure of 5.27 kg/cm2 2 O Pump suction re!!cf 3 15.1 kg/cm abs setting; 23 m /h at 10%

valve (R)l7) accumulation Cooling water relief Sired to prevent over pressurizing valve (F030) piping, valves, and equipment in the coolant loop in the event of failure of pressure control valve F013 Pump test return vahe Capable of throttling control (F008) against differential pressures up to 773 kg/cm 2and closure against differential pressure at 98.4 kg/cm2 I

i l l l O A.nendment 15 5.4 M

                                                  -_       -     . .     - -        . ~      . _ _ .-.

1 ABM ns6ioarn Standard Plant arv c Table 5.4 2 (]o DESIGN PARAMETTRS FOR RCIC SYSTEM COMPONEh*TS l (CONTINUED) Pump suction valve, Capable of opening and closing suppression pool against 14 kg/cm2 differential (F006) pressure Testable check valve Open and/or close against full equvizing valve (F026) differential pressure of 82.8 kg/cm2 Outboard check valve Accessible during plant operation (F005) and capable oflocal testing Turbine exhaust isola- Opens and/or closes against 11.2 kg/cm2 tion valve (F039) differential pressure at a tempera. ture of 1700C, physicallylocated in the line on a horizontal run as close to the containment as practical Isolation valve, steam Opens and/or closes against warmup line (F048) differential pressure of 82.8 kg/cm2, Barometric Condenser These valves operate only when RCIC Condensate Drain 1.ine System is shutdown allowing drain-Isolation Valves age to CRW system and they must . (F031 & F032) operate against a differential pressure of 5.3 kg/cm2 Condensate storage tank This valw isolates the condensate isolation valve (F001) storage tank so that suction may be drawn from the suppression pool; valve must operate against a differential pressure of 14 kg/cm2 O Amendment t$ 5.4 35

ABM 2346imin Standard Plant REV.C Table 5.4 2 g DESIGN PARAMETERS FOR RCIC SYSTEM COMPONENTS (CONTINUED) Vacuum breaker check Full flow and open with a minimum pressure ' valves (F054 & F055) drop (less than 0.04 kg/cm 2across the valves) Steam inlet drain pot These valves allow for drainage of system isolation the steam inlet drain pot and must (FD40 & F041) operate against a differential pressure of 82.8 kg/cm2 Steam inlet trip This valve bypasses the trap D008 bypass valve (F058) and must operate against a differ-ential pressure of 82.8 kg/cm2 Cooling loop shutoff This valve allows water to be vahe (F012) passed thiough the auxiliary equip-ment coolant loop and must operate against a differential pressure of 98.4 kg/cm2 Pump test return This valve allows water to be valve (F0W) returned to the suppression pool during RCIC system test and must operate against a differential l pressure of 98.4 kg/cm2 Steam supply bypass Open and/or close against full valve (F045) differential of 82.8 kg/cm2 within 5 seconds l l O Amendment 15 $.4 36 1 1

ABWR mamn Standard Plant Rrv c . Table 5.4 2 DESIGN PARAMETERS FOR RCIC SYSTEM COMPONESTS (CONTINUED) Turbine exhaust check Capable of with standing impact vahe (F038) loads due to

flapping" during startup.

Vacuum pump discharge Open and/or close against 3.2 kg/cm2 L isolation vah c differential pressure at a temp- _ (F047) ature of 1700C, Vacuum pump discharge Located at the highest check vahe (F046) point in the line, (5) Instrumentation . For instruments and control definition refer to Subsection 7.4.1.1, J (6) Condensate Storage Requirements Total resen'e storage for RCIC and liPCF system is $70 m 3, (7) Piping RCIC Water Temperature The maximum water temperature range for continuous system operation shall not exceed 600C; however, Jue to potential short term operation at higher temperatures, piping expansion calculations were based on 770C. (8) Turbine Exhaust Vertical Reactor Force The turbine exhaust sparger is capable of withstanding a vertical pressure unbalance of 1.4 kg/cm2 , This pressure unbalance is due to turbine steam discharge below the suppression pool water level. (9) Ambient Conditions Relative Temperature 11umidity(%) Normal plant operation 10 to 400C 10 to 90 (10) Suction Strainer Sizing _ The suppression pool suction strainer shall be sized so that: (a) Pump NPSil requirements are satisfied when strainer is 50 percent plugged; and particles over 2.4 mm diameter are restrained from passage into the pump and the feedwater sparger. O Amendment 15 $.437

ABWR mama Standard Plant arv. c Table 5.4 3 RHR PUMP /VAISE IDGIC Automatic Lamic or Permissives Valve Valve Normal Automatic Number Function Positior. Condition & C001 N/A STOPPED Note A Start Automatic start also requires adequate bus A.B,C tower permissht and employs time delays as necesaryto load standby pomer sources. F001 Pump Suction OPEN Permiuhts: To OPEN requires P012 A.D.C Vahts to be fully cicsed. IV12 Shutdown Suetson CLOSED Perminive: To OPEN requires F001, TVOR, A,B,C isolathn Whts and P019B C to be fully closed. IVH 11: Tube Side OPEN Note A OP!LN A.II,C Outlet Whts 1913 lin Dypass CLOSED Note A CLOSE A,H,C Wives 1910 Inboard Shutdomu CLOSED Note D CLOSE To prevent the reactor frorr A,B,C Cooling Suetion draining or filhng isolation IVil Outtward Shutdown Cic4ed Note B C1DSE To present the reactor In>m

A,B,C Cooling Suction draining or filhng l Isolation F008 - 5/P Return CIDSED Note A CLOSE Permissive: To OPEN require A,B,C Wives 1905 and 1912 to be fully closed IV21 Minimum flow OPLN Note C OPEN with Note C A.D.C Wives CLOSE without Note C i 190$ low Pressure CLOSED Note A OPEN With low reactor preuvre A,B,C Ilooder permissht of 30.66 Kg/cm2 , ,

j Injection Vahts 1

           ~ 1917    Dr)well Spray      CLOSED    Note D           CLOSE         Perminive: To OPEN requires high dr)well B,C      W hts                                                       pressure and IVo$ fully closed, or to OPEN test requires F018 fully closed 1V18     Dr)well Spray      CLOSED    Note D           CLOSE         Perminive: To OPEN requires high dr>well

!- B,C ' Isolation Vahts preuvre and IV05 fully closed, or 'o OPEN for test requires IV17 fully closed IVl9 Wetwell Spray CLOSED NOTE A CLOSE Perminive: To OPEN requires TV12 fully B,C - Isolation Wives closed and either the absence of LOCA or + f00$ fully closed closed Amendment 13 3.4 38

1 23A6100AD Standard Plant arv c Table 5.4 3 {

RHR PUMP / VALVE LDGIC (Contirued) _ Aut---'le f a or Permissives v a}ve Valve Normal Automatic Number function Position Condition Action F006 Testable Closed Permissive: To OPEN for test requires A.II.C Check Valve 1905 fully closed and f% unrm-up valve, fv!!y open. , CD02 N/A Run Note A Stop A.fi,C 1929 1. squid Waste Closed Note E Ckse A.II.C Dush Valve 1930 1.iquid Waste Ckwed Note E Clone A.II.C Mush Valve NOTES:

   ,G
   ,                 A.       LOCA signal of low reactor water level (L1) or high drywell pressure.

B, 1 ow reactor water level (L3) or high vessel pressure or RHk equipment area high temperature trip. C, Ptimp is running and low loop flow signal. D. LOCA condition as indicated by a not fully cicsed injection valve F005. i E. Low reactor water level (L3) or high drywell pressure, 4 Amer.dment 15 5.4 39

m ,. 1 m

                           .MM                                                                                                   23A6100AD Standard Plant                                                                                          RIV C Table 5.4-4 RHR HEAT EXCHANGER DESIGN AND PERFORMANCE DA'l A                                           9l        '

l l Number of units 3 Seismic CategoryI design and analysis Typek of exchangers Horizontal U Tube /Shell 1; ' l

{

Maximum primary / secondary side pressure 35.0 kg/cm2 g/14.0 kg/cm2g Design Point Function Post LOCA Containment Cooling l Primary side (tube side) performance data ' 3 (1) Flow 954 m /h [., (2); Inlet temperature ' 1820C maximum i X (3). Allowable pressure drop (max) . 0.7 kg/cm2 1 (4) Type water Suppression Pool or .- - !. Reactor Water li ' c' -(5) Foulingfactor d 2 1.024 x 10 m hoc /kcal' Secondary side (shell side) performance data o 3

                                          . (1) . Flow 1200 rn /h '

u

                                          ;(2) Inlet temperature.'                   32.60C maximum s     o            t         , ;(3) ? Alloweble pressure drop -              ' 0.7 kg/cm2 jl                     ,m maximum
                                          .(4)L Type water                           Reactor Building Cooling Water:                                                                                          -(
                                        ;(5)' Fouling factor                                   d 2                                                   '
                                                                                 '1.024 x 10 m h C/kcal-

l o7 r . l L

c 4

m \. - l[ k, / Amendment 15 '. . 5.440 I I ' 'a

                                             .;.     .n       u           _                                    . _ . ,    . __ _

ABWR. uuiman Standard Plant Rev. c f- Table 5.4 5

      .i}x'                           ('OMPONENT AND SUBSYSTEM RELIEF VALVES Relief    Relief Relief         Pressure  Flow MPL No.        ' Service       Route (l)      (kg/cm2g) (m3 /h)                  ,

C12 F004A B Condensate - B 14.0 C12 F018 Condensate B 14,0 C41 F038A B SLC Liquid C 109.7 C41 F014 SLC Liquid A 14.0 E11 TV28A C. Reactor A 35,1 Water - E11 IV39A C Reactor E 87.9 Water 6 E11 F042A C Reactor A 14.0 Water E11 F051A C ' Reactor A 35.1 i Water i

                      . E22 F020B C . Condensate          A            14.0      2,3 E51 Fr.7.        Condensate      B          - 14,0 -    2.3 -

e t 5 lr O' NOTES (1) A Suppression pool ol1 Bi Equipment drain sump l ' C SLCS pump suction i J i_ D Reactor vessel-

                                .E     Across a valve to same line                                           l F     Floor drain sump i2             Amendment is                                                                       5.4 81 b.:        i If b                                                                                                        .1 l

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l 1 ABWR maman i Standard Plant REY.C I Table 5A 5(Cont.) O! i l 1 l 1 i l 1

l l Deleted l i l l-I 1 .t [.. u 1t s.

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                                      -'ABWR                                                                                                meioari, Standard Phnt                                                                                         REV. C

_ Table 5.44 l L)I

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REACTOR WATER CLEANUP SYSTEM EQUIPMENT DESIGN DATA { l SYSTEM FLOW RATE (kg/hr) 152500 L PUMPS

                                            !              TYPE                                       Vertical Sealless centrifugal pump                    l Number Required                            2-l              Capacity (% of CUW system flow each)S0                                                           l Design temperature (OC                    66 Design pressure F #em]g)                   104
                                                         ' Discharge hea/ at shutoff (m)              160 HEAT EXCHANGERS Regenerative       Nonregenerative Number Reuired-                            1(3 shells per     2(2 Oells per unit)               unit)

Capacity (% CUW system flow each) 100 50 Shell design pressure (kg/cm'g) 104 14 Shell design temperature (OC 302 85

                                                        . Tube design pressure               (kg/cm{g).

90 90 p ,_ Tube design temperature (OC) 302 302: -i 'W .LType Horizontal Horizontal

     'd 3

u-tube u. tube l Exchange capacity (kcal/hr)- 2.74 x 10' 4.8 x 10' (per unit) FILTER.DEMINERALIZERS

Type- pressure precoat gj Number Required . . .2

        ,,.cb                            ' l.    '
                                                       ; Capacity (% of CUW system flow each)50-Flow rate per unit (kg/ht)'                76250 i                            Design temperature (OC -

66 T [ ; Dc:Jgn pressure (kg/cm]g) ~ 104 , 1 , Linear velocity'(m/hr) ~2.5 A l~. Differential Pressures (kg/cm') . Clean: 035 Annunciate 1,75

             ,,                                                 He-kwash ,                        2.1
                                                                                                                                                       -l CONTAINMENT ISOLATION VALVES                                                   .
   ,                           y, c Closing time (sec)                      < 30 u/d-i,Uy,                   ti.
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                         ,                            _ Figure 5.4 6 MAIN STEAMLINE FLOW RESTRICTOR-                                     f
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                        - ABWR                                                                                                                  maooxo Standwd Plant                                                                                                              ac I

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                                                                                                                         'N* AIR CYLINOER k ,'

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                                                                                                         !                      $PRING GUIDE
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                                      $ PEED CONTROL VALVE                     d

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ACTUATOR SUPPORT AND HELICAL $PRINGS i! - N $PRING GUIDE SH AFT ll ll . i I h / I l $l. t .' e [ ;l ij $PRING $E AT MEMBER d g i la STEM PACKING s W

h Q LE AK OFF CONNECTION { '

                                                                                              ,          j                      BONNET BOLTS g,                                    '.

L ANTERN RING

                                                                                                                       -c       BONNET CLEARANCE.                                                                                                                      j
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                                                                                ~                      "
                                      -PILOT $PRING                                                  -

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                                                                                                                                    . POPPET (PLUG, MAIN Di$K) g                j           $

v M AIN VALVE SE AT - PILOT SE AT 4 M l

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Figurr 5.4 7 MAIN STEAM ISOLATION VALVE

   ,      1               Amendment 8                                                                                                                 3.4 50 l

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l 23A6100AB Standard Plant uw. c CHAPTER 6

             ;O V-                                                                    TABLE OF CONTENTS t

Section .11tle Page 1

6. ENGINEERED SAFETY FEATURES 6.01 6.0 - GENERAL 6.0-1 6.1 ENGINEERED SAFETY FEATURE MATERIAIS 6.11 6.1.1 Metallic Materials 6.11 ,

i 6.1.2 Organic Materials 6.12

                        <                               6.2      CONTAINMENT SYSTEMS                                            6.21 D                                                                6.2.1            Containment Functional Design                  6.2 1 '

6.2.2 ' Containment Heat Removal System - ,6.2 16-6.23 Secondary Containment Functional Design 6.2 18

m. .

[ 6.2.4 . - Contain...ent Isolation System 6.2 23 6.2.5 Combustible Gas Control in Containment 6.2 31 4 6.2.6 Containment Leakage Testing- 6.2-39 -l l-

                                                               ;6.2.7            References                                     6.2-44                            i
                   '%                                  6.3      EMERGENCY CORE COOLING SYSTEMS                                  63-1 1

63,1' Design Bases and Summary Description 63 1 l c

                 ?
                                                                                                                                                                -i f ' ,'                                     4 63.2            System Design                                   63-3 633'           . ECCS Performance Evaluation                    6.3-9

g' ,

63.4 , Tests and Inspections ' 6 3-13 ir ' M< .63.5 Instrume.ntation Requirements '- 6 3-14 ' - 63.6 Interfaces - 6 3 1<

             'l                                               : 63.7-           Referenec                                       6 3-15 z

lq 6.4 H ABITABILITY SYSTEMS 6.4-1

~                                  '
   . ,                                                          6.4.1           Desigr. ,w-                                     6.4 2                             !
                                                                                               $ 9-ii L,                                                                                                                                                                 .
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l 23A6100AB

 &                               Standard Plant                                                                                                   uv. c CHAPTER 6 TABLE OF CONTENTS (Continued)                                                         h'!

Section Die Eass 1

                                        ,                       6.4.2                  System Design                             6.4-3 6.43                   System Operation Procedures               6.45 6.4.4                  Design Evaluations                        6.4-6                                 ;
                                                                                                                                              ~

6.4.5 Testing and Inspection 6.4-7 6.4.6 Instrumentation Requirements 6.47 6.4.7 Interfaces 6.47 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.51

                                                               '6.5.1                 Engineered Safety Features Filter Systems  6.5-1 6.5.2 .               Containment Spray Systems                  6.55 6.53               ~ Fission Product Control Systems.           6.5-5
                                                              '6.5.4               ' Ice Condenser as a Fission Product Contrel System                            ._6.5-6                       --
                                                      . 6.6 . PRESERVICE AND INSERVICE INSPECTION AND TESTIN.G OF CLASS 2 AND 3 COMPONENTS                                     6.61.                                  i E                ,                                             6.6.1 -               Class 2 and 3 System Boundaries           6.6                                                                                                         '

W _6.6.2 Accessibility 6.6-2 r p ' 6.63' Examination Categories and Methods 6.6-2 '

                                                              - 4 6.4 '               Inspection latervals                      6.6-4 6.6.5                 Evaluation of Examination Results         6.6 4 -

6.6.6 System Pressure Tests - 6.64 ' y; i 6.6.7 Augmented Inservice Inspection 6.65 , 6.6.8 . Code Exemptions . 6.6-5 !

              ,                                                 6.6.9              - Relief Requests                            6.6-5 i
                               -Amendment 13                                                                                                                          5 g
                                  ~
                      -.l.                .l---_--l_l_.~._.-.           --                                       . + <

n , 1 1 ABWR m6 oorn 1 Sigadard Plant nrv. c l

      ,o                         ,

n SECTION 6.3 - CONTENTS (CONTINUED) Section Title East 633/i.7 . Line Breaks Outside Containment 63 12 ) 633.7.8 Bounding Peak Peak Cladding Temperature 63 12

                                              -633.8                 LOCA Analysis Conclusions                  63 13         j 63.4                 Tests and Insnections                      63 13 63.4.1             - ECCS Performance Tests                     63 13 63.4.2               Reliability Tests and Ingections           6 3-13 63.4.2.1             11PCF Testing                              63 14 i

63.4.2.2 ADS Testing 6 3-14

                                                                                                                              ~
                                              -63 4.23               RHR Testing                                63 14 m

Li- .. 63.4.2.4 RCIC Testing 63 14

        %.                                      63.5-               Instrumentation Reaulrements               -6 3-14 63.6                Interfaces                                  6.5-15 63.7                References                                  63 15 p

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                    ,                Amendment 13                                                                             )

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i l MM 23A61ooAn Standard Plant uv. c - SECTION 6.3 TABLES- h Table Dilt East 63-1 Significant Input Variable Used in tbc Loss of Coolart Accident Analysis 63 16 1 63 2 Operational Sequence of Emergency Core l Cooling Systems, Maximum Core Flooder Line ! Break 6 3-19 63-3 Single Failure Evaluation 6 3-20 63-4 Summary of Results of LOCA Analysis 63-21 i-63 5 Key to Figures 63 22 63 6 . Plant Variables with Nominal and Sensitivity Study Values 63 23 63 7- V APLHGR Versus Exposure 6 3-24 ILLUSTRATIONS Figure Ellt East I p 63 1 HPCF Process Flow Diagram 63 25 < 1 63-2. Dc!cted 6.F45 -: i

                                    '63 3              Deleted -                                     63 28' 63-4             Vessel Pressure Versus High Pressure                                  $

Core Flooder Flow (per Sptem) Used in - r , . LOCA Analysis 63 30 :'

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V' -63 5 Vessel Pressure Versus Reactor Core isolation Cooling Flow Used in L' LOCA' Analysis 63 1 .., lie

63 6' Vessel Pressure Versus Low Pressurc Mb '

                                                    - Flooder Flow (Per System) Used in
              , .-                                    LOCA Analysis                                  63 32 lI "                  '

63 7 HPCF P&ID '63 33 L1 ii . 1

                                                                                                                          . l 63 v '                                          ,

l l~- LAmendment 13 I h i d~ '

ABWR m im^n Standard Plant RFV.C n otherwise could result in a reduction of ECCS 6.3.3 ECCS Performance Evaluation 4

       ! cffectiveness to an unacceptable level. (See Section 3.6 for criteria on pipe whip.)              Performance of the ECCS system is determined by evaluating the system response to an instant-The component supports which protect against   aneous break of a pipe. This evaluat? 4n is damage from movement and from seismic events are  performed using models either approved by the discussed in Subsection 5.4.14. The methods used  USNRC or which have met the change criteria in to provide assurance that thermal stresses do not 10CFR50.46.

cause damage to the ECCS are described in Subsection 3.9.3. The analyses included in this subsection demonstrates the ABWR ECCS system performance 63.2.7 Provisions for Performance Testing for the entire spectrum of postulatec break sizes. The analyses are based upon the core Periodic system and component testing loading shown in Figure 4.31 and was performed provisions for the ECCS are described in with the NRC approved LAMB / SCAT and SAFER / Subsection 6.3.2.2 as part of the individual GESTR models. Plants with different core system description, loadings, including those with some blank fuel bundles similar to that shown in Figure 4.3 2, 6.3.2.8 Manual Actions will show the same system sensitivities. The MAPLHGR, peak cladding temperature (PCT) and The ECCS is actuated automatically and oxidation fraction results will be provided for requires no operator action during the first the limiting break (for each bundle design) and 30 min following the accident, although operator this meets the criteria documented in Appendix action is not prevented. During the long term 4B. These results will be provided by the cooling period (after 10 min), containment utility referencing the ABWR design to the USNRC cooling occurs as a normal consequence of RHR for information. See Subsection 6.3.6. l ' U[] _ LPFL operation because the RHR heat exchangers The accidents, as listed in Chapter 15, for are in series with th'c pumps. Although not prevented from doing so earlier, the operator is which ECCS operation is required are: not required to select another RHR modes, such as Subsection Ti!k suppression pool cooling, until after the 30 m in u t e s.- 15.2.8 Feedwater Line Break 15.6.4 Spectrum of BWR Steam System The operator has multiple instrumentation Piping Failures Outside available in the control room to assist him in Containment assessing the post LOCA conditions. This 15.6.5 Loss of Coolant Accidents instrumentation provides reactor vessel pressures, water levels, containment pressure, Chap'er 15 provides the radiological temperature and radiation levels, as well as consequences of the above listed events for the indicating the operation of the ECCS. ECCS Flow core loading in Figure 4.31. indication is the primary parameter available to assess proper operation of the system. Other 63.3.1 ECCS Bases for Technical indications, such as position of valves, status Specifications of circuit breakers, and essential power bus voltage, are available to assist him in The MAPLHGRs calculated in this performance determining system operating status. The analysis provide the basis for the Chapter 16 electrical and instrumentation complement to the Technical Specifications designed to ensure ECCS is discussed in detail in Section 7.3. conformance with the acceptance criteria of Other available instrumentation is listed in the 10CFR50.46. Minimum ECCS functional P&lDs for the individual systems. Much of the requirements are specified in Subsections monitoring instrumentation available to the 6.3.3.4 and 6.3.3.5, and testing reuirements

   /]    operator is discussed in more detail in Chapter 5 are discussed in Subsection 6.3.4. Limits on

(/ and Section 6.2. Amendment 15 6M

                  - ABWR                                                                                                 m 3mi, Standard Plant                                                                                         an e minimum suppression pool water level are discussed in Section 6.2.

6.3.3.2 Acceptance Criteria for ECCS Performance The applicable acceptance criteria, extracted e: o I 15 - i.e

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                   - Amendment 13                                                                                           g3,9,3 L

ABWR m6mn Standard Plant RB' C from 10CFR50.46 are listed, and, for each Criterion 5: Lone Term Cooline criterion, applicable parts of Subsection 6.3.3 (where conformance is demonstrated) are 'After any calculated successful initial opera-l indicated, tion of the ECCS, the calculated core temperature shall be maintained at an acceptably

Criterion t Peak Claddine Temperature (PCT) low value and decay heat shall be removed for i the extended period of time required by the i 'The calcuWed maximum fuel element cladding long lived radioactivity temaining in the temperature shall not exceed 22000F." core." Conformance to Criterion 5 is l Conformance to Criterion 1 is shown for the demonstrated generically for GE BWRs in system respot.c analyses in subsections 6.3.3.7.3 Reference 1, Section III.A. Briefly summarized, l (Break Spectrum),6.3.3.7 4 (Large Breaks), for any LOCA the water level can be restored to 6.3.3.7.5 (Intermediate Breaks),6.3.3.7.6 (Small a level above the top of the core and maintained Breaks),6.3.3.". 7 (Outside Containment Breaks), there indefinitely.

6.3.3.7.8 (upper 95% Probability PCT) and specifically in Table 6.3 4 (Summary of LOCA 6333 Single Fallure Considerations Analysis Results). Conformance for each plant j will be assured for the limiting break. See The functional consequences of potential Subsection 6.3.6 for interface requirements, operator errors and single failures (including those which might cause any manually controlled Crherbn 2 Maximum Claddinc Oxidation electrically operated valve in the ECCS to move

to a position which could adversely affect the "The calculated total local oxidation of the ECCS) and the potential for submergence of valve cladding shall nowhere exceed 0.17 times the motors in the ECCS are discussed in Subsection

total cladding thickness before oxidation." 6.3.2. There it was shown that all potential

                                                                ' Conformation to Criterion 2 is shown in Figure single failures are no more severe than one of 6.310 (break spectrum plot) and Table 6.3 4 the single failures identified in Table 6.3 3.

(Summary of LOCA Analysis Results) for the system j response analysis. This limit will be assured it is therefore only necessary to consider for the limiting break. See Subsection 6.3.6 for each of these single failures in the ECCS interface requirement. performance analyses. The worst failure for any LOCA event is the Isilure of one of the diesel Criterion 3: Maximum Hydrocen Generation generators which provide electrical power to one HPCF and one RHR/LPFL. This failure results in "The calculated total amount of hydrogen the elimination of the greatest amount of generated from the chemical reaction of the flooding capability at both high and low reactor

    =                                                             cladding with water or steam shall not execed       pressures.

0.01 times the hypothetical amount that would be

          -                                                       generated if all the metal in the eladding 633.4 System Performance During the Accident cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, we:c to        in general, the system response to an

_ react." Conforr .nce to Criterion 3 is shown in accident can be described as:

            .                                                l Table 6.3 4 foi the system analysis.

(1) receiving an initiation signal; Criterion 4: Coolable Geometry (2) a small lat; time (to open all valves and

              !                                                    ' Calculated changes in core geometry shall be            have the pumps up to rated speed); and such that the core remains amenable to cooling."

l As described in Reference 2, Section Ill.A, (3) finally, the ECCS flow entering the vessel. conformance to Criterion 4 is demonstrated by conformance to Criterion 1 and 2. nr y ECCS actuation setpoints and time delays for all the ECCS systems are provided in Table N' Amendment 15 6.3 10

            -                                  ---imimumsim-

1 . ABWR msten Standard Plant REV.C

                                      .'631. The minimization of the delay from the.                                                                                                 i

_ receipt of signal until the ECCS pumps have ' reached rated speed is _llmited by the physical' , const r aint s on accelerating the

                     ,                 diesel generators and pumps. The delay time due
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ABWR n^62=^n Standard Plant prv. c l to valve motion in the case of the high pressure either the ADS initiating sign or the l 7 system provides a suitably conservative allowance overpressure signal opens tb- ,. relief ) (V for valves available for this application. In the case of the low pressure system, the time valve, no conflict exists, i l delay for valve motion is such that the pumps are The LPFL subspten is configured from the RHR l at rated speed prior to the time the vessel pumps and some of the RHR valves and piping. I pressure reaches the pump shutoff pressure. When the reactor water level is low, the LPFL I subsystem (line up) has priority through the The ADS actuation logic locludes a 29 second valve control logic over the other RHR delay timer to confirm the presence of low water subsystems for containment cooling. Immediately level 1 (LWL 1) initiation signal. This timer is following a LOCA, the RHR system is directed to initiated upon receipt of a high drywell pressure the LPFL mode. When the RHR shutdewn coc!!ng signal (which is scaled.in) and a LWL 1 signal, mode is utilized, the transfer to the LPFL mode The timer setting is consistent with the startup must be remote manually initiated. time of the ECCS which also must be running before ADS operation can occur. Once the ADS 6.3.3.6 Limits on ECCS System Parameters timer is initiated, it is automatically reset if the reactor water levelis restored above the LWL Limits ou ECCS parameters are given in the I setpoint before ADS operation occurs. The ADS sections and tables refernced in Subsections control system also provides the operator with an 63.5.1 and 633.7.1. Any number of components ADS inhibit switch which the operator can use to in any given system may be out of service, up to prevent automatic ADS cperation as covered by the the entire system. The maximum allowable engineering operating prowdures. out of service time is a function of the level l of redundancy and the specified test intervals, li The flow delivery rates analyzed in Subsection 633 can be determined from the head. flow curves 633.7 ECCS Analyses for LOCA in Figure's 6.3 4, 6.3 5 and 6.3 6 and the l. (O) pressure versus time plots discussed in Subsection 6.3.3.7. Simplified piping and 6.3.3.7.1 LOCA Analysis Procedures and input Variables l instrumentation and process diagrams for the ECCS I are referenced in Subsection 6.3.2. The The methods used in the analysis have been i operational sequence of ECCS for the limiting approved by the NRC or meet the change criterion case is shown'in Table 6.3 2. in 10CFR50.46. For the system response analysis, the LAMB / SCAT and SAFER / GESTR models l Operator action is not required, except as a approved by the NRC were used. The significant l ' monitoring function, during the short term input variables used for the response analysis cooling period following the LOCA. During the are listed in Table 6.31 and Figure 6.311. long. term cooling period, the operator may need

to.take action as specified in Subsection 6.2.2.2

63.3.7.2 Accident Description to place the containment cooling system inio operation for.some LOCA events. The operation svguence of events for the L . . limiting case is shown in Table 6.3 2. 63.3.5 Use of Dual Function Components for ECCS' 63.3.7.3 Break Spectrum Calculations l With the exception of the LPFL systems, the A complete spectrum of postulated break systems of the ECCS are designed to accomplish sizes and locations were evaluated to only one function; to ' cool the reactor core demonstrate ECCS system performance. For case following a loss of reactor coolant. To this of reference, a summary of figures presented in

      ' extent, components or portions of these systems     Subsection 633 is shown in Table 63 5.

(except for pressure relief) are not required for

operation of other systems which have emergency A summary of rest,lts of the break spectrum (T core cooling functions,' or vice nrsa. Because calculations is shown in tabular form in Table b Amendment 15 6.3 11

                         . .                                                                                 l

ABWR mama Standard Plant REV,C 6.3 4 and graphically in F' ure 6.310, Conformance to the acceptance criteria (PCT =

2200 F, local-oxidation = 17%' and core wide 1 - metal' water reaction = 1%) is demonstrated for - the core loading in Figure 4.31. Results for the limiting break for each bundle design in a plant will be given for infromation to the USNRC by the utility referencing the ABWR design (see

                      . Subsection 6.3.6). Details of calculations for                         '

specific breakr are included in subsequent , paragraphs. J 633.7.4 Large Line Betaks Inside Containment Since the-ABWR design has no recirculatign lines, thefeedwate: maximum maximum 'iacs steamline break (.903 ftbreak,and-(1.g)6 ft ), thq maximum RilR shutdown suction line break ( 853

ft') become the large break cases. Important output variables from the sensitivity study of these events are shown.in Figures 6.312 through 6.3-36.

                      - There variables are:
                      > (1) core flow as a function of time; i.

(2) minimum critical power ratio as a function of time;.

(3) water ,1evel in the luel channels as a

                             ' f unction of time; .  -

3 _'(4)/ water level inside the shroud'a3 a function a

}               .              -of time;

[' (5) -water level outside the shroud as a function.

                              ' of time; f

(6) Messel pressure as a function :)f time; l (7)- flows out of the' vessel as a function of

time;. 1 o

(8) . flows into the vessel as a function of time;

                                                                                                           -1 l

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if-. ?! . Anundtvent 15 6.3 11.1 !

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ABM 2nsioorn Standard Plant REV C g (9) peak cladding temperature as a function of draintiae break case. Therefore, the drainline

    ;         time,                                           break analysis is also bounding for any credible
     %                                                        break within the reactor internal pump A conservative assumption made in the analysis    recirculation system and its associated motor is that all offsite AC power is lost simul-          housing and cover, taneously with the initiation of the LOCA. As a further conservatism, all reactor internal pumps          As expected, the core flooder line break is were assumed to trip at the start of LOCA event the worst break location in terms of minimum even though this in itself is considered to be an transient water level in the downcomer. In accident (See Subsection 15.3.1). The resulting elevation it is the lowest break on the vessel rapid core flow coastdown produces a calculated except for the drainline break. Furthermore, the departure from nucle te boiling in the hot           worst break / failure combination leaves the bundles within the first few seconds of the          fewest number of ECC systems remaining and no transient,                                           high pressure core flooder systems. LOCA analyses using break areat less than the maximum LOCA analyses using break areas less than the    values were also considered. The cases analyzed maximum values were also considered for the          are indicated on the break spectrum plot (refer steamline, feedwater line, and RHR shutdown         to Figure 6.310). From these results it is suction line locations. The cases analyzed are      clear that the overall most limiting break in indicated on the break spectrum plot (refer         terms of minimum transient water level in the Figure 6.310), in general, the largest break at     downcomer, is the muimum core flooder line each location is the worst in terms of minimum       break case.

transient water level in the downcomer. 633.7.7 Line Breaks Outside Containment 633.7.5 Intermediate Line Breaks inside ! Containment This group of breal, is characterized by a p rapid isolation of the break. Since a maximum V For this case the maximum RHR/LPFL injection steam line break outside the containment pro-line break (0.221 ft2 ) was analyzed. Important duces more vessel inventory loss before isola-variables from this analysis are shown in Figures tion than othcr breaks in this category, the re-6.3 37 through 6.3 43, suits of this case are bounding for all breaks in this group. Important variables from these 633.7.6 Small Line Breaks inside Containment analyses are shown in Figure 6.3 60 through 6.3 66. For these cases the maximum high presure core flooder line break (0.099 ft 2) ?nd the maximum As discussed in Subsection 6.3.3.7.4, the l bottom head dran. .ine break (0.0218 ft2 ) were trip of all reactor internal pumps at the start I analyzed. Important variables from these of the LOCA produces a calculated departure from analyses are shown in Figures 6.3 44 through nucleate boiling for all LOCA events. Further. l- 6.3 59. A break in a reactor internal pump more, the high void content in the bundles fol-l would involve either the welds or the casing. lowing a large steamline break produces the l If the weld from the pump casing to the PRV stub earliest times of loss of nucleate boiling for tube breaks, the stretch tube will prevent the any LOCA event. Thus, the summary of sesults in pump casing from moving. The stretch tube Table 6.3 4 show that, though the PCTs for all clamps the diffuser to the pump casing, where its break locations are similar, the steamline i nut seats. The land is located below the casing breaks result in higher calculated PCTs and the l attachment weld and therefore the stretch tube outside steamline break is the overall most l- forms a redundant parallel strength path tc, the limiting case in terms of the highest calculated pump casing restraint which is designed to PCT. Results of the as.ysis of this break will l provide support in the event of weld failure. be provided for each bundle design for l In case the pump casing and the stretch tube information by the utility referneing the ABWR l break, the pump and motor will move downward design. 1g) until stopped by the casinc vstraints. The U L pump is part of the stretch ta In either case the break flow would be much less than the Amendment 15 63 12 l

a : 23A6100AD Standard Plant Rev.c 63.3.7J Bounding Peak Cladding Temperature Calculation - Consistent with the SAFER application methodology in Reference 2, the Appendix K peak cladding temperatures calculated in the previous sections must be compared to a statistically calculated 95% probability value. Table 6.3 6 i presents the significant plant variables which I were considered in the determination of the 95% [ probability PCT or the sensitivity study Again, since' the ABWR LOCA results have a large margin to the acceptacce criteria, a conservative PCT calculation was performed which bounds the 95% probability PCT. This bounding PCT was calculated by varying allplant variables in the

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ABWR m6im^n i Standard Plant _ E .I O V conservati?;e dc A. nimultaneously. The results of this emulation for the limiting case 63.4.2 Rellability Tests and Inspections are given in Figure 6.3 67 through 6.3 75 and The average reliability of a standby Table 6.3 4. Since the ABWR results have large (nonoperating) safety system is a function of margins to the 10CFR50.46 licensing acceptance the duration of the interval between periodic criteria, the ABWR licensing PCT can be based on functional tests. The factors considered in the bounding PCT which is well below the 22000F determining the periodic test interval of the PCT limit. ECCS are: (1) the desired system availability (average reliability); (2) the number of 633.8 IDCA Analysis Conclusions redundant functional system success paths; (3) the failure rates of the individual components Having shown compliance with the alalicable in the system; and (4) the schedule of periodic acceptance criteria of Section 6.3.3..', it is tests (simultaneous versus uniformly staggered concluded that the ECCS will perform its inction versus randomly staggered). in an acceptable manner and meet all d the criteria in Appendix 4B, given operation at or All of the active components of the HPCF below the MAPI.HGRs provided by the utility for System, ADS, RHR and RCIC Systems are designed each fuel bundle. See Subsection 6.3.6. so that they may be tested during normal pit <t operation. Full flow test capability is 6.3A Tests antlinspections provided by a test line back to the suction source. The full flow test is used to verify 63.4.1 ECCS Performance Tests the capacity of each ECCS pump loop while the plant remains undisturbed in the power i All systems of the ECCS are tested for their generation mode, in addition, each individual I operational ECCS function during the valve may be tested during normal plant (q) preoperational and/or startup test program. Each component is tested for power source, range, operation. l direction of rotation, setpoint, limit switch All of the active components of the ADS setting, torque switch setting, etc. Each pump System, except the safety / relief valves and is tested for flow capacity for comparison with their associated solenoid valves, are designed vendor date. (This test is also used to verify so that they may be tested during normal plant flow meastring capability). The flow tests operation. The SRVs and associated solenoid involve the same suction and discharge source valves are all tested during plant initial power (i.e., suppression pool), ascension per Appendix A, Paragraph D.2.c of Regulatory Guide 1.68. SRVs are bench tested to All logic elements are tested individually and establish lift settings. ! then as a system to verify con.plete system response to emergency signals including the Testing of the initiating instrumentation and ability of valves to revert to the ECCS alignment controls portion of the ECCS is discussed in from other positions. Subsection 7.3,1. The emergency power system, l which supplies electrical power to the ECCS in L Finally, the entire system is tested for the event that offsite power is unavailable, is response time and flow capacity taking suction tested as described in Subsection 8.3.1. The from its normal source and delivering flow into frequency of testing is specified in the Chapter tha reactor vessel. This last series of tests is 16 Technical Specifications. Visual inspections pc.iormed with power supplied from both offsite of all the ECCS components located outside the l power and onsite emergency power. drywell can be made at any time during power operation. Components inside the drywell can bc l See Chapter 14 for a thorough discussion of visually inspected only during periods of access ! preoperational testing for these systems.

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Amendment 15 6.1 13

ABWR- = m^n i Sandardflant RIN. C to the drywell. When the reactor vessel is open, 63A.23 RHR Testing , the spargers and other internals can be l inspected. The RHR pump and valves are tested ' periodically during reactor operation. With the 63A.2.1 HPCI' Testing injection valves closed and the return line open to the suppression pool, full flowing pump The llPCF can be tested at full flow with capability is demonstrated. The injection valve suppression pool water at any time during plant and the check valve are tested in a manner operation except when a system initiation signal similar to that used for the HPCF valves. The is present, if an initiation signal occurs while system test conditions during reactor operation the llPCF is being tested, the system returns are shown on the RilR system process diagram automatically to the operating mode. The (Figure 63 3). motor operated valve in the line to the condensate storage system is interlocked closed 6.3A.2A RCIC Testing when the suction valve from the suppression pool is open. The RCIC loop can be tested during reactor operation. To test the RCIC pump at rated flow, A design flow functional test of the HPCF over the test bypass line valve to the suppression the operating pressure and flow range is pool and the pump suction valve from the perfmmed by pumping water from the suppression suppression pool are opened and the pump is pool back through the full flow test return line started using the turbine controls in the to the suppression pool. control room. Core.ct operation is determined by obs;rving the instruments in the control The suction valve from the condensate storage roma. tank and the discharge v:!n io the r-actc.. remain closed. These two valves are tested if an initiation signal occurs during (N separately to ensure their operability. test, the RCIC system returns e the ope:ating mode. The valves in the test bypass lines are Tiie llPCF test conditions are tabulated on the- closed automatically and the RCIC pump discharge llPCF process flow diat; ram (Figure 631). valve is opened to' assure flow is correctly ' routed to the vessel. 63A.2.2 Alls Testing ? .

                                                            . 6.3.5 Instrumentation Requirements dan ADS logic system functional test and simulated automatic operation of all ADS logic              Design details including redundancy and logie channels are to be performed at least once per          of the ECCS instrumentation are discussed in         -
    . plant operating interval between reactor Section 7.3.

refuelings. Instrumentation channels are demonstrated operable by the performance of a All instrumentation required for automatic chnnnel' functional' test and a trip unit and manual initiation of the HPCF, RCIC, RHR and - 4 calibration ut least once per month and a ADS is discussed in Subsection 7.3.1 and is e transmitter calibration at least once per designed to meet the requirements of IEEE.279 operating interval. and other applicable regulatory requirements. The HPCF, RCIC, RHR and ADS can her manually -

        'All SRVs, which incluac those used for ADS are      initiated from the control room.
-l#  bench tested to establish lif t settings 'in compliance with ASME Code Section XI.                      The RCIC, HPCF, and RHR are automatically initiated on low reactor water level or high-drywell pressure. The ADS is automatically actuated by sensed variables for reactor vessel O

AnwodmeM 7 G 14 ,

23A6100AD Standard Plant _ REV.C 1

            'v}                                . Iow water level and drywell high pressure plus indication that at least one RHR or HPCF pump is operating. The HPCF, RCIC, and RHR automatically return from system flow test modes to the emergency core cooling mode of operation following receipt of an automatic invitation signal. The RHR LPFL mode injection into the RPV begins when reactor pressure decreases to the RHR's pump discharge shutoff pressure.

HPCF injection begins as soon as the HPCF pump . W '

is up to speed and the ' injection valve is open, since the HPCF is capable of injection water into the RPV over a pressure range from 1177 psid to -

100 psid or pressure difference between the vessel and drywell.

          >                                     6.3.6 Interfaces P

63.6.1 ECCS Performance Results ~

                                                 ' The exposure dependent MAPLHGR, peak cladding                           j temperature, and oxidation fraction for each fuel

, 1bundl.c design bas'ed'on the limiting break size

               '4 i

will be provided by the utility to the USNRC for Information. L L. W( ).L6.3,7 Referea e l'- J , 1. ; General Electric Company Analytical Model 1 [,9[;,, , for Loss of Coolant Analysis in Accordance . X ;with 10CFR.54 Appendit K, (NEDE 20566 P-A), i September 1986.- ;; !~ r:, ' 1 k( *

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         -o:                                                                Table 6.31 Q,                                               SIGNIFICANT INPUT VARIABLES USED IN THE LOSS.OF. COOLANT ACCIDENT ANALYSIS (Continued)

Variable llalla Value laitiating Signals ft above TAF I 10.6 Low Water Level and High Drywc!) Pressure psig 22.0 DelayTime from AllInitiating see 1 29 Signals Completed to the Time ,

 ,                                       Valves are Open -

C. FUEL PARAMETERS

Variable Units Velac FuelType --- Initial Core Fuel Bundle Geometry -------

8x8 lv Lattice ~ ~ . . C ;

                                      . Number of Fueled Rods per Bundle                - --              62 Peak Technical Specification                kw/ft                 13.4

. Linear Heat Generation Rate t: a 1 Initial Mi .imum Critical -- 1.13. 'g j Pc ;st Ratio' * {.

                ,                       Design Axial Peaking                        - - - - -             1.40 l',                                     Factor                                                                                   i V
               , +.                                                                                                             ,

[ 1 . 1' Au: I

                            * ;The system response analysis is based upon the core loading in Figure 4.31. The sensitivities

(- Jdemonstrated are valid for other ' core loadings. l$ I. .. I" j .

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ABWR msius l Standard Plant nrw. c u Table 6.3 2 ) OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEM MAXIMUM CORE FLOODER LINE BREAK Gi ' h, 1std Etents 1 0 Design basis LOCA assumed to start; normal anxiliary power assumed to be lost (1)

                      ~5       Reactor low water level 3 is reached. Reactor scram occurs. (2)
                      ~ 10     Drywell high pressure is reached. All diesel generators, RCIC, HPCF, RHR/LPFL signaled to start. (3)
                      ~ 18 -   Reactor low water level 2 is reached. RCIC receives second signal to start.
                      ~ 48 _ RCIC injection valve open and pump at design flow which completes RCIC startup.
                      ~ 65     Reactor low water level 1.5 is reached. All diesel generators and HPCF receive second
                             . signal to start. Main steam isolation valves signaled to close.
                      ~ 78     All diesel generators ready to load; RHR/LPFL and HPCF loading sequence beings,
                      ~ 102    HPCF injection valves open and pumps at design Dow, which completes HPCF startup
                      ~118     Reactor low water lem! 1 is reached. RHR/LPFL receives second signal to start.

ADS delay timer initiated.

                      ~ 148 . ADS delay timer timed out. ADS valves actuated.
                      ~344 -Vessel pressure decreases below shutoff head cf RHR/LPFL RHR/LPFL injection vakes open and Gowinto vessel beings.

Core effectively reflooded assuming worst single failure; heatup terminated I i 1- MIpjj; . 3 . For the purpose of all but the next to last entry on this table, all ECCS equipment is I (1) _ assumed to function as designed. Performance analysis calculations consider the t effects of single equipment failures (Subsection 6333).

                    .(2). For the LOCA analysis, the water level is initiated at the scram water level.'

m -- (3) For the LOCA analysis, the ECCS initiation on high drywell pressm e is not considered.

                                                                                                                                     ^1
                     ' See Figure 63 46 l

Amendment 15 6.3 19 t ,'f]' g

<n

ABM useimro l Standsrd Plant REY.C I Table 6.3-4

SUMMARY

OF RESULTS OF LOCA ANALYSES l~ l 1-1 1 i l' .I GE PROPRIETARY - provided under sepa. ate cover l . r li ,_0 ' i. ,> e

-\

1~ 1 i to- . i-

      ,I:,         [

9

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i-e' , I i i l k ANN Mhh . L 4, !g. .

k . l. . FAB M ursioorn Signdard Plant REV.C j- Table 6.3 7 J MAPLHGR VERUS EXPOSURE

t

!- Exposure MAPLHG R** l} IMWd/O kw_f.fL HIGH ENRICHMENT 200 12.0 1,000 12.2 5,000 12.7 10,000 12.9 15,000 12.9 20,000 12.6 25,000 11.7 ; I

                                                                                           - 30,000                           10.8 lc u-v p                               ,                              MEDIUM ENRICHMENT 200-                         11.9 m                                                                                              1,000                           12.0 l!'                     *
                                                                                             '5,000                           12.1
                            .                                                                10,000                           12.2 -                              ,
       .lI(7                                                                                 15,000 20,000 12 3 12,1

, .'; (d 25,000 11.6 - l r , ii:L ' " 30,000 113- )_ l 1.OW ENRICHMENT '- p 2. l3 200 11.5 1,000' 11.4 5,000 113 W"- , 10,000 L 11.5

                       "?                                                                  = 15,000
                         -    <                                                                                              11.5 .

20,000 11.0- i 25,0001 10.4 [ q i fl c r

   . ,N                                                                                                                                                        1

[$~ For the core loading in Figurc 43-1~ - t" ' f. W These values are limited by the peak LHGR of 13.4 kw/ft and not by ECCS performance f'

i.

                                                                                                                                                               -l g                       ":

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+: \ >

                                         ) Amendment 13                                                                                               6.3 24

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ABWR. mame

               ~-Standard Plant                                                                   aev n 7A.2 MULTIPLEXING SYSTEMS GE PROPRIETARY - provided under separate cover O                                                 (includes pages 7A.2-1 through 7A.2-22)

(pages 7A.2-3 and 7A.2-4 Amendment 9, all other pages Amendment 6) i I O Amendment 15 7A.21 22

1 23A6100All Standard Plant n ,.. ii m SECTION 9.1 .

   !   Y                                                                                             l V                                   CONTENTS (Continued)

Sectio.n 31Lis Eagt 9.1.4.2.10.2.5 Vessel Closure 9.1 61 9.1.4.2.103 Departure of Fuel from Site 9.1-6j 1 9.1.43 Safety Evaluation of Fuel Handling System 9.16j 9.1.4.4 laspection and Testing Requirements 9.1 6k 9.1.4.4.1 Inspection 9.1-6k Testing 9.1.4.4.2 9.1 9.1.4.5 Instrumentation Requirements 9.1 61 9.1.4.5.1 Automatic Refueling Ma: hine 9.161 9.1.4.5.2' Fuel Support Grapple 9.1 61 9.1.4.53 Other 9.161

 -- .A-()              9.1.4.5.4      Radiation Monitoring                              9.1-61 9.1.5          Overhead Heavy Load Handline Systems              9.17          ;

9.1.5.1- Design Bases 9.17 l 9.1.5.2 System Description 9.1 7 i-9.1.5.2.1 Reactor Building 9.17 l l 9.1.5.2.2 Reactor Building Laydown Areas 9.1 8 l 9.1.5.23 New Fuel Storage Vault 9.18 1: 9.1.5.2.4 Spent Fuel Storage Pool 9.18-9.1.5.2.5- Storage Pit 9.1-8 l l l 9.1.5.2.6 Upper Drywell 9.1 9 E 9.1.5.2.7 Lower Drywell 9.19 l ,

9.1vi Amendnwns 6 l

JN N 23A6100All Standard Plant nev n SECTION 9.1 1 CONTENTS (Continued) Ol Section Title Eag I l 9.1.5.2.7.2 Reactor Internal Pump Senicing 9.19 ; 9.1.5.2.7.3 Fine Motion Control Rod Drive Senicing 9.1 10 9.1.5.2.7.4 Neutron Monitor Sensor Senicing 9.1 10 l 9.1.5.2.8 Special Senicing Room / Areas 9.1 10 9.1.5.2.9 Heasy lead Handling Equipment 9.1 10 9.1.5.2.10 Transportation Paths / Routing 9.111 9.1.5.2.11 Equipment Operating Procedures, Maintenance and Senice 9.1 11 9.1.5.3 Safety Evaluations 9.112 9.1.5.4 Inspection and Testing 9.1 12 '- 9.1.5.5 Instrumentation Requirements 9.112 9.1.5.6 Operational Responsibilities 9.1 12 l' y i L H l-- O 9.1 vii 1. l- Amendment 15 l l! 1 1-_ _-__ - - -- _ _- _- _ - - _-________ _ _ ____ _

ABWR 2346iooxii Standard Plant _ nev n i compatible with the environment of treated water compartments, which provide lateral restraints J n' _ and provides a design life of 60 years. over the entire length of the fuel assembly or bundle. 9.1.2.2 Facilities Description (Spent Fuel Storage) (4) The racks are fabricated from materials used for construction and are specified in accordance (1) The spent fuel storage racks provide storage in with the latest issue of applicable ASTM the reactor building spent fuel pool for spent specifications at the time of equipment order, fuel received from the reactor vessel Juring the refueling operation. The spent fuel storage (5) Lead in guides at the top of the storage spaces racks are top entry racks designed to preclude provide guidance of the fuel during insertion, the possibility of criticality under normal and abnormal conditions. The upper tieplate of the (6) The racks are designed to withstand, while fuel elements rests against the rack to proside maintaining the nuclear safety design basis, the lateral support. The lower tieplate sits in the impact force generated by the vertical free fall bottom of the rack, which supports the weight drop of a fuel assembly from a height of 6 feet. of the fuel. (7) The rack is designed to withstand a pullup force (2) The rack arrangement is designed to prevent of 4000 lb and a horizontal force of 1000 lb, accidentalinsertion of fuel assemblies or There are no readily definable horizontal forces bundles between adjacent modules. The in excess of 1000 lb and in the event a fuel s'orage rack is designed to provide accessibility assembly should jam, the maximum lifting force to the fuel bail for grappling purposes, of the fuel-handling platform grapple (assumes limit switches fail)is 3000 lb. (3) The location of the spent fuel pool and the containment pool within the complex is shown (8) The fuel storage racks are designed to handle l (] in Section 1.2 Irradiated fuel assemblies, The expected 1- () radiation levels are well below the design levels. 9.1.2.3 Safety Evaluation The fuel storage facilities will be designed to 9.1.2.3.1 Criticality Control Seismic Category I requirements to prevent earthquake damage to the stored fuel. The geometry of the spent fuel storage rack is such that the k will be less than 0.95 under normal and The fuel storage pools have adequate water abnormaY conditions. For GE racks, this is shielding for the stored spent fuel. Adequate

         - demonstrated by applying the criterion specified in     shielding for transporting the fuel is also provided, i          Appendix 4B.                                              Liquid level sensors are installed to detect a low pool L                                                                   water level, and adequate makeup water is available 9.1.2.3.2 Structural Design and Material                 to assure that the fuel will not be uncovered should a Compatibility Requirements                               leak occur.

l (1) The spent fuel pool racks provide storage for Since the fuel storage racks are made of l 270% of the reactor core, noncombustible material and are stored under water, L there is no potential fire hazard. The large water (2) The fuel storage racks are designed to be volume also protects the spent fuel storage racks from supported above the pool floor by a support potential pipe breaks and associated jet impingement structure. The support structure allows loads. sufficient pool water flow for natural con-vection cooling of the stored fuel. Since the Fuel storage racks are made in accordance with the modules are freestanding (i.e., no sup ports latest issue of the applicable ASTM specification at l above the base), the support structure also the time of equipment order. The storage tubes are l q(

')

provides the required dynamic stability. permanently marked with identification traceable to the material certifications. The fuel storage tube (3) The racks include indhidual solid tube storage Amendment 15 9.1 2d

ABWR m6mn Standard Plant nev. ii assembly is compatible with the environment of _ treated water and provides a design life of 60 years, . including allowances for corrosion. Regulatory Guide 1.13 is applicable to spent fuel storage facilities._ The building containing the fuel i storage facilities, including the storage racks and pool,is designed to protect the fuel from damage caused by: (1) natural events such as carthquake, high winds and flooding, and (2) mechanical damage caused by dropping of fuel assemblies bundles, or other objects onto stored fuel. 9.1.2A Summary of Radiological Considerations Ily adequate design and careful operational procedures, the safety design bases of the spent fuel storage arrangement are satisfied. Thus, the exposure of plant personnel to radiation is maintained well below published guideline values. Further details of radiological considerations, , including those for the spent fuel storage 1 arrangement, are presented in Chapter 12. 5 l l l O. Amendment 6 9.12c

ABWR -m Standard Plant RW A All major handling equipment components: cords are available. Qualification load and o) ( V cranes, hoist, refueling platform etc., will be performance testing, including nondestructive provided with an operating instruction and main- examination (NDE) and dimensional inspection on tenance manual for reference and utilization by heavy load handling equipment will be performed operations personnel. Handling equipment operat. prior to OA acceptance. Tests may include load ing procedure will comply with the requirements capacity, safety overloads, life cycle, sequence of NUREG 0612 Subsection 5.1.1(2). of operations and functional areas. The operational programs for maintenance and When equipment is received at the site it will servicing are described in Subsection 9.1.5.6. be inspected to ensure no damage has occurred during transit or storage. Prior to use and at 9.1.5.3 Safety Evaluations periodic intervals each piece of equipment will be tested again to ensure the electrical and/or The crancs, hoists, and related lifting de- mechadcal functions are operational including vices used for handling heavy loads either sa- visual and, if required, NDE inspection, tisfy the single failure proof guidelines of NUREG-0612 Subsection 5.1.6, including NUREG. Crane inspections and testing will comply 0554 or evaluations are made to demonstrate com- with requirements of ANSI B30.2 and NUREG 0612, pliance with the recommended guidelines of Sec. Subsection 5.1.1(6). tion 5.1, including Subsection 5.1.4 and 5.1.5. 9.1.5.5 Instrumentation Requirements The equipment handling components over the fuel pool are designed to meet the single failure The majority of the heavy load handling proof criteria to satisfy NUREG 0554. Redundant equipment is manually operated and controlled safety interlocks and limit switches are provided by the operator's visual observations. This to prevent transporting heavy loads by RB crane type of operation does not necessitate the need O V over any spent fuel storage area. for a dynamic instrumentation system. A transportation routing study will be made of Load cells may be installed to provide auto-all planned heavy load handling moves to evaluate matic shutdown whenever threshold limits are ex. and minimize safety risks. ceeded for criticalload handling operations to prevent overloading. Safety evaluation of related light loads and , refueling handling tasks in which heavy load 9.1.5.6 Operational Responsibilities l equipment is also used are covered in Subsection 9.1.4.3. Critical heavy load handling in operation of the plant shallinclude the following documented 9.1.5.4 Inspection and Testing program for safe administration and safe implementation of operations and control of l Heavy load handling equipment is subject to heavy load handling systems: the strict controls of Quality Assurance (OA), l- incorporating the requirements of Federal Regu- (1) Heavy Load Handling System and Equipment lation 10CFR50, Appendix B. Components defined Operating Procedures, j as essential to safety have an additional set of [ engineering specified " Quality Requirements

that (2) Heasy Load Handling Equipment Maintenance j

identify safety related features which require Procedures and/or Manuals. specific OA verification of compliance to drawing / specification requirements. (3) Heasy Load Handling Equipment Inspection and Test Plans; NDE, Visual, etc. Prior to shipment, every lifting equipment component requiring inspection will be reviewed (4) Heavy Load Handling Safe Load Paths and by OA for compliance and that the required re-

  >n V

s Routing Plans. Amendment 2 9.1 12

E M MM 23A6100All Standard Plant Rw. n (5) OA Program to Monitor and Assure - Implementation and Compliance of Heavy Load Handling Operations and Controls. (6) Operator Qualifications, Training and Control Program. A O

b e Amendment 15 9.113

Standard Plant nev. ri SECTION 9.3

1. ,1 U CONTENTS Section Iult East 93.1 Comoressed Air Systems 93-1 93.2 Process and Post Accident Samnlinn System 93-la 93.2.1 Design Bases 93la 93.2.1.1 Safety Design Bases 93-la 93.2.1.2 Power Generation Design Bases 93la 93.2.2 System Description 93la 93.2.2.1 General Description 93-la 93.2.2.2 Sampled Process Streams and Analyze Parameters 93-la 93.2.23 Provisions for Obtaining Representative Samples 93-la n

l id 93.23 Sampling Panels 93-lb , l-' 93.23.1 Reactor Building Sample Station 93-Ib 93.23.2 Feedwater Corrosion Product Monitor 93-1c

                      .93.233    Residual Heat Removal, Fuel Pool and -

Suppression Pool Sampling 931c , 1 [ 93.23.4 Reactor Building 93-Ic 93.23.5 . Radwaste System Sampling ' 93-Ic l 93.2.4 Sample Probe Design - 93-le 93.2.5 Sample Piping Design 93-Ic 93.2.6 Safety Evaluation Operator Safety 93-1d l l: 93.2.7 Tests and Inspection 93-1d (} 93.2.8 Instrumentation Application 93-Id L. !A lU 93 il i Amendment 15 l

ABWR me ui Standard Plant _ am n SECTION 9.3 I CONTENTS (Continued) h Section M East 933 Eauloment and Floor Drainane System 93 2 1 93.4 Chemical and Volume Control System (PWR) 93 2-93.5 Standhv I.Jauld Control System 93 2 93.5.1 Design Bases 93 2 93.5.1.1 Safety Design Bases 93 2 93.5.2 System Description 93 2 93.53 Safety Evaluation 93-4 ! 93.5,4 Testing and Inspection Requirements 93 6 93.5.5 Instrumentation Requirements 93 6 93.6 - Instrument Air System 93 7 93.6.1 Design Bases 93 7 93.6.1.1 Safety Design Bases 93 7 93.6.1.2 Power Generation Design Bases 93-7' 93.6.2' System Description 93 7 93.63 Safety Evaluation 93-7 93.6.4 Inspection and Testing Requirements 93-7a

                                                 '93.6.5                                               Instrumentation Application              93-7a 93.7                                                Service Air SysteH1                      93-8 93.7.1                                              Design Bases                             93 8 93.7.1.1                                            Safety Design Bases                      93 8 93,7.1.2                                            Power Generation Design Bases            93-8 93.7.2                                              System Description                       93-8 93.73                                               Safety Evaluation                        93-8
                                                                                                                                                                  ~1 93.iii Amendment 6

ABWR u ^6tm A11 Standard Plant Rev.H SECTION 9A CONTENTS Section Title P_ast 9.4.1 Control Hulldinn HVAC 9.4 1 9.4.1.1 Control Room Equipment HVAC 9.4 1 9.4.1.1.1 Design Basis 9.4 1 9.4.1.1.2 Power Generation Design Basis 9.4-1 9.4.1.13 System Description 9.4 1 9.4.1.1.4 Safety Evaluation 9.4 la 9.4.1.1.5 Inspection and Testing Requirements 9.4-la 9.4.1.1.6 Instrumentation Application 9.4 la 9.4.1.1.7. Regulatory Guide 1.52 Compliance Status 9.4 lb 9.4.1.1.8 Standard Review Plan 6.5.1 Compliance Status 9.4 lb l 9.4.1.2 . Essential Electrical and Reactor Building - L Cooling Water Equipment HVAC 9.4-1b l' 9.4.1.2.1 Design Basis 9.41b 9.4.1.2.2 Power Generation Design Basis 9.4-lb l L 9.4.1.23 System Description 9.4-Ib - 9.4.1.23.1 Safety-Related Subsystem 1 9.41b 9.4.1.23.2 Safety-Related Subsystem 2 9.41c ll 9.4.1.233 Safety Related Subsystem 3 9.41c

               ,9.4.1.2.4    Safety Evaluation                              9.4-Ic_             ;
               -9.4.1.2.5    Inspection and Testing ?,equirements           9.4-1c 9.4.1.2.6'   Instrumentation Application                  ' 9.4 Id 9.4.2        Spent Fuel Pool Area Ventilation System        9.42
  'D.

i

'w/

9.4-li l Amendment 15 r

MM . 23A6100AH Standard Plant myn SECTION 9.4 CONTENTS (Cantinued) Section Title East 9.43- Auxillary Area Ventilation System 9.4 2 9.4.4 Turbine Area Ventilation System 9.4 2 9.4.4.1 Design Bases 9.4 2 9.4.4.1.1 Safety Design Bases 9.42 9.4.4.1.2 Power Generation Design Bases 9.4 2 9.4.4.2 Description 9.42 7 9.4.4.2.1 General Description 9.4 2 9.4.4.2.1.1 Turbine Building Exhaust System 9.4-2a 9.4.4.2.1.2 Turbine Building Battery Room Exhaust Subsystem 9.4-2a 9.4.4.2.13 Turbine Building Electrical Switchgear and Air Compressor Rooms Ventilation Subsystem 9.4 2a , 9.4.4.2.1,4 Turbine Building Lube Oil Area Exhaust Subsystem 9.4 2a 9.4.4.2.1.5 TBVS Unit Coolers and Electric Unit Heaters 9.4 2a 9.4.43 Evaluation 9.4 2a - 9.4.4.4 Tests and Inspections 9.4 2b 9.4.4.5 Instrumentation Application ' 9.4 2b '

         . 9.4.5 '       Reactor Buildinst Ventilation System                 9.4 2c 9.4.5.1       Secondary Containment ilVAC System '                 9.4 2c -

9.4.5.1.1 Design Bases 9.4 2c 9.4.5.1.1.1 Safety Design Bases 9.4 2c 9.4 iii Amendment 6

ABWR usamsn Standard Plant nev. n 9A.6 FIRE HAZARD ANALYSIS

     .,        DATABASE l

l 1 l l j GE PROPRIETARY - provided under separate cover (locludes pages 9A.61 through 9A.6-182)- Eagg ' . Amendment

i 9A.6-1 t'.au 9A.6-3 14 I 9A.6-4 thru 9A.6 92 6 9A.6-93 thru 9A.6-162 14 9A.6-163 thru 9A.6171 Reserved for Control Building

9A.6-172 thru 9A.6-102 14 i

l-O Amendment 15 9A.61 182

Mkh 23A6t00AJ Standard Plant am A (m) 10.3 Main Steam Supply Systent 103.2.1 cenerai Description The function of the main steam supply system The main steam supply system is illustrated in is to convey steam generated in the reactor to the Figure 1031. The main steam piping consists of four turbine plant. This section discusses that portion of 28 inch diameter lines from the outboard main steam the main steam system which ranges between, but line isolation valves to the main turbine stop vahes. does not include, the ortermost containment The four main steam lines are connected to a header isolation vanes and the turbne stop valves. upstream of the turbine stop valves to permit testing of the main steam line isolation valves during plant The main steam line pnssure relief system, operation with a minimum load reduction. This main steam line flow restrictor& main steam line header arrangement is also provided to ensure that isolation valves (MSIVs), and mJn steam piping the turbine bypass and other main steam supplies are from the reactor nonles through the ovboard main connected to operating steam lines and not to idle steam isolation valve (MSIV) are described in Sub- lines. The main steam process downstream of the sections a.2.2,5 4.4,5.4.5, and 5.4.9 respectively, turbine stop valves is illustrated in Figure 103 2. 10.3.1 Design Bases The design pressure and temperature of the I main steam piping is 1250 psig and 600 F, j 103.1.1 Safety Design Bases respectively, the same values as the design parameters I of the reactor. The main steam lines are classified as l The main steam supply system is not required discussed in Section 3.2. to effect or support safe shutdown of the reactor or to perform in the operation of reactor safety A drain line is connected to the low points of features, however, the main steam supply system is each main steam line, both inside and outside the designed: containment. Both sets of drains are headered and f) connected, with isolation valves to allow drainage to To comply with applicable codes and standards the main condenser. To permit intermittent draining V l (1 in order to accommodate operational stresses of the steam line low points at low loads, orificed tires such as internal pressure and dynamic loads are provided around the final valve to the main I without risk of failures and consequential condenser. The steam line drains maintain a continu. l releases of radioactivity in excess of the ous downward slope from the steam system low established regulatorylimits; points to the orifice located near the condenser. The drain line from the orifice to the condenser also l (2) To accommodate normal and abnormal envi- slopes downward. To permit emptying the drain lines ronmental conditions; and for maintenance, drains are provided irom the line low points, going to the radwaste system. l (3) With suitable accesses to permit inservice testing and inspections. The drains from the steam lines inside contain-i ment are connected to the steam lines outside con. 103.1.2 Power Generation Design Bases tainment to permit equalizing pressure across the main steam line isolation valves during startup and Power Generation Design Basis One - The system following a steam line isolation. is designed to deliver steam from the reactor to the , turbine generator system for a rangt of flows and 103.2.2 Component Description l pressures varying from warmup to rate I conditions. It also provides steam to the reheaters, the steam jet The main steam system lines are made of air ejectors, the turbine gland sealiag and the carbon steel and are sired for a normal steady state deacrating section of the main conden.cr and the velocity of 150 feet per second, or less. The lines are turbine bypass system. designed to permit hydrotesting following construc-tion and major repairs without addition of temporary

 /q      10.3.2 Description                                    pipe supports.

V Amendment 8 10.3 1

i 33A6)(OA.I Standard Plant nev ^ 103.2.3 System Operation l NORMAL OPERATION At low plant O1 power levels, the main steam system may be used to 10.3.4 Inspection and Testing supply steam to the turbine gland steam seal system. Requirements At high plant power levels, turbine gland sealing steam is normally supplied from the one turbine inspection and testing will be in accordance with crossaround pipe. the requirements of Section 3.2. The main steam line will be hydrostatically tested to confirm leak. Steam is supplied to the crossaround steam re- tightness, beaters in the T G system when the T G load ex-cceds 15 percent and supply steam pNasure is con. 10.3.5 Water Chemistry (PWR) trolled by regulating valves in the 15 to 60 percent load range. This section applies to a pressurized water reactor (PWR), and is therefore not applicable. If a large, rapid reduction in T G load occurs, steam is bypassed directly to the condenser via the 10.3.6 Steam and Feedwater System turbine bypass system (see Subsection 10.4.4 for a Materials description of the turbine bypass systern). Steam and feedwater component materials are 10.3.3 Evaluation identified in Section 3.2. All components and piping for tLe main steam 103.6.1 Fracture Toughness of Class 2 Components supply system are designed in accccdance with the codes and standards listed in ',cetion 3.2. This The fracture toughness properties of the ferritic ensures that the main steam s'.pply system accom. materials of these components will meet the require-modates operational strese.s resulting from static ments of NC 2300, Fracture Toughness Requirements and dynamic loads, inci., ding steam hammer and for Materials, (Class 2) of ASME Code, Section III, normal and abnornial environmental conditions. of edition and addenda as invoked by Regulatory The break of a main steam line or any branch line Guide 1.26, Onality Group Classification and will not result in radiation exposures in excess of the Standards for Water , Steam , and Radioactirc-Waste limits of 10 CFR 100 ta persons located offsite Containing Components of Nuclear Power Plants. because of the safety features designed into the system. The main steam line pipe break accident is addressed in Chapter 15, and high energy pipe failure is discussed in Section 3.6. 103.6.2 Materials Selection and Fabrication The materials specified for use in Class 2 components will conform to Appendix I to Section III of the ASME Code, and to Parts A,B, and C of Section 11 of the Code. Regulatory Guide 1.85, Code Case Acceptability ASME Section HIMaterials, describes acceptable code cases that wi!! be used in conjunction with the above specifications. The following criteria are applicable to all compo-nents: (1) Regulatory Guide 1.71, Wc/dce Qualification for Areas of Limited Accessibility, provides the followirig criteria for assuring the integrity of Amendment 15 10.3-2

                                          -    - -               - .-      ..   -.   .         .-~

M 23A6100AK . Standard Plant nev.n

p. 11.2 LIQUID WASTE MANAGEMENT SYSTEM I O 1 1

GE PROPRIETARY - provided under separate cover j (includes pages 11.21 through 11.2 45) Eggt Amendment Eggs Amendment 11.2 1 15 11.2-23 6 11.2-1.1 14 11.2 24 6 11.2 2 6 11.2 25 6 11.2 3 6 11.2 26 6 11.2-4 6 11.2 27 6 11.2-5 13 11.2 28 6

l. "( 11.2-6 6 11.2 29 6 11.2-7 6 11.2 30 6 11.2 8 6 11.2 31 6 11.2 9 6 11.2-32 6 11.2 10 6 11.2 33 6 11.2 11 6 11.2-34 6 11.2-12 6 11.2 35 6 11.2 13 6 11.2 36 6 11.2 14 6 11.2 37 6 '

11.2415- 6 11.2 38 6 11.2 16 6 11.2-39 6 11.2;17 6 11.2-40 6 11.2 18 6 11.2-41 6 11.2 19 6 11.2-42 6 11.2 20- 6 11.2-43 6 11.2 21 6 11.2-44 6 11.2 22- 6 11.2 45 6 l P f Amendment 15 11.2-1

ABWR 2mimx Standard Plant n,v n 11.3 GASEOUS WASTE MANAGEMENT SYSTEM GE PROPRIETARY - provided under separate cover (includes pages 11.31 through 113 24) fagg Amendment 113 1 6 11.3 2 6-113-3 6

11 3-4 6 11 3-5 14

      .                                 11 3-6                         6 r                                        11 3-7                         6 113-8                          6 11 3-9                         7 113 10                         7 113 11                         6 113-12                         15 l:                                       113-12.1                       15                          .i L.                                       113 13                         7-L                                        113-14                         6' L                                        11 3-15                        13 P                                        11.3 16                        6 113 17                         6 11 3-18                        6

j. 113 19 6
. 113-20 6 i 113-21 6 11 3-22 6 11 3-23 6 l s 113-24 6 Amendment 15 11.31 24 l

             . . - .                 .     . . . .          - . . . . . - . . . . . . . - - . . .  .  . . .     . .~ ..

nA6100AM Standard Plant RIN.H 13.6.3 - Intefaces

Qn. i l i t SAFEGUARDS INFORMATION Provided under separate cover. j (Includes pages 13.G2 through 13.G2.2) l

     ,O':-

l ll -) i l I O Amendment 15 13/

1 23A61MAM Standard Plant RW H O 1. Y i l r L-1 SAFEGUARDS INFORMATION Provided under separate cover. (Continuation of Subsection 13.63, " Interfaces-) 4 i O 13.6-3 f ' Amendment 15

       .ABWR                                                                                        m62mn         l
       - Standard Plant                                                                              uv. c        j CHAPTER 15 TABLE OF CONTENTS                                                     l Section                            Tult                                   East 15     ACCIDENT ANALYSES                                                                               i 15.0 GENERAL 15.0.1         Nuclear Safety Operational Analysis                  15.0-1 15.0.2         Event AnalyticalObjective                            15.0-1 15.0 3        Analytical Categories                                 15.01 15.0.4         Event Evaluation                                    15.0-1 15.0.5         Interface Requirements                              15.0-3          l 15.1 DECREASE IN REACTOR COOLANT TEMPERA'llBE 15.1.1        Loss of Feedwa'.cr Heating                           15.1 1 15.1.2         Feedwater Controller Failure--

l Maximum Demand 15.1 2

  -(                    15.1 3         Pressure Regulator Failure- Open                     15.1 5 15.1.4         Inadvertent Safety / Relief Valve Opening            15.1 7              ,

15.1.5 Spectrum of Steam System Piping Failures I inside and Outside of Containment i inaPWR 15.1 8 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1 8 ; 15.2 INCREASE IN REACTOR PRESSURE I' 15.2.1 Pressure Regulator Failure -Closed 15.2 1 l l 15.2.2 Generator Load Rejection 15.2 2 15.2 3 Turbine Trip 15.2 5 15.2.4 MSIV Closures 15.2-8 l O 15li

     ' Ameridment 15

_. _ . . . . . . ~_. ~ABWR usamin' Standard Plant Rf!V D CHAPTER 15 TABLE OF CONTENTS (Continued) Section Title Eage 15.2.5 Loss of Condenser Vacuum 15.2 11 15.2.6 Loss of Non Emergency Offsite AC Power to Station Auxiliaries 15.2 12 l 15.2.7 Loss of Feedwater Flow 15.2-15 l 15.2.8 Feedwater Line Break 15.2 16 l 15.2.9 Failure of RHR Shutdown Cooling 15.2 16 l 15.2.10 References 15.2 17 153 DECREASE IN REACTOR COOLANT SYSTEM FLOW R ATE 15 3.1 Reactor Internal Pump Trip 15 3-1 3 15 3.2 Recirculation Flow Control Failure-- Decreasing Flow 15 3-3 1533 Recirculation Pump Seizure 153 5 , 15 3.4 Recirculation Pump Shaft Break 15 3-6 15 3.5 Referenices 153 7 - 15.4 REACTIVITY AND POWER DISTRIHtrrION ANOMALIES 15.4.1 Rod Withdrawal Error--Low Power l 15.4-1

                   .15.4.2         . Rod Withdrawal Error at Power            15.4 3
                                                                                                    -l 15.4 3          Control Rod Maloperation (System                                   l Mkifunction or Operator Error)            15.4-4 l

l 15.4.4 AbnormalStartup ofIdle Reactor ' Internal Pump 15.4 4 15.4.5 Recirculation Flow Control Failure with l Increasing Flow , 15.4 5 O 15lii Amendment 2

ABWR mamo Standard Plant - RT?V H CHAPTER 15 TABLE OF CONTENTS (Continued) Section Tjl].c P_ags 15.4.6 Chemical and Volume Control System Malfunctions 15.4 7 15.4.7 Misplaced Bundle Accident 15.4 7 15.4.8 Rod Ejection Accident 15.4 8 15.4.9 Control Rod Drop Accident 15.4 9 l 15.4.10 References 15.4 10 l 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 inadvertent HPCF Startup 15.5 1 l 15.5.2 Chemical Vclume Control System Malfunction (or Operator Error) 15.5 2 15.5 3 BWR Transients Which increase Reactor Coolant Inventory 15.5 2 15.6 DECREASE IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Safety Relief Vahe Opening 15.6-1 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6 1 15.6 3 Steam Generator Tube Failure ~ 15.6 2 15.6.4 Steam System Piping Break Outside Containment 15.6-2 15.6.5 Loss of Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary)- Inside Containment 15.6-5 15.6.6 Feedwater Line Break Outside Containment 15.6 15.6.7 References 15.6-10 9 15 iv Amendment 2

23A6100AD Standard Plant Rrv. c CHAPTER 15 TABLE OF CONTENTS (Continued) h Section 11Lig Eage 15.7 JLADIOACTIVE RELEASE FROM SUBSYS1TMS AND COMPONENTS 15.7.1 Radioactive Gas Waste System Leak or Failure 15.71 15.7.2 ' Lionid Radioactive System Failure 15.7 2 15.7.3 Postulated Radioactive Released Due to Liquid Radwaste Tank Failure 15.73 15.7.4 Fuel Handling Accident 15,7 4 l

! 15.7.5 Spent Fuel Cask Drop Accident 15.7 8 15.7.6 References 15.7 8 15.8. ANTICIPATED TRANSIENTS WITHOUT SCRAM 15.8.1 Requirements - 15.8-1 15.8.2 Plant Capabilitics 15.8-1 APPENDIX 15A Plant Nuclear Safety Operational Analysis (NSOA) APPENDIX 15H Failure Modes and Effects Analysis (FMEA) APPENDIX 15C Probability Analysis of Reactor Internal Pump Trip APPENDIX ISD Probability Analysis of Pressure Regulator Downscale Failure APPENDIX 15E A*IWS Performance Evaluation ; t 15-v l Amendment 15

    -AB M                                                                              2346 oorn Standard Plant                                                                       arv.c
 ..                                          SECTION 15.0 dC                                             CONTENTS Sestion                                          11tle              P.ast                               ;

i 15.0.1 Nuclear Safety Onerational Analysis 15.0-1 [ 15.0.2 Event Analytical Ohlective 15.0 1.1 15.0 3 Analdical Cateqtorl% 15.0-1.1 15.0.4 Event Evaluation 15.0-1.1 15.0.4.1 Identification of Causes and Frequency Classification 15.0 1.1 15.0.4.2 1dentified Results 15.0-1.1 i i 15.0.4 3 Sequence of Events and Systems Operations 15.01,1 15.0.4.4 . Analysis Basis ' 15.0-1.1 15.0.4.4.1 Evaluation Models 15.0-1.1-15.0.4.4.2 - Input Parameters and Initial Conditions for Analped Events 15.0-2 15.0.4.4.3 Initial Powey Flow Operating Cordtraints 15.0-2 l 15.0.4.5 Evaluation of Results 15.0-2.1 15.0.4.5.1 Effect of Single Failures and Operator Errors 15.0-3 15.0.4.5.2 Analysis Uncertainties 15.0-3 1 5 . 0 - 11 Amendment 15

ABM uuioonn Standard Plant REY.C SECTION 15.0 CONTENTS (Continued) 0 Section M East l l 15.0.4.5 3 Barrier Performance 15.0-3 15.0.4.5.4 Radiological Consequences 15.0-3 15.0.5 Interface Reaulttments 15.0-3 l 15.0.5.1 Anticipated Operational Occurences (AOO) 15.0-3 15.0.5.2 Operating Limits 15.0-3 15.0.5 3 Design Basis Accidents 15.0-3 l TABLES Table Illit East ,

                                                                                         .15.0-1                    Input Parameters and initial Conditions for System Response Analysis Transients   15.0-4                         .

s 15.0-2 Results of Summary of System Response Analysis Transients Events . 15.0-7 15.0-3 Summary of Accidents 15.0-11 i 1504 - Core Wide Transient Analysis Results 15.0 11 15.0-5 Scram Reactivity Curves 15.0-12 1 15.0-6 ABWR FMCRD Scram Time 15.0 13

                                                                                                                              - ILLUSTRATIONS-Figure                                    Iille                    East
                                                                                        '15.0-1                     System Response Analysis Power / Flow Map  15.0-14 1 5 . 0 - i11                                            -

Amendment 15

.l

_ m . ~ + A:mm. A 5 .a. ABWR usuoorn Standard Plant RW C , 15.0 GENERAL 15.0.1 Nuclear Safety Operational 7: Analysis (d in this chapter the effects of anticipated process disturbances and postulated component In Appendix 15A nuclear safety operational failures are examined to determine their analyses (NSOA), all unacceptable safety results

consequences and to evaluate the capability built and all required safety actions are identified.

into the plant to control or acwommodate such in addition, an evaluation of the entire failures and events. The system response spectrum of events is :onsistently carried out analysis is 1 ased upon the core loading shown in for all plant designs to demonstrate that a ' Figure 4.31 and is used to identify the limiting consistent level of safety has been attained. events for the ABWR. Other fuel designs and core loading patterns including loading patterns The NSOA acceptance criteria are based on similar to Figure 4.3 2 will not affect the event probability. This means that events more i sensitivities demonstrated by this study. likely to occar are tested against more > Evaluation of these limiting events for each restrictive limits. This is consistent with plant cycle will assure that the criteria in industry practics and the applicable regulatory Appendix 4B are met. requirements. GE has developed a unique systematic approach The starting point for the NSOA is the to plant safety consistent with the GE boiling establishment of Luacceptable safety results, water reactor technology base - The key to the GE This concept enablec the results of any safety-approach to plant safety is the Nuclear Safety analysis to be compaced to applicable criteria. Operational Analysis. A generic nuclear safety Unacceptable safety cesults represent an operational analysis has been developed for each extension of the nuclear design criteria for of the recent GE boiling water reactor product plant systems and compom nts which are used as

lines. It has then been modified to be the basis for system design. The unacceptable , n compatible with the specific plant configuration safety results have been selected so that they j- ] being evaluated. Key inputs into the nuclear are consistent with applicable regulations and 1 -safety operational analysis are derived from the industry code and standards. L applicable regulations and through industry codes and standards,- The generic nuclear safety The focal point of the NSOA is the event operational analysis for ABWR is presented in analysis. In the event analysis, all essential Appendix 15A'. protection sequences are evaluated until all required safety actions are successfully l GE has evaluated the entire spectrum of events completed. The event analysis identifies all in the nuclear safety operational analysis to required front line safety systems and their establish the most limiting or design basis - essential auxiliaries, events in a meaningful manner. It is the design . basis events that are quantified in this chapter. The full spectrum of initial conditions limited by the constraints placed on planned The scope of the situations analyzed includes operation for AOOs, accidents, and plant anticipated (expected) operational occurrences capability demonstrations are evaluated. All (AOos);(e.g., loss of electrical load), events are analyzed until a stable condition is off design abnormal (unexpected) transients that obtained. This assures that the event being induce system operations condition disturbances, evaluated does not have a characteristic for postulated accidents of low probability-(e.g., long term consideration which is important. the sudden loss of integrity of a major component), and finally, hypothetical events of In the event analysis all essential system, extremely low probability (e.g., an anticipated operator actions, and limits to satisfy the transient without the' operation of the entire required safety actions are identified. Limits

controt rod drive system), are derived only for those parameters g continuously available to the operator. Credit

   ;                                                            for operator action is taken only when an Amendment 15                                                                                     15.0 1

ABM 2346ioorn Standard Plant REV.C operator can be reasonably expected to perform groups is assigned to one of three frequency

  - the required action based on the information groups defined in Regulatory Guide 1.70.                        .

available to him. 15.0.4.2 Identined Results In the NSOA a complete and consistent set of safety actions has been developed. These safety Events analyzed for each plant must meet the actions are those required to prevent criteria in Appendix 4B. unacceptable results. For transients and accidents, a single failure proof path to plant 15.0.4.3 Sequence of Events and Systems shutdown must be shown. The application of a Operations single failure criterion to these events is imposed as an additional measure of conservatism Each transient or accident evaluated in the in the nuclear safety operational analysis sensitivity study is discussed and evaluated in process, terms of: 15.0.2 Event AnalyticalObjective (1) a step by step sequence of events from initiation to final stabilized condition; The spectrum of postulated initiating events developed from the NSOA was divided into (2) the extent to which normally operating plant i categories based upon the type of disturbance and instrumentation controls are assumed to the expected frequency of the initiating function; occurrence. The limiting events in each I combinetion of category and frequency were (3) the extent to which the plant and reactor l cvaluated using the core loading in Figure 4.31 protection systems are required to function; to determine the limiting events. The plant safety analysis evaluates.the ability of the (4) the credit taken for the functioning of plant to operate without unacceptable safety normally operating plant systems; results within regulatory guidelines. This objective is met by satisfying the criteria in (5) the operation of engineered safety systems Appendix 4B. that is required. 15.0.3 AnalyticalCatego:les This sequence of events is supported by the NSOA for the transient or accident. The effect Each event analyzed is assigned to one of of a single equipment failure or malfunction or eight categories listed in Chapter 15 of- an operator error on the event is shown in the l- Regulatory Ouide 1.70. NSOA. 15.0.4 Event Evaluation 15.0.4.4 Analysis Basis 15.0.4.1 Identification of Causes and The sensitivity study results given in this l Frequency Classification chapter are hsed upon the core loading given in 1. Figure 4.31. These sensitivities are valid for Situations and causes which lead to the ether fuel higns and core loadings. initiating event analyzed in the sensitivity study are described within the categories 15.0.4.4.1 Evaluation Models 7 designated in Subsection 15.0.3. The frequency of occurrence of each event is summarized based The models used.to analyze the core and upon the nuclear safety operational analysis and system performance have been approved by the currently available operating plant history for USNRC or meet the criteria in Appendix 4B. the transient. event. Events for which inconclusive data exist are discussed separately within each event section. Each initiating event within the major Amendment 15 15.0 1.1

k 23A6100A3 Standard Plant RI?V C 15.0.4.4.2 loput Parameters and initial 15.0.4.4.3 Inltlal Pomr/ Flow Operating Conditions for Analyzed Events Constraints in general, the evects analyzed for the The power / flow map used for the system sensitivity analysis documented within this response enalysis is shown in Figure 15.01, section have values for input parameters and The analyses basis for most of the sensitivity initial conditions as specified in Table 15.01. analyses is 102% thermal power at rated core Analyses which assume data inputs different than flow (100%). This operating point is the apex these values are designated accordingly in the of the operating power / flow map which, in l appropriate event discussion, response to any classified abnormal operational transients, will yield the minimum pressure and The normal maximu n allowable reactor operating thermal margins. Referring to Figure 15.01, condition is the 100%/100% power / flow condition, the apex of the bounded power / flow map is point The maximum power measurement uncertainty is A, the upper bound is the design flow control l usually ~2%. Therefore, the sensitivity analyses line (102% tod line A D), the lower bound is the are based on 102% power level. The transient zero power line H' J the right bound is the results at this condition are more severe than maximum flow line A' H', and the left bcund is that at rated condition, the natural circulation line D.J. The power / flow map (A D-J H:A') represents the operational region covered by abnormal operational transient evaluations. Any other constraint which may truncate the bounded power / flow map must be observed, such as the moisture carryover protection region, the licensed power limit and other restrictions based on pressure and thermal margin criteria. I See Subsection 4.4.3.3 for power / flow map operating instructions. The upper operating The analytical values for some system power / flow limit of a reactor is predicated on characteristics, like SRV delay / stroke time, the operating basis of the analysis and the reactor internal pump coastdown time constant, corresponding constant rod pattern line. This etc., bound the design specification for that boundary may be truncated by the licensed power system. These values will be checked during and the GETAB operating limit, startup tests. Certain localized events are evaluated at All setpoints for the protection system other than the above mentioned conditions, assumed in the analyses are conservative, which These conditions are discussed pertinent to the includes instrument uncertainty, calibration appropriate event, error and instrument drift. The nominal and allowable values for these setpoints, shown in The power / flow operating map for a plant may the technical specifications assume that the differ from that used in the system response setpoint . will not exceed what are assumed in the analysis given in this chapter. Differences in analyses. the map will not change the designation of limitir.g events. The operating map used at a in conclusion, the input parameters and plant will be provided by the utility applicant initial conditions used in the sensitivity study referencing the ABWR des'gn to the USNRC for documented in this chapter are conservative information. See Subsection 4.4.2.1. values, which include uncertainties, and bound l the operating band. O Amendment 15 15.0-2

         . $$                                                                                 2mim^n Standard Plant                                                                        RIN. C 15.0A.5 Evaluation of Results
            - The results of the system response analyses are presented in Table 15.0 2. Based on these
         .results, the limiting events have been identified. Reasons why the other events are not limiting are given in the event O

{ l '. j .- l O

       ' Amendment 15 I

(s'

ABWR 2 mime Standard Plant uvc documentation. The limiting events which 15.0.4.5.3 Marr6er Performance

   . establish CPR operating limit:

The significant arcos of interest for (1) Limillag Pressurlastion Esents: Inadvertent internal pressure damage are the high pressure closv.re of one turbine control valve, and portions of the reactor coolant pressure genere load rejection with all bypass boundary (the reactor vessel and the high valve failure, pressure pipelines attached to the reactor vessel). The pisst shall meet the criteria in Appendix 4B. (2) Limiting Decrease in Core Coolant Temperaturt Events: Runout of Two Feedwater Pumps. 15.0.4.5.4 Radiological Consequences For the core loading in Figure 4.31, the in this chapter, the consequences of resulting initial core MCPR operating limit is radioactivity release for the core loading in 1.17. The operating limit based on the plant Figure 4.31 during the three types of events: loading pattern will be provided by the utility (a) incidents of moderate frequency (anticipated applicant referencing the ABWR design to the operational occurences)l (b) infrequent USNRC for information, see Subsection 15.0.5.2 incidents (abnormal operational occurences); and for interface requirement. (c) limiting faults (design basis accidents), are given. For all events whose consequences Results of the transient analyses for are limiting, a detailed quantitative evaluation individual plant reference core loading patterns is presented. For nonlimiting events, a . will differ from the results shown in this qualitive evaluation is presented or results are chapter. However, the relative results between referenced frorr a more limiting or enveloping core associated events do not chang. Therefore, case or event. only the results of the identitled limiting events given in Tables 15.0 4 will1 c r.rovided by 15.0.5 Interface Requirements

 '     the utility applicant referencing the A0WR design to the USNRC for information. Fee Subsection         15.0.5.1 Ant8 cipated Operational Occurences 15.0.5.1.                                            (ADO) 15/1.4.5.1 Effect of Singic Fallures and                 The results of the events identified in Operator Errors                                      Subsection 15.0.4.5 for plant core loading will be provided by the ultlity applicant referencing The effcet of a single equipment failure or       the ABWR design to the USNRC for information.

malfunction or operator error is provided in Appendix 15A. 15.0.5.2 Operating Limits 15.0.4.5.2 Analysis Uncertainties The operating limit resulting from the analyses normally provided in this subsection The analysis uncetainties meet the criteria in will be provided by the ultiity applicant Appendix 4B. referencing the ABWR design to the USNRC for information. In Table 15.0 3. a summary of applicable accidents !s piovided. This table compares GE 15.0.5.3 Design Basis Accidents Alculated amount of failed fuel to that used in worst case radiological calculatiuons or the Results of the design basis accidents core shown it. Figure 4.31. Radiological including radiological consequences will be calculations for a plant initial core will be provided by the ultiity applicant referencing ptovided by the utility to the USNRC for the ABWR design to the USNRC for information. Informatlod. (See Subsection 15.0.5 for interface requirements). Amendment 15 15A3

1 214610aAll l RimmAard Plant myc 1 ( Table 15.01 INPUT PARAMETERS AND INITIAL CONDITIO!?3 FOR SYSTEM RESPONSE ANALYSIS TRANSIENTS

1. Thermal Power I atl(MWt) !

Warranted V4ue 3926 Analysis Value 4005 1

2. Steam Flow (kg/ht) 6 Warranted Value 7.64 x 10 6

Analysis Value 7.84 x 10

3. Core Flow (kg/hr) 6 Rated $2.2 x 10 Maximum $8.0 x 10

4. Feedwater Flow Rate (kg/sec) !

Warranted Value 1122 Analysis Value 2179

5. Feedwater Temperature ('C) 217 ,

6. VesselDomiPressure(kg/cm g) 73.1 I 7. Vessel Core Pressure (kg/cm g) 73.7

8. Turbine Bypass Capacity (% NBR) 33

9. Core Coolant inlet Enthalpy . 294.1 (kcal/kg) 2

10. Turbine Inlet Pressure (kg/cm a) 69.9

11. Fuel Lattice N

12. Core Leakage Flow (%) 11.67

13. Required MCPR Operating Limit 1.17

14. MCPR Safety Limit 1.07

15. Doppler Coefficient (-)//'C Analysis Data for Power increase Events (REDY only)* 0.429 Analysis Data for Fower Decrease Events (REDY only)* 0.180 }

For transients simulated on the OD)W computer model, this input is calculated by OD1W.

Amendment 13 !$.0-4

ABM uss m n : Standard Plant nrv. c i Table 15.01 h I?'PUT PARAMETERS AND INITIAL CONDITIONS FOR SYSTEM RESPONSE ANALYSIS TRANSIENTS (Continued)

16. Void Coefficient (.)(/% Rated Voids '

Analysis Data for Power increase Events (REDY only)' 11.6 Analysis Data for Power Decrease Events (REDY only)* 2.5

17. Core Average Rated Void Fraction (%) (REDY only)' 43.4

18. Seram Reacthity, Sak Table 15.0 5 Analysis Data (REDY only)'

19. Control Rod Drive Table 15.0 6 Position versus time I

20. Nuclear characteristics used in EOEC" ODYN simulations

21. Number of Reactor Internal Pumps 10

22. Safety / Relief Valve Capacity (WNBR) 913 at 80.5 Kg/cm7g l

Quantity Installed 18

23. Relief Function Delay (sec) 0.4

24. Relief Function Opening Time (sec) 0.15 t

25. Safety Function Delay (sec) 1.0 (1)

26. Safety Function Openirs Time (sec) 0.3

27. Set Points for Safety / Relief Valves l Safety Function (Kg/cm2g) 82.8, 83.5, S4.2, 84.9,85.6 RelicIFunction(Kg/cm 2g) 80.5,81.2, 81.9,82.6,833,S4.0

28. Number of Yahc Groupings Simulated Safety Function (No.) 5 Relief Function (No.) 6 i- ' For transients simulated on the ODYN model, this input is calculated by ODYN l " E0EC = End of Equilibrium Cycle (1) This is a programming convenience number, i

Amendment 15 15.0 5

ABWR mnun samadard Plant an c I Table 15.01 INPUT PARAMETERS AND INITIAL CONDITIONS FOR SYSTEM RESPONSE ANALYSIS TRANSIENTS (Continued)

29. S/R Vaht Reclosure Setpoint Both Modes

(% of setpoint) Maximum Safety limit (used in analysis) 98 Minimum Operationallimit 93

30. High Flux Trip (% NBR)

Analysis Setpoint (125 x 1.02) 127.5 2

31. High Pressure Scram Setpoint (kg/cm g) 77.7

32. VessellevelTrips (m above bottom of separator skirt bottom)

Level 8 -(12) (m) 1.73 level 4 - (IA)(m) 1.08 Level 3 -(13)(m) 0.57 level 2 -(L2)(m) 0.75

33. APRM Simulated Thermal Power Trip Scram % N11R Analysis Setpoint (115 x 1.02) 117.3 ;

Time Constant (sec) 7 l 34. Reactor Internal Pump Trip Delay (sec) 0.16

35. Recirculation Pump Trip Inertia Time Constant for Analysis (sec) '" 0.62 3

36. Total Steamline Volume (m ) 113.2

37. Set presfure of Recirculation purep trip (kg/cm g) 79.1

(, l

For transients simulated on the ODYN model, this input is calculated by ODYN.

I: " EOEC = End of Equilibrium Cycle.

                "* The inertia time constant is defined by the expression:

2PJ n t = ....... 9........ , where t = inertia time constant (see); gT " = pump motor inertia (kg.m); J" n = pump speed (rps); g = gravitational constant (m/sec ); and T, = pump shaft torque (kg.m) O t. Amendment 15 15.0-6

ABM 2a6 ion Standard Plant Riv c Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYEIS TRANSIENT EVENTS Max Core No. Max. Man. Average et Duntion Ma n. Max. Vessel Steam Surface Yahts et Sub Neutron Dome Bottom 1.ine flest Ilus A Freq. Ibt Blom 40mm Section figure flux Presure Preuvre Presure (% of in Cate- Blow. (seconds) M 1LL Descriotion M (Egiggi 8) 2 2 (b&lGI18) Ch8!C32g) gg[,jgj) g gg g 15.1 Deertue in core coolant temperature 15.1.1 Ima of I:ced. 112.8 73.1 75.9 71/. 112.8 0.07 a 0 0 ester heating 15.1J 15.12 Runout of one 1N.5 73.2 75 8 71.7 1018 0.06 a 0 0 teed =3ter pump 15.1.2 15.13 Runout of two 139.0 83.3 M9 82.8 .05.9 0.10 a+ 10 6 feedwater pumps 15.1 3 15.14 Opening ot 102.1 73.1 75.6 71.6 100.0 " a 0 0 one Hypass Vaht . 15.1.3 15.1 5 Opening ot all 102.0 80.4 81.8 80.1 100.0 " a+ 0 0 Control and Dypau Vahts i 15.1.4 Inadvertent open SE *ITXT ing of One SRV 15.1.6 Inadvertent RilR Sm 'ITAT Shutdown Cooling 15.2 Increase in Reactor Pressure 15.2.1 15.2 1 Closure et One 138.8 75.7 78.2 "/4.5 103.1 0.10 a 0 0 Turbine Control Yaht 15.2.1 15.2 2 Pres. Regu- 154.8 85.8 87.4 85.1 103.0 N/A c 18 6 lator Downscate I' ail. i 15.2.2 15.2-3 Genetator Imd 148.1 830 64 7 82.7 100.2 0.06 a 10 5 Rejection, Dypau on Frequency depnition is discussed in Subsection 13.0.4.1 Not limiting (Sec Subsection 15.0.4.5.)

a Moderate Frequency b infrequent c Limiting Fault N/A Not applicable s + This event should be classiped as a limitingfault. However, criteriafor rnoderatefrequent incidents are conservatively applied.

Amendment 15 15.0 7

l

               \                                                                                                    2M6100AD Standard Plant                                                                                                 nry e Table 15.0 2 V)           RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.) Max Core No. Max. Max. Average of Duration Mar. Max. Vessel Steam Surface Valves of Sub Neutron Dome Bottom Une IIcat flux A Freq. First Ilk =down Section Figure flus Preuvre Pressure Preuvre (% of in Cate. Blow. (seconds) Q Q Description ,ci;ldlB (WCm 2g) f'At/Cm2g) (Kt/Cm2g) gjtgg) gij g jgggy. d2*.31 13.2.2 15.2-4 Generator land 153.3 84.2 85.8 83.6 100.5 0.07 a+ 14 $ Rejection,l'ailurs of One Bypau Valve 15.2.2 15.2 5 Generator land 184 6 86.1 87.7 85.6 1 02.3 0.10 a+ 18 6 Rejection with failure o.* all Dypass Valves 15.2.3 15.2-6 Turbine Trip 122.1 83.0 84 6 82.6 100.0 0.05 a 10 $ Dypsu-On . 13.2.3 15.2 7 Turbine Tnp 131.9 84.1 85.6 83 4 100 0 0.0$ a+ 14 $ l w/ Failure of One Dypass Valve 15.2.3 15.2-8 Turdir e Trip 158.6 86.1 87.7 85 4 100.6 0.08 a+ 18 6 r'] (g with failure of all Dypau Valves 13.2.4 15.2 9 Inadvertent 102.1 M.6 86.4 84.1 100.1 " a 18 $ MSIV Closure 13.2.5 15.2 10 loss of 122.3 83.0 M.6 82.6 100.0

                                                                                                         "         a        10        $

Condenser Vsevum 15.2.6 15.2 11 less of AC 113.2 82.9 84 4 82.7 100.0 0.05 a 10 $ Power ( , 15.2.7 15.2 12 less of All 102.0 73.1 75.7 - 71.6 100.1

                                                                                                         "         a       0          0 l                            Feedwater flow 1528               Feedwater Piping Dreak               SEE         TT!XT Frequency definition is discussed in Subsection 150.4.1
         "       Not firniting (See Subsection 150.4.5) a       Moderate Frequency b       infrequent c       Lirniting Fault N/A Not applicable
  ,,     +       This ewnt should be classiped as an infrequent event or a lirnitingfault. Howcrer, criteriafor

( , rnoderate frequent incidente are conscivativey applied. Amendment 15 1$44

A:BWR meima Standard Plant nry c Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.) Max Core No. Max. Max. Average of Duration i Max. Max. Veuel Steam Surface Yahts of Sub Neutron Dome Ibttom line flest Ilus A Preq. first Dk= dows Section figure flux Preuvre Preuure Preuvre (87c of in Cate. Bk=- (seconds) d G Desenedqn 5.l@B (ht/Cm El 2 SElfm 21) $1/f.m 2g) ggjggg g y g gggy. d2ED 1519 Pailure of RIIR SEE TEXT Shutdomm Cooling 15.3 Deercase in Reactor Coolant System flow Rate 15.3.1 15.3 1 Tnp of Three 102.0 73.3 76.0 71.7 100.1 0.41 a 0 0 Reactor Internal Pumps 15.3.1 15.3 2 Trip of All 102.0 83.2 64.1 82.7 100.2 "' c Reactor Internal Pumps 15.3.2 15.3-3 Past Runback 102.0 73.0 75 9 71.6 100.0 " a 0 0 ot One Reactor Internal Putnp 1 l 15.3.2 !$.3-4 rest Runback 102.0 73.1 76.0 71,6 100.0 " a+ 0 0 O of All Resetor Interal Pumps 15.3.3 15.3-5 Selrure of One 102.0 73.1 75.9 71.6 100.0 " c 0 0 Reactor Internal Pump 15.3.4 One Pump Shaft SEE T11XT Dreak 15.4 Reaethiry and Power Distribution Anomalies 15.4.1.1 RWE Refueling SEE TEXT hequency definition is discussed in Subsection 15.0.4.1

   "        Not lirniting (See Subsection 15.0.4.3.)

CPR criterion does not apply. PCT <593.30C a Moderate Frequency b Infrequent c Lirniting Fault

   +        This event should be classtped as a livnitingfaul:. :lowever, criteriafor rnoderatefrequent incidents are conservatively applied.

Amendment 15 15.0 9

2M6100AD Standard Plant RIN.C Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.) Max Core No. Max. Max. Average of Duration Max. Max. Vesact Steam Surface Vanes ot Sub Neutron Dome Bottom 1Jne flest nux A Fres nrst Blowdown Section Figure Mux Pressure Preasure Pressure (% of in Cate. Blow- (seconds) 1.11 M Description $1TB (hz!Cm 21) (Kt/Cm 8) 2 (EE!f E) InintD EE}L EQty.' 1211 15A.12 RWE Startup SEE TTXT 15A.2 RWE at Power $EE TTXT 15 4.3 Control ht=1 SEE TTXT Misoperation 3A4 Abnormal Startup 1E TEXT ot One Reactor laternal Pump

                                                                                                                                                                           ""                    0         0 15A.5   15.4 2    rest Runout           89.8        71.1      72.3                                               70.6                         116.1           a of One Reactor Internal Pump 15A3 1543         Fast Runout          135.0       723       74.7                                               713                           168.5          a+              0         0 of All Reactor Internal Pumps 1547              Misplaced Bundle                                                                                SEE                        '!Tyr Accident 15.5              Increase in Reactor Coolant inwntory
                                                                                                                                                                            "     a+              0         0 15.5.1  15 3-1    Indvertent            102.0       73.1       75.6                                               71.6                        100.0 ll *CF 5tartup Frequency definislon is discussed in Subsection :S.O.4.1 Not limiting (See Subsection 15.0.4.5.)
                     **"    Transients initiatedfrom low power.

a Moderate Frequency b infrequent c Limiting Fault

                      +      This event should be classified as a limitingfault. However, criteriafor moderatefrequent inciaents are conservatively applied.

Amendment 15 15.010

MM 2M6100AD Standard Plant RIV C Table 15.0 3

SUMMARY

OF ACCIDENTS FAILED FUEL RODS GE NRC SUBSECTION CALCULATED WORST CASE 1., "iTLE VALUE ASSUMITION 15 3.1 Trip of All Reactor Internal Pumps Ncne -60% 1533 Seizure of one Reactor Internal Pump None None 15 3.4 Reactor Internal Pump Shaft Break None None 15.6.2 Instrument Line Break None None 15.6.4 Steam System Pipe Break Outside None None Containment 15.6.5 LOCA Within RCPB None 100 % 15.6.6 Feedwater Line Break None None 15.7.1.1 Main Condenser Gas Treatment N/A N/A System Fallare ' 15.73 Liquid Radwaste Tank Failure N/A N/A 15.7,4 Fuel Handling Accident < 125 125 15.7.5 Cask Drop Accident None All Rods in Cask Table 15.0 4 CORE WIDE TRANSIENT ANALYSIS RESULTS TO HE PROVIDED FOR DIFFERENT CORE DESIGN MAX. CORE MAX. AVERAGE NELTRON SURFACE FLUX HEAT FLUX DELTA TRANSIENT (%NBR) (%NBR) f2R FIGURE Closure of One Turbine Control Valve X X X X Load Rejection with all Bypass Valves X X X X Failure Runout of 2 Feedwater Pumps X X X X l

 . Amendment 15                                                                                          15.0 11

23A6100All Standard Plant MVA Table 15.0 5 SCRAM REALTIVin' CURVES SCRAM REACMTIY W CONTROL FRACTION BOClt EOEC* 0.0 0.0 0.0 0.05 -0.235 -0.082 0.10 0.473 0.170 0.20 0.945 0.360 030 1.611 0.656 0.40 2.576 1.122 0.50 4.295 1.875 OI4 7.160 3.366

  /                                   0.70                 13.60 -6.728 l                                      0.80                 25.44   14.08 0.90                 33.44  27.05 i

1.00 34.56 31.20

    *BOC1    = Beginning of Qtle i                                                                "

EOEC = End ofEquilibrium Qvie l' l 15.0 12

3146100 Aft Standard Plant uvc Table 15.0-6 h' ABWR FMCRD SCRAM TIME ! Scram Tlaw (seconds) (includian Solenoid De enerulaation) Rod Insertion (%) Used in A=-Ivsis 10 0.46 40 1.208 (4 1.727 100 3.719 O O Amendment 13 15.0 13

ABWR nuime Standard Plant nov c O 120 i l l l A' 100 - B'- C B BD y OESIGN F LO.'.' CONTROL LINE

 -3 e
                                                     \

5 m E 60 - -

 .E Oj                                                            K' 40   -

y - RATED CORE FLOW NATURAL CIR CU L ATION 20 - - MOISTUR E CARRYOVER PROTECTION LINE J , , , H H' O O 20 40 60 B0 100 120 CORE FLOW (% rated) Figure 15.0-1 SYSTEM RESPONSE ANALYSIS POWER / FLOW MAP Amendment 15 13.0 14

ABWR zwiman Standard Pant nix c SECTION 15.1 L' CONTENTS (Continued) Sectlon Dile East 15.1.6.2.2 Systems Operation 15.1 8 15.1.6.3 Core and System Performance 15.19 15.1.6.4 Barrier Performance 15.1 9 15.1.6.5 Radiological Consequences 15.1 9 TABLES Table Ilde Eage 15.11 Sequence of Events for less of Feedwater Heating 15.1 10 - 15.1 2 Loss of $5.60C Feedwater Heating 15.1 10 l 15.13 Single Failure Modes for Digital Controls 15.111 , ['( 15.1 4 Sequence of Events for Figure 15.12 15.112 15.15 Sequence of Events for Figure 15.13 15.1 12

              !$.16            Sequence of Events for Figure 15.14                15.1 13
             '.5.1 7           Sequence of Events fo: rigure 15.1-5               15.1 13 15.18             Sequence of Events for inadvertent Safety / Relief Valve Opening                                      15.114            F 15.19            Sequence of Events for inadvertent RHR Shutdown Cooling Operations                                 15.1 14 I                                                  15.1 4i Amendment 15

ABWR awun Standard Plant avc I i ILLUSTRATIONS i Figure lhlt East 9i ! 15.1 1 Simplified Block Diagram of Fault Tolerant l Digital Controller System 15.1 15 l l l 15.1 2 Runout of One Feedwater Pump 15.1 16 15.1 3 Runout of Two Feedwater Pumps 15.117 l 15.1-4 Inadvertent Opening of One Bypass Valve 15.1 18 15.15 Inadvertent Opening of All Control and Bypass Valves 15.1 19 i l f 0 15.1 di Amendment 15 l

ABWR wima Standard Plant nrv c 15.1 DECREASE IN REACI'OR COOLM TEMPERATURE Because this event is very slow, the operator r - 1.4.1.1 Lass of Feedwater Heating action will terminate this event. Therefore,

the worst event is the loss of feedwater heating 15.1.1.1 Identincation of Causes and resulting in a temperature difference just below

_ Frequency Classincation the AT setpoint. However, a loss of 55.60C l feedwater temperature is analyzed to bound this 15.1.1.1.1 Identincation of Causes event. A feedwater heater can be lost in at least two 15.1.1.1.2 Frequency Classincation ways: The probability of this event is considered (1) steam extraction line to beater is closed; low enough to warrant it being categorized as an or infrequent incident. However, because of the lack of a sufficient frequency data base, this (2) steam is bypassed around heater, transient disturbance is analyzed as an incident 7 of moderate frequency. The first cass produces a gradual cooling of the feedwater. in the second case, the steam 15.1.1.2 Sequence of Events and Systems bypasses the heater and no heating of that Operation feedwater occurs. In -ither case, the reactor vessel receives cooler feedwates. The maximum 15.1.1.2.1 Sequence of Events number of feedwater heaters which can be tripped or bypassed by a single event represents the most Table 15.11 lists the sequence of events severe transient for analysis considerations, for this transient. This event has been conservatively estimated to O incur a loss of up to 55.60C of the feedwater heating capability of the plant and causes an 15.1.1.2.1.1 Identification of 0perator Actions increase in core inlet subcooling. This increases core power due to the negative void Because no scram occurs during this event, no reactivity coefficient. However, the power immediate operator action is required, As soon increase is slow, as possible, the operator should verify that no e operating limits are being exceeded. Also, the The feedwater control system (FWCS) includes a operator should determine the cause of failure l logic intended to mitigate the consequences of a . prior to returning the system to normal. _ loss of feedwater beating capability. The system will be constantly monitoring the actual 15.1.1.2.2 Systems Operation feedwater temperature and comparing it with a reference temperature. When a loss of feedwater In establishing the expected sequence of heating is detected (i.e., when the difference events and simulating the plant performance, it between the actual and reference temperatures was assumed that normal functioning occurred in exceeds a AT setpoint, which is currently set the plant instrumentation and controls, plant at 16.70C), the FWCS sends an alarm to the protection and reactor protection systems, operator. The operator can then take actions to mitigate the event. This will avoid a scram and The high simulated thermal power trip (STPT) reduce the o CPR during the event, scram is the primary protection system trip in mitigating the consequences of this event. However, the power increase in this event is not high enough to initiate this scram. Operation of engineered safeguard features (ESP) is not O Amendment 15 15.1 1 . l

21A610aAD Standard Plant nry c expected for the transient. 15.1.2 Feedwater Controller Failure . Maximum Demand 15.1.1.3 Core and 5,vstem Perfonnance 15.1.2.1 Identincatloa of Causes and 15.1.1.3.1 Input Paranaters and initial Frequency Classincation Conditions 15.1.2.1.1 Identincation of Causes The transient is simulated by programming a change in feedwater enthalpy corresponding to a This event is postulated on the basis of a 55.60C loss is feedwater beating. single failure of a control device, specifically one which can directly cause an increase in 15.1.1.3.2 Results coolant inventory by increasing the feedwater flow. Because the power increase during this event is relatively slow, it can be treated as a quasi The ABWR feedwater control system uses a steady. state transient. The 3.D core simulator, triplicated digital control system, instead of a has been used to evaluate this event for the single channel analog system as used in current equilibrium cycle. The results are summarized in BWR designs (BWR 2 6). The digital systems Table 15.12. consist of a triplicated fault. tolerant digital controller, the operator control stations and The MCPR response of this event is small due displays. The digital controller contains three to the mild thermal power increase with shifting parallel processing channels, each containing l axial shape. The worst A CPR response is 0.07, the microprocessor based hardware and associated software necessary to perform all the control No scram is initiated in this event. The calculations. The operator interface provides l increased core inlet subcooling aids thermal information regarding system status and the margins. Nuclear system pressure does not change required control functions. significantly (less than 0.4 Kg/Cm 2) and consequently, the reactor coolant pressure Redundant transmitters are provided for key boundary is not threatened. process inputs, and input voting and validation are provided such that faults can be identified 15.1.1.4 Barrier Performance and isolated. Each system input is triplicated , internally and sent to the three processing As noted previously the consequences of this channels. (See Figure 15.11) The channels event do not result in any temperature or will produce the same output during normal pressure transient in excess of the criteria for operation. Interprocessor communication which the fuel, pressure vessel or containment provides self diagnostic capability. A two out-are designed; therefore, these barriers maintain of three voter compares the processor outputs to } their integrity and function as designed. generate a validated output to the control actuator. A separate voter is provided for each 15.1.1.5 Radiological Consequences actuator. A 'ringback* feature feeds back the I final voter output to the processors. A voter Because this event does not result in any fuel failure will thereby be detected and alarmed, failures or any release of primary coolant to in some cases a protection circuit will lock the either the secondary containment or to the actuator into its existing position promptly environment, there are no radiological after the failure is detected. consequences associated with this event. O Amendment 13 13.1 2

l ABM 23uioun h=Aard Plant RrV. C Table 15.13 lists the failure modes of a remaining feedwater pump will decrease to offset O triplicated digital controt system and outlines the effects of each failure. Because of the triplicated architecture, it is possible to take the increased flow of the failed pump. The effect on total flow to the vessel will not be significant. The worst additional single fail-one channel out of service for maintenance or ute would cause both feedwater pumps to run out repair while the system is on line. Modes 2 and to their maximum capacity. However the proba. 5 of Table 15.13 address a f ailure of a bility of this to occur is extremely low (less component while an associated redundant component than 7 x 10** failure per reactor year). is out of service. This type of failure could potentially cause a system failure. However, the 15.1.2.1.2 Frequency Classification probability of a component failure during servi. cing of a counterpart component is considered to 15.1.2.1.2.1 Runout of One Feedwater Pump be so low that these failure modes will not be considered incidents of moderate frequency, but Although the frequency of occurrence for this will be considered limiting faults. event is less than once per 100 reactor years, this event is conservatively evaluated as an Adverse effects minimization is mentioned in incident of moderate frequency, the effects of Mode 2. This feature stems from the additional intelligence of the system 15.1.2.1.2.2 Runout of Two Feedwater Pumps provided by the microprocessor. When possible, the system will be programmed to take ection in The frequency of occurrence for this event is the event of some failure which will reduce the estimated to be less than once per 10000 years. severity of the transient. For example, if the it should be classified as a limiting fault as total steam flow or total feedwater flow signals specified in Chapter 15 of Regulatory Guide failed, the feedwater control system will detect 1.70. Nonetheless, since the corisequence of this by the input reasonability checks and this event has no significant impact on the automatically switch to one element mode (i.e., operating CPR limit, the criteria of moderate O' control by level feedback only). The level control would essentially be unaffected by this frequent incidents are conservatively applied to this event. failure. 15.1.2.2 Sequence of Events and Systems The only credible single failures which would Operation lead to some adverse affect on the plant are-Modes 6 and 7, a failure of the output voter and 15.1.2.2.1 Sequence of Events a control actuator failure. Both of these failures ~would lead to a loss of control of only 15.1.2.2.1.1 Runout of one Feedwater Pump one actuator (i.e., only one feedwater pump with _ increasing flow). A voter failure is detected by With momentary increase in feedwater flow, the ringback feature. The FWCS will initiate a the water level rises and then settles back to lock up of the actuator upon detection of the its normal level. Table 15.14 lists the failure. The probabilities of failure of the sequencing of events for Figure 15.12. variety of control actuators are very low based on operating experience (less than 0.0088 15.1.2.2.1.2 Runout of Two Feedwater Pumps failures per reactor year), in the event of one pump run.out, the FWCS would then reduce the With excess feedwater flow, the water level demand to .the remaining pump,- thereby rises to the high Icvel reference point, at automatically compensating for the excessive flow which time the feedwater pumps and the main from the failed pump. Therefore, the worst turbine are tripped and a scram is initiated, single failure in the feedwater control system Table 15.15 lists the sequence of events for causes a run out of one feedwater pump to its Figure 15.13. The figure shows the changes in maximum capacity. However, the demand to the important variables during this transient. O Amendment 15 15.1 3

ABWR muun Ramadad Plant _ uvc j l 15.1.2.2.1.3 Identification of Operator l M ions 15.1.2.2.13.1 Ruseet of One Feedwater Pump O , 1 Because no scram occurs for runout of one feedwater pump, no immediate operator action is l l l l l O l t li l 1 O Amendment 15 13,1,3 3 l l

ABM ux6iman Standard Plant uvc required. As soon as possible, the operator assumed to be 75% of spted flow at the design [_] V should verify that no operating limits are being pressure of 74.9 kg/cm'g. The total feedwater exceeded. Also, the operator should determine flow for both pumps runout is assumed to be 130% l the cause of failure prior to returning the of rated at the design pressure of 74.9 system to normal. kg/cm2g. 15.1J.2.1J.2 Runout of Two Feedwater Puseps 15.1.2.3J Results The operator should: 15.1J.3.2.1 Runout of One reednter Pump m observe that high feedwater pump trip has The simulated runout of one feedwater pump terminated the failure event; event is presented in Figure 15.12. When the increase of feedwater flow is sensed, the (2) switch the feedwater controller from auto to feedwater controller starts to command the manual control to try to regain a correct remaining feedwater pump to reduce its flow output signal; and immediately. T1 c vessel water level increases slightly (about 6 inches) and then settles back , (3) identify causes of the failure and report to its normal leve!. Tge vessel pressures only all key plant parameters during the event. increase about 0.1 kg/cm . MCPR remains above j the safety limit. 15.1.2.2.2 Systems Operation 15.1.2.3.2.2 Runout of Two Feedwater Pumps 15.1.2.2.2.1 Runout of One Feedenter Pump The simulated runout of two feedwater pumps Runout of a single feedwater pump requires no accident is shown in Figure 15.13. The high l protection system or safeguard system operation, water level turbine trip and feedwater pump trip This analysis assumes normal functioning of plant are initiated at approximately 18 seconds. l -( p) instrumentation and controls. Scram occurs and limits the neutron flux peak v and fuel thermal transient so that no fuel 15.1.2.2.2.2 Runout of Two Feedwater Pumps damage occurs. It is calculated that the MCPR is right at the safety limit. Therefore, the To properly simulate the expected sequence of design limit for the moderate frequent incident events, the analysis of this event assumes normal is met. The turbine bypass system opens to functioning of plant instrumentation and limit peak pressure.,in the steamline near the controls, plant protection and reactor protection SRVs to 82.8kg/cm*g and the pressure gt the systems. Important system operational actions bottom of the vessel to about 84.9 kg/cm g. for this event are high level tripping of the main turbine and feedwater pumps, scram and The level will gradually drop to the Low recirculation pump trip (RPT) due to turbine Level reference point (Level 2), activating the trip, and low water level initiation of the RCIC system for long term level control. reactor core isolation cooling (RCIC) system to maintain long term water level control following The applicant will provide reanalysis of this tripping of feedwater pumps. event for the specific core configuration. 15.1.2.3 Cost and System Performance 15.1.2.4 Barrier Performance 15.1.2.3.1 Input Parameters and initial As previously noted the consequence of this Conditions event does not result in any temperature or pressure transient in excess of the criteria for The runout capacity of one feedwater pump is which the fuel, pressure vessel or containment l are designed; therefore, these barriers maintain Amendment t$ 15J.4

r 21A6100AH Et==dard Plant uv c their integrity and function as designed. turbine control valves and bypass valves could (n be f. illy opened. However, the probability of 15.1.2.5 Radiological Consequences this event to occur is extremely low (less than 7 x 10.s failure per reactor year), and hence , While the consequences of this event do act the event is considered as a limiting fault. ! result in any fuel failures, radioactivity is nevertheless discharged to the suppression pool 15.1 3.1.2 Frequency Classincation as a result of SRV actuation. However, the mass input, and hence activity input, for this event 15.1 3.1.2.1 Inadvertent Opealog of One is much less than those consequences identified Turbine Spass Valve in Subsection 15.2.4.5 for Type 2 events. Therefore, the radiological exposures noted in This transient disturbance, estimated to Subsection 15.2.4.5 cover the consequences of occur less than 0.0088 times per year, is this event, conservatively categorized as one of modsate l frequency. 15.1.3 Pressure Regulator Failure- Open 15.1J.l.2.2 Inadvertent Opening of all 15.13.1 Identincation of Causes and Frequency Turbine Control Valves and Bypass Valves l Classincations The frequency of occurance for this event is 15.1 3.1.1 Identincation of Causes estimated to be less than once per 10000 years. It should be classified as a limiting fault as The ABWR steam bypass and pressure control specified in Chapter 15 of Regulatory Guide system (SB&PCS) uses a triplicated digital 1.70. Nonetheless, since the consequence of control system instead of an analog system as this event has no significant impact on the used in current BWR desians (BWR 2 6). The operating CPR limit, the criteria of moderate SB&PCS controls turbine control valves and frequent incidents are conservatively applied to l

 -(   turbine bypass valves to maintain reactor         this event.

pressure. As presented in Section 15.1.2.1.1, no credible single failure in the control system 15.13.2 Sequence of Events and Systems will result in a maximum demand to all actuators Operation for all turbine control valves and bypass valves. A voter or actuator failure may result 15.13.2.1 Sequence of Events in an inadvertent opening of one turbine control valve or one turbine bypass valve. In this case, 15.1 3.2.1.1 Inadver1ent Opening of One the SB&PCS will sense the pressure change and Turbine Bypass Yalve command the remaining control valves to close, and thereby automatically mitigate the transient Table 15.16 lists the sequence of events and maintain reactor power and pressure, for Figure 15.14. Because the effect of sudden opening of one 15.1 3.2.1.2 landver1ent Opening of All bypass valve, which bypasses about 11% of rated Turbine Control Valves and Bypass Valves , steam flow when full opened is more severe than sudden opening of one turbine control valve, Table 15.17 lists the sequence of events for i which is almost wide open at rated power, it is Figure 15.15. assumed for purposes of this transient analysis ; that a single failure causes a single bypass 15.13.2.13 Identincation of 0perator ! valve to fall open. Actions I As presented in Section 15.1.2.1.1, multiple 15.13.2.13.1 Inadvertent Opening of one failures might cause the SB&PCS to erroneously Turbine Bypass Valves issue a maximum' demand to all turbine control valves and bypass valves. Should this occur, all Because no scram occurs during this event, r o Amendment 15 l' .l.5 1

    - _ -   .   . ..        _. _ . . . - . . ~ . -                   + . - -   . ~ . . , , ...     .          ._.

J ABM zwiooxii ffandard Plant REY C l immediate operator action is required. As soon l as possible, the operator should verify that no i operating limits are being exceeded. Also, the operator should determine the cause of failure j prior to returning the system to normal. l 15.1J.2.13.2 Inadvertent Opening of All Turbine Control Valves and Bypass Valves e, (' e, l Amendment 15 15.1 5.1 l

  .                       ,                        _ . . . _ . _ _ _         _                            . . _ . ~

ABM 234sioorti Standard Plant RTV C If the reactor scrams as a result of the setpoint. Normal startup and actuation can O- isolation caused by the low pressure at the take up to 30 seconds before effects are turbine inlet (58 kg/cm'g) in the run mode, the realized, following sequence of operator actions is expected during the course of the event. Once if these events occur, they will follow isolation occurs, the pressure willincrease to a sometime after the primary concerns of fuel point where the SRVs open. The operator should: thermal margin and overpressure effects have occurred, and are expected to be less severe (1) monitor that all rods are in; than those already experienced by the system. (2) monitor reactor water level and pressure; 15.1.3.3 Coet and Sptem Performance (3) observe turbine coastdown and break vacuum 15.1.3.3.1 Input Parameters and initial before the loss of steam seals. Check Conditions turbine auxiliaries; A five second isolation valve closure (4) observe that the reactor pressure relief instead of a 3.0 second closure is assumed when valves open at their setpoint; the turbine pressure decreases below the l turbine inlet low pressure setpoint for main (5) observe that RCIC initiated on low. water steamline isolation initiation. This is within level; the specification limits of the valve and represents a conservative assumption. (6) secure RCIC when reactor pressure and level are under control; 15.1.3.3.2 Results (7) monitor reactor water level and continue 15.1.3.3.2.1 Inadvertent Opening of One cooldown per the normal procedure; and Turbine Bypass Valve (8) complete the scram report and initiate a The simulated inadvertent opening of one maintenance survey of the SB&PCS before turbine bypass valve is presented in Figure reactor restart. 15.14. When the decrease in reactor pressure is sensed, the pressure control system starts 15.1.3.2.2 Systems Oleration immediately to command turbine control valves to close to maintain the reactor pressure. The 15.1.3.2.2.1 Inadgertent Opening of one Turbine vessel water level increases slightly (about Hypass Yahe four inches) and then settles back to its normallevel. Ryctor pressure decreases by This event doen not require any protection about 0.7 kg/cm'. MCPR remains above the system or safeguard system operation. This safety limit, analysis assumes normal functioning of plant instrumentation and contrels. 15.1.3.3.2.2 Inadsertent Opening of All Turb!ne Control Valves and Bypass Yahes 15.1.3.2.2.2 Inadvertent Opening of All Turbine Control Valves and Bypass Valves Figure 15.15 presents graphically how ihe high water level turbine trip and the isolation To properly simulate the expected sequence of valve closure stops vessel depressuriration and events, the analysis of this event assumes normal produces a normal shutdown of the isolated functioning of plant instrumentation and reactor, controls, plant protection and reactor protection systems, except as otherwise noted, Depressurization results in formation of voids in the reactor coolant and causes a Initiation of RCit gstem functions occurs decrease in reactor power almost immediately, O when the ve'.sel water leul reaches the L2 The depressurization rate is large enough such Amendmen. t$ 15.16

l 21A6100AD mem=dard Plant myc that water level swells to the sensed level trip this event. (3 V l setpoint (L8), feedwater pumpinitiating mainswitches trips. Position turbine on the and 15.1.4 Inadvertent Safety / Relief Valve Opening l turbine stop valves initiate reactor scram and a trip of 4 RIPS. i 15.1.4.1 Identincation of Causes and l After a presssurization resulting from the Frequency Classification I turbine stop valve closure, pressure again drops , and continues to drop until turbine iniet 15.1.4.1.1 Identification of Causes l pressure is below the low turbine pressure ' isolation setpoint when main steamline isolation Cause of inadvertent opening is attributed to

          - finally terminates the depressurization. The                 malfunction of the valve or an operator turbine trip and isolation limit the duration and           initiated opening. It is therefore simply severity of the depressurization so that no postulated that a failure occurs and the event                                    l significant thermal stresses are imposed on the             is analyzed accordingly. Detailed discussion of reactor coolant pressure boundary. No the valve design is provided in Chapter 5.

significant reduction in fuel thermal margins occur; therefore, this event does not have to be 15.1.4.1.2 Frequency Classification analyzed for specific core configurations. This transient disturbance is categorized as 15.13.4 harrier Performance an luftequent incident. Barrier performance analyses were not required 15.1.4.2 Sequence of Events and Systems because the consequences of this event do not Operation result in any temperature or pressure transient in excess of the criteria for which fuel, 15.1.4.2.1 Sequence of Events t l pressure During thevessel event of or containment inadverteut openingare of all designed, Table 15.18 lists the sequence of events for

  \                                                                      this event, l           turbine control and bypass valves, peak pressure in thq bottom of the vessel reaches 81.8 kg/cm'g. wh{ch is below the ASME code limit of                15.1.4.2.1.1 Identification of Operator 96.7 kg/cm g for the reactor coolant pressure Actions                                                                         !

boundafy Vessel dome pressure reaches 80.4 kg/cm g, below the setpoint of the second The plant operator must reclose the valve as pressure relief group. Minimum vessel dome soon as possible and check that reactor and T O l pressure of 50.6 kg/cm'g occurs at about 40 output return to normal. If the valve cannot be seconds, closed, plant shutdown should be initiated. 15.1.3,5 Radiological Consequences 15.1.4.2.2 Systems Operation While the consequences of this event do not This event assumes normal functioning of , result in any fuel failures, radioactivity is normal plant instrumentation and controls, nevertheless discharged to the suppression pool specifically the operation of the pressure as a result of SRV actuation. However, the mass regulator and level control systems, input, and hence activity input, for this event is much less than those consequences identified 15.1.43 Cost and System Performance ,- in Subsection 15.2.4.5 for Type 2 events, t Therefore, the radiological exposures noted in The opening of one SRV allows steam to be Subsection 15.2.4.5 cover the consequences of Amendment l$ 15.1 7

ABWR mmu Standard Plant krv c discharged into the suppression pool. The sudden 15.1.6 Inadvertent RHR Shutdown Cooling increase in the rate of steam flow leaving the Operation reactor vessel causes a mild depressurization transient. 15.1.6.1 Identification of Causes and Fnquency Classincation l decrease and within a few seconds closes theThe SB&PCS senses the nuclear system 15.1.6.1.1 Identification of Causes l turbine control valves for enough to stabilize the reactor vessel pressure at a slightly lower At design power conditions, no conceivable l value and the reactor settles at nearly the malfunction in the shutdown cooling system could initial power level. Thermal margins decrease cause . .cmperature reduction, only slightly through the transient, and no fuel darnage results from the transient. MCPR is in startup or cooldown operation, if the essentially unchanged and , therefore, the safety reactor were critical or near critical, a very limit margin is unaffected and this event does slow increase in reactor power could result. A not have to be reanalyzed for specific cose shutdown cooling malfunction leading to a configurations, moderate temperature hrease could result from misoperation of the cooling water centrols for , 15.1.4.4 Itarrier Performance the RHR heat exchangers. The resulting temperature decrease would cause a slow i As presented previously, the transient insertion of positive reactivity into the core, resulting from a stuck open relief valve is a If the operator did not act to control the power mild depressurization which is within the range level, a high neutron flux reactor scram would l of normal load following and therefore has no terminate the transient without violating fuel significant effect on RCPB and containment design thermal limits at j without any measurable pressure limits. increase in nuclear system pressure. 15.1.4.5 Radiological Consequences 15.1.6.1.2 Frequency Classincation l result While the consequence of this event does not Because no single failure could cause this in fuel failure, it does result in the event, it is categorized as a limiting fault, discharge of normal coolant activity to the l suppression pool via SRV operation. Because this 15.1.6.2 Sequence of Eients and Systems activity is cor.tained in the primary contalnment, Operation there will be no exposures to operating personnel. Because this event does not result in 15.1.6.2.1 Sequence of Eients an uncontrolled release to the environment, the plant operator can choose to leave the activity A shutdown cooling malfunction leading to a bottled up in the containment or discharge it to moderator temperature decrease could result from the environment under controlled release misoperation of the cooling water controls for conditions. If purging of the containment is RHR heat exchangers. The resulting temperature I chosen, the release will be in accordance with decrease causes a slow insertion of positive the established technical specifications; reactivity into the core. Scram occurs before , therefore, this event, at the worst, would only any thermal limits are reached if the operator I result in a small increase in the yearly does not take action. The sequence of events integrated exposure level, for this event is shown in Table 15.19. l 15,1.5 Spectrum of Steam System Piping 15.1.6.2.2 System Operation l Failures inside and Outside Containment in a PWR A shutdown cooling malfunction causing a moderator temperature decrease must be This event is not applicable to BWR plants. considered in all operating states. However, O Amendment 15 15.1 4

i MM ElainA= d Plant 2M610MD RN C l l= -s this event is not considered while at power j operation because the nuclear system pressure is i l'(}' too high to permit operation of the shutdown cooling mode of the RHRs. , i No unique safety actions are required to avoid unacceptable safety results for transients as a ] result of a reactor coolant temperature decrease induced by misoperation of the shutdawn cooling heat exchangers. In startup or cooldown operation, where the reactor is at or near ! critical, the slow power increase resulting from ' the cooler moderator temperature is controlled by the operator in the same manner normally used to l l control power in the startup range. 1 15.1.6.3 Cort and System Performance The increased subcooling caused by

l. misoperation of the RHR shutdown cooling mode ;

could result in a slow power increase due to the I reactivity insertion. This power rise is terminated by a flux scram before fuel thermal limits are approached. Therefore, only l qualitative description is provided here and this i event does not have to be analyzed for specific j f7 core configuration.

l. &

I 1 15.1.6.4 Barrier Performan<v As previously presented, the consequences of l this event do not result in any temperature or pressure transient in excess of the criteria for ; j which the fuel, pressure vessel or containment are designed; therefore, these barriers maintain l their integrity and function as designed. i [ 15.1.6J Radiological Consequences l Because this event docs not result in any fuel f ailures, no analysis of radiological consequences is required for this event, j! i l l O ! I Amendment 15 t$.19 '{ l

msimin A.BWR

           .a . e . ..                                                                                                                    m.c Table 15.1 1

( ' SEQUENCE OF EVENTS FOR IESS OF FEEDWATER HEATING TIME (nasl EVENT 0 Initiate a 55.60C temperature reduction in the feedwater system 5 Initial effect of unheated feedwater starts to raise core power level 100(est.) Reactor variables settle into new steady state Table 15.12 LOSS OF $5.60C FEEDWATER HEATING l IlliC* to EOC'

4 I

tO Change in V- Core Power (%) 12.8 l Change in MCPR 0.07 l

l. BOC = Beginning of Qvle EOC = End of Qvie i-l l

l [ b Amendment 15 15.1 10

ABM zwiooxii Etandard Plant RIV A Table 15.13 SINGLE FAILURE MODES FOR DIGITAL CONTROLS 9t MODES DESCRII' TION EFFECTS

1. Criticalinput failure None.

Redundant transmitter takes over. Operator informed of failure

2. Input failure while one Possible system failure. Adverse ,

sensor out of senice effects minimized when possible

3. Operator switch single None-contact failure Trigicated contacts

4. Processor channel failure None-Redundant processors maintain control; Operator informed of failure

5. Processor failure while one System failure channel out of senice

6. Voter failure loss of controf of one actuator (l.c., one feedwater pump only) FWCS willlock up actuators

7. Actuator failure less of one actuator (i.e., One feedwater pump only) 1 O

i 15.111 l l

ABWR mama Standard Plana arv c Table 15,14 C

SEQUENCE OF EVEh"Ill FOR FIGURE 15.12 TIME (sec) EVENTS 0 Initiate simulated runout of one feedwater pump (at system design pressure of 74.9 kg/cm's the pump runout flow is 75% of rated feylwater flow) l

                     -0.1          Feedwater controller starts to reduce the feedwater flow from the other feedwater pump 16.6        Vessel water level reaches its peak valve and starts to return to its normal value                                                                           l
                     ~60 (est.)    Vessel water level returns to its normal value.                                          -

l-I.- Table 15.15 SEQUEN( .E OF EVENTS FOR FIGURE 15.13 TIME (nec) EVEN' (' 0 Initiate simulated runout of ty feedwater pumps (130% at system ( design pressure o'74.9 kg/cm g on feedwater flow) r 18.35 L8 vessel icvel setpoint initiates trips main turbine and feedwater pumps. 10.36 Reactor scram and trip of 4 RIPS are actuated by stop valve position switches 18.5 Main turbine bypass vahrs opened due to turbine trip l 20.1 SRVs open due to high pressure l l > 25 SRVs close l

                      > 40 (est.) Water level dropped to low water level setpoint (f.evel 2)
                      > 70 (est.) RCIC flow into vessel (not simulated)

Amendment 15 15.1 12

MN 23A6100ni Standard Plant uv. e Table 15.16 g SEQUENCE OF EVENTS FOR FIGURE 15.14 TI M E bec) EVENTS 0 Simulate one bypass valve to open

                      ~ 0.5                             Pressure control system senses the decrease of reactor pressure and 4

commands control valves to close 5.0 0.c actor sectics at another steady state Table 15.17 SEQUENCE OF EVENTS FOR FIGURE 15.15 TIME (sec) EVENTS 0 Simulate all t9rbine control valves at.d bypass valves to rpen. e 2.8 Turbine control valves wide open. I 2.87 Vessel water level (L8) trip initiates main turbine and E feedwater turbine trips. - 2.9 Main turbine stop valves reach 85% open position and initiates reactor scram and trip of 4 RIPS. e 2.97 Turbine stop valves closed. 17.2 Vessel water level reaches L.2 wtpoint. The remaining 6 RIPS are tripped. RCICis initiated. 36.2 Low turbine inlet pressure trip initiates main steamline isolation 41.2 Main steam isolation valves closed. Bypass valves remain open,

                                                          . exhausting steam in steamlines downstream ofisolation valves.

47.2 (est.) RCIC flow enters vessel (not simulated). O Amendment 15 15.113

7 O O O w

                                                                                                                                    ;g     -

l a. E . - 3

                         .                                                                                                         l
                                                                                                                                   ~
                                            'C ANALOG AND DISCRETE *

CONTROLLER 1 INPUTS .

                      , , , , ,              m E                      m                        --

ANALOG AND DISCRETE

r ACTUATORS AND CONTROLLER 2 VOTERS r DISCRETE PERMISSIVES INPUTS .m.

m ! OPERATORS PNL eDISPLAYS elNDICATORS !

ALARMS +
RECORDERS ;

e

ANALOG AND DISC 1ETE ANALOG AND

                                          +                            CONVERSION TO EACH                                                     [

DISCRETE *

CONTROLLER 3 t INPUTS = ** INTERPROCESSOR e COMMUNICATION LINKS AND i REDUNDANT CLOCK LOGIC !

sme-u -l

         - Figure 15.1-1 SIMPLIFIED BLOCK DIAGRAM OF FAULT-TOLERANT DIGITAL CONTROLLER SYSTEM y

a t s 5"E l i V

                    ~
                               <   ~
                                              ,    7      F
                                                               ~

F' "

                                                                                         <     w _ _ _ . _- _~-_ _ _ _       ________.n_

                     >                                                                 1 ' NEUTRON F LUX -

j! FA > ' 2 PEAK FUEL CENTER TEMP

3. AVE SURFACE HEAT FLUX Eg a.

150 4 FEEDWATER FLOW 125 2#

~#

{ o i ;5 VESSELSTEAM FLOW n G a Q J.. 1 VESSEL PRES RISE (PSI) bj T tr 100 E'S '

                                                                                                        " E3S                          75                           2 STM LINE PRES RISE (PSI)               F g

3 TURBINE PRES RISE (PSI) ~ 5 g 4 RELIEF VALVE FLO?/ (PCT)- z 5 BYPASS VALVE FLOW (PCT)

                                 $                                                                                                                                  6 TURB STEAM FLOW (PCT) a 50                                                                                        25
                                                                                                                                            - tnm         Pais              pr                  l'36 0            *-      i                                                        1        -25         '                                                        ~

0 4 8 12 16 0 4 8 12 16 TIME (sec) TIME (sec) 1 LEVEL (INCH-R EF-SEP-SKIRT) 1 VOID REACTIVITY 2 W R SENSED LEVEL (INCHES) 2 DOPPLER REACTIVITY 150 3 N R SENSED LEVEL (INCHES) 1 3 SCRAM REACTIVITY 4 CORE INLET FLOW (PCT) 4 'sOTAL.7EACTIVITY 5 PUMP FLOW 3 (PCT) G "S " 4 ' 100 -"-~45 0 '"3 --_ _ . b 1 - _ _ _ 31 . 5 g 73 23 ?3

                                                                                                                                  $q 1                   1              1
                                                 ...j l                                             0        -  t-                                                               *
                                                                                                                                      -2     -  '-   -                                         1

O 4 8 12 16 0 4 8 12 16

                                                                                   -TIME (sec)                                                                    TIME (sec)                                     y I                     E

87-226-14 59 f, 8 Figure 15.1-2 RUNOUT OF ONE FEEDWATER PUMP n>"

O O O

w. . .--

                                                                                                                                                                                                                           ,~s
                                                                                                                                                                                                                          \,_,

r m J s i 2 1% it u y 125 i { c. # o# C o I G f; dh 2 S j .a 5---~ __ _La r,_ 1 "3;- , g

   <   100                                          -.                                  4'(4                                             75         1 VESSEL PRES R:SE (PSI)                  f'                                         -

(O l 2 STM LINE PRES RISE (PSI) j 3 TURBINE PRES RISE (PSI) y 1 NEUTRON FLUX 4 RELIEF VALVE FLOW (PCT) uJ 2 PEAK FUEL CENTER TEMP 5 BYPASS VALVE FLOW (PCT) e ' g 50 3 AVE SURFACE HEAT FLUX 4 - 25 6 TURD STEAM FLOW (PCT) - . --

   @               4 FEEDWATER FLOW                                                                                                                                                                                         k
              - I 5 VESSEL STEAM FLOW                                                                                                          17us             ty_ e            fy                  )

0.........l 0 5 10 15 20 ( " '- -25 0 5 10 15 20 TIME (sec) TIME (sec) 1 LEVEL (INCH-REF-SEP-SKIRT) 2 W R SENSED LEVEL (INCHES) I ' 3 N R SENSED LEVEL (INCHES) 4 CORE INLET FLOW (PCT) 5 PUMP FLOW 3 (PCl) k 100 : 5 0 I*' _~N M j -- h ! o 1 VOID REACTIVITY h

50 ' -' & b -1 2 DOPPLER REACTIVITY I

                                                              #                               N                                      E                3 SCRAM REACTIVITY p                                                                                                                                _

4 TOTAL REACTIVITY 0''''5 0 10 15 20

                                                                                                    *                                   -2 '****

0 5 10 15 20 TIME (sec) TIME (sec) ,, r 90-21509 qf y Figure 15.1-3 RUNOUT OF TWO FEEDWATER PUMPS

                                                                                                                                                                                                                                         <s os

1 I L l 1 VESSEL PRES RISE (PSI) FA ) g 1 NEUTRON FLUX 2 STM LINE PRES RISE (PSI) Eg g 2 PEAK FUELCENTER TEMP 3 TURBlNE PRES RISE (PSI) Oj' 3 AVE SURFACE HEAT FLUX 125 4 RELIEF VALVE FLOW (PCT) 1# 2 150 g G o 4 FEEDWATER FLOW 5 VESSELSTEAM FLOW 5 BYPASS VALVE FLOW (PCT) G TURB STEAM FLOW (PCTI Nh 1

                                                                                                     \F                   8              E              F 2

E 100 Q -

                               '?             -I ? -        -
                                                                  !) ' - - - - - - - -         75 O

k a

      '2  50                                                                                   25 5              5              5            5

_ u u u u

                                                                                                          -          T              T           3P
                                                                                   *                      " '                                                  4 O                                                                              -25     "

0 10 20 30 40 0 10 20 30 40 TIME (sec) TIME (sec) 1 LEVEL (INCH-REF-SEP-SKIRT) 1 VOID REACTIVITY 2 W R SENSED LEVEL (INCilES) 2 DOPPLER REACTIVITY 3 N R SENSED LEVEL (INCilES) 1 3 SCRAM REACTIVITY 150 4 TOTAL REACTIVITY 4 CORE INLET FLOW (PCT) 5 PUMP FLOW 1(PCT)

                                                                                                                      'M8            'Mi         nn 100     '6               '6                '85              %                     $    o   S' 2                                                                                 I U

6

I 50 A -

2 _i

t t t 0- ' "' -2 - 0 10' 20 30 4ti 0 10 20 30 40 +J TIME (sec) TIME (sec) 5b E i 87-22sts h<$a j Figure 15.1-4. INADVERTENT OPENING OF ONE BYPASS VALVE } O O O a

                                                                                                                                                                           ,a . _                       .2                                                                .,%

[ 1 NEUTRON FLUX '

                                                                                                                              ' 2 PEAK FUEL CENTER TEMP 1 VESSEL PRES RtSE (PSI)  'N 2
                                                                                                                                                                                                                                                 - 2 STM LINE PRES RISE (PSI)

{. a 150 3 AVE SURFACE IEAT FLUX - 150 m 3 3 TURBNE PRES RISE (PSI) 5.Wj 4 FEEDWATER FLOW - 4 RELIEF VALVE FLOW (PCT) C 5 VESSEL STE AM Fn O', Q N . 5 BYPASS VALVE FLOW (PCT) c o , e -iI \_ s 6 TUR8 STEAM FLOW (PCT) N $

                                                          .m                                                                                                                                                                     Yd\ ,         nn             6  eN            E
                                                                                                                                                                                                                /                                                              E E

o 50 -

i

                                                                                                                                                                                               -150                                       \          s 5                                                         -

0 0

                                                               ....                             k.-  4 10 i

20 h 30

                                                                                                                                                . ls-a m 40

_300 ......... 0 10 20 30 40 TIME (sec) TIME (sec) 1 LEVEL (NCH-REF-SEP-SKIRT) 2 W R SENSED LEVEL (INCHES) 150 3 N R SENSED LEVEL (INCHES) 1 4 CORE INLET FLOW (PCT) 7 3 ._

5 PUMP FLOW 1 (PCT) 100 - ; $ 0 j x j s. 8 P 50 7{ 1 k -l 1 VOID REACTIVITY 2 DOPPLER REACTIVITY 3 jr p 3 SCRAM REACTIVITY L_ 4 5 L s - 4 TOTALREACTIVITY l 0 '- 3 3 -2~. *- - l

O 10 20 30 40- 0 to 20 30 40 TIME (sec) TIME (sec) f3 90-215-10

Figure 15.1-5 OPENING OF ALL CONTROL-BYPASS VALVES o?

23A6100AB Standard Plant , an e

   ,s                                                              SECTION 15.2 i    \

U CONTENTS (Continued) . Section 31 Llc P.ase 15.2 3.4.2 Turbine Trip with Failures of One Bypass Valve 15.2 7 , 15.23.43 Turbine Trip with Failure of All Bypass Valves 15.2-7 15.2 3.5 Radiological Consequences 15.2 7 15J.4 MSIV Closures 15.2 7 15.2.4.1 Identification of Causes and Frequency Classification 15.2-8 15.2.4.1.1 Identification of Causes 15.2-8 15.2.4.1.2 Frequency Classification 15.2 8 15.2.4.1.2.1 Closure of All Main Steamline Isolation Valves 15.2-8 15.2.4.1.2.2 Closure of One Main Steamline Isolation Valve 15.2 i 15.2.4.2 Sequence of Events and Systems Operations 15.2 8

  ,A-l?- Q                15.2.4.2.1                      Sequence of Events                              15.2-8 15.2.4.2.1.1                    Identification of Operator Actions              15.2 8 15.2.4.2.2                      Systems Opcration                               15.2-8 15.2.4.2.2.1.                   Closure of All Main Steamline Isolation Valves  15.2 l                    15.2.4.2.2.2                    Closure of One Main Steamline Isolation Valve ~ 15.2 9 l-15.2.4 3                        Core and System Performance                     15.2 9 ~
                  - 15.2.4 3.1                      Input Parameters and initial Conditions         15.2 9 -

15.2.4 3.2 Results 15.2-9 15.2.4 3.2.1 Closure of All Main Steamline Isolation Valves 15.2 9 L l i 15.2 vi Amendment 15 s -

                                  , , - - - - ~

M 23A6100AD Standard Plant RIW. C1 _, SECTION 15.2 CONTENTS (Continued) h Sectlon M Egge

                       - 15.2.4 3.2.2                    Closure of One Main Steamlie Isolation Valve -    15-2-9
                       ' 15.2.4.4                        Barrier Performance                               15.2-10 15.2.4.4.1                     Closure of All Main Steamline Isolation Valves    15.2 10 15.2.4.4.2                     Closure of One Main Steamline Isolation Valve -   15.2 10 15.2.4.5                      Radiological Consequences                          15.2 10 15.2.4.5.1                    General Observations                               15.2 10.
                       -15.2.4.5.2                      Depressurization Shutdown Evaluation               15.2 10
  ;                      15.2.4.5.2.1-                  Fission Product Release from Fuel                  15.2 10                 .

l :15.2,4.5.2.2 Fission Product Release to Environment 15.2 10.1-

                       '15.2.4.5 3 -

Radiological Exposures 15.2-11

                                                                                                                                 -t 15.2.5                         Loss of Condenser Vacuum -                         15.2-11 15.2.5.1                      Identification of Causes and Frequency Classification                                      15.211 ~
    /

15.2.5.1.1 Ideatification of Causes 15.2 11' k 15.2.5.1.2 Frequency Classification '15.2-11 15.2.5.2 Sequence of Events and Systems Operations' 15.2 j 15.2.5.2.1 Sequence of Events ~ 15.2 ,

];.

15.2.5.2.1.1 Identification of Operaor Actions 15.2-11 15.2.5.2.2 . Systems Operation 15.2 11 ! [ , 15.2.5.3 Core and System Performance 15.2-11' g

        ,                15.2.5.3.1 '                  Input Parameters and initial Conditions '          15.2-11 l

l: 15.2 vil , l Amendment 9 5

MM 23A6100AD Standard. Plant mvc : SECTION 15.2 ,

   ;ilh                                         CONTENTS (Continued) 1 Section                                  Dilt                                East                        i 15.2.5.3.2         Results                                                   15.2 12 15.2.5.4          Barrier Performance                                        15.2 12                     l l

15.2.5.5 Radiological Consequences 15.2 12 l m ; 15.2.6 . Imss of Non Emernency Offshe AC Power to i Station Auxiliades 15.2 12 J 15.2.6.1 Identification of Causes and Frequency Classification 15.2 12 15.2.6.1.1 Identification of Ca ses 15.2 12 15.2.6.1.1.1 Loss of Unit AuxiliaryTransformers . 15.2 12 15.2.6.1.1.2 Loss of Grid Connections 15.2 13 15.2.6.1.2 Frequency Classification 15.2 13 - 1 15.2.6.1.2.1 Loss of Unit Auxiliary Transformer 15.2 13 v-

                     - 15.2.6.1.2.2       Loss of Grid Connections                                  15.2 13 b

c 15.2.6.2 - Sequence of Events and Systems Operations 15.2 13 -, 15.2.6.2.1 Sequence of Events 15.2 13 - 15.2.6.2.1.1' Loss of Unit AuxiliaryTransformer 15.2 0

15.2.6.2.1.2. Loss of Grid Connections 15.2-13

                                                                                                                           .i i

15.2.6.2,1.3 .!dentification of Operator Actions 15.2 14 t 15.2.6.2.2' - Systems Operation ~ 15.2 14 i

  ' ,                                                          15.2 viii I   :

Amendment 15 - t

( MM. 23A6100AB

        ~ Standard Plant                                                                         nrv. c SECTION 15.2 CONTENTS (Continued)                                               h Section                               Ihlt                         Eajs '

15.2.6.2.2.1- Loss of Unit Auxiliary Transformer 15.2 14 ' l 15.2.6.2.2.2 Loss of Grid Conncetions 15.2-14 15.2.6 3 Core and System Performance 15.2 14 15.2.6 3.1 Loss of Unit . Auxiliary Transformer 15.2-15 l l< 15.2.6 3.2 Loss of Grid Connections 15.2 15 15.2.6.4- Barrier Performance 15.2-15 15.2.6.4.1 Loss of Unit Auxiliary Transformer 15.2 15 15.2.6.4.2 Loss of Grid Connections' 15.2 15

,                  15.2.6.5      Radiological Consequences                           .15.2-U                    ,

I 15.2.7- Loss of Feedwater Flow 15.2 15 15.2.7.1' Identification of Causes and Frequency; Classification 15.2-15' 15.2.7.1.1 Identification of Causes 15.2 15 15.2.7.1.2 Frequency Classification . 15.2 15 15.2.7.2 Sequence of Events and Systems Operations 15.2-16

15.2.7.2.1. Sequence of Events 15.2-16

                  .15.2.7.2.1.1  Identification _of Operator Actions.                 15.2-16 '
       , Amendment 15 i

ABM us63aoso Standard Plant RN A

w. SECTION 15.2 U CONTENTS (Continued)

Section Tillt East 15.2,7.2.2 Systems Operation 15.2 16 15.2.7.3 Core and System Performance 15.2 16 i 15.2.7.4 - Barrier Performance 15.2 16 15.2.7.5 Radiological Consequences 15.2 16 15.2.8 Feedwater une Break 15.2 16 15.2.9 Fallust of RHR Shutdown Coollnn 15.2 16 15.2.10 References 15.2 17 TABLES Table 1111C East

   >M                   15.2 1             Sequence of Events for Figure 15.21.                 15.2 18 V-.                                                                                                                 '

15.2 2 Sequence of Events for Figure 15.2 2 15.2 18 l 15.2 3 Sequence of Events for Figure 15.2 3 15.2 19 15.2-4 Sequence of Events for Figure 15.2 4 15.2-19 15.2 5 Sequence of Events for Figure 15.2-5 15.2-20 ( 15.2-6 < Sequence of Events for Figure 15.2-6 15.2 20 . 15.27 Sequence of Events for Figure 15.2 7 15.2 21 l1 15.2-8 Sequence of Events for Figure 15.2-8 15.2 21 15.2 9 Sequence of Events for Figure 15.2 9 - 15.2 22 l 15.2 10 Post-Transient Release Rate to the Containment

u. with Suppression Pool Cleanup 15.2-23 15.2 11 Activity Released to the Environment 15.2 24 l: ~

L 3: 1x) 1 l 15.2-x

                                                                                                                         -l l

l. l: -

MM 23A6100AD Standard Plant Rev. c SECTION 15.2 TABLES (Continued) ma m e.ge 15.2 12 Estimated Doses and Atmospheric Dispersion Factors 15.2 24

                                 '15.2 13     Typical Rates of Decay Ior Condenser Vacuum    15.2 25
                       ,           15.2 14     Sequence of Events for Figure 15.2-10         15.2 25 15.2-15     Trip Signals Associated with the loss of Condenser Vacuum                              15.2-26 15.2 16     Sequence of Events for Figure 15.211          15.2 27 15.2-17     Sequence of Events for Figure 15.212          15.2 28 ILLUSTRATIONS Figure                            Title .                  Ease 15.2 1-      Clo.ute of One Turbine Control Valve          15.2 30 15.2-2      Pressure Regulator Downscale Failure -        15.2-31 15.2-3      Generator Load Rejestion with' Bypass         15.2 32 15.2-4      Load Rejection with One Bypass Valve Failure  15.2-33 15.2 5      Load Rejection with All Bypass Failure        15.2 34 [

15.2-6 Turbine Trip with Bypass 15.2 35 15.2 7 Turbine Trip with One Bypass Valve Failure 15.2 36

                                   .15.2 8 -  ' Turbine Trip with All Bypass Failure 15.2-37 15.2-9 .   .'MSIV Closure Direct Scram                    15.2-38
                                   '15.2 10      Loss of Condenser Vacuum                     15.2 39 15.2    Loss of AC Power                              15.2-40 15.2 xi Amendment 15

ABM 2miman Standard Plant RTIV. C v-SECTION 15.2 uf ' ILLUSTRATIONS (Continued) Figure Dile East i 15.2 12 loss of All Feedwater Flow 15.2 42 It o j; ( , y

l ':: 1' fi l' l' ! t 15.2.xii Ms -

                                 - Amendment IS l

MM 23A6tooAn Standard Plant RN C 15.2 INCREASE IN REACTOR T RESSURE control processors, called ' pressure regulator downscale failure.' However, the probability of 15.2.1 Pressure Regulator Failure. Closed this event to occur is extremely low (less than 7x10 failure per reactor year), and hence 15.2.1.1.identincation of Causes and Frequency the event is considered as a limiting fault. Classincation 15.2.1.1.2 Frequency Classincation 15.2.1.1.1 Identincation of Causes 15.2.1.1.2.1 Inadvertent Closs.e of One Turbine The ABWR steam bypass and pressure control ControlValve system (SB&PCS) uses a triplicated digital control system, instead of an analog system as This event is conservatively treated as a used in BWR/2 through BWR /6. The SB& PCS controls moderate frequency event, although the turbine control valves and turbine bypass valves voter / actuator failure rate is very low (0.0088 to maintain reactor pressure. As presented in- failure per reactor year). Subsection 15.1.2.1.1, no credible single failure in the control system will result in a minimum 15.2.1.1.2.2 Pressure Regulator Donscale demand to all turbine control valves and bypass Failure l valves. . A voter or actuator failure may result in an inadvertent closure of one turbine control The probability d muce of this event is valve or one turbine bypass valve if it is open calculated to be less than 7x10-5 per year as - at the time of failure, in this case, the SB&PCS shown in Appendix ISD. This event is treated as will sense the pressure change and command the a limiting fault, remaining control valves or bypass valves, if needed, to open, and thereby automatically 15.2.1.2 Sequence of Events and System mitigate' the transient and try to maintain Operation reactor power and pressure. L 15.2.1.2.1 Inadvertent Closure of One Turbine Because turbine bypass valves are normally Control Valve

                             - closed during normal full power operation, it is assumed for purposes of this transient analysis         Postulating a voter / actuator feilure of the that a single failure causes a single turbine SB&PCS as presented in Subsection 15.2.1,1.1 control valve to fait closed. Should this event will cause one turbine control valve to close, occur at full power, the opening of remaining- The pressure will increase, because the reactor control valves may not be sufficient to maintain is still generating the initial steam flow. The the reactor pressure, depending on the turbine SB&PCS will open the remaining control valves
                                                                             ~

design. Neutron flux will increase due to void and some bypass valves. This sequence of events collapse resulting from the pressure increase, A is listed in Table 15.2-1 for Figure 15.2-1. reactor' scram will be initiated when the high flux scram setpoint is exceeded. 15.2.1.2.1.2 Pressure Regulator Downscale Failure . No single failure will cause the SB&PCS to issue erroneously a minimum demand to all turbine Table 15.2 2 lists the sequence of events control valves and bypass valves. However, as- for Figure 15.2 2. discussed in Subsection 15.1.2.1.1, multiple failures might causr the SB&PCS to fail and - 15.2.1.2.1 3 Identincation of Operator erroneously issue a minimum demand. Should this Actions

                              -occur, it would cause full closure of turbine controls valves as well as an inhibit of steam        The operator should:

bypass flow and thereby increase reactor power and pressure. When this occurs, reactor scram (1) monitor that all rods are in; will be -initiated when the high reactor flux scram setpoint is reached. This' event is (2) monitor reactor water level and pressure; analyzed here as the simultaneous failure of two Amendment 15 15.2 1 1

ABM m6mn ,

 ,              Standard Plant                                                                   wv. c     .l a.

(3) ~ observe turbine coastdown and break vacuum ; before the loss of steam seals (check L f 3 i-i t i

i e; Amendment 15 g, l 1

ABM 23xsioorn Standard Plant uv. c valve is pesented in Figure 15.21. The

    'O~            turbine auxiliaries);                        analysis assumes that about 85% of rated steam G                                                        flow can pass through the remaining three i-(4). observe that the reactor pressure relief turbine control valves.

valves open at their setpoint; Neutron flux increases rapidly because of the , (5) monitor reactor water level and continue void reduction caused by the pressure increase, cooldown per the normal procedure; and When the sensed neutron flux reaches the high neutron flux scram setpoint, a reactor scram is (6) complete the scram report and initiate a initiated. The neutron flux increase is limited maintenance survey of pressure regulator to 139 % NBR by the reactor scram. Peak fuel before reactor restart, surface heat flux does not exceed 103.1% of its-initial value. MCPR for this transient is still 15.2.1.2.2 Systems Operation above the safety MCPR limit. Therefore, the design basis is satisfied. See Subsection 15.2.1.2.2.1 Inadvertent Closure of One Turbine 15.0.4.5 for additional evaluation of results. Control Valve 15.2.1.3.2 Pressurt Regulator Downscale Normal plant instrumentation and control are Failure assumed to function. This event takes credit for high neutron flux scram to shut down the reactor. A pressure regulator downscale failure is simulated at 102% NBR power as shown in Figure After a closure of one turbine control valve, 15.2 2. the team flow rate that can be transmitted through the remaining three turbine control Neutron flux increases rapidly because of the valves depends upon the turbine configuration. void reduction caused by the pressure increase, , For plants with full arc turbine admission, the When the sensed neutron flux reaches the high N. steam flow through the remaining three turbine neutron flux scram setpoint, a reactor scram is O control valves is at least 95% of rated steam initiated. The neutron flux increase is limited flow. On the other hand, this capacity drops to to 155% NBR by the reactor scram. Peak fuel about 85% of rated steam flow for plants with surface heat flux does not exceed 103% of its partial are turbine admission. Therefore, this initial value. It is estimated less than 0.2% transient is less severe for plants with full arc of rods will get into transition boiling, turbine admission. In this analysis, the case Therefore, the design limit for the limiting with partial arc turbine admission is analyzed to - fault event is met.

           - cover all plants.

15.2.1.4 Barrier Performance b l 15.2.1.2.2.2 Pressurt Regulator Downscale Failure 15.2.1.4.1 Inadvertent Closure of One Turbine s Control Valve Analysis of this event assumes normal functioning of plant instrumentation and ' Peag pressure at the SR valves reaches 74.5 0 . controls, and plant protection and reactor pro- kg/cm g. The peak yessel bottom pressure

            .tection systems. Specifically, this event takes reaches 78.2 kg/cm g, beloy the transient credit for high neutron flux scram to shut down pressure limit of 96.7 kg/cm g.

the reactor. High system pressure is limited by >. the pressure relief valve system operation. 15.2.1.4.2 Pressure Regulator Downscale Failure l 15.2.1.3 Core and System Performance Peag pressure at'the SRVs reaches 85.1 kg/cm g. The peagnuclear system pressure 15.2.1.3.1 Inadvertent Closure of one Turbine reaches 87.4 kg/cm g at the bottom of the Control Valve vessel, below the nuclear barrier pressure ' limit.' g A simulated closure of one turbine control l Amendment 15 15.2 2 1 > c

MM 2aAstooAn Standard Plant ,_ av. e ij 15.2.1.5 Radiological Consequences

While the consequences of this event do not result in any. fuel failures, radioactivity is nevertheless discharged to the suppression pool as a result of SRV actuation. However, the mass input, and hence activity input for this event is much less than those consequence identified in Subsection 15.2.4.5 (for a Type 2 event). Therefore, the radiological exposures ncted in l t i O Amendment 15 15.2-2.1 - l

MM 2.wlooAn Standard Plant Rt'v. c Subsection 15.2.4.5 cover the consequences of 15.2.2.1.2.3 Generator lead Rejection with this event. Failutt of All Bypass Valves 15.2.2 Generator lead Rejection 4 Frequency: <3.6x10 / plant year 15.2.2.1 Identification of Causes and Frequency Frequency Basis: Thorough search of domestic Classification plant operating records have revealed three instances of bypass failure during 628 bypass 15.2.2.1.1 Identification of Causes system operations. This gives a probability of bypass failure of 0.0048. Combining the actual Fast closure of the turbine control valves frequency of a generator load rejection with the

                  '(TCV) is initiated whenever electrical grid failure rate of bypass yields a frequency of a disturbances occur which result in significant generator load rejection with bypass failure of loss of electrical load on the generator. The 0.0036 event / plant year. With the triplicated turbine control valves are required to close as       fault. tolerant design used in ABWR, this failure      r rapidly as possible to prevent excessive              frequency is lowered by 'at least a factor of overspeed of the turbine generator (T G) rotor.       100. Therefore, this event should be classified Closure of the main turbine control valves will       as a limiting fault, however, criteria for
                  -cause a sudden reduction in steam flow, which         moderate frequent incidents are conservatively results in an increase in system pressure and         applied.

reactor shutdown. 15.2.2.2 Sequence of Events and System After sensing a significant loss of electrical Operation load on the generator, the turbine control valves p - are commanded to close rapidly. At the same 15.2.2.2.1 Sequence of Events l, time, the turbine bypass valves are signaled to i

   -h .         : open in the ' fast" opening mode by the Steam         15.2.2.2.1.1 Generator Load Rejection-Turbine Bypass and Pressure Control System (SB&PCS),

Q' which uses a triplicated digital controller. As Control Valve Fast Closure

                 . presented in Subsection 15.1.2.1.1, no single            A loss of generator electrical load from high failure can cause all turbine bypass valves fall' power conditions produces the sequence of events to open on demand. The worst single failure can       listed in Table 15.2 3.

only cause one turbine bypass valve fail to open ' on demand. Therefore, the probability of this to 15.2.2.2.1.2 Generator lead Rejection with

               ~ occur is very low (less than one failure every 11      Failure of One Bypass Valve
               - year). Therefore, generator load rejection with L                . failure of one turbine bypass valve is considered         A loss of generator electrical load from            '

an infrequent event; while generator load

high power conditions with failure of one bypass l- . rejection with failure of all turbine bypass . valve produces the sequence of events listed in _ valves is a limiting fault. Table 15.2 4. 1 15.2.2.1.2' Frequency Classification 15.2.2.2.1.3 Generator Load Rejection with

                                 .                                     Failure of All Bypass Valves                             l l                  15.2.2.1.2.1 Generator Load Rejection l:                                                                           A loss of generator electrical load at high R                      This event is categorized as an incident of power with failure of all bypass valves produces l mode' 'c frequency,                                  the sequence of events listed in Table 15.2-5, 15.2.2.1.2.2 Generator Load Rejection with           15.2.2.2.1.4 Identification of Operator
               ' Fallure of One Bypass Valve                           Actions This event should be categorized as an The operator should:
   -Cy           infrequent event. However, criteria for moderate
   'V            frequent incidents are conservatively applied.        (1) verify proper bypass valve performance; Amendment 15                                                                                           112-3

_____=-.__-_.____x_____--__--__-- - -

ABWR m63mn Standard Plant anv. c (2) observe that the feedwater/ level controls i

          ' have maintained the reactor water level at a satisfactory value;'

(3) observe that the pressure regulator is controlling reactor pressure at the desired value; (4) observe reactor peak power and pressure ; and (5) verify relief valve operation. l l' l 1 0 L t-I= 0; Amendment 15 15.2-1 1 l^

                -AB M
               , Standard Plant -

zwiooxo uv. c

               ' 15.2J 2.2 System Operation                             assumed such that the valves operate in the full
      /-                                                                arc (FA) mode and have a full stroke closure (x)           15.2.2.2.2.1 Generator lead Rejection with Bypass time, from fully open to fully closed, of 0.15 sec.

To properly simulate the expec'ed sequence of events, the analysis of this event assumes normal functioning of plant instrumentation and controls, plant protection and reactor protection

               ~ systems unless stated otherwise.

The reactor is operating in the manual Turbine control valve (TCV) fast closure flow. control mode when load rejection occurs. initiates a scram trip signal for power levels Results do not significantly differ if the plant greater than 40% NB rated. In addition, a trip had been operating in the automatic flow control of four of ten RIPS is, initiated. Both of these mode, trip signals satisfy the single. failure criterion and' credit is taken.for these protection The bypass valve opening characteristics are features, simulated using the specified delay together - with the specified opening characteristic The pressure relief system, which operates the required for bypass system operation, relief valves independently when system pressure exceeds relief valve instrumentation setpoints, Events caused by low water level trips such is assumed to function normally during the time as an initiation of the RCIC core cooling system

period analyzed. function is not required. Should this event occur, it will follow sometime after the primary.

All plant control systems maintain normal concerns of fuel thermal margin and overpressure l ; operation unless specifically designated to the effects have occurred, and is expected to bc

      ;  [      contrary.                                              less severe than those already experienced by l

A.-- _ the system. l: 15.2.2.2.2.2 Generator Load Rejection with l Failure of One Dypass Valve ~ 15.2.23.2 Results l: Same as Subsection 15.2.2.2.2.1, except that 15.2.23.2.1 Generator Load Rejection with

           ; failure of one main turbine bypass valve is _ Bypass                                                                          -
              -assumed for the entire event.

Figure 15.2 3 shows the results of the 15.2.2.2.2 3 Generator Load Rejection with ' generator trip from the 102% rated power

           ' Failure 6f All Bypass Valves                             conditions. Peak neutron flux rises 48% above ~

4 ,

                                 .                                    NB rated conditions.

L Same as Subsection 15.2.2.2,2.1, except that

                                   ~
           = failure of all turbine bypass valves;is assumed i

The a'verage fuel surface heat flux shows no- l for the entire event. increase from its initial value, and MCPR does not significantly decrease below its initial

           ' 15.2.23 Core and System Performance .                    value. Therefore, this event does not have to 15.2.2 3.1 Input Parameters and initial 1 Conditions                                             15.2.23.2.2 Generator Load Rejection with Failure of One Bypass Valve
                  ' The turbine electrody'draulic control system
           -(EHC) detects load rejection before a measurable              Figure 15.2 4 shows that, for the case of one turbine speed change takes place.                      bypass valve failure, peak neutron flux reaches about 155% of rated, and the average fuel O-           The closure characteristics of the TCVs are        su face heat flux still shows no increase from -

I Amendment 15 15.2-1

ABWR m 6>oorn

Standard PInnt arv. c value. 15.2.4.5. Therefore, the radiological exposures j .

noted in Subsection 15.2.4.5 for Type 2 exposure

         -v             The MCPR for this event is above the safety        cover these consequences of this event, limit. Therefore, this event does not have to be analyzed for a specific core configuration,           15.2.3 TurbineTrip 15.2.23.23 Generator Lead Rejection with              15.2J.1 Identification of Causes and Frequency Failure of All Bypass Valves                          Classification Figure 15.2 5 shows that, for the case of all      15.23.1.1 Identification of Causes bypass valves failure, peak neutron flux reaches about 185% of rated, and average surface heat            A variety of turbine or nuclear system flux reaches 102.3% of its initial value. The         malfunctions will initiate a turbine trip. Some MCPR for this event is right at the safety limit     examples are moisture separator and heater drain                   <

and meets the criteria for moderate frequent tank high levels,' large vibrations, operator incidents. The event should be analyzed for a lockout, loss of control fluid pressure, low specific core configuration, condenser vacuum and reactor high water level. ; 15.2.2.4 Barrier Performance After the main turbine is tripped, turbine bypass valves are opened in their fast opening 15.2.2.4.1 Generator Load Rejection mode by the SB&PCS. As presented in Subsection 15.2.2.1.1, any single failures can only cause Peal pressure at the SRVs reaches 82.7 one bypass valve fail to open on demand. Only kg/cm;g. The peak vessel bottom pressure multiple failures can cause all bypass valves 2

                  . reaches 84.7 kg/cm .g,;below the transient fail to open on demand.

2 pressure. limit of-96.7 cg/cm g. c f 15.23.1.2 Frequency Classification ,

            !      '15.2.2.4.2 Generator Load Rejection with                         .

Failum of One Rypass Valve 15.23.1J,1. Turbine Trip ,

.)

L Peng pressure at the SRVs reaches 83,6 This transient is categorized as an incident kg/cm g. The peak vessel pressure at tpe of moderate frequency In defining the

bottom of the vessel reaches 85.8 kg/cm g, frequency of this event, turbine trips which below the pressure limit. ~ occur as a byproduct of other transients such as L= loss of condenser vacuum or reactor high level 15.2.2.4 3 Generator Load Rejection with trip events.are not included. However, spurious Failum of All Bypass Valves low vacuum or high level trip signals which -

cause an: unnecessary turbine trip are included Peap pressure at the SRVs reaches 85.6' in' defining the frequency. To get an accurate L kg/cm g. The peak nuclear2 system pressure event.by-event frequency breakdown, this type of

                 . reaches 87.7 kg/cm g at the bottom of the di5ision of initiating causes is required.                                     ;

vessel, below the pressure limit.. 15.23.1.2.2 Turbine Trip with Failure of One

      ,.            15.2.23 Radiological Consequences                     Bypass Valve -

While the consequences of the event's This' event is conservatively considered as an f

                ' identified previously do not result in any fuel        incident of moderate frequency f ailures, radioactivity is nevertheless
                  .' discharged to the suppression pool as a result of   15.23.1.23 Turbine Trip with Failure of All SRV actuation. However, the mass input, and          Bypass Valves ,

hence activity input, for this event is much less ;

                 . than.those consequences identified in Subsection          This disturbance should be categorized as a l

Amendment 15 15.2 5

MM 23A61ooAn Standard Plant RN C limiting fault. Frequency is as follows: maintain pressure for restart purposes; Frequency: <6.4x10 / plant year (5) put the mode switch in the startup position before2the reactor pressure decays to <59.8 Frequency Basis: As presented in Subsection kg/cm g; 15.2.2.1.2.3, the failure rate of the bypass is 0.0048. Combining this with the turbine trip (6) secure the RCIC operation if auto initiation frequency of 1.22 events / plant year yields the occurred due to low water lever; frequency of 0.0064/ plant year. The ABWR design reduces this frequency by at least a factor of (7) monitor control rod drive positions and the 100. However, criteria for moderate frequent SRNMS; l incidents are conservatively applied. (8) investigate the cause of the trip, make 15.23.2 Sequence of Events and Systems repairs as necessary, and complete the scram Operation report; and 15.2 3.2.1 Sequence of Events (9) cool down the rector per standard procedure if a restart is not intended. 15.2 3.2.1.1 Turbine Trip 15.2 3.2.2 Systems Operation Turbine trip at high power produces the sequence of events listed in Table 15.2 6. 15.2.3.2.2.1 Turbine Trip 15.2.3.2.1.2 Turbine Trip with Failure of One All plant contr01 systems maintain normal Bypass Valve operation unless specifically designated to the contrary. Turbine trip at high power with failure of one bypass valves produces the sequence of events Turbine stop valve closure initiates a listed in Table 15.2 7. reactor scram trip via position signals to the protection system. Credit is taken for 15.2.3.2.1 3 Turbine Trip with Fallure of All successful operation of the reactor protection Bypass Vahes system. Turbine trip at high power with failure of all Turbine stop valves closure initiates a trip bypass valves produces the sequence of events of four RIPS, thereby reducing the core flow, listed in Table 15.2 8. The pressure relief system which operates the 15.2.3.2.1.4 Identification of Operator Actions relief valves independently when system pressure exceeds relief valve instrumentation setpoints The operator should: is assumed to function normally during the time period analyzed. (1) verify auto. transfer of buses supplied by generator to incoming power (if automatic 15.2.3.2.2.2 Turbine Trip with Failure of One transfer does not occur, manual transfer Bypass Valve must be made); Same as Subsection 15.2.3.2.2.1, except that (2) monitor and maintain reactor water level at a failure of one bypass valve is assumed. required level; 15.2.3.2.23 Turbine Trip with Failure of All (3) check turbine for proper operation of all Bypass Valves auxiliaries during coastdown; Same as Subsection 15.2.3.2.2.1, except that (4) depending on conditions, initiate normal operating procedures for cooldown, or Amendment 15 15.2-6

ABM 234siooxa Standard Plant REV.C failure of all main turbine bypass valves is value, and averge surface heat flux reaches A assumed for the entire transient time period 100.6% of its initial value. Therefore, this

  -h        : analyzed,                                             transient is less severe than the generator load.

rejectiors with failure of bypass transient 15.2.3.3 Core and System Performance ' presented in Subseetion 15.2.2.3.3.2. . 15.2.33.1 Input Parameters and laltlal 15.2.3.4 Barrier Performance L Conditions 15.2.3,4.1 Turbine Trip Turbine stop valves full stroke closure time is 0.1 seconds. Peak pressure ig the bottom of the vessel reaches 84.6 kg/cm g whichis below the ASME A reactor scram is f aitiated by position code limit of 96.7 kg/cm g for the reactor switches on the stop valves when the valves are coolant pressure boundary, yessel dome pressure > less that 85% open, does not exceed 83.0 kg/cm g. The severity of l turbine trips from lower initial power levels Reduction in core recirculation flow is decreases to the point where a scram can be initiated by position switches-on the main stop avoided if auxiliary power is available from an valves, which actuate trip circuitry which trips external source and the power level is within the four of the reactor internal pumps, bypass capability. 1

           ' 15.2.3.3.2 Results                                    15.2.3.4.2 Turbine Trip with Failure of One Bypass Valve 15.2.33.2.1 Turbine Trip Peak pressure af the bottom of the vessel A turbine trip with the bypass system            reaches 85.6 kg/cm g while yssel dome pressure .

operating normally is simulated at 102% NBR power does not exceed 84.1 kg/cm g. 2B th are below

  ' VO        conditions as shown in Figure 15.2-6.                the pressure limit of 96.7 kg/cm g.

l- Neutron flux increases rapidly because of the 15.2.3.4.3 Turbine Trip with Failure of All L void reduction caused by the pressure increase. Bypass Valves However, the flux increase is limited to 122% of rated by the stop valve scram and the trip of The S/R valves open and close sequentially as

        .l    four RIPS.' Peak fuel surface heat flux does not     the stored energy is dissipated and the pressure exceed its initial.value. : Therefore, this event     falls below the setpoints of the valves. Pcak 2

does not have-to be reanalyzed for a specific nuclear system pressure reaches 87.7 kg/cm g at l

           ' core configuration ~                                ' the vessel bottom; therefore, the overpressure l      ,                                                            event is below the-reactor coolant pressure                        i o

15.233.2.2 Turbine Trip with Failure of One boundary pressure limig. Peak dome pressure does - ,

Bypass Valve . not exceed 86.1 kg/cm g.. i Same ' as S ubsection '15.2.3.3.2.1, except the peak ncutron flux is 132% of rated. This event 15.2.3.5 Radiological Consequences ) is shown in Figure 15.2-7. While the consequences of this event do not . result in any fuel failures, radioactivity is- ll L 15.2.3.3.2.3 Turbine Trip with Failure of All nevertheless discharged to the suppression pool _as " ,j Bypass Valves a result of SRV actuation. However, the mass input, and hence activity input,' for this event is 'I A' turbine trip with failure of the bypass less than those consequences identified in system is simulated at 102% NBR power conditions Subsection 15.2.4.5 for a Type 2 event. There-in Figure 15.2 8.~ fore, the radiological exposures noted in Section 15.2.4.5 cover the consequences of this event. t p= Peak neutron flux reaches 159% of its rated V 15.2.4 MSIVClosures Amendment !$ 15,2 7

                                                                                                                                       .l
                                                                                                                                    '1

                                                                                                                 -1 ABM                                                                                             2346ioorn 1

l Standard Plant RIM C ' 15.2.4.1 Identification of Causes and 15.2.4.2 Sequence of Events and Systems Frequency Classification Operation j 15.2.4.1.1 Identification of Causes 15.2.4.2.1 Sequence of Events i l Various steamline and nuclear system Table 15.2 9 lists the sequence of events for j malfunctions, or operator actions, can initiate Figure 15.2 9. main steamline isolation valve (MSIV) closure. l Examples are low steamline pressure, high 15.2.4.2.1.1 Mentification of 0perator Actions steamline flow, high steamline radiation, low water level or manual action. The following is the sequence of operator actions expected during the course of the event, 15.2.4.1.2 Frequency Classification assuming no restart of the reactor. The operator should: 15.2.4.1.2.1 Closure of All Main Steamline Isolation Yalves (1) observe that all rods have inserted; This event is categorized as an incident of (2) observe that the relief valves have opened for moderate frequency. To define the frequency of reactor pressure control; l this event as an initiating event and not the byproduct of another transient, only the (3) check that RCIC auto starts on the impending following contribute to the frequency: manual low reactor water level condition; action (purposely or inadvertent); spurious signals such as low pressure, low reactor water ;4) switch the feedwater controller to the manual level, low condenser vacuum; and, finally, position; equipment malfunctions such as faulty valves or operating mechanisms. A closure of one MSIV may (5) secure RCIC when the reactor vessel level has cause an immediate closure of all the other MSIVs recovered to a satisfactory level; depending on reactor conditions. If this occurs, it is also included in this category. During the (6) initiate RHR operation when the reactor main'steamline isAtion valve closure, position pressure has decayed sufficiently; L switches on the mes provide a reactor scram if the valves in-two or more main steamlines are (7) determine the cause of valve closure before less than 85% open (except for interlocks which resetting the MSIV isolation; permit proper plant startup). Protection system logic, however, permits the test closure of one (8) observe turbine coastdown and break vacuum valve without initiating scram from the position before the loss of sealing steam (check T.G switches. auxiliaries for proper operation); 15.2.4.1.2.2 Closure of one Main Steamline (9) check that conditions are satisfactory prior Isolation Valve to opening and resetting MSIVs; and This event is categorized as an incident of (10) survey maintenance requirements and l moderate frequency. One MSIV may be closed at a complete the scram report, time for testing purposes; this is done

manually. Operator error or equipment 15.2.4.2.2 Systems Operation malfunction may cause a single MSIV to be closed l inadvertently. If reactor power is greater than-15.2.4.2.2.1 Closure of All Main Steamline about 80% when this occurs, a high flux scram may Isolation Valves l ~ result (if all MSIVs close as a result of the single closure, the event is considered as a MSIV closures initiate a reactor scram trip l closure of all MSIVs). via position signals to the protection system. O Amendment 15 112-8

ABM 2u6ioaro Standard Plant Rn c Credit is taken for successful operation of 15.2.4 3.2.1 Closure of All Main Steamline (] the protection system. Isolation Valves V The pressure relief system which initiates Figure 15.2 9 shows the changes in important opening of the relief valves when system pressure nuclear system variations for the simultaneous exceeds relief valve instrumentation setpoints is isolation of all main steamlines while the reactor assumed to function normally during the time is operating at 102% of NBR power. Neutron flux period analyzed, increases slightly, and fuel surface heat flux shows no increase. All plant control systems maintain normal operation unless specifically designated to the Four RIPS are tripped due to high pressure. l contrary. Water level decreases sufficiently to cause a trip of remaining 6 RIPS and the initiation of the RCIC l 15.2.4.2.2.2 Closure of One Main Steamifne system on the Level 2 (L2) trip at some time Isolation Valve greater than 10 seconds. However, there is a delay up to 30 seconds before the water supply l A closure of a single MSIV at any given time enters the vessel. Nevertheless, there is no will not initiate a reactor scram directly. This change in the thermal margins. Therefore, this is because the valve position scram trip logic is event does not have to be reanalyzed for specific designed to accommodate single valve operation core configurations. and testability during normal reactor operation at limited power levels. Credit is taken for the 15.2.4 3.2.2 Closure of One Main Steamline operation of the pressure and flux signals to Isolation Valve initiate a reactor scram.

Only one isolation valve is permitted to bc l All plant contral systems maintain normal closed at a time for testing purposes to prevent operation unless specifically designated to the scram. Normal test procedure requires an initial g$ contrary, power reduction to approximately 75 to 80% of

 ;d                                                      design conditions in order to avoid high flux 15.2.43 Core and System Performance                scram, high pressure scram, or full isolation from high steam flow in the

live" lines. With a 3 l 15.2.4 3.1 Input Parameters and Initial second closure of one main steam isolation valve Conditions during 102% rated power conditions, the steam flow disturbance may. raise vessel pressure and reactor The main steam isolation valves close in 3 to power enough to initiate a high neutron flux 5 seconds. The worst case (the 3 second closure scram. This transient is considerably milder than i time) is assumed in this analysis, closure of all MSIVs at full power. No l.

L quantitative analysis is furnished for this Position switches on the valves initiate a event. However, no significant change in thermal reactor scram when the valves are less than 85% . margins is experienced and no fuel damage occurs, open. Clostre of these valves causes the dome Peak pressure remains below SRV setpoints, pressure to increase. Four RIPS are tripped when Therefore, this event does not have to be the high pressure setpoint is reached. reanalyzed for specific core configurations. Inadvertent closure of one or all of the isolation valves while the reactor is shut down , (such as operating state C, as defed in Appendix l 15A) will produce no significai.t transient. l Closures during plant heatup (operatn g state D) will be less severe than the maximum po."er cases 15.2.43.2 Results (maximum stored and decay heat) presented in b V Amendment 15 15.2 9

1 l ABM ux6ioorn Standard Plant RIV C Subsection 15.2.4 3.2.1. to planned operation, little radiological impact results. 15.2.4.4 Barrier Performance (2) Where major RCPB equipment failure requires 15.2.4.4.1 Closure of All Main Steamline immediate plant shutdown and its attendant Isolation Valves depressurization under controlled shutdown time tables (4 hours), the radiologicalimpact The nuclear system relief valves begin to open is greater. l at approximately 2.9 seconds after the start of isolation. The valves close sequentially as the To envelope the potential radiological impact, stored heat is dissipated but continue to a worst case like example No. 2 is described discharge the decay heat intermittently. Peak below. However, it should be noted that most pressurp at the vessel bottom reaches 86.4 transients are like exarnple (1) and the kg/cm*g, below the pressure limits of the radiological envelope conservatively overpredicts reactor coolant pressure boundary. cak pressure the actual radiologicalimpact by a factor greater l in the main steamline is 84.1 kg/cm[g.than 100. 15.2.4.4.2 Closuit of One Main Steamline 15.2.4.5.2 Depressurization Shutdown isolation Yalve Evaluation No significant effect is imposed on the RCPB, 15.2.4.5.2.1 Fission Product Release from Fuel since, if closure of the valve occurs at an unacceptable high operating power level, a flux While no fuel rods are damaged as a consequence i or pressure scram may result. The main turbine of this event, fission product activity associated l bypass system continues to regulate system with normal coolant activity levels as well as pressure via the other three open steamlines, that released from previously defective rods will be released to the. suppression pool as a 15.2.4.5 Radiological Consequences consequence of SRV actuation and vessel depressurization. The release of activity from 15.2.4.5.1 Gener l Observations previously defective rods is based in part upon measurements obtained from operating BWR plants The radiologicalimpact of transients involves (Reference 1).

   . consequences which do not lead to fuel rod damage as a direct result of the event itself. Addi-            Because each of those transients identified tionally, many events do not lead to the depres-     presiously (which cause SRV actuation) will result surization of the primary system but only the        in various vessel depressurization and steam venting of sensible heat and energy via fluids at    blowdown rates, the transient evaluated in this coolant loop activity through relief valves.to section is that one which maximizes the the suppression pool. In the case of previously radiological consequences for all transients of             i defective fuel rods, a depressurization transient this nature. This transient is the closure of all will result in considerably more fission product main steamline isolation valves. The activity carryover to the suppression pool than hot- airborne in the containment is based on the standby transients. The time duration of the analysis presented in Reference 1. The results of t,2nsient varies from several minutes to more        these analyses are presented in Table 15.210, than four hours.                                     which was used in evaluating the radiological dose consequences in this section.

These observations lead to the realization that radiolo ,ical e aspects can-involve a broad 15.2.4.5.2.2 Fission Product Release to spectrum of results. For example: Environment (1) Transients where appropriate operator action Because this event _does not result in the (seconds) results in quick return (minutes) immediate need to purge the containment, it is O Amendment 15 t$.210

4 ABM 2mioorn mandard Plant nw c

                         .                  ;                                                                                l assumed that purging of the containment through    (1) verify auto transfer of buses supplied by         1 the SGTS occurs under average annual (Vo 1    meteorological conditions and commences 8 hours generator to incoming power if automatic transfer has not occurred, manual transfer l

I after initiation of the event. The SGTS must be made; I

             ' efficiency for lodine is 99% for organic forms                                                                l and 99.9% for other forms Reference 2 contains (2) monitor and maintain reactor water level at a description of the containment purge release         required level; model used. The lategrated release to the                                                                    !

environment is presented in Table 15.211. (3) check turbine for proper operation of all I auxillaries during coastdown; I 15.2.4J.3 Radiolofral Exposures (4) depending on' conditions, initiate normal The offsite radiological doses for this event operating procedures for cooldown, or maintain are presented in Table 15.212. It should be pressure for restart purposes; noted that the radiological doses in the table are exposures per event. For the isolation (5) put the mode switch in the STARTUP position transient, this event is not expected to occur beforey the reactor pressure decays to <59.8 more than 2.5 times per year; therefore, it is kg/cm g; conservative to assume the yearly commitments for these transients will be ~2.5 times the (6) secure the RCIC operation if the auto. Individual values. initiation occurred due to low water level; 15.2.514ss of Condenser Vacuum (7) monitor control rod drive positions and the SRNM; 15.23.1 Identification of Causes and Frequency Classification - (8) investigate the cause of the trip, make tA . repairs as necessa.y, and complete the scram Q 15.23.1.1 Identification of Causes report, and; Various system malfunctions which can cause a (9) cooldown the reactor per standard procedure if. 1 loss of condenser vacuum due to some single a restart is not intended, l equipment failure are designated in Table l

              '15.2 13.                                           15.23.2.2 Systems Operation                            ;

15.23.1.2 Frequency Classification in establishing the expected sequence of , events and ' simulating the plant performance, it c 1Although the frequency of occurrence of this was assumed that normal functioning occurred in i event is expcted to be infrequent, this event is the plant instrumentation and controls, plant categorized as an incident of moderate frequency, protection and. reactor protection systems. 15.2.5.2J Sequence of Events and-Systems . Tripping functions incurred by sensing main LOperation. turbine condenser vacuum are presented in Table .; 15.2 15. r 15.2.5.2.1' Sequence of Events 15.2.5.3 Core and System Performance

                  . Table .15.2414 lists the sequence of events for Figure 15.210..                                15.2J3.1 Input Parameters and Initial                  1

Conditions 15.2.5.2.1'.1 Identification of Operator Actions Turbine stop valves full stroke closure time l :The Operator should: is 0.1 seconds.

O N j Amendment 15 15.2-11

L L ABM us6ioorn Standard Plant . RW C l l A reactor scram is initiated by position identified previously 40 not result in any fuel I switches on the stop valves when the valves are failures, radioactivity is revertheless discharged l: less that 85% open. This stop valve screm trip to the suppression pooi as a result of SRV signal is automatically bypassed when the reactor actuation. However, the mass input, and hence is below 40% NBR power level, activity input, for this event is much less than those consequences identified in Subsection The analysis presented here is a hypothetical 15.2.4.5; therefore, the radiological exposures l case with a conservative 5.1 centimeter Hg/sec noted in Subsection 15.2.4.5 for Type 2 events vacuum decay rate. Thus, the bypass system is cover the consequences of this event. l available for several seconds, because the bypass l I is signaled to cicse at a vacuum level of about 15.2.6 Loss of Non ernergency AC Power I' l 25.4 centimeters Hg less than the stop valve to Station Auxillaries closure. 15.2.6.1 Identification of Causes and Frequency 15.23.3.2 Results Classification l Under this hypothetical 5.1 centimeters 15.2.6.1.1 Identification of Causes L Hg/sec vacuum decay condition, the turbine bypass valves and main steamline isolation valve closure The non. emergency AC power to the station would follow main turbine trip about 5 seconds auxiliaries are provided by two unit auxiliary after it initiates the transient. This transformers. The unit auxiliary transformers are transient, therefore, is similar to a normal powered by the unit turbine / generator via a low turbine trip with bypass. The effect of main voltage generator breaker. Each unit auxiliary steamline isolation valve closure tends to be transformer provider power to two electrical buses l minimal, because the closure of main turbine stop which provide half of the unit's auxiliary loads, valves and subsequently the bypass valves have including the five reactor internal pumps (RIPS). already shut off the main steamline flow. Figure Following a generator trip arid during plant 15.210 shows the transient expected for this startup, the low voltage generate breaker is open . l event. It is asrumed that the plant is initially but the high voltage breaker t.s the =itchyard l operating at 102% of NBR power conditions. Peak remains closed to backfeco ,wer frora the grid to neutron flux reaches 122% of NBR power, while the unit auxiliary transformers, average fuel surface heat flux shows no increase. SRVs 'open to limit the pressure rise, then sequentially reclose as the stored energy is dissipated. Therefore, this event does not have to be reanalyzed for specific core configurations. 15.23.4 Barrier Performance Peak nuclear system pressure is 84.6 kg/cm#g at the vessel bottom. The overpressure 15.2.6.1.1.1 Loss of Unit AuxillaryTransformer transient is below the reactor coolant pressure bounfary transient pressure limit of 96.7 Causes for interruption or loss of power from kg/cm g. Vyssel dome pressure does not exceed the unit auxiliary transformer can arise from L 83.0 kg/cm g. A comparison of these values to transformer (main and unit auxiliary) malfunction L those for turbine trip at high power shows the and isolated phase bus failures. l similarities between these two transients. The l- l prime difference 3 the subsequent main steamline isolation. 15.2.5.5 Radiological Consequences l While the consequences of the events 1. l Amendment 15 15.2 12 l l l; L

1 ABM 234aoorn menadard Plant av c , ' g M/G sets are capable of holding the RIPS at their original speeds for one second. Then the RIPS (v) will coast down at a speed of 10%/sec for two seconds, and trip at three seconds after the opening of the M/G set supply breakers. Therefore, this event is less severe than the transient described in Subsection 15.2.6.1.1.1. . 15.2.6.1 F equency Clasalfication 15.2.6.1 osa of Unit Auxiliary Transformer l j A loss of one unit auxiliary transformer is assumed to result in a generator trip and the Although the frequency of this event is low , opening of the generator and high volatge enough to be an infrequent event, this transient breakers. The generator trip will cause a disturbance is analyzed as an incident of moderate reactor scram and trip of four RIPS not connected frequency. ; to M/G sets immediately. The opening of ' generator and high voltage breakers will result 15.2.6.1.2.2 tms of Grid Connections i t' s loss of power to both unit auxiliary trar.sformers. However, the remaining six RIPS Although the frequency of this event is low , will be powered by M/G sets. The M/G sets are enough to be an infrequent event, this transient capable of holding the RIPS at their original disturbance is analyzed as an incident of moderate speeds for one second then the RIPS will frequency. l coastdown at a speed of 10%/sec for two seconds, ' and trip at three seconds after the start of the ! event. t l 15.2.6.2 Sequence of Events and Systems Operation 15.2.6.1.1.2 less of Grid Connections 15J.6.2.1 Sequence of Events , Loss of grid connection can result from major 15.2.6.2.1.1 leen of Unit Auxiliary Power shif ts in electrical loads, loss of loads, Transformer lightning, storms, wind, etc., which contribute to electrical grid instabilities. These Table 15.216 lists the sequence of events for instabilities will cause equipment damage if Figure 15.211. unchecked. Protective relay schemes

automatically disconnect electrical sources and 15.2.6.2.1.2 Loss of Grid Connections
loads to mitigate damage and regain electrical L grid stability. This event is similar to a loss of unit at.xiliary transformer as discussed in Subsection Should this occur, it would result in a 15.2.6.2.1.1.

generator load rejection, which subsequently initiates a reactor scram and trip of 4 RIPS. The remaining six RIPS are powered by M/G sets. l' Initially, the M/G sets continue to receive power l from the main generator (as long as 12 seconds). l When the main generator voltage decays sufficiently to cause a reversal and outflow of O V electrical power from the M/G set motor, the M/G set supply breakers trip open. At this time the Amenoment 15 15.2 13

ABWR m-n Standard Plant nry c 15.2.6.2.1.3 Identification of Operator Actions power. Estimates of the responses of the various reactor systems (assuming loss of the unit / The Operator should maintain the reactor auxiliary transformer) provide the following ff" water level by use of the RCIC system and control simulation sequence: reactor pressure by use of the safety / relief 9 valves. Verify that the turbine d c oil pump is * (1) A generator trip occurs at time = 0, which operating satisf actorily to prevent turbine initiates a scram and trip of 4 Rips (already bearing damage. Also verify proper switching and tripped at time = 0, see (2) below); loading of the emergency diesel generators. (2) All electrical pumps, including feedwater The following is the sequence of operator pumps, including 4 RIPS not connected to the actions expected during the course of the events M/G sets are tripped at a reference time, t = when no imbedi' 6e restart is assumed. The 0, with normal coastdown times for the reactor operator should internal pumps; (1) verify all rods in; (3) The remaining 6 RIPS powered by M/G sets are capable of maintaining their original speeds (2) check that diesel generators start and carry for one second, then coast down at a speed of the vital loads; 10%/sec for two secor.ds, and trip at three seconds after the start of the event. (3) check that the RCIC starts when reactor vessel level drops to the initiation point (4) Within f seconds, there is a loss of the main after the relief valves open; condenser circulating water pumps which is conservatively assumed to cause the condenser (4) break vacuum before, the loss of scaling vacuum to drop to the main turbine trip steam acev'st setting, causing stop valve closure, assuming 1.27 centimeters lig/sec vacuum decay rate. (5) check T G A.Sirries during coastdowr.; ' (6) secure the RCIC when both reactor pressure condenser vacuum is expected to reach the MSIV and level are under control; and bypass valves closure setpoint and initiate steamline isolation. (7) continue cooldown per the norrnal procedure; and Operation of the RCIC system function is not simulated in this analysis as its operation occurs (8) complete the scram report and survey the at some time beyond the primary concerns of fuel maintenance requirements. thermal margin and overpressure effects of this analysis. 15.2.6.2.2 Systems Operation 15.2.6.2.2.2 Loss of Grid Connections 15.2.6.2.2.1 Loss of Unit Auxillary Transformer Same as Subsection 15.2.6.2.2.1. This event, unless otherwise stated, assumes and takes credit for normal functioning of plant instrumentation and controls, plant protection and reactor protection systems. 15.2.63 Core and System Performance The reactor is subjected to a complex sequence of events when the plant loses all auxiliary 9 Amendment 15 15.2 14

M 21A6100AD me..a..a pi... m ) Safety / relief valves open in the pressure relief ! [_} v 15.2.63.1 Lens of Unit Auxiliary Power Transformer mode of operation as the pressure increases beyond their setpoints. The pressure at the bottom of the ve sel is limited to a maximum value of 84.4 Figure 15.211 shows graphically the simulated kg/cm g, below the vessel pressure limit of 96.7 transient. The initial portion of the transient kg/cm g. 1 is similar to the load rejection transient. At 8 ! seconds the turbine trips, on low condenser 15.2.6.4.2 Less of Grid Connections vacuum. Main steamline isolation valves and turbine bypass valves close at 28 seconds on Same as Subsection 15.2.6.4.1. their condenser vacuum setpoint. Sensed level drops to the RCIC initiation setpoint at 4p timately 21 seconds after loss of auxiliary pot. . l There is no significant increase in fuel temperature or decrease in the MCPR value, fuel thermal margins are not threatened and the design basis is satisfied. Therefore, this event does 15.2.6.5 Radiological Consequences not have to be remnalyzed for specific core configurations. While the consequences of the events identified previously do not result in any fuel 15.2.63.2 Loss of Grid Connections failures, radioactivity is nevertheless discharged to the suppression pool as a result of SRV l Same as Subsection 15.2.6.3.1. This event actuation. However, the mass input, and hence c does not have to be reanalyzed for specific core activity input, for this event is much less than I configurations, those consequences identified in Subsection ' 15.2.4.5; therefore, the radiological exposures noted in Subsection 15.2.4.5 for Type 2 events cover these consequences of this event. l 15.2.7 Less of Feedwater Flow 15J.7.1 Identincation of Causes and Frequency Classincation 15.2.7.1.1 Identincation of Causes N 15.2.6.4 Barrier Performance A loss of feedwater flow could occur from pump l failures, operator errors, or reactor system j 15.2.6.4.1 Less of Unit Auxillary variables such as high vessel water level (L8) l Transformer trip signal. I . The consequences of this event do not result 15.2.7.1.2 Frequency Classincation in any significant temperature or pressure I, transient in execss of the criteria for which the This transient disturbance is categorized as L fuel, pressure vessel or containment are an incident of moderate frequency. I designed; therefore, these barriers maintain . l their integrity and function as designed. , l 1 (D 1+ Q Amendment t$ 13.21$

ABWR = =^n

$13ndard Plant av c 15.2.7.2 Sequence of Events and Systems 15.2.7.3 Core and System Perforinance Operation 15.2.7.2.1 Sequence of Eventr, The results presented in Figure cf thisFeedwater 15.212. transient flow simulatio terminates at approximately 5 seconds. Subconting Table 15.20.7 lists the sequence of events decreases causing a reduction in core power level for Figure 15.212. and pressure. As power level is lowered, the turbine steam flow starts to drop off because the 15.2.7.2.1.1 Identification of 0perator Actions pressure regulator is attempting to maintain pressure for the first 10 seconds. Water level The operator should ensure RCIC actuation so continues to drop until, first, the recirculation that water inventory is maintained in the reactor flow is rut back at level 4 (L4) and then the vessel. Monitor reactor water level and pressure vessellevel(L3) scram trip setpoint is reached, control and T G auxiliaries during shutdown. whereupon the reactor is shut down and the four l RIPS are tripped. Vessel water level continues to The following is the sequence of operator drop to the L2 trip. At this time, the remaining actions expected during the course of the event six RIPS are tripped and the RCIC operation is when no immediate restart is assumed. The initiated. MCPR remains considerably above the operator should: safety limit, because increases in heat flux are not experienced. Therefore, this event does not (1) verify all rods in, following the scram; have to be reanalyzed for spccific core configurations.

l (la) verify trip of four RIPS; 15.2.7.4 Itarrier Performance (2) verify RClC initiation; The consequences of this event do not result (3) verify that the remaining recirculation in any temperature or pressure transient in excess pumps trip on reactor low level (L2); of the criteria for which the fuel, pressure vessel or containment are designed; therefore, (4) continue operation of RCIC until decay heat these barriets maintain their integrity and diminishes to a point where the RFIR system function as designed, can be put into service; 15.2.7.5 Radiological Consequences (5) monitor turbine coastdown, break vacuum as necessary; and The consequences of this event do not result in any fuel failure. Therefore, no analysis of (6) complete scram report and survey maintenance the radiological consequences in required, requirements. 15.2.8 Feedwater Line Break 15.2.7.2.2 Systems Operation Refer to Subsection 15.6.6. Loss of feedwater flow results in a reduction of vessel inventory, causing the vessel water 15.2.9 Failure of RHR Shuidown Cooling level to drop. The first corrective action is the low level (L3) scram trip actuation. Reactor The RHR system performs low pressure core protection system responds within one second cooling, containment heat removal, containment after this trip to scram the reactor. The low spray and shutdown cooling funct.ons. The RHR level (L3) scram trip function meets the system has three independent divisions each of l single failure criterion. Four of the RIPS are which contains the necessary piping, pumps, tripped at level 3. valves, heat exchangers, instrumentation and O' Amendment t$ 15.2 16 l l

MMpi... me. a .. nA610arn myc electrical power for operation. Each division also has its own cooling water supply, diesel

   .              generator and room cooling system. For the shutdown cooling function each division has its own suction line from and return line to the RPV. Thus each of the three RHR divisions is completely independent of the other divisions in its shutdown cooling function. The RHR system reduces the primary system temperature to 51.70C within 24 hours of plant shutdown.

Normally, in evaluating component failure considerations associated with RHR system shutdou cooling mode operation, active pumps, valves or instrumentation would be assumed to fall. If the single active failure criterion is applied to the RHR system, one of the three RHR divisions would be inoperable. However, the two operable RHR divisions could achieve cold l shutdown shutdown. to 1000C within 36 hours after reactor Failure of offsite power is another case which could affect the shutdown cooling function. The plant will have two independent offsite power supplies, if both offsite power supplies are lost, each RHR division has its own diesel generator which will permit operating that division at its rated capacity. Application of

                  . the single active failure criterion would still leave two RHR divisions operational.

The RHR system description and performance evaluation in Subsection 5.4.7 describes the models, assumptions and results for shutdown cooling with two RHR divisions operational. 15.2.10 References

1. ~ F. G. Brutchscy, et al., Beharlor oflodine in Reactor Water During Plant Shutdown and Startup, August 1972 (NEDO 10585).

2. H. Careway, V Nguyen, and P. Stancausge, Radiological Accident The CONACO3 CODE, December 1981 (NEDO 211431).

O Amendment 15 15.2 17

L ABWR mamu Standard Plant myc ' Table 15.21 I (.)) SEQUENCE OF EVENTS FOR FIGURE 15.21 IIMEfawi EVENT 0 Simulate one main turbine control vaht to close 0- Failed turbine control vahr starts to close 13 Neutron flux reaches high Dux scram setpoint and initiates a reactor scram 1.5 Turbine bypass valves start to open. 63 Water level reaches level 3 setpoint. Four RIPS are tripped. Table 15.2 2 SEQUENCE OF EVENTS FOR FIGURE 15.2 2 TIME fsui EVENT 0 Simulate zero steam now demand to main turbine and bypass valves. O Turbine control valves start to close. 1.0 Neutron Aux reaches high Dux scram setpoint and initiates a reacter scram. 2.4 Four RIPS are tripped due to high dome pressure. 2.6 - Safety / relief valves open due to high pressure. 8.9 Safety / relief valves close. 9.4 Group 1 safety / relief valves open again to relieve decay heat 9.8 Group 2 safety / relief vahrs open again to relieve decay heat. 15 (est.) Safety / relief valves close. l v Amendment 15 15.2 16 i

ABM 2mimin ! me .d ed Plant myc l 1 Table 15.2 3 l SEQUENCE OF EVENTS FOR FIGURE 15.2 3 TI M E (nec) D'ENT (-)0.015 Turbine generator detection of loss of electrical load. 0.0 Turbine generator load rejection sensing devices trip to initiate turbine control valves fast closure and main turbine bypass system operation. l 0.0 Fast control valve closure (FCV) initiates reactor scram and a trip of 4 RIPS. 0.07 Turbine control valves closed. 0.1 Turbine bypass valves start to open. 1.9 Safety / relief valves open due to high pressure. 7.0 Safety / relief valves close. Table 15.2 4 SEQUENCE OF EVENTS FOR FIGURE 15.2 4 l Tl M E (sec) EVENT (-)0.015 Turbine generator detection ofloss of electricalload. O 0.0 Turbine generator load rejection sensing devices trip to initiate turbine control valves fast closure and main turbine bypass system operation. 0.0 On ti bine bypass valve fails to operate on demand. l0.0 Fast control valve closure (FCV) initiates reactor scram and a trip of 4 RIPS. 1 1 0.07 Turbine control vahts closed. 0.1 Remaining bypass vahrs start to open. 1.6 Safety / relief valves open due to high pressure. l6.9 Safety / relief valves close. O Amendment 15 15.2 19

MN 2WlooAn Standard Plant _ uv c Table 15.2 5 SEQUENCE OF EVENTS FOR FIGJJRE 15.2 5 TIME (anc) EVENT (-)0.015 Turbine generator detection ofloss of ele ricalload. 0.0 Turbine generstor load re}cetion sensing devices trip to initiate turbine control vahts fast closure. 0.0 Turbine bypass valves fail to operate. l0.0 Fast control valve closure (FCV) initiates reactor scram and trip of 4 RIPS. 0.07 Turbine control valves closed. 13 Safety / relief valves open due to high pressure. 7.6 Safety / relief valves close. 83 Safety / relief vahts open again to relieve decay heat.

                             > 15.0 (est.) Safety / relief valves close again.

Table 15.2 6 SEQUENCE OF EVENTS FOR FIGURE 15.2 6 TI M E (sec) EVENT 0.0 Turbine trip initiates closure of main stop valves. 0.0 Turbine trip initiates bypass operation. 0.01 Main turbine stop valves reach 85% open position and initiate reactor scram and trip of 4 RIPS. 0.1 Turbine stop valves close. 0.1 Turbine bypass valves start to open to regulate pressure. 2.0 Safety / relief valves open due to high pressure. 6.9 Safety / relief vahts close. O Amendment 15 13.220 '

MM 21A6ttnAn Stannlard Plant RtY C Table 15.2 7 g SEQUENCE OF EVEN"I3 FOR FIGURE l$.2 7 TIME faec) E\TST 0.0 Turbine trip initiates closure of main stop valves. 0.0 One turbine bypass valve fails to operate. 0.01 Main turbine stop valves reach 85% open position and initiate reactor scram and trip of 4 RIPS. 0.1 Turbine stop valves close. 0.1 Remaining bypass valves start to open. 1.7 Safety / relief valves open due to high pressure. 6.8 Safety / relief valves close, Table 15.2 8 SEQUENCE OF EVENTS FOR FIGURE 18.2 8 IIMJi.(ats) EVENT 0.0 Turbine trip initiates closure of main stop valves. 0.0 Turbine bypass valves fail to operate. 0.01 Main turbine stop vahrs reach 85% open position and initiate reactor scram and a trip of 4 RIPS.

  ' O.1            Turbine stop valves close.

1.4 Safety / relief valves open due to high pressure. t 7.4 Safety /relieIvalves close. 8.2- Safety / relief valves open again to relieve decay heat, t

   > 14.0 (est.)   Safety / relief valves close again.

l h Amendment 15 15.2-21 l

23A6100All RimmAard Plant mye Table 15.2 9 SEQUENCE OF EVENTS FOR FIGURE 15.2 9 TI M E (sec) EVENT 0.0 Clos, 'e of all main steamline isolation valves (MSIV). 0.45 MSIVs reach 85% open. 0.45 MhlVs position trip seram initiate <l. 2.6 Four RIPS are tripped due to high reactor pressure. 2.9 Safety / relief valves open due to high pressure. 7.8 Safety / relief valves close. 9.0 Safety / relief valves open again to relieve decay heat.

                   > 10.0 (est.)                                                                  Vessel water level reaches L2 setpoint. RCIC is initiated.

The remaining six RIPS are tripped. l

                   > 13.0 (est.)                                                                  Group 1 safety /tclief valves close again.

40.0 (est.) RCIC flow into vessel (not included in simulation).

O : i 3 i O Amendment 15 15.2 22

ABM 23462 man Standard Plant REY B Table 15.210 POST. TRANSIENT PRIhiARY COSTAINhfEST INVESTORY (AIR PLUS WATER) (CURIES) g ISOTOPE Lh1L' 10SIIN 1 HOUR 2 HOUR fllQl@ B HOUR

12 HOUR 1 D AY 4 DAYS 30 D AYS 3 131 2.1E + O2 3.0E + 02 19E + O2 2.7E + O2 2.3E + O2 1.7E + 02 IJE + 02 S.2E + 01 23D01 1132 4JE + 02 4.2E + 02 2.1E + 02 2.2E + 02 1.0E + 02 2JE+01 52E + 00 5.9E42 IJE 13 1 133 6.4E + 02 6.3E + 02 5.8E + 02 5.2E + 02 4.3E + 02 1.8E + 02 1.9E + 02 $4E+01 2.9D02 1134 7.0E + 02 6.2E + 02 3 0E
02 1.3E + 02 1.3E + 01 73E41 2.3E42 7.4E47 l 115 6.1E + 02 $.9E + 02 3.1E + 02 4.3E + 02 3.0E + 02 IJE + 02 7.$E + 01 9.0E + 00 2.8E45 TOTALS 2.7E + 03 2.6E + 03 2.0E + 03 1.6E + 03 1.1E + 03 63E+ 02 4.0E + 02 1.1E + 02 2.7E41 MR43M 2.0E + 02 1.9E + 02 1.4 E + 02 9.6E
01 43E+01 9.9E + 00 1.2E + 00 2.1E 03 KR43M 4 4E+ 02 4.3E + 02 18E + 02 3.2E + 02 14E+02 1.3E + 02 3.8E + 01 9.7D01 2.8510 KR43 2.0E + 01 2.0E + 01 2.0E+01 2.0E+01 2.0E+01 2.0E + 01 1.1E + 01 18E40 33E45 KR47 8 4E + 02 7.7E 4 02 4 9E + 02 2.8E + 02 9JE+01 1.1E + 01 6.7E41 1.6E N KR48 1.2E + 03 1.2E + 03 94E 02 7.3c + 02 43E + 02 1.7E + 02 3JE + 01 3.1E41 1.4 D13 KR49 1.2E+03 1.7E+02 3.0E 03 6.0E M XE131M 1.0E + 01 1.0E + 01 1.0E + 01 1.0E + 01 1.0E + 01 1.0E+01 5JE 00 8.8E 01 1.$E45 XE133M 13E + 02 13E + 02 15E + 02 13E + 02 1.4 E + 02 1.4E+02 7.0E + 01 9.90 M 7.6D05 XD133 3.6E + 03 3 6E + 03 3 6E + 03 16E + 03 33E+03 33E+ 03 1.8E + 03 2.9E + 02 38E-03 XE135M 63E+02 4 4E + 02 4.8E + 0! 3.3E + 00 1.6E42 4.0E47 5.3D12 XD135 4.7E + 02 4.6E + 02 4.3E + 02 4.0E + 02 3 4E + 02 23E+02 1.0E + 02 6.7E 00 $3E47 XD137 2.7E + 03 5.2E + 02 61DC: 1.2E46 4 4D16 XD138 2.9E + 03 1.8E
03 1.6E + O2 8JE + 00 2.4 E 02 1.9E47 8.3D13 TOTALS 14E + N 9.8E + 03 6 4 E + 03 5 6E + 03 4.9E + 03 4.2E + 03 2.1E + 03 3.1E + 02 39E43 O

l

Beginning of Containment Purge

(. l l Amendment 2 13 S'23 O

ABMR ummin Standard Plant RD' B

    /~l LJ Table 15.211 ACTWITY RELEASED TO THE ENVIRONMENT (CURIE)

Isotope 12 HOURS 1lM 4 DAYS 30 DAYS 1131 2.8E 03 6.6E 03 8.6E 03 8.6E 03 1132 23E 04 2.9E 04 2.9E 04 2.9E44 1 133 4.4E 03 9.2E 03 1.1E 02 1.1E 02 1 134 4.0E 06 4.1E 06 4.1E 06 4.1E 06 1135 2.1E 03 35E 03 3.7E 03 3.7E 03 i TOTALS 9.5E 03 2.0E 02 23E 02 23E 02 i KR 83hi 2SE+ 00 2.8E + 00 2.8E + 00 2.8E + 00 i KR 85hi 4.4E + 01 6.2E + 01 63E + 01 63E+01 KR 85 8.9E + 00 1.8E + 01 2.0E + 01 2.0E + 01 KR 87 2.2E + 00 23E + 00 23E+ 00 23E + 00 KR4B 5.1E + 01 65E + 01 63E+01 6.5E + 01 l KR 89 INSIGNIFICANT XE131h1 4.6E + 00 9.1E + 00 1.0E + 01 1.0E + 01 , XE133h1 6.0E + 01 1.2E + 02 1.2E + 02 1.2E + 02 j XE 133 1.6E + 03 3.1E + 03 33E + 03 33E + 03 XE135hi 2.1E 0S 2.1E 08 2.1E OS 2.1E OS XE 135 1.0E + 02 1.6E + 02 17E + 02 1.7E + 02 XE 137 INSIGNIFICANT ( XE 13S 93ti M 93E 09 93E 09 93E 09 k TOTALS 1.8E + 03 3.5E + 03 3.8E + 03 3.8E + 03 Table 15 212 DOSE EVALUATION AND METEOROLOGY ~ DISPERSION THYROID W BODY BETA SKIN SEC/m* mrem mrem mrem mrem 1.0E 5 3.0E 2 8.5E 1 13E+0 2.2E + 0 5.0E 6 ISE 2 43E1 6.6E1 1.1E + 0 1.0E 6 3.0E 3 85E 2 13E1 2.2E1 5.0E 7 1.5E 3 43E 2 6.6E 2 1.1E 1 1.0E 7 3.0E 4 8.5E 3 13E 2 - 2.2E 2 L Amendment 2 152 24

i MM 23A61ooAli Standard Plant nry c Table 15.213 - TYPICAL RATES OF DECAY FOR CONDENSER VACUUM ! CAUSE ESTIM4TED VACUUM DECAV RATE (1) Failure ofIsolation of <2.5 centimeters Hg/ min Steam Jet Air Ejectors (2) Loss of Scaling Steam Approximately 2.5 to 5.1 centimeters Hg/ min to Shaft Gland Seals (3) Opening of Vacuum Breaker Valves Approximately 5.1 to 30.5 centimeters Hg/ min , (4) Loss of One or More Circulating Approximately 10.2 to 61 centimeters Hg/ min Water Pumps Table 15.214 SEQUENCE OF EVENTS FOR FIGURE 15.210 TIME (sec) EVENT 3.0 Initiate simulated loss of condenser vacuum at 5.1 centimeters Hg/sec 0.0 Low condenser vacuum main turbine trip actuated. s 0.01 Main turbine trip initiates scram and a trip of 4 RIPS. 2.0 Safety / relief valves open due to high pressure. 5.0 Low condenser vacuum initiates main steamline isolation valve closure, , 5.0 Low condenser vacuum initiates main bypass valve closure. 7.0 Safety / relief valves close, 8.9 Safety / relief vahes open again to relieve decay heat, 13.7 Vessel water level reaches L2 setroint and initiates RCIC,

                     > 16.0 (est.)      Safety / relief valves close again.
                     > 40.0 (est.) -    RCIC Dow enters vessel (not included in simulation).

O Amendment 15 15.2 25

ABWR m aman .. madmed plant myc Table 15.215 O raie sicsits issoci<ito wira toss or cosoessen vicoua VACUUM (cm. af Hz) ERQ1I M 69 to 71 Normal Vacuum Range Si to $8 Main Turbine Trip (Stop Valve Closures) 18 to 25 Main Steam Line Isolation Vaht (MSLIV) Closure and Bypass Vahe Closure u 7 i i I l O Amendment 15 15,2-26 i

ABM 2miooxii Standard Plant RIV C Table 15.216 SEQUENCE OF EVENTS FOR FIGURE 15.211 TIME (sed EVENT 0.0 loss of unit auxiliary power transformer, which initiates a generator trip, t 0.0 Turbine control valve fast closure is initiated. 0.0 Turbine control valve fast closure laitiates main turbine bypass system operation. l 0.0 Four RIPS which are not connected to the M/O sets are tripped. 0.0 Fast control valve closure (FU/) initiates a reactor scram. 0.0 Feedwater pumps are tripped 0.07 Turbine control valves closed.

               -0.1             Turbine bypass valves start to open 1.9            Safety / relief valves open due to high pressure.

3.0 Remaining six RIPS powered by M/G sets are tripped. 63 Turbine bypass valves close, after operating for 6 seconds. 7.1 Safety / relief valves close. 7.8 Safety / relief valves open again to relieve decay heat. 8.0 Turbine trips due to low condenser vacuum.

15.6 Safety / relief valves close again.

183 Sa.*ety/ relief vahrs open again to relieve decay heat. 21.0 (est.) Vessel water lesel reaches L2 setpoint. 24.0 (est.) Safety / relief vahrs close again. . 28.0 (est.) Closure of MSIV and turbice bypass valves is initiated via low condenser vacuum (not simulated). 51.0 (est.) RCIC flow enters vessel (not simulated). , 1 O Amendment t5 15.2 27 l

  - - ~ .           _ . - . . . _. _ _ . _ _ . _ . . _ _ _ _ _ _ _ _ _ _ , , _ _ _ _ _ _                     ,

1 l handard Plant uA61%All )' REV c 1 O l l l i 1 1 Deleted O O Amendment 15 15.2 28

ABM 2mi=4n Standard Plant my c Table 15.217 SEQUENCE OF EVENTS FOR FIGURE 15.212 - TIME (sec) EVENT 0 Trip of all feedwater pumps initiated. 2.8 Vessel water level reaches Level 4 and initiates recirculation flow runback, 5 Feedwater Dow decays to rero. 7.5 Vessel water level (L3) trip initiates reactor scram and trip of four RIPS. 19.5 Vessel water level reaches level 2. The remaining six RIPS are tripped. RCIC is initiated. 49.5 RCIC flow enters vessel. O , b E O-Amendment 15 15.2 29

j. ,

O O O 1 NEUTRON FLUX 1 VESSELPRESRISE(PSI) 'S $ 2 STM LINE PRES RISE (PSI) $ 2 PEAK FUEL CENTER TEMP [ 3 150 3 AVE SURFACE HEAT FLUX 125 3 TURBINE PRES RISE (PSI) y 4 FEEDWATER FLOW 4 HElfEF VALVE FLOW (PCT) -v E C 5 BYPASS VALVE FLOW (PCT) $ o 5 VESSEL STEAM FLOW

                                                                             \                             6 TURB STEAM FLOW (PCT)                g
                                                                                          "                                                       3 4
              -                                                         75                               -

Q 100

       !50                   .     %                                    ,,        44%                %                                -
                                                                                     /     2                           ~ .-       .

L- ' '= -25

O*-

0 2 4 6 8 0 2 4 6 8 TIME (sec) TIME (sec) 1 LEVEL (INCHREF-SEP-SKIRT) 2 W R SENSED LEVEL (INCHES) , # 150 3 N R SENSED LEVEL (INCHES) 4 CORE INLET FLOW (PCI)

}.

5 PUMP FLOW 3 (PCT) k

                           =                         't             $    0      1         V T

100

                                                        \,          t b

0 1 50 '" 5

                                                                       -1                                     1 VODREACTIVITY 2 DOPPLER REACTIVITY k.

f 3 GCRAM REACTIVITY 4 TOTALREACTIVITY "F '

r..... l l .

0- 8- -2 .. 8 0 2 4 6 8 0 2 4 6 TIME (sec) TIME (sec) w215-11 ,

                                                                                                                                                   ?9
                                                                                                                                                   ,g 5                                    Figure 15.2-1 CLOSURE OF ONE TURBINE CONTROL VALVE 8
 . _ _                                                                                                                      .i.

g 1 VESSEL PRES RtSE (PSI) 4 RELIEFVALVEFLOW(PCT) r

                                                 ;;                                                                                 2 STM LINE PRES HISE (PSI) 5 BYPASS VALVE FLOW (PCT)           E y                                                                                  3 TURBINE PRES RISE (PSI) 6 TUR8 STEAM FLOW (PCT)              Qj 0                                                                                               3                                                 m#

150 1 NEUTRON FLUX 2 PEAK FUEL CENTER TEMP 125 m [ p 3 AVE SURFACE IIEAT FI tIX 4 FEEDWATER FLOW '

                                                                                                                                                                                            /      3 o                                    5 VESSEL STEAM FLOW                                           %            .k             .'
                                                                                                                                                                                          /
                                                                                                                                    \

100 M i 75 ' - e s NN , N 's i

o 50 x ,, '

                                                                                                                                           \   ~

25

tN ' '

_ l is s s , G \s ou o - i %. ./ - ' -- - 0 2 4 6 8 0 2 4 6 8 TIME (sec) TIME (sec) 1 LEVEL (INCH-REF-SEP-SKIRT) 2 W R SENSED LEVEL (It3 cites) 1 <

3 N R SENSED LEVEL (INCliES) 4 CORE fNLET}}

ML20065L313

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