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{{#Wiki_filter:_ - ___- - --- - ------- - - - - - - - - - - - - - - - - - - - bh eMT--JB3rf- M ZFS AMENDMENT 12 WN. H. EIEMER NUCLEAR POWER STATION CONTENT AND
SUMMARY
OF INSTRUCTION SHEETS FOR AMENDMENT 12 VOLUME 1 SECTION - TABLE OF CCEFENTS Fases 11,15,17,18,19 through 22 Pages 2.0-1x, 2.0-x111, 2.0-xv, 2.0-mix, 4,0-vit, 5.0-1, 5.0-11, 5.0-iv, 5.0-v, 7.0-1x, 7 0-1x.1, 7.0-ziv, 7.0-xviii, 9.0-11, 9.0-111, 12.0-1, ' 12.0-111, 12.0-v 13.0-iv, 13,.0-vi, 14.0-viii, 14.0-ix, D,0-miv. F.0-1, and F.0-11. SECTION 1.0 - INTRODetlrION AND
SUMMARY
Pages 1.0 1 and 1.10-1 SBCTION 2.0 - SITE l l Pages 2.0-iz, 2.0-xiii. 2.0-xv, and 2.0-xix Fages 2.5-86, 2.5-91, 2.5-91.1 through 2.5-91.4, 2.5-98 through 2.5-98.2 Figures 2.4-7, 2,5-52, and 2.5-53 Pages 2.3-1 through 23-7,2.4.5-7 through 2.4.5-10, 2.4.5.1-1, 2.4.5.3-1, 2.4.5.3-2, 2.6-1 through 2.6-18 . SECTION 3.0 - REACTOR f , Pages 3.3-1 through 3.3-5, 3. 7.4. 2. 3-1 through 3. 7. 4. 2. 3-3 POLIDIE 2 l Fages 11,15,17,18,19 through 22 1 e t- os f SECTION 4.0 - REACT 0t GIOLANT SYSTEM ' s 7 e Page 4.0-vii Page 4.10-1 through 4.19-6
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Figures 4.10-1 through 4.10-4 {~ JUN I 51971* -y Pages 4.3-1, 4.3-2 and 4.4-2
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2PS AMENDMENT 12 SECTION 5.0 - CONTAINMENT
- Pages 5.0-1, 5.0-11, 5.0-tv, and 5,0-v Pages 5.2-6 through 5.2-8, 5.2-13, 5.2-14, 5.2-16 through 5.2-16.2, 5.3-6, and 5.3-6.1 Pisures 5.2-1, 5.2-3,- and 5.3-1 Pages 5,2.3.1-2 through 5.2.3.1-5, 5.2.3.8-1 through 5. 2.3. 8-4, 5.3.3.3.3-3 through 5.3.3.3.3-7, 5.3.4.3-1, 5.3.4.4-1 SECTION 6.0 - 00RE STANDBY COOLING SYSTEM (CSCS)
Pages 6.4-1, 6.4-2, 6.5-1 through 6.5-45 TOLUMg 3 Pages 11, 15, 17, 18, 19 through 22 8ECTION 7.0 - CONTROL AND INSTRUMENTATION f Pages 7.0-iz, 7.0-iz.1, 7.0-ziv and sviii Pages 7.2-2, 7.2-6, 7.2-7, 7.2-14, 7.2-23, 7.2-24, 7.2-25, 7.3-8, 7,3-10, 7.3-12, 7.3-15, 7.3-17, 7.3-19 3 7.3-20, 7.3-22, 7.3-23, 7.3-24, 7.3-26, 7.3-28, 7.3-53, 7.4-3, 7.4-6, 7.4-9, 7.4-11, 7.4-15, 7.4-19, 7.4-27, 7.4-28, 7.5-9, 7.5-11, 7.5-19, 7.5-21, 7.5-23, 7.12-1 through l 7.12-14, 7.13-3 through 7.13-6 Pigures 7.12-1.1 and 7.12-1.2 Pages 7.0-1 through 7.0-4, 7.2.3.1-1 through 7.2.3.1-6, 7. 2.3.3-1, 7.2.3.9-2, 7.3.4.8-1, 7. 7-1, 7.9-1 through 7.9-17, 7.10-1, 7.12.5.3-1 VOLtME 4 Pages 11,15,17,18,19 through 22 , SECTION 9.0 - RADI0 ACTIVE (RADWASTE) SYSTEMS Pages 9.0-11 and 9.0-111 . Pages 9.3-1, 9.4-1 through 9.4-9 h s h SECTION 10.0 - AUKILIARY SYSTEMS j! Pages 10.0-11 and 10.0-vii Pages 10.6-1, 10.6-1.1 through 10.6-2, 10.8-1, 10.8-1.1 through 10.8-2, d( 10.10-4, 10.19-1 , 10.19-2, and 10. 20-1 li:! Pigures 10,8-1, 10.8-2, 10. 8-3, and 10.10-2 Pages 10.0-2, 10.19-2, 10.19-3, and 10.20-1 :' l' l b i I
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EPS AMENDMENT 12 SECTION 12.0 - STATION STRUCTURES AND SHIELDING Pages 12.0-1,12.0 111, and 12.0-v Pages 12.2-9, 12.2-13, 12.2-14 through 12.2-16, 12.3-6, 12.4-2, 12.4-4, 12.4-6 through 12.4-10, 12.5-2 and 12.5-3 - Figures 12.2-1, 12.3-3 and 12.3-8 Pages 12. 2.1.1-3, 12. 2.1.1-4 through 12.2.1.1-7, 12.2.2-3, 12.2.2.4-1, 12.2.2.5-1, 12.2.2.5-2, 12.2.2.5-3, 12.3.2.2-1, 12.3.2.2-2, 12.3.2.3-1, 12.3.2.3-2, 12.3.2.5-1, 12.3.4.2-1, 12.3.6-1, 12.3.7-1, 12.3.8-1, 12.5.1-1, 12.5.6-1 SBCTION 13.0 - CONDUCT OF OPERATIONS Pages 13.0-iv and 13.0-31 , 4 Pages 13.6-16 sad 13.6-17 . Figure 13.2-1 gECTION 14.0 - PLANT SAFETY ANALYSIS Pages 14.0-viti and 14.0-iz
; Page 14.6-16 I
Fiat res 14.5-15, 14.5-16, 14.6-10, 14.6-11, and 14.6-12 l' Pages 14.5 -1, 14.5.5-1, 14.5.5-2, 14.5.5-3, 14.5.5-4, 14.5.5-5, 14.5. 6-1, l 14.6.3-4, 14.6.3-5, 14.6.3-6, 14.6.3-10, 14.6.3-11, 14.6.3-12, 14.6.3-13, i 14.6.3-14 through 14.6.3-18, 14.9-1-3 through 14.9.1-5, 14.9.2.3-1
! through 14.9.2.3-3 7
,1 VOLUME 5 l Pages 11,15,17,18,19 through 22 {
l I I APPEFDIX 5.0 - OUTLINE OF PROPOSED TECHNICAL SPECIFICATIONS l i Fages 5.5-16, 5.6-1 through B 6-5 Pages 3.1-1 and B.1-2 ! APPENDIX C.O - EQUIPMENT DESIGN CRITERIA i j Pages C.3.1-4 through C.3.1-7 APPENDIX D.O - 00ALITY CONTROL SYSTEM Page D.0-miv Figure D.2-1 h Page D.6-14 APPENDIX F.O - CONFORMAN,CE TO AEC DESIGN CRITERIA Pages F.0-1 and F.0-11 Pages F.1-1, F.2-1 through F.2-7, F.2-9, F,2-10, F.2-12, F.2-15, F.2-18, ; F.2-25, F.2-27, F.3-1 through F.3-20 , ! 4 APPENDIX H.O - IDENTIFICATION - RESOLUTION OF AEC - ACES AND STAFF CCSNCERNS f Page H.0-1
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g AMENDNENT 12 T
' APPENDIX !.0 - PROCEDURES FOR THE SEISNIC ANALYSIS OF CRITICAL NUCI2AR
$ _F0WER PLANT STRUCTURES. SYSTEMS AND EQUIPMENT l
Page I.0-1 T ' e 4 e e ie b. I l l I - 1 4 s l { l I 4 4
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F i. ZPS I AMENDMENT 12 l l ANSWERS TO AEC QUESTIONS OF APRIL 9,1971 Mi. H. ZIMMER NUCLEAR POWER STATION P
,' The answers to t?.e April 9,1971 AEC Questions have been provided in
'i Amendment 12. Table of Contents Pages 21 and 22 indicates the way in which the AEC questions have been renumbered to reflect their relevance to particular sections of the PSAR. The system for renumbering the . questions is provided below:
i 1 AEC . f QUESTIONS RENUMBERED AS I, NUMBER QUESTION 14.1 14.6.3-1 . 14.8 14.6.3-2 14.5 14.6.3-3 The numbers before "~ f Where more than one AEC
! the dash (-) refer to question pertains to a l a particu.utr section, particular section, sub-subsection or para- section or paragraph, graph in the PSAR these additional ques-which is directly tions are indicated by related to the AEC a number which follows tuation, the dash (-).
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,f AMENDMENT 12 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 1 SECTION - TABLE OF CONTENTS All changes have been indicated by a vertical line and the amendment number (12) in the right margin of the page. i
- 1. In Volume 1, SECTION - TABLE OF CONTENTS, remove and destroy Pages 11,15,17,18,19 and 20 and replace with amended Pages 11, 15, 17 j 18, 19 and 20. Af ter Page 20 insert new Pages 21 and 22. ;
- 2. In Volume 1, SECTION - TABLE OF CONTENTS, remove and destroy the following pages and replace them with the appropriate amended pages listed below:
Remove Pages Replace With Amended Pages 2.0-ix 2.0-1x 2.0-xiii 2.0-x111 2.0-xv 2.0-xv I
) 2.0-xix 2.0-xix l 4.0-vii 4.0-vii.
5.0-1 5.0-1 5.0-11 5.0-11 l 5.0-iv 5.0-1v i 5.0-v 5.0-v 7.0-ix 7.0-xiv 7.0-ix and 7.0-ix.1 i 7.0-xiv 7.0-xviii 7.0-xviii 1 i 9.0-11
- 9.0-11 9.0-111 9.0-111 12.0-1 12.0-1
< 12.0-111 12.0-111
- j. 12.0-v 12.0-v i
13.0-iv 13.0-iv 13.0-vi 13.0-vi 14.0-viii 14.0-viii 14.0-ix 14.0-ix D.0-xiv D.O-xiv ll F.0-1 F.0-1 {- F.0-li F.O-ii 1 ' l
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( ZPS j AMENDMENT 12 VOLUME 5 _ TABLE OF CONTENTS, (Continued) PAGE D.0 APPENDIX D.O - QUALITY CONTROL SYSTEM TABLE OF CONTENTS D.0-1 - a
- D.1 INTRODUCTION '
D 1-1
.I D.2 CINCINNATI CAS & ELECTRIC CO. QUALITY ASSURANCE PROGRAM D.2-1 D.3 SARGENT & LUNDY QUALITY ASSURANCE SYSTEM D.3-1 l
I D4 GENERAL ELECTRIC QUALITY SYSTEM FOR BWR NUCLEAR STEAM
,I SUPPLY PROJECTS D.4-1 D.5 KAISER ENGINEERS INC. (CONSTRUCW)RS) QUALITY ASSURANCE -
l QUALITY CONTROL PROGRAM D.5-1 D.6 APPLICABILITY OF QUALITY ASSURANCE PROGRAM TO COMPONENTS,- 3 l SYSTEMS AND STRUCTURES D.6-1 J E.0 APPENDIX E.O - STATION ATMOSPHERIC RELEASE LIMIT CALCULATIONS ; E.1-1 j F.0 APPENDIX F.O - CONFORMANCE TO AEC DESIGN CRITERIA l TABLE OF CONTENTS F.0-1 F.1
SUMMARY
DESCRIPTION F.1-1 ! F.2 CRITERION CONFORMANCE (THE "70" CRITERIA) F.2-1 12 l F.3 CRITERION CONFORMANCE (THE "64" CRITERIA) F.3-1 ! G.0 APPENDIX G.0 - STATION NUCLEAR SAFETY OPERATIONAL ANALYSIS TABLE OF CONTENTS G.0-1 G.1 ANALYTICAL OBJECTIVE G.1-1 G2 APPROACH TO OPERATIONAL NUCLEAR SAFETY G.2-1 C.3 METHOD OF ANALYSIS G.3-1 G.4 DISPLAY OF OPERATIONAL ANALYSIS RESULTS G.4-1 11
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) AMENDMENT 12 LIST OF ZPS, OCTOBER 13. 1970 '
AEC QUESTIONS, (Continued) ; i
,t AEC QUESTION RENUMBERED VOLUME .
- j. NIMBER AS QUESTION PACE OF PSAR . ji -
-l 5.5 5.2.5-1 5.2. 5-1 2 $ [ 5.6 5.0-1 g 5.0-1
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f 5.7 5.2.5.1-1 5.2.5.1-1 2 5.8 5.2.3.1-1 5. 2. 3.1-1 2 l 5.9 12.3.2.5-1 12.3.2.5-1 4 1 1 5.10 5. 2. 3.8 -1 5.2.3.8-1 2 b2 12.1 12.2.1.1-1 12.2.1.1-1 4 12.2 , 12.3.1-1 12.3.1-1 4 12.3 12.2.1.1-2 12.2.1.1-2 4 12.4 12.2.2-1 12.2.2-1 4 1 12.5 12.2.2.5-1 12.2.2.5-1 4 12.6 12.4.4.1-1 12.4.4.1-1 4 y 7 12.7 12.4.4-1 12.4.4-1 4 i 12.8 12.3.2.3-1 12.3.2.3-1 4 . i 12.9 12.3.2-1 12.3.2-1 ' 4 l
- I 12.10 1 i' 12.3.2.2-1 12.3.2.2-1 4 ! [
12.11' 12.3.2.2-2 12.3.2.2-2 4 12.12 12.4.3.3-1 12.4.3.3-1 s 4 ! 12.13 12.3.3-1 e 12.3.3-1 4 ' i 12.14 12 .2 .2 .4 -1 12.2.2.4-1 4 i
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12.15 12.3.6-1 12.3.6-1 4 ! 12.16 12.2.2-3 12.2.2-3 4 12.17 12.2.2-2 12.2.2-2 4 l 12.18 12.5.1-1 12.5.1-1 4 I 12.19 ii 12.3.4.2-1 12.3.4.2-1 4 ,! ' 12.20 12.3.2.3-2 12.3.2.3-2 4 [ 12.21 12.2.1.1-3 12.2.1.1-3 4 ,
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. _ . _ , - - _ _ _ _ - _ _ _ - _ - - _ = _ - _ - - _ _ - - { l 2PS AMENDMENT 12 LIST OF ZPS. FEBRUARY 23 1971 AEC QUESTIONS AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR 2.12' 2.2.3-2 2.2.3-12 1 2.13 2.3.2.1-2 2.3.2.1 1 2.14 2.3.2.1-3 2.3.2.1-3 1 2.15 2.6-1 2.6-1 2,16 1 l 12 2.3.8-1 2.3.8-1 1 4.9 4.7-2 4.7-2 2 9 4.10 4.7-1 4.7-1 2 4.11 4.9-1 4.9-1 2 9 4.12 4.0-1 4.0-1 2 5.11 5.2.3.7-1 5.2.3.7-1 2 7 5.12 10.19-1 10.19-1 2 - 5.13 5.2.3.8-2 5.2.3.8-2 2 - 3 5.14 5.3.3.3.3-3 5.3.3.3.3-3 2 5.15 5.3.3.3.3-4 5.3.3.3.3-5 2 12 5.16 5.3.3.3.3-5 ' 5.3.3.3.3-7 2 - 5.17 5.2.3-1 5.2.3-1 2 7.1 5.3.3.3.3-1 5.3.3.3.3-1 1 2 7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 ! 7.4 7.1-1 7.1-1 3 11 7.5 7.2.3.1-1 7. 2. 3.1- 1 3 l 12 7.6 4.4-1 4.4-1 2 7.7 7.2.3.3-1 7.2.3.3-1 l 12 3 7.B 7. 2. 3. 6-1 7.2.3.6-1 3 7.9 7.2-1 11 7.2-1 3 7.10 7.2.3.9-1 7.2.3.9-1
. 3 7.11 7.2.3.9-2 7. 2. 3. 9-2 3
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ZPS AMENDMENT 12 LIST OF ZPS, FEBRUARY 23, 1971 I AEC QUESTIONS, (Continued) AEC QUESTION RENUMBERED NUMBER VOLUME AS QUESTION PAGE OF PSAR I 7.12 7.2-2 - i 7.2-2 ^3 l 7.13 1.0-1 1.0-1 f11 1 1 t 7.14 7.12.5.3-1 7.12.5.3-1 3 i I 7.15 7.4.3-1 7.4.3-1 3 7.16 7.5.7.3.3-1 7.5. 7.3. 3 -1 3 7.17 7.8.5-1 7.8.5-1 3 11 7.18 7.5.8-1 7.5.8-1 3 7.19 7.6.3-1 7.6.3-1 3 7.20 7.8.5.2-1 7.8.5.2-1 3 7.21 7.9-1 7.9-1 3 I 7.22 7.3.4.8-1 7 3.4.8-1 3 12 i 7.23 7.10-1 7.10-1 3 ! 7.24 D.6-2 D.6-14 5 7.25 j D.0-1 D.0-1
'7.26 10.10.3-1 5 l11 10.10.3-1 4 7 7.27 7.2-3 7.2-5 7.28 7.0-1 7.0-1 3
l11 3 l 12 s 7.29 10.19-2 10.19-2 4 11 7.30 7.0-2 i 7.0-4 3 > 7,31 7.7-1 7.7-1 3 . 12 8.1 8.3.2.1-1 8.3.2.1-1 4 8.2 8.3.2-1 8. 3. 2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 8.5 8.5.4-1 ; 8.5.4-1 4 8.6 8.4.3-2 8.4.3-2 4 8.7 8.5.3.1-1 ,j 8.5.3.1-1 4 8.8 8.0-1 8.0-1 4 18 9
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ZPS i AMENDMENT 12 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued) l AEC QUESTION RENUMBERED VOLUME NUMBER AS OUESTION PAGE OF PSAR 8.9 8.0-2 8.0-2 4 8.10 8.9-1 8.9-1 4 8.11 8.10-1 8.10-1 4 f 9.1 9.2.4-1 , 9.2.4-1 4 ! 9.2 9.2.4.6-1 9.2.4.6-1 4 ! 9.3 9.4-1 9.4-1 4 9.4 9.4.6-1 9.4.6-1 4 l11 9.5 9.2.4.7-1 9. 2.4. 7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 10.0-2 10.0-2 4 l12 , 10.2 10.5-1 10.5-1 4 ! 10.3 10.0-1 10.0-1 4 l 11 10.4 10.11.2-1 10.11.2-1 4 7 l 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 , , 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6-1 4 13.3 13.2.1.2-1 13.2.1.2-1 4 9 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 11 14.12 14.9.1-1 14.9.1-1 4 14.13 14.9.1-2 14.9.1-3 4 12 i 19 9
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ZPS AMENDMENT 12 LIST OF ZPS. FEBRUARY 23, 1971 AEC QUESTIONS, (Continued) AEC QUESTION RENIMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR 15.16 A.2-1 A.2-1 5 7 15.17 A.2-2 . A.2-5 5 15.18 A.2-3 - A. 2- 6 5 ,! 15.19 B.1-1 B.1-1 5 l12 I 1 i i 2 I t i 20 l9 f
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l ZPS i AMENDMENT 12 l l LIST OF ZPS, APRIL 9.1971 1 AEC QUESTIONS l AEC QUESTION RENUMBERED VOLUME I NUMBER AS QUESTION PAGE OF PSAR 1.1 1.10-1 1.10-1 1 L.
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1.2 H.0-1 , , H.0-1 5 2.17 2.3-1 ' 2.3-1 1 4 1 . I 2.18 2.4.5-2 2.4.5-7 1 ! 2.19 2.4.5.1-1 2.4.5.1-1 1 l l 2.20 2.4.5.3-1 2. 4.5 . 3-1 1
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2.21 2.4.5.3-2 2.4.5.3-2 1 2.22 2.4.5-3 2.4.5-10 1
.{ 3.5 3.3-1 3.3-1 1 i' 3.6 3.3-2 3.3-2 1 f 3.7 3.3-3 3.3-4 1 T '
12 1;
- i. 4.13 4.3-1 4.3-1 2 4.14 4.4-2 4.4-2 2 5.0 10.20-1 10.20 -I 4 6.1 6.5-1 6.5-1 2 6.2 6.5-2 6.5-5 l
2 i 6.3 6.5-3 6.5-7 2 i 1 6.4 6.5-4' 6.5-13 2
]
6.5 6.5-5 6.5-16 2 l 6.6 6.5-6 6.5-21 2 6.7 6.5-7 6.5-22 2 6.8 6.5-8 6.5-32 2 6.9 6.5-9 6.5-38 2 6.30 6.5-10 6.5-39 2 6.11 6.5-11 6.5-41 2 . 6.12 6.5-12 6.5-43 2 , ! 6.13 6.4-1 6.4-1 2 1 s 21 l 1 1
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'ZPS AMENDMENT 12 LIST OF ZPS, APRIL 9.1971 I (
l AEC QUESTIONS, (Con tinued) AEC QUESTION f, 1 RENUMBERED < NUMBER A$ QUESTION ^ ,' VOLUME ~j' PAGE OF PSAR j. 7.32 7.9-2 ~ > 7.9-6 3 i 7.33 7.4-3 . 7.9-6 3 ' ~ 12.24 12.2.2.5-2 12.2.2.5-2 4 12.25 12.2.1.1-4 12.2.1.1-4 4 I 12.26 12.3.8-1 12.3.8-1 4 14.14 4.3-2 4.3-2 2 0 14.15 3.7.4.2.3-1 3.7.4.2.3-1 2 14.16 7.9-4 7.9-9 3 14.17 7.9-5 12 1
+
7.9-11 3 ; 14.18 7.9-6 ; 7.9-13 3 : 14.19 7.9-7 7.9-14 3 14.20 14.5-1 14.5-1 l' 14.21 14.5.5-1 4 - j, 14.5.5-1 4 14.22 14.5.5-2 ! 14.5.5-3 4 14.23 14.5.6-1 14.5.6-1 4 fI 14.24 5.3.4.3-1 i 5.3.4.3-1
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3 r 2 l, 14.25 5.3.4.4-1. 5.3.4.4-1 2 ' 14.26 14.9.2.3-1 , 14.9.2.3-1 4 15.19 I.0-1 I.0-1 jj 5 15.20 c.3.1-4 C.3.1-4 5 . i
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ZPS AME*DMENT 12 TA_BLE OF CONTENTS, (Continued) I PACE 2.5.4.4.3.1 Evaluation of Liquefaction Potential 2.5-88 2.5.4.4.3.2 Soil stability Analysis 2.5-91 2.5.4.4.3.3 Bearing Capacities 2. 5 - 91.'- 2.5.4.4.3.4 Static Settlement , i[ ' '2. 5 -91.4. 2.5.4.4.3.5 Dynamic Settlement ,'- -
, . , 2.5-94 2.5.4.4.3.6 Rock - Soil - Structure Interaction 2.5-94 2.5.4.5 Subsurface Walls 2.5-94 5 2.5.4,6 River Bank Stability 2.5-98 2.5.4.7 Effects of Nearby Quarry Blasting on Plant Construction and Operation 2.5-98 2.5.4.8 Future Units 2.5-99 2.5.4.9 References 2.5-100 2.5.4.10 Soil Liquefaction Analysis Report No. 43 2.5-102 I
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ZPS
-i AMENDMENT 12 LIST OF TABLES, (Continued)
TABLE NUMBER - TITLE PAGE 2.5-1 Bulk Sample Descriptions 2.5-6 2.5-2 Direct Shear Tests on Recompacted Sand Samples 2.5-7 2.5-3 RocP Unconfined Compression Test Results 2.5-8 l 2.5-4 Cyclic Triaxial cosapression Test Data - Moduli and Damping 2,5-10 2.5-5 Cyclic Triaxial Compression Test Data - Liquefaction
- 2.5-12 2.5-6 Resonant Column Test Results 2.5-14 2.5-7 Shockscope Test Results 2.5-16 2.5-8 Relative Density Test Results 2.5-19 2.5-9 Specific Gravity Test Results 2.5-20 2.5-10 Penneability Test Results 2.5-21 2.5-11 Soil Conditions in Plant Construction Area 2.5-34 2.5-12 Earthquake Epicenters - 82.5*-86.5* West '
Longitude - 37 *-41' North Latitude 2.5-56 t
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2.5-13 Earthquake Epicenters Intensity V and Greater 1699-1969 - 84*-90* West Longitude - 35*-39' North Latitude 2.5-60 2.5-14 Moduli and Damping values 2.5-76 5 2.5-15 Modified Mercalli Intensity (Damage) Scale of - ' 1931 2.5-77 2.5-16 } Structural Loading conditions 2.5-82 3 2.5-17 Subsurface Conditions 2.5-85 2.5-18 Generalized Subsurface Properties l 2.5-90 ; 2.5-18.1 Results of Stability Analyses During Liquefaction 2.5-91. 2 l 12 i 2.5-19 j Ultimate Bearing capacities 2.5-92 i 2.5-20 Estimated Total and Differential Settlements 2.5-93 2.5-21 EstimatedStructures Adjacent Differential Settlements Between .' 2.5-95 2.5-22 , Parameters for Analysis of Rock-Soil-Structure Interaction 2.5-96 2.5-23 Estimated Lateral Pressures 2.5-23.1 2.5-97 Results o { Slope Stability Analysis Natural River Ban 2.5-24 2.5-98.2 l12 Properties of Analytical Model f' 2.5-105 l5 . 2.0-xiii '
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2PS AMENDMENT 12 I I: l LIST OF FIGURES , (Continued) ; 2.4-1 Nap of the Ohio River Drainage Basin and the Proposed Site of the Wm. H. Zimmer Nuclear Power Station 2.4-2 Location of Public Ground Water Supplies 2.4-3 Location of Private Wells
- 2.4-4 Ohio River Basin Hydrologic Sub-Area Map 2.4-5 Seven-Day Average Low-Flow Frequencies on the Ohio River
] l 2.4-6 Relative Flood Stage Profiles - Ohio River l l 2.4-7 Ohio River Profiles 12 l 2.4-8 Ohio River Stage Frequencies 2,4-9 l ( Precipitation: Area-Depth Duration i 2.4-10 Standard Project Flood Hydrography 2.4-11 Probable Maximum Flood Hydrography 2.4-12 Locations of Cross Sections in Markland Pool 2.4-13 Typical Cross Sections in Markland Pool 2.4-14 Cross Section at Wm. H. Zimaner Site 2.4-15 Cround Water Contour Map 4 7 l 2.0-xv i
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ZPS AMENDMENT 12 LIST OF FIGURES, (Continued) FIGURE NUMBER TITLE 2.5-29 Thickness of Glacial Drift I 2.5-30 Wisconsin Glacial Moraines 2.5-31 Effects of Glaciation 2.5-32 Variation of Shear Modulus I
. i 2.5-33 Lumped Mass Model 'I 2.5-34 Shear Stress at 10 Feet - Taft 2.5-35 Shear Stress at 30 Feet - Taft 2.5-36 Shear Stress at 45 Feet - Taft !
2.5-37 Shear Stress at 55 Feet - Taft 2.5-38 Shear Stress at 65 Feet - Taft 2.5-39 Shear Stress at 75 Feet - Taft 2.5-40 Shear Stress at 92 Feet - Taft $ 2.5-41 Shear Stress at 110 Feet - Taft 2.5-42 f Results of Cyclic Triaxial Tests ' 2.5-43 Stress vs cycles to Liquefaction - 10 Feet ! 2.5-44 Stress vs Cycles to Liquefaction - 30 Feet ' 2.5-45 Stress vs cycles to Liquefaction - 45 Feet 2.5-46 Stress vs Cycles to Liquefaction - 55 Feet l 2.5-47 Stress vs Cycles to Liquefaction - 65 Feet
'}
2.5-48 Stress vs Cycles to Liquefaction - 75 Feet i 2.5-49 Stress vs Cycles to Liquefaction - 92 Feet 2.5-50 Stress vs Cycles to Liquefaction - 110 Feet { 2.5-51 l Liquefaction Potential - Taft 2.5.52 ! Results of Stability Analyses Radwaste Building ' (Case 12) 2.5-53 Results of Stability Analyses Natural River Bank 12 ; (Case 3) l
}
- 2. 0-xix f
ZPS AMENDMENT 12
i LIST OF FIGURES, (Continued)
FIGURE NUMBER TITLE 4.10-1 Typical Temperature Monitoring Leak Detection System 4.10-2 i Leak Detection Differential Temperature Indication Schematic 4.10-3 Leak Detection Absolute Temperature Indication Schematic 12 4.10-4 ) Leak Detection Reactor Water Cleanup Differential Flow I I i t
'l l l
i o 4.0-vil
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ZPS AMENDMENT 12 SECTION 5.0 - CONTAIle(ENT - TABTE OF CONENTS I PAGE 1 d
~
5.0 ' CONTAIle(ENT 5.1-1 j - 1 5.1 StHMARY DESCRIPHON 5.1-1 5.1.1 General 5.1-1 ;t 5.1.2 Primary Containment 5 1-1 f 5.1.3 secondary containment 5 1-1 ! l 5.2 PRIMARY CONTAIlemer 5.2-1 ! ;
} -i 5 2.1 Safeey objective 5.2-1 ; /
5 2.2 Safety Design Basis 5.2-1 , 5.2.3 Description . 5.2-2 j 5.2.3.1 General 5.2-2 j 5.2.3.2 Drywell - 5.2-3 j q 5.2.3.3 Pressure Suppression Chamber and Vent System 5.2.3.4 Penetrations 5.2-4 fJ 5.2-4 { 5.2.3.4.1 General 5 2-4 l 5.2.3.4.2 Pipe Penetrations 5.2-6 l 5.2.3.4.3 Electrical Penetrations 5.2-6 5.2.3.4.4 TIP Penetrations 5.2-9 l 5.2.3.4.5 Personnel and Equipment Access 5 2-9 I 5.2.3.4.6 Access into the Pressure Suppression Chamber } 5.2-12 , 5.2.3.4.7 Access for Refueling Operations 5.2-12 { 5.2.3.5 Isolation valves 5.2-12 i 5.2.3.5.1 General criteria 5.2-12 ' 5.2.3.5.2 Specific criteria 5.2-14 f 5.2.3.6 Primary Contalmeent Venting and Vacuusi Relief System 5.2-15 l 5.2.3.7 Primary Containment Normal Heating, Ventilation and 5.2-16 , Air Conditioning System
- 5.2.3.8 Provisions for t.dditional Primary Containment :
Equipment 5.2-16 12 5.2.4 Safety Evaluation 5.2-16.2 ; 5.0-1 i
ZPS AMENDMENT 12 h, TABLE OF CONTENTS, (Con tinued) j PAGE . 5.2.4.1 General 5 2-16.2 12
- 5. 2.4. 2 Primary Containment Characteristics During Reactor 5.2-17 Blowdown .
d 5.2.4.3 Primary Containment Characteristics Af ter Reactor 5.2-18 l Blowdovn a 1 5.2.4.4 Primary Co. tain.:ent capability ' 5.2-18 5.2.4.5 Primary Containment Leakage Analysis 5.2-18 . 5.2.4.6 Missile Protection 5.2-20 I 5.2.4.7 Penetrations 5.2-21 5.2.4.8 Isolation valves 5.2-23 5.2.5 Inspection and Testing 5.2-26 5 2.5.1 Primary Containment Integrity and Imak Tightness- 5.2-26 5.2.5.2 Penetrations ' 5.2-26
- 5. 2. 5. 3 Isolation valves 5.2 27 h 5.2.6 Operational Nuclear Saf'ety Requirements 5.2-27 -
5.3 SECO!OARY CONTAINMENT SYSTEM 5.3-1 5.3.1 Safety Objective i 5.3-1 5.3.2 Safety Design Basis 5.3-1 5.3.2.1 General 5.3-1 5.3.2.2 Reactor Building , 5.3-1 5.3.2.3 Reactor Building Heating, Ventilation and Air 5 3-1 Conditioning System 5.3.3 Description 5.3-2 5.3.3.1 Reactor Building 5.3-2 5.3.3.2 Reactor Building Penetration 5.3-2 5 3.3.3 Reactor Building, Heating, Ventilation and Air 5.3-2 Conditioning Systems 5.3 3.3 1 Normal Ventilation Design Features 5.3-2 5.3.3.3.2 Abnormal Ventilation Design Features 5.3-3 5.3.3.3.3 Standby Gas Treatment System 5.3-4
. I 5.0-11
-ZPS !
AMENDMENT 12 SECTION 5.0 - coNIAIleSNT LIST OF TABIES TABIE Niag5R _TJ_1]E, PAGE , 5.2-1 Primary Containment - Drywell and Pressure 5.2-5 Suppression Chamber - Principal Design Parameters and Characteristics 12 5.2-3 Electrical Penetrations - Environmental Design 5.2-10 Conditions 5.2-4 Primary Containment - Pressure suppression 5.2-19 System _- Maximus Blowdown Pressure comparison . 9
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ZPS AMENDMENT 12 SECTION 5.0 - CONTAINMENT. LIST OF FIGURES FIGURE NUMBER TITLE a 5.2-1 Primary and Secondary Concrete Containment l l 7 12 ' Structures 5.2-2 Column and Wall Base Detail at Floor Liner Plate 5.2-2.1 Drywell Floor Joint At Containment Wall
~
l7 5.2-3 Typical Section at Buttress
'12 l 5.2-4 Tendon Access Gallery f 5.2-5 Typical Leak Test Chamber f 5.2-6 Drywell Heat Attachment Detail - Tendon l Anchor at Drywell Head 7 5.2-7 Primary Containment System Hot Process Line Penetration i 5.2-8 Primary Containment System Cold Process Line i Penetration 5.2-9 Typical Electrical Penetration Assembly 5.2-10 Personnel Access Lock 5.2-11 Drywell Cooling and Ventilation System 5.2-12 Emergency Lock and Equipment Hatch 5.2-13 Typical Layout of Hoop Tendons 5.2-14 Typical Layout of Vertical Tendons 5.2-15 Reactor Containment Development Elevation 11 5.3-1 Standby cas Treatment System l5 12 5.3-2 CSCS Equipment Area Cooling System 5.3-3 Reactor Building Ventilation System 7
5.3-4 Standby Gas Treatment System Equipment Train -
'5.3-5 Schematic-Showir.g Mixing Effect of Supply Outlet 5.0-v
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p <<' l ejy } l \; r: i ZPS 1 AMENDMENT 12 1 1 TABLE OF CONTENTS, (Continued) PAGE 7.12.1.4 Safety EYaluntion 7.12-4 l l l 7.12.1.5 Inspection And; Nesting 7.12-4 ; L 7.12.1.6 Operational Nucicar Safet; Requirements 7.12-4 { l 7.12.2 Air Ejector Offgets Rndiation Monitoring and Sampling System 7.12-5 fi 7.12.2.1 Power Generation Objective 7.12-5 l 7.12~2.2
.' Power Generation Design Basis 7.12-5 7.12.2.3 Description -
7.12-5
,.. 7.12.2.4 Safety Evaluation 7.12-6 7.12.2.5 Inspection and Testing 7.12-6 '
- 7.12.2.6 Operationni Nuclear Safety Requirements 7.12-6 7.12.3 Offsis Ven*;, Pipe Radiation Monitoring System 7.12-8 7.12.3.1 Power Generation Objective 7.12-8 7.12.3.2 Power Generation Design Basis 7.12-8 i 7.12.3.3 Description 7.12-8 7,12.3.4 Inspection and Testing. 7.12-8 7'.12.3.5 Operational Nuclear Safety Requirements 7.12-8 ;.
7.12.4) Process Liquid Radiation Monitors 7.12-10 12 f 7.12.4.'id Power Generation Objective 7.12-10 ! 7.12.4.2 Power Generation Design Basia 7.12-10 f 7.12.4.3 Description 7.12-10 l 1 7.12.4.4 Power Generation Evaluation 7.12-11 ' 7.12.4.,5 Inspection and Testing 7.12-11 V7.12 A.6 ' Operational Nuclear safety Requirements 7.12-11 f g : 7.'12.5 Reactor Building Ventilation Exhaust Radiation l Monitoring System 7.12-12 l 7.12.5.1 Power Generation Objective 7.12-12
'I 7.12.5.2 Power Generation Design Basis 7.12-12 ll 7.12.5.3 Description 7.12-12 .
7.12.5.4 Inspection and Testing 7.12-12 , f 3 f 7.0-ix l , f .
ZPS
~
AMENDMENT 12 TABLE OF CONTENTS, (Continued) l PACE 1
, 7.12.5.5 Operational Nuclear Safety Requirements 7.12-12 ,
7.12.6' Fuel Pool Ventilation Exhaust Radiation Monitoring System 7.12-13 7.12.6.1 Safety objective + - 7.12-13 l , 7.12.6.2 Safety Design Basis .
- r ,. 7 , 7.12-13
' 1,4 l 7.12.6.3 Description '
7.12-13 l 7.12.6.4 Safety Evaluation , 7.12-14 7.12.6.5 Inspection and Testing 7.12-14 l 9 L 1 i 7.0-ix.1
r ZPS
-i AMENDNENT 12 LIST OF TABLES, (Continued)
TABLE NUMBER ,MTLE, PAGE 7.6-1 Refueling Interlock Effectiveness 7.6 -6 7.7-1 Reactor Manual Control System Instru- 7.7-11 { ment Specifications - i 7.8-1 Reactor Vessel Instrumentation . ' 7.8-3 Instrument Specifiestions I
. ('
' {
7.12 1 Characteristics of Process Radiation 7.12-3
' Monitoring Systems 7.12-2 Process Radiation Monitoring System 7.12-7 Environmental and Power Supply Design 12 Conditions 7.13-1 Area Radiation Monitoring System Environ- 7.13-2 mental and Power Supply Design Conditions .
7.13-2 Locations for Area Radiation Monitoring 7.13-4 12 l Sensors 4 7.16-1 Instrumentation Input Sununary Neutron 7.16-4 I
. Monitoring System I 7.16-2 Instrumentation Output Summary Signal 7.16-12 .
Output Description 7.17-l' Acceptable Ultimate Performance Limits 7.17-5 I' 7.17-2 Acceptable Operational Design Limits 7.17 8 i
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i i i i e e f 7.0-xiv f
1> l1 l i ZPS AMENDMENT 12 i L. '
)
LIST OF FIGURES, (Continued) I lj FIGURE NUMBER' TITLE o
- t 7.7 l- Manual Control Self-Test Provisions l l 7.7-5 Rod Block Interlocks from Feutron Monitoring System 4 7.7-6.1 Reactor Control Bench Board -.Part 1 I! 7.7-6.2 Reactor Control Bench Board - Part 2 >
'l !
l} 7.7-7 Rod Block Functions l i l 7.8-1 - Nuclea'r Boiler System Piping & Instruments-tion Diagram l lj 7.8-2.1 Nuclear Boiler System Instrumentation !5 ,i. Diagram - Part 1 , ll 7.8-2.2 Nuclear Boiler System Instrumentation
, Diagram - Part 2
] 7.8-3 Reactor Vessel Thermocouple Locations j 7.9-1 . Recirculation Flow Control Illustration I 7.9-2.1 Reactor Recirculation System Valve Flow Control Functional Control Diagram 6 7.9-2.2 Reactor Recirculation System Valve Flow Control Functional Control Diagram 7.9-3 Reactor Recirculation System Flow Control Instrument Engineering Diagram 7.10-1 Feedwater Control System Instrument s Engineering Diagram l j 7.12-1.1 Process Radiation Monitoring System Instrument Engineering Diagram 12 I 7.12-1.2 Process Radiation Monitoring System Instrument Engineering Diagram 7.13-1 Area Radiation Monitoring System Instrument Engineering Diagram 7.17-1 Damping Coefficient Versus Decay Ratio l (Second Order Systems) (. 7.17-2 Hydrodynamic and Core Stability Model 7.17-3 Comparison of Test Results with Analysis 7.17-4 Total System Stability Model 7.17-5.1 10 Cent Rod Reactivity Step at 68% Power, 12 Natural Circulation 7.0-xviii
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zPs AMENDMENT 12 TABLE OF CONTENTS, (Continued) PAGE 9.3.6 Inspection and Testing 9.3-3 9.3.7 Operational Nuclear Safety Requirements ' 9.3-3 9.4 . GASEOUS RADWASTE SYSTD( 9.4-1 [ 9.4.1 Process Safety objective 9.4-1 , 9.4.2 Process Safety Design Bases
- 9.4-1 9.4.3 Sources of Radioactive Gas 9.4-2 ,
9.4.3.1 Process Off-Gas 9.4-2 l 9.4.3.2 Mechanical Vacuum Pump Off-Gas 9.4-3 9.4.3.3 Drywell Ventilation 9.4-3 9.4.3.4 Other Potentially Radioactive Gases 9.4-3 9.4.3.5 Gland Seal Condenser off-Gas 9.4-4 9.4.4 Description of System 9.4-4 12 9.4.4.1 Recombiner Subsystem 9.4-4 9.4.4.2 Short-Term Holdup Subsystem 9.4-6 9.4.4.3 Long-Term IIoldup Subsystem 9.4-6 f 9.4.5 Process Safety Evaluation 9.4-7 ! 9.4.6 Instrumentation and Control 9.4-8 l
,9.4.7 Inspection and Testing 9.4-8 .
9.4.8 Operational Nuclear Safety Requirements 9.4-8 s e a 9 9.0-11 k
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ZPS AMEN 0t4ENT 12 l I. 1 SECTION 9.0 - RADIG\CTIVE WASTE lf LIST OF TABLES l1 TABLE NLHBER TITLE PAGE ; t 9.2-1 Estimated Quantities of Principal Fission 9.2-6 Product Isotopes Released to the Environs per Year from the Liquid Radwaste System. l, if 9,2-2 Estimated of the Maximum Quantity of Fission 9.2-10 ll
'j Product Activity Present in the Liquid Rad- 7 waste System.
I j- ' 9.4-1 Estimated Quantities of Fission Product 9.4-5 Isotopes Released to the Environs per Year 12 from the Off-Gas Processing System. i 6 9.0-111
ZPS AMENDMENT 12 TABLE OF CONTENTS, (Continued) PAGE 10.5 FUEL POOL COOLING AND CLEANUP SYSTEM 10.5-1 10.5.1 Power Generation Objective , 10.5-1 ! 10.5.2 Safety Design Basis - 10.5-1 , 10.5.3 Power Generation Design Basis ' ' 10.5 10.5.4 Description 10.5-1 10.5.5 Safety Evaluation . 10.5-3 10.5.6 Inspection and Testing 10.5-4 10.6 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 10.6-1 j 10.6.1 Safety Objective 10.6-1 10.6.2 Safety Design Basis 10.6-1 10.6.3 Power Generation Design Basis 10.6-1 10.6.4 Description 10.6-1 10.6.5 Safety Evaluation ' f
.0.6.6 Inspection and Testing 10.6-1.2 l12 10.6-2 10.6.7 Operational Nuclear Safety Requirements 10.6-2 10.7 TURBINE BUILDING CLOSED COOLING WATER SYSTEM 10.7-1 i i
10.7.1 Power Generation Objective 10.7-1 10.7.2 Power Generation Design Basis 10.7-1 , 10.7.3 Description 10.7-1 ; 10.7.4 Inspection and Testing , 10.7-1 l4 l t 10.8 SERVICE WATER SYSTEM 10.8-1 } 10.8.1 Safety Objective 10.8-1 f 10.8.2 Safety Design Basis 10.8-1 10.8.3 Power Generation Design Basis 10.8-1 10.8.4 Description 10.8-1 ! 10.8.4.1 Equipment 10.8-1 : 10.8.4.2 Operating Conditions 10.8-1 3 10.8.5 Safety Evaluation 12 10.8-1.4 10.8.6 Inspection and Testing 10.8-2 10.8.7 Operational Nuclear Safety Requirements 10.8-2 7 10.0-11 L
r { ZPS AMENDMENT 12 ) l i SECTION 10.0 - AUXILIARY SYSTEMS i LIST OF FIGURES
?
i '! FIGURE NUMBER TITLE 10.2-1 Fuel Storage Arrangement l4 l t i 10.2-2 New Fuel Storage Rack
, 10.3-1 Spent Fuel Storage Rack 10.5-1 Fuel Pool Cooling and Cleanup System
'f 10.6-1 Reactor Building Closed Cooling Water System 11 10.7-1 Turbine Building closed Cooling Water System 4 10.8-1 Service Water Pump Structure Arrangement 1 t 10.8-2 Servie.e Water Pump Structure Arrangement 12 10.8-3 Service Water System
)
10.10-1 Control Room HVAC System 4 10.10-2 Station Ventilation System 12 10.10-3 Diesel-Generator Ventilation System, Service Water Pump House Ventilation System and Switchgear Heat Removal System 7 10.10-4 Service Water Pump House Ventilation, Make-Up and Service Water Pump Rooms Heat Removal Systems 10.12-1 Station Service Air System 10.12-2 Control and Instrument Air System 4
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10.0-vii l
ZPS AMENDMENT 12 SECTION 12.0 - STATION STRUCTURES AND SHIELDING TABLE OF CONTENTS PAGE 12.0 STATION STRUCTURES AND SHIEIDING 12.1-1 ,' 12.1
SUMMARY
DESCRIPTION 12.1-1 !I 12.1.1 Principal Structures and Classification 12.1-1 12.2 DESIGN CRITERIA 12.2-1 12.2.1 ' General Desigin Data 12.2-1 i 12.2.1.1 Interconnected Class I and' Class II Structures 12,2-1 12.2.2 Loading Combination and Stress criteria 12.2-7 12.2.2.1 General Notations for Class I Structures 12,2-7 , 12.2.2.2 Allowable Stresses 12.2-7 12.2.2.3 Structural Design Basis 12.2-12 12.2.2.4 Containment Proof Testing and Instrumentation 12.2-13 12.2.2.5 Seismic Recording System 12.2-15 12 12.2.2.6 Tendon Surve111nce Program 12.2-16 . 1 12.3 ANALYTICAL TECHNIQUES 12.3-1 12.3.1 Seismic Analysis 12.3-1 j 12.3.2 Asymetric Shear-Beam and Shear Wall Structures 12.3-1 l 12.3.2.1 Mass calculation 12.3-2
- 12.3.2.2 Spring Calculation 12.3-2 j_
12.3.2.3 Loading 12.3-3 ' 12.3.2.4 Shear Wall and Framed Structures 12.3-3 f 12.3.2.5 Shell Structuras 12.3-3 i 12.3.2.6 Horizontal Floor. Response Spectra 12.3-4 , 12.3.2.7 Forcing Functions 12.3-4 ' 12.3.2.8 Pesponse Spectra curves 12.3,4 7 12.3.2.9 4 tical Amplification Effects 12.3-4 ' 12.3.2.10 Analysis by owner 12.3-4 ' 12.3.2.11 Responsibility 12.3-4 , 12.3.2.12 Review of Vendors' Seismic Analysis 12.3-5 , 12.3.2.13 Damping 12.3-5 ' 12.3.2.14 Dynamic Soil Pressure.= 12.3-5 12.0-1
ZPS i AMENDMENT 12 1 TABLE OF CONTENTS. (Continued) PAGE 12.4.4.1 Specification for splicing Reinforcing Bar Using the Cadwell Process 12.4.3 l7 12.4.5 Structural Steel 12.4.6 12.4.6 Liner and Penetrations - 12.4-6 12.4.7 Construction Codes of Practice 12.4-6 12.4.7.1 Control Tests for concrete 12.4- 12 12.4.7.2 Evaluation of Test Results 12.4-12.5 PRINCIPAL STRUCTURES DESCRIPTION 12.5-1 1 12.5.1 Reactor Building 12.5-1 12.4.2 Auxiliary Building 12.5-1 12.5.3 Turbine Room 12.5-2 12.5.4 - Radwas te Facilities 12.5-2 7 12.5.5 Service Building 12.5-2 g 12.5.6 Service-Water Pump liouse Intake Structure 12.5-2 12.6 SHIELDING 12.6-1 12.6.1 Design Basis 12.6-1 12.6.2 Primary Sourees 12.6-2 12.6.3 Surveillance and Testing 12.6-4 s 1 l f l l 12.0-111
ZPS AMENDMENT 12 SECTION 12.0 - STATIpN STZ3CTURES AND SHIEIDING LIST OF FICORES FIGURE NIMBER TITLE 12.2-l' Reactor Containment Strafa Instrumentation y l12 - 12.3-1 Dead Load of containment only 12.3-2 Dead Load of containment med Building 12.3-3 Vertical PosC-Tensioning' .[ 12.3-4. l12 Horizontal Post-Tensioning i 12.3-5 Water in Suppression Pool 12.3-6 Liner Expansion 12.3-7 Internal Pressure 45 PSI 12.3-8 Earthquake 0.10g 12.3-9 l712 Flooding of Containment 12.3-10 Load Shift l I 12.3-11 Pipe Break Load (Concentrated, Eccentric) \ 12.3-12 Temperature Distributions Through Drywell Containment Wall ' 12.3-13 Permanent Loads (During Operation) 12.3-14 Permanent Loads + Temperature Effects ~ 7 12.3-15 Permanent Loads + Test Pressere 12.3-16 Interaction Diagram Axial fM vs Bending Moument 12.3-17 ., Interaction Diagram Axial TM vs Bending Moment 12.3-18 Intermediate Anchor for Vertical Tendon 12.3-19 Tendon Layout Around Equipment Hatch 12.3-20 Tornado Pressure vs Distance i 12.3-21 Resultant Surface Pressures IDae to Tornado 12.3-22 Resultant Dynamic Pressures 1pue to Tornado 12.5-1 Plot Plan 12.5-2 General Arrangement, Main Fleer Plan "A-A" l7 j 12.5-3 General Arrangement, Mezzanime Floor Plan "B-B" 12.5-4 General Arrangement, El . 496 '--O" and El. 503 ' -6" Plan "D-D" 12.5-5 General Arrangement, Sec t ions "A-A" and "B-B" l 12.5-6 Service Water Pump House Intale Structure 7 1 - 12.0-v J
I ZPS AMENDMENT 12 l i. TABLE OF CONTENTS. (Continued) PAGE i 13.6.4.1 General 13.6-2 I 13.6.4.2 General Description of Site and Surrounding Terrain '13.6-3 13.6.4.3 Emergency Organization 13.6-4
.1' 3.6.4.4 Liaison : local and civil Authorities and other Agencies 13.6-7 l 13.6.4.5 Protective Measures : On and off Site 13.6-9 l' 13.6.4.6 Emergency Treatament; Decontamination; Transportation 13.6-14 13.6.4.7 Training 13.6-15 13.6.4.8 Recovery and Re-Ehtry 13.6-15 13.6.4.9 Flood Protection 13.6-16 l12 i t 13.7 RECORDS 13.7.1 Initial Tests and operations 13.7-1 13.7-1 ll
,) l 3
13.7.2 Normal Operations 13.7-1 l 13.7.3 Maintenance And Testing 13.7-1 g i 13.7.4 Other Records 13.7-2 P 13.8 OPERATIONAL REVIEW AND AUDITS 13.8-1 l I 13.8.1 Administrative Control 13.8-1 , i 13.8.2 Routine heviews 13.8-1 13.8.3 Operations Review t'n-ni etee 13.8-1 ; 13.9 REFUELING OPERATIONS 13.9-1 ' 13.9.1 General 13.9-1 13.9.2 Training for Refueling Operations ,13.9-1 4 13.9.3 Inspection Procedurcs 13.9-1 13.9.4 Emergency Procedures 13 9-1 13.9.4 Emergency Procedures 13 9-1 l . 13.0-iv
1 'o ZPS AMENDMENT 12 SECTION 13.0 - CONDUCT OF OPERATIONS LIST OF FIGURES FIGURE NUMBElt ILILE l 13.2-1 Wm. H. Zinsner Nuclear Power Station Organization J 9 12 Chart I i 1 i l
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I i
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f f f I (
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l b i l 13.0-vi 1
ZPS AMENDMENT 12 SECTION 14.0 - PLANT SAFETY ANALYSIS LIST OF FIGURES FIGURE NUMBER TITLE 14.4-1 Plant Safety Analysis - Method for Identifying and Evaluating Abnorinal Operational Transients 14.4-2 Plant Safety Analysis - Method for Iden'tifying and Evaluating Accidents 14.5-1 Transient Results, Turbine Trip from High Power with Bypass 14.5-2 Turbine Trip Without Bypass I. 14.5-3 Transient Results, Turbine Trip from Low Power Without Bypass
'!- 14.5-4
{ Transient Results, Closure of All Main Steam Line Isolation Valves f' 14.5-5 Transient Results, closure of One Main Steam Line Isolation Valve
!- 14.5-6 Transient Results, Ioss of Feedwatre Heater i 14.5-7 Transient Results, Continuous Rod Withdrawal During Power Range l- 14.5-8 Transient Results, Pressure Regulator Failure 14.5-9 Transient Results, Inadvertent Opening of A Relief Valve or Safety Valve +
14.5-10 Transient Results, Loss of Feedwater Flow 14.5-11 Transient Results, Loss of Auxiliary Power 14.5-12 Water Level vs Time Following Loss of Auxiliary Power (RCIC Only)
- 14.5-13 Transient Results, Trip of Two Recirculation Pumps 14.5-14 Transient Results, Recirculation Pump Seizure 14.5-15 Recirculation Flow Control Failure - Increasing Flow 14.5-16 transient Results, Startup of Idle Recirculation Pump 12 14.5-17 F6edwater Controller Fai1.ure - 115% Demand l 14.6-1 Maximum Rod Worth Versus Moderator Density 14.6-2 Maximum Rod Worth Versus Power Level 14.6-3 Rod Drop Accident (Cold, Critical) Peak Fuel Enthalpy 14.6-4 Rod Drop Accident (Hot, Critical) Peak Fuel Enthalpy I
) I 1 I 14.0-viii l5
ZPS AMENDME.'.T 12 LIST OF FIGURES,(Continued) FIGURE NUMBER TITLE ~ 14.6-5 Rod Drop Accident (Power Range) Peak Fuel Enthalpy 14.6-6 Loss of Coolant Accident, Humboldt Bay Primary Containment Pressure Response 14.6-7 Loss of Coolant Accident, Bodega Bay Primary Containment ; Pressure Response l. 14.6-8 Loss of Coolant Accident, Bodega Bay Primary Containment ! Pressure Response ?
,14.6-9 Loss of Coolant Accident, Comparison of Calculated and Measured Peak Drywell Pressure for Bodega Bay and Humboldt Bay 14.6-10 Primary Containment Pressure Response 14.6-11 Drywell. Temperature Response 14.6-12 Suppression Pool Temperature Response 14.6-13 Loss of Coolant Accident, Primary Containment Leak Rate 14.6-14 Primary. Containment Capability Index to Metal-Water Reaction 14.6-15 Main SteamLine Break Accident, Break Location * ) 14.6-16 Steam Line Break Outside Drywell, Break Mass Flow Rate Transient-14.6-17 l Steam Line Break Outside The Primary Containment Core Inlet Flow (
[ 14.6-18 Minimum Critical Heat Flux Ratio Following A Steam Line Break I j; Outside The Primary Containment j' 14.8-1 Fuel Rod and Fuel Bundle Details i I f e 6 0
- 14. 0- ix 5 ,
I L_______-
ZPS AMENDMENT 12
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APPENDIX D.0 - QUALITY CONTROL SYSTEM LIST OF FIGURES 3 Figure Number Title I i l 12 ,' [ D.2 1 Engineering & Construction Organization ' l' Wm. H. Zinener Nuclear Power Statiois , 7 i D.2-2 The Cincinnati Gas & Electric Company
- Quality Assurance Interfaces i
D.3-1 Sargent & Lundy Project Organization ! D.4 1 BWR Nuclear System Project Quality Systems Organizational Structure i D.5-1 Kaiser Engineers, Inc. Organization - l Construction Services D.5-2 Xaiser Engineers, Inc. Project Functional Chart D.5-3 Functional Diagram QA/QC Program I
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l l D.0-xiv
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ZPS AMENDMENT 12 APPENDIX F.0 - CONFORhANCE TO AEC DESIGN CRITERIA TABLE OF CONTENTS PAGE F.0 CONFORMANCE TO AEC DESIGN CRITERIA F.1-1 F.1
SUMMARY
DESCRIFIION F.1-1 I F.1.1 Proposed General Design for Nuclear. Power Plant Construction ' Permits, July 16,1967 (The "70" Criteria) F.1-1 F.1.2 10CFR50 Appendix A - General Criteria for Nuclear Power Plants, May 21, 1971 (The'"64" Criteria) F.1-1 12 F.2 CRITERION CONFORMANCE (Ihe "70" Criteria) F.2-1 l F_. 2.1 Group I - Overall Plant Requirements (Criteria 1-5) F.2-1 F.2.2 Group II - Prctaction by Multiple Fission Product Barriers (Criteria 6-10) F.2-2
)
F.2.3 Group III - Nuclear and Radiation Controls (Criteria 11-18) F.2-4 F.2.4 Group IV - Reliability and Testability of Protection System F.2-6 (Criteria 19-26) F.2.5 Croup V - Reactivity Control (Criteria 27-32) F.2-9 F.2.6 Group VI - Reactor Coolant Pressure Boundary (Criteria 33-36) F.2-11 ; F.2.7 ' Group VII - Engineered Safety Features (Criteria 37-65) F.2-14 F.2.8 Group VIII - Fuel and Waste Storage System (Criteria 66-69) f F. 2-17 ; F.2.9 Group IX - Plant Effluents (Criterion 70) F.2-26 F.3 CRITERION CONFORMANCE (THE "64" CRITERIA) I F.3-1 8 F.3.1 Group I - Overall Requirements (Criteria 1-5) $ F.3-1 ; F.3.2 Group II - Protection by Multiple Fission Product Barriers (Criteria 10-19) F.3-3 I 3 F.3.3 Group III - Protection and Reactivity Control Systems F.3-7 12 (Criteria 20-29) F.3.4 Group IV - Fluid Systems (Criteria 30-46) F.3-11 F.3.5 Croup V - Reactor Containment. (Criteria 50-57) F.3-13 F.3.6 Croup VI - Fuel and Radioactivity Control , F.3-19 (Criteria 60-64) l i F.0-1 j
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2PS 1 AMENDMENT 12 i APPENDIX F.0 - CONFORMANCE TO AEC DESIGN CRITERIA LIST OF TABLES - f TABI.E NUMBER TITLE
- PAGE I F. 2-1 1967 AEC General Design Criteria - Group I (Overall F.2-3 f Plant Requirements)
F.2-2 1967 AEC General Design Criteria - Group II (Protec- F.2-5 tion by Multiple. Fission Product Barriers) F.2-3 1967 AEC General Design Criteria - Group III (Nuclear F.2-7 and Padiation Controls) F.2-4 1967 AEC General Design Criteria - Group IV (Relia- F.2-10 bility of Protection Systems) F.2-5 1967 AEC General Design Criteria - Group V (Reac-tivity Control) F.2-12 F.2-6 1967 AEC General resign Criteria - Group VI (Reac- ' F.2-15 tor Coolant Pressure Boundary) F.2-7 1967 AEC General Design Criteria - Group VII (En- ' gineered Safety Features) F.2-18 F.2-8 1967 AEC General Design Criteria - Group VIII (Fuel i F.2-25 12 1 and Waste Storage Sys tems) F.2-9 1967 AEC General Design Criteria - Group IX (Plant F.2-27 Effluents) F.3-1 1971 AEC Design Criteria - Group I (Overall Plant Requirements) F.3-2 F.3-2 1971 AEC Design Criteria - Group II (Protection F.3-5 by Multiple Fission, Product Barriers) F.3-3 1971 AEC Design Criteria - Group III (Protection F.3-9 and Reactivity Control Systems) i F.3-4 1971 AEC Design Criteria - Group IV (Fluid Systems) i F.3-14 F.3-5 1971 AEC Design Criteria - Group V (Reactor Containment) F.3-17 F.3-6 1971 AEC Design Criteria - Group VI (Fuel and F.3-20 Radioactivity Cor. trol) i l F.0-ii
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BLANK PAGE 1l s
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?PS AMENDMENT 12 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 1 SECTION 1.0 - INTRODUCTION AND
SUMMARY
This section' has been amended to incorporate answers to AEC-DRL ques tions. i_ All changes have been indicated by a vertical line and the amendment number (12) in the right margin of the page. - All pages (text, tables, figures) with changes have also been marked in l
' the upper right corner of the page with "AMENEMENT 12".
Figures that have been altered in any way are indicated by the amendment
- number in the upper right corner of the figure; note that there are no other marks. that would indicate changes in figure. On the page marked " LIST OF FIG-f URES", figures that have changed in any way are designated by a vertical line I
with the amendment number alongside the title of the figure. See example below:
} FIGURE NUMBER TITLE 1
[ 2.2-1 Station Site Area Topography i 12
't To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 1, _SECTION 1.0 - INTRODUCTION AND
SUMMARY
behind the red tabbed divider page titled "Amendmenta to Section 1.0" insert Pages 1.0-1 and 1.10-1. k
4-BLANK PAGE
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i 6 O 1 _ _ _ - _ _ _ _ _ _ _ _. -.~ ~~~~~~~~~~~~~~~
2PS AMENDMENT 12 1.0-1 (ZPS - February 23, 1971, AEC Oues tion 7.13) QUESTION Tables in the PSAR attempt to establish minimum requirements for operation, survelliance (testing) frequencies, and calibration frequencies for the reactor protection system, and engineered safety feature instrumen-tation an; control systems. Our review of these tables has revealed that 3 they (1) contain errors, (2) are inconsistent with system designs, and (3) ! are less conservative than current technical specifications. These tables l in order to remain a part of the application, should be corrected to remove errors and inconsistencies. Additionally, submit an analysis along with ! the design test and evaluation programs which have and will be used to sup- I port the requirements of these tables. ANSWER The infomation requested by the question and the analyses to sup-port the test programs will be supplied in the FSAR. The purpose of submitting these tables and the skeleton Operational Nuclear Safety Requirements (ONSR) paragraphs was to outline the approach to establishing technical specifications. Instead of removing all ONSR paragraphs and tables at this point, it is requested that they be consid-ered examples rather than design documents. The final version of the ONSR's will be provided in the FSAR.
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1.0-1
ZPS AMENDMENT 12 1.10-1 (ZPS - February 23, 1971, AEC Question 1.1) QUESTION Provide an up-to-date summary description of the status and results of the research and development programs underway in support of the Um. B. Zimmer Nuclear Facility. ANSWER ' Subsection 1.10 substitted in Amendment 7 and Appendix H.0 submitted in Amendment 9 supply summaries of the research and development program status f in support.of the Wm. H. Zineser Facility. 4 l i l 1 1 i f
'L.10-1 I
zPS AMENDMENT 12 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 1 SECTION 2.0 - SITE This section has been revised to incorporate new information and minor editorial changes as well as answers to AEC questions. All changes have been indicated by a vertical Ifne and the amendment number (12) in the right margin of the page. All pages (text, tables, figures') with changes have also been marked in
;i the upper right corner of the page with " AMENDMENT 12".
i Figures that have been altered in any way are indicated by the amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", I figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below: FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography l 12 , To update your copy of the En. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 1, SECTION 2.0-SITE, remove and destroy Table of Contents Pages 2.0-ix, 2.0-xiii, 2.0-xv and 2.0-xix and replace with the amended Pages 2.0-ix, 2.0-xiii, 2.0-xv and 2.0-xix.
- 2. In Volume 1, SECTION 2.0-SITE remove and destroy the following text pages and replace with the amended pages listed below:
REMOVE PAGE REPIACE WITH AMENDED PAGE 2.5-86 2.5-86 2.5-91 2.5-91 2.5-91.1 2.5-91.1 through 2.5-91.4 2.5-98 2.5-98 through 2.5-98.2
- 3. In Volume 1, SECTION 2.0-SITE, remove and destroy Figure 2.4-7 and replace with Amended Figure 2.4-7. Behind Figure 2.5-51 insert new Figures 2/5-52 and 2.5-53.
- 4. In Volume 1, SECTION 2.0-SITE, behind the red tabbed divider page titled " Amendments to Section 2.0"
4 /' 1
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ZPS AMENDMENT 12 P i a. Behind Page 2.2.5-1 insert Pages 2.3-1 through 2.3-7.
- b. Behind Page 2.4.5-6 insert Pages 2.4.5-7 through 2.4.5-10
- c. Behind Page 2.4.5-10 insert Page 2.4.5.1-1.
- d. Behind Page 2.4.5.1-1 insert Pages 2.4.5.3-1 and 2.4.5.3-2.
- e. Behind Page 2.5.4.7-1 insert Pages 2.6-1 through 2.6-18.
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ZPS AMENDMENT 12 TAR'.E OF CONTENTS, (Continued) PACE 2.5.4.4.3.1 Evaluation of Liquefaction Potential 2.5-88 5 2.5.4.4.3.2 Soil Stability Analysis 2.5-91 2.5.4.4.3.3 Bearing capacities 2.5-91.4 2.5.4.4.3.4 Static settlement 2.5-91 A i 2.5.4.4.3.5 Dynamic sectiement 2.5-94 2.5.4.4.3.6 Rock - Soil - Structure Interaction 2.5-94 2.5.4.5 Subsurface Walls 2.5-94 5 2.5.4.6 River Bank Et4111ty - 2.5-98 2.5.4.7 Effects of Nearby Quarry Blasting on Plant Construction and Operation 2.5-98 2.5.4.8 Future Units 2.5-99 2.5.4.9 References 2.5-100 2.5.4.10 Soil Liquefaction Analysis Report No. 43 2.5-102 1 I l 2.0-ix j 4 _ o
ZPS AMENDMENT 12 LIST OF TABLES, (Continued) TABLE NUMBER TITLE P_ ACE 2.5-1 Bulk Sample Descriptions 2.5-6 2.5-2 Direct Shear Tests on Recompacted Sand Samples 2.5-7 2.5-3 Rock Unconfined Compression Test Results 2.5-8 2.5-4 Cyclic Triaxial Compression Test Data - Modulf and Damping 2.5-10 2.5 5 Cyclic Triaxial Compression Test Data - Liquefaction 2.5-12 2.5-6 Resonant Column Test Results 2.5-14 2.5-7 Shockscope Test Results 2.5-16 2.5-8 Relative Density Test Results : 2.5-19 4 2.5-9 Specific Gravity Test Results 2.5-20 2.5-10 Permeability Test Results 2.5-21 2.5-11 Soil Conditions in Plant Construction Area 2.5-34 ! 2.5-12 Earthquake Epicenters - 82.5'-86.5 ' Wes t * ' Longitude - 37*-41' North Latitude 2.5-56 2.)-13 Earthquake Epicenters Intensity V and Greater 1699-1969 - 84*-90* West Longitude - 35'-39' North Latitude ' i 2.5-60 2.5'-14 Moduli and Damping values l 2.5-76 5 ! 2.5-15 Modified Mercalli Intensity (Damage) Scale of l ! 1931 2.5-77 l 2.5-16 Structural Loading Conditions s 2.5-82 l 2.5-17 Subsurface conditions 2.5-85 2.5-18 { Generalized Subsurface Properties 2.5-90 j 2.5-18.1 l Results of Stability Analyses During Liquefaction 2.5-91.2 l 12 2.5-19 Ultimate Bearing Capacities 2.5-92 l 2.5-20 Estimated Total and Differential Settlements 2.5-93 I 2.5-21 EstimatedStructures Adjacent Differential Settlements Between 2.5-95 f 2.5-22 Parameters for Analysis of Rock-Soil-Structure 5 - In terx tion 2.5-96 2.5-23 Estimated Lateral Pressures 2.5-97 ' 2.5-23.1 Results o { Slope Stability Analysis Natural ) River Ban
.2.5-24 2.5-98.2 l 12 ,;
Properties of Analytical Model 2.5-105 l5 2.0-xiii ( .. . . . . . . . . . .
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!5 ZPS. ' '
,j AMENDMEW 12
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.: -LIST- OF FIGURES -
, (Continued) / )
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.j, / j l 2.4-1 I Map of the Ohio River Drainage Basin and the Proposed Site of the Wm.. H. Eisumer Nuclear Power Station j
2.4-2 Location ~ of Public Grou'nd Water Supplies
- ) 2.4 ,.
Location of Private Wells
- I i 2,4-4 Ohio River Basin Hydrologic Sub-Area Map i
{ l 2.4-5 Seven-Day Average Low-Flow Frequencies on the Ohio River ) ;
' ~
2.4-6 ' Relative Flood Stage Frofties - Ohio River I 2.4-7 Ohio River Profiles 12
.2.4-8 Ohio River Stage Frequencies i
2.4-9 Precipitation: Area-Depth Duration , j 2.4-10 Standard Project Flood Hydrography t 2.4-11 Probable Maximum Flood Hydrography 4 2.4-12 Locations of Cross Sections in Markland Pool 2.4-13 Typical Cross Sections in Markland Pool 2.4-14 Cross Section at Wm. H. Zinaner Site 2.4-15 Cround Water contour Map t , 4 7 l l 9 f_ 2.0-xv
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ZPS i AMENDMENT 12 L LIST OF FIGURES, (Continued) FIGURE NUMBER _ _ TITLE p 2.5-29 Thickness of Glacial Drift 2.'5-30 Wisconsin Glacial Horaines 2.5-31 Effects of Glaciation 2.5-32 Variation of Shear Modulus 2.5-33 ' Lumped Mass'Model
'g 2.5-34 Shear Stress at 10 Feet - Taft Ea 2.5-35 l Shear Stress at 30 Feet - Taft 2.5-36 Shear Stress at 45 Feet - Taft 2.5-37 Shear Stress at 55 Feet - Taft 2.5-38 Shear Stress at 65 Feet - Taft 2.5-39 Shear Strens at 75 Feet - Taft 2.5-40 Shear Stress at 92 Feet - Taft 2.5-41 Shear Stress at 110 Feet - Taft 2.5-42 Results of Cyclic Triaxial Tests 2.5-43 Stress vs cycles cc, Liquefaction - 10 Feet 2.5-44 Stress vs Cycles to Liquefaction - 30 Feet 2.5-45 Stress vs cycles to Liquefaction - 45 Iwt j 2.5-46 Stress vs Cycles to Liquefaction - 55 Feet o-2.5-47 jj Stress vs Cycles to Liquefaction - 65 Feet !
2.5-48 Stress vs cycles to Liquefaction - 75 Feet 2.5-49 Stress vs Cycles to Liquefaction - 92 Feet : ,
', .2.5-50 Stress vs Cycles to Liquefaction - 110 Feet i
l j
-' 2.5-51 Liquefaction Potential - Taft {
2.,5.52 j Results of Stability Analyses Radwaste Building ' j ! (Case 12) j 2.5-53 Resnilts of Stability Analyses Natural River Bank
' ' 12 ! ;
(Case 3) ' 1 1 2.0-xix ! I j i
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ZPS AMENDMENT 12 1 2.5.4.4 Foundation Considerations I
- j. 2.5.4.4.1 General l The results of field explorations, laboratory tests and foundation en-i gineering analyses indicate that the site is suitable from a foundation stand-point, for construction of the nuclear power station facilities. The main build-ing complex, including all Class I structures, will be supported on mat founda-tions established on a prepared subgrade at or near the grades shown on Table 2.5-16 with the exception of the river intake structure, which will bear on rock, and the service water system, which will be supported on piling. 12 Major considerations in foundation analysis have been the evaluation of liquefaction and consolidation potential of the on-site sands. In order to pro-vide an adequate margin of safety against liquefaction, and to limit settlement of structures, all sands between foundation level and elevation 450 will be ex-cavated and recompacted to a relative density of at least 85 percent.
2.5 4.4.2 Site Preparation 2.5.4.4.2.1 General backfilling Site preparation will consist of stripping, excavating, dewatering, and operations. 5 Subsurface soils throughout the structure area will be densified between foundation level and elevation 450. Densification will be accomplished by con-ventional earthwork procedures. Detailed quality assurance will be maintained { throughout all phases of site preparation. If sands are excavated and recom-Facted, each layer of recompacted material will be tested by field density test methods prior to the placement of a subsequent layer. The subsequent discussion of the various site preparation operations per-tains to densification of the sands by conventional earthwork operations. 2.5.4.4.2.2 Stripping Trees, bresh, grass, roots and other deleterious materials will be stripped from areas to be occupied by structures and from all areas to be filled. All top-soil will be removed prior to general ev.cavation operations. l5 2.5.4.4.2.3 Excavating Excavations for major structures will extend to elevation 450. Excavated clayey silt and silty clay soils will be used as site fill remote from the area of structures. Excavated sands above elevation 450 will be stockpiled for use as backfill. 5 Excavations for structural foundations will extend a minimum lateral distance beyond foundation lines equal to the vertical distance between 2.5-86
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K ZPS J' _ AMENDMENT 12 2.5.4.4.3.2 Soil stability Analysis , y Soil liquefaction is a phenomenon which occurs when loose, saturated, granular soils are subjected to vibratory loading. Using the soil properties foundhas enon at the beensite together with the appropriate seismic excitation, this phenom-investigated. It has been determined that, in all cases investi-i gated, the induced stresses across the soil profile are less than those required for liquefaction.
. a.
No. 43, "For Soila Liquefaction better understanding Analysis"the full text of the Sargent & Lundy Report included at the end of Subsection 2.5, Paragraph 2.5.4.10.for En. H. Zimmer'5Nuclear i a Bar. J on the above analysis, granular soils at the site are considered to 450.have an insufficient margin of safety against liquefaction above elevation 4 Soils below elevation 450 indicate an acceptable margin of safety by the ' above outlined analysis, and reference to other published data confirms this conclusion. (Ambraseys and Sarma, 1969: Castro, 1969; Seed, 1969). Conclusions resulting from the above outlined study have been used in 5
-- selecting the method of foundation support; namely, mat foundation support on a < prep'ared and extending bas toe of granciar450.
elevation soils compacted to a relative density of 85 per cent Generalized properties of soil strata at the 1 site after compaction of foundation materials and placement of fill and backfill materials adjacent to structures, are presented in Table 2.5-18
}
Although the compacted soils immediately below mat foundations. will be l stable under carthquake loading, the soils above elevation 450 and outside the compacted basis zoue were assumed to be susceptible to liquefaction daring the design earthquake. 11 Analyses were performed to evaluate the possibility that the } - If quefaction of these foundations supported thereon. loose adjacent soils might affect the stability of the i The analytical procedure used in these studies was a pseudo-dynamic t slope stability analysis based on the method of slices. The general method of l{ analysis is to approximate the actual failure surface by an arc of a citcle. y The failure zone is divided into vertical slices. For eac.h slice, moments are
-: developed about the circle center by the internal forces and external loads in-cluding the weight of the soil mass, the weight of the portion of the structure I resting on the zone of rotation, and a dynamic force equal to the horizontal acceleration times the vertical load at any point in the soil profile, applied 12
- as an extra static force in the direction of slope instability. The algebraic !
; sum of these moments is the overturning moment tending to cause failure. The l resisting moment is provided by the soil shearing resistance developed along
{ soilsfailure the on thesurface face ofplus the the hydrostatic force exerted by the adjacent liquefied slope. l factors of safety calculated for each.Several trial circular surfaces are assumed and 5 factor of safety is the most critical. The circle that yields the minimum For studies outlined herein, a computer 2.5-91 ; j
1
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ZPS AMENDMENT 12 program was utilized to analyze a large number of failure circles with different l centers and radii. 12 j The above outlined analytical procedure was used with the following prin cipal assumptions:
- a. Ground surface elevation at 520; water level elevation at 508.6.
- b. No shearing resistance in the liquefied soils.
- c. Weight of saturated liquefied soils equal to 110 pcf. 11
!. d. Horizontal acceleration of 0.2g throughout the soil profile.
- e. Horizontal dynamic force equal to 0.2g vertical load at any point in the soil profile, included within the liquefied layer.
- f. Maximum foundation pressures for dynamic loading equal to 150 per cent of dead plus live static loading tabulated in Table 2.5-16.
- 3. Boundary between undisturbed compacted fill and liquefied soils assumed to be a sloping line extending from the foundation base to elevation 450. Analyses were performed assuming various points of intersection 12 of this line and elevation 450.
De results of the stability analyses indicate that all Class I structures, will be stable during the postulated design basis earthquake provided that the compacted fill underlying foundations extends a minimum later.a1 distance beyond foundatice lines equal to the vertical distance between elevation 450 and the planned building. base elevation of the foundation for all structures except the reactor, For the Reactor Building, the bottom of the compacted fill will extend an additional 30 feet from the distance indicated above. The resulting factor of it safety for this condition is 1.0 for the Reactor Buf1 ding, and in excess of 1.0 for other Class I structures. A factor of safety equal to or greater than 1.0 is considered acceptable given the very conservative nature of the assumptions including: (a) the assump-tion of a zone of liquefied soils extending from elevation 450 to the ground surface; profile. (b) the use of a horizontal acceleration of 0.2g throughout the soil (A horizontal acceleration of 0.2g will occur only for one or two peak cycles during the postulated earthquake anc only at or near the ground surface.) 2.5-18.1.ne results of the above stability analyses are summarized in Table l l A typical stability analysis (Case 12) for the Radwaste Buf1 ding showing the soil profile, the assumed soil parameters, and the computed critical failure surface is presented on Figure 2.5-52. - 12
)
A factor of safety equal to or greater than 1.0 is considered acceptable 2.5-91.1
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ZPS AMENDMENT 1? { given the very conservative nature of the assumptions including: (a) the assump-tion of a zone of liquefied soils extending from elevation 450 to the grour.d j surface. Even if liquefac':fon were to develop within zones of the loose natural l silty sands, it is highly improbabic that complete li"uefaction q of the natural { silts, clays, and the adjacent compacted fill would also occur; (b) the use of 12 ] a horizontal acceleration of 0.2g throughout the soil profile. A horizontal ac-celeration of 0.2g will occur only for one or two peak cycles during the posto-lated earthquake and only at or near the ground surface; (c) the horizontal dy-namic force applied to the liquefied soils. Liquefied soils cannot transmit the horizontal shear stresses induced by the postulated earthquake, p.
- 2. 5-9 . 3
ZPS AMENDMENT 12
- 2. 5.4.4. 3. 3 Bearing capacities Major structures will be supported on mat foundation established on struc-tural fill at the elevations tabulated' in Table 2.5-16. Ultimate bearing capaci-ties and' indicated factors of safety for mat foundations are presented in Table 2.5-19. We tabulated factors of safety have been determined by assuming that 11 each structure, or portion of it with a variable sat elevation is isolated from the adjacent structure.
All structures will have foundations proportioned such that the peak i foundation loading during seismic loading condition will not exceed 150 per cent of. the foundation loads tabulated in Table 2.5-19. Thus, factors of safety under short duration seismic loading will not'be less than about two-thirds of the values tabulated in Table 2.5-19. From a bearing capacity standpoint the most unfavorable condition during gg I dynamic loading will occur if soils above elevation 450 and outside' the compacted zone are assumed to liquefy. Under this condition, the bearing capacity of the ! soils will be reduced. Factors of safety during this conuition will .not be less 1 than;about one-third of the values tabulated in Table 2.5-19. ' 2.5.4.4.3.4 Static Settlement Total and differential settlements which the proposed structures will-undergo due to the static Iceds have been examined by the following methods: 5
- a. One dimensional conventional settlement analysis using the results i
L 2.5-91.4 12
ii zps - li fj. AMEND E 12-
- i R Surcharge pressures from adja:ent . structure > or loads will be adde
- to the tabulated lateral pressures in the design of all walls.
The lateral pressures for rigid walls subjected to static loading ' ave r been assumed equal to earth pressure at rest. ' For cantilever salls subjected to stahc loads, active earth pressure has been assumed. In destp, a factor of safety of 1.5 will be applied in the use of these values to allow for residaal soil pressures which could result froa the high degree of compaction whi h the backfill soils will receive. 1 The lateral pressure for both rigid walls and cantilever walls subjected to dynamic loading have been determined in accordance with the methods outlined by Seed and Whitman (1970). In design, a factor of safety of 1.1 will be applied in the use of these values. 2.5.4.6 River Bank Stability There is evidence of sloughing along the Ohio River bank on the west edge of the ~ site. Blocks of sitt and sand have slipped down to the river shoreline carrying trees and brush. It is believed that this instability is a result of normal erosion due to undercutting of the bank by wave action. No deep-seated stability failures have been reported or observed along this reach of the Ohio River during the past 200 years. There was no reported damage of the river banks in the site vicinity during the 1811-1812 New Madrid earthquakes, and it is believed that the intensities felt in the site vicinity from these shocks were the largest from any known seismic event. On this basis, i deep-seated bank stability problems are not expected under normal conditions. II Analysis of the river bank stability has been performed for both static and pseudo-dynamic loading conditions. the analytical procedure used in these studies was the circular failure analysis using the method of slices outlined in Paragraph 2.5.4.4.3.2. The results of the stability analyses f:r the river bank are summarized in Table 2.5-23.1. A section through the river bank showing the principal assumptions incP ling soil parameters, soil profile, river level and the computed critical circle for Case 3, assuming a horizonta! a celeratien of . 0.2g throughout the soil profile, is shown on Figure 2.5-53. A minimum factor of safety of 1.5 against deep-seatea failure has been 12
- obtained for static loading conditions. Ltder pseudo-dynamic 1cading conditions factors of safety against deep-seated failures of 1.0 and 1.2 tave been obtained for horizontal accelerations of 0.2g and 0.lg respectively, i
A factor of safety equal to or greater than 1.0 is consi;ered acceptable given the very conservative nature of the assumption of a horiz:mtal acceleration of 0.2g throughout the soil profile, and the rather conservative soil para aters assumed in the analysis. 2.5-96 \ ' l l r ___ _ _ _ _ - - _ - - - _ - I '
9 ZPS AENDMENT 12 The effect of liquefaction of the loose sands adjacent to. the river bank was not included in the above analysis. Due to the pre-earthquake stress condi- l 1' tions in earth banks, it has been shown by Seed (1968) that liquefaction of a gy 1 sand layer adjacent to a slope is less likely than well behind the slepe. The mode of failure of the entire soil mass including the river embankment under these conditions would be as described in Paragrraph 2.5.4.4.3.5. 2.5.4.7- Effects of Nearby Quarry Blasting on Plant Construction and Coeration The Black River Mining Company operates an underground limestone mine at Carntown, Kentucky, approximate 1" two miles upriver from the proposed site of the Zimmer Nuclear Power Plant, Moscow, Ohio, I t l-1 I I l 2.5-98.1 _ _ _ _ _ _ _ - _ _ _ _ _ - - - _ _ - - - - - - - - - -i
ZPS l ! i AMENDPET 12 TABLE 2.5-23.1 { j RESULTS OF SLOPE STABILITY ANALYSES 1 NATURAL RIVER BA3_K f CASE-CRITICAL CIRCLE DATA FOR DEEP-SEATED TY CEN1ER COORDINATES RADIUS NO. X FACTOR OF Y IN FEET _SAFETT 1 CONDITION 165.0 498.0 62.0 1.5 Static Loading Water Level @ - 2 Elevation 457 170.0 560.0 !
, 135.0 1.2 Pseudo-dynamic . Loading Horizontal Acceleration
= 0.1 g Water Ievel @
3 Elevation 457 183.0 577.0 159.0 1.0 Pseudo-dynamic Loading Horizontal Acceleration
= 0. 2 g ->
Water Level @ l
.; 4 Elevation 457 I 160.0 515.0 80.0 2.4
.
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9 Llevation 508 . l u
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\ >-
2P5 - 1 j AMENDMENT 12
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i 2.3-1 (ZPS - April 9.1971._AEC Question 2.11) Qt'ESTION 1 I During our february 26, 1971 meeting, the Zimmer site meteorologic al ! conditions f sr the loss-of-coolant accident analysis were dircussed. In order for us to select the appropriate Pasquill-Turner type conditions (F vs' C)'for the Zimmer site, we need the following infermatior: . e a. Compare the wind speed data obtained at the lower site lecetion with the Greater Cincinnati Airport (CVC) data for night condi-tions and Pasquill-Turner Type E, F, and G conditions. b. Compare the CVG wind speeds and wind speeds obtained at the upper and lower site location not previously shown.
- c. Provide a figure similar to Iigures 2.3-2 and 2.3-3 that show a comparison of, wind speed data obtained at the site end at CVG.
d. Describe the insertanentation used in the short term meteorology-cal program, the on-site program for obtaining data and an indi-cation of the starting wind. speed on the instruments and elevation of ' the instruments. ANSWER a.
.9 simultaneous data are available for tht. Creater Circinnati Airport (CVG) and the Lower 31te for the period 23 September 1969 through 30 April 1970.
Data at the Lower Site were obtained with at MRI Mechanical Weather Station located at 33 f t height at the location " Lower Site," marked on Figure 2.3-2 of the PEAR Amendment 11. Average wind speeds were as follows: _ Site Conditions Average Speed No. Dat_a CVG Night Lower 4.2 m/s 2163 Night 1.5 m/s CVG 2163 E,F,G 2.8 m/s Lower 1032 E,F,G 0.8 m/s 1032 b. 8 September 1970 thrmzh 18 September 1970. Additional data were obtained o MRI Mechanical Weather Stations were located at the Upper Site (8 f t), Frieberg (8 f t), North Lower Site (8 f t) and South Lower Site (8 f t) as shown on Figure 2.3-3 of the PSAR Amend:sent 11. l An catedMRI Vector in Figure Vane was installed at 33 f t at the " Lower Site" location as ind 2.3-3. i I i 2.3-1 I V
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AMENDMEYT 12 { t 1 Average wind speeds at all locations- for the 11-day . ere: period w i S_i,,,tg i
}
Averaae Speed l CVG No. Da ra_ i Upper 3.1 m/s i
.1.0 214
- North Lower 0.8 214 South Lower 211 0.7
! Lower (Vector Vane) 1.2 214 i
( 214 i i 8 to 18 September 1970. For this briefer are as follows: mparisons period, e o a e
. ,SR,e,' ;
Averaae Speed CVG No. Da ta Upper 2.9 m/s 1.0 91 Frieberg -91 1.2 91 j
- 1970 show the fol?owing: Average wind speeds for all data from 23 Se ugh 30 April' t
B,tpe, N:e'ame Speed _ CVG No. Data Lower 4.8 m/s 2.0 4075 Upper 4075
' 2.0
'3953 than grade Although level the elevation of the Upper Site (800 f t) is consid about, the same as(at the Lower Site (33 f t above gec at surrcending by the Uppertrees. Site was located at 8 f t above ground The .
and instrument y sheltered was partiall exposure.and 100 f t lower in elevatien indicates a higher wind speed than the Uppe spite of being
, should give a more represe. The 50-f t tower to be located near the old Upper S above the valley. ntative measure of the wind speed at the higher levels I t
- c. .
at' CVG and the Lower Site as well as CVG ve, n s pe*48 the Uppe The data sample the lower Site was 33 f t and 8 f t for Instrument the Upper heightSiteis ai of , above and half below the dashed i.e., lines.ures half of theindicate sitDashedthe e data fallmedian v lines in the fig-i ' 4. used for the evaluation of the Zinmer Site. Figure 2.3-3 shows the at were two Mechanical Wather Stations (WS) at the site duriDames and Moore, Inc. , insta ember 1969 chrough 30 April 1970. ng the period of 23 Sep-1 One WS was located at the Upper Site, the
- 2. 3-2 i
i 4 ZPS [ AMENDMENT 12 other was placed at the middle of the plant site area and called the Lower Site,
~l see Figure 2.3-2.
The WS at the Upper Site was mounted on a standard tripod which exposes the sensors at an 8-f t level above the surface. The Upper Site location was significantly sheltered from all wind., by adjacent trees except from a northerly direction. The sheltering of the instrument by the trees and the lower height above ground of the sensors significantly reduced the wind speeds that would have been expected at this site. The WS at the Lower Site was mounted at the top of a utility pole. ) The sensors were exposed at a height of 33 f t. The wind speed and direction data from the two WS locations were reduced
. as hourly averages.
o An intensive on-site measurement program was conducted during the period - of 8 to 18 September 1970. - Surface Measurements - Four WS were located as shown on Figure 2.3-3.
~' All stations were mounted on the standard tripod with the sensors exposed at F 8 f t above the surface. The data for wind speed and wind direction were reduced as hourly averages for every hour of the collection period.
The Vector Vane was installed on the utility pole at the Lower Site location. Data from it were reduced as 10-minute averages at hourly intervals. Airborne Measurements - An aircraf t instrument package was used to mea-sure the vertical distribution of temperatures over the site. The soundings were normally taken at 0700, 0900,1100,1500,1700 and 1900 EST, weather per-mitting, from 50 f t above the river to 3500 ft. Pilot Balloons - The pihals were taken simultaneously with the aircraf t temperature soundings. The data for wind speed and direction were read at one-half minute intervals to 4000 f t. The p'.bal site is indicated in Figure 2.3-3 of the PSAR Amendment 11. Smoke Releases - A hilltop in Kentucky, west of the site, was chosen as the location for smoke releases (see Figure 2.3-3). The Uemical smoke pots
- i. were released from a 50-f t level above the surface at various times of the day' to provide visual representations of the diffusion conditions in the valley.
A few small smoke sources were released at the North and South Sites to provide additional information on wind trajectories under stable conditions, f 1.e., at sunrise or just prior to sunset.
=
2.3-3 ,
i I
-- i ZPS AMENDMENT 12 35 nun photographs were taken of the smoke releases f rom the aircraf t.
The starting speeds 'of the Mechanical Weather Station and the Vector
, Vane have been determined to be 0.75 mph. Specifications for the sensors are shown in the attached brochure under the heading of " Wind Sensors 1074 and 1075"-(same as HWS) and Tkctor Vane. The 1074 (with light chopper) is the instrtunent to be used in the new tower system. Even though the cups of the !'
Mechanical Weather Station will start at about 0.75 aph, there is a slight underes'timating of the wind until a wind speed of 2 to 3 mph is reached.. Thereafter, a time representation of the wind is achieved. For the Vector Vane, this point on the response curve is reached earlier, between 1 and. 2 : mph. For this reason as well as. the starting speed problem, winds at low speeds -(less than 3 mph) are.somewhat 'underes timated. i l L I L L l 2.3-4 1
ZPS AMENDMENT 12 PAGE 2.3-5 1 IS METEOROLOGY RESEARCH, INC'S. METEOROLOGICAL SENSORS AND SYSTEMS I SYSTEliS BROCHURE
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ADD 3 295 StasGLE S 200 DUAL te0UNTED en MhtL $ S00 w as.t,,,,, TSPLE h60VNTED Im MNEL 8700 For more details, call us at (213) 791-1901 Ext. 301
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In some applications, a single Wind Direction and Wind Speec is required. The Model 1C j supply and all signal conditioning in the sensor. The output comecto provides two pairs of wit. a Switch Cbsure for 1/10th mile count. h h h hhh l wino pacCre* caso serra evnce esa n na more - s4o* rou corent:0wcren i/cen weLe Couwr si.oso
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Prices listed are f.o.b. Altadena, Californ ._1.__ _ _ _ _ _ _ _ _ _ _ . _ . ____ - - - - - - - -- a.,ii ......a : - ... - - - - . - - .~- ee .
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.I Wind Direction and Wsul Speed is required. The Model 1075 anemometer hou 1
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COMPARISON OF CVG WIND SPEEl SIMULTANEOUS WIND SPEEDS FOR l THE UPPER SITE 2.3-6 (23 Sept.1969 thru 30 April 1970) ,
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aa: - : : : : : : 3 . . . . . . . . . g (29/Ttu) paads puin ons sonog a WM, H.ZIMMER NUCLEAR POWER STATION l l PRELIMINARY SAFETY ANALYSIS REPORT j FIGURE 2 COMPARISON OF CVG WIND SPEEDS TO SIMULTANE0US WIND SPEEDS FOR THE LOWER SITE 2.3-7 (23 Sept.1969 thru 30 April 1970)
ZPS . AMENDMENT 12 2.4.5-2 (_ZPS - April 9. - 1971. AEC Ques tion 2.18) fl QUESTION 5
-Determine the effects of a radioactive spill during the Probable Maximum Floor (PMF) on potential contamination of public ground water supplies (wells).
The pathways to ground water supplies that should be' considered in your evalua-
- . tion are (1) existing surface-ground water connections such as wells, pits, old site bore holes and the eifects of percolation, and (2) the possibility of the
- ground water gradients reversing during long durations of high river flood flows.
ANSWER f The following assumptions were used in analyzing the effects of a radio-active spill during the Probable Maximum Flood:
- 1. The river flood waters attain free access to the interior of the i
radwaste building at a stage of $24 f t, or 4 f t above grade, through collapse of the doorway. This value was chosen because .t it gives a depth of about 4 ft over the doorsill and would permit '!' mixing of river water with the contents of the building within a 'Is. reasonable time.
- 2. The entire 35,900 gallon content of three radwaste tanks is emptied l into the building at or prior to the time the door collapses. The
+
total rad'oactivity is considered to be the same as that shown in the table supplied with the answer to Question 2.4.5-1 (AEC - October 13, 1970, Question 2.7).
- 3. The radioactivity is considered to escape to the river completely and at a uniform rate in 15 hours. This time period gave consid-eration to the width of the door (19 f t), the size of the building (137 f t by 85 f t), and water depth in the building (about 11 f t).
A shorter time for the pollutant to escape to the river would ] e increase the concentration of the pollutant cloud but decrease its dura tion. The net effect on ground water would be negligible. Table 2.4-2, Amendment 2, lists 11 towns or water districts within I 25 miles of the Wm. H. Zinsner site that receive their water supply from wells, i Five of these groups of wells are upstream of the Wm. H. Zimmer site, the near-est being about six miles. Contaminated surface water could not extend upstream and a ground water gradient could not extend upstream any significant distance. These ground water supplies, which a: e Augusta, Kentucky and Felicity, Higginsport, t Ripicy, and Chilo (Shiloh in Table), Ohio, could not possibly be contaminated by ) a radioactive spill at the Wm. H. Zimmer Site. i Two groups of wells, Indian Hill and Milford, are in the general vicinity of Cincinnati. The answer to Question 2.4.5-1 showed that, for the worst possible 2.4.5-7 l I
__,_,___-a-,--?' " " ^ ^ _ S 2PS AMENDMENT 12 spill condition, the river water at Cincinnati never cwould rea h
, concentration. _ The great additional dilution on the allowablethe g entering y significant ground water pollution in this y rulearea.whateve out any Therefore, they need not be consiAred furtherof .
reach them. wells a downstream from the Wm. H. Zimmer site.The ver remaining Valley four l Tate-Monroe Water Association in Ohio.
, New Richmond Union in OhioTownships and i
and therefore the most critical.of the Tate-Monroe Water Assoc
. H. Zimmer site c eream. They were selected for further study. Any spill at the Um, B. Zimmer site will reaching the main channel. be n when on facing e somewhat in the r down-from the Wm. H. Zimmer site The Tate-!!onroe area is about lesscomputed was in theforTate-Monroea point three miles area,.but, downstream from because Wm of this de 3
I
. H. Zimmer, i , section at Tate-Monroe even though uthere g the" crossis prac ank flood-plainthefor fore, com two of the .five miles between Wm. H. Zimmer and Tate M Monroe area. puted concentrations were doubled for application - onroe. There- to the the river water at Tate-Monroe of 0.68 x 10-7These studies oactivity in indica This is less than the allowable concentration forngdrinkimicrocuries water.
per cubic The concen-
.bytration the following in the wells would be a very small fraction discussion. ofe this va by 25 f t or more of water before theeas floodwill belevel inundatedreach Therefore, any radioactive pollutant, which already e radwaste building.
limits, that might seep into the grocad water would be vwould be below drin ery greatly diluted. In in the arca and available for infiltration. o addition utcJ ' water will there be wil the peak concentration for about 42 hours . Concentration would exceed 0.1 of Monroe area, and other downstream grcops es that theof wells i Tate-be submerged for four days prior to the time of e Ohio any River Valley, will Therefore, all open wells, pits and here holes will bpossible radioactive spill. taminated water well before the arrival of the poll te completely full of uncon-from these sources as well as directly into the groundu ant cloud, with infilt would be submerged for a total period of about greatly ecduced. 30 days a dThe area n to the ground water during the entire period . exposed to infiltration The infiltration rate will be much greater prior to th e arrival of the 2.4. 5-8 I
ZPS l q l J j , AMENDMENT 12 l pollutant cloud than during its passage, and longer period af ter passage. ewhat som { It was estimated that less during the much I Therefore, the average radioactivity of waterrate w
' e pollutant cloud.
( reservoir would be .097 x 10-7 entering the ground water i further very greatly diluted oactive s by the pre-exi decay. This wouldtilimit be there would be a very large further reducti ng ground water. j ground waterthe tive following and by ion exchanges by the time thon by radioactive de flood. s f Association wells.during the Probable Maximum Flood cou
. Zimmer
{ g River Valley are farther from the WmSince the other public water I n the Ohio wells, they also could not be endangered abyspill. such.11. Ziumer site than b i l
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i i 2.4.5-9 i i L--_- i
ZPs AMENDMENT 12 2.4.5-3 (ZPS - April 9.1971. AEC Question 2.22) QUESTION Provide the minimum dilution factors between the site, Markland Dam and the city of Cincinnati water intake (Question 2.7c, Amendment 4), that you used to evaluate the effects of a river spill (Question 2.7d, Amendment 4). , ANSWER The answer to this question has been provided in the res ' sed answer to AEC Question 2.4.5-1 (AEC - October 13, 1970, Question 2.7), Page 2.4.5-5 .
'of the Wm. H. Zimmer Nuclear Power Station PSAR, Amendment 7. {
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1 i i 2.4.5-10
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ZPS AMENDMDIT 12 2.4.5.1-1 (ZPS - April 9.1971._ AEC Question 2.19) f QUESTION Ohio River water will be required to maintain safe emergency shut-down conditions at the plant, and postulated storage in the river cannot be justified without documentation in answer to Question 2.7b, Amendment 4. State the flow r . requirements for'such a condition, and evaluate the probable minimum instantan- .i A eous. flow at the site, based on recorded flows between Cincinnati and Maysville l (or other areas near the site), to determine the adequacy of the Ohio River as !L a water supply source. ANSWER ' 1
-l 1 }
a The magnitude of stream flows is very significantly influenced by re-leases from upstream reservoirs during the dry weather season. Storage in the Ohio River basin is operated for the purpose of augmenting the low flow of 3 j 1. tributaries with incidental increase in' the low flow of the main stem of the i Ohio River. The Corps of Engineers has computed seven day minimum flows for conditions that include for all completed river projects, projects under con- [ struction, and projects in advanced planning with the high degree of assurance of early construction. One-day minimum flows have not been computed because [ they can be significantly affected by gate operation of the navigation dans. t-3 The seven day low flows at the Wm ' H. Zimmer site would be slightly higher than those computed for Maysville, Kentucky which are as follows: 1,ow Flow Maysville. Kentucky Recurrence Interval , 7-day Averaste Flow (years) (cubic feet per second) 10 11,100- 1 20 10,400 50 9,600 100 9,100 A minimum daily flow of 2100 cubic feet per second was recorded at Louisville, Kentucky on August 12, 1930. No record was found of a lower flow at stations closer to the Wm. H. Zimmer site. This low flow occurred prior to , the construction of the upstream reservoirs and local navigation dams and there-fore would not represent present conditions because the large amount of storage ' . in the Ohio River basin greatly increases the minimum flow. However, a flow of f 2100 cubic feet per second greatly exceeds the essential cooling water require- i ment for Wm. H. Zimmer Nuclear Pow'er Station - Unit I which is 28.0 cubic feet per second. ] l, 2.4.5.1-1
1 ZPS AKENDMENT 12 2.4.5.3-1 (ZPS '- April 9.1971. AEC Ques tion 2.20)' QUESTION Your answer to Question 2.7.a.3, Amendment 4, indicated the water level that would result from the PMF. We will need the following additional informa-tion on this matter.
- a. Paragraph 2.4.5.6.12 of Amendment 2 indicates the sensitivity of the flood levels at the plant site was tested to different assuned initi'al conditions. State tne flood used for this test, the PNF or oti.er. Provide further analysis of the PMF backwater to demon-strate the conservatism in'the estimates of the water level at the plant site. For example, a lesser initial slope could be used i
based on extrapolation from other floods of record, or the compu-tations could be started at a point further downstream. I i
- b. Modify Figure 2.4-7 of Amendment 2 to show the computed PMF water surface profile.
_A,[SWER The starting elevation was derived from the average water slope between
; Cincinnati, Ohio, and Madison, Indiana, for the 1937 flood.
i { The s tage-discharge rating curves for Cincinnati, Markland Dam, and Madison, in the above downstream order, shov progressively less increase in flood stages for comparable increases in flood discharge, even if discharge ' is expressed as a percentage of average discharge. This indicates that for the corresponding two reaches, Cincinnati to Markland and Markland to Madison, the slope increases with an increase in flood discharge. The racing curves : for Madison and Louisville, farther downstream, are about parallel, indicating comparable flood slopes. The above observation is confirmed by the Corps of ragineers' Report,
" Ohio River Basin Comprehensive Survey." That report shows in Table 31 that l 3 the river fall between Cincinnati a.nd Louisville is 3.1 ft more for the Stan-
- dard Project Flood modified by storage tsc.n for the 1937 flood. The natural Standard Project Flood shows 3.6 f t more fall. This confirms the steeper slope for larger floods.
The above confirms that the starting slope used to determine backwater curves for the Probable Maximum Flood is conservatively low. Therefore, the elevation at Markland Dam for the start of the backwater curves is conserva-tively high. Figure 2.4-7 has been modified to show the computed PMF water surface profile.
- 2. 4. 5. 3- 1 l
ZPS AM EDMENT 12 2.4.5.3-2 (ZPS - April 9.1971. AEC Ques tion 2.21) QUESTION I Determine what the water level at the plant site would be considering
' the effects of wind wave action superimposed on the PMF. Evaluate the effects of water level on.the structural features of Class I and Il components; con-sider wave action (wave height, wave runup and pressures) that would result
{ from wind speeds of at least 45 mph for the most critical wind direction. MWG site. For wave determination the reach was 12 miles downstream from the plant g, No reduction in reach was made for the town of New Richmond, Ohio, seven iniles downstream, because most of the buildings would be submerged or carried away by the Probable Maximum Flood. Neither was any reduction in reach made for the Beckjord Power Plant about 10 miles downstrean because the tall structures occupy only a small part of the overbank width. A large reduction, however, was made in effective reach because of the narrowness of the river l I compared to the total reach. The river width at Probable Maximum Flood stage was measured at each mile in the 12-mile reach and the average found to be
' 4600 ft. Applying the width-to-fetch ratio of 0.0725 to Figure 1-13 of the Corps of Engineers' Report, " Shore Protection, Planning and Design," gave a f correctson factor of 0.2, which reduced the 12-mile reach to an effective I
reach of 2.4 miles. From Molitor's formula on Page 274 of Hinds, Creager and Justin's
" Engineering for Dams," a wave height of 3.02 f t from crest to trough and 2.01 f t of crest height above still water was determined. Wave runup
' against a vertical surface was found to be 4.02 f t above the still water level.
i I f(aximum wave pressure was found to be 453 pounds per square foot ac a {, height of 0.38 f t above still water level. The maximum total pressure was found to be 1140 pounds per foot of width of the structure with the resultant at a height of 1.13 f t above s till water level. The wave runup and pressures given are for the downstream side of ver-tical structures. Values would be somewhat less on the upstream side because of shorter fetch. They would be appreciably smaller on the river side and much smaller on the landward side. i 1 l 2.4.5.3-2 _ __ l-
_7____.
+
ZPS b A!!ENDMENT 12
- 2. 6-l' (ZPS - February 23, 1971, U
AEC Question 2. 15) ! Q_ESTION
' includes Expand the description of your environment al monitoring program to a.
The frequency, obtain. method, location and type of b. Provide a map showing the sampling networksanples yo The type of analysis planned for each type
- c. of sample taken, Arrangement program. with other government agencie s regarding your monitoring d.
Ohio River below the plant site, n that with section of theyourYo whether chain or of man. not such fish may be inevaluation the critical demonstrating path of the food ANSWER _ a. Wm. H. Zimmer Nuclear Power Station samples obtained for theenviT is indicated in Tables 1 and
- b. 2 and the attached Figure 1.ronmental m cated in the attached Table 1The type of analysis pla of sample taken is indi-c.
Arrangements with other government agenc posed monitoring plan after final coments conce to discuss the vironmental Report have been received and erning therevi Zimmer En-ments and/or changes reconenended will be Program ewed. fact improve-itoring' program and continued li i ored into the final mon-be maintained, pre-operational and post-operation with Public Ha
- d. ealth Officials will Industrial BIO-TEST Laboratories, Inc has made field collections at the sitel indicates the results of these investig .
a o River near the site and tithe following information' relative abundance of the inportant sport andons and emphasize area. cal food chain to man.These abundant fishes are the importentc species in the criti-2.6-1
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l ,
ZPS I AMENDMENT 12 i TABLE 2 ENVIRONMENTAL HONITORING PROGRAM Sampling Locations Categorized by Frecuency of Collection FREQUENCY WEEKLY
- 1. Airborne Particulate locations: Onsite Laurel, Ohio Moscow, Ohio Point Pleasant, Ohio Neville, Ohio California, Kentucky Felicity, Ohio Grants Lick, Kentucky New Richmond, Ohio Mentor, Kentucky Bethel, Ohio FREQUENCY BI-WEEKLY
- 1. Iodine-131 Locations: Onsite Point Pleasant, Ohio Moscow, Ohio Neville, Ohio California, Kentucky FREQUENCY M014HLY
- 1. Hilk Loca tions : Felicity, Ohio Point Pleasant, Ohio Bethel, Ohio California, Kentucky
- 2. Precipi ta tion
, locations: At all eleven (11) airborne particulate samplers
- 3. Well Water locations: Onsite 2.6-8
l l l ZPS AMENDMENT 12 TABLE 2 (Continued) f FREQllENCY QUARTERLY l I
- 1. Ambient Canrna e
Locations: TLD at all eleven (11) airborne particulate samplers
- 2. Well Water Locations: Beckjord Generating Station Tate-Monroe Water District Two (2) private wells - > 2 miles from plant
- 3. Cisterns Locations: Moscow, Ohio Point Pleasant, Ohio 1
Two (2) private cisterns - > 2 miles from plant b 4 Surface Water Locations: Meldahl Lock and Dam Beckjord Generating Station Directly below the site Little Indian Creek Cincinnati Water Works
- 5. Bottom Sediments Locations: Meldahl Lock and Dam Beckjord Generating Station Directly below the site Cincinnati Water Works Little Indian Creek 6 Bottom Organisms Locations: Meldahl Lock and Dam Beckjord Generating Station Directly below the site Cincinnati Water Works Little Indian Creek 7 Slime Locations: Meldahl Lock and Dam Beckjord Generating Station !
Directly below the site Cincinnati Water Works Little Indian Creek l 2.6-9 I
/ - ZPS I
! AMENDMENT 12
'l
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TABLE 2 (Continued) . l FREQt'ENCY SEMI-ANNUALLY
- 1. Fish Locations:. Four (4) > 1 mile from the site 2 Vegetation Locetions: Onsite Point Pleasant, Ohio Felicity, Ohio California, Kentucky Bethel, Ohio NOTE: Same locations as Milk production plus onsite
- 3. Soil Locations: Onsite Point Pleasant, Ohio Felicity, Ohio California, Kentucky Bethel, Ohio NOTE: Same locations as Milk production plus onsite FREQUENCY ANNUALLY
- 1. Ambient Camma locations: TLD at all eleve:u (11) airborne particulate samplers
- 2. Vegetation Locations: Seven (7) for garden veretables, tobacco and for fruit
> 10 miles from site
- 3. Soil Locations: Seven (7) same locations as the annual vegetation samples 4 Heat and Wildlife Locations: Four (4) > 10 miles from site 2.6-10 \
7,PS .
. AMENDMENT 12 Pre-operational.
at present, sampling of the fish population of the area involved Is,
- limited and ' inconclusive.
L shoreline with. a 1/2" mesh minnow seine, 'and 50 emerald shiners. Twenty-fiv
, Notropis ]
a therinoides , . were co. lec ted. J Because of a' lack of field collection of different fish species, this sec; ion of the report is based primarily upon data obtained in - a s tudy by. the Ohio Ri%er Valley Sani'.ation Commission and the University of; Louisville from 1957 to 1959 (ORSANCO,1962). : An investigation of the Ohio River fish population was initiated in 1966, by the Federal Water Quality Administration and is expec ted to continue through 1972. Cocplete data from this more recent studywcre unavailable for the preparation of the present report. However, a cursory examination of the FWQA data af ter the first year of study indicates a
. very close similarity to the results of the ORSANCO-University of I4uisville 1 atudy of 1957-1959.(Cherryholmes,1970).
Historical century indicate records related an abundance to fish of desirable food from fish.the Ohio River in the eithteenth i By the mid 1800's agricul- 1 tural practices and developing communities contributed to siltation.and municipal wastes which had local effects on the fish population. The construction of dams fromt-water swif 1855 todwelling 1929 eliminated species. rapids from the river and destroyed the habitat of The establishment of coal mining and the steel 'o - industry' in the last half of the 1500's resulted'in increased river pollution and
.further alteration of the fish population. . Pollution from checicel indus tries during the twentieth century har led to extensive fish kills and in some cases impairment of the flavor of food fisher. The result of these various types of pollution has caused a change in the fisheries from an abundance of sport and food fishes to an increase in the less desirable rough fish population.
The two most abundant fish species of recreational and cocanercial im-portance were Aplodinotus the channel, catfish, Ictaluras punctatus, and the freshwater drum, grunniens. cepedianium, Other comon species were the gizzard shad, Dorosoma and emerald forage for more valuable species. shiner, Notropis atherinoides, both of inportance as and crapples were not abundant in this reach of the river.Important game fish such a Table 3 shows the results of a creel census (1959) of Kentucky fishermen in the middle third of the Ohio River. side of the river and includes the arca :.djacent to the Zimmer siteThe . Table 4 census was
'ists the general fish species in the area of the Ohio River near the Zimmer site and. provides an indication of the more coamon species and their relative abundance in the general area of the proposed nuclear power plant.
ca tch, abouThet 727.. catfish and freshwater drum constituted the major portion of the Species usually considered game fish, e.g. , black bass, white ; I bass, crapp.ies, and sunfish accounted for approximately 207. of the catch. 2.6-11
2PS AMENDMENT 12 i
UhlE 3
! PERCEhTAGE COMPOSITION OF TE CATCE OF - _ON THE MIDDLE PORTIC5 0F TEE IN OEC 1959IIstE SPECIES
' COMMO*. NAMI APPROX. 7. OF CATCH Ictaturus g and Pylodictis o'ivaria Ca tfish 32 ge Jinotus grunniens Freshwater fruim 40 4.
_Cyprinus caipio Carp 8 Reccus chrysops White bass 6 l Lepocis n Sunfish
!. 14 Pomoris n ' Crappies 4
[ Micropterus a Blac~c bass 5 Misc. spp. 2 O 2.6-12 i
p t'b f j AMENDMENT 13 , I TABLE 4 i CENERAL SPECIES COMPOSITION Species Comon Names 7. by No. 7. by W t . Ictalurus punctatus Channel Cat 26.0 47.0 Dorosoma cepedianum Cizzard Shad 23.0 23.0 Aplodinotus grunniens Drum . 8.5 8.0
~
Notropis atherinoides Emerald Shiner 17.C 0.7 Cyprinus carpio Carp 0.5 8.0 Alosa chrysochloris Skipjack 1.3 1.5 Pylodictis olivaris Flat cat 0.9 4.7 Hybopsis storeriana Silver chub 5.0 - Lepomis megalotis Sunfish 1.7 - campostoma anomalum Stonerolle 1.6 - Ictiobus cyprinellus Large mouth buffalo - 1.8 Ictiobs buballis Small mouth buffalo - 1.2 Carpiodes forbesi Plains carpsucker - 0.7 All Others 15.0 4.1
- Note: Percentages by numbers and wei;; hts of fishes sacpled between Ohio River Miles 400 and 500 < bring 1957-1959 (CitSANCO,1962).
J
)
2.6-13 j l i
, ZPS k
AMENDMENT 12 ki [i During.a three-year crcel census on the middle third of the river,1,908 [ fishermen caught 3,191 fish. The rate of catch was 0.5.7 fish per hour of fishing
; time, and the fish averaged 11.0 in. long. Census results indicated that the l i sport fishery in the Kentucky portion of the Ohio River is a valuable natural re- i f source. Table 5 shows compositions of catches of nine coccercial fishemen in the '
Shio River between Mile 317 and Mile 535 during 1959. S $
\
L As was true of the sport fishery, the catfish and freshwater drum con-
;j f stituted the major portion of the coccercial catch. The three species of cat-
;. y fishes and the drum accounted for 66.57. by weight of the cocnercial catch in the j{ l middle portion of the river in 1959.
It was estimated that the total value of the catch of Kentucky comercial
'm y e fishermen amounted to $410,000 in 1956 and a total of 2,000,000 pounds.
j . Numerous reports claim that the flavor of food fish of the river has been l impaired by industrist and municipal pollution. Conversations with residents in w the region of the proposed power plant indicate that the flavor of fish in this
? area of ten is undesirable. Included below are data on individ .a1 species found ll in the area of the En. H. Zimmer Plant site that are of cocnercial, recreational l
and/or ecological importance (Calhoun,1966 and Carlander,1969):
- a. Channel catfish (Ictalurus punctatus): This is the dominant fish be-l1 ,. tween River Mile 400-500 in terms both of numbers and weights. It is l concercially and recreationally important and is one of the most i sought-af ter fish in the Ohio River. Channel catfish mature when they reach the 12-15-in. size. Spawning occurs when water temperatures
,, reach 70*F. Eggs are laid in darkened nests in holes, under rocks, and in other protected sites. The young feed primarily on insect
, larvae. As fish grow, they feed more on crayfish and forage fish.
1 l 3 b. Freshwater drum (Aplodinotus grunniens): The drum is abundant in the j 3 lower two-thirds of the river and is important as a commercial and
<r sport fish. It prefers quiet waters with a mud bottom. A bottom feeder, the drum prefers mussels, but also takes crustaceans and smaller fish.
b
} c. Gizzard shad (Dorosoma cepedianum): The gizzard shad is extremely 3
common in the River Mile 400-500 region, comprising as much as 237. by j both number and volume of the tors'. fish population. Although it has no food or sport value, it is an important forage fish for more in-
] portant species such as the catfish and black bass. Spawning in this j region begins in May. Heavy die-offs may follow spawning. Gizzard 4 shad are planktophai;ic (feeding on plankton), but occasionally they f 4 eat bottom fauna, such as midge larvae and oligochaetes.
o
. l l
i , 2.6-14 e j M ! I l d ,
ZPS AMENDMENT 12 TABLE 5 PERCENT COMPOST 10N OF NINE COMMERCIAL FISHERMEN IN THE OHIO RIVER BETWEEN MILE 317 AND MILE 535 (1959) Species Comon Name % by No. Wt. in Lb. % by Wt. Ictalurus punctatus Channel catfish 62.0 5,295 37.0 Ictalurus furcatus Blue catfish 8.0 1,730 12.1 Pylodictis olivaris Flathead catfish 4.9 1,431 10.0 Aplodinotus grunniens F.W. drum 7.7 1,057 7.4 Cyprinus carpic Carp 6.6 2,364 16.5 All Others 10.8 2,450 27.0 0 I
- The table lists the more commercially valuable species ned their relative importance in the comercial catch in the middle portion of the river.
I I 1 1 2.6-15 1
l ZPS I g AMEEMENT 12 r i' l*
- d. Entsrald shiner (Notropis atherinoides): This fish is abundast in the l' Ohio River and comprises 17. of the fish population of the River 400-l, 500 region. Its major icportance is as forage for more valuab'.e j
species. Spawning occurs in late May and early June. Major- ads are insects, both aquatic and terrestrial. The emerald shine: travels i t in large schools, numbering in the thousands, and feeds on tin float-
.1 :
ing plankton near the surface. L l e. Flathead catfish (Pylodictis olivaris): The flathead is an excellent i food fish that reaches large sizes and is valuabic to both the sport and commercial fisherman. Spawning begins in early June. Eggs are laid in nests that are guarded by the male. Young fish feed on in-sect larvae. As fish get larger they feed more on crayfish and smal-1er fish. Very large flatheads are solitary, remain in deeper pools and fee.d primarily on other fish, v
- f. Carp (Cyprinus carpio): Carp were not very abundant in tht. River Mile
' 400-500 region in 1959. Kwver, indications are that their future '
populations may increase. This fish is usually considered ur. desirable
)
because of its poor quality as a food fish; hence, its preser.ce tends to make a body of water less desirable for more valuable species. , j Carp move into shallow areas in the spring and begin spawning when the ' water temperature reaches 60*F. Reproductive potential is high; a y five pound female lays as nany as 500,000 eggs. They are bottom feeders and consume a wide variety of both plant and animal crganisms.
< k, This fish is tolerant of very low oxygen levels and therefore can with-stand extreme variations in temperature and r 11ution.
Il I
- g. Skipjack herring (Alosa chrysochloris): This species is not of sport i) !
or commercial value, and becomes too large to be a suitable forage
' fish. Being piscivorous (fecding on fish), it may compete with more valuable species for forage. It is potamodromus (migrates to spawn),
and begins spawning in early May. Young skipjacks feed on ir. sects ( ff and adults feed on other fish. l 1 i p,i h. Silver chub (Hybopsis storeriano): The value of this fish is as
" forage for more important species. Spawning starts at about 65'F and q
occurs in open water. Zooplankton is the major food of the pung; j adults feed on aquatic insect larvae. q j
} j The following species were r.ot abundant but are considered r,a.e fish i and are included because of their importance to the sport fisherman:
! f I i 1
' 1. Spotted bass (Micropterus punctulatus): This was the most abun- l dant of three black basses in the Ohio River. It is highly prized l as a sport fish, but does not attain large sizes. Spawning occurs '
( aj u when water temperature reaches 65'F. Males build nests in mud or I l l? - l l l i ~I ;t t
' /
! 2.6-16 -l I !
1d _ _ _
! ~ ZPS E, 2 AMENDME!C 12 f gravel bottoms. Fever eggs are produced than by other black buses. i 3 Yorng spotted bass feed on zooplankton, graduating to insects ed finally fish. -
- 2. Largemouth bass (Micropterus salmoides): Although not as numersus as spotted bass,' the largemocth attains a larger size. - Spawninl; occurs in the spring when water temperatures reach 60*F. Nes ting c
- substrate such as sand, . gravel or roots are required. Males build
' the . nest and guard the eggs. Fry feed largely on zooplankton.
Adults feed mainly on fish, but will' take a wide variety of living-organisms. ' i
- 3. Crpppies (Pomoxis gp,.): - Although 'neither species was very coamon, the' white crappie, A _ annularis, was more abundant,. but the bla:k crappie, h nigromaculatus, was larger.
black crappie when water temperatures reach 58 to 64 F and 64 tsSpawn i 68'F for' the white crappie. Black crappies nest on gravel or mm! bottoms, white crapples spawn near brush piles or stumps and prefer
- to deposit their. eggs on plant materials. The young feed on _ _
zooplank tor.. Larger fish eat crustaceans and insects, and fish ultimately become the most inportant food.
- ' 4. Sunfish (Lepomis sm): Most abundant game sunfish were the bluegill, E Macrochirus, longear, L2 Megalotis, and green sunfish, L 2
' cyane11us. These fish all spawn in the spring and build nests i:
bottom materials near the shoreline. The young feed on zooplank:on and insects. Adults feed on insects and worms, but the green sm-fish feeds on crayfish and, to some extent, on smaller fish. F i White bass (Roccus chrysops): ' The white ' bass exhibits good sporting qualities and is a good food fish. Beginning in April adults migrate up tributary streams and spawn in riffle areas. The young feed primarily on crustaceans and insects. Adults are mainly piscivoroms (feeding.on fish). I . Samples of channel catfish, Ictalurus punctatus, freshwater drum, Aplodinotus grunniens, or gizzard shad, Dorosuma cepedianium will be analyzed semi-annnally depending upon availability. The gizzard shad however, is the most abuMant species in thi: section of the river and it is anticipated that it will be th predominant species analyzed for radionuclides. The other species 4 will be shalyzed when available. A gamma scan will be performed on the flesh of all edibl{ species while a specific analysis for strontium-89 and -90 will be made on the bones of all samples. t
\ I 2.6-17 7
4 J __ _ > - - - k
15 ML t i ist KEY TC SAMPLING LOCATIONS: g '*'
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'l 4 WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSIS REPORT FIGURE 1 SAMPLING LOCATIONS FOR ENVIRONETAL (RADIOLOGICAL)
MONITORING PROGRAM 2.6-1[ s.
4 1 l BLANK PAGE
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ZPS AMENDMENT 12 INSTRUCTIONS FOR UPMTING YOUR PSAR VOLUME 1 SECTION 3.0 - REACTOR This section has been amended to incorporate answers to AEC questions. - All changes have been indicated by. a vertical line and the amendment ! number (12)- in the nght margin of the page.
- All pages (text , tables, - figures) with' changes have also been marked in 3 the upper right corner of the page with " AMENDMENT 12".
Figures ' that have been altered in any way are indicated by the amendment i number in the upper right corner of the. figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", ] figures that have changed in any way are designated by a vertical line with the , amendment number alongside the title of the figure. See example below: l f- FIGURE ND(BER TITLE 1 2.2-1 Station Site Area Topography 12 ] t To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, [ please use the following procedure:
- 1. In Volume 1, SECTION 3.0 - REACTOR, behind the red tabbed divider -
page titled Amendments to Section 3.0. l (
- a. In front of Page 3.3.4-1 insert Pages 3.3-1 through 3.3-5. I b, Behind Page 3.3.6-1 insert Pages 3. 7.4. 2. 3-1 through 3. 7. 4. 2. 3- 3.
f L 1 I s n ,d, 4
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( ZPS . l
!. AMENDMENT 12 I 3.3-1 (ZPS - April 9,1971, AEC Question 3.5)
QUESTION Taking into account the uncertainties in the calculations and the models, what are the nominal and upper limit values of the LOCA-blowdown loads exerted on I the control rod guide tubes during accidents initiated by a main steamline break and by a double-ended break of a recirculation line? Identify potential uncertain-l
- ties in the calculations and quantify the effect of each uncertainty on the loads experienced by the guide tube.
ANSWER The location of the recirculation line suction and discharge on the re-actor pressure vessel for a jet-pump plant causes the pressure differential across the control rod guide tube to be essentially the same for both normal operation and recirculation line break conditions. As indicated in PSAR Paragraph 3.3.5.3, rather than using nominal and upper limit values as mentioned in the concern, con-servative assumptions are used throughout, which maximized the assumed control rod guide tube loading during the postulated accident. The maximum differential pressure across a control rod guide tube vs. time after a steam line break acci-dent is presented in PSAR Figures 3.3-8 and 3.3-9. (Pressure differential is the same as the core plate differential pressure - Trace 2 of figures.) As shown in Paragraph 3.3.5.2.2 their maximum pressure is 24 psig. Although control rod guide tubes are noe pressure vessels per se, an ASME Code calculation on the control rod guide tube gives 93 psi as the allowable pres-sure differential during normal reactor operation. Under the code, the allowable pressure during an accident would be higher. The minimum factor of safety is greater than three since the maximum differential pressure from superimposing all j worse case loads is 30 psid or less. Lateral loadings on control rod guide tubes are obtained by assuming full recirculation pump flow from one side simultaneous with an earthquake exceeding the Design Basis Earthquake. It Fermi project is understood that concerns expressed recently on the Detroit Edison (Docket 50-341) regarding elastic instability coupled with or the result of column yet unresolved. loads superposed with the various blowdown loads above are as A program is under way to resolve these concerns. The Fermi resolution will be applicable to other BWR's such as the Wm. H. Zimmer Nuclear Power Station. 3.3-1 ! 1
1 ZPS AMENDMENT 12-3.3-2 (ZPS - April 9,1971. AEC Question 3.6) QUESTIONS For both the recirculation line break and the steamline break events that. result in the following largest hydraulic loadings on a control rod guide tube, present the information:
)
a. The pressure differential developed across the control rod guide tube wall versus time, b. the tubelateral versushydraulic time, loadings experienced by a control rod guide c. the column load on the control rod guide tube versus time,
- d. 1 the control rod displacement versus time, when time zero is taken as the instant of the recirculation line break, and e.
assuming the minimum wall thickness and maximum allowable ovality of the guide tube and using the principle of superposition, present an analysis of the resultant effect, taking into account the possi-bility of elastic instability on tube collapse due to the partial effects of the above hydraulically developed loads. Discuss the minimum factor of safety to control rod guide tube collapse. ANSWER a. The' differential pressure across the control rod guide tube wall is described in Paragraph 3.3.5.2.2 and is shown as trace #2 of Figures 3.3-8 and operating 3.3-9 power for main steam line breaks from two different reactor
,onditions.
As described in Paragraph 3.3.5.2.1, the differential prcasures developed during a recirculation line break are essentially the same as during normal operation, i.e. 20 psi. b. As explained in answer to AEC question 3.5 and in Paragraph 3.3.5.5, lateral loads on the control rod guide tube will be developed by assuming full recirculation pump flow from one side simultaneous with earthquake loads at least as large as design basis earthquake, c. As noted in the answer to AEC question 3.5, column loadings are part of a concern to be resolved on another docket. When these are re-solved, they will be applicable to other BWE's, d. As explained in PSAR Paragraph 14.6.3.3.1, the reactor shuts down almost instantaneously with a major line bre.ak due to formation of
- voids in the core. Coatrol rod displacement versus time is given
. in PSAR Paragraph 3.4.5.4 with tine zero taken as the opening of the 3.3-2
ZPS AMENDMENT 12 reactor protection system trip actuator. In order to approximate
, the control rod displacement with time, zero taken as the instant i of the recirculation line break, approximately .75 to .85 seconds i should be added to the information in Paragraph 3.4.5.4. This time delay is distributed as follows:
no more than .1 to .2 sec
.[ from instant of break until drywell pressure exceeds the scram set
[ point; nor more than .6 seconds from pressure exeeding set point-until pressure switch operates; approximately .05 see from pres-
- sure switch operation until reactor protection system trip actuator operation.
i< e. j As indica +ed in answer to question 3.5, the minimum tactor of I safety to control rod guide tubes is greater than 3. The 93 psi 3 differential pressure given in answer 3.5 is calculated using
=. superposition foi with maximum ovality and minimum wall thickness.
Details l this calculation are given in answer 5.4.2 of Amend- 1 i _ ment 12 for Enrico Ferni Unit 2 Docket #50-341. As indicated in g answer j is being to question on 3.5 the the concern docket.regarding elastic instability i
- i. I resolved Fermi applicable to other BWR's. The Fermi answer will be I :=
1 M am M i 2 u a
=;
1
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M c - s 3.3-3 ( ,
' i
ZPS AMENDMENT 12 i 3.3-3 !
. (ZPS - April 9,1971, AEC Question 3. 7)_ I QUESTION :
charging pressure and volume.The scram driving force of the control rods is the Discuss the driving force' of the drive mechanism over its full stroke and the resisting force developed considering 'the potential effects control rod.of the fuel assembly. channel bulging and the guide tube buckling onto ANSWER- 1 Prior to a scram r,ignal the accumulator in the Hydraulic Control Unit has i i approxima side. tely 1450-1510 psig on the water side.and 1050-1100 psig on the i As the inlet scram valve opens, the full water side pressure is available at; the control rod drive acting on a 4.1 square inch area. this lator pressure and the CRD. drops to the gas side pressure less line losses betweenaccumu- theAs CRD n with a resulting gas side pressure of approximately 575 psig.At ting pressure the accumulator only partially discharges with reactorure pressAt react providing tor pressure. the necessary scram force when reactor pressure exceeds scram accum when the reactor pressure is either low or at zero.The Control Rod D ! low, the sertion accumulator of the retains control rod in sufficient the required time.stored energy to insure the complet! i when the reactor is close to or at full operating pressure.The ac the reactor pressure alone will scram the control rod in the requirede.timIn this instan How-ever,inthe sure accumulator providing does provide an additional energy boost to the reactor p scram action. s
~
The greatest potential for fuel channel bulging occurs during a steam line break accident (approximately wherenormal) 2.5 psi above the maximum for a short differential pressure reaches 13 psi Fi 6eres 3.3-8 and 3.3-9 in the PSAR). time (a few seconds as shown on clasti: ally and then return to within a few mils of its original positionThe effect en the adjacent control rod would be insignificant consideringThe the follow-ing: a
- a. '
The control rod is 70 percent inserted before the differential pres-sure reaches its maximum. _
- b. 1 The additional friction force due to channel bulging is an extremely l small percentage of the total force available to insert a control rod.
3.3-4 . d J
- , 1 1
ZPS AMENDMDrr 12
' Die control rod guide tubes in all BWR's are analyzed and designed to l vithstand collapsing forces due to blow down. PSAR Paragraph 3.3.5 discusses the i;
planned analysis of the reactor internals including control rod guide tubes in
, considerable detail, t
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3.3-5 l 1 i i
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ZPS AMENDMENT 12 3.7.4.2.3-1 (ZPS - April 9.1971. AEC Question 14.15) QUESTION Describe all of the mechanical and control features that limit maximum and minimun flow in each' recirculation loop. Define the limits and interlocks associated with the bypass valve, control valve, and recirculation pump for each of the three -opera ting regions as defined in Paragraph 3.7.4.2.2. In region III, define the differences between the limits and interlocks that are active in the manual and automatic modes of operation. ANSWER The following limits are defined in reference to the attachad Figure 1:
- 1. Minimum power limit at high core flows.
6 To prevent cavitation in the recirculation pump, jet pumps, and flow control valves, the system is provided with an interlock to shut the main flow control valves and the discharge block valves if the dif-ference between steam line temperature and recirculation pump inlet temperature is less than a preset value (typically 4.5'F) . This dif-ferential temperature is measured using high accuracy (RTD's) with , sensing error of less than .2*F at the two standard deviation (2a) { confidence level. The valve closure action is initiated electro.:ically { through a 15 second time delay. The interlock is active while in both the automatic and manual operation modes.
- 2. Minimut.1 power limit at low core flow.
During low power, low loop flow operations, the temperature differ-ential interlock may not provide sufficient cavitation protection to the flow control valves. Therefore, the system is provider' t.th an interlock to shut the main flow control valves and dischart lock valves if the feed water flow falls below a pre-set level (t,pically 307. of rated) and the flow control valves are below a pre-set posi- ! tion (typically 357. open). The feed water flow and recirculation j flow control valve position are measured by existing process control l instruments. The valve closure action is electronically initiated. ] This interlock is active during both automatic and manual modes of l , operation. I
'N
- 3. Pump Bearing Limit For pumps as large as the recirculation pumps practical limits of pump bearing design require that minimum pump flow be limited to 207. of rated. To assure this minimum + flow, the system is prmided 3.7.4.2.3-1
l ZPS l AMENDMENT 12 with an interlock which will prevent starting the pump if the bypass valve is typically at less than 257. position. This is a permissive interlock circuit activated by the closure of a position limit switch before starting plant operation. s
- 4. Valve Position 9
l While the bypass valve is positionable, it should procedurally remain 3 in full open position during plant operation. However, to protect j the pump, the pump is tripped if the bypass valve position is {' typically at less than 257. and simultaneously the discharge block valve is at less than 907. position. The pump is also tripped if the
. bypass valve is typically at less than 57. position and the flow con-trol valve is at less than 107 position. These interlocks are active in both the automatic and manual operation modes. Activation of these ; interlocks is by position limit switches acting through a logic if/and ,
discrimination circuit. To prevent structural or cavitation damage ' to the recirculation pump due to pump suction flow starvation, the
! system is provided with an interlock to prevent starting the pumps, ! or to trip the purps if the suction block valves are at less than 907.
I open position. This circuit is activated by a position limit switch l and is active before the plant is started, during manual operation mode and during automatic operation mode. g i t i 3.7.4.2.3-2 ; i _ _ _ _ _ _ _ _ ____ I
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ZPS ( l AMENDMENT 12 k ' t INSTRUCTIONS FOR' UPDATING YOUR PSAR' VOLUME 2 All changes have been indicated by a vertical line and the' Amendment l Number (12) in- the right margin of the page.
- 1. At the beginning of Volume 2 remove and destroy.Pages 11.15,17,
18,19 and 20 and replace with amended Pages 11, 15, 17, 18, 19 and
- 20. Af ter Page 20 insert new Pages 21 and 22.
I j ., i i y 7 E i , 1 e Y s i
BLANK PAGE 4 d
\
1E i
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2PS MiENDMENT I? VOLUME 5
)
_ TABLE OF CONTEICS, (Continued) PAGE D.O APPENDIX D.0 - QUALITY CONTROL SYSTEM TABLE OF CONTENTS D.O-i D.1 INTRODUCTION D.1-1 D.2 CINCINNATI GAS [ ELECTRIC CO. QUALITY ASSURANCE PROGRAM D.2-1 D.3. SARGENT'& LUNDY QUALITY ASSURANCE SYSTD! D.3-1 D.4 GENERAL ELECTRIC QUALITY SYSTEM FOR BWR NUCLEAR STEAM SUPPLY PROJECTS D.4-1 D.5 KAISER ENGINEERS INC. (CONSTRUCTORS) QCALITY ASSURANCE - QUALITY CONTROL PROGRAM D.5-1 D.6 APPLICABILITY OF- QUALITY ASSURANCE PROGLVi TO COMPONENTS, 3 SYSTEMS AND STRUCTURES D.6-1 1 E.0 APPENDIX E.0 - STATION ATMOSPHERIC RELEASE LIMIT CALCULATIONS E.1-1 F.0- APPENDIX F.0 - CONFORMANCE TO AEC DESIGE CRITERIA ; TABLE OF CONTENTS F.0-1 l F.1
SUMMARY
DESCRIPTION F.1-1 F.2 CRITERION C JNFORMP OE (Ti!E "70" CRITERIA) F.2-1 12 ' F.3 CRITERION CONFORMANCE (THE "64" CRITERIA) F.3-1 G.0 APPENDIX G.0 - STATION NUCLEAR SAFETY 0?ERATIONAL ANAI.YSIS TABLE OF CONTEhTS G.O-i G.1 ANALYTICAL OBJICTIVE G.1-1 G.2 APPROACH TO OPERATIONAL NUCLEAR SAFETY C.2-1 G.3 METHOD OF ANALYSIS G.3-1 G.4 DISPLAY OF OPERATIONAL ANALYSIS RESULTS G.4-1 Il ! I
ZPS
) AMENDMENT 12
\
LIST OF ZPS, OCTOBER 13. 1970 AEC QUESTIONS, (Continued) AEC QUESTION RENUMBERED VCLLHE NUMBER' AS OUESTION PACE OTPSAR 5.5 5.2 . 5-1 5.2.5-1 2 5.6 5.0-1 5.0-1 2 5.7 5 . 2 . 5.1-1 5.2.5.1-1 2 5.8 5. 2. 3.1-1 5.2.3.1-1 2 ; 5.9 12.3.2.5-1 12.3.2.5-1 1
,4 ;
5.10 5.2. 3.8 -1 5.2.3.8-1 2 b2 12.1 12.2.1.1-1 12.2.1.1-1 4 12.2. 12.3.1-1 12.3.1 - 1 4 12.3 12.2.1.1-2 12.2.1. 1-2 4 12.4 12.2.2-1 12 .2.2 - 1 4 12.5 12.2.2.5-1 12.2.2.5 -1 , 4 12.6 12.4.4.1-1 12.4.4 . 1-1 4 7 12.7 12.4.4-1 12.4.4 -1 4 12.8 12.3.2.3-1 12.3.2 . 3 - 1 4 12.9 12.3.2-1 12.3.2 -1 4 12.10 12.3.2.2-1 12.3.2.2-1 4 12.11 12.3.2.2-2 12 .3.2 . 2 -2 4 12.12 12.4.3.3-1 12 .4. 3. 3 -1 4 12.13 12.3. 3-1 12.3.3-1 4 2 12.14 12.2.2.4-1 12.2.2 . 4 - 1 4 12.15 12.3.6-1 12.3.6-1 4 12.16 12.2.2-3 12.2.2- 3 4 12.17 12.2.2-2 12 .2.2 -2 4 12.18 12.5.1-1 12 .5.1 - 1 4
, 12.19 12.3.4.2-1 12 .3.4 . 2 -1 4 12.20 12.3.2.3-2 12.3.2.3-2 4 12.21 12.2.1.1-3 12.2.1. 1- 3 -
15 9
l l , I ZPS l AMENDMEbT 12 LIST OF ZPS, FEBRUARY 23,__1971 AEC QUESTimS i AEC QUESTION RENUMBERED VOLLME ' l NUMBER AS QUESTION PAGE ' OF PSAR ; l 2.12 2.2.3-2 2.2.3-12 .1 I 2.13 2.3.2.1-2 2.3.2.1-2 1 2.14 2.3.2.1-3 2.3.2.1-3 1 2.15 2.6-1 2.6-1 1 l'1'
- j. 2.16 2.3.8-1 2.3.8-1 1 4.9 4.7-2 4.7-2 2 9 4.10 4.7-1 4.7-1 2 i 4.11 4.9-1 4.9-1 2 9i j 4.12 4.0-1 4.0-1 2 5.11 5.2.3.7-1 5.2.3.7-1 2 7 i 5.12 10.19-1 10.19-1 2 5.13 5.2.3.8-2 5.2.3.8-2 2 j
,! 5.14 5.3.3.3.3-3 5.3.3.3.3-3 2 5.15 5.3.3.3.3-4 5.3.3.3.3-5 2 12 5.16 5.3.3.3.3-5 5. 3. 3. 3. 3- 7 2 5.17 5.2.3-1 5.2.3-1 2 ;
4 7.1 5.3.3.3.3-1 5. 3. 3. 3. 3 - 1 2
; . 7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 .5.3.3.3.3-2 5. 3. 3. 3. 3 -2 2 1 7.4 7.1-1 7.1-1 3 11 1
7.5 7.2.3.1-1 7. 2 . 3.1- 1 3 l 12) 7.6 4.4-1 4.4-1 2 )
.; 7.7 7. 2. 3. 3-1 7.2.3.3-1 3 l 12 7.8 7. 2. 3.6 -1 7.2.3.6-1 3 111 7.9 7.2-1 7.2-1 3 i l
l 7.10 7.2.3.9-1 7.2.3.9-1 3 7.11 7. 2. 3. 9-2 7.2.3.9-2 3 l 12 e 17 l9 i L _m__ ___m.__..._. - - - 1
ZPS AMENDMENT 12 LIST OF 2PS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued) AEC QCESTION RENUMBERED EDGER VOLUME AS QUESTION PAGE OF PSAR 7.12 7.2-2 7.2-2 7.13 3 1
-1.0-1 1.0-1 7.14- 1 12 '/ .12. 5. 3-1 7.12.5.3-1 1, 7.15 3
, . 4.3 -1 7.4.3-1 3 7.16 o
l.5.7.3.3-1 7.5.7.3.3-1 7.17 3
".8.5-1 7.8.5-1 7.18 3 11 7.5.8-1 7.5.8-1 9 7.19 3
- 7. 6.3 -1 7.6.3-1 7.20 3 7.8.5.2-1 7.8.5.2-1 7.21 3 i 7.9-1 7.9-1 7.22 3
- 7. 3. 4. 8-1 l 7 3.4.8-1 3 '12 7.13 7.10-1 7.10-1 7.24 3 D.6-2 D.6-14
, 7.15 5
D.0-1 l D.0-1 5' 7.26 10.10.3-1 l 11 10.10.3-1 4 7 7.27 7.2-3 7.2-5 7.26 3 7.0-1 l 11 7.0-1 3 7.29 10.19-2 l 12 10.19-2 4 11 7.30 7.0-2 7.0-4 7.31 3 11 7.7-1 7.7-1 3 12 8.1 12 8. 3.2.1 -1 8.3.2.1-1 8.2 4 8.3.2-1 8.3.2-1 12 8.3 4 C.3.3-1 8.3.3-1 8.4 4 8.4.3-1 8.4.1-1 11 4 8.5
- 8.5.4-1 B.5.4-1 8.6 4 8.4.3-2 8.4.3-2 8.7 4
!2 8.5.3.1-1 8.S.3.1-1
, 8.8 4 8.0-1 8.0-1
. 4 18 9
ZPS i AMENDMEN1 12 LIST OF ZPS, FEBRUARY. 23, 1971 l AEC 03STIONS, (Continued).
' AEC QUESTION RENUMBERED VOLtMF-NUMBER AS QUESTIGE PACE OF PSAR j 8.9 8.0-2 t
8.0-2 4 ) 8.10 8.9-1 8.9-1 4 8.11 8.10-I 8.10-1 4 j 9.1 9.2.4-1 9.2.4-1 4 t 9.2 9.2.4.6-1 9.2.4.6-1 4 9.3 9.4-1 9.4 1 4 9.4 9.4. 6-1 9.4.6-1 4 l11: 9.5 9.2.4.7-1 9. 2.4. 7-1 4 9.6 9.4.3-1 9.4.3-1 4 I. 10.1 10.0-2 10.0-2 4
, l12 10.2- 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 4 l 11
; 10.4 10.11.2-1 10.11.2-1 4 7 f 12.22 12.6.1-1 12.6.1-1 4
; _ 12.23 12.5.6-1 12.5.6-1 4 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6-1 4 13.3 13. 2.1.2-1 13.2.1.2-1 4 9
, 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4 1 4 13.7 13.0-3 13.0-3 4 gg 14.12 14.9.1-1 14.9.1-1 4 14.13 14.9.1-2 14.9.1-3 4 12 4
19 9
1 ZPS
' .1 AMENDMENT 12 1 LIST OF ZPS, FEBRUARY 23, 1971 AEC Qt'ESTg',, (Continued)
AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PACE OF P AR 15.16 A.2-1 A.2-1 5' 7 15.17 A.2-2 A.2-5 5 15.18 A.2 *, A. 2- 6 5 I 15.19 B.1-1 B.1-1 ' 5 12 { 4 1 :- 20 l9
.i n _. . _ _ _ _ _ _ _ _ _ _ _ - - .
}
l IPS ( NfENDMENT 12 { i
. i LIST OF ZPS, APRIL 9,19)1 AEC CUESTIONS '
i' AEC QUESTION RENUMBERED VOLUME - NJMBER AS QUESTIQ1 PAGE 01 P3AR 1.1 1.10-1 1.10-1 1 { 1.2 H.3-1 H.0-1 5
, 2.17 2.3-1 2.3-1 1 2.18 2.4.5-2 2.4.5-7 1
! 2.19 1
2.4.5.1-1 2. 4 . 5 .1-1 1 2.20 2.4.5.3-1 2. 4.5 . 3-1 1 y 2.21 2. 4. 5 .' 3-2 2. 4. 5. 3-2 1
, 2.22 2.4.5-3 2.4.5-10 l 1 3.5 3.3-1 3.3-1 1 g 3.6 3.3-2 3.3-2 1
! 3.7 3.3-3 3.3-4 1
- j. 4.13 4.3-1 12 I
4.3-1 2 4.14 4.4-2 4.4-2 2 5.0 10.20-1 10.2 0 -1
' 4
, 6.1 6.5-1 6.5-1 2 6.2 6.5-2 6.5-5 2 6.3 (
'.5-3 6.5-7 2 I
, 6.4 6.5-4 6.5-13 l 2
6.5 6.5-5 6.5-16 2 6.6 6.5-6 6.5-21 2 6.7 6.5-7 6.5-22 2 6.8 6.5-8 6.5-32 2
, 6.9 '6.5-9 6.5-38 2 6.10 6.5-13 6.5-39 2 6.11 6.5-11 6.5-41 2 6.12 6.5-12 6.5-43 2 6.13 6.4-1 6.4-1 2
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ZPS AMENDMENT 12 i LIST OF ZPS. APRIL 9.1971 AEC QUESTIONS, (Continued) AEJ QUESTION RENUMBERED NUMBEk VOI.UME AS QUESTION PAGE OF PSAR 7.32 7.9-2 7.9-6 3 7.33 7.9-3 7. 9 -6 3 12.24 12.2.2.5-2 12.2.2.5-2 4 1 i 12.25 12.2.1.1-4 12.2.1.1-4 4 12.26 12.3.8-1 12.3.8-1 4 14.14 4.3-2 4.3-2 2 14.15 3.7.4.2.3-1 3. 7. 4. 2 . 3-1 2 14.16 7.9-4 { 7.9-9 3 14.17 7.9-5 12 7.9-11 3 14.18 7.9-6 7.9-13 3 14.19 7.9-7 7.9-14 3 14.20 14.5-1 14.5-1 4 14.21 14.5.5-1 ; 14.5.5-1 4 14.22 14.5.5-2 14.5.5-3 4 14.23 14.5.6-1 14.5.6-1 4 14.24 5.3.4.3-1 5.3.4.3-1 2 14.25 5.3.4.4-1 5.3.4.4-1 2 14.26 14.9.2.3-1 14.9.2.3-1 4 15.19 I.0-1 4 1.0-1 5 15.20 C.3.1-4 C.3.1-4 5 4 4
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s ZPS AMENDMENT 12 INSTRlfCfl0NS FOR l'PDATING YOUR PSAR VOLUME 2 SECTION 4.0 - REACTOR COOLANT SYSTEM This section has been amended to reflect engineering changes in the. lean-
] detection system and answers to the April 9,1971 AEC questions.
l 6
, 1 All changes have been indicated 1,y a vcrtical line and the amendment number (10) in the right margin of the page.
All pages (text, tabics, figures) with changes have also beer marked in the upper right corner of the page with " AMENDMENT 10". T, i number Figares that have in the upper rightbeen altered corner of thein any way are indicated by the amendment figure; note that there are no other marks i that would indicate changes in figure. On the page marked " LIST OF FIGURES", a figures that have changed in any way are designated by a vertical line with the amendu:at number alongside the title of the figure. See example below: FIGURE NUMBER TITLE
, 2}
2.2-1 Station Sf *.e i rca Topography i l12 To update your copy of the Wn.
-q pleace use the following procedure: H. Zimmer Nuclear Power Station PSAR, 1
- ! . In "olume 2 SECT 10N 4.0 - REACTOR COOLANT SYSTEM, remove and destroy
; Table of Contents Contents Psge 4.0-vil. Page 4.0-vii, and replace it with amended Table of a
- 2. In Volume 2 SECTION 4.0 - REACTOR COOLANT SYSTEM, remove and destroy j Text pages through 4.10-6. 4.10-1 through 4.10-6 and replace with amended Pages 4.10-1
- 3. In Volume 2 SECTION 4.0 - REACTOR COOLANT SYSTEM, remnve and destroy the following figures, and replace them with appropriate amended figures.
REMOVE FIGURE
~,: ^
-REPLACE WITH AMENDED FIGURE j 4.10-1 through 4.10-4 4.10-1 through 4.10-4 i 4. In Volume 2, SECTION 4.0 - REACTOR COOLANT SYSTEM, behind the red
. 1 tabbed divider page titled " Amendments to Secticn 4.0".
j
- a. Behind page 4.2.6-5 insert pages 4.3-1 and 4.3-2.
i
- b. Behind page 4.4-1 insert pages 4.4-2.
N -
i 1 h BLANK PAGE D I
\
X
l l I . -
-g 2PS AMENDMENT 12 ,
LIST OF FIGURES, (Continued) FIGURE NUMBER TITLE 4.10-1 Typical Temperature Monitorind Leak Detection System 4.10-2,
. Leak Detection Differential Temperature Indication Schematic
)
4.10-3 Leak Detection Absolute Temperature Indication Schematic 12 ' 4.10-4 Leak Detection Reactor Water Cleanup Differential Flow I i I f l 1 j 1
.\
I 4.0-vii I
. I ~
ZPS AMENDMENT 12 . 4.10 NUCLEAR SYSTEM LEAKAGE DETECTION AND LEAKAGE RATE LIMITS 4.10.1 Safety objective Reliable means shall be provided to detect and isolate leakage from the nuclear system process barrier before predetermined limits are exceeded. 4.10.2 Safety Design Basis l j
- 1. A means shall be provided to detect abnoraal leakage where necessary before the results of this leakage become unacceptable. l12 j
- 2. A means shall be provided to isolate abnormal leakage before )
j the results of this leakage become unacceptable.
]
i
- 3. Limits shall be established on abnormal leakage so that '
corrective action can be taken before unacceptable results occur. The unacceptable results are as follows: (a) A threat of significant compromise to the nuclear system process barrier (b) A leakage rate in excess of the coolant make-up capa- ) bility to the reactor vessel (c) Flooding of equipment required for safe operation or shutdown of the plant Definitions : Norr.al Desistn Leakane - Controlled quantity of fluid released from seals or ' sealing syste:as of piping components which are properly assembled and in good j condition. Abnormal Leakage - Fluid released from a small crack or damaged seal in the nuclear system process barrier which has a low probability of rapid growth and does not exceed the guideline limits of Federal regt.lations with respect to accidents. G_r_oss Leakage - Uncontrolled fluid released from a ruptured piping com-ponent at such a rate that the guideline limits of Federal regulations with re-spect to accidents could be violated if isolation is not affected. 4.10.3 Description This subsection describes the leakage detection systems which are provid-
\ '
ed to detect abnormal leakage from the nuclear system process t arrier both inside 1
=
4.10- 1
l i
"PS '
AMENDMENT 12
.i
' and 'outside the primary containment. Also discussed in this subsection are nu-
[ clear system leakage' rate limits and hows they are established. e the systems which detect gross leakage roulting from a pipe rupture and initiate automatic' isolation are considered as part of the Reactor Vessel and Primary Containment Isolation Control System and are discussed in Subsection 7.3
" Reactor Yessel and Primary Containment Isolation Coetrol System". The controls available for unanually initiating isolation are ~also discussed in Subsection 7.3.
In some cases a leakage detection system which provides an automatic ~ isolation
- J. signal also provides an indication or alarm signifying abnormal leakage. In such
; cases the indication or alarm function provided is discussed in this subsection.
'i l 4.10.3.1 Normal Design Leakane
.( !
1he pump packing glands, valve stems and other seals in systems which are part of the nuclear systen process barrier and from which normal design leakage is expected are provided with drains or. auxiliary sealing systems. The valves and pumps (4 in, and larger) in the nuclear system insife the drywell are i equipped with double seals. Leakage from the primarr recirculation pump seals is l 12 piped to tt: equipment drain sump as described in Sdsectien 4.3, " Reactor Recircu- [' . lating System". Leakage from the main steam relief a-d safety valves is identified by; t.amperature senscrs which transmit to the main ce::rol room. Any temperature ; increase detected by.these sensors above the drywell ambient temperature indi-
]- cates valve leakage. Isakage from the reactor vessel head flange gasket is piped to a collection chamber and then to the equipmect drain samp. The collec-
]
tion chamber filling time is periodically timed'during plant operation and the
! flange gasket leakage rate is calculated. j l
i A more detailed discussion is presented in Sensection 7.8, " Reactor i Vessel Instrumentation". Thus, the leakage rates fra pumps, valve seals and i the reactor vessel head seal are measurable during operation of the plant. These leakage rates, plus any other leakage rates measured shile the dryvell is open ] l are defined as identified leakage rates. 4.10.3.2 _ Unidentified Leakage Rate
- The unidentified leakage rate is that portice of the total leakage rate received in the drywell sumps which is not identified as described above. A ;
threat of significant compromise to the nuclear Fystem process barrier exists if i the barrier contains a crack which is large enough te propagate rapidly. The unidentified leakage rate limit must be low because of the possibility that most
- i. of the unidentified leakage rate might be emitted free a single crack in the nu-clear system process barrier.
- A leakage rate of 150 gal / min has been conservatively calculated to be the miniwun liquid leakage from a crack large enough t: propagate rapidly. An allowance for reasonable leakage which does not compranise barrier integrity and b i-
.I i 4.10-2
ZPS AMENDMENT 12 is not identifiable is made for normal ~ plant operation. The unidentified leakage rate limit is established at 5 gal / min which l12 is far enough below the 150 gal / min leakage rate to allow time for corrective action to be taken before the process barrier could be. eignif f cantly compromised. 4.10.3.3 Total Leakase Rate which flows to the drywell rioor drain and equipment The criter- drain sum ion for establishing the total leakage rate limit is based on the make-up capa-bility of the control rod drive (CRD) and the RCIC systems, and independent of the feedwater system, normal ac power, and the core stand-by cooling systems. The CRD system supplies 63 gal / min into the bottom of the reactor vessel; the RCIC system can supply 400 gal / min through the feedwater sparger to the reactor vessel. The total leakage rate limit is conservatively established at 30 gal / l4 , min. l12 The total of the drywell leakage rate limit is also set low enough to prevent overflow sumps. q , drain sump (capacity 500 gal),The e' uipment drain susp (capacity 500 gal) and the ! one 50-gal / min pump. The total which collect leakage rate all leskage, are each drained by limit capacity of the pumps for each sump. is set below the removal 7,3
~
4.10.3.4 Leakage Detection Systems The systems or parts of systems which contain water or steam coming from the reactor vessel or supply water to the reactor vessel and which are in direct i Table 4.10-1 shows the systems in communication with the leakage detection systems which monitor these systems, and the locations or areas monitored. of leakage detection system and its instrumentation.The following paragr type of detection system is used in several locations. In most cases the same 4.10.3.4.1 Detection of leakage Inside Drywell detection systems are necessarily comon.Since the systems within the dry Each of the leakage detection systems inside tablished theleakage drywell rate is desiped limits. with a capability to detect leakage less than es- ) 4.10-3
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l I ZPS AMENDMENT 12 Drywell Pressure Measurement The primary containment is pressurized and maintained at a slightly posi-tive pressure durin; reactor operation. The pressure fluctuates very slightly as a result of barometric pressure changes and outicakage. A pressure rise above the normally indicated values may indicate the presence of a leak in a system l 12 within the drywell. Normal operating pressure will be 0.75 psig + 0.1 psig. Drywell Temperature Measurement The primary containment cooling system recirculates the primary contain-ment atmosphere through heat avehangers (air coolers) to maintain the primary con-tainment at its design operating temperature of 135'F. The drywell ventilation l 12 system chilled water system provides cooling water to the air coolers. An in-crease in the primary containment atmosphere temperature would increase the heat load on the air ecclers and thus result in an increased temperature rise in the cooling water passing _ through the coils of the air coolers. Thus the drywell ventilation system chilled water temperature difference l 12 increase between inlet and outlet to ths air coolers will indicate the presence of a reactor coolast or steam leakage. Also, drywell ambient temperature rise above 135'F will indicate the presence of reactor coolant or steam leakage, pro-viding the drywell coolers are in operation. Drywell Floor Drain Sump Flow Hensurement 12 The drywell floor drain sump has a displacement type, continuous level, monitoring system that will alarm in the main control room when the unidentified ! leakage-rate limit is reached, 'f4 The normal design leakage collected in the floor drain sump consists of leakage from the control rod drives, valve flange leakage, drywell air coolers condensate and chilled water leakage is identified during preoperational tests, additionally, the leakage from the chilled water system is identified during plan : uperation by changes in the chilled water system surge tank level. Any increase above these identified values in detected by the floor drain sumps. 12 brywell Eosipment Drain Sump The drywell equipment drain sump has a displacement type, continuous-level monitoring system that will alarm in the main control room when the un-identified leakage-rate limit is reached. ! a L 4.10-5
- _ - _ _ _ _ . l 1
l ZPS AMENDMENT 12 l12 l l The normal leakage collected in the equipment drain sump consists of 1eakage from the reactor vessel . head flange, the recirculation pump seals, the recirculation system valves, the RCIC system valves, the cleanup system valves, l driveshutdown the system valves. system valves, the main steam isolation valves and the control rod i This leakage is identifled during operational tests and !2 any sumpincrease level. above these values is detected by monitoring the equipment drain Drywell Air Sampling I i
! The drywell air sampling system is used to supplement pressure, and flow variation method described above to detect leaks in thethe temperature, nuclear system process barrier.
g !' with air sampling poinha one of which is in service continuously.The drywel i Air is dawa outside the drywell to a continuous air sampling monitor which counts and records i gross particulate, noble gases and iodine levels. activity. The remaining sample points provide a means of taking manual airAn alar samples leak. from specific areas in order to determine the approximate location of the 1. 2 This system is described more fully in subsection 10.19. 8 4.10.3.4.2 Detection of Abnormal Leakage Outside Primary Containment { Outside the primary containment the piping within each system monitored
- for leakage is in compartments or rooms separate from other systems. wherever feasible so that leakage may be detected in drains or by area temperature indi-cations.
designed to detect leakage rates less than the established leaka
. Room Ventilation or Standbv Cooler Temperature A differential temperature sensing system is installed in each room con-talning equipment which is pt.rt of nuclear system process barrier. This in-ciudes the RCIC, RIE and reactor water cleanup systems equipment rooms, and main steamline tunnel. Table 4.10-1 shot.
the parts of systems which compose the 12 nuclear system process barrier and the room or areas 1,n which dif ferential tem-perature detection systems are installed to monitor various parts of systems. Temperature provide normal sensors are placed in the inlet and outlet ventilation ducts which ventilation. Additionally, temperature sensers are installed in
' are provided (see Table 4.10-1). coolers in the rooms where standby coolers the inlet and outlet of the standby A differential temperature switch between each set of sensors initiates and alarm in the control room when the temperature difference equal to the reaches leakage a point rate which limit. indicates a leakage within the monitored room, in Figures 4.10-1 and 4.10-2. The instrument arrangement is illustrated The alarm point is determined analy-tically by calculating the increase in dif ferential temperature would which ree". I t if a leak equal to the abnormal Icakage rate
_ _ _ _ _ _.- - ~ - - - - - - - - - - - ' _ 4 ql0-6
i AMENDMLNT 12 l l l l Tc ALARM ISOLATION TS-A A __ _ ..___ _ _ 3 y SIGNAL j VENT AIR VENT AIR INLET OUTLET i, i i dTS
!12.
___,_ _,___~ T, l t _ _ _L _ _ _ ALARM' y ISOLATION y
~~
SIGNAL
' ' STANDBY COOLER CHILLED ~~
WATERINLET - - - STANDBY COOLER CHILLED WATER OUTLET 3 T A l A a's l __ _2_ _ .t_ _ _ _ _ A g l ALARM 'ISOL ATION L _ _ ._ _) y SIGNAL 2 1 WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSTS REPORT FIGURE 4.10-1 T(PICAL TEMPERATURE MONITORING LEAK DETECTION SYSTEM
ia AMENDMENT 12. 1 i
-j AREA DIFFERENTIAL TEMPERATURE
-TY PICAL- l
{ VENTILATION VENTILATION INLET OUTLET j TE TE <
' I REACTOR WATER
. _ _ _ . _ _ _. CLEANUP EQUIPMENT.
ROOMS RWC dTR HPCS A 7 -K lHPCS ROOM VENT DIFF'. TEMP. RHR
,, ,, .: K 1RHR ROOMS VENT DIFF'. TEMPS.
- =( lRCIC R60M VENT DIFF'. TEMP.
STEAM sellii A TUNNEL i":( l STEAM TUNNEL VENT DIFF. TEMP. DWWELL COOLEE yH l1 .( lDRYWELL AIRTEMP, COOLER CHILLED WATER DlFF. CA BER :=( I
"^ ^ ^
AREA D FF TEM l. WM. H. ZIMMER NUCLEAR POWEA STATION PRELIMINARY SAFETY ANALY5t's REPORT 1 l FIGUT.E 4.10-2
. ]
mjEAgj{ECII0f DIFFERENTIAL _
, MGW t
I t I $ i ( !
~ --
AMB ENT l TIR I ! l i .TE l F l A l
! TYPfCAL l lIliI l RHR llll l 1 g -l RHR ROOMS RCIC .._{ i l l ll _
.( l RCIC ROOMS I
SUPPR. CHAMBER l l l L- { AREA { l SUPPRESSION POOL ARE ! l L l Il ' - - - N l STEAM TUNNEL ,
~
BENT
~
I l ^ bI " VALVE LEAKOFF AH - - {. l VALVE LEAKOFF i WM. H. ZIMMER Ntr.IIAR POWER STATIO! PRELIMINARY SAFET) ANALYST $ REPORT FIGURE 4.10-3 __ LEAK DETECTION ABSOLUTE
AMENDHENT 12 ,
' \
l
. TO REACTOR
( FEEDnATER ( i : n FT '-- , FI --- '
. h A
i i FZ _ dFS [ISOL ATION PRIMARY CONTAINMENT b SIGN AL l HO h-- ll. ( E H0 g I E R EG EN E R ATIVE FROM HE AT EXCHANGERS
> >o3L8,wl : q ; . ..
REACTOR !E 8 TY PIC AL
] F1- F
=
---_J V -
o r ---- ' i g I " : l ' FI 1 l F-
~
, N j_
=
9_ _
}. .
,[
._ TO MAIN ,
Y -- CONDENSER I h A fF '
- ~ ~ ~
NON REGEN ER ATlVE
, , ,o HEAT EXCH ANGERS
" CLEAN-UP FILTER DEMINERALIZED WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY 5ATETY ANALYSIS REPORT FICRE 4.10-4 LEAK DETECTION REACTOR WATER CLEA'i'.: DIFFERENTI AL BLOW
a
)
. I I
i ZPS ! AMENDMENT 12 i 4.3 (ZPS - April 9, 1971,- AEC Question 4.13) QUESTION Discuss the extent to which protection will be provided in th'e design of the containment and " core standby cooling system"against missiles that' might originate from the failure of the recirculation pumps. Discuss tne potential for overspeed of the recirculation pump motors as a result of a LOCA. In the event that the consequences of the recirculation pump-motor overspeed are considered indetermi-nate, what design measures could be taken to prevent overspeed beyond design values? ANSWER This concern was the subject of questions 12.3.7, 8 and 9 in Amendments 3,10 and 15 of the -Erico Fermi Atomic Power Plant Unit 2 (AEC Docket #50-341). As reported in those amendments, extensive studies are currently under way to determine the potential for missile generation, types of possible missiles, and design measures to mitigate such an occurrence. The results of these studies will apply to the Wm. H. Zimmer as well as to the Fertni plant. It. is anticipated that these studies will be completed and that a satisfactory resolution to this j concern will be obtained before the end of the year 1971. l 1 i 4.3-1 '
^
i ZPS AME%ThST 12 l. 8 I 4.3-2 (ZPS - April 9,1971, AEC Question 14.14) QUESTION [ The Wm. i H. Zinsner plant is the first BWR facility to utilize flow control
! valves instead lation flow of a variable recirculation pump speed control system for recircu-control. This is a new design feature of' the GE-1969 BWR plant. The recirculation loop flow control valves are 20 inches in diameter. The largest valve- previounty used in BWR's was 14 inches.
tests do you plan on the flow control valve? Consequently, what performance l If the tests are not full scale, I discuss how you will extrapolate the results of the tests to a 20-inch valve.
.l ANSWER i
f. full production The tested sizeprototype actuator. valve is an eight inch valve and the actuater is a Extrapolation of the tet. results on the 8" valve I' to 20" valve size is done by means of conventional hydraulic scaling techniques similar to " affinity laws" used in scaling pump proportions. It is planned that the first production flow control valve and its actua-4 tor will be for. actual service. tested in conjunction with the pump that will be installed with it l
, These tests will be performance tests based on functional requirements.
General Electric and the applicant are confident that the scaling methods are adequate and that the performance tests will confirm this adequacy. ! l 4 e
\
\
4.3-2
ZPS AMESDMENT 12 4.4-2 (ZPS - April 9,1971, AEC Question 4.14) ~ QUESTION Discuss the potential for common failure modes for the pressure relief valves. In developing your response consider the valve problems discussed in the literature (e.g. , " Conditions When Sticking of High Pressure Steam Valves May
' Occur" by C, L. Head,1955 American Power Conference, pp. 511-525) and experiences with valves in nuclear power plants (e.g., Reactor Safety Operating Experiences, ROE 69-8).
ANSWER The pressure relief valves used on Boiling Water Reactors have a long history of successful operation on conventional plants. Whereas the newer fossil plants have. imposed severe conditions of pressure, temperature and water impurities for which improved valves had to be developed, the BWR operates well within the state of the art with a very moderate environment. The paper by' Head discusses sticking of high-pressure steam valves (stop valves or throttle valves) on turbines, not pressure relief valves. The important dif- ; ferences are these: (1) Stop valves are normally open and depend on stem packing , to keep leakage low; Relief valves are normally closed and the stem is not sub- { jected to the steam environment, and the stem packing is not depended upon to con- ' trol leakage. (2) the buildup of the hard oxidide which led to sticking on stop valves did not occur at temperatures below 1000*F. A BWR operates at temperatures below 600* F. (3) Head's paper'was presented in 1955; the preblem he discusses has been solved by a proper selection of material and clearance, and daily valve stroking exercise. The best estimate of stop valve failure rates for BWR turbines is one failure in approximately 100 years. ROE 69-8 discusses various valve failures encountered on operating reactors. Most of these failures involved dif ficulty with the stem sticking either due to deposits carried over from borated water, or due to improper choice of a packing material for the environment. In one case, the spring was improperly sized and the valve would not close against rated steam pressure. The references cited are good illustrations of common modes of failure that can encroach on valve integrity; however, they are not applicable to pressure , relief valves. A credible common failure mode in the "f ailure to open" direction j has not been identified. The good history of operation through long years of appli- ; cation of the pressure relief valves continues to be the most convincing evidence of their integrity. l-1 L 4.4-2 ,
. _ _ _ _ _ - _ _ ___ _ a
r l ZPS AMENDMENT 12
! INSTRUCTIONS FOR UPDATING YOCR PSAR l
1 V0lWE 2 SECTION 5.0 - CONTAINMENT This Section has been amended to incorporate new information and answer to AEC pestions. I All changes have been indicated by a vertical line and the amendment num-i ber (12) in the right margin of the page. i i All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 12". Figures that have been altered in any way are indicated by the amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below: FIGURE NUMBER TITLE l-2.2-1 Station Site Area Topography 12 To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 2, SECTION 5.0 - Col..'AINMENT, remove and destroy Table of Contents Pages 5.0-1, 5.0-11, 5.0-iv and 5.0-v and replace with amended Pages 5.0-1, 5.0-11, 5.0-iv and 5.0-v.
i
- 2. In Volume 2, SECTION 5.0 - CONTAINMENT, remove and destroy the follow-ing text pages and replace with the appropriate pages listed klow:
REMOVE PAGE REPLACE WITH AMENDED PAGE 5.2-6 through 5.2-8 5.2-6 through 5.2-8 5.2-13 5.2-13 5.2-14 5.2-14 i 5.2-16 5.2-16 through 5.2-16.2 l- 5.3- 6 5.3-6 5.3-6.1 5.3-6.1
- 3. In Volume 2, SECTION 5.0 - C0hTAINMENT, remove and destroy Figures 5.2-1, 5.2-3 and 5.3-1 and replace with amended Figures 5. 2-1, 5. 2-3 and 5.3-1.
t L_____.__ _ -. _ - - - - -
gpg AMENDMEhT 12
- 4. In Volume 2, SECTION 5.0 - CONTAINMENT, behind the red tabbed divider page titled " Amendments to Section 5.0":
'~
- a. Remove and destroy Pages 5.2.3.1-2 and 5.2.3.1-3 and replace with amended Pages 5.2.3.1-2 through 5.2.3.1-5.
- b. Remove and destroy Pages 5.2.3.6-1 through 5.2.3.6-3.
- c. Behind Page 5.2.3.7-1 insert Pages 5.2.3.8-1 through 5.2.3.8-4.
i 1 +- d. Behind,Page 5.3.3.3.3-2 insert Pages 5.3.3.3.3-3 through 5.3.3.3.3-7.
- e. Behind Page 5.3.3.3.3-7 insert Page5.3.4.3-1.
- f. Behind Page 5.3.4.3-1 insert Page 5.3.4.4-1.
k i; l ) . i i Si l -
- ZPS AMENDMENT 12 SECTION 5 0 - CONTAI!EENT TABIE OF CON'IEffTS M
5.0 CONTAllHENT 5.1-1 5.1 S1HMARY DESCRIPTION 5.1-1 ! 5.1.1 Ceneral 5.1-1 5.1.2 ' Pr. ary containment 5.1-1 5.1.3 Secondary Containment 5.1-1 5.2 PRIMARY CONTAINMENT 5.2-1 f 5.2.1 Safety Objective 5.2-1
-5.2.2 Safety Design Basis 5.2-1 5.2.3 Description 5.2-2 5.2.3.1 General 5.2-2 5.2.3.2 Drywell 5.2-3 5.2.3.3 P essure Suppression Chamber and Vent System 5.2-4 f
5.2.3.4 Pene trations 5.2-4 5.2.3.4.1 General 5.2-4 5.2.3.4.2 Pipe Penetrations ] 5.2-6 { 5.2.3.4.3 Electrical Penetrations 5.2-6 l 5.2.3.4.4 TIP Penetrations 5.2-9 l 5.2.3.4.5 Personnel and Equipment Access 5.2-9
- 5. 2.3.4. 6 Access into the Pressure Suppression Chamber 5.2-12 5.2.3.4.7 Access for Refueling Operations 5.2-12 5.2.3.5 Isolation Valves 5.2-12 5.2.3.5.1 General Criteria -
5.2-12 5.2.3.5.2 Specific Criteria 5.2-14 5.2.3.6 Primary Containment Venting i d Vacuum Relief System 5.2-15 5.2.3.7 Primary Containment Normal lienting, Ventilation and 5.2-16 Air Conditioning System q 5.2.3.8 Provisions for Additional Primary Containment Equipment 5.2-16 12 5.2.4 Safety Evaluation \ ~ 5.2-16.2 ; l 5.0-1 l i _ _ _ _ _ l __ _ __
ZPS ATRIENT 12 R i TABLE OF CONIENTS, (Continued) PACE j 1 5.2.4.1 General 5.2-16.2 12 5.2.4.2 Primary Containment Characteristics During Reactor .5.2-17 Blowdown i 5.2.4.3 Primary Contairuneat Characteristics Af ter Reactor 5.2-18 Blowdown 5.2.4.4 Primary Containment capability 5.2-18 5.2.4.5 Primary contairunent Isakage Analysis } 5.2-18 5.2.4.6 Missile Protection 5.2-20 I 5.2.4.7 Pene trations 5.2-21 5.2.4.8 Isolation valves f 5.2-23 5.2.5 Inspection and Testica 5.2-26 5.2.5.1 Prirary containment Integrity and Leak Tightness - 5.2-26
, 5.2.5.2 Pene trations 5.2-26
- 5. 2. 5. 3 Isolation valves 5.2-27 5.2.6 Operational Nuclear Safety Requirements 5.2-27 5.3 SECONDARY CONTAINMENT SYSIDI 5.3-1 5.3.1 Safety Objective 5.3-1 5.3.2 Safety Design Basis 5.3-1 s 5.3.2.1 General 5.3-1
- 5. 3. 2. 2 Reactor Building 5.3 -1 I
- 5. 3. 2.3 Reactor Building Heating, Ventilation and Air 5.3-1 Conditioning System 5.3.3 Description 5.3-2 5.3.3.1 Reactor Building 5.3-2 5.3.3.2 ' Reactor Building Penetration 5.3-2 j 5.3.3.3 Reactor Building, Heating, Ventilation and Air 5.3-2 Conditioning Syatems q
5.3.3.3.1 Normal Ventilation Design Features 5.3-2 . 5.3.3.3.2 Abnormal Ventilation Design Features 5.3-3
- 5. 3. 3.3. 3 Standby Gas Treatment System 5.3-4 l J
\
t 1 5.0-11 ) i _ - _ - - - _ - - - _. - J
EPS AMENDMENT 12 BECTION 5 0 - ColfrAINNENT UST OF TABTES l l TABIE NtMBER , TITLE PA2 l l 5.2-1 Primary Containment - Drywell and Pressure 5.2-5 j Suppression Chamber - Principal Design j Parameters and Characteristics j 12 5.2-3 Electrical Penetrations - Environmental Design 5.2-10 Conditions 5.2-4 Primary Contstament - Pressure Suppression 5.2-19 System - Maximun Blowdown Pressure Comparison I 1 1 1 1 l l 5.0-iv
p. 1 ' l l- . ; ZPS
~
V l
\
AMENDhENT 12 '
. SECTION 5.0 - CONTAINMENT
{ _ LIST OF FIGURES i FIGURE NUMBER ) TITLE 5.2-1 Primary and Secondary concrete containment ! Structures 7 12 { 5.2-2 Column and wall Base Detail at Floor Liner Plate 5.2-2.1 Drywell Floor Joint At Contain. tent Wall 5.2-3 l7 Typical Section at Buttress i 12 ! 5.2-4 Tendon Access Gallery 5.2-5 Typical Leak Test Chamber 5.2-6 Drywell Heat Attachment Detail - Tendon l Anchor at Drywell Head
)
7 5.2-7 Primary Containment System Hot Process Line Penetration 5.2-8 Primary Containment System Cold Process Line Penetration 5.2-9 Typical Electrical Penetration Assembly 5.2-10 Personnel Access Lock 5.2-11 Drywell Cooling and Ventilation System 5.2-12 Emergency Lock and Equipment Hatch 5.2-13 Typical Layout of Hoop Tendons 5.2-14 Typical Layout of Vertical Tendons 5.2-15 Reactor Containment Development Elevation i 11 5.3-1 l Standby Gas Treatment System 5.3-2 l5 12 j
\
CSCS Equipment Area Cooling System 5.3-3 Reactor Building Ventilation System 5.3-4 Standby Gas Treatment System Equipment Train 5.3-5 Schematic-Showing Mixing Effect of Supply Outlet
ZPS AMENDMENT 12
- a. They are capable of withstanding the peak transient pressure, bi- They are capable of withstanding. the. forces caused .by impingement of
.the fluid f rom the rupture of the largest local pipa or cont.ection j without failure.
[ c. They are capable of acconnodating the thermal and mechanical stresses [ which may be encour.tered during all modes of operation without fail-j ure. . Refer to Table 7.3-1 for approximate number and size of these penetra-tions and Figure 5.2-15 for the location of the penetrations. l2 l11. l 5.2.3.4.2 Pipe Penetrations I Pipe penetrations will be of the type as shown in Figures 5.2-7.and , 5.2-8 for all process lines penetrating the containment. Piping penetration l locations will be consistent with the requirements of the safety design basis 7 ! for the piping, control and instrament systems. The pipe will be welded' directly to the sleeve whir.o is imbedded into the concrete as it penetrates the i containment. Insulation atid air gaps are' provided around the pipe to reduce thermal stress in the containment during normal operations. In addition ta their function as a primary containment barrier, the -I
.j. penetrations serve as anchors to the pipes. Thermal growth and movement will be
.taken up in the piping system. Guided supports will be used where required to direct pipe expansion. The piping system will be designed such that the result-ant combined stress in the pipe and penetration components under normal and ac-cident conditions do not exceed the code allowable design limits, ASME Nuclear Vessel Code, Section III, Subsection B.
5.2.3.4.3 Electrical Penetratig_s Electrical penetrations will be designed to accommodate the electrical requirements of the plant. These are functionally grouped into low voltage power and control cable penetration assemblies, high voltage power cable pene-tration assemblies, and shielded cable penetration assemblies. . T.ach penetration seal will have the same basic configuration. An assembly will be sized to be inserted in the 12 inch schedule 80 penetration nozzles which are furnished as part of the containment s t ructure . Installation of the penetration assembly will be accomplished by inserting it from outside of the containment into the penetration nozzle. Three. field welds are required to complete the installation i of the assembly in the penetration nozzle. It is intended that the penetration canister assemblies be supplied from commercially available designs (Figure L 5.2-9). , Headerplates conforming to the inner diameter of the penetration nozzle l will be provided at each end of the penetration assembly, forming a double pres-sure barrier. Radiation shieldi:g will be attached to the penetrations on the 5.2-6 l __C __ _ _ _ _ _ _ . . - -
i i l 1 ZPS
. M!ENDMENT 12 4
1 1 1 i l, THIS SHEET LEFT BIRIK l l 5.2-7 ..
) l' - 1 la l ! ZPS l. A".IND>1ENT 12 h t I i l i l i l THIS SHEET IIFT BLANK i. 9
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5.2-8
\
-- _ _ _ _ _ _ _ __ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I
r ' f C j
- l 1
x ZPS : AMENDMENT 12
,c 3 ,
Valves in thig category will be designed to close automatically from. selected sigdals and will be capable of remote manual actuation 'from
- . the main control room (Refer to Table 7.3-1). l 12
( The valve's are physically. separa ted. On lines connecting to the re-actor. primary system, one valve lis loca ted inside the primary con-f , tainment and the second cutside the primary contairament as close to
; the primary containment as practical.
I
- b. . Lines which pr.etrate the primary containment and which neither con-
.nect to the reactor primary system nor which open into the primary l containment, ars'provided with at least one valve which may be lo-
.cated outside the primary containment, valves in this category are capable of remote actuation from- the main control room.
y ,t
, c. Motive. power for the valves on process lines which require two valves are physically independent sources to provide a high proba-l, bility that no single accidental event could interrupt motive power to both closure devices. Upon loss of valve actuation power and ,
when containment closure action of the valve is called for, the valve will fail. closed. Loss of valve actuation power is detected
>; and' annunciated.
s -
- d. Main steam line' isolation valve closure time is such that for any de-sign basis break', the coolant loss is restricted so that the reactor
.y. core would not be' uncovered.
p' e. Valves, sensors, mand other automatic devices ecsential to the isola-
. tion of the contatranent are provided with,yeans to periodically test PN functional performance of the equipment. ,,Svch tests include imenscration of goper working conditions, correct set point of sensors, proper speed responses, and operability of fail safe fea-tures. .
The following are exceptions to the above is 21ation valve criteria:
- a. Automatic isolstion valves, in the usual sensa, are not used on the inlet lines of the reactor core and containment cooling systems, and reactor feedwater systems, since operation of these systems is es- j sential following ~a loss-of-coolant accident. Since normal flow of water in these systems is inward to the reactor vessel or to the primary containment, check valves located in these lines will pro-vide automatic isolation, when necessary.
Automatic isolation ' valves are not provided on the outl,et lines from the pressure suppression chamber to the core spray and shutdown-1 containment spray pumps. These lines return to the containment and S.2-13 ad
ZPS AMENDMEhT 12 h lI are required to be opened during post-accident conditions for . opera-tion of these systems.- ! b. No automatic isolation valves are. provided on the ' control rod ' drive
, hydraulic; system lines. These lines 'are isolated by means of the-normally closed hydraulic system control valves located ira the reac-.
f tor building, and by means of check valves comprising a' part of the
, drive mechanism.
c. TIP isolation valves and small diameter instrument lines. Table 7.3-l is 's typical listing of the prine,ipal isolation valees.
' - table indicates the _ size, closure time, motive power, service and number of The l 12 valves for each service. -
5.2.3.5.2 ' Specific criteria l< Effluent lines such as main steam' lines which connect to the reactor vessel or which are open to the primary containment have air powered valves. r Studies have shown this arrangement to.have a high reliability with respect to
- j. functional performance. ' These valves are closed -automatically by the signals indicated in Table 7.3-1. Any ac operated valve motor will be connected to the y
" emergency bus to assure available power at all times. Any check valve will l12 [ close automatically by reverse flow through the pipe. Any motor operated valve can be~ closed by remote manual signal. 4 TIP system guide tubes are provided with an isolation valve which closes automatically upon- receipt'of proper signal and after the TIP cable and fission ' chamber have been retracted. A back-up isolation shear . valve is included in - series with the isolation valve. Both valves are located outside the drywell. The function of the shear valve is to assure integrity of the containment even in the unlikely event that the other isolation valve should fail to close or the chamber drive cable should fail to retract if it should be extended in the. guide tube during the time that containment isolation is required. This valve is de-signed to shear the cable and seal the guide tube upon an actuation signal.
. Valve position (full open or full closed) of the automatic closing valves is indicated in the control room. Each shear valve is operated independently. The valve activating each is expected to beprovided.
circuit an explosive type valve, de operated, with monitoring of - In the event of a contairunent isolation signal, the TIP system receives a command to retract the traversing probes for the several mechanisms. Upon full retraction, the isolation valves are then closed automatically. If a probe were jammed in the tube run such that it could not be retracted, instruments would supply this information to the operator, who would in turn investigate to L d :termine if the shear valve should be operated. i: 5.2-14
Z?S 4 AMENDMENT 12 closed isolation valve and will be interconnected to the drywell purge system at a point between the two isolation valves in the drywell purge exhaust line. Nominally, the suppression chamber will have the same purge rate as the drywe l l . Thechamber.
, suppression above systems will be designed to purge either the drywell or the Provision is .not made to purge both areas simultaneously.
fore, The purge system is not required to operate following accident. The re-the drywell and suppression chamb(r purge exhaust fans are not connected to the essential bus. q i 5.2.3.7 Primary System Containment Normal Heatieg. Ve n t ila t ion and Air Conditioning The drywell cooling system design for each primary containment is based on recirculating chilled water through the drywell air handling units to main-tain the required ambient temperature. Figure 5.2-11 shows the design. The de-sign vi11 include two (100%) air handling units located within the dryw(11 with
. one (1) unit operating normally and one serving as spare. Air will be distributed through ductwork and/or up through the annular space between the reactor vessel insulation and the biological shield.
cooling such as the recirculation motors, CRD area, and the bellows area. Air w Re ttxrn air is ducted back to the operating units' operation, and air distribution balance of the system.This design will simplify the desig:m The design is based on the use of two (2) 100"/. centrifugal water chillers and associated chilled water pumps f or producing and recirculating the chilled water and one supplied to the drywell air handling units. One unit will normally operate unit is standby. be supplied to the water chiller condensersReactor building closed cooling water system wa normal and loss of offsite power conditio:s. to dissipate absorbed heat only under The drywell cooling system is not required for safe shutdown, but is de-
- signed with redundant equipment and powered from essential buses t insure con-tinuous operation in order to satisfy the power generation design objective.
'Ihe drywell cooling system is not designed to operate following a IDCA.
All hand switches for operating the equipment will be located in the main control room. 5.2.3.8 Provisions for Additional Primary Containment Equipment The primary containment will be designed so that a nitrogen containment inerting and purge system ar.d a hydrogen recombiner can be added to the plant if it is determined that such systems are necessary. Penetrations for these systecs 12 will of thesebe provided sys tees. and the design of the plant will in no way preclude the additio= The necessity of making provisions for these systems is due tc' the possibility of excessive hydrogen generation following a loss of coolant acci-dent. 5.2-16 { - _ ]
ZPS AMENDMENT 12 Hydrogen generation in BWR's is discussed in Dresden 3. Amendment 23.
, Details of the program for resolution of this concern are given below. The Cir.-
)
' cinnati Cas & Electric Company is following this problem. The station design will also include provisions for venting of the containment through the SBGTS (see Paragraph 5.3.3.3.3). Equipment will be installed which will analyze the contain-ment atmosphere for dangerous levels of hydrogen (see Paragraph 10.20). I The program underway by General Electric is to develop a recombiner using l the following bases:
- a. CH2 - 0.20 molecules /100ev in the pool and 0.44 in the core based on experimental observations from tests at ORN1..
- b. Maximum allowable H2 concentration in drywell - 47 in drywell..
- c. An arbitrary imposed 0.757. metal water reaction occurs at 30 minutes j af ter the IDCA. This is over 10 times that calculated with the ECCS. I i
- d. A decontamination factor of 2 is taken on the halogens to the pool '
} (consistent with AEC dose assumptions). This results in 257, of the j
halogen fission products from the core being deposited in the sup-pression pool when using TID-14814 fission product assumptions. l Analysis of the ORNL radiolysis data indicates that for our conditions, l the G f actor ;or a suppression pool-type situation is usually about 0.1 and is { I always less than 0.2 (the recommended value). Similarly, the data indicate that for a core-type situation, the G facter is between 0.2 and 0.44. Hence, a value 12 of 0.44 will be taken as the G value in the core region, with 27. of the decay ) energy in the core abaorbed by the water. If the effects of water temperature, back reaction due to metal water hydrogen, core flow, core voids and equilibrium concentrations are considered, the choice of G=0.2 in the pool and 0.44 in the core are very conservative. The recombiner system should be designed to handle as much es 0.75 initial metal water reaction hydrogen, which shall be considered to be added to the con-tainment at a uniform rate for the first 30 minutes following the accident. Requirements that are placed on the recombiner system design are:
- a. Only 1 recochiner system is needed. The back-up system is containment venting through the standby gas treatment system. j
- b. All active components and instruments will meet single failure cri- l terion. Instrumentation and controls will meet IEEE 279 requirements. l
- c. A redundant hydrogen continuous sampling system will be provided.
- d. A redundant activity sacpling system will be provided (not continuous). '
Existing systems for radiation activity sampling are acceptable, pro-vided they e:;ually meet the same cuality requirements as the recombiner system. 4 5.2-16.1
L I ZPS AMEhTMENT 12
- e. The system will be capable of handling anywhere from 0 to 47. hydrogen by volume. (This corresponds to about 17. metal water reaction.)
- f. The system will meet the Class I seismic requirements.
- g. The system will maintain the atmosphere below 47. hydrogen by volume.
- h. The system will not result in a net decrease in the original quantity of atmosphere, i.e., the noncondensible gases from the containment must be replaced to protect pump NPSH.
- 1. The system will be testable, at least during plant shut-down, to prove it can remove hydrogen and oxygen.
- j. The containment atmos-here will have a 20 to 1007. relative humidity and a temperature range of 100 to 250*F.
- k. The system will work for containment pressures at atmospheric pressure to 15 psig.
- 1. The system will either be properly isolated, or designed to the 45 psig 12 containment design pressure.
- m. The system will be capable of being manually actuated 1/2 hour af ter the IDCA.
The tentative schedule fer completion of this program is: Preliminary design - mid 1971 Testing completed - end 1971 j Final design - beginning 1972 5.2.4 Safety Evaluation ' 5.2.4.1 General The primary containment and its associated safeguards systems are designed to accomplish four principal furetions, namely: h 5.2-16.2
)
ZPS { l AMENDMEN'i 12 i
- h. A high efficiency particulate filter identical to the one described i in d. above. 5 )
{i l q All of is required. the above equipment will be evaluated to determine 11 shielding The components of the Standby Cas Treatment System will be de-signed to withstand the doses received from the radioactive isotopes released as a consequence of any accident. The TID 14844 source model will be used to cal-culate the quantity of ' activity released as a result of these accidents. The results of this analysis will be used to design and specify all standby gas treatment system equipment components.
\
All .of the above equipment will be evaluated to determine if shielding is . required Retention capacity of the filters for particulate and iodides will be, as a minimum, that amount which might reasonably be expected to be released in the reactor building during a postulated design basis accident. I l Ducts and equipment of the standby gas treatment system will be designed ! i to handle saturated air at pressures and temperature corresponding to those of the containments to which they will be connected. The standby ! j {
, gas treatment system will be designed to withstand pressures up to 2 psi.
Administrative procedures will be incorporated in the technical specifi-cations to prevent purging the drywell when drywell pressures are in ex-t
-I cess of 2 psi. The corresponding design pressure will apply to ducts
- and equipment up to and including the isolation valves on each duct.
The system will be designed for high reliability. The systes equipment will not contain radioactive matter except af ter an accident, and hence, will bc accessible' and can be tested and maintained during normal plant operation. System design will meet seismic Class I requirements. 1 s The standby gas treatment system will start (and the normal reactor k building ventilation system will be taken out of service automatically) in response to any one of the following signals:
- a. High radiation in the fuel pool ventilation exhaust. 12
- b. High pressure in the drywell.
c ., Manually initiated signal f rom the main control room. Any one of these signals, through a relay circuit for each of the stand-by gas treatment systems, starts the equipment train simultaneously. A signal from either relay circuit initiates reactor building isolation by closing the j sir Operated isolation valves in the suppl: and exhaust headers of the ventila-tion system. These valves will be specified and tested to insure a maximum ten 7 second closure time. These valves are operated by air cylinders with instrument 5.3-6 (
EPS AMEN &ENT 12 air being controlled by an air solenoid valve for each isolation valve. Pres-sure switches monitor the instrument air supply to each valve operator and open their contacts (thereby closing the isolation valves) when air pressure decreases 7 to (later) psi. An air tank mounted locally near each valve will insure a suffi-cient air supply to close the isolation valves under all plant operating condi-tions. Each tank is instrumented with pressure switches which cause alarms on the main control board if tank pressure is not sufficient to allow two operations of each valve. In the worst event of buildiag pressure equalizing during changeover from normal ventilation system to SBGT System, it will take approximately 30 seconds to reduce the building pressure' from 0 psig to 0.1 inch negative, and approxi- 12 mately two minutes from 0 psig to 0.25 inch negative pressure. The response of the radiation monit'or and electric relays is almost immediate. A maximum time of about 5 seconds will be required for the SBG7 System to obtain full operating speed af ter high radiation or high drywell pressure is sensed. The standby gas treatment system is designed to automatically start both equipment trains simultaneously. Idhen both trains are operating an audible and 7 visual alarm on the main control board will warn the operator to shut-down one of the trains. 1 1 i i 'i l l t 5.3-6.1 ) ( . I j .- -- .----- ---- o
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I AWENDMENT12
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ZPS AMINDMENT 12 5.2.3.1-2 (AEC - October 13, 1HO, Question 14.2) QUESTION Describe the bases for. determining the height of the downcomers above the drywell floor. If it is assumed that the blowdown liquid separates from the blow-d own s team , can the downcomers be flooded with water from broken recirculation line? Discuss the relationship of the break area to vent area ratio to the peak drywell pressure and to the peak drywell floor differential pressure. Are the downcomers designed to withstand waterhanrner and other possible dynamic events that may occur during blowdown? ANSWER All evidence is that the flow from the reactor would explosively decom-press upon entering the drywell. This would lead tu good mixing of the air, 4 i water vapor and water droplets within the containment rather than to an accumula-
!' tion of liquid on the drywell floor.
A y A consideration of the vent geometry and flow rate leads to the conclu-f sion that it would be difficult for the vents to become flooded. During the time ,) 4j ; .; j tha t the vents are still being cleared of water, the flow rate into the vents is ii l relatively slow and it could be postulated that some water would accumulate on the drywell floor. {; this time only 200 f However, the vents are cleared in s0.6 seconds and during ^ t 3 of liquid
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I would enter the drywell. Even if all of this accumulated on the floor, the liquid depth would be less than 5/8 inches, con- i servatively less than the S10 inch projection of the downcomers above the drywell floor. i Thus, flooding of the vents could not occur during this part of the
} transient.
k Following vent clearing, flow is established in the vents; the velocity i j, at which the mixture of air, water and vapor would enter the downcomers is in 12 i { ; ! excess of 250 ft/see at the start of the accident and would not be less than i 5 200 that f t/sec et anytime prior to the occurrence of the peak drywell pressure. By i time, pression 807. of the reactor liquid would have been carried over to the sup-chamber. ! With such high vent entrance velocities and remembering that j' because of the baffles the flow entering the vents will be moving horizontally, i there would appear to be no mechanism by which gross flooding of a significant
!j h j number of vents could occur. Figure 1 shows the relationship between vent flow area 1l sure. and both the peak containment pressure and the peak deck differential pres- lj It can be seen that nearly 407. of the vents could be plugged before the containment design pressure would be reached. Further, even if 507. of the reactor ;j liquid entering the drywell during the blowdown were carried over to the suppres- ')
g sion chamber, the liquid depth accumulated on the floor would be less than 8.5 inches. For the over-under containment design being used for the Wm. H. Zimmer l Nuclear Power Station the peak drywell pressure is relatively insensitive to the
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1 5.2.3.1-2 ,1 -_.._-_1._.
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ZPS i i AMENDMENT 12 ' k i ! i I i
' break to vent area ratio. 'Ibe peak drywell pressure in this type of containment
( would occur at the end of the blowdown period and occurs when all the noncon-i densibic gas originally in the drywell have been washed over to the suppression chamber air space. h The attached Figure 2 shows both the peak drywell pressure and peak deck j differential as a function of primary system break area. The design basis acci-dent for the Wm. H. Zinner Unit results in a blowdown area of 2.238 f t2. thus,
! the values presented on Figure II for break areas in er. cess of 2.238 f t$ are for j illustrative purposes only.
It can be seen that the primary system break area would have to i,e doubled before Jhe peak containment pressure would equal the design value of 45
psig. Similarly the break area could be almost tripled before the peak deck dif ferential would approach the design value of 25 psi. For the range of break areas shown on Figure 2 the peak deck differential pressure occurs at the time the vents are cleared of water i.e., at s.6 seconds af ter the postulated instan- I taneous break of the recirculation line. The differential pressure during the 12 vent flow transient is less than that which occurs at the time of vent clearing.
The only dynamic conditions which the vent system would see during a loss of coolant accident are the rapid increase in drywell pressure and the resultant clearing of the liquid originally in the vents. The vents are designed for these conditions. I Water hammer effects associated with condensing steam are localized in nature and geometry dependent. In the region where steam condensing will be occurring, the vent system for the Wm. H. Zimmer unit is geometrically similar to the Bodega Bay and Humboldt test apparatus (See document referenced in the answer j' to Question 14.3). Some of these tests were considerably more severe than the design basis accident being considered here. The Wm. H. Zicner containment has a 1 primary system break area to vent flow area ratio of .0081 whereas some of the j tests had a ratio as high as .06. None of these tests indicated any water hammer l or rapid condensation related damage to the vent system. I . Y l
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1 l ZPS B' E fg- , AMENDMENi 12 5.2.3.8-1-(AEC - October 13, 1970, Question 5.10) 12 L'
;; { QUESTION Means- should be provided to control the concentration of hydrogen below.' the lower flamnable limit in the primary containment following a loss-of-coolant accident. In developing these means, consideration should be given to the potential for hydrogen generation during the snort term due to metal-E^ water reaction and. over the long term due to rediolysis effects. Since venting
* - of the containment atmosphere to the outside environment should not be consider-
; ed as a primary means of controlling the hydrogen concentration, consideration should be given to other methods of primary hydrogen control which provide for E retention of the LOCA-produced atmosphere within the containment, and are designed as engineered safety features. -Provide a description of your proposed means- for hydrogen control. for the Zimmer units.and give the design basis for the hydrogen generation source terms, a
g ANSWER The answer to this question can be found in Paragraph 5.2.3.8 of the Wm. H. Zimmer Nuclear Power Station PSAR Amendment '12. 12 m M M M 1 E" h . 5.2.3.8-1 z -_--- . _ - . -
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l f ZPS l l l AMENDMENT 12 j l 1 5.2.3.8-2; (ZPS - February 23, 1971, AEC Question 5.13) j f- QUESTION On the basis of our evaluation of the potential consequences of hydrogen generated in the containment following a LOCA, we have concluded that means should be provided for mixing, sampling and control of combustible gases following a lhCA which do not necessarily involve releases of radioactive materials to the environment. It is, therefore, requested that your response to Question 5.10 as presented in Amendment 4, be expanded to indicate the !- design bases and conceptual designs of systems intended to control the concen-
, tration of hydrogen generated following's LOCA as a result of metal water reaction, radiolytic decomposition of water and corrosion. The proposed ;
systems should meet the design, quality assurance, redundancy, energy source and instrumentation requirements for an engineered safety feature. The table below lists values of the parameters that we ci nsider should be used to determine the concentrations of hydrogen and oxygen in the primary containment and to evaluate the design of the systems to control these gases. Based on our studies and discussions with your representatives, we have concluded that use of these listed values would assure reasonably conservative system design parameters. If in developing your preliminary design, parameters other than those listed are used they should be documented and justified. , Your response should include the engineering criteria to be followed and the details of a specific design, other than venting, to control the concen-tration of hydrogen in the containment following a LOCA. - Hydrogen Generation Assumptions l
- 1. Fraction of fission product radiation energy absorbed by the coolant. (a) Beta (1) Betas froc fission i products in the fuel rods: 0 t
I (2) Betas from fission products intimately mixed with coolant : 1.0 (b) Gan=a (1) Gammes from fission products in the fuel rods, coolant in core region: 0.1 5.2.3.8-2
l ZPS : I AMENDMENT 12 (2) Ga nas from fission l products intimately mixed with coolant, all coolant: 1.0
- . 2. G(H2 ) 0.5 molecules /100cv
- f. 3. G(0) 2 0.25 molecules /100ey
- 4. Extent of metal water reaction 5 (percentage of fuel cladding that reacts.with water)
- 5. Aluminum corrosion rate for aluminum 200 mils /yr exposed to solution.
- 6. Fission product distribution model (a) 507, of the halogens s and 1*/. of the solids present in the core are intimately mixed with the coolant wcter.
(b) All noble gases are released to the containment. (c) All other fission products remain in fuel rods.
- 7. (a) Hydrogen concentration limit. 4 volume percent (This limit should not be exceeded if more than 5 volume percent oxygen is present)
(b) Oxygen concentration limit. 5 volume percent (This limit should not be exceeded if more than 4 volume percent hydrogen is present) ANSWEh Safety Guide No. 7 outlines the criteria for evaluating the hydrogen generation problem that follows a IDCA. These criteria are arbitrary and apparently overly conservative when co= pared to the results of recent ORNL tests. 5.2.3.8-3
~
l ZPS I 1 AMENDMENT 12 1
.l The. criteria considered appropriate for an analysis of the' problem and 1
for the developmer.t of hydrogen recombination equip:acnt are given in Paragraph j
.5.2.3.8.
' The need for eixing of the containment atmospher.: to preclude. the un-likely occurrence of hydrogen pocketing vill; be studied. It is not anticipated d that a gas such as hydrogen could pocket in an enviornment as' violently _ as I' .
that which would result from LOCA conditions. However, the design of the I 4 containment will not preclude the. addition of such mixing equipmenti should, that become a requirement. If_ such equipment is required, it vill be designed -
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to ' operate in.the accident envi.ronment and to meet the- separation and redundancy .j requirements for safety systems. l The equipment for sampling the post-accident containment environment is described-in new Subsection 10.20 (Amendment 11.). I l I-
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AMENDMD*T 12 5.3.3.3.3-3 (ZPS - February 23, 1971, AEC Question 5.14) _ QUESTION It is indicated on Page 5.3-4 that an electric heating coil is used in the standby gas treatment system to reduce relative humidity to less than 70".. Please indicate whether or not temperature sensors are used with this device. Describe the consequences of the electric coil overheating during the operation of the standby gas treatment system following a loss-of-coolant. It is stated on Page 13.4-27 that the high efficiency filters in the standby gas treatment system will be checked with DOP. How many grams of EOP could be trapped in the HEPA filters due to routine testing? Could the DOP released f rom the HEPA filters be trapped in the charcoal filters as a result of an inc.rease in the HEPA filter temperature following a design basis ac;:ident? Discuss the effects of DOP on iodine retention and/or conversion to an organic form in the charcoal and on the possible change in the ignition temperature cf the charcoal due to the presence of this contaminate with a flash point of 410'r. ANSWER The standby gas treatment of approximately 8 kW. system electric heater will have a capacity
) This capacity is just sufficient to increase the dry-bulb temperature to reduce the relative humidity of inlet air to less the 707..
Under the worst conditions, assuming the air coming f rom the secondary centain-ment is as high as 150'F, the temperature rise across the heater will be approx-imately 17*F resulting in a maximum temperature entering the filter sectien of approximately 167'F. The heater will be designed and operated as a one-stage device such that the heater will operate whenever the standby gas trea:nent blower is operating. i The heater will be equipped with internal high temperature j cutouts and a downstream high limit thermostat which will serve to deenergize l the electric heater in the event of heater or system c alfunction. Af ter initial installation of the HEPA filters, and periodically there- I after, the filter bank leak integrity will be determined using DOP. Dr. t'e basis of testing each HEPA filter element individually, a maximum of appr:xi-mately 2.82 ugm of DOP will be retained in each HEPA filter element. The stand-by gas treatment will tentatively have four (4) HEPA filter elements and there-fore, a total of approximately 11.28 ugs of D0P can be retained af ter testing. It is possible that the DOP can be released from the HEPA filter as the filter temperature increases; however, the charcoal absorber, being a rather peer particulate filter, would retain a negligible quantity of DOP released in this l manner. ! It is more likely, and a more sevegsituation, if the DOP is con-verted to methyl iodine by combining with I . If the worst case is ass ned, that is, the accider.t occurs immediately af ter the HEPA filters have been tested and 11.28 ugm of DOP is available for conversion, it is es timated that not more than 0.045 grams of methyl iodine will be formed, assuming cocelete
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-l AMENDMENT 12 conversion. Since approximately 100 grams of radioactive iodine enters the SBCT systec. in a 30-day period of which approximately 107 is nomally assumed to be t in the organic form as methyl iodine, the additional generation of 0.045 grams .I of methyl iodine is seen to be insignificant in terms of the basic assumptions. .I The filter system will be designed with sufficient charcoal capacity to ' absorb the maxi m loading of both the radioactive and nonradioactive isotopes of iodine
[. and bromines, and therefore, the performance of the standby gas treatment system will not be impaired due to the additional methyl- iodine that could be generated. 4 The operating temperature of the standby gas tr.tatment system, even after an accident, is substantially below the 410*F flash pos.nt of DOP. Any retained
DOP on the HEPA filter will either combine with fodine to form methyl iodine, I' which will be absorbed by the charcoal and/or will be vaporized as a result of increasing ' the system temperature and will pass on through the system. In either case, a change in the charcoal ignition tmperature is not likely.
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j ZPS I AMENDMENT 12 5.3.3.3.3-4 (ZPS - February 23, 1971, AEC Question 5.15) QUESTION
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I Provide preliminary design information on the emergency.. gas filter system, ! including arrangement and spacing of components. For the charcoal absorbers, I specify the number, type, and arrangement of drawers in each bank, number of banks per unit, weight of charcoal, depth of charcoal. in each unit, type of char- '
- coal to be used. Outline the proposed test procedures to determine bypass flow
-for each component for each unit. Specify the minimum and maximum. filter face i air velocities with (a) one unit operative and (b) with both units operative.
Provide an analysis of the range of expected charcoal filter temperatures, assuming operation of. the electric heating coil and a TID release fraction of iodine from the core and all iodine absorbed on the first charcoal filter of the single unit, including the loss of flow case. Specify the provisions to prevent iodine reabsorption or charcoal ignition. i ANSWER ' the tesign configuration of the standby gas treatment system is present-ed schematically in Figure 5.3-1 of the PSAR and the tentative physical layout appears in Figure 5.3-4. This preliminary layout is predicted .on providing access into the filter housing between each component of the standby gas treat-ment train. This access will be used for visual inspection, removal and instal- ' lation of filter elements and for inplace leak testing purposes. Tentatively, each charcoal absorbing bank will be of the all welded design, bulk filled type to prevent the bypass of unfiltered air. The quantity and arrangement of the , charcoal will be designed to provide a contact time to assure the minimum char- l coal absorption efficiency of not less than 99%. The quantity of charcoal will i be not less taan that required to absorb all of the radioactive and nonradio-active isotopes of iodine based upon a maximum of 25; of the core inventory be-ing released as airborne, which can leak to secondary containment. The type of charcoal used will be an impregnated coconut shell of the best grade available ! i at the time of installation, but of a quality not less than Barneby-Cheney type 727, or equivalent. l At present, it is envisioned that the standard DOP test will be used to l i l-- test the inplace leak integrity of HEPA filter bank and the refrigerant R-112 test will be used to test the inplace leak integrity of each charcoal absorber , bank. The maximum filter velocities for both HEPA and charcoal absorbers will ! not exceed the values normally considered be good design practice. In the case of the HEPA filters, the maximum velocity is approximately 300 fpm. In ' the case of the charcoal absorber, the face velocity is a function of bed depth and required contact time. As stated previously, the total contact time will be not less than that necessary to insure ;.he minimum absorption efficiency, and since the bed depth has not veen determined, a limit on the charcoal absorber face velocity has not been established. The maximum velocities would 5.3.3.3.3-5 l
1 s , ZPS 4 L AMENDMENT 12 be expected to occur when only one standby gas treatment syster: is in operation. The minimum velocities will occur when both the standby gas treatment units are operative. These velocities are a function of the flow rate through each unit which will be dependent upon the total system resistance. This resistance is a function of the detailed physical design which has not been established. At velocities less than the desig*1 veloc.ities, there will be no decrease in per-forraance of either the HEPA filter or charcoal absorber. filter section is not expected to exceed 167'F.As indicated for Question 5.1 being released from core, the maximum heat dissipation Based upon 251 of the iodine on the HEPA filters and in the first bank of the charcoal absorber nominal 2300 cfm standby gas treatment flow rate, is approximately 4600 Btu /hr . Fer a both the HEPA filter and charcoal absorber will be 3'F.the temperature rise across The actual charcoal air and this temperature will be determined later. temperature is expec ) It is apparent, howe er, 1 that due to the relatively low magnitude of heat released and low temperature rise of the standby gas treatm:nt air, that the charcoal temperatures are well below the charcoal ignition temperature of approximately 350*C (662*F). As sum-ing a loss of filter flow, cooling air of will he provided from an independent blower. approximately 100 to 200 e fm magnitude l Assuming the minimum flow rate of 100 e fm, concurrently with a total maximum decay heat of 4600 Btu /hr, the maximum I air temperature leaving the charcoal will be approximately 218'F. Note that under these conditions the electric heat coil is inoperative and that the enter-ing cooling air temperature is assumed to be 150*F. To further minimize the ped with the necessary temperature detectors to alarm and protec tion system. These temperature detectors will have a setpoint to be established later, which will be determined so as to prevent unnecessary initia-tion of the fire protection system, but which will assure an adeouate sa fe ty margin to preclude the possibility of actual charcoal ignition. The possibility of iodine desorption is remote; however, the standby gas treatment system employs two charcoal banks in series and in the event that iodinebank, second deabsorbs from the first bank, this iodine will be reabsorbed by the i 5 . 3. 3. 3. 3-6
ZPS AMENDMENT 12 be expected to occur when only one standby cas treatment system is in operation. The minimum veIocities will occur when both the standb f gas t reatmen t units are operative. The se veloc i t ies ;re a function of the flow rate through each unit which will be dependent upon the total system resistance. This resistance is a function veloc i tiesof less the detailed physical design which has not been established. At than the design velocities, there will be no decrease in per-formance of et t'ner the HEPA filter or charcoal absorber. filter section is not expected to exceed 167'F.As indicated for Question 5.15, t Based upon 25*/. ci the iodine being released f rom core, the maximum heat dissipation on the HEPA fi? ters and in the first bamk of the charcoal absorber is approximately 4600 Btu /hr. For a nominal 2300 c fn standby gas treatment both the HEPA filter and charcoal absorber will be 3'F. flow rate, the temperature rise ac The ac tual charcoal temperature is expected to be only a couple of degrees warmer than the leaving air and this tec::perature will be determined later. that due to the It is apparent, howe er, rise of the standby gasrelatively low magnitude of heat released and low temperature treatment air, that the charcoal temperatures are well below the charcesal ignition temperature of approximately 350*C (M2*F). As sum-ing a loss of filter flow, cooling will be provided from an independent blower. air of approximately 100 to 200 c fm magnitude 100 air cfm, concurrently temperature with a total maximum decay heat of 4600 Btulhr, Assuming the maximum leaving the charcoal will be approximately 218'F. Note that under these conditions the electric heat coil is inoperative and that the etter-ing cooling air temperature is assumed to be 150*F. To further m!niinize the ped tion protec with theThese systen. necessary temperature detectors to alarm and temperature detecters will have a setpoint to be established tion of the fire later, which will 1,e determined so as to prevent unnecessary initia-margin to preclude the possibility of actual charcoal ignition. protection syste
'Ihe possibility of iodine desorption is remote; however, the standby gas treatment iodine deabsorbs system fromemploys two charcoal banks in series and in the event that the first bank, second bank. this iodine will be reabsorbed by the 4
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.j ZPS AMENDMENI 12 5.3.3.3. 3 jZPS - February 23, 1971, AEC Question 5.16)
QUESTION Please indicate how the charcoal is protected against ignition due to
. decay heat if a ventilation fan in one of the filter trains becomes inoperative af ter trapping a large quantity of radioactive iodine. Page 1.5-11, Paragraph
- 1.5.2.15 indicates that there is a device to measure the infiltration of the standby gas treatment system. Please describe this system used to monitor the ef festiveness of the standby gas treatment filters. What are the. limits of ,
detectability? .If an iodine. monitor is not included, please justify the . 1 absence of such instrumentation. Also, justify not providing the capability l j for automatic activation of the redundant filter train system by signal from w an iodine monitor should the charcoal filter system fail to perform its , intended f unction. Specify the total time required- to switch from the normal ! j containment ventilation system to the standby gas 1 treatment system upon detection of high' radiation. Also, specify.the time required to bring the - containment to the design ~ negative pressure. Indicate the assumptions made in ' ' this calculation, such as response time of the monitor, vent closure time,- ' and fan startup time.. What tests will be conducted in the charcoal to deter- '
.; mine that it is activated or impregnated and meets the design specifications? 7 3 ANSWER # -!i
[j 2 1. Refer to the answer to Question 5.15 for a discussion of the ' projections ( taken to prevent ignition of the charcoal bed in the case of the failure
, of a ventilation fan.
i m 2. j The flow ecasuring instrumentation mentioned in Paragraph 1.5.2.15 is 4 not used to measure the infiltration 'of the SBGT System but rather the flow through the systems which is the infiltration across the secondary containment boundary. This instrumentation is not used for monitoring the effectiveness of the filters, f 3. A' system to monitor the operational performance of the SBGT System is di not presently planned since a radiation monitor of the required sensi-31 tivity and response is not known to exist. If such a monitor becomes available consideration will be given to adding this to the SBGT System. l 4.- Refer to Amendment 12 Paragraph 5.3.3.3.3 , i 5. The quality control procedures or the standby gas treatment system
- will include documentation to insure that the charcoal is certified by the manufacturer to meet the design specifications.
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h ZPS AMENDMENT 12 5.3.4.3-1 (ZPS - April 9.1971. AEC Question 14.24) QUESTION Provide an analysis to show that, following a refueling accident, there is adequate time available to isolate the reactor building and actuate the standby gas treatment system before the first " puff" of contaminated air-borne particulate and gases reaches the ventilation ducts. The analysis should include the following:
- a. an estimate of the volume of the radioactive gas released to the surface of the pool;
- b. an indication of the expected bubble size used to determine the water decontaisination factor for radioactive iodine;
- c. a detailed description of the radiation detection instrumentation used to initiate the operation of the standby Sas treatrent system (SGTS);
d. an indication of the air velocities across the surface of the fuel i pool prior to the actuation of the SGTS, and as a result of the operation of the system; and
- e. a description of the path of the released gas to the SGTS.
ANSWER
< a. The volume of radioactive gas released to the surface of the pool is not needed to determine the time available to isolate the reactor b'uilding and actuate the standby gas treatment system,
- b. A DF of 10 is assumed for iodine.
- c. Refer to Paragraphs 5.3.4.3 and 7.12.6.
- d. During normal operation the surface velocity of air at the center of the pool will be' approximately 50 fpm. The surface velocity near the exhaust intake will be a maximum of 500 fpm. Af ter the SBGT is started the supply and exhaust air flows will be reduced by 207, and likewise the surface velocities will be 207. Iower.
- e. Refer to Paragraph 5.3.4.3 and Figure 5.3-3.
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ZPS 1 AMENDMENT 12 {j i 5.3.4.4-1_ (ZPS - April 9,1971, AEC Question 14.25) l QUESTION f Your analyses of the radiological consequences of the various design basis accidents are based on a containment' building mixing of 807.. In order to determine the dose reduction factors for an accident condition, provide the following containment mixing information.
- a. preliminary layout drawings (not schematics) that show the mixing system in the containment.
- b. a detailed description of the containment building mixing system including flow rates, exit -velocities, iree air volumes, number and location of duct work at each level (include blowers, intake and exit locations).
ANSWER Refer to Paragraph 5.3.4.4 and Figures 5.3-3 and 5.3-5. The remainder of information is yet unavailable but will be provided with the FSAR when the system design is completed. ' 0
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FOR l'PDATINC YOUR PSAR. , - 1. 4.<', sl' YOLtHE 2 ,. SECTION 6.0 -~ CORE STANDEY COOLING SYSTEM (CSCS) b. i~^ This section has been amended to incorporate c.aswe j'.co AEC question.a. _ All' changes have been indicated by.' a vertical line and the amendment - number (12) in the right cargin 'of the papcf.- All- pages (text, tables, figures) with changes have also been marked in the upper right corner of the paE, with / " AMENDMENT 12". , v Figures that have been alter 4d in any way are indicated:by the amendment - '
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number .in - the upper right corner of the figure;. note that thc:re are no other : narks that would indicate changes in figure. On the page marked " LIST OF FIGURES", ;{ figures that have changed in any way are designated by a vertical line with the j amendment number alongside the title of the figure. See example below: l FIGURE NLHBER g TITLE 2.2-1 Station Site Area' Topography 12 To update your copy oi the W. H. Zimmer Nuclear Power. Station PSAR, please use thelfollowing procedure: 3
- 1. In Volume. 2, SECTION 6.0 - CORE STA'OBY COOLING SYSTEM (CSCS) #~
behind the red tabbed divider page titled "Acendments to
.Sectica 6.0" insert the following pages in the order given below:
Paec Number 6.4-1 6.4-2 6.5-1 through 6.5-45 l
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ZPS s AMENDMENT 12
.6.4-1 (ZPS - April 9.1971. AEC Question 6.13)
QUESTION Describe the design features and the results of appropriate analyses re-garding mechanical integrity of piping and vital components of safety related systems including the core cooling systems in the event of excessive hydraulic forces (e.g., water hammer and steam compression) resulting from system initia- ! tion with the pu=p discharge lines not completely filled with fluid. l ANSWER One design requirement of any core cooling' system is that cooling watcr flow to the RPV be initiated rapidly when the system is called on to perform its f ur.c tion. This quick start system characteristic is provided by quick opening valves and quick start pumps. By always keeping the core cooling pump discharge lines full, the lag between the signal for pump start and the initiation of flow into the RPV can be minimized. If for some reason these lines were empty when 3 the systeams were called for, not only would the lag time be increased, but the i large momentum forces associated with accelerating fluid into a dry pipe might cause physical damage to the piping. Since the core cooling pumps are located in the sub-basement of the
- reactor building, approximately 75 feet below the point where the discharge piping enters the RPV, check or stop-check valves are provided near the pumps ! to prevent back flow from emptying the lines into the suppression pool. Past 1
experience has shown that these valves will leak slightly, producing a small l back flow that will eventually empty the discharge piping. To provide a posi-t !
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tive means to make up this leakage, design features will be incorporated as part of the IMR 69 product line for maintaining the CSCS lines completely filled j with fluid. Alarms will be provided to signal failure of these provisions. The arrangement of the CSCS filling system is' presently being designed. It is planned, h:=cver, that the systen will consist of three small pumps (gen-erally referred to as " Water leg pumps") which take suction from the suction lines of the core cooling pumps and discharge forward of the check valves in the main discharge lines of the core cool'ng pumps. One water leg punp takes suction frcxn RHR pump C and discharges to RER pump B and C discharge lines. A second water leg pump takes suction from the LPCS pump suction and discharges to the LPCS pump discharge and the RHR A pump discharge. The third water leg 3 pump takes suction f rom the llPCS pump suction discharges to the IIPCS pump dis-charge. The water leg pump motors are connected to the same electrical bus as the systems they service. The water leg pump electrical loads are classified as essential when the sys tems they service are not in operation. A pressure switch is provided in each RHR line, the LPCS and HPCS line. This pressure switch initiates an alarm when the monitored pressure drops below a prescribed value. The trip point for the alarm is selected to assure that the CSCS lines are maintai.ned in a full condition. The motor loads fi.e these pumps will be ' 6.4- 1
ZPS AMENDMENT 12 included in the sizing of the appropriate diesel generators af ter the design of the filling system is completed. These puap motor 1 cads have been estimated and are included in Table 8.5-1. The piping and components associated with. the " water leg pumps" are con-sidered a part of the gnre cooling system and as such are classified as Group B as defined in Appendix A.O of the PSAR. I Because of the addition of the " water leg ptmps" the momentum forces as-sociated with the dry pipe condition will not be considered a normal condition for the design of the core cooling piping and its supports and/or restraints, i llowever, because of the quick start characteristic of the core cooling pumps, severe traxsient mcnentum forces might be developed during pump start with the piping initially full as intended. The core cooling systems will be designed to withstand full pipe momentum forces caused by rapid fluid acceleration due to pump start. I l t To determine restraint forces and piping deflections and stresses two types of analyses will be made. The first analysis determines time dependent momentum ferees in liquid filled piping networks. Thse transients can be initiated either by opening or closing of one or more valves, by a prescribed change in source pressure, by starting a pump, or by pump power failure. The second analysis uses as input the force time histories generated in the first analysis. The second analysis determines time dependent piping stresses, piping deflections, and restraint loads in response to the loading function. i 1 k 6.4-2 4
ZPS AMENDHENT 12 6.5-1 (ZPS - April 9.1971. AEC Ques tion 6.1) _ QUESTION Discuss the criteria used to establish the core standby cooling system i (CSCS) pump performance requirements (head-flow characteristics). -Compare these criteria with those used for core standby cooling systems on earlier GE-BWR , designs, i.e., the 1967 product line plants. Provide results of analyses to show ! the sensitivity of CSCS performance to pump capacity and head performance. The analyses should include the case' for which the pump capacities are identical to those fuel used clad for a similar size 1967 INR and a discussion relating to acceptable peak temperatures. ANSWER The criteria to establish the head-flow requirements for the CSCS are discussed in Subsection 6.2 of the PSAR and the Topical Report NEDO-10183, page 9. With respect to the clad temperature criteria, sufficient margin was designed into the pump heads to ensure peak clad temperatures below 2300*F across the break spectrum even though the fragmentation threshold is 2700*F. the 1969 Itsystems. should also be noted that additional margins were involved in sizing Since cooling during blowdown and water remaining af ter blow-down were not fully defined at the time, conservative es timates in these areas resulted in even greater margins. A discussion of acceptable peak clad temperatures has been presented covering both the fragmentation aspects upon cooldown and the turnaround tempera-ture allowable for core spray and flooding in Pilgrim Amendment 14, Dresden Amend-ment 7-8 and NEDO-10179 "Effect of Clad Temperature on ECCS." Basically, for the exposure times involved here, we believe that a peak of 2700'T represents a fragmentation threshold. Core spray will turn around temperatures in excess of 2300*F and possibly as high as near the cladding melting point but tests in excess of 2300*F are not definitive. Flooding will turnaround temperatures in excess of 2300*F. future under the FLECHT Program demonstrateMore recent tests to be published in the near 2600*F. flooding turnaround temperatures of Although definitive threshold limits do not exist, turnaround temperatures well in excess of those anticipated in the reactor have been experimentally demon-strated for both core spray and flooding. An additional criterion affecting head-flow characteristics was used which in breaks. essence states that clad perforations will be prevented for all but large This was not a criterion for the 1967 product line plants and in many cases clad perforations occurred for intermediate breaks. As shown in the Topical Report NEDO-10183, given a break and single failure, no perforations occur for the 1969 P.L. except for liquid breaks in excess of about 1 square foot. 6.5-1
ZPS AMENDMENT 12 For' the HPCS, a criterion was used (in addition that it depressurize properly in conjunction with the low pressure systems) which prevents clad heating for breaks less than a 1 inch pipe when functicaing alone. This was done to ensure maintenance of level at rated vessel pressure for the more probable leaks that might occur over plant. life. Since 1 inch lines predominate, this provided a good basis fer such a criterion. This flow is also orders of magnitude in excess cal size. of leakage that would occur for cracks in large pipes approaching criti-The only dif fercnces in pump capacities between the 1967 P.L. and the 1969 P.L. are in the High Pressure Coolant Injection (Spray), and the LPCI sys tem. Figure 1.A sensitivity study showing the effect of LPCI pump capacity is shown in It can be seen that even if the pump capacity is equal to that of the 1967 P.L. the difference in peak temperature is only 150*F. Note that the peak temperature for the 3 LPCI case is only 1750*F. The curve assumes nucleate boiling coefficient until MCHFR is less than 1.0; then h=0 until plenum flashing; h=100 flooding.during lower plenun flashing until flashing stops and finally h=25 upon j The pump head characteristic affects the intermediate break temperatures. Sensitivity studies indicate that over the intermediate break spectrum the clad temperatures will var,v approximately 6*F per psi in LPC1 pump shut off head fer breaks in the .005 f t' size; 4*F per psi for breaks in the 0.01 f t 2 to .04 f t2 range and 6'F in the range of .05 f t 2 to .06 f t2 range. j pump head becomes increasingly less significant. The effect Beyond the .06 ft2 range, of changing the LPCI flow rate 207. at all heads is shown in Figure B-14 NED0-10329. The difference in temperature is about 200*F. of the Models Topical Report The HPCS flow is 4625 gpm at 200 psig and 1860 gpm at 1000 psi. For the same size 1967 plant, the HPCI flow is 4250 gpo at all pressures above 150 psi. The HPCS flow at high pressure affects the liquid break size for which no clad temperature rise occurs and the temperature of the intermediate break region. The break size for which no temperature rise occurs increases by about 0.05 f t' , per 1000 gpm of flow at 1000 psig. The HPCS flow sensitivity for temperature 1. shown in Figure B-15 of NEDO 10329, the BWR ECCS Models Report. It is shown that 2 for the worst in flow rate. break ize (0.1 f t ) the tecarature changes 200*F for 107 change However, increases in HPCS flow are not warranted nor are they simpic to accommodate. flow. The HPCS at low pressure is already in excess of the HPCI Any increase would have to occur at the high pressure end. However, it would be difficult to do this and still match the core spray flow limits at the {' low pressure end without unduly complicating the entire control system. Dies el size and all associated equipment is sensitive to such changes. These were con-siderations in attaining the current system which is optimum from an overall viewpoint. j
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Other sensitivity studies .for a Zimmer class plant regarding component capacities are discussed in detail in ND0-10329 as follows: i
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6.5-2
f:: i ZPS AMENDMENT 12
.4 (a) Figure B-11 Auto Relief Capacity (b) Figure B-12 !!PCS Capacity (c) Figure B-13 Core Spray Capacity In conclusion the entire 1%9 P.L. network must be viewed as a complete system end specific comparisons to the 1967 P.. can be misleading. As a total networ'.;, the 1969 P.L. more than meets all existing criteria and in addition, its overall reliability is superior to the 1967 P.L. systems. It should be judged on that basis.
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I i 2P3 l' AMENDMENT 12 6.5-2 (ZPS - April 9.1971. AEC Ques tion 6.2) _ QUESTION Provide the criteria that will be used is the design of auxiliary systems that are required to function during Icss-of-ccelant accident (LOCA). Identify the protection that will be provided for postula:ed failures of active or passive components within these auxiliary systens. ANSWDt - l General design criteria S3.26 and S3.27 poted below from PSAR Table 1.3-3 i apply to the " engineered safeguards." It is implicit that required auxiliary systems must be operable in order to assure operability of the required equipment. i S-3.26 Sufficient engineered safeguards shall be provided to assure a safe shut-down in the event of a loss-of-coolant accideat (IDCA) acccuipanied by the follovirg: A. Loss off-site ac power, and B. Earthquake, and C. A single active failure in any system that is required to function in order to accomplish a safe shut-down. S-3.27 Sufficient sys tems and/or components will be provided to assure that under conditions where the reactor coolant pressure boundary is intact and where no other accident exists, a single failure in any system that is required for a safe shut-down will not preclude that system's effective operat' ion with or without off-site ac power being available. The required auxiliary systems consist of standby ac power systems, de battery systems, CSCS equipment and equipcient space cooling systems and emergency service water systems. Sufficient independence auf redundancy is provided to assure meeting the general design criteria rited. The effects of postulated equipment failures on containment temperatures and pressures are shown on PSAR Figures, 14.6-10,11.6-11 and 14.6-12 for the design basis loss-of-coolant accident. During a loss-of-coolant accident (LOCA) certain core cooling sys tems, as described in Section 6.0, are required to operate. In support of the core cooling systems are a number of auxiliary service systems which must function in order for the various core cooling systems to sta t up or continue to function during sys tems are as follows: the period required. Those systems which are auxiliary to the core cooling 6.5-5
ZPS AMENDMENT 12 Standby diesel generators Switchgear (4kV, 480V and 480VMCC's) 125V.dc control power (batteries, etc.) Core cooling pump room switchgear IWAC and service water pump room ventilation and cooling equipment Control room and electric equipme'nt roon air conditioners I Service Water System Reactor Building CCW System As noted in Section 8.0, the Standt:y Diesel generators, switchgear and DC power sources are grouped in division so that each ESS division has its own source of 4kV, 480,120/208 VAC and 125Vdc power. This is a complete " split bus" physical arrangement of the electrical supply and distribution equipment. The loss of any single power source cannot disable more than one ESS division, thus leaving two safeguards divisions available at all times. The core cooling pump room ventilation systems are similarly oriented. ! This equipment is powered from the same switchgear, etc. as the pimp which it serves. Upon actuation or start up of a core cooling pump, the ventilation equipment associated with it is started. , Control room HVAC, service water and Reactor Building Closed Cooling water systems are arranged so that, under an IDCA condition, there are two (2) 100*/. redundant sets of equipment in each system. Redundant sets are powered from diesel generator busses which are in different divisions.1 Therefore loss of a power source or set of equipment will not negate the availability of 100*/. of any required service. Piping systems will be designed to meet criteria set forth in Section 3, Criteria S-3.26 and S-3.27 including amendments 4 and 6. 4 i t
.I a 6,5-6 1
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I 3 I
-l ZPS
'i
)
AMENDMENT 12 6.5-3 (2PS - April 9,1971, AEC Question 6.3) l _ QUESTION I j Identify the margin inherent in the CSCS performance capability to accom-J 1 modate the following situations:
- a. the occurrence of a LOCA during various operating modes, such as:
j with one recirculation pump in operation, natural recirculation and following a transient, (e.g., following a turbine trip) and
- b. a postulated delay in time to actuate the CSCS equipment.
ANSWER (a) The effects of a design basis accident with one recirculation pump in operation and during natural circulation have been investigated. Because of the reduced maximum operating power level, the consequences of a recirculation
] line break during either one of these modes of operation is significantly less j than during two pump operation. During one pump operation, the maximum attain able core power and core flow is 72% and 60% of rated respectively; of this 60%
flow, approximately 34% is due to natural circulation effects. In the event of a rupture of the operating loop, the pumped flow contribution is rapidly lost and the ' core flow will rapidly drop to the natural circulation value. This f natural. circulation flow will decrease as the water level outside the shroud drops due to the loss of water out the break. When' the water level drops below the top of the jet pumps, the core flow will stop. This sequence of events is I basically the same that occurs during the break with two pump operation; with the exception that the core flow af ter the rupture will be slightly higher due to the flow contribution from 'the one remaining pump that is coasting down. The core flow following a rupture of a recirculation loop during one pump operation is shown in Figure 2. It should be noted that the time and magnitude of lower plenum flashing is virtually unaffected by one pump operation since the thermo-dynamics of the system will remain approximately the same i.e. , lower plenum subcooling and the vessel depressurization rate are approximately the same as for two pump operation. Even though the amount of core flow following the rupture is less for the.one pump operation than trae two, nucleate boiling vill continue l
to exist in the core until the jet pumps uncover. This is due to the reduced i heat fluxes in the core (72% of rated). Figure 3 depicts the MCHFR during the l t rans!.en t . In Figure 4 the peak clad temperature history is shown for the
" worst" single failure condition, i.e. failure of the LPCS/LPCI diesel. The analysis incorporates the use of the most recent core spray cooling model and a constant lower plenum flashing heat transfer coefficient of 100 B/hr-f t 2 *F out to the point of which the core uncovers. The resulting clad peak temperature ;
of 1280 F is significantly less than the DBA during two pump operation (SEE { Question AEC 6.12 of this amendment); in fact no fuel rod perforations result for this accident. The peak clad temperatures following a DBA at natural l 6.5-7
ZPS AMENDMEhT 12 circulation will be less than for one pump operation since the maximum operating power will be only 62% of rated and the core cooling during the blowdown will be identical to that shown in Figure 3. It can therefore be concluded that the consequences of a design basis accident during either one pump operation or natural circulation is less ; severe than during two pump operation. (b) The effects of postulated delays in times to actuate the CSCS equip- l ment is shown in Figure 5 for the design basis accident. The condition analyzed i
' is for single failure of the LPCS diesel. This represents the " worst" single failure.
Delays of 30 and60 seconds in operation of both core spray and LPCI l pumps were considered. The results indicate that there is considerable margin inherent in the CSCS performance. Even a 60 second delay in operation of both l the core 1880*F. spray and the LPCI pumps results in a peak clad temperature of only 4 I e I 1 i e f 1 6.5-8
l
!I J AMENDMENT 12 I l 1 1
1 l 1 100 , , , , , , f80 W -- J' 8 CORE FLOW DUE TO
- l 60 - NATUR AL CIRCULATION
~
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- f. -
! 0 l 0 5 10 15 20 TIME AF'TER ACCIDENT 45EC
, 1 t WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSr5 REPORT FIGURE 2 CORE FLOW FOLLOWING A RUPTURE 6.5 9 0F A RECIRCULATION LOOP DURING ONE PUMP OPERATION
-1
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ZPS ' AMENDMENT 12 6.5-4 (ZPS - ~ April 9,1971, AEC Question 6.4) QUESTION List the postulated piping break sizes, types and locations specifically analyzed for' the un. H. Zimmer facility.
; g. .
ANSWER _. For classifying the consequences of a loss-of-coolant on a BWR it is convenient to classify the breaks according to the location of the penetration on the reactor vessel. The break types will fall into one of three categories. These along with the lines that fall into these categories are as follows: j I. l STEAM TYPE BREAKS These are breaks in which the reactor vessel penetration is exposed to the steam regions inside the vessel.
- 1. Steam Lines
- 2. RCIC Injection i
i
- 3. Some Instrument Lines II. STEAM / LIQUID TYPE BREAKS i
.l These are breaks in which the reactor vessel penetration is either ex- l
( i posed to the two-phase regions inside the vessel or to regions which are exposed ! f to ligoid, but are near the water level and would therefore turn into steam l breaks very shortly after the break occurred. These are located above the core. j
- 1. FEEDWATER Lines
- 2. CORE SPRAY Lines l
j l 3. LPCI Injection Lines j
~
\
4 l SOME INSTRUMENT Lines III. LIQUID TYPE BREAKS These are breaks in which the reactor vessel penetration is well below the vessel water icvel, and below the top of the core.
- 1. Recirculation Pump Suction Line
- 2. Recirculation Riser Line 1
6.5-13
- ~)
ZPS ( AMENDMENT 12 I
- 3. Drain Line For a given size break'the peak clad temperatures will be higher the l lower the line penetration is located on the vessel, i.e., the peak clad temper-j ature for a given size break will be higher for those lines in Category III than in Category II and those in II will be higher than those in I. In demonstrating
.l '
the performance and capability of the CSCS, recirculation line breaks are ana-lyzed size. break since these will result in the highest peak clad temperatures for a given The rupture and consequences of a main steam line have also been analyzed. The results of these analyses are presented in Section 6.0 of the PSAR.
. The specific line sizes, etc., analyzed are listed in Table 1.
I l By analyzing breaks in the main steam line, the effects of all other
. steam type breaks are covered. For liquid type breaks, the spectrum analysis performed on the recirculation line break covers the effects of all other type liquid breaks such as the RHR suction and return lines, and recirculation riser lines.
The peak clad temperatures for the stenm/ liquid type breaks will be less than for the comparabic size liquid breaks. This was shown in part in Millstone i. Amendment Breaks in the14cort in which the effects of various size feedwater breaks were analyzed. spray line and LPCI injection line will result in lower tem-peratures than for comparable liquid breaks. The size of either the core spray line orisLPCI spray line break achieved. is such that the core is reflooded before rated core Hence, the effect of either of the breaks on the core spray distribution is irrelevant since the spray cooling mechanism does not even enter j into determining the peak clad temperature for the accident. It can therefore be concluded that the consequences of all possible l breaks has been covered by the analysis presented in the PSAR. a I u. 6.5-14 '
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a ZPS APEKDMENT 12 6.5-5 (ZPS - April 9, 1971, AEC Question 6.5) O!!ESTION Discuss the study undertaken showing the advantages and disadvantages of bottom versus top core injection of the low pressure flooding system. We are not able to agree that the proposed top injection mode has been justified without detailed engineering evaluation of all related aspects including updraft effects on core spray water. (See Questions Nos. 6-10.) Will the integrity of the injection lines of the lowpressure flooding systems be monitored during operation in a manner similar to the core spray linest f ANSWER l Before selection of the proposed system several alternates were ers-ined. { The mechanicci problems were examined by making layouts of five different ways of injecting water into the shroud as shown in Figure 6 and described briefly i} below. The advantages and disadvantages are shown in Table 2. These conclusions were arrived at from layout studies by experienced mechanical designers ammare of what detailed design of each concept would in.olve. Case I - direct injection into bottom plenum Case II - injection through a riser pipe from a nozzle in the bottom j - head leading into the shroud region outside the fuel cJiannels. Case III - penetrate vessel above the core and downward injection into l a riser pipe within the annulus between vessel and shroud and into the bottom plenum. l-j Case IV - injection into a riser pipe from a nozzle just above the
; bottom plenum and into the core region outside the channels.
Case V - penetrate vessel near the top of the core injecting directly into the shroud outside the core channels. It was determined from a systems analysis that to increase the overall t? liability significantly, separate injection points for each pump were required to eliminate both the single valve common to all the LPCI pumps and the loop selection logic. This precluded injection into the recirculation loops dvem-selves. l Direct in'ection into the lower head either from an external pipe or through an ' intern.N riser were eliminated early because rupture of the LPCI injection pipe wou.1 cause high pressure loadings across the lower shroud, core plate, and guide tubes. This arises because the bottom plenum is a pressurized, 3 subcooled region and pressure drops rapidly upcn being exposed to a break to the outside. This was considered a sufficiently serious disadvantage that far out- , j l I i 6.5-16
r ZPS l l AMENDMENT 12 weighed any mechanical advantages offered. These considerations reduced the - mechanical choices to Cases II, IV and V. , i Phenomenologically it is important to note that either injection into the shroud or into the bottom plenum can be made to work. There are no over-whelming phenomenological advantages for either one. The important point is that only with shroud injection could the logic system and single valve be eliminated
'l i .without incurring the very serious mechanical loading disadvantages commensurate with direct injection into the bottom plenum.
However, a few items are of interest. For bottom injection no core cooling occurs until the bottom has been filled thus introducing an inherent i
' time delay in the LPCI system. Injection into the shroud results in immediate cooling of the fuel channels since the water rises to the top of the channels within a few seconds.
rate of the core. CalculationsThis provides a heat sink which will attenuate the heatup indicate that if significantly reduced heat j transfer during blowdown is assumed to occur to the point where the peak clad
)' temperatures effect would approach would reduce 2400'F before being flooded, the outside cooling this by 200*F.
predicted, i.e., However, for the temperature being actually less than 2000'F, the outside cooling effect reduces the tem-perature before flooding by 50*F. Therefore, the outside cooling effect tends j to attenuate any uncertainty in the blowdown heat transfer phase of the accident I and is a distinct advantage of shroud injection. f j Opdraft effects due to LPCI water entering the vessel are not severe in should occur.because either case the amount of water injected is so large that no bulk boiling This is discussed more fully in ABC Question 6.10. The amount of experimental data which exists for bottom flooding is more extensive than for shroud injection. However, no credit is taken for core cool-ing until the core is being flooded even with the shroud injection. Thus the data for bottom flooding is applicable. One disadvantage of shroud injection is that because of the colder heat sink sooner. outside the channels the level swell drops below the top of the active fuel However, as discussed under long term couting, this does not occur until 3 hours after the LOCA. Thus the operator has more than enough time to put the LPCI pumps in the shutdown cooling mode one into each recirculation loop. Con-nections quired long exist to both recirculation loops for this purpose. Only 1 pump is re-term. In this state long term cooling is identical to the 1967 P.L. and the temperatures are such that the core can remain in that state indefinitely. In conclusion, the question of shroud injection versus bottom injection cannot be viewed in a restricted sense. All the aspects, mechanical and phenome-nological must be weighed together. Shroud injection becomes the most practical method to accomplish the objective of increased reliability through elimination of the single injection point and loop selection logic. It also adds the least 6.5-17
..d
l 2PS l AMENDMENT 12 hardware within the vessel for the entries which do not present blowdown force problems in the event of external pipe failure. The integrity of the internal interconnecting pipe will be monitored if a way can be found to accomplish this. However, it will only be possible to de-tect large leaks the size of which will depend on details idiich are not yet 1 available. i I f l
)
I h I
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l 1 l l l l } 6.5-18 -
i 2PS AMENDMENT 12 TABLE 2 ARRANGEMENTS STUDIED CASE ADVANTAGES DISADVANTAGES l I 1. No internal components. '1. High internal AP for breaks in l 2. Simple mechanically. ._ any of the 3 LPCI lines.
- 3. Low Pressure drop. 2. Core is not refloodable for Pipe break.
Il 1. Permits core reflooding. 1. Higher pressure drop. 2 Space constrictions and layout difficulties. 3 High irradiation effects in riser pipe. 4 Piping underneath shroud diffi-
; cult to design due to high loads and s tiffness.
. III 1. Permits care reflooding. 1 High differential expansion dif-ficult to accocunodate in con-stricted space.
2,. High pressure drop
- 3. Proximity to core spray piping.
l 4 High internal AP for LPCI breaks. I
! IV 1 Permits core reflooding. s 1. Same as Case III but less diffi-l 2 Some gain due to earlier "" * ,i cooling but not used in 2. High Pressure. . **'8"'
- 3. Proximity to core spray piping.
V 1. Low pressure drop, 1 Proximity to core spray piping.
- 2. Permits re. flooding.
- 3. More room for piping.
- 4. Differential expansions accommoda ted easier.
6.5-19
AMENDMENT 12 l CASE I CASE !!
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1 WM H. ZIMMER NUCLEAR POWEA STATION PRELIMINARY SAFETY ANALYSTS REPORT l W I i FIGURE 6 i FIVE DIFFERENT WAYS ! 0F INJECTING WATER i j 6.5-20 INTO THE SHROUD j
p a I
\
ZPS l i AMENDMENT 12 6.5-6 (ZPS - April 9,1971, AEC Question 6.6) QUESTION Provide the physical separation criteria for 'the CSCS equipment (both
! inside and outside of primary containment).
ANSWER j ,, ;
- l The ' separation criteria for the Core Standby Cooling System (CSCS) )
l equipment outside the primary containment has been described in the answer to - AEC Question 5.3, dated October 13, 1970. In Addition, we have the following information: 1 j
-4 Redundant CSCS piping will be routed in separate pipe chases outside I. the primary containment, and will be physically separated as much as possible
' within the primary containment. Pipe whip will be kept to a minimum by the i use of check valves, normally closed motor operated valves, and pipe whip
.j . restraints.
P1, e whip will not be permitted where one CSCS line could damage l' anothe r. For more information on pipe whip protection, refer to the answer I to AEC Question 12.25 dated April 21, 1971. Electrical equipment forming } j part of core standby cooling syst6ms will be separated in accordance with IEEE 279. -]
)
I 4 4 q l. 1 4 I I l i f 1 I 4 4 6.5-21 \ . __J
, m i i u [ ) I "p ZPS f AMENDMENT 12 6.5-7 _(ZPS ~ April 9,1971, AEC Question 6.7) QUESTION Provide a description of the model and the results of an analysis to l de nonstrate that the core would remain adequately cooled following the initial
; LOCA transient assuming that only flooding systems are available. The concern is for I:he long-term effect.
ANNER
)
f 1.0 Summary I
-l An analytical study was conducted to determine rLe capability of the LPCI system accident. !;o provide adequate long-term cooling follM.ng a design basis
)
Long-term cooling is defined as cooling after the initial thermal i i transient has been ' terminated until the fuel can be safely removed. The analysis which is supported by experimental data shows that maximum fuel f cladding temperatures will be maintained less than 1000*F by the LPCIS alone j for the long term period.
; This temperature is sufficiently low that the fuel could remain in this condition indefinitely. j 2.0 Short Term Cooling Effectiveness
- Immediately upon flooding the core, sufficient level swell takes place 'to cool the fuel bundle to saturation over the entire length. An l example of this behavior is shown in Figure 7. In this figure, the water level inside the shroud represents the water level in the average power fuel bundle.
Water level in the highest power assembly is actually higher and thus the predicted temperatures for this short term period are conservative . i In the lowest power fuel assembly, the water level will not be as high and
! after the initial period, a slow heatup will begin as the voide collapse and the top few inches of this low power assembly begin to uncover in l about 3 hours. i i
From this point on, long term cooling becomes a consideration. The LPCI after this time period will be injecting through the RHR heat exchangers into the recirculation loops rather than into the top of the core shroud . Thus, the phenomena after this period are the same for the earlier product line plants. , j i 3.0 Duration of Level Swell Cooling Phenomenon ( and analytically over a wide range of conditions.The 1cvel swell cooling p k of the 8. Figure icyc1 swell phenomenon during the long term condition is shown inThe model ul I i 6.5-22 ! _ {
ZPS AME:NDME:NI 12 l i The fuel bundle is represented by 12 axial nodes. For each node f '
' W N* N-1 +
fg, 1) E **** b = steam flow leaving node N, gy = dmy heat for, node N, and hf= enthalpy change due to vaporization The solution to equation 1 defines the steam flow at any position up the bundle, The void fraction at each elevation is given by 1 1 N + b -1 yg 2 1
,= y 2)
Where: Vg = specific volume of steam, A = channel cross-sectional flow area, VN = bubble rise velocity j' The bubble rise velocity is a function of pressure h
, and void fraction and is given by the Wilson correlation.kl)ydraulic diameter 3 Equations 1) and 2) define the void fraction up the channel. Level ; is determined by the boundary condition on the collapsed liquid level. That ! . is, 3 3 I (1-aN) (hN ) Pf =C 3pg 3)
N=1 Where : hN = length of node, pf = density of saturated water, p = density of subcooled water, 6.5-23
ZPS AMENDHENT 12
\
l C
= collapsed level of subcooled water, l j = node number at top of mixture.
included by iteration.The effect of subcooled water coming into the bottom of the bu the subcooled water to saturation temperature isThat is, the amount of bundle P3 8 W) @f - h,) O Where h, = enthalpy of inlet water Equations 1), 2), 3), 4) and the Wilson corre:ation are solved simultaneously for the five unknowns W'8 N N' N' " II"" I' " I' "" #S The above model can be used to predict the duration of level swell the swollen level covers the top of the active fuel. cooling by assum The model has been verified9.in Figure this manner by comparison to experimental data (2,3) as shown in ' model. The excellent agreement verifies the adequacy of the level swell
\
As is apparent from Figure 9, the reactor pressure is of prime importance to the long volume at increased term cooling capability due to the reduced void pressure. ' be nearly equal to containment pressure.The quasi-steady state reactor pressure will valves remain open for all reactor pressures.This is true because the relief by LPCI will quench any steam generated in the core.Also, the cold water injected As shown in the PSAR, the long term containment pressure is never
. Thus, a reactor pressure of 30 psia can be used in the calculat power for Figure the 3010psia shows the long term swollen level as a function of bundle condition.
.) mixture are cooled to saturation. The portions of rods covered by two-phase '
.j ,
I l ) i! l 6.5-24 l
i > . ZPS AMENDMENT 12
~
4.0 Heat Transfer Analy, sis for Exposed Rods The portions of fuel rods not covered by mixture are cooled by.
' convection to the steam generated below. the two-phase mixture. Becaus'e of the relatively small amount of steam generated, the flow is laminar which results in a Nusselt number in the rod array of at least 6.(4) It is expected I
that this value is somewhat low, but can be used to assure a conservative re-i sult. Taking no credit for the 'ncrease in steam conductivity with tempera-ture results in a convective fi :oefficient of at least 1.74 Btu /hr-ft 2.ye, j,a The temperature rise of the stea. s given by 3 1 12
=t-r b QI i=1+1 l (Tout - Tin) " . . .. wjP v i
, 5) '
Where Tout = Steam temperature leaving bundle Tin = Steam Temperature leaving jth Node (top of mixture) C = Specific heat of steam ; P 4 Rod temperatures are then de'termined from I i f 1. 74 (TR~2 out + in}3 = 6 + 1 i s
- 6) ,
i Where: ! i l TR = Rod surface temperature. f
- i
#pg = Surface heat flux at the first mode.above the two-phase mixture.
Note that since the surface heat flux is based on the condition just , above water 1cvel and the steam temperature is based on an average for the ' total exposed portion of the rods, further conservatism is introduced. More refined calculations can be done by considering a continuously increasing sink temperature and a continuously decreasing surface heat flux. The results of the calculations are shown in Figure 11. Note that maximum temperatures are always less than 1000'F and, thus, the rods can remain 't in this condition indefinitely without resulting in core damage. J 6.5-25
ZPS
) AMENDMENT 12 REFERENCES
- 1. Wilson, J. F., et.al. , "The Velocity of Rising Steam in a Bubbling .
, Two Phase Mixture" ANS Transactions, Vol. 5, No.1, pg.151 (1962). .
s
$ 2. Duncan, J. D. and Leonard, J. E. , " Response of a Simulated BWR j Fuel Bundle Cooled by Flooding Under Loss-of-Coolant Conditions", .
)
, GEAF10117, December 1969. >
{ l
;- 3. Shraub, F. A. and Leonard, J. E* ., " Core Spray and Core Flooding Heat 1 Transfer Effectiveness in a Full-Scale Boiling Water Reactor Bundle",
; APED 5529, June,1968.
l t 4. Kays, W. M. , " Convective Heat and Mass Transfer", New York: McGraw-l Hill, 1966. 1
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, e, auntvuu nai ov,3 uvu n.21 o~o, WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSIS REPORT FIGURE 11 6.5-31 LONG TERM PCT STEAM COOLING 1
- .. _ - - _ - _ _ __ l
ZPS AMENDMENT 12 6.5-8 _(ZPS - April 9,19 71, AEC Question 6.8) QUESTION { l
)
inadvertent operation of the CSCS during variousThemodes results of, an evaluation of the potential nuclear reactivity incident due to - actuation uation. of the high pressure core spray system should be included in the eval-occurrence. Describe the protective features. to be provided to preclude such an ANSWER The low pressure core cooling systems can only inject water into the
. core while operating at low pressures and therefore at low power. The largest reactivity insertion for inadvertent operation of low pressure core cooling would occur with the reactor operating at heating power of approximately 5% of rated map (PSAR Figure 3./-1).which would result in about 25% core flow as shown the LPCS pump and the HPCS pump all came on and simultaneo into the reactor vessel.
(
~
The reactivity inserted by such action is of little 5.5% of rated power. consequence and it is estimated the reactor power would change The change would be slow and inherently self regulating and is small Subsection compared to other reactivity insertion transients analyzed in PSAR 14.5. into the core and the HPCS system alone is capable of injecti The HPCS will be actuated pressure signal. by either a low reactor water level signal or a high drywell ; fore inadvertent operation is hot likely.Each of these signals meetThe the re- single f Furthermore, as described below, the effects which would occur due to inadvertent operation are so mild that in our judgement additional protective features are not requ* red. Analyses were made of cases of inadvertent HPCS operation at the 105% of nuclear boiler (NB) warranted power condition and at 30% of NB warranted power with core recirculation flow at about 45% of design flow. The later case is representative of the power level required to clear the recirculation valves NPSH limit at the maximum flow on the recirculation bypass valves, just prior to the opening of the main recirculation flow control valves. System depressurizes the reactor and causes aThe on the reactor are quite sfid. drop in reactor effects At the 105% of NB warranted power level the anal-ysis assumed an HPCS capacity equal in value to about 10% of NB warranted feed-water flow at a temperature of 70* F. flow that the HPCS pumps can inject into the reactor atThis is an amount in excess of the a the normal operating 6.5-32
l
/
l 2PS I AMENDMENT 12 l l l reactor pressure of 1040 psig. The analysis at 30% power assumed EPCS flow of. 12.5% of warranted feedwater to account for the increase in HPCS flow that would .I result from the lower reactor pressure compatible with the Iwer power leve.l.
' The results of the analysis with the reactor initially at the 105% of NB warranted power level are shown on Figures 1 and 2. De introduction of the cold HPCS flow into the upper plenum region of the reactor drops the vessel pres-sure from a value 1040 psig to approximately 1024 psig within 25 seconds. Reactor power drops slightly from the initial 105% value to about 98% within 40 seconds. 1 No threat to the fuel exists during the t'ransient as MCHFR rises by a small amount from its initial value of 1.9. Electrical load drops off to about 93% .rhich is l
less than the reactor power as a portion of the reactor power must heat the HPCS 4 flow introduced to saturation temperature. The feedwater control system regu- I lates the reactor water level during the transient. Feedwater flow settles to a value below the steamflow leaving the vessel by an amount equal to the HPCS ! flow. A slight increase in reactor water level results because of the action of I the three element feedwater control system responding to the mismatch between steamflow and feedwater flow. The results of the analysis with_ the reactor initially at the 30% of NB warranted power level are shown on Figures 3 and 4. Although the HPCS flow is slightly larger than the previous case, because of the lower initial reactor pressure (9602 psig), the reduction in vessel pressure is somewhat s= miler.
'The reduction in reactor power is more gradual than the previous case j
and occurs over a longer time span. By 50 secs, reactor power has dropped from
' its initial 30% value to about 18%. ne slow drop in power from 15 seconds on j
is due to the lower feedwater flow and the decreasing subcooling which results { both from low feedwater and from lower pressure.
' As vessel steam flow has dropped below the total value of HPCS flow and feedwater flow, reactor level will slowly increase until such time that the '
{ high vessel water level trip point is reached. This trip will cause closure of
- the stop valves on the main turbine and on the feedwater turbine (s) and will cause closure of the HPCS valves. Since the transient is mild and gradual, it is expected that the operator would have had time to assess that inadvertent HPCS actuation had occurred and would shut down the HPCS system before reactor j level rises to the trip point.
4 1 I 6.5-33
/
AMENDMENT 12 a l i l i , .. . 2540 MWT l [/[' ' INA0VERTANT ifCS j g, ACTUATION
! e NEUTRON FLUX
'l
- AVG SURFACE HERT FLUX
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AMENDMENT 12
=
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#* ACTUATION i HPCS FLOW f/ t F HTD FW)
N R SENSED LEtELfINCHES) a CORE PRESSURE RISE (PSil
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WM. H. ZIMMER NUCLEAR IOWER STATION PRELIMINARY SAFETY ANALYSIS REPORT FIGURE 2 CONSEQUENCES RESULTING FROM INADVERTENT OPERATION OF HPCS 2540 MIT 6.5-35
AMENIPfENT 12
=
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- RCTUATION i i EUTRON FLUX s RVG SURFRCE WRT FLUK e CORE INLET FLCW
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- OF HPCS 731 fMT
_-______-______-____________-____r,.s.u ---
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WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY SAFETY ANALYSIS REPORT I FIGURE 4 CONSEQUENCES RESULTING FROM INADVERTENT OPERATION OF HPCS 731 MWT 6.5-37
I ZPS ' AMENDMENT 12 6.5-9 _(ZPS - April 9,1971, AEC Ques tion 6.9) QUESTION We understand that features have been incorporated in the recirculation , system throttle-valve to prevent valve motion in the event of e LOCA. Provide ; preliminary drawings to demonstrate how this is accomplished and indicate to ' what extent these features meet criteria appropriate for an engineered safety ) feature. If these features are not designed as engineered safety features, pro- j vide analyses of a LOCA assuming that the valve closes during the accident. { ANSWER It now appears that the valve throttling characteristics can be made such that the flow coastdown flow will not be affected, that is the flow coasts . down more rapidly than the valve throttling even if closure is started at the ! time of the accident. Therefore it will not be necessary to add any features to prevent valve closure fo13owing either a recirculation line break or a steam line break. All details in this area have not been worked out; final results I will be available before the FSAR. The objective will be to not let the valve closing affect the coastdown flow. i 5 l 6.5-38
ZPS AMENDMENT 12 6.5-10 (ZPS - April 9,1971, AEC Question 6.10) QUESTION i Trom the description of the methods used to inject the water from the flooding and core spray systems, it appears that more steam can be generated with both flooding and spray systems in operation than for spray operation alone. Accordingly, provide results of analyses and tests that show how the effects of
~
j; the additional bution steam of the core generation spray wa:er. have been censidered in evaluating the distri-
$ f< . , ,
_ ANSWER i
.Y together.Additional steam is not generated with LPCI and core spray operating P
)
; 140*F. The LPCI water will enter the shroud in the subcooled s. tate at about !
;y Most of the coolant flow passes around the channels and down through the bundle bypass orifices into the bottom plenum. The high flow races coupled with the low heat capacities of the fuel channels does not permit any steam genera-y tion and in fact only heats the wate.r from 140*F to 149'F, even if the stored heat in allboiling.
nucleate the channels (approximately 36,000#) were to be absorbed assuming Thus the LPCI water enters the bottom plenum still subcooled. , l
', The only guide heat source in the bottom plenum is the heat capacity of the shroud and tubes.
, by the thermal diffusivity of the metal. Assuming nucleate boiling, the rate of heat addit , Calculations indicate that it would take nearly an hour to release all the stored heat in the bottom plenum. Even assuming no addition of fluid, all the heat within the bottom plenum including , the vessel walls would heat up the LPCI water to only 217*F compared to the boiling point of 240*F at 25 psia. Hence LPCI cannot add more steam. . Therefore it follows that the worst possible case for additional steam generation to occur would be for complete failure of the LPCI. In such a case the water would remaining occur from th,e in the bottom stored heat in of thethe vessel would be saturated and boiling metal. Even in this case the upward velocity of steam is negligible. Assuming the entire spherical head is full of saturated water, detailed steaming calculations of this nature which included all the components within the head were conducted several years ago for the Browns Ferry class of reactors by BMI.
- 8. These were reported in BMI-1841, Table 6, page This is reproduced below with a column normalizing to flow per bundle added.
6.5-39
/ l 1
i ZPS AMENDMENT 12 i TIME BOIL OFF BOIL OFF STEAM AVAILABLE* _SEC RATE PER BUNDLE PER BUNDLE 3 f/sec f/sec #/sec 30 13.0 .017 .0085
.f 35 12.43 .016 .0080 j 55 10.64 .014 -
.0070 105 8.02 <f. .011 .0055' J[. 205 5.75 -
.0075 .0038 f 305 4.70 , .0062 .0031 J' 405 4.07 .0054 .0027 505 3.62 '
.0047 .0024 605 3.26 .0043 .0022 705 2.97 .0039 .0020 805 2.72 .0036 .0018 985 2.33 .0031 .0016 The maximum amount of steam generated available to the core is only 0.0085f/see per bundle initially and 0.0031#/see in 5 minutes. The maximum steaming rate for core spray with no accumulation occurs in about that time period when sufficiently overall good heat transfer has been established (See
', APED 5529 Figure 45 a,b,c,d) . Assuming the core spray is saturated when it enters the hot bundle and that only the minimum core spray enters, the maximum steaming rate is 0.078#/sec/ bundle. The core spray distribution is routinely tested at 0.095f/sec/ bundle and deleterious effects are usually not noted until 0.12 to 0.13#/sec/ bundle. The combined maximum steaming rate is only 0.081#/see/ bundle. This is below the routine updraft tests and well below the threshold where core spray distribution will be adversely affected. 5 I 1 i I l 1 l i i 4
*0nly about half of the steam can go through the core since the resistance from the bottom plenum to the region outside the shroud is split about equal through the core and through the jet pumps. H l
6,5-40
-- --__- -U
'. I ZPS l AMENDMENT 12 i
f 6.5-11 3 i (ZPS - April 9,1971. AEE Question 6.H) j QUESTION l Tabulate the reactor vessle volume (cubic feet), as a function of eleva- l tion (feet) to the bottom and top of the active fuel regions of the core. ' The volumes reques ed are tabulated below and in the attached Figure 12: INSIDE SHROUD: VOL LEVEL 0 0 266 3.69 1516 9.27 2871 17.4 4223 29.4 5075 35.8 5375 43.7 OUTSIDE SHROUD:
, VOL LEVEL 0 9.27 2810 35.8 4753 43.7 i
5211 46.5 6631 52 7916 54.3 9970 60.23 11540 69.31 6.5-41
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I i ZPS i AMENDMENT 12 6.5-12 (ZPS - April 9, 1971, AEC Question 6.12)
.l QUESTION l
Provide the results of your evaluation of the c6re cooling system per-formance considering the use of the most recent CE core spray cooling model for the spectrum of postulated piping breaks. Include the results of parametric studies of the sensitivity of the results to the assumed parameters, including fuel channel wetting times, heat transfer coefficients during the transient (even
' for those cases where the coefficient is assumed to be zero), and times and magni-tudes of lower plenum flashing for the design break of the large recirculation lines.
,A, NSWER The most recent change in the core spray cooling model will not affect the predicted peak clad temperature for small breaks because the core becomes reflooded before rated spray flow is achieved. Therefore, the spray cooling mechanism does not even enter into determining the peak clad for small breaks and the temperatures
. presented in the PSAR are applicable. The new core spray cooling model will only affect the large breaks. In Figure 13 the peak clad temperature history is shown (solid line) for the design basis accident and the " worst" single failure of the LPCS diesel. his analysis includes the new core spray cooling model and conser-vatively assumes a heat transfer coefficient of 100 B/hr-ft2 *F from the time of
.i lower plenum flashing to the time at which the core uncovers (30 seconds). Actually because of the high flow rates that occur during lower plenum flashing, the heat transfer coefficient will initially be similar to that presented in the PSAR and gradually decreasing as the core flow decreasing as the core flow de-creases. The channel wet time is calculated as described in topical report NEDO-10329. As requested, the effects of certain assumptions in the analysis is also shown in Figure 13. In particular the effect of an earlier core un overy (20 seconds instead of 30 with a lower plenum flashing heat transfer coefficient of 100 B/hr-f t 2 ..F is shown. The effect of channel wet times is shown befow in Table 3. The effects of premature CHF has been described in NEDO-10329 Supplement
- 1. It can therefore be concluded the consequences of a DBA is not sensitive to heat transfer assumptions used during the blowdown.
6.5-43
- - ~
i ZPS i AMENDMENT 12 ' TABLE 3
^ PEAK CLAI i_. HPERATURES EFFECT OF HEAT TRANSFER ASSUMPTIONS DURING
_ DESIGN BASIS ACCIDENT
, pr 2 LPCI + HPCS 1
- -i c' Lower Plenum Flashing Heat Time of Core Transfer Coeff. Channel Eet Time Ur.covery
___(B/hr-ft2 - F) (sec) Calculated Calculated + 2 min. 0 - 1924*F 1926*F 100 20 1745'F 1764* F 100 30 1550*F 1587'F i I 4 l l 6.5-44 1
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.1 ZPS AMENDMENT 12
, INSTRUCTIONS FOR UPDATING YOUR PSAR i VOLUME 3 l All changes have been indicated by a vertical line and the Amendment I
/ Number (12) in the right margin of the page.
i
} 1. At the beginning of Volume 2 remove and destroy Pages 11, 15, 17, a
18, 19 and 20 and replace with amended Pages 11,15,17,18,19 and [
~
- 20. After Page 20 insert new Pages 21 and 22.
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r I m i ZPS AMENDMENT 12 j YOLUME 5 TABLE OF CONTENTS, (Continued) PAGE D.0 APPENDIX D.O - QUALITY CONTROL SYSTEM TABLE OF CONTENTS '
, , D.0-1 D.1 INTRODUCTION D.1-1 D.2
- CINCINNATI GAS & ELECTRIC CO. QUALITY ASSURANCE PROGRAM D.2-1 D.3 SARGENT & LUNDY QUALITY ASSURANCE SYSTEM D.3-1 D.4 GENERAL ELECTRIC QUALITY SYSTDI FOR BWR NUCLEAR STEAM SUPPLY PROJECTS D.4-1 D.5 KAISER ENGINEERS INC. (COESTRECTORS) QUALITY ASSURANCE -
QUALITY CONTROL PROGRAM D.5-1 D.6 APPLICABILITY OF QUALITY ASSURANCE PROGRAM TO COMPONENTS, 3 SYSTEMS AND STRUCTURES
, D.6-1 . . E.0 APPENDIX E.0 - STATION ATMOSPHERIC RELEASE LIMIT CALCULATIONS E .1-1 !
F.0 APPENDIX F.0 - CONFORMANCE TO AEC DESIGN CRITERIA TABLE OF CONTENTS F.0-1 F.1
SUMMARY
DESCRIPTION 1 F.1-1 F.2 CRITERION CONFORMANCE (THE "70" CRITERIA) F.2-1 F.3 12 CRITERION CONFORMANCE (THE "64" CRITERIA) F.3-1 , G.0 APPENDIX G.0 - STATION NUCLEAR SAFETY OPERATIONAL ANALYSIS TABLE OF CONTENTS G.0-1 , i G.1 ANALYTICAL OBJECTIVE G.1-1 G.2 APPROACH TO OPERATIONAL NUCLEAR SAFETY G.2-1 G.3 METHOD OF ANALYSIS ' G.3-1 . C.4 DISPLAY OF OPERATIONAL ANALYSIS RESULTS l G.4-1 i 11 l 1
w-ZPS' I AMENDMENT 12 LIST OF 2PS, OCTOBER 13. 1970 AEC QUESTIONS, (Continued)
'AEC QUESTION RENUMBERED VOLUME NtNBER , AS QUESTION PAGE OF PSAR 5.5 5.2.5-1 5.2.5-1 2 l 5.6
~
5.0-1 ,. # 5.0-1 ' "'C' l ' ,'- 2- i 5.7 5.2.5.1-1 g,'. } , ',
, 5 . 2 . 5 .1-l '. 2 5.8 5.2.3.1-1 5.2.3.1-1 2 5.9 12.3.2.5-1 12.3.2.5-1 4 5.10 5.2. 3.8 -1 5.2.3.8-1 2 b2 ] i i 12.1 12.2.1.1-l' 12.2.1.1-1 4 j
,' 12.2 . 12.3.1-1 12.3.1-1 4 i
! s I i 12.3 12.2.1.1-2 12.2.1.1-2 4 il 12.4 12.2.2-1 12.2.2-1 4 12.5 12.2.2.5-1 12.2.2.5-1 4 !
3 1 s 12.6 12.4.4.1-1 12.4.4.1-1 4 7 ,! 12.7 12.4.4-1 12.4.4-1 4 5 12.8 12.3.2.3-1 12.3.2.3-1 4 12.9 12.3.2-1 , 12.3.2-1 4 12.10 12.3.2.2-1 12.3'.2.2-1 4 12.11 12.3.2.2-2 12.3.2.2-2 4 12.12 12.4.3.3-1 12.4.3.3-1 4 l 12.13 12.3.3-1 12.3.3-1 4 i 12.14 12.2.2.4-1 12.2.2.4-1 4 12.15 12.3.6-1 12.3.6-1 4 12.16 12.2.2-3 12.2.2-3 4 ,f 12.17 12.2.2-2 12.2.2-2 4 12.18 12.5.1-1 12.5.1-1 4 h 12.19 12.3.4.2-1 12.3.4.2-1 4 12.20 12.3.2.3-2 12.3.2.3-2 4 . 12.21 12.2.1.1-3 12.2.1.1-3 4 15 '9 '
[:
\
L. ZPS AMENDMENT 12 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS
'i AEC QUESTION RENUMBERED VOLLHE NUMBER AS QUESTIDW PAGE OF'PSAR 2.12 , 2.2.3-2 2.2.3-12 1
- t 2.13 .
2.3.2.1-2 ,_ 2.3.2.1-2 1
..;>a .
; 2.14 2.3.2.1-3 -f[(j'2.3.2.1-3 . 1
; 2.15 2.6-1 2.6-1
. y;..j 1 l 12 2.16 2.3.8-1 2. 3. 8 -1 1 4.9 4.10 4.7-2 4.7-1 4.7-2 2 l9 4.7-1 2 4.11 4.9-1 4.9-1 2 9
. 4.12 4.0-1 4.0-1 2 5.11 5.2.3.7-1 5. 2.3. 7-1 2 7 5.12 10.19-1 10.19-1 2 5.13 5.2.3.8-2 5.2.3.8-2 2 5.14 5.3.3.3.3-3 5.3.3.3.3-3 2 5.15 5.3.3.3.3-4 5.3.3.3.3-5 2 12 5.16 , 5. 3. 3. 3. 3-5 5.3.3.3.3-7 2 5.17 5.2.3-1 5.2.3-1 2 7.1 5.3.3.3.3-1 5.3.3.3.3-1 2 7.2 s 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 I
7.4 7.1-1 7.1-1 3 11 7.5 7.2.3.1-1 7.2.3.1-1 3 l 12 7.6 4.4 1 4.4-1 2 : 7.7 7.2.3.3-1 7.2.3.3-1 3 l 12 7.8 7. 2. 3. 6-1 7.2.3.6-1 3 11 7.9 7.2-1 7.2-1 3 7.10 7.2.3.9-1 7.2.3.9-1 3 7.11 7.2.3.9-2 7.2.3.9-2 3 l 12 17 l9 ( __ _ _ - _ _ _ - _ _ _ . - - _ - _ _ . - -.
- 1 ZPS AMENDMENT 12 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS. (Continued)
AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PACE OF PSAR 7.12 7.2-2 7.2-2 3 . 11 7.13 1.0-1 1.0-1 1 7.14 , 7.12.5.3-1 7.12.5.3-1
, 3 i 7.15 7.4.3-1 7.4.3-1 3 i 7.16 7.5.7.3.3-1 7. 5. 7. 3. 3 -1 3 7.17 7.8.5-1 7.8.5-1 3 11 7.18 7.5.8-1 7.5.8-1 3 7.19 7.6.3-1 7.6.3-1 3 7.20 7.8.5.2-1 7.8.5.2-1 3 7.21 7.9-1 7.9-1 3 7.22 7.3.4.8-1 73.4.8-1 3 12 7.23 7.10-1 7.10-1 3 7.24 D.6-2 D.6-14 5 l
7.25 D.0-1 D.0-1 5 l11 7.26 10.10.3-1 10.10.3-1 4 7 7.27 . 7.2-3 7.2-5 7.28 3 l11 7.0-1 7.0-1 3 l 12 7.29 10.19-2 10.19-2 4 11 7.30 7.0-2 l 7.0-4 3 7.31 7.7-1 7.7-1 3 12 8.1 8.3.2.1-1 8. 3. 2.1 -1 4 8.2 8.3.2-1 8.3.2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 8.5 8.5.4-1 8.5.4-1 4 8.6 8.4.3-2 8.4.3 2 4 8.7 8.5.3.1-1 8.5.3.1-1 4 8.8 8.0-1 8.0-1 4 18 9 f I
l ZPS AMENDMENT 12 LIST OF 2PS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued) AEC QUESTION RENUhBERED NUMBER VOLUME AS QUESTION PAGE OF PSAR l 8.9 8.0-2 8.0-2 4 8.10 8.9-1 8.9-1 4 8.11 8.10-1 8.10-1 4 9.1 9.2.4-1 , 9.2.4-1 4 9.2 9.2.4.6-1 9.2. 4. 6-1 4 9.3 9.4-1 9.4-1 4 9.4 9.4.6-1 9.4.6-1 4 l 11 9.5 9.2.4.7-1 9. 2.4. 7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 10.0-2 10.0-2 4 l12 10.2 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 ! 4 l 11 10.4- 10.11.2-1 10.11.2-1 4 7 12.22 12.6.1-1 12.6.1-1 4
! 12.23 12.5.6-1 12.5.6-1 i 4
'~
, 13.1 13.0-1 13.0-1 4
13.2 13.2.1.6-1 13.2.1.6-1
' k.
13.3 13.2.1.2-1 13.2.1.2-1 4 9 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 11 14,12 14.9.1-1 14.9.1-1 4 14.13 14.9.1-2 14.9.1-3 4 12 {
)
19 9
l ZPS l i AMENDMENT 12 l LIST OF 2PS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued) AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR 15.16 A.2-1 A.2-1 5 7 15.17 A.2-2 A.2-5 5
." l 15.18 A.2-3 '
A. 2- 6 5 15.19 B.1 B.1-1 5 l12 t i { 4 l i i l l 4 l l i O e i e 20 l9 ; bl
l ZPS i AMENDMENT 12 LIST OF 2PS. APRIL 9.1971 AEC QUESTIONS AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PACE OF PSAR 1.1 1.10 ,, 9 .- 1.10-1 1 y.g 1.2 .B.0-1 '
. ;= 8.0-1
' 5
,-.t[
2.17 2.5-1 < 2.3-1 1 ! 2.18 2.4.5-2 2.4.5-7 1 .' 2.19 2.4.5.1-1 2.4.5.1-1 1 l-2.20 2.4.5.3-1 2. 4.5 . 3-1 1 2.21 2. 4.5 . 3-2 2.4.5.3-2 1 2.22 2.4.5-3 2.4.5-10 1 3.5 3.3-1 3.3-1 1 3.6 3.3-2 3.3-2 1 l 3.7 3.3-3 3.3-4 1 - 12 3 4.13 4.3-1 4.3-1 2 4.14 4.4-2 4.4-2 2 5.0 10.20-1 10.20-1 , 4
,' 6.1 6.5-1 6.5-1 2 6.2 6.5-2 6.5-5 2 6.3 6.5-3 6. 5-7 2 6.4 6.5-4 6.5-13 2 6.5 6.5-5 6.5-16 2 6.6 6.5-6 6.5-21 2 6.7 6.5-7 6.5-22 2 6.8 6.5-8 6.5-32 2 6.9 6.5-9 6.5-38 2 6.10 6.5-10 6.5-39 2 6.11 6.5-11 6.5-41 2 6.12 6.5-12 6.5-43 2
{ 6.13 6.4-1 6.4-1 2 21 i
i i l ZPS
. AMENDMENT 12 t
LIST OF ZPS. APRIL 9,1971 AEC QUESTIONS, (Continued) l
, j
~:p v , ; 'ipa - .- - , -
AEC QUESTION RENUMBERED
~~
1 NUMBER VOLUME 1 AS QUESTION PAGE m 0F PSAR f 7.32 7.9-2 7.9-6 3 7.33 7.9-3 7.9-8 3 < 12.24 12.2.2.5-2 12.2.2.5-2 4 12.25 12.2.1.1-4 12.2.1.1-4 4 12.26 12.3.8-1 12.3.8-1 4 14.14 ~ 4.3-2 4.3-2 2 14.15 3. 7.4. 2. 3-1 3. 7. 4. 2. 3-1 2 14.16 7.9-4 7.9-9 3 ' 14.17 7.9-5 12 7.9-11 3 14.18 7.9-6 7.9-13 3 ! 14.19 7.9-7 7.9-14 3 14.20
- 14.5-1 ~
14.5-1 4 14.21 14.5.5-1 f' 14.5.5-1 4 ' 14.22 14.5.5-2 14.5.5-3 4 14.23 14.5.6-1 1 14.5.6-1 4 14.24 5.3.4.3-1 l 5.3.4.3-1 2 ! 14.25 5.3.4.4-1 5.3.4.4-1 2 : 14.26 14.9.2.3-1 . 14.9.2.3-1 4 15.19 I.0-1 I.0-1 5 15.20 C.3.1-4 C 3.1 -4 5 9 O e
I zrS AMENDMENT 12
' INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 3 SECTION 7.0 - CONTROL AND 7*lSTRUMENTATION This section has been amended to incorporate new information, system changes, minor editorial comments and answers to AEC questions.
All changes have been indicated by a vertical line and the ' amendment number (12) in the right margin of the page. All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 12". Figures that have been altered in any way are indicated by the amendment number in the upper right corner of the figure;' note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below: FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography l12 ] To updateprocedure: your copy of the Wm. H. Ziauner Nuclear Power Station PSAR, please ] use the following ; , i
- 1. In Volume 3, SECTION 7.0 - CON 11tOL AND INSTRUMENTATION, remove and.
destroy the Table of Contents Pages 7.0-ix, 7.0-xiv 'and 7.0-xviii and replace with amended Dages 7.0-ix, 7.0-ix.1, 7.0-xiv and xviii.
- 2. In Volume 3, SECTION 7.0 - CONTROL AND INSTRUMENTATION remove and de-stroy the following pages and replace with the appropriate pages listed below:
)
REMOVE PAGE REPLACE WITH AMENDED PAGE 7.2-2 7.2-2
, 7.2-6 -
7.2-6 1 7.2-7 7.2-7 ) 7.2-14 7.2-14 h ! 7.2-23 l 7.2-23 7.2-24 7.2-24 7.2-25 o ! 7.2-25 7.3-8 7.3-8 f1' 7.3-10 7.3-10 7.3-12 7.3-12 1 1 J
T. PS j, AMENDMENT 12 REMOVE PAGE REPLACE WIDI AMENDED PAGE 7.3-15 7.3-15 7.3-17 7.3-17 7.3-19 7.3-19 7.3-20 7.3-20 7.3-22 7.3-22 7.3-23 7.3-23 7.3-24 7.3-24 7.3-26 7.3-26 7.3-28 , . . 7.3-28 7.3-53 7.3-53 7.4-3 7.4-3 7.4-6 7.4-6 7.4-9 7.4-9 7.4-11 7.4-11 7.4-15 - 7.4-15 7.4-19 7.4-19 7.4-27 7.4 27 7.4-28 7.4-28 7.5-9 7.5-9 7.5-11 7.5-11 7.5-19 7.5-19 7.5-21 7.5-21 7.5-23 7.5-23 7.12-1 through 7.12-21 7.12-1 through 7.12-14 7.13-3 through 7.13-5 7.13-3 through 7.13-6 ,
- 3. In Volume 3, SECTION 7.0 - CONTROL AND INSTRUMENTATION remove anddestroy Figure 7.12-1 and replace with amended Figures 7.12-1.2 and 7.12-1.2.
- 4. In Volume 3, SECTION 7.0 - CONTROL AND INSTRUMENTATION, behind the red tabbed divider page titled " Amendments to Section 7.0".
- a. In front of Page 7.1-1 insert Pages 7.0-1 through 7.0-4.
- b. Behind Page 7.2-5 insert Pages 7.2.3.1-1 through 7.2.3.1-6
- c. Behind Page 7.2.3.16 insert Page 7.2.3.3-1.
- d. Behind Page 7.2.3.9-1 insert Page 7.2.3.9-2.
- e. Behind Page 7.2.3.9-2 insert Page 7.3.4.8-1.
- f. Behini Page 7.6.3-1 insert Page 7.7-1.
- g. Behind Page 7.8.5.2-1 insert the following Pages in the order given below.
Page Number 7.9-1 through 7.9-17 7.10-1 , 7.12.5.3-1 j l s ___ " - ~ - ' - - - ' - - ' - - ' - - - - - - - ~ ~ -
9 e
, ZPS 5-
'j AMENDMENT 12 I
i l TABLE OF CONTENTS. (Continued) PAGE
, 7.12.1.4 Safety Evaluation 7.12-4 l 7.12.1.5 Inspection and Testing 7.12-4 7.12.1.6 Operational Nuclear Safety Requirements ' *
"~
7.12-4
- j 7.12.2 Air Ejector Offgas Radiation Monitoring and Sampling System
" 7.12-5 7.12.2.1 Power Generation Objective 7.12-5 l 7.12.2.2 Power Generation Design Basis 7.12-5 I 7.12.2.3 Descriptf.on -
7.12-5 7.12.2.4 Safety Evaluation 7.12-6 { 7.12.2.5 Inspection and Testing 7.12-6 l 7.12.2.6 Operational Nuclear Safety Requirements 7.12-6 7.12.3 offgas vent, Pipe Radiation Monitoring System 7.12-8 , 7.12.3.1 Power Generation Objective 7.12.3.2 Power Generation Design Basis 7.12-8 fI 7.12-8 fi 7.12.3.3 Description 7.12-8 ; 7.12.3.4 Inspection and Testing 7.12-8 l 7.12.3.5 Operational Nuclear Safety Requirements 7.12-8 7.12.4 Process Liquid Radiation Monitors 7.12-10 12 i 7.12.4.1 Power Generation Objective 7.12-10 l 7.12.4.2 Power Generation Design Basis 7.12-10 7.12.4.3 Description 7.12-10 7.12.4.4 Power Generation Evaluation 7.12-11 7.12.4.5 Inspection and Testing 7.12-11 7.12.4.6 Operational Nuclear Safety Requirements 7.12-11 7.12.5 Reactor Building Ventilation Exhaust Radiation Monitoring System 7.12-12 i 7.12.5.1 Power Generation Objective 7.12-12 7.12.5.2 Power Generation Design Basis 7.12-12 7.12.5.3 Description 7.12-12 l 7.12.5.4 Inspection and Testing 7.12-12 , 7.0-ix
ZPS AMENDMENT 12
, TABLE OF CONTENTS, (Con tinued)
PACE 7,12.5.5 Operational Nuclear Safety Requirements 7.12-12 7.12.6 Fuel Pool Ventilation Exhaust Radiation Monitoring System 7.12-13 7.12.6.1 Safety Objective - 7,12-13 7.12.6.2 Safety Design Basis 7.12-13 12 7.12.6.3 Description '
~
7.12-13 7.12.6.4 Safety Evaluation , 7.12-14' 7.12.6.5 Inspection and Testing 7.12-14 I i i I 1 f i l l I
)
I l f
.I 7.0-ix.1
.i
1 i l ZPS j AMENDMENT 12
, LIST OF TABLES, (Continued)
TABLE NUMBER LTL_E PAGE , 1 7.6-1 Refueling Interlock Effectiveness 7. 6 -6 f 7.7-1 Reactor Manual control System Instru- 7.7-11 f-ment Specifications j , 7.8-1 Reactor Vessel Instrumentation 7.8-3 ; l Instrument Specifications , 7,12-1 Characteristics of Process Radiation 7.12-3 Monitoring Systems , 7.12-2 Process Radiation Monitoring System 7.12-7 , Environmental and Power Supply Design 12 ; Conditions ,, ; 3 7.13-1 Area Radiation Monitoring System Faviron- 7.13-2 mental and Power Supply Design Conditions 7.13-2 Locations for Area Radiation Monitoring 7.13-4 12 ' Sensors
- 7.16-1 Instrumentation Input Summary Neutron 7.16-4 Monitoring System 7.16-2 Instrumentation Output Sunanary Signal 7.16-12 l l' Output Description .
l 7.17-1 Acceptable Ultimate Performance Limits 7.17-5 7.17-2 Acceptable Operational Design Limits 7.17-8 I
. i as 7.0-xiv
ZPS AMENDMENT 12 LIST OF FIGURES, (Continued) FIGURE NUMBER TITLE , n 7.7-4 Manual Control Self-Test Provisions i
)_
7.7-5 Rod Block Interlocks from Neutron , Monitoring System 7.7-6.1 Reactor Control Bench Board - Part 1 7.7-6.2 Reactor Control Bench Board - Part 2 7.7-7 Rod Block Functions 7.8-1 Nucles'r Boiler System Piping & Instruments-tion Diagram 7.8-2.1 Nuclear Boiler System Instrumentation Diagram - Part 1 7.8-2.2 Nuclear Boiler System Instrumentation Diagram - Part 2 7.8-3 Reactor vessel Thermocouple Locations 7.9-1 Recirculation Flow Control Illustration . 7.9-2.1 Reactor Recirculation System Valve Flow Control Functional Control Diagram 6 7.9-2.2 ' Reactor Recirculation System Valve Flow j Control Functional Control Diagram j 7.9-3 Reactor Recirculation System Flow Control l Instrument Engineering Diagram : 7.10-1 ' Feedwater Control System Instrument l Engineering Diagram 8 7.12-1.1 Process Radiation Monitoring System I Instrument Engineering Diagram 12 7.12-1.2 Process Radiation Monitoring System f' Instrument Engineering Diagram 7.1 -i Area Radiation Monitoring System Instrument Engineering Diagram ,- 7.17-1 Damping Coefficient Versus Decay Ratio . (Second Order Systems) 7.17-2 Hydrodynamic and Core Stability Model 7.17-3 Comparison of Test Results with Analysis , 7.17-4 Total System Stability Model 7.17-5.1 10 Cent Rod Reactivity Step at 687. Power, 12 ; Natural Circulation ' 7.0-xviii ' f
ZPS 1 A.'ENDMENT 12 4
, essential variables that have spatial dependence.
7.- The following bases provide assurance that the reactor. protection L system is designed with sufficient reliability to fulfill safety design bases 1, 2, and 3:
^
- a. Any one failure, intentional bypass, maintenance operation, calibration operation, or test to verify operational avail-ability shall not impair the functional ability of the pro-7 tection system to respond correctly. jl
- b. The system shall be designed for a high probability that iI
- j when any monitored variable exceeds the scram serpoint, the I event shall either result in an automatic scram or shall not !
impair the ability of the system to scram as other monitored i variables exceed their scram trip points. l
- c. Where a plant condition that requires a reactor scram can 'I be brought on by a failure or malfunction of a control or j regulating system, and the same failure or malfunction pre- : l Vents action by one or more reactor protection system channels ; j designed to provide protection against the unsafe condition, j !
the remaining Portions of the reactor protection system shall ' meet the requirements of safety design bases 1, 2,3, 7a, and 7f. ' 12 ,
- d. The power supply for the reactor protection system shall be 1 j l
arranged so that loss of one supply neither causes a reactor scram nor prevents an orderly plant shutdown. ; I
, e. The system shall be designed so that, ence initiated, a reactor protection system action goes to completion. Return to normal operation af ter protection system action shall require deliberate operator action.
f. s There shall be sufficient electricC., and physical wiring yy and piping separation between trip channels and between trip logics monitoring the same variable to prevent environmental factors, electrical transients, and physical events from impairing the ability of the system to respond correctly. g. Earthquake ground motions shall not impair the ability of the reactor protection system to initiate a reactor scram. 8. The following bases are specified to reduce the probability that reactor protection system operational reliability and precision vill be degraded by operator error: a. Access to all trip settings, component calibration controls, t s. 7.2-2 l
I i ZPS ) 1 l AMENDMENT 12 5 de-energized, the rods are not scrammed. If a trip then occurs in any of the trip
. logics of the other trip system, the remaining scram pilot valve solenoid for cach rod is de-energized, venting the air pressure from the scram valves, and al-lowing control rod drive water to act on the control rod drive piston. Thus, all control rods are scrammed. The water displaced by the movement c,f each rod pis-ton is vented into a scram discharge volume. Figure 7.2-1 shows that when the solenoid for each backup scram valve is energized, the backup scram valves vent the air supply for the scram valves; this action initiates insertion of every control rod regardless of the action of the scram pilot valves.
A scram can be manually initiated. There are two scram buttons, one for trip logic A3 and one for trip logic B3. Depressing the scram button on trip I logic A3 de-energizes trip ' actuators A3 'and opens corresponding contacts in trip ' cetuator logics A. A single trip system trip is the result. To effect a manual scram, the buttons for both trip logic A3 and trip logic B3 must be depressed. j The manual scram buttons are physically close together so that one hand motion ccn effect a scram. By operating the manual scram button for one trip logic at e time, followed by reset of that trip logic, each trip system can be tested for manuti scram capability. It is also possible for the control room operator to ceram the reactor by interrupting power to the reactor protection system. This ccn be done by operating power supply breakers in the control room. The manual scram capability provided in the control room meets safety design basis 9. To restore the reactor protection system to normal operation following cny single trip system trip or scram, the trip actuators must be manually reset. Rsset is possible only if the conditions that caused the trip or scram have been cisared and is accomplished by operating switches in the control room. Figure 7.2-3 shows the functional arrangement of reset contacts for trip system A. This meats safety design basis 7e. Whenever a reactor protection' system sensor trips, it lights a printed red window, common to all four trip cha inels for that variable, on the reactor control panel in the control room to indicate the out-of-limit variable. Each trip system lights a red window indicating the trip system which was tripped. A reactor protection system trip channel trip also sounds a buzzer or horn, which can be silenced by the operator. The annunciator window lights latch un-til manually reset; reset is not possible until the condition causing 'the trip has been cicared. In addition to the computer printout of individual channels in a " tripped status as described below the physical positions of reactor protection 12 l system relays may be used to identify the individual sensor that tripped in a group of sensors monitoring the same variable. The location of alarm windows pro- .' vides the operator with the means to quickly identify the cause of reactor pro-tection system tri.ps and to evaluate the threat to the fuel or nuclear system pro- ' cess barrier. ; V To provide the operator with the ability to analyze an abnormal transient I during which events occur too rapidly for direct operator comprehension, all : recctor protection system trips are recorded by an alarm typewriter controlled ' by the process computer system. All trip events are recorded. The first 40 ! l 7.2-6 '
] .
ZPS AMENDMENT 12 are recorded in chronological sequence except that evente occurring within 4 milliseconds of each other are treated as having ocer.rred simultaneously. Use of the alarm typewriter and computer is not required for plant safety, and information provided is in addition to that immediately available from other annunciators and data displays. The printout of trips is of particular useful-nass in routinely verifying the proper operation of pressure. level, and valve position switches as trip points are passed during startups, shutdowns, and maintensoce operations. Reactor protection system inputs to annunciators, recorders, and the computer are arranged so that no malfunction of the annunciating, recording, or computing equipment can functionally disable the system. Signals directly from the reactor protection system sensors are not used as inputs te annunciating or dacs logging equipment. Isolation is provided between the primary signal and the information output. The arrangement of indications y,tinent to the status and response of the reactor protection system satisfies safety design bases 10s and 10b. 7.2.3.6 Scram Functions and Settings The following discussion covers the functional considerations for the variables or conditions monitored by the reactor protection system. Table 7.2-1
- lists the specifications for instruments providing signals for the system. Figure as t
7.2-4 shows the scram functions in block form. Testability during operation is h provided for each of the functions listed below except item 11 (mode switch). 12
- 1. Neutron monitoring system trip. To provide protection for the fuel against high heat generation rates, neutron flux is monitored and used to initiate a reactor scram. The neutron monitoring system setpoints and their bases are discussed in the "Eeutron Monitoring System" Subsection 7.5. ,
l6
- 2. Nuclear system high pressure. High pressure within the nuclear system poses a direct threat of rupture to the nuclear sy' stem -
process barrier. A nuclear system pressure increase while the reactor is operating compresses the steam voids and results in a positive reactivity insertion causing increased core heat generation that could lead to a violation of the core thermal-hydraulic safety limit. A scram counteracts a pressure increase by quickly reducing the core fission heat generation. The nuclear system high pressure scram setting is chosen slightly above the reactor vessel maximum normal operating pressure to per-mit normal operation without spurious scrams, yet provide a wide margin to the nuclear system pressure safety limit. The locatic,n of the pressure measurement, as compared to the location of highest nuclear system pressure during transients, was also considered in the selection of the high pressure scram setting. The nuclear J l l 7.2-7
l I I ZPS
~
AMENDMENT 12 side the primitry containment and inside the reactor building; they ara physically separated from each other and tap off the reactor vessel at widely separated points. The reactor protection system pressure switches, as well as instruments for other systems, sense pressure and level from these same pipes. The physical separation and signal arrangement assure that no single physical event can prevent a scram ' due to reactor vessel low water level. Temperature equalizing columns ! are used to increase the accu: cy of the level measurements. Fas t de-pressurization transients do not affect the accuracy of the instruments 3 i as described in subsection 6.13 of Amendment 17 to the Monticello Nu- 12 ' clear Plant (Docket #50-263). L
- 4. Turbine stop valve closure inputs to the reactor ptotection I
system are i ' from valve stem position switches mounted on the four turbine stop : valves. Each of the double-pole, single-throw switches is arranged to ' open before the valve is more than 10% closed to provide the earliest positive indication of closure. Either of the two trip channels as-sociated with one stop valve can signal valve closure. The logic is arranged so that closure of any two valves causes a single trip system trip, and closure of three or more valves initiates a scram.
- 5. Turbine control valve fast closure inputs to reactor protection system are from four (4) pressure switches on the control valve hydraulic fluid [
discharge header. The pressure switches monitor the loss of hydraulic j fluid pressure which will result in the fast closure of the controi 11 valves. The pressure switches provide signals to the four reactor pro-tection system trip channels. The arrangement is a one out of two [ i taken twice logic. '
- 6. Main steamline isolation valve closure inputs to the reactor protection '
system are from valve stem position switches mounted on the eight main steamline isolation valves. Each of the double-pole, single-throw ; i switches is arranged to open before the valve is more than 10% closed ' to provide the earliest positive indication of closure. Either of the two trip channels associated with one isolation valve can signal valve l closure. To facilitate the description of the logic arrangement, the ' position sensing channels for each valve are identified as follows:* l l6 Position Sensing Valve Identification Channels Main steam line A, inboard valve a,b Main steam line A, outboard valve c,d Hain steam line B, inboard valve e,f I Main steam line B, outboard valve g,h {
-{
Main steam line C, inboard valve j,k l
;q i
i CAdditional information is available in GE Topical Report NED-10139 N 7.2-14 6
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l ZPS AMENDMENT 12 The following section covering inspection and testing of the reactor protection system demonstrates that safety design basis 11 is satisfied. - 7.2.5 Inspection and Testing s The reactor protection systeu can be tested during reactor operation by five separate tests. .The first of these is the manual trip actuacor test. By ( depressing the manual scram button for one trip system, the manual trip logic l actuators are de-energized opening contacts in the trip actuator logics. After resetting the first trip system tested, the second trip system is tripped with the other manual scram button. The total test verifies the ability to deenergize all eight groups of scram pilot valve solenoids by using the manual scram push button switches. Scram group indicator lights verify that the trip actuator contacts have opened. The second test is the automatic trip actuator test which is accomplish-ed by operating, one at a time, the keylocked test switches for each automatic h trip logic. The switch de-energizes the trip actuators for that trip logic, caus-l ing the associated trip actuator contacts to open. The test verifies the ability 3 of each trip logic to de-energize the trip actuator logics associaled with the j parent trip system. The actuator and contact action can be verified by observing the extinguishment of trip actuator output lights. l 12 s The third test includes calibration of neutron monitoring system by means cf simulated inputs from calibration signal units. The section titled
" Neutron Monitoring System" describes the calibration procedure.
The fourth test is the single rod scram test which verifies capability of each rod to scram. It is accomplished by operation of toggle switches on the protection system operations panel. Timing traces can be made for each rod scrammed. Prior to the test, a physics review must be conducted to assure that the rod pattern during scram testing does not create a rod of excessive reactivity worth. The fifth test involves applying a test signal te each reactor protection system trip channel in turn and observing that a trip logic tr'.p results. This test also verifies the electrical independence of the trip cnannel circuitry. The test signals can be applied to the process type sensing instruments (pressure and differential pressure) through calibration taps. The test is conducted as follows:
- 1. An instrument technician following instructions of authorized personnel unlocks or cuts the seal on the instrument shutoff valves to a specific instrument and shuts off the instrument line.
- 2. The instrument is isolated using the instrument valve (or instru-ment manifold valve) and a calibration set is attached to the in-7.2-25
_. / I TABLE 7.3-1 (Cont I ' ]" l_ Approx. Valves Valve
---- Location No. of Pipe Per Ref. to Valve Powe8 l Line Isolated Lines Size (in . ) Line Class Drywe ll Ty pe (6) cpen Rx. Eldg. Closed Cooling 1 6 1 C Inside 70 Gate d Water Inlet ,
1 - C Outside NO Cate d Rx. B1dg. Closed Cooling 1 6 1 C Inside NO Cate O Water Dutict 1 C Outside NO Cate O Demineralized Weer In 1 4 1 3 Inside Check - 1 8 Outside Check - Drywell Purge Inlet
- 1 18 1 B Outside HD Butterf'ly M 1 B Inside HD Butterfly M Drywell Main Exhaust 1 18 1 B Inside HD Butterfly AC Drywell Exh. Val. Bypass 1 2 1 B Outside A0 Globe Air /d
- Supp. Chamber PurBe Inlet 1 18 2 B Outside H0 Butterfly A@ 1 B Inside PO Butterfly AC Supp. Chamber Exh. Vm1. , Bypass 1 2 B Outside 1 AO Globe. Air /Sf Supp., Chamber Hain Exh. I 18 1 B Inside 10 Butterfly 4C Drywell Air Sample 1 % 2 B Outside A0 Globe Air /a Supp. Chamber Air Samples 1 2 Outside 1 B AO Globe Afriff Drywell Air Samples 2 h 2 B Outside AO Globe Air /AQ Supp. Chamber Air Samples 2 % 2 B Outside AO Globe Air /63 Supp, Chamber Air Samples 2 1 2 Outside B A0 Globe Air /M Drywell 6 Supp. Chamber Purge Exh. Fan Suction 1 18 1 B Outside HD Butterfly AC Standby Cas Treatment System Suction 1 18 1 B Outside }O Butterfly AC SO - Solenoid Opem HD - Motor Operatec AO - Air Operated I . I
- ZPS AMENDENT 12 inued)-
6n Hin. Closing Normal to Power to Isolation Rate or Status } Close (5) (6) Sinnal Time (7.11.12) (9. 10) Remarks C AC RM S tandard Open g C. AC RM Standard Open S C AC RM Standard Open g C AC RM Standard Open O Process Rev. Flow - Closed Process Rev. Flow - Closed C AC F,A,Z, (8) S tandard Closed g C AC F,A,Z (8) S tandard Closed S C AC F,A,Z (8) S tandard Closed Note (13) S C Spring .F,A,Z (8) Standard Closed Note (13) SC AC F , A ,2, (8) Standard Closed ! AC F,A,Z (8) Standard Closed l C S'p ring F,A,Z (8) Standard Closed Note (13) @ C F,A,Z AC (8) Standard Closed Note (13)
/AC Spring F,A,Z (8) Standard Open E 'AC Spring F,A,Z (8) Standard Open 3 AC Spring F,A,Z (8) Standard Closed S AC Spring F,A,Z (8) Standard Closed S AC Spring F,A,Z (8) S tandard closed 3
C AC F,A,Z (8) Standard Closed AC F,A,Z (8) Standard Closed ersted
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- l ZPS AENDENT 12
, .Min.
I Closing Normal - ( Isolation Rate or Status Sinnal Time (7.11.12) (9. 10) Remarks !
~RM Standard Open :
RM Standard Open - RM Standard Open . RM Standard Open Rev. Flow - Closed , , Rev. Flow - Closed .
- i F,A,Z,(8) Standard Closed i F,A,Z (8) Standard Closed F,A,Z (8) Standard Closed Note (13) l .F,A,Z (8) Standard Closed Note (13)
F,A,Z (8) Standard Closed f F,A,Z (8) Standa rd Closed .; F,A,Z (8) Standard Closed Note (13) 11 F,A,Z (8) Standard Closed Note (13) F,A,2 (8) Standard Open , F,A,Z (8) Standard open i F,A,Z (8) S tandard Closed 12 F,A,Z (8) Standard Closed , F,A,Z (8) Standard Closed i F,A,Z (8) Standard Closed f F,A,Z (8) Standard closed 3 i . ap 7.3-8 l6 h .
ZPS AMENDMENT 12 TABLE 7.3-1, (Continued) ISOLATION SIGNAL CODES FOR TABLE 7.3-1 F l Sinnal Description Z* High radiation, fuel pool ventilation exhaust. 12
- RM* Remote manual switch from control room. (All regular Class A and Class B isolation valves are capable of remote manual ,
operation from the control room.) f
- These are the isolation functions of the primary containment and reactor vessel isolation control system; other functions i are given for information only.
l 4 7.3-10 4
ZPS AMENDMENT 12 TABLE 7.3-1, (Continued) NOTES FOR TABLE 7.3-1 ' 1
- 8. Fuel pool vent exhaust high radiation signal "Z" is generated by l 12 'l two trip channela, each channel has two trip units. This rewires one '
unit at high trip or one unit at downscale (instrument failure) trip, on i one trip channel and one unit at high trip or one unit at dommscale trip 11 on the other trip channel in order to initiate isolation. ;
- 9. Valves identified by an asterisk' in the " Normal Status" colves: can be i opened or closed by remote manual switch for operating convenience during any mode of reactor operation except when automatic signal is present. '
- 10. Normal status position of valve (open or closed) is the position during normal power operation of the reactor (see " Normal Status" column).
- 11. 1he specified closure rates are as tequired for containment isolation j ;
only. ,
- 12. Minimum closing rate is based on valve and line size. ~
- 13. A manual switch overrides all automatic signals on the drywell exhaust valve bypass valve, suppression chamber exhaust valve bypass walve, 11 drywell exhaust valve and suppression chamber exhaust valve, s
h .. aff 7.3-12 __ _ _ _ _ _ _ _ ~ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _
- - - ^ - - - ^ - ' - - - - - - - - ^ - ' - ' ~ ^ ' - - ' - " - - ~ - - - - - ~ ~ -, - - - - - - ' - - - ' - -
l ' ZPS AMENDMENT 6 I satisfies safety design bases ils and lib. 7.3.4.6 Isolation Valve Closing Devices and Circuits Table 7.3-1 itemizes the type of closing' device provided for each isolation valve intended for use in automatic or remote manual isolation of the primary containment or reactor vessel. To meet the requirement that
- automatic Class A valves be fully closed in time to prevent the reactor vessel {
water level from falling below the top of the active fuel as a result of a ; i break of the pipeline which the valve isolates, the valve closing mechanisms are designed to give the minimum closing rates specified in Table 7.3-1. In ll I many cases a standard closing rate of 12 inches per ainute is adequate to meet isolation requirements. Using the st.andard rate a 12 inch valve is closed in ', i 60 seconds. Conversion to actual closing time can be made by using the size of f the line to be isolated. Because of the relatively long time required for ~ fission products to reach the containment atmosphere following a break in the nuclear system process barrier inside the primary containment, a standard closure rate (12 inches / minute) is adequate for the automatic closing devices on Class B isolation valves. ] i Motor operators for Class A and class B isolation valves are selected ; with capabilities suitable to the physical and environmental requirements of 4 service. The required valve closing rates were considered in selecting motor ; ! operators. Appropriate torque and limit switches are used to ensure proper i valve seating. Handwheels, which are automatically disengaged from the motor operator when the motor is energized, are provided for local-manual operation. Direct solenoid operated isolation valves and solenoid air pilot valves are chosen with electrical and mechanical characteristics which make them suitable for the service for which they are intended. Appropriate watertight l12 or weathertight housings are used to ensure proper operation under accident conditions. .- The pneumetic actuator used for testable check valves is designed to allow opening the valve at near zero psi differential pressure across the valve. The actuator cannot close the valve against forward flow or prevent the closing of the valve against reverse flow. Thus,the check valve will nei-ther hinder forward fluid flow nor fail to stop reverse flow regardless of the 4 condition of the actuator. The main steam isolation valves are spring-closing, pneumatic, piston-operated valves designed to close upon loss of pneumatic pressure to the valve I operator. This is fail safe design. The control arrangement is shown in n Figure 7.3-2. Closure time for the valves is adjustable between 3 and 10 seconds. Each valve is piloted by two, three-way, packless, direct-acting, solennid-operated pilot valves - both powered by ac. An accumulator is ,, located close to each isolation valve to provide pneumatic pressure for l6 l' valve closing in the event of failure of the normal air supply system. 7.3-15 !
ZPS AMENDMENT 12 I t " TABLE 7.3-2 i l PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTDI i INSTRUMENTATION SPECIFICATIONS Isola tion - Function Trip Sensor h Settina AccurJc2 - Reactor vessel differential 0-210" later low water level pressure switch 527.5 inches above vessel sero Reactor vessel differential 0-210" later low water level pressure switch 474 inches above vessel zero Main steam line radiation high radiation see " Main Steam Line Radiation monitor Monitoring System" Main steam line te:.ipera ture 0-600*F l space high +2% 200'F 4 switch ; temperature Main steam line ' differential 0-150 psi +2% high flow ~ 140% rated flav ' Main steam line pressure switch 0-1500 psig +1% low pressure 850 psig -! Primary containment pressure switch high pressure later later 2 psig l4 RCIC turbine steam tempera ture 0-300'F line space high +2% 200*F 4 switch b temperature
! i
} RCIC turbine steam differen tial 0-200" +2%
,- line high flow ~ 150" H O pressure switch HO (45,003 lb/hr)
- 2
<,. RCIC turbine steam pressure switch line low pressure 0-1500 psig +*2% 50 psig Fuel Fool radiation ventilation exhaust monitor see " Fuel Pool Ventilation l 12 high radiation Exhaus t Radiation Monitoring System"
~
i L 7.3-17 1 o
ZPS l AMENDMENT 12 I function plays in initiating isolation of barrier valves or groups of valves is illustrated in the functional control diagram on Figure 7.3-3a and 7.3-3b.
' 1.
Reactor vessel low water level. A low water level in the reactor vessel could indicate that reactor coolant is being lost through )
' a breach in the nuclear system process barrier and that the core '
l is in danger of becoming overheated as the reactor coolant in-ventory diminishes. i Reactor vessel low water level initiates closure of vanous Class j A valves and Class B valves. The closure of Class A valves is 1 intended to either isolate a breach in any of the pipelines in { which processvalves lines.are closed or conserve reactor coolant by closing off 1 e The closure of Class B valves is intended to 4 prevent the escape of radioactive materials from the primary i containment through process lines which are in communication with the primary containment free space. j j
\ )
Two reactor vessel low water level isolation trip settings are l
- used to complete the isolation of the primary containment and the
' reactor vessel.
The first reactor vessel low water level isola-tion trip setting, which occurs at a higher water level than the second setting, initiates closure of all Class A and Class B valves in major process pipelines except the main steam lines. i 1 The main steam lines are left open to allow the removal of heat from the reactor core. The second and lower reactor vessel low water level isolation trip setting completes the isolation of the primary containment and reactor vessel by initiating closure of the main steam isolation valves and any other Class A or Class B t valves must be shut to isolate minor process lines. The first low water level setting, which is coincidentally the same as the reactor vessel low water level scram setting', was , I selected to initiate isolation at the earliest indication of a possible breach in the nuclear system process barrier yet far enough below normal operational levels to avoid spurious isola-tion. Isolation of the following pipelines is initiated when l i reactor (Table 7.3-1,vessel low water signal A): level falls to this first setting RHR reactor head spray l11 Reactor water cleanup Drywell equipment drain discharge Drywell floor drain discharge ! Drywell purge inlet Drywell purge exhaust Suppression chamber purge inlet i l 12 1 Suppression chamber purge exhaust 12 Suppression chamber exhaust bypass valve 7g f
.' )
7.3-19
t q w gpg AMENDMEhT 12 Drywell cxhaust bypass valve Drywell 02 analyzer sample f12 13 l Sunnression chamber 02 analyzer sample i 12 1 j The second and lower of the reactor vessel low water level 1 isolation settings, which is coincidentally the same water level i setting at which the RCIC system is placed into operation, was t selected low enough to allow the removal of heat frois the reactor for a predetermined time following the scram and high enough l' to complete isolation in time for the operation of core standby cooling systems in the event of a large break in the nuclear system process barrier. Isolation of the following pipelines is initiated when the reactor vessel water level falls to this i second setting (Table 7,3-1, signal B): ;; All four main steam lines , Main steam line drain i; Reactor water sample line j
- 2. Main steam line high radiatio _n. l High radiation in the vicinity )
of the main steam lines could indicate a gross release of fission products from the fuel. High radiation near the main steam lines initiates isolation of the following pipelines (Table 7.3-1, signal C): < f i All main steam lines ' Main steam line drai'n i f Reactor water sample line ; l i' The high radiation trip setting is selected high enough above I f i. background radiation leve,1s to avoid spurious isolation, yet low enough to promptly detect a gross release of fission products
! from the fuel. Further information regarding the high radiation l
setpoint is available in the " Process Radiation Monitoring" section. l 3. Main steam line snace high temperature. High temperature in the space in which the main steam lines are located outside of the primary containment could indicate a breach in a main steam line. { l The automatic closure of various Class A valves prevents the
'- t ercessive loss of reactor coolant and the release of significant i amounts of radioactive material from the nuclear system process barrier. When high temperatures occur in the main steam line space, the fo11owin8 Pipelines are isolated (Table 7.3-1, signal D):
i [ 7.3-20
ZPS ,3
) AMENDMENT 12 i
All four main steam lines Main steam drain line i Reactor water sample line ,
)
i l The low steam pressure isolation setting was selected far enough i i below noriaal turbine inlet pressures to avoid spurious isolation j yet high enough to provide , timely detection of a pressure regu-lator malfunction. Although this isolation function is not I required to satisfy any of the safety design bases for this sys- l ! tem, this discussion is included here to make the listing of isolation functions complete. l t l
- 6. Primary containment idrywell) hinh pressur_e_
High pressure in the drywell could indicate a breach of the nuclear # system process barrier inside the drywell. The automatic closure of various Class B valves prevents the release of significant l amounts of radioactive material from the primary containment. Upon ! detection of a 'high drywell pressure, the following pipelines are i isolated (Table 7.3-1, signal F): I RER reactor head spray l Drywell equipment drain discharge Drywell floor drain discharge Traversing in-core probe tubes Drywell purge inlet Drywell purge exhaust Suppression chamber purge inlet l 12
' Suppression chamber purge exhaust Drywell exhaust bypass valve .
12 Drywell 02 analyzer sample Spooression chamber 02 analyzer sample it Suppression chamber exhaust bypass valve !12 The primary containment high pressure isolation setting was i selected lation trips. to be as low as possible without inducing spurious iso-7. RCIC turbine signal K) steam line space high temperature (Table 7.3-1, High temperature in the vicinity of the RCIC turbine steam line outside the primary containment could indicate a break in the RCIC steam line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant process amounts of radioactive material from the nuclear system barrier. When high temperature occurs in the RCIC steam 7.3-22
l , ZPS t AMENDMENT 12 line space the RCIC turbine steam ifne is isolated. The high temperature isolation setting was selected far enough above anticipated normal RCIC system operational levels to avoid spurious operation but low enough to provide timely detection of an RCIC turbine steam line break. 8. RCIC turbine high steam flow (Table 7.3-1, signal K) RCIC turbine high steam flow could indicate a break in the RCIC turbine steam line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive materials from the nuclear system process barrier. Upon detection of RCIC turbine high steam flow the RCIC turbine steam ifne is isolated. The high steam flow trip setting was selected high enough to avoid spurious isolation yet low enough to provide timely detection of an RCIC turbine steam line break. The logic arrangement used for this function is shown on Figure 4.7-2A and is an exception to the usual logic requirement be- l 11 cause high steam flow is the second method of detecting an RCIC turbine steam line break.
- 9. RCIC turbine steam line low pressu,rc (Table 7.3-1, signal K)
RCIC turbine steam line low pressure is used to automatically close the two isolation valves in the RCIC turbine steam line so that steam and radioactive gases will not escape from the RCIC turbine shaft seals into the reactor building after steam pres-sure has decreased to such a low value that the turbine cannot be operated. The isolation setpoint is chosen at a pressure be-
, low that at which the RCIC turbine can operate effectively.
- 10. Fuel pool}}