ML20042F227
| ML20042F227 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 05/02/1990 |
| From: | GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19302E098 | List: |
| References | |
| MFN-037-90, MFN-37-90, NUDOCS 9005080015 | |
| Download: ML20042F227 (532) | |
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v , I May 23,1990 MFN No.037 90 i Docket No. SIN 50 605 i Document Control Desk l U.S. Nuclear Regulatory Commission Washington,D.C. 20$55 i
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l Attention: Charles L. Miller, Director - Standardization and Non Power Reactor Project Directorate i
Subject:
Submittal of Amendment 11, Non. Proprietary Information, to ! GE's ABWR SSAR , I
Reference:
Submittal of Amendment 11, Proprietary Info mation, to OE's ABWR SSAR, MFN No. 038 90, dated May 2,1990 - l
Dear Mr. Miller:
Enclosed are th:rty.four copies of selected sections of Chapter 1, Introduction and General i Description .f Plant, Chapter 3, Design of Structares, Components, Equipment, and Systems, _ ! Chapter 5, Reactor Coolant System and Conneled Systems, Chapter 6, Engineered Safety Features, l Chapter 7, instrumentation and Control Systems, Chapter 9, Auxiliary Systems, Chapter 10, Steam and Power Conversion, Chapter 11, Radioactive Waste Management, Chapter 13, Conduct of i Operation, Chapter 14, initial Test Program, Chapter 19, Response to Severe Accident Policy Statement, Chapter 20, Question and Response Gulc'e, of the Standard Safety Analysis Report (SSAR) for the Advanced Bolling Water Reactor (ABWR). This submittal includes additions covering the following: Response to standby gas treatment system questions, includes modifications to the design; Ac.dition of a flammability control system to the atmospheric control system; Addition of the mitial test program for the turbine island and : radwaste facdities; Addition of the reactor service and turbine water systems; Resolution of draft - safety evaluation report open items for Chapters 4,5 and 6 (except for preservice/ inservice 1 inspection plan); Resolution of preliminary draft safety evaluation report items for Chapters 3 , (except for inservice testing plan); Expansion of Chapter 7 that includes turbine island and i l radwaste facility instrumentation and controls; Balance of Chapter 7 and 10 question responses; . l Update of apphcable USls/OSis; and Closecut of design related emergency preparednesslssue, l ' n
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l l I i Document Control Desk Docket No. STN 50 605 : U.S. Nuclear Regulatory Commission MFN No.037 90 j May 2,1990 Page 2 j i In addition there are other changes to various sections of the SSAR which are identified, along with the above changes, on the page change instruction sheet. ! Please note that all or parts of the following sections contain information that is designated as , General Electric Company proprietary information: 7.7,19.5,19B and 20.3. '!his information is l being submitted under separate cover. ,
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Sincerely, OLS cling Manager Licensing end Consulting Services i ec: D.R.Wilkins (GE) T.E.Murley (NRC) D.M.Crutchfield (NRC) D.J.McGoff (DOE) F. J. Miraglia, Jr. (NRC) D.C.Scaletti (NRC) K.E.Stahlko3f (EPRI) ' F. A. Ross (DOE) EEN9018 O. \ l
ABWR SSAR { Amendment 11. Page change instruction l The folloudng pages have been changed, please make the specined changes in your SSAR. Pages are listed as f page pairs (front & back). Bold page numbers represent a page that has been changed by Amendment 11. I REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. F CHAPTER 1 3218,19 3.2 18,19 ) 3 2 21b,22 3 2 21b,22 l 1 11,111 - 1 11,111 3.2 23,24 3.223,231 1 IV 1Iv 3.2 24.1 3224,241 ; 3225,26 3.2 25,26 + 1.72,3 172,3 3.2 31,32 3.2 31,32 j 174,5 174,5 3.2 33,34 3.2 33,34 , 1.8 51b,51c 1.8 51b,$1c 3.71,2 3.71,2 1.8 51d,51e 1.881d,51e 3.7 23,24 3.7 23,24 ! 1.8 5th,511 1.8 $1h,511 ' 3 .9 11,111 3.9 1i,111 1.92,3 1 9 2,3 3.9 iv,v 3.9 iv,v l 1.931 1.931 3.9 x,xi 3.9x,xt ; 3.93,4 3.93,31 1A iv,v [ 1A Iv,v Add 3.94 6 (~ 1A.2 7,8 1A.2 7,8 3.95,6 3.95,6 ; 1A 2 Sa,9 1A.2 Ba,9 3.9 9,10 3.9 9,10 ; 3.9 17,17.1 3.9 17,17.1 l- 1AA li,ill 1AA II,Ill 3.8 18 3.9 18,18.1 1AA.21,2 1A A.2 1.2 3.9 31,32 3.9 31,31.1 1AA31,2 1AAS 1,2 Add 3.9 32 1A A.41,2 1AA.41,2 3,9 33,34 3.9 33,34 Add 1AA.4 3 Add 3.9341 ' 1AA.51,2 1AA.51,2 3.9 35,36 3.9 38,36 1AA53 1AA.5 3 3.9 39,40 3939,40 Add 1AA 5-4,5 3.9 45 3.945 Add 1AA.5-6,7 3.9 58,5Sa 3.9 58,58a Add 1AA.5 8 Add 3.9 64 1AA.6 2,3 Deleted ' 1AA.6 4,5 Deleted 30.4 2,3 30.4 2,3 1AA.6 6,7 Deleted CHAPTER S CH APTER 3 5.2 9,10 5.29,9.1 l 3.1 1,2 3.11,2 Add 5210-5.211,11a 5.211,11a 3.26,7 3.26,7 l i 3.28,9 3.28,81 5 3 11,111 53 il,Ill l Add 3.29 $31,2 53 1,2 3.2 13,14 3.2 13,14 Add 5321 3 214a,15 3.214a,15 53 3,4 53 3,4 , 1
f ABWR SSAR $ Anwmement 11. Page abange lastreetion (Coat-) ) The telleming pages have been abanged, please make the spectned abanges la year 88AR. Pages an listed as page pairs (front & boek). Bold page sambers apresent a page that has been ebaard by Amendment 11. AEMOVE ADD REMOVE ADD PAMF No PAGE No. PAMF No. Pant No. 5.35,6 555,6 Add Appendix 6A Cover 5.3 6a $J en Add 6A 1,2 Add 6A 3,4 5.4 19,20 5.4 19,20 Add 6A 8,6 Add 6A 7,8 CHAPTER 6 Add 6A 9,10 6.0iv 6.0lv Add Appendix 6B Cover Add 6B 1,2 6.2 vi,vii 6.2 vi,vil Add 6B 3,4 6.2.x,xi 6.2 x,xl Add 68 8,6 Add 6.2 xil Add 6B 7 6.2 17,18 6.2 17,18 l 6.2 31,32 6.2 31,32 CHAPITR 7 6.2 33,33.1 6.2 33,33.1 62 M 6.2 34 7 IV 7IV,y 6.2 35,36 6.2 38,36 7.15,5a 1.18,Sa L' 6.2 36.1 6.2361 7.16 7.16,6.1 D 6.2 44 6.2 44 7.17,8 7.17,8 6.2 50.25,50.26 6.2 80.28,50.26 7.1 24,25 7.1 24,25 6.2 50.39,50.40 6.2 50.39,50.40 6.2 50.43,50A4 6.2 50.43,50.44 7 .3 11,111 7 5 l1,111 7.3 31,32 7.3 31,32 6.2 90 6.2 90,91 Add 6.2 92 7.7 il,lii 7.7 11,111 Add 6.2 93 ?.7iv 7.7lv,v 1.7 33,34 7.7 33,34 6513,14 6513,14 Add 7.7 34.1,34.2 Add 7.7 M 3 ,M .4 651i,lii 6 5 11,111 Add 7.7 34J,M.6 6.51,2 651,2 Add 7.7 M.7,34.8 ; 653,4 653,4 Add 7.7 M.9 ' Add 6.5-41 7.7 35,36 7.7 35,36 ! 655,6 658,6 7.7 37,38 7.7 37,38 657,8 657,8 Add 7.7 38.1,38.2 659,10 6.5 9,10 7.7 39 7,7 39 6.5 11,12 6511,12 Add 7.7 88 6513 6513 Add 7.7 89,90 Add 6514 Add 7.7 91,92 2 0 I
i ABWR SSAR s Amendment 11. Page ebange instructica (Cont-) } l L ne following pages have been chanpd, please snake the specifbed changes in your SSAR. Pages are listed as pay pairs (front & back). Bold page numbers represent a page that has been changed by Amendement 11. j 1 REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGF & Add 7.7 93,94 CHAPIER 10 Add 7.7 95,96 10.2 7,8 10.2 7,8 Add 7.8 11 10.2 12,13 10.2 12,13 7.81 7.8 1 103 5,6 103 5,6 + CHAPTI'R 9 10.4 3,4 10.4-3,4 . 10.4 5,6 10.4 5,6 ; 9 li,ill 9 11,111 10.4 11,12 10.4-11,12 - 9 IV 91v 10.4 13,14 10.4 13,14 10.4 15,16 10.4 15,16 9.12,2a 9.12,2a 10.4 17 10.4 17 9.16,6a 9.1-6,6a 10.4 18,19 10.4 18,19' 9.113a,13b 9.1 13a,13b 10.4 24,25 10.4-24,25 10.4 26,27 10.4 26,27 9.2iv,v 9.2Iv,1va 10.4 30,31 10.4 30,31 O 4 k 9.2 vi 9.2v,vl Add 10.4 31.1 9.21,la 9.21,la 103 32,33 10.4 32,33 9.23,4 9.23,4 10.4 34 10.4 34,35 9.2 10 9.2 10 Add 10.4 36 9211,12 9.2 11,12 Add 9.2121,12.2 CHAITER 11 9.2 13 9.2 13 i Add 9.2 25c 11.5 ii,ili 1 1 .5 11,111 Add 9.2 45,46 1151,2 11.51,2 11.5 7,8 11.5 7,8 l 93-1 93 1 11.5 9,10 1159,10 93 4,5 93-4,5 11 5 11,12 11 5 11,12 93 12,13 93 12,13 Add 93131 CH APTER 13 Add 93 24 133 1 133 1,2 9.4 il,iii 9 .4 11,11i Add 133 3,4 9.41,la 9.41,la Add 133 5 9.41b,1c 9.4 lb,1b.1 i Add 9.4-1c CHAPTER 14 14.2 vi,vil 14.2 vi,vil Add 14.2 vil.1,vil 2 14.2 7,8 14.2 7,8 O 3
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i i 1 ABWR SSAR gl l Amendment 11. Page change Instructica (Cont.) ! The following pages have been changed, please make the speelflod changes in your SSAR. Pages are listed as , page phlrs (front & beck), Bold page numbers represent a page that has been changed by Amendment 11, j REMOVE ADD REMOVE ADD . PAGE No.' PAGE No. PAGE No. PAGE No. ; E 14.2 9,10 14.2 9,10 CHAFTER 20 ! -14.2 11,12. 14.2 11,12 -l 14.2 15,16 14.2 15,16 20.1 1.2,13 20.1 1.2,13 14.2 19,20 14.2 19,20 20.11.4,15 20.1 1.4,1.5 14.2 25,26 14.2 25,26 20.1 1.6,1.7 20.116,1.7 14.2 27,28 14.2 27,28 20.1 1.8,1.9 20.1 1.8,1.9 14.2 29,30 14.2 29,30 20.1 1.10,1.11 20.1 130,1,11 , 14.2 31,32 14.2 31,32 20.1 1.12,1.13 20.1 1.12,1,13 14.2 39,40 14.2 39,40 20.1 1.14 20.1 1.14,1,15 l 14.2 41,42 14.2 41,42 Add 14.2 42.1 20.2 13.2,133 20.2 13.2,13.2.1 14.2-43,44 14.2 43,44 Add 20.2 133 Add 14.2 44.1,44.2 20.2 13.6,13.7 20.2 13.6,13.7 ; Add 14.2 44.1,44.4 20.2 13.12 20.2 13.12,13.13 Add 14.2 44.5,44.6 Add 20.2 13.14 Add 14.2 44.7,44.8 20.2 16 20.2 16,16.1 Add 14.2-44.9,44.10 Add. 20.2 16.2,163 Add 14.2 44.11,44.12 Add 20.2 16.4,16.5 Add 14.2 44.13,44.14 Add 14.2 M.15.44.16 14.2-45,46 14.2 45,45.1 203 vi 203 vi,vil Add 14.2-46 203 13,14 203 13,14 14.2 57,58 14.2 57,58 20 3 15,16 203 15,16 Add 14.2 58.1 203 54,55 20.3 54,55 14.2 63,64 14.2 63,64 203 56,57 20 3 56,57 14.2-65 14.2 65 203-60,61 203-60,61 , 203-61.1 203 61.1,61.2 CHAl'TER 19 20 3 126,127 203 126,126.1 Add 203 127 19iv 19tv 20 3 128,129 20 3 128,128.1 Add 203-129 19.5 1 19.5 1 20 3 130,131 203 130,130.1 Add 20 3 131 , 19B il,lil 19B11,111 20 3 132,133 20 3 132,133 , Add 19BIV Add 20 3 133.1 i 19B.11 19B.11 20 3-134,135 20 3 134,134.1 19B.2 ii,iii 19B.2 il,ill Add 20 3-135 19B.21 19B.21 203-138,139 20 3 138,139 , I 19H.41 1911.4 1 31 40.1 14 140.1 O.
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O Asnendment 11. Page change instruction (Coat-) The following pages have boem changed, please snake the specified changes in your SSAR. Pages am listed as page pairs (front & back). Bold page numbers represent u page that has been changed by Asmondseemt 11. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. 20 S 142,142.1 20 3 142,142 1 20B 22,23 20B 22,23 ; Add 203 143.1.143.2 20B 24,25 20B 24,25 , 20 3-153.1,153 2 20.3 153.1,153.2 20B 26,27 20B 26,27 i 203 206 20 3 206 20B 28,29 20B 28,29 20 } 212,213 20 3 212,213 20B 30,31 20B 30,31 203 214,21$ 20 3 214,214 1 Add 20B 32,33 Add 203 215 Add 20B 34,35 20 3 232,233 20 3 232,233 Add 20B 36,37 Add 20 3 233.1 Add 20B 38,39 20 3 255 290 20.3 255 290 Add 20B 40,41 Add 203 291,292 .\dd 20B-42,43
- Add 203 293,294 Add 20B 44,45 Add 20.3 295,296 Add 20B 46,47 Add 203 297,298 l .
Add 20 3 299,300 Add 203 301,302 l Add 203 303,304 Add 203 305,306 Add 203 307,308 Add 203 309 Add 203 310 Add 203 311,312
- Add 203 313,314 l Add 20 3 315,316 Add 203 317,318 Add 20 3 319 i 20,4 1 20.4 1 1
20B 2,3 20B 2,3 20B 4,5 20B 4,5 20B 6,7 20B 6,7 20B 8,9 20B 8,9 20B 10,11 20B 10,11 20B 12,13 20B 12,13 20B 14,1$ 20B 14,15 20B 16,17 20B 16,17 20B 18,19 20B 18,19 20B 20,21 20B 20,21 (} .$.
MM me- ' = ' pa- ' *sAstooAc uv c CHAPTER 1 O rintsorcoureurs
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1.1 INTRODUCTION
1.11 1.1.1 Format & Contents 1.1 1 1.1.2 ABWR Standard Plant Soope 1.11 1.13 8-j :'6 Documentation 1.1 1 1.1.4 Type of Ucense Required 1.11 1.1.5 Number of Plant Units 1.11 1.1.6 Description oflocation 1.1 1 1.1.7 Type of Nuclear Steam Supply 1.11 1.1.8 Type of Coatainment 1.11 1.1.9 Core Thermal Power levels - 1.11 1.1 GENERAL PLANT DESCRIPrlON 1.21 i 1.2.1 PrincipalDesign Criteria 1.21 1.2.2 Plant Description 1.25 1.3 ROMPARISON TABLES 13 1 13.1 Nuclear Steam Supply System Design Characteristics 1.31 13.2 Engineered Safety Features Design Characteristics 13 1 133 Containment Design Characteristics 13 1 13.4 Structural Design Characteristics 13 1 13.5 Instrumentation and Electrical Systems Design Characteristics 13 1-1.4 IDENTIFICATION OF AGENTS AND CONTRAC"IDRS 1,41 0 1.ii wa i
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*WietWC m,e CHAPTER 1 TABLE OF CONTENTS (Continued)
- = r, 1.s RaourREurim rOR puR1MERnCHNICAL INFORMmON 1.$1 14 MATERIALDICORPORAM RY REFERENCE 1.61 1.7 DRAWINS$ ANp 01MERIIETAttXD INFORMA110N 1.71 1.7.1 Piping and lastruunastation Diagrams 1.71 1.7.2 Elmaric. lastrtinentation and Control Drawings 1.71 1A CONFORMANCE MTpt RTANDARD REVIEW PLAN AND APPLICAR11JTY OF CODES AND STANDARDS 1.81 1.8.1 Conformance With Standard Review Plan 1.81 1.8.2 Applicability of Codes and Standards 1.81 1.8.3 laterfaces 1.81 1.9 IN11RFACES 1.91 APPENDIX 1A RESPONSES TO TMi RELATED MAT 1ERS 1 lii
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ABWR 2Wi100AC Eenndard Plant mm e : Table 1.71 l O PIPING AND INS 11tUMENTATION AND PROCESS F14W DIAGRAMS P
$$AR Fig. No. Page No. Title 1)pe i 4.6-8 4.6 24 CRD System P&tD +
i 4.69 4.6 26 CRD System PFD , 5.13 5.15 Nuclear Boller System P&tD 5.44 5.4 47 Reactor Recirculation System P&ID 5.45 5.4 48 Reactor Reeltculation System PFD , 5.48 5.4 51 Reactor Core Isolation Cooling System P&lD 5.49 $ 4 53 Reactor Core Isolation Cooling System PFD 5.4 10 5.4 55 Residual Heat Removal System P&lD 5.4 11 5.4 59 Residual Heat RemovalSystem PFD , I 5.4 12 5.4 61 Reactor Water Cican Up System P&ID 5.4-13 5.4 63 Reactor Water Clean Up System PFD , 6.2 39 6.2 90 Atmospherie Control System P&ID " 6.2-40 6.2 92 Flamibility Control System P&ID I . 6.31 6.3 25 High Pressure Core Flooder System PFD l 6.37 6.3 33 High Pressure Core l'1oMer System P&ID l l 6.51 6.5 13 Standby Gas Treatment System P&lD . i ! 6.71 6,74 Nitrogen Gas Supply System P&ID 9.11 9.1 23 Fuel Pool Cooling and Cleanup System P&ID 9.12 9.125 Fuel PoolCooling and Cleanup System PFD 9.2 la 9.2 26 Reactor Building Cooling Water System P&ID 9.2 laa 9.3 34a Reactor Building Cooling Water System PFD O 1.7-2 Amendment 11
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ABWR 8AH00AC Standard Plant m., e Toble 1.71 PIPING AND INSTRUMENTATION AND PROCESS FLOW DIAGRAMS (Continued) SSAR Fig. No. Page No. Dtle Type 9.22 9.2 35 HVAC NormalCooling Water System Pa!D 9.23 9.2 37 HVACEmergency Cooling Water System PatD 9.24 9.2 39 Makeup Water System (Condensate) P&tD 9.24 9.2 40 Makeup Water System (Perified) P&TD 9.31 9.316 Standby Liquid Control System P&lD 9}6 93 21 Instrument Air System Pa!D 9.37 93 22 Service Air System P&ID O 9.48 9.48 Drywell Cooling System P&ID 9.51 9.5 11 Suppression Pool Cleanup System P&ID Amendment 11 1.73 9' j
Mkk 2M6100AC Remndard Plant mme Table 1.71 PIPING AND INSTRUMENTATION AND PROCESS FLOW DIAGRAMS (Continued) 8$AR Fig Wo. Page No. Title Type 11.2 1 11.2 12 Uquid Radwaste System PFD 11.2 2 11.2 13 Uquid Radwaste System P&lD 113 1 113 21 Offgas System Prn 113 2 113 23 Offgas System P&lD O Amendment 11 1.74 i
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ABWR imi oxe we l c Remndard Plant l t Figure 1.7 2 ! INSTRUMENTATION ENGINEERING, IN'IERIDCK . BIDCK AND SINGLE LINE DIAGRAMS l esa n u 7e 7ae ; 5.24 5.244 14ak Detection and laolation System IED
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- 7.29 7.2.M Reador Protection System IED ;
7110 7.2 47 Roomor Protectica System. IBD 751 7S$2 High Pressure Core Flooder Sptem IBD -
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7S2 7159 Nuclear Boiler System IBD ; 753 73 74 Reactor Core Isolation Cooling System IBD 4 754 7S90 ResidualHeat Removal Sy tem IBD l 7.35 7S104 1.4ak Detection and Isolation System IBD . l
- .i 7S6 7S105 Standby Oas Treatment System IBD
7S7 7.3 116 Reactor Building Cooling Water System IBD , t 73 8 7&l32 Essential HVAC System IBD 7.3-9 71133 HVAC Emergency Cooling Water System IBD 7 S 10 71141 High Pressure Nitrogen Gas Sptem IBD 4 7.41 7.4-19 Standby Liquid Control System IBD ' 7.42 7.4 24 Remote Shutdown System IED - 7.43 7.4 26 Remote Shutdown System IBD ; 7.61 7.6 24 Neutron Monttoring System IED , 7.62 7.6 28 Neutron Monitoring System IBD 7.65 7.6-46 Process Radiation Monitoring System IED 7.6-6 7.6 59 Fuel PoolCooling and Cleanup System IBD 7.67 7.6 64 Containment Atmosphere Monitoring System IED 7.6-8 7.6 69 Containment Atmosphere Monitoring System IBD 7.6 11 7.6 80 Suppression Pool Temperature Monitoring System IED v
ABM ar ' ' plant swi.uc ! an e TABIJE 1.8 21(Contissed) : O INDUSTRIALCODES AND STANDARDS APPLICABIJE TO ABWR , I Code er Standard - Number Year TWe- . ANS i 2.3 1983 Standard for Entissatlag Tornado'and Other Entreme Wind l CharamerWes at Nuclear Power Sites e 2.8 1981 Determinlag Dnige Basis Flooding at Power Reactor ! Sites ; 5.1 1979 Decay Heat Poweein LWRs , , 18.1(N237) 1984 Radioactin Source Term for Normal Operation of LWRs l 32.1 1983 Nuclear Safety Design Criteria for tbc Design of l Stationary Bolling Water Reactor Plants 55.4. 1979 Gaseous Radioactive Waste Processlag Systemt for Light Water Reactors 57.1 1980 Design Requirements for LWR Fuel Handling Systems 57.2(N270) 1976 Design Requirements for LWR Spent Fuel Storage Facilities at NPP 58.2 1988 Design Basis for Protection of Light Water NPP Against l Effects of Postulated Pipe Rupture 59.51 (N195) 1976 Fuel Oil Systems for Standby Diesel Generators t t
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2&A610QAC ReamAmed Pla=* ntev. c TARG 1.8 21(Continued) INDUS'IRIAL CODES AND STANDARDS APPLICABLETO AB%R Code er Standard Number Year Title ANSI A58.1 1982 Design leads for Buildings and other Structures, Minimu:n 23.5 1960 American Standard Tolerance for Ball and Roller Bearings B30.2 1983 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Gridor, Top Running Trolley Holst) B30.11 1980 Monorailand Underhung Cranes B31.1 1986 Power Piping lC1 1985 Specifications of General Requirements for a Quality Program D975 1981 DieselFuel oils, Spec for HE! 1970 Standards for Steam Surface Condenser,6th E., Heat Exchangers Institute MC11.1 1976 Quality Standard for Instrument Air N$.12 1972 Protective Coatings (Paint) for Nuclear Industry N13.1 1%9 Guide to Sampling Altborne Radioactive Materials in Nuclear Facilities N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials N509 1980 Nuclear Power Plant Air Ge==ta: Units and Components N510 1980 Testing of Nuclear Alt f%= 2 ; Systems N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities. N101.4 1972 QA for Protective Coatings Applied to Nuclear Facilities N195 (See ANS 59.51) N237 (See ANS 18.1 0 , Anwndown: 11 1351e
ABWR siasi.a$c ; m...a..a pi.., my c ; i TABLE 1.8 21(Continued) , 4 O INDUSTRIAL CODES AND STANDARDS APPLICABLE 10 ABWR ' , l Code or i Standard Number Year Title - i
- ANSI (Com't.) .
N270 (See ANS $2.2) }
- OM3 1987 Requirements for preoperational and laltlal Startup Vibration Test Programs for Water <coled Power Plants OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant Piping Systems [ September 1986 ,
- (Draft RevWon 7)]
API i 620 1986 Rules for Design and Construction of Large, Welded, Low Pressure Storage Tanks , 650 1980 Welded SteelTanks for Oil Storage > O l f l l l I i i O Amendnient 11
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MM ReasmAmira l PRami 3AA6100AC REY. C TABM1&21(Coattamed) INDUSTRIAL CODES AND STANDARDS mucomm-Code er
. Beandard Noenbar Year 'nele ASHRAE 30 1978 , Methods of Testing LAquid Chilling Packages 33 1978 Methods of Teatlag Forced Circulation Air Cooling and Air Heatlag coils
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MM meandard Plane zwlouc am c TABM 1.8 21(Continued) O ixou 2=>itcooB ixo Tixoi o. APPLICABM TO AB%= Code er Standard Number Year litle. aus l See B 1989 BPVCSectice H Material SHrenninn. SecHI: 1989' BPVC Section III, Rules for Construction of Nuclear Power Plant Components Sec VH1. M89 BPVC Section VIII, Rules for Construction of Prestire Vessel SecIV- 1989 BPVC Section IX, Qualification Standard for Welding and Brating Procedures Welder, Brazers and Welding and Brazing Operators Sec XI 1989 BPVC Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components - ACTM (See ASME BPVC Section HI)
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ABWR m ime Renadard Plant REV. C TABLE 1A21(Coatlaned) . INDUSTRIAL CODES AND STANDARDS APPLICABLE 10 ABWR Codeer Senadard Number Year TWe. AWS A4.2 1986 _ Proceduren for Calibrating Magnetic Is.struments to Measure the Delta Ferrite content of Amstealtic Stainless Steel Weld Metal D1,1 1986 SteelStructuralWelding Code l D14.1 1985 Welding of Industrial and Mill Cranes and other Material Handling Equipment i i l l 4 i h l i
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T Amendment 8 2.8-5H
ABM >>^sioo^c
' Remndard Plant a- c
SUMMARY
OF ABWR ST PLANTINTERFACES WITH REMAINDER OF PIANT FIEM INTERFACE NO, SUBJECT TYPE SUBSECTION 1.1 Standard review plan sections for r-" of ConGrmatory 1.8.2 plant IdaadW as ' Interface
- la Table 1A19 1.2 Applicability of reguictory guides for remainder ConGrmatory 1.8.2 ,
of plant ladaarlw as" Interface'in Table l L 1 & 20 ' . 13 Emergency procedures and emergency procedures Procedural 1A3.1:
- kh%FWm 1.4 Procedures for removing safety related systems - Procedural 1A3.2 from service 1.5 Inplant radiatiom monitoring Procedural 1A33 I 2.1 Envelope of ABWR Standard Plant Site Design Design & 2.2.1 -
Parameters Connrmatory 4 O O 2.2 Standard Review Plan Site Characteristics Confirmatory 2.2.2 3.1 Site-Specific Design Basis Wind Conarmatory 333.1 e 3.2 Site-Specific Design Basis Tornado Confirmatory 333.2
- . 33 - Effect of remainder of plant structures, Connrmatory 3333 '
systems and components not designed to tornado l loads i 3.4 Flood Elevation Design 3.43.1 l l 3.5 Ground Water Elevation Daign 3.43.2 { 3.6 Protection of ultimate heat sink Confirmatory 3.5.4.1 i !' 3.7 Missels generated by natural phenomena from Confirmatory 3.5.4.2 > 4 remainder of plant 3.8 Site proximity missiles and aircraft hazards Confirmatory 3.5.43 3.9 Protection against secondary missiles inside ConGrmatory 3.5.4.4 j containment !f) 3.10 Details of pipe break analysis results Confirmatory 3.6.4.1 V and protection methods i
. Ameadawat11 1.9-2 i
-~ , . - - , , - e. , r v-e.n, % , --
L l MM - 23A6100AC Standard Plaat au c Table 1.9-1
SUMMARY
OF ABWR STANDARD PIANT INTERFACES WITH REMAINDER OF PLANT (Continued) FIEM INTERFACE I NO. SUBJECT 'IYPE SUBSECI10N 3.11 14ak.before-break analysis reaults Confirmatory . 3.6.4.2 i 3.12 Foundation WaterprooGag Conf.rmatory 3.8.6.1 3.13 Site Specific Physical Properties and Confirmatory 3.8.6.2 Foundat'u= Settlement 3J4 Reactor laternals Vibration Analysis, Confirmatory 3.9.6.4 j Measurement and laspection Pragrams !, 3.15 Equipment qualification records .Confttmatory 3.10.S.1 3.16 Dynamic qualification report Con!irn:atory 3.10.5.2 3.17 Environmental Qualification Document Confiru:atory 3.11.6.1 3.18 Enviromental Qualification Records Confirmatory 3.11.6.2 4.1 CRD Inspection Program Proredura: 4.53 5.1- Water Chemistry Design 5.2.6 l t 6.1 ExternalTemperature Con!irmatory 6.4.7.1 6.2 Meterology(X/Os) Confema:ory 6.4.7.2 l 63 Toxic Gases Confirmaiory 6.4.73-7.1 Effects of Sation Blackout on HVAC Confirma:oq 7.8.1 7.2 Electrostatic Discharge on Exposed Confirma:ory 7.8.2 Equipment Components 73 localized High Heat Spots in Semiconductor Confirmatory .7.83 Material for Computing Devices 8.1 Stability of offsite power system Confirmatory 8.1.4.1 8.2 Diesel Generator Reliability Procedural 8.1.4.2 83 ClassIE Feeder Circuits Design 8.23.1 8.4 Non classIE Feeders Design 8.23.2 O Amendment 11 1.93
J 21A6100AC Standard Plant m.m e i Table 1.91
SUMMARY
OF ABWR STANDARD PIANT INTERFACES WITH REMAINDER OF PIANT(Continued) 1 FIEM INTERFACE NO, SUDJECT TYPE SUBSECTION 8.5 Specific ABWR Standard Plant / remainder of plant Design 8.233 power sysytem interfaces - 8.6 Interupting Capability of Electrical Confirmatory - 83.4.1 Distribution Equipment 8.7 Diesel Generator Design Details Confirmatory - 83.4.2
' 8.8 Certified Proof Tests on Cable Samples Confirmatory 83.43 l
l 8.9 Electrical Penetration Assemblies Confirmatory . 83.4.4 l 8.10 Analysis Testing far Spatial Seperation Cadirmatory 83.4.5-l per IEEE 304 l 8.11 DC Voltage Analysis Confirmatory 83.4.6 8.12 Seismic Qualification of Eyewash Equipment Confirmatory 83.4.7 ! f 8.13 Diesel Generator Load Table Changes . Confirmatory 83,4.8 L 8.14 Offsite Power Supply Arrangements Procedural 83.4.9 l 8.15 Diesel Generator Qualification Tests Confirmatory 83.4.10 8.16 Defective Refurbished Circuit Breakers Confirmatory 83.4.11 ~ 8.17 Minimum Starting Vokages for Class Confirmatory 83.4.12 1E Motors 9.1 Ultimate heat sink capability Design - 9.2.17.1 9.2 Makeup water system capability Design 9.2.17.2 i 93 Contamination of DG combustion air intakes - Confirmatory 9.5.10.1 O Amendment 11 1.93.1
MM 2sA61ooAc ' neward Plane arv. c -
. APPENDIX 1A -
Q TABLE OF CONTENTS (Continued) Section Ildt East 1A.2.25 Report on Outages of Emergency Core Cooling SystemsI h Report . and Proposed Technical Specification j Changes [II.K.3(17)] 1A.216 ! 1A.2.26 Modification of Automatic Depressuri-zation System Logic . Peasibility for l Increased Diversity for Some Event . 1 Sequences [II.K.3(18)] . 1A.216 1A.2.27 Restart of Core Spray and LPCI Systems on Low LevelDesign and Modification [II.K3(21)] 1A.217 - f+ 14.2.28 Automatic Switch Over of Reactor Core Isolation Cooling System Suction - Verify Procedures and Modify De:ign [II.K3(22)) 1A.217 1A.2.29 Confirm Adequacy of Space Coo'ing for - High Pressure Coolant Injection and
'l Reactor Core Isolation Cooling Systems
[II.K3(24)) - 1A.2-18 l 1A.230 Effect of Loss of Ahernating Current I Power on Pump Seals [II.K3(25)] . 1A.219 1A.231 Verify Qualification of Accumulators on Automatic Depressurization System Valves . l [lI.K3(28)] 1A.2-19 1A.232 Revised Small Break loss Of Coolant- i Accident Methods To Show Compliance With 10 CFR PART 50, Appendix K [II.K3(30)] 1A.2 20 l 1A.233.1 Plant Specific Calculations to Show Compliance With 10-CFR Part 50.46 [lI.K3(31)] 1A.2-19 { 1A.233.2 Evaluation of Anticipated Transients with I Single Failure to Verify No Fuel Failure l [lI.K3 (44)] 1A.219 - ( l i 1Aiv O , Amendment 8 i
. _ ~ _ . _ .. . _ . - . _ _ . _ _
.i 22A6100AC I Riandard plani . REV. c APPENDIX 1A $
TABLE OF CONTENTS (Continued)_ ; i M N ! 1A.2333 Evaluate E# __ _ '-- - other than Full ADS [II.K3 (45)J 1A.219a 1A.233.4 Responding to M*Alman Conceras [II.KJ (46)] _ 1A.2-19a l 1A.234 Primary enala=# Sources Outside ra=8ain. ment Structure [III.D.1.1(1)] 1A.219a ; 1A.235 In Plant Radiation Moaltoring [1113 3(3)) 1A.2-21 ' . I 1A.236 . Control Room Habitability (Ill D3.4(1)] 1A.2 21 1A3 INTERFACES , 1A3.1 Emergency Procedures and Emergency ~ Procedures Training Program 1A31 1A3.2 Review and Modify Procedures for Removing Safety Related Systems - From Sevice 1A3-1 1A33 Implant Radiation Monitoring - "1A31 ! 1A.4 REFERENCES 1A.41-1AA ATTACHMENT ATO APPENDIX 1A: l PLANT SHIELDING TO PROVIDE ACCESS TO VITAL . AREAS AND PRO 1TCT SAFE 1Y EQUIPMENT FOR POST. ACCIDENT OPERATION [II.B.2] l 1 1AA.1 INTRODUCTION 1A A'.1 1 l 1AA.2
SUMMARY
OF SHIELDING DESIGN .1AA.2-1 REVIEW l 1AA3 CONTAINMENT DESCRIPTION AND POST ACCIDENT OPERATIONS 1AAS-1 1AA.4 DESIGN REVIEW BASES 1AA.41 (! 1A.v O . Ameadownt 11 i
ABM Sta=hrd Plant isasioore nvc 1A.2.7 Post Accident Santpling [II.B.3) position with the following exception. The upper O NRC Position limit of Activity in the samples at the time they are taken is as follows-A design and operational review of the reactor liquid sample 1 Cg /ml acciant and containment atmosphere sampling line , 5 ' syntans shall be performed to determine the capabil- gas sample 10 pC;/ml-iry of personnel to promptly obtain (less than 1 hour) ' a sample mader accident condalons without incurring
- a radianie= exposure to any individual in excess of 3 and 183/4 rom to the whole body or outremities, re-spectively. Accident conditions should assume a .
Regulatory Guide 13 or 1.4 release of fission prod-ucts, if the review indicates that personnel could not promptly and safely obtain the samples, additional ! design features or shielding should be provided to meet the aiteria. 1 A design and operational review of the radio- _ logical spedrum analysis facilities shall be performed to determine the capability to promptly quantify (in ' less than 2 hours) certain radionuclides that are indi-entors of the degree of core damage. Such radionu-clides are noble gases (which indicate cladding failure),lodines and cesiums (which indicate high '
- fuel temperatures), and nonvolatile isotopes (which indicate foci melting). The initial reactor coolant
- spectrum should correspond to a Regulatory Guide vO 13 or 1.4 release. The review should also consider the effects of direct radiation from piping and-components in the auxiliary building and possible contamination and direct radiation from airborne i effluents. If the review indicates that the analyses I
required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria. In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reac- ' tor conditions. Procedures shall be provided to perform boron and chloride chemical analyses as. suming a highly radioactive initial sample (Regula-tory Guide 13 or 1.4_ source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).
Response
The post accident sampling ssystem described in Subsection 93.2 meets the requirements of this Amendment 9 1A.2-7
ABWR .
>>4slooie Standard Plant arv c 1A.2.8 Rule Making Proceeding or the tests. The SRVs tested were typical of models Degraded Corc Accidents [II.B.8] used in domestic BWRs. The results of this testing verified the adequacy of SRV operation and the -
Response to this TM1 action plan item is ad. integrity of the SRV and discharge pining under
~
dressed in Appendix 19A. expected liquid discharge conditions. IA.2.9 Coolant System Valves. Testing The water discharge test results noted above will Requirements [II.D.1] be used as the qualification basis for plant specific SRV models and discharge piping that are NRC Position sufficiently similar to those reported in Reference &. Plant specific SRV models and discharge piping that Pressurized. water reactor and boiling water are not similar will be tested in accordance with reactor licensees and applicants shall conduct testing NUREG-0737 requirements. to qualify the reactor coolant system relief and safety valves under expected operating conditions for 1A.2.10 Relief and SafetyValve Position . design basis transients and accidents. Indication [II.D.3] : Response NRC Position The ABWR safety / relief valve (SRV) is postu. Reactor coolant system relief and safety valves lated to discharge steam only, not liquid or two phase shall be provided with a positive indication in the ; flow, control room derived from a reliable valve position 1 detection device or a reliable indication of flow in the The ABWR system logic for response to high discharge pipe, water level conditions is described in Subsection 7.3.1.1.1.1(3) and is considered to be sufficiently Response ; redundant that the probability of steam line flooding by ECCS is extremely low. There is no high drywell- The ABWR Standard Plant safety relief valves - pressure signal that would inhibit this logic system. are equipped with linear variable differential trans- , formers (LVDT's) which' are qualified as Class 1E - In the ABWR design, each of three RHR shut- components. These LVDT's are mounted on'the down cooling lines has its own separate containment valve operators and are highly reliable sensors for penetration and its own separate source of suction monitoring valve position. from the reactor vessel. Alternate shutdown using the SRV is therefore not required for ABWR in In addition, the downstream pipe from each valve order to meet single failure rules. line is equipped with thermocouples which signal the , annunciator and the plant process computer when ' Reference 7 documents a series of water the temperature in the tailpipe exceeds the predeter- ; discharge tests of SRVs used in BWRs. These tests mined setpoint. were conducted to demonstrate the operational , adequacy of the SRV and integrity of the SRV These sensors are shown on Figure 5.13 ) discharge piping for expected operating conditions (Nuclear Boiler System P&ID). for transients and accidents. The tests were performed to satisfy requirements of II.D.1 of. 1A.2.11 Systems Reliability [II.E.3.2] . NUREG 0737. In these tests the SRVs were mounted on representative steam lines, and This TMI action plan item superseded by USI representative discharge lines were routed to a pool A 45. USI A-45 is addressed in Appendix 19B. - j of water. The open and closing of the SRVs was 4 monitored, Fluid conditions and flows were 1A.2.12 Coordinated Study of Shutdown , measured along with strains, accelerations, Heat Removal Requirements [II.E.3.3] temperatures and pressures in the SRVs and associated piping. The SRVs were operated under This TMI action plan item superseded by USI flow conditions of both steam and water flow during A 45. USI A 45 is addressed in Appendix 19B. , Amendment 11 1A.2-8
ABWR - DAH00AC Standard Plant arv c 1A.2.13 Containment Design. Dedicated Penetration [lI.E.4.1) NRC Position For plant designs with external hydrogen recombiners, provide redundant dedicated contain. ment penetrations eo that, assuming a single failure, the recombiner systems can be connected to the con. tainment atmosphere.
Response
A flammability control system (FCS T49) is provided to control the concentration of oxygen in the primary containment. The FCS utilizes internal recombiners with externally located blowers to provide any required mixing between the drywell and wetwc!!. The FCS will be operable in the event of a single active failure. The FCS is described in Subsection 6.2.5. 1A.2.14 Containment Design. Isolation Dependability [lI.E.4.2) hTC Position (1) Containment isolation system designs shall com.
. ply with the recommendations of Standard Re. -
view Plan Subsection 6.2.4 (i.e., that there be di. l versity in the parameters sensed for the ini. Listion of containment isolation). (2) All plant personnel shall give careful consid. eration to the definition of essential and non. essential systems, identify each system determined to be nonessential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC. i (3) All nonessential systems shall be automatically l
' isolated by the containmentisolation signal. !
l (4) The design of control systems for automatic I containment isolation valves shall be such that reseting the isolation signal will not result in the Amendment 11 1A.2-8a l 1
ABM >>4si.ase i Remadard Plant nve automatic reopening of containment isolation - (5) The ABWR Standard Plant design is consistant : valves. Reopening of containment isolation with this position. I valves shall require deliberate operator action. ) (6) All ABWR costainment purge valves meet the - ; (5) The containment setpoint pressure that initiates criteria provided in BTP CSB 6 4. 'the main 22* aa=*=&aaient isolation for nonessential penetra- purge valves are fall closed and are maintained tions must be reduced to the minimum compat- closed through power operation as denned in the l ble with normal operating aaadiriaan plant technical specifications. All purge valves j are remote operated, fait closed and receive i (6) raneal==aat purge valves that do not satisfy the cantal==ent isolation signals. Certain valves can I operability criteria set forth in Branch Techale=1 be opened in the presents of an isolation signal, Position CSB 6 4 or the Staff interim Position of but these valves are keylocked sad under October 23,1979 must be sealed closed as de- ed=lal=*rative control. l Aned in SRP 6.2.4, Item II.3.f during operational . conditions 1,2,3, and 4. Furthermore, these (7)In ABWR design, the HVAC system isolation i valves must be verified to be closed at least valves in the secondary containment close I every 31 days. automatically on high radiation signals from the 1 refueling handling area or the reactor building : j (7) Containment purge and vent isolation valves exhaust lines going to the stack. The closure of ' I must close on a high radiation signal, the HVAC supply and exhaust lines restrict any radiation going out to the environment. Also. on > Response high radiation signal, the standby gas treatment system is initiated and stops any untreated gas or (1) The isolation provisions described in the Stan- radiation going out to the environment.
~
dard Review Plan, Subsection 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation) were re- . siewed in conjunction with the ABWR Standard Plant design, it was determined that the ABWR Standard Plan is designed in accordance with these recommendations of the SRP. (2) This request appears to be directed primarily toward operating plants. However, the classifi-cation of structures, systems and components for the ABWR Standard Plant design is addressed in Section 3.2 of this SSAR. The basis for classi- .( ! fication is also presented in Section 3.2. The ABWR Standard Plant fully conforms with the NRC position so far as it relates to the new equipment supplier. (3) All non essential systems comply with the NRC position to automatically isolate by the contain-ment isolation signals, and by redundant safety grade isolation valves, s (4) Control systems for automatic containment iso-lation valves are designed in accordance with this position for the ABWR Standard Plant Design. G: l Amendment 9 1A.2-9
.____._ __ . _ - . . - _ . _ - - . _ _ _ _ . - - _ _ _ _ _ _ - - - - - - _ _ _ - - _ _ _ _ _ - ~ - - - - - - - _ _ _
l ABWR . ziasi. orc I Standard Plant am c -
] ATTACHMENTA CONTENTS i l
Section 21tig Eass j 1AA PIANT SHIELDING M PROVIDE ACCESS 'ID VITAL AREAS - AND PROTECT SAFETY EQUIPMENT FOR POST. ACCIDENT OPERATION [II.BJ) ! 1AA.1 Intradmetiam 1AA.11 . 1AA.2 Emannary Of Shiniding Design Rerisw 1AA.21 ,
- 1AAJ Cantalement Descriptian and Past Accident Operntlans 1AA3-1 l - 1AA3.1 - Description of Primary / Secondary .
Containment - 1AA3-1 ( i 1A A 3.2 Vital Area and Systems ' 1AA3 . 1AA33 Post Accident Operations- 1AAJ1 l l 1AA.4 Design Review Bases 1AA.4-1 O ' Q 1AA.4.1 Radioactive Source Term and Dose Rates 1AA.41 1AA.4.2 Accidents Used as the Basis for the Specified , Radioactivity Release - 1AA.4-1 1AA.43 ' .vailability of Off Site Power 1A A.4-1 1AA.4.4 Radiation Qualification Conditions 1AA.41 1AA.5 Results of the Review 1AA51 1AA.S.1 Systems Required Post Accident 1AA51 , 1AA.5.1.1 Necessary Post Accident Functions and Systems IAA51
- l. 1AA.5.1.2 Emergency Core Cooling and Auxiliaries 1AA.5-1 1AA.S.13 Combustible Gas ControlSystems and 1 i Auxiliaries 1AA52 1AA.5.1.4 Fission Product Removal and Control Systems and Auxiharles 1AA.5-2 1AA il Aseendment 11
-- . - . , , a,..
-i M anAstcQAC n- c .
Reandard Plant l l I ATTACHMENT A - l CONTENTS (Continued)' l i' assden Bus East 1AA.5.1.5 Instrumentatin= and Control, Power and Habitability Systems and Audliaries 1AA52 j TABLES' - IAlllt Edf ESEE 1
- 1AA.41 Radiation Source Comparision 1AA.4 3 '-
1AA51 Post Accident Emergency Core Cooling Systems and Auxiliaries 1AAS4 1AA52 ~ Post Accident Combustible Gas Controf Systems and Auxiliaries 1AA56 e' 1AA53 Post Accident Fission Product Removal and Control Systems and Auxiliaries 1AA57 1AAS4 Post Accident lastrumentation and Controls, Power and ~ - Habitability Systems and Auxiliaries 1AA58 1AA lii > Amendment 11 . I t L - - - . . . - . _ ,, , .
.AB M 234siooxc Standard Plant mu e IAA.2
SUMMARY
OF SHIELDING (1) The period of interest begins with the plau in a DESIGN REVIEW safe shutdown condition. Thus, the various safety related systems needed to achieve safe shutdown Several alternatives are potentially available to conditions have performed, and only the the designer to assure continued equipment engineered safety features systems (Chapter 6) availability and performance under post accident and auxiliaries, as described later, are required to conditions. One is to provide redundant systems maintain this condition. and/or conoponents which are qualified to operate in the expected environment. Another is to provide (2)-. Based upon the accident source terms of operator access to conduct the operations and to Regulatory Guides 1.3,1.7 and Standard Review maintain the equipment. This latter alternative' Plan 15.6.5, and normal operations the vital would generally be accompanied by appropriate equipment exposures will be within a paximum ! ahielding and in many cases would be difficult if not required envelope exposure of 3.6 x 10 Rgds for ; impossible to carry out, equipment in primary coatainment,9 510 Rads for equipment in ECCS rooms,9x10 in SGTS : General Electric has taken the first approach rooms and pumps and valves per Tables 31.3-6, j and furthermore has designed the plant so that most 31.3-7,31.315, and 31.316 where the integrated responses to transient conditions are automatic, exposure is for six months. All vital equipment including achieving and maintaining safe. shutdown will be environmentally qualified. This exposure conditions.; The design basis for the ABWR envelope is not time dependent after 100 days, j Standard Plant is to require safety related equipment ! to be appropriately environmentally qualified and (3) ; it is not necessary for operating personnel to have operable from the control,toom, As a result of this access to any place other than the control room, design philosophy and as shown by this review, no the technical support center, the post accident changes ate necessary to assure that personnel sampling station, the sample analysis area, and the access is adequate or that safety equipment is not safety related nitrogen supply bottles to operate degraded because of post accident operation. the equipment ofinterest during the 100 day period. The control room, technical center and As part of the design of the ABWR Standard sample analysis area are designed to be accessible Plant it was necessary to establish the environmental post accident. The latter areas are considered conditions for qualification of safety related accessible on a controlled exposure basis. Direct < equipment. A result of this design work was an shine from the containment is less than 0.5 R/hr I environmental requirement establishing the within four houts post. accident. integrated dose that the equipment must be able to l withstand. These values are listed in Appendix 31. (4) Access to radwaste is not required, but the , radwaste building is accessible since primary l Another aspect of the review was the manner in containment sump discharges are isolated and which the safety related equipment is arranged and secondary containment sump pump power is shed operated during normal and abnormal operation and at the onset of the accident. Thus, fission postulated accidents.' The essence of the ABWR products are not transported to radwaste. The Standard Plant is to achieve and maintain a safe combustible gas control system is operated from shutdown condition for all postulated accident the' control room; the ABWR does not have a conditions with all operator actions being conducted containment isolation reset control area or a from outside the primary and secondary containment manual ECCS alignment area. These functions zones, principally from the control room. are provided in the control room. The purposes of this review is first to verify that (5) Following an accident, access is available to
& c equipment access is required, it is reasonably electrical equipment rooms containing motor accessible outside the primary and secondary control centers and corridors in the upper reactor containment zones.' Secondly, the review should building Section 12.3.6. This is based on ,
verify that inaccessible equipment is environmentally radiation shine from the ECCS rooms and qualified and is operable from the control room, primary containment; there is no airborne The results of the review are: Amendment 11 1AA.2-1 l 1
$$ **^'"c Standard Plant = c radiation source in the electrical equipment
. rooms and ECCS corridor areas. While not necessary to maintala safe shutdown, such access can be useful in extending system .
"'""^y and is plant m.
1 (6)' The emergency power. supplies (diesel I gesorators) ars -h However, access is -1 act accessary since the equipment is 1 environmentaDy gn.11rmt i 0
- f i
O Almendment 10 M2
ABMR sasi=4e , Standard Plant n,; c 1AA.3 CONTAINMENT failures can be repaired thereby improving systems , n DESCRIPTION AND POST availability. 1 V ACCIDENT OPERATIONS 1AA.3.2 Vital Area and Systents 1AA.3.1 Deuription of Primary / i Secondary Contalement A vital area is any area which will or may - require occupancy to permit an operator to aid in the The ABWR design includes many features to mitigation of or recovery from an accident. Areas assure that personnel occupancy is not unduly which must be considered as vital after an accident limited and safety equipment is not degraded by are the control room, technical support center, i post accident radiation fields. These features are sampling station, sample analysis area and the HPIN detailed in the SSAR and only a brief summary nitrogen supply bottles. description and SSAR reference are provided here . for emphasis. The vital areas also include consideration (in accord with NUREG. 0737, II.B.2) of the post 4DCA . The configuration of the pressure suppression hydrogen control system, maralament isolation reset primary containment with the suppression pool , control area, manual ECCS alignment area, motor , mart =1*s the scrubbing action of fission products by control center and radwaste control panels. However, the suppression pool. The particulate and halogen the ABWR design does not require a containment > content of the primary containment atmosphere isolation reset control area or a manual ECCS -
- following an accident is thereby substantially reduced alignment area as these functions are available from compared to the Regulatory Guide 1.3 source terms, the control room. Those vital areas which are normally areas of mild environment allowing Primary containment leakage is limited to less unlimited access are not reviewed for access.
than one half percent of the primary containment
- atmosphere per day. The surrounding secondary Essential systems specific to the ABWR to be I containment is kept at a negative pressure with- considered post. accident are those for the ECCS, respect to the environment permitting monitoring fission product and combustible gas control and the j and treating all radioactive leakage from the primary auxiliary systems necessary for their operation (i.e.,
i containment. Instrumentation, control and monitoring, power, I cooling water, and air cooling). The standby gas treatment system (SGTS) operates automatically from the beginning of an 1AA.3.3 Post Accident Operation accident to control the secondary containment pressure to (.)1/4' w.g. The large volume of this Post accident operations are those necessary to . portion of the reactor building acts as a mixing 1) maintain the reactor in a safe shutdown condition, chamber to dilure any primary containment leakage 2) maintsin adequate core cooling,3) assure before processing by the SGTS and discharge to the containment integrity and 4) control radioactive environment. Discharge of radioactivityis thus releases within 10CFR100 guidelines. controlled and reduced. Radioactivity content of secondary containment atmosphere is reduced with Many of the safety related systems are required time by SGTS treatment as well as by decay. for reactor protection or to achieve a safe shutdown , (However, prior removal of halogens by scrubbing in - condition. However, they are not necessarily needed the suppression pool offsets the necessity of this once a safe shutdown condition is achieved. Thus, the treatment), systems considered herein are the engineered safety features (ESP) (see Chapter 6) used to maintain the i Each ECCS pump and supporting equipment is plant in a safe shutdown condition. l ' located in an Individual shielded, watertight compartment. Spread of radioactivity among For purposes of this review the plant is assumed compartments is thus limited. Radiation to the other to remain in the safe shutdown condition, equipment areas and corridors of the reactor building is limited to shine through the walls; there is The basis fo this position is that the foundation O no airborac radiation in these other areas. As these become accessible after an accident, any component of plant safety is the provision of sufficient-
' Amendment 10 1AA.31 l
a.
i 1 ABM . innsinasc-
- hndard Plant w c- ,
I redundancy of systems and logic to assure that the plant is abut down and that adequate core cooling is
)
l maintained. Necessary shutdown and post acddent _j operations are performed from the control room, ;
- except for the post accident sampling station, the i sample analysis area, and two manual nitrogen lresem supg% h l-i t
i c Anwadawat 11 1AA32 .,~. .- -
ABWR siasionac Standned Plant a- c 1AAA DESIGN REVIEWBASES The postulated control rod drop accident I O- 1AA.4.1 Radioactive source Term and (Suhauvina 15.4.9) is one which occurs without a pipe break and so may require depressurization to attain Dose Rates long term core cooling with the RHR system. Normal 3y, this accident is terminated by a seam, and De radioactive source term used is equivalent the plant is cooled and recovers. De performance of to the source terms recommended in Regulatory the separation detection devices and the rod block Guides 13 and 1.7 and Standard Review Plan 15.6.5 interlocks virtually preclude the cause of a rod drop q with appropriate decay times. Depressurized eaalant accident. This accident is not further considered. 4 is assumed to contain no noble gas. There is no leakage outside of secondary containment other than The DBA LOCA is the accident producing the o via SG11i. limiting constitinas of interest for this design review. j In this accident the reactor is depressuriced and ; Done rates for areas requiring continuous reactor water mixes with suppression pool water in -p occupancy may be averaged over 30 days to achieve the process of keeping the fuel covered and cooled, the desired <15 mrem / hour. Fission products are assumed to be essentially instantaneously released and mixed in the Design dose rates for personnel in a vital area containment atmospheres and suppression are such that the guidelines of General Design pool reactor water volumes. Criteria (GDC) 19 (i.e., <5 Rem whole body or its equivalent to any part of the body) are not exceeded 1AA.4.3 Availability of Off Site Power for the duration of the accident, based upon expected l occupancy and protection. The availability of off site power is not influenced by plant accident conditions. Loss of 1AA.4.2 Accidents Used as the Basis for off site power may be assumed as occurring the Specified Radioactivity Release coincident with the beginning of the accident sequence, however, continued absence of off site Table 15.0 3 summarizes the various design power for the accident duration is not realistic, While basis accidents and associated potential for fuct rod restoration of off site power is not a necessary failure. This chapter also provides the accident condition for maintaining core cooling,its availability parameters. Of those accidents only the can permit operatie. of othr plant systems which DBA LOCA may produce 100% failed fuel rods would not otherwise be p.mitted by emergency i under NRC worst case assumptions. The rod drop power restrictions, e.g. operation of the pneumatic air i accident and fuel handling accident are the only system, non safety related HVAC systems and other ' other accidents postulated as leading to failed fuel systems useful to plant cleanup and recovery, rods with the potential consequence of radioactivity releases. Based on Table 19D3 3 of Section 19DS.5, the probability for off site power recovery is 0.983 in 8 For the fuel handling accident, the reactor is hours. This is con:.ervative since the bagest time for either shutdown and cooled or is operating normally restoration of off site power was six hours for the if the accident is in the spent fuel storage pool, Pennsylvania New Jersey Maryland interconnection, Based on the conditions of Regulatory Guide 1.25, it the grid used as a basis for the probabilistic risk-is assumed that the airborne activity of the reactor . == ara ==ent presented in Section 15D3. building (Table 15.7 9) is released to the environment over a 2 hour period via a 99% iodine efficient SGTS. The total activity released to the 1AA.4.4 Radiation Qualification environment is presented in Table 15.710 and the Conditions calculated exposure in Table 15.711. The exposures are within the guidelines of 10CFR100. Thus, The safety related :quipment requiring review recovery is possible well within the specified 100 day for qualification is only that necessary for equipment qualification period. ECCS equipment is post accident operations and for providing not affected by this accident and radiation in the information for assuring post accidcnt control. O ECCS area is not increased. This accident is not considered further. Amendment 11 1AA.41 i
I MMRennAard Plant asAstooAc a- e la 10CFR$0 the long term cooling capability is ' given as follows:" decay heat shall be removed for - (. the estended period of time required by the long 1 lived radioactivity remaining in the core? A 100 day l period period has been permitting site and selected astoa sufficient extended facility response rmaate ihe e,s.t. As part of the design review process a set of reference conditions is necessary for comparing expected post accident radiation exposures. Appendix 31 defines the cavironacatal conditions for safety related equipment zones for periods of 60 years normal operaticas, including anticipated tests and abnormal events, and six months following the ! a DBA LOCA. These conditions are upper bound ' envelopes need to establish the environmental design and qualification bases of safety related equipment, In effect these are specification values, and ! equipment will be qualified to meet or exceed these values. The environmental zones shown in the above tables are defined in Table 31.21. Radiation sources la the secondary 1 containment (especially the ECCS rooms of the reactor building) are the same as the Table 1AA 1 design basis values for water sources. For airborne radiation sources the plant design basis of Table { j 1AA 1 for air is used. Primary containment leakage is assumed to occur in each of the individual d
' secondary containment compartments. This leakage is limited by the fission product control systems (Subsection 6.5.3). As previously noted, no credit has been taken for the radio halogen scrubbing ,
which is an inherent feature of the BWR. 1 Amendment 11 1AA.4 2
MN
- Rinnalard Plant -
z w tooAc m, e l Table 1AA.41 - RADIATION SOURCE COMPARISON Attivity % Core lavsetery Released Group R.G. I.3 R.G,1.7 Plant Design i Basis ; g-N oble G ases 100 100 100* Halogens - 25 -- 25' All Remaining - - .. f WalCI Noble Gases 0 .. 100 Halogens - 50 50** All Remaining - 1 -1**
- Uniformly mixed within the primary containment boundary
*
- Uniformly mised in the suppression pool and reactor coolant t
4 O : Asnandesat 11 1AA.&3 4 t s -, -- ,- s .e e- . .
ABWR =oore
- Standard Plant n~ c 1AA.5 RESULTS OF THE REVIEW Tablu 1AA 2 thru 5 are generated: l 1AA.5.1 Systents Required Post Accident (i) to show what ador equipment and systems are required to function and thereby define the This section establishes the various systems systems for review,and l equipment which are required to function following an accident along with their 6*6 The espected (ii) to show the redundnat equipment locations by habitability conditions and access and control needs eivisional isolated rcom or erea aad are beh4 for the required post accident period. - containmect er kW 1AA.8.1.1 Necessary Post Accident Fametless and 1AA.S.1J Emergency Core Coollag Systems and Systems Ausiliaries Following an accident and assuming that Table 1AA 2 iists various systems related to immediate plant recovery is not possible, the cooling the fuel under post accident conditions as following functions
- are necessary, described in Section 6.3 and Subsection 9.4.5.2 HVAC. This table shows ECCS equipment and -
(1) Reactivity control equipment coolers in an ECCS room.- Instrumentation transmitters are in adjoining areas. (2) Reactor core coohng The required power and cooling water in the sante division are described in Section 1AA.5.1.5. All (3) Reactor coolant system integrity perform together to provide an ECCS function. (4) Primary reactor containment integrity, and The automatic depressurization system (ADS) ; function is described in Subsection 1.2.2.4.8.2. A (5) Radioactive effluent control postulated non break or small break accident could require continued need for the depressurization Reactivity controlis a short term function and function until the RHR system is placed in the is achieved when the reactor is shutdown. The shutdown reactor cooling mode. In the case of a , remaining functions are achieved in the longer term non break or a small break accident, the majority of post accident period by use of: the fission products would be released via the safety : relief valves to the suppression pool and hence to the . (s) The emergency core cooling system (ECCS) containment rather than direct mixing through the and their auxiliaries (for reactor core cooling). supersession pool vents as would occur following a DBA LOCA. In either cisse the' distribution of fission , (b) The combustible gas control system (CGCS) products is assumed to be the same as for the ! and auxiliaries (for primary containment and DBA LOCA even though realistically a significant ., reactor coolant system integrity), portion of halogens and solid fission products would be retained in the reactor pressure vessel. Thus, the (c) The fission product removal and control system rer,ults as they apply to the ADS are very conservative. and auxiliaries (for radioactive effluent The pneumatic nitrogen supply for the ADS and control), and other containment valves is included in Table 1AA 3 ' as a portion of the combustible gas control. The hand (d) Instrumentation and controls and power for operated nitrogen reserve supply valves P54 F017C accident monitoring and functioning of the and D are accessible outside the secondary necessary systems and associated habitability containment, if needed, to mitigate a large leak. systems. The high pressure core flooder (HPCF) and tlie low pressure flooder (LPFL) functions are described
- ANSI /ANS 4.5 Criteria for Accident Monitoring Functions in in Subsection 1.2.2.4.8.1.1 and 1.2.2.4.8.3 respectively.
Lisht Water Ructon The coc, ling function can also satisfy the containment cooling function in that by cooling suppression pool O water, which is the source of water flowing to the reactor, the containment source of heat is also Amendment 11 1AA.51
ABM Standard Plant s w.iarc a- c removed. The wetwell/drywsil sprays are described Engineered safety feature Alter sy.tems are the in Subsecdon 1.2.2.4.9.4. atandby gas treatmout system (SGTS) and the control building outdoor air cleanup system. Both consist of Tbc feel pool cooling function (Subsection redundant systems designed for accident conditions 1.2.2.8.2) is also included ce the basis that a recently and are controlled from the control room, ne SOTS unloaded fuel betch could require continued cooling filters the gaseous effluent from the primary and during the post. accident period. The equipment is secondary containment when required to limit the environmentally qualified so access la not required discharge of radioactivity to the environment. The and redundancy is included in system components. system funaion is denaribed in Sukanetaa 1.2.2.4.1.6. The location of selected associated valves and A portlos of the control building heating lastrument transmitters are included. These do not ventilating and air-conditioning (HVAC) provides represent all of this type of equipment which is - detection and limits the introduction of radioactive - environmentally quahfied, safety related, or included material and smoke into the control room.- This la the systems of Table 3.21. It does however, portion is described Subsectica 9.4,1.1.3. represent principal components which are needed to operate, generally during post accident operations. The CAMS described in the previous section For example, most ECCS system valves are normally also measures and records containment area radiation l open, and only a pump discharge valve needs to open under post accident conditions. A post accident ' to direct water to the reactor. Similarly, the sampling system (PASS) obtalas containment lastrument transmitters shown are those which atmosphere and reactor water samples for chemical would provido information on long term system and radiochemical analysis in the laboratory. Delayed performance post accident. Control roo:n sampling, shielding, remote operated valves and instrumentation is not listed since it is all in an sample transporting casks are utilized to reduce accessible area where no irradiation degradation radiation exposure.' The samples are manually would be expected. Passive elementa such as transported between the PASS room in the reactor thermocouples and flow sensors are not listed building and the analysis laboratory in the service although they are environmentally qualified. The building. The system is' described Subsection components listed under main steam (B21) are those 9.3.2.3.1. Table 1AA 4 lists the fission product for ECCS function or monitoring reactor vessel level. removal contr01 components and locations. t 8 Suppression poollevelis included with the HPCF instrumentation. 1AA.5.13 Instrumentation and Control, Power, and l
- Habitability Systems and Annularies 1AA.S.I.3 Combustible Gas Control Systems and Auxillaties Most of the post accident instrumentation and control system equipment is listed with the applicable i Flammability control in the primaty equipment in Table 1AA 2,1AA 3 and 1AA 4. The containment is achieved by an inert atmosphere remaining instrumentation and control equipment is during all plant operating modes except operator included wit'h the power and habitability systems access for refueling and maintenance and a equipment listed in Table 1AA 5. Instrumentation is recombiner system to control oxygen produced by consistent with the post accident phase variables radiolysis. The high pressure nitrogen (HPIN) gas monitored by the post accident monitoring (PAM) supply is described in Subsection 1.2.2.8.8. The system listed in Table 7.5.2.
containment atmospheric monitoring system (CAMS) measures and records containment Standby AC power is supplied by three diesel oxygen / hydrogen concentrations under post accident generators in separate electrical divisions as described conditions. It is automatically initiated by detection in Subsection 1.2.2.4.1.8. The diesel generators, ofloss of coolant accident (LOCA) and is described switchgear and motor control centers are included in in Subsection 7.6.1.6. Table 1AA 3 lists the the unit Class 1E AC power system described in combustible gas control principal components and Subsection 1.2.2.5.1.2. Storage batteries are the their locations. standby power source for the unit Class 1E DC power system described in Subsection 1.2.2.5.1.7. The safety 1AA.5.1.4 Fission Product Removal and Control system logic and control power system is described in Systems and Auxillaries Subsection 1.2.2.5.1.3. Amendment It 1AA.5-2
.AB M a u .io m e Se==dard Plant am c Habitability systems ensure that the operator O
can remain in the control room and take appropriate. Action for post accident operations. The control
~ buildii I =*a==gincludes
>r= or==tiasthallo rthe instrumentation a =>raa=ir d = d- and controls
!- post acddest aanh he control room, control and reactor building HVAC essential equipecat are a portion of the plant environmental control of temperature, pressare, humidity and airborne contaalaation descrbed in Subsection 1.2.2A.10(1), (4), (5), (7) and (8). HVAC units controllag the local room environments are lacluded with respective equipment in Tables 1AA 2,1AA 3 and 1AA 4. The major HVAC equipment and locations are listed in Table 1AA 5. The reactor building cooling water (RBCW) I system provides cooling water to designated equipment in the reactor building including containment as described in Subsection 1.2.2.8.1. The HVAC emergency cooling water (HECW) system provides chilled water to designated equipment in the control building as described in Subsection 1.2.2.8.4. O Amendment 11 1AA.5-3
. = - - ..- . - . . - .. .-.- -.- .. - .
ABWR *$^'2" standard Plant "" C Table IAA.51 j POST ACCIDENT EMERGENCY CORE
+
COOLING SYSTEMS AND AUXILIARIES Egalpment MPL Laention ADS & Trannniners SR Valm - B21 F010A,C,F,H,L,N,R,T Upper Drywell (PC) > Solenoid Valve- B21 F101L thru U . Upper Drywell (PC) l
$+aaM Valve B21 P102L thru U _ Upper Drywell (PC)
SR Accumulator B21.A004A,C,F,H,L,N,R,T Upper Drywell (PC) - Rx Water level (ADS,RHR) B21 L1003A thru H Instrument Rack Rm (SC) Rx Waterlevel(NPCF) . B21 LTV01A,B,CD Instrument Rack Rm (SC) Rx Press. (RHR) B21.IT301A,B,C,D - Instrument Rack Rm (SC) DW Press. (HPCF, RHR) B21 ITl06A,B,C,D : Instrument Rack Rm (SC) s a HP.2 i Pumps E22 C001B,C HPCF Rm. B,C (SC) ; SP Suction Valve . E22 F006B,C HPCFRm B,C(SC) -1 RxInjection Valve E22 F003B,C Valve Rm. B,C (SC) CST Suction Valve E22-F001B,C Valve Rm. B,C (SC) . EssentialHVH(HVAC) . U41 D102,106 - HPCF Rm. B,C (SC) CST Water Level P13 L1V01A,B,C,D HPCF Rm. B,C (SC) SP Water Level T31 LT061,062,063,064 By HPCF Rm. B,C (SC) ', Flow E22 FT008B1,B2,C1,C2 By HPCF Rm B,C (SC)- lO 4 Suction Pressure lajection Pressure E22-PTV02,003;B,C E22 PT006,007;B,C By HPCF Rm B,C (SC) By HPCF Rm B,C (SC) LP.2 Pump' E11 C001A,B,C RHR Rm. A,B,C (SC) Heat Exchanger Ell l> '1A,B,C - RHR Rm. A,B,C (SC) 7 i RCW Dischargs Valve P21 F013A,B,C ' RHR Rm. A,B,C (SC) - ! SP Suction Valve E11 F001A,B,C RHR Rm. A,B,C (SC); SP Return Valve E11 F088A,B,C _ RHR Rm. A,B,C (SC) -
- l Rx Injection Valve E11 F005A,B,C Valve Rm. A,B,C (SC) ;
Rx Return Valve E11 F010,011,012;A,B,C ' Valve Rm A,B,C (SC) l DW Spray Valve E11 F017,018;B,C Valve Rm B,C(SC) ! WW SprayValve E11 F019B,C Valve Rm. B,C (SC) _ FPC Supply Valve E11 F015B,C Valve Rm. B,C (SC) FPS Supply Valve E11 F101,102,103 Valve Rm. B,C (SC) . I Essential HVH(HVAC) U41 D103,104,105 RHR Rm. A,B,C (SC) Flow' E11 FT008A1,B1,C1 By RHR Rm. A,B,C (SC) Flow E11 FT008A2,B2.C2 By RHR Rm. A,B,C (SC) * , RCW Flow P21 FT008A,B C By RHR Rm. A,B,C (SC)
- HxI/O Temperature E111T006,007;A,B,C By RHR Rm. A,B,C (SC)
Discharge Pressure . E11 P1V04A thru F By RHR Rm. A,B,C (SC) DW Temperaturr T31 TT/SSA051,053 Inst. Rack Rm. (SC) - DW/WW Pressure Ratio T31 PT057,059 Inst. Rack Rm. (SC) WW Pressure 131-P1V58 Inst. Rack Rm (SC) DW Pressure 131 PT054 Inst. Rack Rm. (SC)
~
SP Temperature T53-1Txxx By Supp. Pool (SC) Anwndment 11 1AAS4
MM Reandard Pimat
' amicoAc
- m, e i
. Table 1AA.51 ,
g POST ACCIDENT EMERGENCY CORE - ! COOLING SYSTEMS AND AUXILIARIES (Contiamed) ! t l
- Egelpment MPL tmeties EECS Pump G410001A,B ' FPC Pump Rm. (SC)
Heat Exchanger 041 B001A,B FPC Hz Rs. (SC) i Pump Discharge Valve L G41 P021A,B FPC Valve Rm. (SC) i Essential HVH (HVAC) - U41 D107,108 - FPC Valve Rm. (SC) , Flow -- _ G41171)06A,B By PPC Panp Rm. (SC) 1 Suction Pressure G41 P11X)3A,B By FPC Pump Rm. (SC) ' Sklanner STlevel ' G41 L'It:20A,B Refueling Floor (SC) i l' i F i O" l' ,
~!
s b (PC)- Primary containment (SC) Secondarycontainment O Amendment 11 1AA.5-5 l
MM swlooAc Standard _ Plant m.v. e Table L4A.5-2 , POST ACCIDENT COMBUSTIBLE GAS CONTROL SYSTEMS AND AUXILIARIES Equipment MPL laesties - HI21 Nitrogen Storage Bottles P%A001A'!hru V ByVals Rm (RB) Bottle SupplyValm(Hand) P%P017C,D By Valw Rm (RB)- Supply Valm - P%P018A,E ByValw Rm (RB) Supply Pressure 'P % Pitl06A,B By Valve Rm (RB) 112i Recombiner & Audliaries T49-A001A,B (PC) RHR Cooling /IsolValve T49-P008,010;A,B . (PC)(SC) Flow T49-Fitl02,004;A,B Inst. Rack Rm. A,B (SC) ' Pressure- T49 PTV03A,B Inst. Rack Rm. A,B (SC) .. C&MS Hydrogen, Oxygen Elements D23 Hf,0 Rack 3 A,B ^ CAMS Rm.' A,B (SC) Gas Measurement D23-GIs CE. Rack A,B CAMS Rm. A,B (SC) . Gas Elements D23-Gas Cal. Rack A,B CAMS Rm A,B (SC) . DW Gas Valve D23 F004A,B- . CAMS Rm. A,B (SC) WW Gas Valve D23 F006A,B CAMS Rm.-A,B (SC) U41 D113,114 O Essential HVH(HVAC) Gas Supply CAMS Rm. A,B (SC) D23-Gas Cyl Rack A,B CAMS Rm. A,B (RB) (PC)-Primary containment (SC). Secondary containment (RB)- Reactor building outside (Secondary Containment) O , Amendewat 11 1AAS6 l
M ' 28A6100AC [ Sinndard Plani n.m c Table IAA.5 3 h.' POST ACCIDENT FISSION PRODUCT REMOVAL AND CONTROL SYS'IEMS AND AUXILIARIES j Egelpment MPL Imeeties 3023 , Exhaust Fan T22 0001A,B Fan / Dryer Rs. (SC) , i Dryer 'IDD001A,B Fan /Dryor Rm. (SC)' , CharcoalFiber. ~ '!22 D002 - Filter Trsin Rm. (SC) - PClelet Valve T221902A,B Fan / Dryer Rm. (SC),
~ SClalet Valve : T22 P001A,8 . Fan / Dryer Rm. (SC) l- Stack Outlet Valve 722-F004A,8 ' . Piker Train Rm. (SC)
PC(DW,WW) Isolation Valves T31.P004,006,006 Valve Rm. (SC) Fawnslal HVH(HVAC) U41 D111,112 - . ' SG'!1 HVH Rm.'(SC)'
. Radiation (Ion /Scint.) . D11 RB002,011;A,B ~ SGTS Monitor Rm.(RB) -1 i
Sampling Rack H22 P250 BySGTS (SC) Plow T22 FIV18A,B By Filter Train.Rm. (RB) i Pilter Moisture - T22-MTV11A,B,C,D ' By Filter Trala Rm. (RB) _j CR HVAC , Emerg. Recirculation Fan ' U41 C603A,B CR HVAC Rm. A,B (CB) Emerg. Charcoal Filter Unit . U41 B - A,B CR HVAC Rm A,B (CB) Air Intake Iso. Valves U41 F. A,B - CR HVAC Rm A,B (CB) Conditioning / Holding Rack P91 (SC) O
- . Sampling / Casks Rack P91 .
FASS Rack Rm. (RB) i LPCF Supply Valve E11.PO45,046;A (SC) DW/WWGas (CAMS) Vale D23 (SC) ll. Controllanel(FT,'IT) H22 FASS Rack Rm (RB) , Chemical Radiological Analysis . Laboratory (SB) SIACE j Radiation (Ion /Scint.) D11 RE041,043;A,B Stack (RB) Monitor Racks, Control Rod H21,H22 ' Stack Monitoring Rm.(RB) i l l (CB) Controlbuilding
-(SC) Secondarycontainment (RB) Reactor building outside (Secondary Containment)
(SB) Servicebuilding i Aswadment 11 1AAS7 )
ABWR zw1mc Standard Plant a- c Table 1AA.5 4 POST ACCIDENT INS'IRUMENTATION AND CONTROLS, POWER AND HABITABILITY SYSTEMS AND AUXILIARIES Equipenest MPL Leestles lastranentation & Controls Post Accident I & C H11 Post Accident Control & Panet Res. (CB) EWEF.it DC Supply R42 Storage Batteries Battery Rm. (CB) ESP HVALV Switchgear R22 Post Accident Eng. Electric Rs. A,B,C(RB) _ ESF Motor ControlQater R24 Post Accident Eng. Electric Rs. A,B,C(RB) , DieselGenerator & Auxiliaries R4M001A,B,C DG Rm. A,B,C (RB) - ' DG Motor Control Center R43 P001A,B,C - DG MCC Rm. A,B,C (RB) - SupplyFan(HVAC) U41 C201,204,207;A,B DG Supply Fan Rm. A,B,C(RB) ' Exhaust Fan (HVAC)- U41202,205,208;A,B DG Exhaust Fan Rm. A,B,C(RB) Esan. Fresh Air Fan (HVAC) U41 W W.W;A,B
. ' DG Essen. Fan Rm A,B,C(RB)
RCW Discharge Valve P21 P055A Thru F - DG Rs. A,B,C (RB) - Control Panel (H21) R43 P002,003,004;A,B DG Control Pal. Rm. A,B,C(RB) CB HVAC Supply Fan U41 C504,606,608;A,B. E/HVAC Rm. A,B,C (CB) Exhaust Fan .U41 C505,607,609;A,B E/HVAC Rm. A,B,C (CB) _ O MCR Supply Fan MCR Exhaust Fan U41 C601,604;A,B: U41 C602,605;A,B
' CR HVAC Rm. A,B (CB)
CR HVAC Rm. A,B (CB)- j' RQY Pump P21 C001A Thru F Pump Rm, A,B,C (CB) Hx Return Valve P21-F004D,E,F Hx Rm. A,B,C (CB) Temperature Control Valve P21 F006,010;A,B,C Hx Rm. A,B,C (CB) Non Post Accident Supply Valve P21 F074A,B,C - (RB) Non Post Accident Return Valve P21 F082A,B,C (RB) Flow P21 FT006A,B,C By Pump Rm.A,B,C (CB) , Pressure P21 FT004A,B,C' By Pump Rm.A,B,C (CB) Surge Tank 1.evel P21 LT013A,B,C By Surge Tank A,B,C(RB) ! HECW l Pump _ P25 C001A,B,C,D,E,F Chiller Rm.A,B,C (CB) - Refrigerator P25 D001A,B,C,D,E,F Chiller Rm.A,B,C (CB) ; Pressure Control Valve P25 F012A,B,C HVAC Rm. A,B,C (CB) Temperature ControlValve P25 F005A,B,C HVAC Rm. A,B,C (CB) Temperature Control Valve P25-F016A,B,C HVAC Rm. A,B,C (CB) , Temperature Control Valve P25 F022A,B,C (RB) RCW Temp. Control Valve P21 F025A,B,C,D,E,F (CB) Differential Pressure P21 PT007A,B,C HVAC Rm. A,B,C (CB) INST AIR Compressor PS2-C001,002 Inst. Air Rm. (RB)
~
(RB). Reactor building outside (Secondary Containment) (CB) Controlbuilding ; Amesdawn 11 1AA.5-8 l
ABWR . :wious gandaggytant mya TABLE 3.21 Table CLASSIFICATION
SUMMARY
(Continued) hble 9
- 3J.1 MPL 34 1 MPt .
handila. Manhar** Illig M Number" 2 Big J NmelaarFast P Beatlam Ausliimrs sentens Seeitem 31 P1 P13 MakeupWater System (Condensate) a :
- m. o .,
P2 P21 Reactor Building Coolig Water System
- K1 K11 Radionaht Drain Transfer System P3 P22 Turbine Building Cooling Water System P4 -P24/P25 HVAC Cooling Water Systems' l P5 P41 Reactor Cooling Water System P6 P42 Turbine Cooling Water System P7 P$1/P52 Instrument /Senice Air N Penerhele Erstems /P54 /High Pressure Nitrogen Systems N1 N11/N21 Power Conversion System N22/N25 N26/N27 R Station Electrical Systems N31/N32 N33/N34 R1 R42 DC Power Supply' N35/N36 N37/N38 R2 R10/R11/ Auxiliary AC Power System' N39/N41 R22 N42/N43 N44/N51 R3 R43 Emergency Diesel Generator N61/N71 System' N72 R4 R$2 Lighting and Senicing Power N2 N62 Offgas System Supply l, These systems or subsystems thereof. have a crimarv function that is saferv-related. As shown in the balance of this Table. some of these systems contain non safety related components and.
converselv. some systems whose primary funettons are non. safety related contain comnonents that have been dosionated saferv-relgggf
** Mactor Parts Iler Number antanar*A for the system O
Amendment 11 3.27 i
l awiowt A..BWR
.a..a pi..r may .
s 3.1 CONFORMANCEWITH NRC to safety shall be designed, fabricated, GENER/1 DESIGN CRITERIA erected, and tested to quality standards f, commensurate with the importance of the safety 3.1.1
SUMMARY
DESCRIPTION functions to be performed. Where generally recognited rodes and standards are used, they This section costalas an evaluation of the shall be identified and evaluated to determine principal design eriteria of the ABWR Standard their applicability, adequacy, and sufficiency Plant as measured agalnat the NRC General Design and shall be supplemented or modified as Criterla for Nuclear Power Plants,10CFR50 necessary to assure a quality product in keeping , Appendix A. Thr, general design criteria, which with the required safety function. A quality are divided lato six groups with the last assurance program shall be established and r criterion numbered 64, are latended to establish implemented in order to provide adequate minimum requirements for the principal design assurance that these structures, systems, and ' criteria for nuclear power plants, components will satisfactorily perform their safety functions. Appropriate records of the , I The NRC General Design Criteria were latended design, fabrication, erection, and testing of . to guide the design of all water. cooled nuclear structures, systems, and components important to : power plans; separate BWR. specific criteria are safety shall be maintained by or under the not addressed. As a result, the criteria are control of the nuclear power unit licensee subject to a variety of interpretations. For throughout the life of the valt, this reason, there are some cases where conformance to a particular criterion is not 3.1.1.1.1J Evcluation Against Cdtedon t ; directly measurable. In these cases, the conformance of the ABWR design to the interpreta. Safety.related and nom. safety.related i tion of the criteria is discussed. For each structures, systerns, and components are criterion, a specific assessment of the plant identified on Table 3.21. The total quality design is made and a complete list of refeates assurance program is described in Chapter 17 and i O is included to identify where detailed %n information pertincut to that criterion is treated in this safety analysis report (SAR). is applied to the safety related items. The quality requirements for non safety related Items are controlled by the quality assurance program described in Chapter 17 in accordance Based on the content herein, the design of the with the functional importance of the item. The ABWR design fully satisfies and is in compliance intent of the quality assurance program is to with the NRC General Design Criteria. assure sound engineering lu all phases of design and construction through conformity to 3.1.2 EVAI.UATION AGAINSTCRITERIA regulatory requirements and design bases described in the license appilcation. In , 3.1.2.1 Group ! . Overall Requirements addition, the program assures adherence to ' specified standards of workmanship and ' 3.1.2.t.1 Criterion 1. Quality Standards and implementation of recognized codes and standards Records in fabrication and construction. .it also ; includes the observance of proper preop'erational ' 3.1.2.1.1.1 Criterion 1 Statement and operational testing and maintenance procedures as well as the documentation of the Structures, systems, and components important foregoing by keeping appropriate records. The total quality assurance program is responsive to and in conformance with the Intent of the .; quality.related requirements of 10CFR50 Appendix ', l i l l O . Amendment !! 3.11 l l, . _ _ _ . - _ . . _ . . . . . . . . . . -, __ _,_ . . _ _ _ _ _ - ._ -_ -
EN a...a.a pi...
. zwiams l l TABLE 3.21 g
CIASSIFICA110N
SUMMARY
(Continued)
- Qualh l Ge**P Qual % !
Satsty Emen. Qassi. Assursace Seisntle i Pdadpal Campamente ggge ggge 808083d Requirmaante (;MARMII Malta l
- 2. Vessel . air accumulators 3 C C B I (for ADS and SRVs) j 1
- 3. Pipinglacluding supports. C ! a 2/3 R/C B (b) )
nafety/ relief valve discharge f i l j i i 1 0' i l l 1 l l t 1 l l l O Aswadment 11 3.2-8.1 _ . _ . . ~ . . , _ . . _ . . _ . . . . - . , _ . - . _ , _ . _ .
. . . . _ . . _ - ~-- _. -.-. - - - - . . . - - - - - - - - ~ - -- - - - - .
2&A6100AE RemmAmsd Plant nry. m l TABLE 3.21 CLASSIFICATION
SUMMARY
he classincation laforination is presented System wise *** in the following order: ! TaWe TaWe 3.21 MPL 3J.1 MPL , l kamlia. Maahr** Ikig kamlia, Nam p 2 Mig ! B Mnetaar Stamm Anamir Swataans E2 E22 High Pressure Core Flooder i System' - B1 B11/J10 Reactor Pressure Vessel . J11/J12 System'/ Fuel' E3 E31 leak Detection and holation l System
- 32 B21 Nuclear Boiler System' E4 E51 RCIC System' 33 B31 Raactor Recirculation System .
F Raneter Sanicing ' C Central and lastrnet Swataans F1 Fil Fuel Servicing Equipment ; C1 C11/C12 CRD System' , F2 F13 RPV SerMeing Equipment C2 C31 Feedwater Control System , C3 C41 Standby Liquid Control Equipment System F4 F15 Refueling Equipment . O C4 C5 C51 C61 Neutron Monitoring System' Remote Shutdown System F5 F16 FuelStorage Equipment G Reactor Auxilian Systems C6 C71 Reactor Protection System'
- G1 G31 Reactor Water Cleanup System D Radiatlan Monitoring Systems G2 G41 FuelPoolCooling and Cleanup ,
D1 D11 Process Radiation Monitoring' System System G3 G51 Suppression Pool Cleanup D2 D23 Containment Atmospheric System i Monitoring System' H Centrol Panels , E Core Coalla, Systems l H1 H11 Main Control Room Panel' i El E11 RHR System' H2 H21 1.ocalControl Panels' These systems or subsystems thereof, have a primary function that is safety related. As shown in the balance of this Table, some of these systems contain non safety related components and, ' conversely, some systems whose primaryfunctions are non safety related contain components that have been designated safety related.
" Master Parts List Number designatedfor the system
*" Only those systems that are in the ABHR Standard Plant scope are included in this table.
. _ , m
i BUINAE Re amA.=d Pla=* mw a l TABLE 3.21 } I i CLASSIFICATION
SUMMARY
(Coattamed) l OmmKW , i Gr**P Quahy Imss. Qassi. Assuranse Saisade ; i flhdBalimasamt* Man' halland Basdantal' M Maias ! ! B1 Beneter Pressere VesselSpeesm/ Feel a. un 1 ; i 4
- 1. Reamer vessel - -
1 C A B ! I t l~ 2. Reamer vesselsupport skirt 1 C A B I and stabiliser i ! I
^
i 3. Reamer vessel appurtenances 1 C A B (3) , Pre 45=re retainin8 Portims
- 4. Supports for CRD bousing, 1 C A B 1 .
l in. core housing and recircu. lation laternalpump ! 5. Reactor internal struaures . 3 C B B ! ! , foe.dwater, RHR/ECCS high
- pressure core Dooder sparsers
- 6. Reactor laternal structures- 3 C - B 1 safety related components ,
, including core support structures l (See Subsection 3.9.5) ; i l 7. Reactor laternal structures . N C - - - l non. safety related components j (See Subsection 3.9.5) ! 8. Controt rods 3 C -- B I
- 9. Power range detector hardware 3 C - B I l- including startup range detector i
- 10. Fuel assemblies 3 C - B I
- 11. ReactorInternalPump 1 C A B ! l Motor Casing 82 NeclearBollerSystem
- 1. Vessels . lewlinstrumenta. 2 C B B I tion maad*asing chambers .
l Amendment 11 124 t i 3 5 e
_.y l 1
- MM si.si. tat ;
Standard Maat aw a i TABLE 3.21 l CLASSIFICA110N
SUMMARY
(Contissed) @ ; OEGIk Gesup QualW > s.. ammi. Assernmee ashale , EtiadanLCamaanas!* ana' Sandea d sh emus' Mesa j l D2 Centaineset Atmospbsic Mealtategsysum
- 1. Composest with safety related ' 3 C,$C - B I El RHRSystem
- 1. Heat enchangers prianary side 2 SC B B I
, 2. Heat enchangersincluding 3 SC C B I supports. secondary side
- 3. Pipingincluding supports
- 1/2 C,$C A/B B 1 (g) l within outermost isolation I valves ,
- 4. Coatalament spray piping 2 C B B I including supports and 8 spargers, within and I including the outer.
most isolation valves e s 4a. Pipingincluding supports 2/3 SC B/C B 1 (g) l beyond outermostisolation valves
- 5. Main Pumpsincludingsupports 2 SC B B I ;
- 6. Main Pump motors 3 SC - B I
- 7. Valves. Isolation,(1.PFL 1 C,SC A B I (g) line) including shutdown suction line isolation valves +
- 8. Valves. lsolation, other 2 C,SC B B I (g)
(pool suction valves and
- pool test return valves) ,
- 9. Valves beyond isolation 2/3 SC B/C B I (g) valves
- De RHR/ECCS lowpressureflooder spegers arepet of the reactorpressure nsselsystem, see hem BLS.
Amendment 11 33 l
. . . _ . _ _ . . _ _ _ _ _ . _ , , _ _ . _ _ . , . _ - . _ _ _ _ . _ _ _ _ 1_ _-
ABM
.a.c...
swiaaim m,. TABLE 3.21 CLASSIFICATION
SUMMARY
(Contissed) Group Quauty Base. Qasel. Asseresse Bdessic BrlastaaLCangnaamata Qg aBB Nest finanand W Caesmar/ Maist 32 Neelser Beuse Sysessa (comessed)
- 4. Pipingincluding supports. 1 C,5C A B I main ana==% (MSL) and feed-water (PW)linewkhis outermost liaissina valm S. Pipingimendingsupports. 2 SC B, B 1 MSL and FW from outermost isolstice valve to andincluding seismicinterface restraint and FWfrom outermostisolation to andincluding shutoff valve
- 6. . Pipingincluding supports . MSL N T B ~ ~
fron the seismicinterface s restraint to the turbine stop valve R$
- 7. Deleted
- 8. Piping. FW beyond seismic N T D -- ~
interface restraint
- 9. Seismic laterface Restraint - 2 SC B B !
MSL/FW 10_ Pipewhiprestralats.MSL/FW 3- SC,C - B ~ l 11. Piping including supports . other within outermostisolation valves
' a. RPV head vent 1 C A B 1 (g)
- b. RPV head spray 1 C A B I (g)
- c. Main steam drains 1 C,SC A B I (g) 4 12. Pipingincudag supports. other I beyond outermost isolation valves
- a. RPV head vent N C D - -.
- b. RPV head spray N SC D ~ -
- c. Main steam drains N SC D - -
Amoeduwat 11 3.2-9
. . . - ~ . - - - . . . - . - . - . . . . . - - . . . _ . _ . . . - . - . . . - . . . . . . . ~ . . .
ABWR numa l Riandlard Plant am a ' l TABLE 3.21 CIASSIFICATION
SUMMARY
(Coattmoed) O. : l i o=utr GmeP Omnuty : IAm* Q8004 A00Erthet EdShic ! Pdadpal Campamenta Citat Slage 30 R g im m ange g gggg ; E2 High Proesure Care Fleeder
- 1. Reactor pressure vessel 1/2 C,5C A/B B I (g) =
i lejoetionline and connected g l pipingincluding supports with- ; la outermost isolation valve' :
- 2. Allother pipingincluding 2/3 SC,0 B/C B I (g) supports " '
, 3. Mala Pump 2 SC B B I ,
- 4. Main Pump motor 3 SC - B I
- 5. Valves . outer isolation 1 C,SC A B I (g) 3 and within the reactor pressure vesselinjectionline and >
connected hoes . l6. All other valves 2/3 SC B/C B I (g) 4
- 7. Electricalmodules with safety- 3 C,$C,X - B I related function
- 8. Cable with safety related 3 C,$C,X - B I function E3 14ak Detection and Isolation System
- 1. Temperature sensors 3/N C,SC -
B/- I/- (r) i
- 2. Temperature switches 3/N X -
B/- I/- (z)
- 3. Pressure transmitters 3/N C,SC --
B/- 1/- (r) .
- 4. Pressure switches 3/N X -
B/- I/- (r) - l
- 5. Differentialpressure 3/N C,SC -
B/- I/- ' (r) l transmitters (flow) The ECCS high pressure core flooder spargers are part of the Reactor Pressure Vessel System, see .! Item B1.5. i
" Poolsuction piping, suction pipingfrom condensate storege tank, test line topool, pump dischap pipingandretum line topool.
Amendment 11 3.2-15
M s w to w ! mer 'N P^-' mw a TAB 123 21 CIASSIFICATION
SUMMARY
(Continned) Groep Quality Leen. Onesl- Asseresse h heatipal caranamente QBas don' Bandaad WM Halet Cs esmeeesheedemaspetan e ,r- of tbla system are included under b2, Ei, E4,03, H2, and P2.
- 1. Electricalmodules with 3 C,5C,RZ, - B 1 anfety related functica X
- 2. Cable with safety related 3 RZ ~ B 1 fumaica i, C4 Reacter Protection systems
- 1. Electrical modules with 3 SC,X,T, - B 1 anfety related function RZ
- 2. Cable with safety 3 SC,X,T, - B 1 O 3.
related functions ElectricalModules,other N T,X RZ (u) l
- 4. Cable,other N T,X - - !
(u) D1 Process Radiation Moaltorlag System (lacludes gaseous and liquid amuent mealtorlag)
- 1. Eledricalmodules with 3 SC,X,RZ - B 1 with safety related functions (lacludes monitors)
- 2. Cable with safty related 3 SC,X,RZ - B 1 functions
- 3. ElectricalModules, other N T,SC,RZ, - - -
(u) X,W
- 4. Cable, other N T,SC,RZ, - - -
(u) X,W O ' Amendment 3 3.2 13
-asamAnd MM Plant switoAE arv. m TABLES.21 CLASSIFICATION
SUMMARY
(Continued) g
- Quality Gr**P Asserener Seisene Pdadnal Casanament" Onsel. d Ramdemment' CateparrI Beatlan Matas 01 Basseer Water Cleaney Spetem (Centissed)
- 4. Pipingincluding supports and 1 C,$C A B 1 (3) was withis andincludies outermost oostainment isolation valves on pump suction
- 5. Pump suaion and discharge N SC C ~ - (g) pipinginduding supports and valves from containment isole.
tion valves back to shut off valves at feedwater line enmaneta .
- 6. Pipinginduding supports and 2 SC B B I (g) valves from feedwater lines to and including shut off valves
- 7. Pipinginduding supports and N SC,T C ~ - (g) valas to main condenser
- N
- g. Non regenerative beat exchanger N SC C - - (g) tube inside and pipingincluding supports and valves carrying process water
- 9. Non regenerative beat exchanger N SC D - -
shell and pipingincluding supports carrying closed cooling water
- 10. Filter /demineralizer N SC D - -
precoat subsystem
- 11. Filter demin holding pumps N SC C - -
includingsupports vahes and pipingincluding supports Anendswat 3 3.2 19 G I
i zwian i l A.BWR
....a pi... .m .
I t TABIE 3.21 1 i l CIAS$!FICATION
SUMMARY
(Contioned) ; I G'88P Quably . Safstr Emen. Oasel. Assuranse Ostende ! Primelant canapammmen Data' don', Sandead WM Meta l El RNR System (Contissed)
- 10. Mechanicalmodules with 3 SC C B I I safety related functions -
j
- 11. Electricalmodule4 with 3 C,SC,X - B I ;
l safety related function i l *
- 12. Cable with safety related 3 C,$C,X -- B !
l function
- 13. Other mechanical and N C SC,X - - ~
clectrical modules 4 14. Jockeypumps including 2 SC B B ! k Supports
- 15. Jockeypumpmotor N SC ~ - --
- i k
i 1 5
?
Amendewat 11 3.214e f
. . . - - . . . - _ - . .,-_..,,..e. - . - - . . . . . - - - - - - . - . - ._ - - . - . - - - - - - - , _ - . , , . . - - , - . - - , - . . . . _ . . . . ~ . < . - -
i MM . autom standard Maat an a i TABE 3.21 I l CLASSIFICATION
SUMMARY
(Contissed) o.amy h J Group Quality ; Saq Laos. Qassi. Assessace aanamie , ! Pdadpal canapammate skg* gaat m e Enemimment' Cdanar/ Mgu i i l P1 Maheep Water system (comesanate) l
- 1. Pipingincluding supports and 2 C B B ! -
valvas foradas part of the non-talement boundary i
- 2. enmAename* storage tank N O D - -
(w) I 4 including supports L
- 3. raad==an's header piplag 2 SC B B I including supports and valves I
- 4. Pipingincluding supports and N O D - -
valves
- 5. Other components N O D - -
r F2 Beactor Bulldlag Coollag Water System O
- 1. Piping and valves forming part 2 SC,C B B I (g) 'l' of primary containment boundary .
f 2. Other safety related piping, including supports pumps and 3 SC,C C B I ' valves
- 3. Electricalmodules with 3 SC,C,X .- B. I with safety related function
- 4. Cable with safety related 3 SC,C,X - B 1.
! function i
- 5. Other mechanical and N SC,C,X,M - - -
i electricalmodules I P3 'hrbine Building Cooling N T D - - Water System l 9 i G i j Anwedmont 11 3.2 22 j >
. . _ _ . _ _ . . ~. ._ . , . . , . , . . . _ . _ . . . _ _ . . . _ _ . . . . .. __1..._.--
MM menaAnd Plant 3&A6100AE an a TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) 2 Quality ,
. Group Quality i Set Emes. Classi. Assuranse Selsede Priadpal Cemannasta gggg ggge gd Ramelemment' M Helgg .j F2 RPVServiciasEgelpment ,
e i 1. Steamline plugs N- SC ~ - - I
- 2. Dryer and separator N SC ~ - - j strongback and head strongback i
F3 RPV nt malServicingEquipment i i
- 1. Controlrod grapple N SC ~ - -
F4 Refueling Equipment
- 1. Refueling equipment N SC I
( platform assembly
.(bb)
{ i l
- 2. Refueling bellows N SC ~ - -
l F5 Fuel Storage Equipment
- 1. Fuel storage racks. N SC ~ - 1 (bb) .;
new and spent C ) N l j 2. Defective fuel storage N SC - - -- " (bb) I container ' l i } G1 ReactorWaterCleanupSystem I i 1. Vessels including supports N SC C - - (filter /demineralizer) i l
- 2. Regenerative heat exchangers N SC C -- - ~
including supports carrying reactor water ). 3. Cleanup recirculation N SC C -- - pump, motors ] Amendment 11 3.2-1B
23Ad100AE M..M..e m.. m. TABIE 3.21 CI.ASSIFICATION
SUMMARY
(Contissed) 4 Qasilty Gr**P Quality Pdadant caranamente g glggg ime. gggge clanel. DEBilGR0 SRAElEEBBBl' Assurames BM M E0184
- 2. Bletrical modules and cables -
N P,0,T - - - with non safety related funesion l F7 lastrument/Serdes Air Systems
- 1. Containment isolationinclud- 2 C B B I lag supports valves and piping
- 2. Pipingincluding supports with 3 SC,C C B 1 safety related function
- 3. Electric modules with . 3 RZ,X --- B I safety related functions
- 4. Cable with safety related 3 SC,RZ, . B I function X
- 5. Other non safety related N SC,RZ, -- -- -
mechanical and electrical X components R1 DCPowerSupply.NuclearIsland R
- 1. 125 volt batteries, battery .3 SC,C,X, 6 - B I racks,batterychargers, and RZ distribution equipment i
Asundment 11 3.2 23.1
i l ABM swioaus i Standard Plant an a i
- i . TABIE3.21
- t CIASSIFICA110N
SUMMARY
(Continued) i Quality i cmuy Queuty I Imm. Oassi. Assurance Seismic i W campananta Ggl Bat lies' ikallaad W C at W E I Hatti
- 4. Turbine bypass piping N T D - -
{ including supports ;
- 5. Turbinestopvahe, turbine N T D - -
(1)(n)(o) ; bypass vehen, and the main steam leads from ib turbine . controlvalve to the turbine !
- 6. Feedwater system components N T D N/A N/A- !
beyond outboard shutoff valve [
- 7. Turbine generator N T - - - i
- 8. Condenser N T - .. ~
g 9. Air ejector equipment N T - - -
- 10. Turbine gland sealing N T D - -
system components N1 Power Conversion System (Later) N2 Oftps System l
- 1. Pressure vesselsincluding N T - - - (p)(q) -
j supports
- 2. Atmospheric tanks including N T -- -- - (p)(q) supports
- 3. 015 psig tanksincluding N T - - - (p)(q) supports
- 4. Heat exchangersincluding N T - -- .. (p)(q) supports 1
- 5. Pipingincluding supports N T - - - (p)(q) j and valves i l
- 6. Pumps including supports N T - -- -
(P)(9) ) A*t 9 3.2-21t> 1 i '
21A610 M B
..a..a pi..e m TABLE 3.21 h1 , CLASSIFICATION
SUMMARY
(Continued) Quauty ; Group Quauty Pdadpal Campamange g gggg Imen. ggge ' Clasel. ggggggd Asserene Bassissant' M Sdesde Estas , 9. M otors 3 sC,C,X,RZ - B I j
- 10. Imad sequencers 3 SC,X,RZ - B I j l I 11. Protective relays and aantrol 3 SC,X,RZ - B I l Panels 1
12, Valve operators 3 SC,C,X,RZ - B I I Bs smassocy06 Ia w tersystem
- 1. Starting air recolver tanks 3 RZ C B I (y)
PIlP agincluding supports from andincluding check valve and downstream pipingincluding supports and valves l 2. Starting air compressor and N RZ - - - m.o,s
- 3. Combustion alt intake and 3 RZ,0 C B I exhaust system t
- 4. Safety.related pipingin- 3 RZ,0 C B I ciuding supports valves. fuel oil system, diesel cooling water system, and lobe oil system RZ,0 l 5. Pumpmotors fueloil 3 -- B I system, diesel cooling water system andlobe oil system l
l 6. Dieselgenerators 3 RZ - B I (y) l 7. Mechanical and electrical 3 RZ,0,X - B I modules with safety-related fumaions 8 Cable with safety related 3 RZ,0,X - B I . functions
- 9. Other mechanical and N RZ,0 - - -
electrical modules Asentment 10 3.2-34.1 e -*------__--___------_mm._ - ,-v-ee- -wve- "r-- - , . , - - r* ,-w<---w- w w - ,-w w---g.e.-e -.--weyw,-mes,= -.--- g ywa -- g - -mi- e w-v-
N M.a.e... alAHelAE m. TABLE 3.21 CLASSIFICATION
SUMMARY
(Coattamed) Quahty Gramp Quahey , se gaan. fmal. Assuranse ama mk EldasiaalEsmesmanga Q808 $8B' Omd8Bd WM Maha N lastrument/Servise Air /High pmasere Nkrossa syseen.
- 1. ca .i .
kai w i i.d. 2 C B B 1 ins supports valm and piping
- 2. Assumalators,and downstream 3 SC,C C B I pipinglaciuding supports with safety related in.cek '
- 3. Blectric modules with 3 C,RZ,X - B I safety related functions
- 4. Cable with safety related 3 C SC,RZ, - B I ,
function X '
- 5. Other non safety related N C.SC,RZ, - - -
mechanical and electrical X components P8 Makeup Water System (Condensate)
- 1. Pipingincluding supports and 2 C B B I valves forming part of the con-tainment boundary
- 2. Condensate storage tank N O D --- -
(w) including suppets
- 3. Condensate header piping 2 SC B B 1 including supports and valves l
- 4. 'Pipingincluding supports and N O D - -
valves
- 5. Other components N O D - -
R1 DC Power Supply. Nuclear Island
- 1. All components with safety- 3 SC,C,X, -- B I related function RZ Amendment 11 3.323
-.j
3hunclAE RemenAmed Pl==* m. m I TABIE 3.21 , CLASSIFICATION
SUMMARY
(Coetlaned) l 1 QuallV Groep Quelly saast ima. Chand. Asseramee Sdsele I Primelmal cammaman,a gggge des' Scadead RMairamante Calmu2 I h l l 12 Centalassent laternal Steveteres (Contiamed) j a 2. Support structures for safety. 2 C ~ B I I g related pipingincluding 1 supports and equipment i 13 RPY Fedestaland Shield Wall 1
- 1. RPV pedestaland shield wall 3 C - B I l l \
- 2. Diaphragm Door 3 C - B 1 l l
T4 Standby Gas TWestment System
- 1. All equipment except 3 SC,C,RZ - B I deluge piping and valves
- 2. Deluge piping and valves N SC - ~ .
8 l L I ! i l I 1 1 l O-Amendment 11 3.2 26 l i
.t
1 ABM swi== , Standard Plant aw a j l J TAB 123.21 l J CLASSIFICATION
SUMMARY
(Continued) l 1 Quath 4 oreep Qualh !
! ass- Onesi- Assernam N !
M adP al h anente P g%ang game 33gg34 W CatamanI Esta l
- 2. Centrol power cables ('aciud- 3 SC,CX - B I !
under ground emble system, RZ , aplicos,aonnemors and ! terminalblocks) !
- 3. Conduk and emble ersys and 3 SC,C,X, - B ! l their supports RZ - '
- 4. Protestive relays and control 3 SCXRZ - 3 I ;
i Penals j l t
- 5. Contalament elearical pene. 3 SC,C - B ! !
, trations assemblie.s
- 6. Motors 3 SC,C,X, - B I )
RZ R2.hssulary AC Power System
- 1. 6900 volt switch gear 3 SCXRZ - B !
- 2. 480 mh load centers 3 SC,X,RZ - B I
- 3. 480 volt motor control 3 SC,X,RZ ~ B !
eenters 7
- 4. 120 VAC safety related dis- 3 SC,X,RZ - B I l tribution equipment includ-inginverters ;
- 5. Control and power cables 3 SC,C,X - B 1 (irduding underground RZ cable systems, cable splices, conmemors and terminalblocks)
- 6. Conduit and cable trays and 3 SC,C,X - B I i their aupports RZ l
- 7. Containment electrical 3 SC,C,X - B I penetration assemblies RZ
- 8. Transformers 3 SC,C,X - B 1 RZ Amenement 10 3.2 24
.a. e. .. ..
MQIM (Contioned)-
- 4. Allother instrument lines:
i 1,,ougs e _ .s - abaoeof e name .s - as . .,ates to which they are attached. Il Beyond the eoot valve,if used to navate a safety system, the lines aball be of the same clanalfication as the system to which they are attached, til Beyond the root valve,if not used to actuate a safety system, the lines may be Code Group D.
- 5. All sample lines from the outer isolation valve or abe process root valve through the remainder of the sampling system may be Code Group D.
g 6. All safety relsed instrument sensing lines shall be in conformance with the criteria of l 5 Regulatory Guide 1.1$1.
- h. Relief valve discharge piping shall be Quality Group B and Seisak Category 1.
Safety / relief valve discharge line (SRVDL) piping from the safety / relief valve to the quen. chers in the suppression pool consists of two parts: the first part is attached at one end to the safety / relief valve and attached at its other end to the diaphragm floor penetration.' This first portion of the safety / relief valve discharge piping is analysed with the main steam piping as a complete system. The second part of the safety / relief valve discharge piping extends from the penetration to the quenchers in the suppression pool. Because of the penetration on this part of the line,it is physically decoupled from the main steam piping and the first part of the SRVDL piping and is, therefore, analyzed as a separate piping system,
- l. Electrical devices include components such as switches, controllers, solenoids, fuses, junction boxes, and transducers which are discrete components of a larger subassembly /
module. Nuclear safety.related devices are Seismic Category 1. Fall. safe devices are non Scismic Category 1. J. The control rod drive insert lines from the drive flange up to and including the first valve on the hydraulic control unit are Safety Class 2, and non safety related beyond the first valve.
- k. The hydraulic control unit (HCU) is a factory. assembled engineered module of valves, tubing, piping, and stored water which controls two control rod drives by the application of pressures and flows to accomplish rapid lasertion for reactor scram.
Although the hydraulic control unit, as a unit,is field installed and connected to process piping, many of its internal parts differ markedly from process piping components because of the more complex functions they must provide. Thus, although the codes and standards invoked by Groups A, B, C, and D pressure integrity quality levels clearly apply at all levels to the interfaces between the HCU and the connection to conventional piping components (e.g., pipe nipples, fittings, simple hand valves, etc.), it is considered that they do not apply to the specialty parts (e.g., solenoid valves, pneumatic components, and instruments). Amenamen: 6 3.2 32
I ABWR h '--d ]>lant imam _ REV. A TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) QuaHty Group Quality Sag Classi.d Assurance Selsele I homenpal canapannes' gnag %as .gganan g Raami m ment' Catanerv Hggs B4 ughtleg and Serviclag Power Supply
- 1. Emergency Lighting N SC,C,X - ~ ~
ItZ T1 Pvtmary Containment System
- 1. Primary containment veuci 2 C B B I (PCV). reinforced concrete ma'al== cat veuel (RCCV)
- 2. Vent system (vertical flow 2 C B B 1 channels and horizontal discharges
- 3. Suppression chambee/drywell 2 C O
B B I vacuum breakers
- 4. PCV penetrations 2 C B B 1 and drywell steel head
- 5. Upper and lower drywell altlocks 2 C,SC - B i
- 6. Upper and lower drywell 2 C.SC - B 1 equipment hatches
- 7. Lower drywell access tunnels 2 C -- B I
- 8. Suppression chamber 2 C,SC - B 1 access hatch
- 9. Safety related instrumentation 2 C,SC - B 1 T2 Costalament Internal Structures
- 1. Reactor vessel stabilizer 3 C - B 1 truss O Amanhnent 2 3.2 25 I
1
314s101AB M..M..m..t -a : 21038 (Ceadased) ; , n. All cast pressure retalalag parts of a sire and configuration for which volumetric methods er sffective are emaalsed by radiographic methods by qualified personnel. Ultrasoalc saamination to ! equivalent standards is v6ed as an sitsta&te to r6diographic methods. Eaasiastion procedures and ! acceptance standards are at least equivalent to those defined la Paragraph 136.4, Nomboiler External i Pipag, ANSI 331.1. ;
- o. De foBowing qualifkations as met with respect to the eerehmeta= requirements: j
- 1. The manufacturer of the turbine stop valves, turbine control valves, turblee bypass valves, and mala steam leads from turbine control valve to turblae eastag utilises quality control procedures equivalent to those defined in GE Publication GEZ 4962A, Gescral Electric Large Steam Turbine r Generator Quality Control Program.
' r
- 2. A certificatloa obtalmed from the manufacturer of those valves and steam loads demonstrates that i the quality control program as defined has beca accomplished. ;
p, Regulatory Guide 1.143 furnishes complete design guidance relating to seismic and quality group [ classification and quality assurance provisions for radioactive waste management systems, structures + and components. l
- q. Detailed seismic design criteria for the offgas system are provided in Section 11.3. ,
i
- r. The portions of the MSL from the second isolation valve to the turbine stop valve and the first valve in branch lines shall be designed so that the SSE does not cause structurallateraction or failure >
that could degrade the functioning of a Seismic Category I structure system or component to an > unacceptable safety level,
- s. Not used !
- t. There is a limited quality assurance program for the Fire Protection System,
- u. Special seismic qualification and quality assurance requirements are applied,
- v. Not used, i i w. The condensate storage tank wl!! be designed, fabricated, and tested to meet the intent of API Standard API 650. In addition, the specification for this tank will require: (1) 100% surface ;
i examination of the side wall to bottom joint and (2) 100% volumetric examination of the side wall weld joints.
- x. The crancs are designed to hold up their loads under conditions of OBE and to m S tain their ,
positions over the units under conditions of SSE. l
- y. All off engine components are constructed to the extent possible to the ASME Code, Section III, Class
- 3. :
-1
- z. Components associated with a safety.related function (e.g., isolation) are safety related.
an. Structures which support or house safety related mechanical or electrical components are safety related, bb, A quality assurance requirements shall be applied to ensure that the design, construction and testin requirements are met. Anwedsent 11 3.2.M
. _ . _ . _ _ _ _ _ _ _ . __...__ _ ._ _. . -_ o
M M=u.ei==. austaus ma MmES
- a. A module is an assembly of laterconoscted composeats which constitute an identifiable device or plete of equipment. For example, electrictl modules naciude sensors, power supplies, and signal processors and mechanical modules include turbines, atrainers, and orifices.
- b. 1,2,3, N = Puelaar safety related fmnian @a deAmed in Subseslams $13 and $1$.
- c. C = Prim ary t'a=tal==='
H = Servi:e Building M= any W rlaca*1a= 0 = Cutdoors omake RE = Reactor Building Qeaa 2ces (balance portion of the tsamor building outside the Secondary Ccatainment Zone) sC = secondary Costalement portion of the reemor building T = Turbine Building W= Radwante Building X = ControlBuilding F = Firewster Pump House U = Uhlmate Heat Sink Pump House P = Power Cycle Heat Sink Pump House
- d. A,B C.D = Quality groups defined in Regulatory Guide 1.26 and Subsection 3.2.2. The structures, systems and components are designed and constructed in accordance with the requirements identified in Tables 3.2 2 and 3.2 3.
~
= Ouality Group Classification not applicable to this equipment,
- s. B = the quality assurance requirements of 10CFR50, Appendix B are applied in accordance with the quality assurance program described in Chapter 17.
~
= Requirements of 10CFR50, Appendix B are not applicable.
- f. 1 =
The design requirements of Seismic Category I structures and equipment are applied as described in Section 3.7, Seismic Design.
- = The seismic design requirements for the safe shutdown earthquake (SSE) are not applicable to the equipment. However, the equipment that is not safety related but which could damagt Slesmic Category I equipment if its structural lategrity failed is checked analytically and designed to assure its integrity under seismic loading resulting from the SSE.
- g. 1. 1.ines one inch and smaller which are part of the reactor coolant pressure boundary shall be ASME Code Section !!!, Class 2 and Felsmic Category 1,
- 2. Allinstrument llacs which are connected to the reactor coolant pressure boundary and are utilized to actuate and monitor safety systems shall be Safety Class 2 from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation,
- 3. Alllastrument lines which are connected to the reactor coolant pressure boundary and are not utilized to actuate and monitor safety systems shall be Code Group D from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation.
Amendment 11 3.2 31
ABM...a pi... siansaan mu values of the vertical $$E spectra applied at the The magantede of the $$E design time history ground surface la the free field for damplas is equal to twice the magaltede of the destga rallos of 2.0, $.0, 7.0, and 10.0% of critical OBE time history. The OBE time histories and damping where the maalmum vertical ground response spectra are ased for dynamic analysis acceleration is 0.30 g at 33He, same as the and evaluation of the structural Seismic Systeal maalmum borirostal ground acceleratloa. the OBE results are doubled for evaluatlag the structural adequacy for $$E. For developesent of the design iralues of the OBE response spectra Not response spectra for Seismic Subsystem are one.balf* of the spectra shown la Figures analysis and evaluatloa, aos Subseetloa 3.7.2.5. 3.71 and 3.7 2. These spectra are abown la Figures 3.7 3 through 3.7 20. The response spectra produced from the OBE dealga time histories are shown la Figures 3.7 3 The design spectra are constructed la through 3.7 20 along with the design OBE accordance with Regulatory Guide 1.60. The response spectra. The closeness of the two normalisation factors for the meninsum values in spectra la all cases ladicates that the two borleostal directions are 1.0 and 1.0 as synthetic time histories are acceptable, applied to Figure 3.71 For vertical direction, the normalleatlos factor is 1.0 as applied to The response spectra from the systbetic thee. Figure 3.7 2. histories for the damping values of 1,2,3 and 4 percent conform to the requirement for an 3.7.1.2 Desiga 71me History enveloping procedure provided in Item !!,1.b of Section 3.7.1 of NUREO.0000 (Standard Review The design time histories are synthetic Plan, SRP). However, the response spectra for acceleration time histories generated to match the higher damping values of 7 and 10 percent the design response spectra defined in Subsection show that there are some deviations from the SRP 3.7.1.1. requirement. This deviation is considered loconsequential, because (1) generatlag an The design time histories considered in GESSAR artificial time history whose response spectrs (Reference 1) are used. They are developed based would envelop design spectra for five different on the method proposed by Vanmarcke and Cornell damping values would result in very r.onservative (Reference 2) because of its intrinsic capability time histories for use as design basis input, of imposing statisticalIndependence among the and (2) the response spectra from the synthetic syntbestred acceleration time history time histories do envelop the design spectra for components. The earthquake acceleration time the lower damping values. This is very history components are Identified as H1, H2, and important because the loads due to SSE on V, The H1 and H2 are the two horizontal structures should use 7 percent damping for components mutually perpendicular to each other, concrete components, but are obtained by Both H1 and H2 are bued on the design horizontal ratiolog up the response from the OBE ana'r8s ground spectra chown in Figure 3.71. The V is involving the lower damping. The OBE analf ,s the vertical component and it is based on the uses only the lower damping values (up to 4%), design vertical ground spectra shown la Figure which are consistent with the SRP requirements i 3.72. (See Subsection 3.7.1.5).
'I
- The OBE given in Chapter 2 is one third of the SSE, i.e., 0.10 g, for the ABWR Standard Nuclear Island design. However, as discussed in Chapter 2, a more conservative value of one half of the SSE, i.e., 0.15 g, was employed to evaluate the structural and component response.
Amendment 1 352 O
g . . _ _ _ _ . __ l i ABWR imi. ora ! StandardPlant ~ an a gggg (r n..ma) J l The design and construction specifications for the HCU do invoke such codes and standards as can be j
\ reasonably applied to ladividual parts in developing required quality levels, but of the remaining i parts and details. For example: (1) all welds are LP inspected; (2) all socket trelds are inspected for gap between pipe and socket bottom; (3) all welding is performed by qualified welders; and (4) t all work is done per written procedures. Quality Group D is generally applicable because the codes and slandards invoked by that group contata clauses which permit the use of n'anufacturer standeds
- and proven design techniques which are not explicitly defined within the codes for Quality Grou,os A, l B, or C. This is supplemented by the Oc techalque described. i L De turbine stop valve is designed to withstand the SSE and =mintala its lategrity. [
- m. The RCIC turblae is not lacluded la the scope of standard codes. The assure that the turbine is fabricated to the standards commeasurate with safety and performance requirements, General Electric has established specific design requirements for this component which are as follows:
- 1. All welding shall be qualified in accordance with Section IX, ASME Boiler and Pressure Vessel Code.
- 2. All pressure containing castings and fabrications shall be hydrotested at 1.5 times the design pressure
- 3. All high pressure castings shall be radiographed according to: ;
ASTM E 94 i E 141 E 142 maximum feasible volume O E 71,186 or 280 Severitylevel3 i 4. As-cast surfaces shall be magnetic particle or liquid penetrant tested according to ASME Code, , Section 111, Paragraphs NB.2575, NC 2576, or NB 2576, and NC 2576. 1 -
- 5. Wheel and shaft forgings shall be ultrasonically tested according to ASTM A 388.
- 6. Butt welds shall be radiographed and magnetic particle or liquid penetrant tested according1o the ASME Boiler and Pressure Vessel Code. Acceptance standards shall be in accordance with ASME Boiler and Pressure Vessel Code Section Ill, Paragraph NB 5340, NC 5340, NB 5350, or NC 5350, respectively.
- 7. Notification shall be made on major repairs and records maintained thereof, f
- 8. Record system and traceability shall be according to ASME Section !!I, NCA 4000,
- 9. Control and identification shall be according to ASME Section III, NCA 4000. I
- 10. Procedures shall conform to ASME Section III, NB 5100 and NC 5100.
- 11. Inspection personnel shall be qualified according to ASME Section III, NB 5500 and NC 5500.
)
A art au 10 3.2 33 l l l 1
.. . . . - - . , -- -- -~ . . . . - . . - - . - - . .
1 3M61onAB ! - M..N...a pi... m j
. Where small, non Selsmic category piping is (6) recording and playback equipment; and directly attached to Seismic Category I piping, l its effect on the Seismic Category I piping is (7) annummators. j accounted for by lumping a portlos of its mass f with the seismic Category 1 piping at t'se point The location of seismic instrumentation is l of attachment. cuttined la Table 3.7 7. :
l Furthermore, son. Seismic Category I piplag 3.7A.1.1 1kne History Amesteropophs (particularly high energy piplag as defined la : gection 3.6) is designed to withstand the SSE to Time hlstory accelerographs produce a record avoid jeopardirlag adjacent Seismic Category I of the time.varylag acceleration at the sensor piplag if it is not feasible or practical to locatloa. This data is used directly for analy. isof ate tbese two pipiog syateas. als and comparison with refereace taformation and may be, by calculational methods, converted to 3.7114 Sensate Analysis for Roseter response spectra form for spectra comparisons v laternals with design parameters. l' l i The modeling of RPV laternals is discusse.d la Each triarlal acceleration sensor unit con. ; Subsection 3.7.2.3.2. The damping values are tains three accelerometers wuated la an ortho. given la Table 3.71. The seismic model of the gonal array (two horizontal and one vertical). RPV and laternal is shown in Figure 3.7 32. All acceleration seits have their prlaclpal axes oriented identically. The mounted units are ' 3.73.18 Analyals Procedures for Damplag oriented so that their axes are aligned with the building major axes used in development'of the Analysis procedures for damping are discussed mathematical models for scismic analysis, in Subsection 3.7.2.15. One THA is located on the reactor building 3.7.4 SelsnticInstrumentation (RB) foundation mat, El ( ) 13.2 M, at the base , of an RB clean zone for the purpose of measuring ; 3.7A.1 Comparison with NRC Regulaton Gulde the input vibratory motion of the foundation 1.12 mat. A second THA is located in an RB clean zone at El (+) 26.7 M on the same azimuth as the The seismic lastrumentation program is foundation mat THA. They provide date on the consistent with Regulatory Guide 1.12. frequency, amplitude, and phase relationship of ^ i the seismic response of the reactor building l 3.7A.214 cation and Description of structure. A third THA is located in the free lastrummatation field at the finished grade approximately 160 M from any station structures with axes oilented in The following instrumentation and associated the same direction as the reactor building ' equipment are used to measure plant response to accelerometers. earthquake motion: Two seismic triggers, connected to form redun. (1) three triaxial time.hlstory accelerographs dant triggering, are provided to start the THA (THA); recording system. They are located in the free field at the finished grade 160 M from the reac. (2) three peak recording accelerographs (PRA); tor building. The trigger unit consists of or. thogonally mounted acceleration sensors that act-(3) two triaxial seismic triggers; unte relays whenever a threshold acceleration is exceeded for any of the three axes. The trigger - (4) one seismic switch (SS); is engineered to discriminate against false starts from other operating inputs such as traf. ; (5) four response spectrum recorders; fic, elevators, people, and rotating equipment. Amendment 1 3.7-34 O l
23A6100AE Standard Plant arv.a 3.7 SEISMICDESIGN that earthquake which produce vibratory ground motion for which those features of the nuclear
\O All structures, systems, and equipment of the power plant necessary for costleued operation facility are defined as either Seismic Category I without undue risk to the health and safety of or non Seismic Category 1. The requirements for the public are designed to remain functional.
Seismic Category I identification are given in During the OBE loading condition, the safety. Section 3.2 along with a list of systems, compo- related systems are designed to be capable of seats, and equipment which are so identified, costlaved safe operation. Therefore, for this loading condition, safety related structures, All structures, systems, components, and equip- and equipment are required to operate within mest that are safety related, as defined in sec. design limits, tion 3.2, are designed to withstand earthquakes as defined herein and other dynamic loads includ. The seismic design for the $$E is latended to ing those due to reactor building vibration (RBV) provide a margla la design that assures caused by suppression pool dynamics. Although capabill ty to shut down and malatain the this section addresses seismic aspects of design nuclear facility la a safe condition. In this case,it is only necessary to ensure that the l and 1.70,analysis in accordance the methods of this section with Regulatory are also Guide and components do not lose required systems applicable to other dynamic loading aspects, their capability to per form their
- except for the range of frequencies considered, safety related function. This is referred to es The cutoff frequency for dynamic analysis is 33 the no loss of function criterion and the Hz for seismic loads and 80 ZHz for suppression loading condition as the SSE loading condition.
pool dynamic loads. The definition of rigid system used in this section is applicable to Not all safety related components have the l seismic design only, same functional requirements. For example, the reactor containment must retain capability to The safe shutdown earthquake (SSE) is that restrict leakage to an acceptable level, O earthquake which is based upon an evaluation of Therefore, based on present practice, clastic the maximum carthquake potential considering the behavier of this structure under the SSE loading regional and local geology, selsuology, and condition is ensured. On the other hand, there i specific characteristics of local subsurface are certain structures, components, and systems material, it is that earthquake which produces that can suffer permanent deformation without the maximum vibratory ground motion for which loss of function. Piping and vessels are Seismic Category I systems and components are examples of the latter where the principal designed to remain functional. These systems and requirement is that they retain contents and ; components are those necessary to ensure: allow fluid flow. (1) the integrity of the reactor coolant pressure Table 3.21 identifies the equipment in boundary; various systems as Seismic Category I or non. Seismic Category I. (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; and 3.7.1 Seismic Input , (3) the capability to prevent or mitigate the 3.7.1.1 Design Response Spectra consequences of accidents that could result in potential offsite exposures comparable to The design earthquake loading is specified in the guideline exposures of 10CFR100, terms of a set of idealized, smooth curves ; called the design response spectra in accordance The operating basis earthquake (OBE) is that with Regulatory Guide 1.60. " ! carthquake which, considering the regional and local geology, seismology, and specific charac. Figure 3.71 shows the standard ABWR design teristics of local subsurface material, could values of the horizontal SSE spectra applied at reasonably be expected to affect the plant site the ground surface in the free field for damping during the operating life of the plant. Itis ratios of 2.0, 5.0, 7.0 and 10.0% of critical Amendment 11 3.7-1 4 k
, - , - - - - - . . - ~ - ~ - , . . - - .--.-,n.~ ,- -. - - - . , , , , , . , , - , . , - - - w s
MM meanda.d pinne ' awle64E arv a a SEC110N3.9 CONTENTS (Continued) g
- m e.
3.9.1/.* ASME Qase 2 and 3 Vanaals SAS 3.9.1.4.10 ASME Qass 2 and 3 Pumps SA3 3.9.1.4.11 ASME Qass 2 and 3 Vahes 3.93 3.9.1.4.12 ASME Qass 1,2 ar.d 3 Piping SAS 3.9.1.5 lastantic Analysis Methods 3A3 3.9.2 Dunamie Tessing and Anahmis 3.93.1 3.9.2.1 Piplag Vibration Thermal 87 % and - Dynamic Effects 3.93.1 33.2.1.1 Prooperational and initial Startup Vibration 3.9-4 and Dynamic Effects 3.9.2.1.1.1 Measurement Tochalques 3.94 3.9.2.1.1.2 Monitoring Requirement 3.94 3.9.2.1.13 Test Evaluation and Acceptance Criteria for Main Steam Piping 3.95 l 3.9.2.1.1.4 Reconciliation and Correalve Actions 3.9.$ 3.9.2.1.2 ThermalRi-- t aeTesting 3.96 3.9.2.1.2.1 Meuurement Techalques 3.94 3.9.2.1.2.2 Monitoring Requirements 3.9-6 3.9.2.1.23 Test Evaluation and Acceptance Criteria 3.94 3.9.2.1.2.4 Reconciliation and Corrective Actions 3.97 3.9.2.2 Seismic Qualification of Safety Related l Mechanica1 Equipment (locluding Other RBV i Inducedleads 3,99 3.9-111 Asneaanwnt It O
ABM nasiaans - itenndard P' ant am n 3.73.g.2.2 Ettoct of D!ttersatial Baudlag adequately accounted for in the analysis. In Monuments case of buried systems sufficiently flex. The relative displacement between anchors is
. Ible relative to the surrounding or under.
lying soll, it is assumed that the systems determined frori the dynamic analysis of the will follow essentially the displacements and structures. The results of the relative anchor- deformations that the soll would have if the point displacement are used in a static analysis systems were absent. When applicable, to determine the additional stresses due to procedures, which take into account the relative anchor point displacements. Further phenomena of wave travel and wave reflection details are given in Subsection 3.7.3.8.1.8. In c ampacting soll displacements from the - ground displacements, are employed.' 3.7.3.9 Multiple Supported Egalpment Compements s Witb Distinct inputs (2) The effects of static resistance of the 7 surrounding soll on piping deformations or ; The procedure and criteria for analysis are displacements, differential movements of j described in Subsections 3.7.2.1.3 and piping anchors, bent geometry and curvature ' 3.7.3.3.1.3. changes, etc., are considered. _ When-applicable, procedures utilizing the 3.7.3.10 Use of Constant Vertical Static principles of the theory of structures on Factors clastic foundations are used. All Sei;mic Category I subsystems and compo. (3) When applicable, the effects due to local nents are subjected to a vertical dynamic soll settlements, soll arching, etc., are pnalysis with the vertical floor spectra or time also considered in the analysis, hhtories defining the input. A static analysis is performed in lieu of dynamic analysis if the 3.7.3.13 lateraction of 0ther Piping with , peak value of the applicable. floor spectra times Seismic Categon! Piping ' a factor of 1.5 is used in the analysis. A 9- factor of 1.0 instead of 1.5 can be used if the In certain instances, non. Seismic Category I equipment is simple enough such that it behaves piping may be connected to Seismic Category I essentisily as a single degree of freedom piping at locations other than a piece of equip-3 system, if the fundemental frequency of a compo- ment which, for purposes of analysis, could be M r i in the vertical direction is greater than or represented as an anchor. - The transition points qual to 33 Hz, it is treated as seismicsJly typically occur at Seismic Category I valves rigid and analyred statically using the which may or may not be physically anchored. zero pe sponse spectrum. Since a dynamic analysis must be modeled from
.. pipe anchor point to ancher point, two options -
3.7,3.11 Torsional ElTects of Eccentric Masses exist: Torsional effects of eccentric masses are (1) specify and design a structural anchor at included for Seismic Category I subsystems the Seismic Category I valve and analyze the similar to that for the piping systems discussed Seismic Category I subsystem; or, if in Subsection 3.7.3.3.1.2. Impractical to design an anchor, 3.7 3.12 Puried Seismic Category I Piping and (2) analyze the subsystem from the anchor point Tunnels in the Seismic Category I subsystem through the valve to either the first anchor point
- For buried Category I buried piping systems in the non. Seismic Category I subsystem; or ,
and tunnels the following items are considered in to sufficient distance in the non Seismic - the analysis: Category I Subsystem so as not to significantly degrade the accuracy of (1) The inertial effects due to an earthquake analysis of the Seismic _ Category I piping. upon buried systems and tunnels will be O Anendment 11 3.7-23
3SA6100AB StanA ed Plant anv. a
' SECTION 3.9 [
section CONTENTS (Continued) h 2hls Easa 33.2.2115- Other ASME RIEquipment 3.9 13 3.912.2.16 Supports 3.9 14 3.9.23 Dynamic Response of Reador Internals Under : Operational Flow Transients and Steady State t'nnaletana 3 S.14 3.9.2.4~ Prooperational Flow Induced Vibration - Testing of Reactor Internals 3S 16 3.9.2.5 Dynamic System Analysis of Reactor Internals . t Under Faulted Conditions 3.9 17 3S.2.6 . Correlations of Reactor Internals Vibration Tests With the Analytical Results . SS-17.2 3S3 ASME Code t'tana 1. 2. and 3 Cam---- '= Caltaantat.Stagelskand Com Support Structures . 3.9 18 ! 'I 3.93.1 loading Combinations, Design Transients, and 1 Stress Limits 3.9 18 i 3S3.1.1 Plant Conditions 3.9 18.1 i. 3.93.1.1.1 NormalCondition 3.9 18.1 3.93.1.1.2 Upset Condition 3.9 18 3.93.1.13 Emergency Condition 3.9 19 i 3.93.1.1.4 Faulted Condition 3.9 19 3S3.1.1.5 Correlation of Plant Condition with Event l Probability 3.9 19 3S3.1.1.6 ' Safety Class Functional Criteria 3.9-19 3S3.1.2 Reactor Pressure Vessel Assembly 3.9-20 393.13 Main Steam (MS) System Piping 3 9-20 Amendment it
. - - - ..-. . -. . ~ _ - - - - - - - - - _ - _ _ _ _ . _ _ _ _ _ _ _ - - _ - - _ _ _ -
J MM : sautomAs lEtamAard Plant REV.A SECTION3.9 ; O = c ""
= n. !
i 3,9,1 spadat Topies far Machanimi canonsats 3Al 3.9.1.1 Design?malents 3.91 3.9J.2 Computer Programs Used in Analyses 3A1_ , 3.9.1.3 ExperimentalStress Analysis 3Al. 3.9.13.1 Piping Snubbers and Restraints 3.91, 3.9.1.3.2 Pine Motion Control Rod Drive (FMCRD) 3.9-1 3.9.1.4 Considerations for the Evaluation of Fauhed Conditions 3.9-1 I 3.9.1.4.1 Control Rod Drive System Components 3.92 3.9.1.4.1.1 Fine Motion Control Rod Drive . 3.92 [ 3.9.1.4.1.2 Hydraulic Control Unit. 3.92 39.1.4.2 . Reactor Pressure Vessel Assembly 3.92 3.9.1.4.3 Core Support Structures and Other Safety Reactor . InternalComponents 3.92 3.9.1.4.4 RPV Stabilizer and FMCRD and Incore Housing Restraints (Supports) 3.9-2 3.9.1.4.5 Main Steam Isolation Valve, Safety / Relief !.- Valve and Other ASME Class 1 Valves .3.92 t 3.9.1.4.6 ECCS and SLC Pumps, RRS and RHR Heat hh= yrs, RCIC Turbine and RRS Motor 3.92. 3.9.1.4.7 Fuel Storage and Refueling Equipment 3.9 3.9.1.4.8 Fuel Assembly (Including Channel) 3S3 3 . 9 -11 Aswadment 1 8 _ _ . . . . - _ _ , . _ , __ _ , . _ . . _ . . . _ _ ..i.
ABM w.sim Standad Plant - anv. n SECTION3.9 TABLES h ro, nu, e l 3.91: Plant Events ' 3.9 46' 3.92 Imad Combinations and Acceptance Criteria for - l l Safety-Related, ASME Code Class 1,2 and 3 , 1- Components, Component Supports, and Class CS Structures' 3.9 49 33 3 Pressure Differentials Across Reactor Vessel laternals 3S 53
'3.94 l Deformation Limit, for Safety Class Reactor Internal Structures Only 3.9 54 ,
3.9-5 Primary Stress Ilmit, for Safety Class Reactor laternal Structures Only 3.9-55 3.9-6 Buckhng Stability limit, for Safety Class ReactorInternal Structures Only 3.9-57 33 7 Fatigue Limit, for Safety Class Reactor , Internal Structures Only - 3.9 58 . m.
' ILLUSTRATIONS Figure Ihlt East 33-la Transient Pressure Differential Following a Steam Line Break 3 S-58.1 3.9-1b Reactor Vessel and Internal Assembly 3.9-59 3.92 Resclor Internal Flow Paths and Minimum Fbodable Volume 3.9 60 :
-3.9-3 ABWR Recirculation Flow Path 3.9 3.9-4 FuelSupport Pieces 3 S-62 3S-5 Pressure Nodes for Depressurization Analysis - 3.9 '
3S-6 Stress Strain Curve for Blowout Restraints 3.9-64 > J 3.9 xi Amendment 11 l
.i
MM ' 2sAstooAB RimmAnzd Plant _ REV. A - SECTION3.9 CONTENTS (Continued) Eastian 2Rin Ense
' 3 3 1 2.1 Tests and Analysis Criteria and Methods 3S.9 3.9.22.1.1 Random Vibrationinput _3.9-10 33.2.2.1.2 - Application ofInput Modes _ 3S 10 3S 12.13 Fluture Design 3.9 10 -
33.211.4- Prototype Testing 3 S-10 33.2.2.2 Qualification of Safety-Related Mechanical ! Equipment 3.9 10 j 3S.2.211 CRD and CRD Housing 3S-10 33.2.2.2.2 Core Support (Fuel Support and CR Guide Tube) ' 3S 11 3S.2.2.23 Hydraulic Control Unit (HCU). 3.9 11. 3.9.2214 Fuel Assembly (lacluding Channe.1) 3S 11 3S12.2.5 Reactor Internal Pump and Motor Assembly - 3.9-11 3S1216 ECCS Pump and Motor Assembly 3.911 = l i 3.9.2.2.2.7 RCIC Pump and Turbine Assembly 3 S-11' l 3.9.2.2.2.8 Standby Liquid Control Pump and Motor Assembly 3.9-12 3.9.2.2.2.9 RMC and RHR Heat Exchangers 3S 12 - 3 9.2.2.2.10 Standby Liquid ControlTank- 3.9 12 3.9.2.2.2.11 Main Steam Isolation Valves 3.9 12 i 3.9.2.2.2.12 Standby Liquid Control Valve (Injection Valve) 3.9 12 3.9.2.2.2.13 Main Steam Safety / Relief Valves 3.9 12 3 9.2.2.2.14 Fuel Pool Cooling and Cleanup System Pump and a Motor Assembly 3.9 12 f Amendsunt 1 i f
4
=
MM panAnd Mant saA610QAE REV. B The loading combinations and design criteria . 3.9J.I Piping Vibration. Thermal Espansion, for pipe whip restraints utilized to mitigate the and Dynamic Effects ! effects of postulated piping failures are ' provided in subsection 3.6.2.3.3. The overall test program is divided into two phases; the preoperational test phase and I In the case of the RIP motor casing failure the initial startup test phase. Piping vibra-event, there are specific restraints applied to tion, thermal expansion and dynamic effects test-mitigate the effects of the failure. The ing will be performed during both of these f mitigation arrangement consists of lugs on the phases as described in Chapter 14. Subsections , RPV bottom head to which are attached two long 14.2.12.1.51,14.2.12.2.10 and 14.2.12.2.11 re- ! rods for each RIP. The lower end of each rod late the specific role of this testing to the ov. { engages two lugs on the RIP motor / cover. He use erall test program. Discussed below are the gen- j of inelastic analysis methods is linilted to the oral requirements for this testing. It ; middle slender body of the rod itself. The i attachment lugs, bolts and clevises are shown to l be adequate by clastic analysis. The selection i of stainless steel for the rod is based on its high ductility assumed for energy absorption j during inelastic deformation. I The mitigation for the CRD housing attachment weld failure is by somewhat different means than are those of the RIP in that the components with regular functions also function to mitigate the weld f ailure effect. The components are specifically: (1) Core support plate (2) Controlrod guide tube [ (3) Control rod drive housing : (4) Controlrod drive outer tube i i (5) Bayonet fingers 1 Only the cylindrical bodies of the control rod guide tube, control rod drive housing and control rod drive outer tube are analyzed for energy absorption by inelastic deformation. Inclastic analysis for there latter two events together with the criteria used for evaluation are consistent with the procedures ; described in Subsection 3.6.2.3.3 for the different components of a pipe whip restraint. Figure 3.9 6 shows the stress strain curve used for the blowout restraints. 3.9.2 Dynamic Testing and Analysis ' l Amendment 11 3.93.1
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SECTION3.9 g Sacdon CONTENTS (Continued)- 310A East 3.93.2 1.meding Conditions - 3.9 41 3.9.5.2.1 Events to be Evaluated 3.9 41 33.5.2.2 l Pressure Differential During Rapid Depressurization 3.9 41 , t 3.9.5.23 - . Feedwater IJae and Main Steam 1Jae Break 3.9 42 33.5.2.3.1 . Accident Definition 3.9-42 : 3S.5.23.2 Effects oflaltlal Reactor Power and Core Flow . 3.9 42 l 3.9.5.2.4 Seismic and Other Reactor Building Vibration Events 3.9 42.
'3.9.53 Design Bases 3.9 43 i j 3.9.53.1 Safety Design Bases' 3S 43 39.53.2 Power Generation Design Bases 3.9-43
.3.9.533 Design 1.cading Categories 3.9 43 a 3.9.53.4 Response oflaternals Due to Steam Line Break Accident 3 S-43 y
39.53.5 Stress and Fatigue Limits for Core Support Structures d 39 43 '; 3.9.53.6 Stress, Deformation, and Fatigue Limits for Safety Class Reactor Internals (Except Core Support Structures) 3.9-44 - ! 3.9.6 lamice Testina of Puman and Valves -3.9 44 i 33.6.1 , laservice Testing of Pumps 3.9-44
# 39.6.2 laservice Testing of Valves 3.9-44 5
3.9.7 Interfaces 3 S 45 3S.7.1 Reactor Internals Vibration Analysis, 3 S-45 Measurement and Inspection Programs = 3.9.7.2 ASME Class 2 or 3 or Quality Group 3S 45 Components with 60 Year Life l 3S.8 References 3.9-45 3.9-x Amundment 11
ABM 2msiaa$s enandard Plant nev. n analyzed for the faulted loading conditions. The 33.1A.10 ASME Class 2 and 3 Pumps j ECCS and SLC pumps are active ASME Class 2 compo- _
'U nents. The allowable stresses for active pumps . Elastic analysis methods are used for evaluat.
are provided in a footnote to Table 3.9 2. lag faulted loading conditions for Class 2 and 3 pumps. The equivalent allowable stresses for i The reactor coolant pressure boundary compo- nonective pumps using elastic techniques are ob-nents of the reactor recirculation system (RRS) tained from NC/ND 34000f the ASME Code Section pump motor assembly, and recirculation motor cool. - III. These allowables are above clastic lim- : ing(RMC) subsystem heat exchanger are ASME Class its. The allowablea for active pumps ere pro. . 1 and Class 3, respectively, and are analysd for vided in a footnote to Table 3.9 2. l the faulted loading conditions. All equipment 1 stressea arc within the clastic limits. ' 3.9.1A 11 ASME Class 2 and 3 Valves , i 33JA.7 Fuel Storape and itefbellag Equipment Elastic analysis methods and standard design rules are used for evaluating faulted loading Storage, refueling, and servicing equipment conditions for Class 2, and 3 valves. The which is important to safety is classified as es- equivalent allowable stresses for nonnetive sential components per the requirements of- valves using clastic techniques are obtained 10CFR50 Appendix A. This equipment and other from NC/ND 3500 of ASME Code,: Section III. , equipment which in case of a failure would de- These allowables are above clastic limits. The { grade an essential component is defined in Sec- allowables for active valves are provided in a , tion 9.1 and is classified as Seismic Category footnote to Table 3.9 2. ! I. These components are subjected to an clastic l dynamic finite element analysis to generate load. 3.9.1A.12 ASME Class 1,2 and 3 Piping ings. This analysis utilizes appropriate floor response spectra and ccmbines loads at frequen. Elastic analysis methods are used for evaluat-cies up to 33 Hz for seismic loads and up to 60 ing faulted loading conditions for Class 1,2, Hz for other dynamic loads in three directions, and 3 piping. The equivalent allowable stresses Imposed stresses are generated and combined for using clastic techniques are obtained from Appen-normal, upset, and faulted conditions. Stresses dix F (for Class 1) and NC/ND 3600 (for Class 2 are compared, depending on the specific safety and 3 piping) of the ASME Code Section III. ; class of the equipment, to Industrial Codes, - These allowables are above clastic limits. The ' ASME, ANSI or Industrial Standards, AISC, allowables for functional capability of the es-
. allowables, sential piping are provided in a footnote to Table 3.9 2, 39.1A.8 Fuel Assembly (including Channel) .
3.9.13 laelastic Analysis Methods , GE BWR fuel assembly (including channel) de- ' sign bases, and analytical and evaluation methods Inelastic analysis is only applied to ABWR including those applicable to the faulted condi- components to demonstrate the acceptability of tions are the same as those contained in Refer- three types of postulated events. Each event is , ences 1 and 2. an extermly low probability occurence and the equipment affected by these events'would not be 3.9.1A.9 ASME Class 2 and 3 Vessels reused. These three events are: Elastic analysis methods are used for evaluat- (O Postulated gross piping failure. ing faulted loading conditions for Class 2 and 3 vessels. The equivalent allowabic streses using -(2) Postulated blowout of a reactor internal. elastic techniques are obtained from NC/ND 3300 recirculatica (RIP) motor casing due to a and NC 3200 of the ASME Code Section III. The,se weld fahrt allowables are above clastic limits. (3) Postulated olowout of a control rod drive (CRD) housing due to a weld failure. O Amendment 11 3.9-3 l 1
IsA6100AE , flilf andard Plant an n l (1). the piping system durrag system heatup and measurements, using a hand held scale or ruler, cooldown is free to expand and move without against a fixed reference or by recording the j unp!n.ned obstruction or restraint in the x, position of a snubber or spring can. A more .- ! y, and z directions; precise method would be using permanent or ' temporary instrumentation that directly measures - (2)L the piping system does shakedown after a few displacement, such as a lanyard potentiometer, thermal expansion cycles; that can be monitored via a remote indicator or' recording device. The technique to be used will (3) the piping system is working in a manner con- depend on such factors as the amount of movement , sistest with the assumption of the stress predicted and the assessability cf the piping. , analysis; ; Measurement of piping temperature is also of - (4) there is adequate agreement between calcu- importance when evaluating thermal expansion. letW values and measured values of displace. This may be accomplished either indirectly via L mems; and ' the temperature of the procese fluid or by ; direct measurement of the piping wall tem- ' (5) there is consistency and repeatability in perature and such measurements may be obtained
- thermal displacements during heatup and either locally or remotely. The choice of tech-cooldown of the systems, alque use'd shall depend os such considerations as the accuracy required and the assessability i The general requirements for thermal expan- of the piping, i sion testing of piping systems are specified in
, Regulatory Guide 1.68, 'Preoperational and Ini- 3.9.21.2.2. Monitoring Requiresnents - - , E tial Startup Testing Programs for Water Cooled P. Power Reactors.' More specific requirements are As described in Subsections 14.2.12.1.51-i defined in ANSI /ASME OM7 ' Requirements for and 14.2.12.2.10 all safety related piping shall i Thermal Expansion Testing of Nuclear Power Plant be included in the thermal expansion testing pro-( Piping Systems." Detailed test specifications gram. Thermal expansion of specified piping sys- . l will be prepared in full accordance with this tems should be measured at both the cold and hot : - standard and will address such issues as extremes of their expected operating condi-prerequisites, test conditions, precautions, tions. Physical walkdowns and recording of measurement techniques, monitoring requirements,- hanger and snubber positions should also be con. ? test hold points and acceptance criteria. The ' ducted where possible considering assessability _, development and specification of the types of and local environmental and radiological condi. 'l
- measurements required, the systems and locations tions in the hot and cold states. Displacements !
l to be monitored, the test acceptance criteria, and appropriate piping / process temperatures - ; ! and the corrective actions that may be necessary shall be recorded for those systems and condi-L are discussed in more detail below, tions specified. Sufficient time shall have
- passed before taking such measurements to ensure l 3.9 2.1.2.1 Measurement Techniques the piping system is at a steady state condi-t tion. In selecting locations for monitoring pip.
3 Verification of acceptable thermal expan- ing response, consideration shall be given to sion of.specified ' piping systems can be accom. the arximum responses predicted by the piping i plished by several methods. One method is to analysis. Specific consideration should also be
- physically walkdown the piping system and verify given to the first run of pipe attached to 1 by visual observation that free thermal movement component nozzles and pipe adjacent to.
, is unrestrained. This might include verification structures requiring a controlled gap. that piping supports such as snubbers and spring hangers are not fully extended or bottomed out 3.9.2.1J.3 Test Evaluation and Acceptance Crl-and that the piping (including branch lines and teria t instrument lines) and its insulation is not in 4 hard contact with other piping or support To ensure test data integrity and test structures. Another method would involve local safety, criteria have been established to fa-Amendment 11 3.9-6 l t l
MN . IsA6100AE 3 fdapdard Pinnt arv.m i should be noted that because one goal of the dy- to more rigorous testing and precise instrumenta-namic effects testing is to verify the adequacy tion requirements and, therefore, will require O of the piping support system, such components are remote monitoring techniques. Local measurement addressed in the subsections that follow. How- techniques, such as the use of a hand held ever, the more specific requirements for the de- vibrometer, are more appropriate in cases where sign and testing of the piping support system are it is expected that the vibration will be less described in Subsection 3.9.3.4.1. complex and oflessor magnitude. Many systems that are assessable during the preoperational 3 9.2.1.1 Vibration and Dynamic Effects Testing test phase and that do not show significant latersystem interactions will fall into this cat. The purpose of these tests is to confirm egory. Visual observations are utilized where that the piping, components, restraints and sup- vibration is expected to be minimal and the need { , ports of specified high and moderate energy sys- for a time history record of transient behavior ' tems have been designed to withstand the dynamic is not anticipated. However, unexpected visual j effects of steady state flow indue:d vibration observations or local indications may require i and anticipated operational transient condi- that a more sophisticated technique be used, tions. The general requirements for vibration Also, the issue of assessability should be con- ,' and dynamic effects testing of piping systems are sidered. ' Application of these measurement tech. specified in R egula tor y G uid e.1.68, niques is detailed in the appropriate testing
'Preoperational and Initial Startup Test Programs specification consistent with the guidelines con-for Water Cooled Power Reactors". More specific tained in ANSI /ASME OM3, [
vibration testing requirements are defined in ! ANSI /ASME OM3,' Requirements for Preoperational 3 9.2.1.1.2 Monitoring Requirements and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems'. Preparation of As described in Subsection 14.2.12.1.51, detailed test specifications will be in full 14.2.12.2.10 and 14.2.12.2.11 all safety related accordance with this standard and will address piping systems will be subjected to steady state O-k such issues as prerequisites, test conditions, and transient vibration measurements. The scope precautions, measurement techniques, monitoring of such testing shall include safety related in-requirements, test hold points and acceptance - strumentation piping and attached small bore pip-M criteria. The development and specification of ing (branch piping). Special attention should the types of measurements required, the systems be given to piping attached to pumps, compres- ! and locations to be monitored, the test sors, and other rotating or reciprocating equip-acceptance criteria, and the corrective actions ment. Monitoring location selection consider-that may be necessary are discussed in more ations should include the proximity of isolation detail beltw. valves, pressure or flow control valves, flow ,! orifices, distribution headers, pumps and other - 3.9.2.1.1.1 Measurement Techniques elements where shock or high turbulence may be h of concern. Location and orientation of instru-There are essentially three methods avall- mentation and/or measurements will be detailed able for determining the acceptability of steady in the appropriate' test specification, state and transient vibration for the affected Monitored data should include actual deflections systems. These three measurement techniques are and frequencies as well as related system operat-visual observation, local measurements, or re- Ing conc 'nns. Time duration of data recording motely monitored / recorded measuremerts. The tech. should be .aicient'to indicate whether the vi-nique used in each case will depend on such fac. bration is continuous or transient. . Steady tors as the safety significance of the particular state monitoring should be performed at critical system, the expected mode and/or magnitude of the conditions such as minimum or maximum flow, or vibration, the assessability of the system during abnormal combinations or configurations of designated testing conditions, or the need for a system pumps or valves. Transient monitoring time history recording of the vibratory behav. should include anticipated system and total lor. Typically, the systems where vibration has plant operational transients where critical the greatest safety implication will be subject piping or components are expected to sher Amendment 11 3,94 m ~
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MM...a as.. pi..s - 2SA6100AE :
.m the case of other RBV induced loads is (3) the input has sufficient intensity and discovered, dynamic tests and/or mathematical - duration to excite'all modes to the analyses may be used to verify operability and required magnitude so that the testing structural integrity at the required dynamic response spectra will envelop the- _ '
L input conditions. corresponding response spectra of the L . individual modes. - , When the equipment is qualified by dynamic ' test, the response spectrum or time history of 3.922.12 Appliention efinput Modes ; the attachment point is used in determialag input, motion. When dynamic tests are performed, the ~ Li .nput motion is applied to one vertical and Natural frequency may be determined by running one horizontal axis simultaneously. However, a continuous sweep frequency search using al If the equipmes; response along the vertical - sinusoidal steady. state input of low magnitude. Dynamic load conditions are almulated motion by testing along the ~' direction is not sensitive horizontal direction and vice- to the vibra! using random vibration input or single frequency versa, then the input motion is applied to one - input (within equipment capability) over the - direction at a time. In the case of single i frequency range of interest. Whichever method is frequency inputithe time phasing of the used, the input amplitude during testing . inputs in the vertical and horizontal envelopes the actual input amplitude eccted. direc*.lons are such that a purely rectilinear 1 during the dynamic loading condition._ - resultant input is avoided. , a The equipment being dynamically tested is
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3.9.22.1.3 Fistum Design mounted on a fixture which simulates the intended service mounting and causes no dynamic coupling The fixture design simulates the actual to the equipment, service mounting and causes no dynamic , L coupling to the equipment. Equipment having an extended structure, such
- as a valve operator, is analyzed by applying 3.9.2.2.1A Prototype Testing - - ;
static equivalent dynamic loads at the center of .
. gravity of the extended structure. _ Jn cases Equipment testing is conducted on where the equipment structural complexity makes. prototypes of the equipment to be installed in. ' mathematical analysis impractical, a static bend the plant.
test is used to determine spring constant and i operational capability at maximum equivalent 3.9.2.2J Qualification'of Safety Related dynamic load conditions. Mechanical Equipment 3.9.2J.1.1 Random Vibration input The following subsections discuss the , testing or analytical qualification of the When random vibration input is used, the- ~ safety.related major mechanical equipment, and 3 actualinput motion envelopes the appropriate other ASME III equipment, including equipment floor input motion at the individual modes. supports. However, single frequency input such as sine .
.i beats can be use provided one of the following 3.9.2.2.2.1 CRD and CRD Housing conditions are met:
The qualification of the CRD housing (with (1) the characteristics of the required input . enclosed CRD) is done analytically, and the motion is dominated by one frequency; stress results of their analysis establish the structural integrity of these components. (2) the anticipated response of the equipment is Preliminary dynamic tests are conducted to adequately represented by one mode; or verify the operability of the control rod Amendment i O 19-10 _ _ ___ -- _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . .,_ _ 4-_ _ _
masi ass A.B a..aM . pin.s mm a l significant response. Steady state conditions During the course of the tests, the remote j and transient events to be monitored will be snessurements will be regularly checked to verify 1 detailed in the appropriate testing specification compliance with acceptance criteria. If trends ' consistent with OM3 guidelines. Indicate that criteria may be violated, the men. I surements should be monitored at more frequent 3.9J.1.1J Test Evalnation and Acceptance Crite. Intervals. The test will be held or terminated ria as soon as criteria are violated. As soon as possible after the test hold or termination ap-The piping response to test conditions propriate investigative and corrective actions shall be considered acceptable if the review of will be taken. If practicable, a walkdown of l the test results indicates that the piping re. the piping and suspension system should be made sponds in a manner consistent with predictions of in an attempt to identify potential obstructions
- the stress report and/or that piping stresses are or improperly operating suspension components.
, within ASME Code Section III (NB 36000) limits. Hangers and snubbers should be positioned such l Acceptable limits are determined after the comple- that they can accommodate the expected deflec- i tion of piping systems stress analysis and are tions without bottoming out or extending fully.
' provided in the piping test specifications. All signs of damage to piping supports or an-chors shall be investigated. ;
, To ensure test data integrity and test
- safety, criteria have been established to fa. Instrumentation indicating criteria failure ,
! cilitate assessment of the test while it is in- shall be checked for prc'per operation and ' progress. For steady state and transient vibra- calibration including con parison with other in-tion the pertinent acceptance criteria are usu- strumentation located in the proximity of the ex-i ally expressed in terms of maximum allowable dis- cessive vibration. The assumptions used in the placement / deflection Visual observation should calculations that generated the applicable lim- ) only be used to confirm the absence of sig- its should be verified against actual conditions ! nificant levels of vibration and not to determine and discrepancies noted should be accounted for acceptability of any potentially excessive vibra- in the criteria lir its. This may require a i tion. Therefore,in some cases other measurement reanalysis at actual system conditions. - a
- techniques will be required with appropriate quan-l titative acceptance criteria. Should the investigation of instrumentation
' and calculations fall to reconcile the criterie There are typically two levels of acceptance ' violations, then physical corrective actions seay criteria for allowable vibration displacet be required. This might include identification ments/ deflections. Level I criteria are bounding and reduction or elimination of offend!ng fore-type criteria associated with safety limits while ing functions, detuning of resonant piping spans Level 2 criteria are stricter criteria associated by appropriate modifications, addition of brac-with system or component expectations. For ing, or changes in operating procedures to avoid steady state vibration the Level I criteria are troublesome conditions. Any such modifications based on the endurance limit (10,000 psi) to as- will require retest to verify vibrations have sure no failure from fatigue over the life of the been sufficiently reduced. plant.. The corresponding Level 2 criteria are based on one half the endurance limit (5,000- 3.9 2.12 'Ihermal Expansion Testing psi). For transient vibration the Level 1 crite-i- ria are based on either the ASME III code upset A thermal expansion preoperational and startup !' primary stress limit or the applicable snubber testing program performed through the use of vi-
~
l load capacity. Level 2 criteria are based on a sual observation and remote sensors has been ! given tolerance about the expected deflection established to verify that normal unrestrained value. thermal movement occurs in specified safety-related high and moderate energy piping sys-l 3.9.2.1.1.4 Reconcillation and Corrective Ac. tems. The purpose of this program is to ensure l tions the following: lO M -at3 l 3.95 l i l
M 2sA6100AE EfenAned Plant am n i laternals to the RBV is also determined with the reactor and internals are performed. The dynamic model and dynamic analysis method results of these analyses are used to generate , described below for seismic analysis. the allowable vibration levels during the vibration test. The vibration data obtained -i (4) 10CAlands The Assumed LOCA also resuks in during the test will be analyzed in detail. ' RBV due to suppression pool dynamics as , described in Appendix 3B and the response of the reactor internals are again determined ; with the dynamic model and dynamic analysis method used for seismic analysis. Various types of LOCA loads are identified on Table 3.92. (5) Selsele Leads The theory, methods, and
- computer codes used for dynamic analysis of the reactor vessel, internals, attached !
piping and adjoining structures are described in Section 3.'1 and Subsection 3.9.1.2. Dynemic analysis is performed by ( coupling the lumped mass model of the reactor vessel and laternals with the building model to determine the system { .i natural frequencies and mode shapes. The i 2 relative displacement, acceleration, and
- load response is then determined by either the time history method or the resonse spectrum method. The load on the reactor internals due to faulted event SSE are obtuined from this analysis. 1 The above loads are considered in combination as defined in Table 3.9 2. The SRV. LOCA (SBL, IBL or LBL) and SSE loads as defined in Table 3.9 2 are all assumed to act in the same direction. The peak colinear respcases of the ,
reactor internals to each of these loads are added by the square root of the sum of the squares (SRSS) method. The resultant stresses in the reactor internal structures are directly added with stress resulting from the static and steady state loads in the faulted load' combination, including the stress due to peak
. reactor internal pressure differential during the i
LOCA. The reactor ;uernals satisfy the stress deformation and fatigue limits as defined in Subsection 3.9.5.3. 3.9.2.6 Correlations of Reactor laternals Vibration Tests With the Analytlent Results Prior to initiation of the instrumented vibration measurement program for the prototype plant, extensive dynamic analyses of Asneedment 8 3.9 17.1
ABM BeamAnssi Plane
. siasisaus .
any. n . 1 33J2 melamle Qomuscaties et O Safety Balated Mechanleal Egalpensat Gacinding Other RBVladesed Leeds)' This subsection describes the criteria for dynamic qualification of safety related'
. mechanical equipment and associated supports,:
and also describes the qualification testing and/or analysis applicable to the major: composeats on a component by componest basis. Seismic and other events that may induce' -i reactor building vibration'(RBV)-(see Appendia
~
> l
- 38) are considered. In some cases, a module or assembly consistlas of mechaalcal and' I,
' electrical equipment is qualified as a unit .
-(e.g., ECCS pumps). These modules are generally discussed la this subsection and-
' Subsection 3.9.3.2 rather than providing discussion'of the separate electrical parts in Section 3.10. Electrical supporting equipment _. i a such as control consoles, cabinets, and panelsi !
g are discussed in Section 3.10. l
.. f '
3.9.2J.1 Tests and Analysis Criteria and . Methods , The ability of equipment to perform its safety function during and after the 'j application of a dynamic load is demonstrated by tests and/or-analysis.; The analysis la performed in accordance with Section 3.7. - ! Selection of Testing.:analy' sis or a ; combination of the two is determined by the .
' type, size, shape, and compiculty of the equiptnent being considered. When practical, .
the equipment operability is demonstrated by t e s ting. Otherwise, operability is demonstrated by mathematical analysis. Equipment which is large, simple, and/or consumes large amounts of power.is usually . ; qualified by analysis or static bend test to i show that the loads, stresses and deflections are less than the allowable maximum. Analysis-and/or static bend testing is also used to
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show there are no natural frequencies below 33 , Hz for se.:smic loads and 60 Hz for other RBV
- The 60 Hz frequency cutoff for dynamic loads'. 'If a natural frequency lower than 33
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analysis of suppression pool dynamic loads is the Hz ls the case of seismic loads and 60 Hz in minimum requirement based on a generic Reference . 8, using the' missing strain energy method, performedfor representative BWR equipment under high. frequency input loadings. O Amendment 11 3.99-d a ' -i
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P MM 21A6100AE ' nev. n Etandmed Plant
' 3.93J.1.1- Normal Condition -
Normal conditions are any conditions in the course of system startup, operation in the design power range, normal hot standby (with condenser available), and system shutdown other than upset, emergency, faulted, or testing. 3.93.1.1J Upset Condities An upset condition is any deviation from normal conditions anticipated to occur often . enough that design should include a capability to-withstand the conditions without operational
. impairment The upset conditions include system operational transients (SOT) which result from eny single operator error or control malfunction, from a fault in a system component requiring its isolation from the system, from a loss of load or power, or from an operating basis carthquake, ,
Hot standby with the main condenser isolated is an upset condition. 1 D G: r i 4 i l i 4
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.j Amendment 11 3.9 18.1 .
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1: ABWR z wiaans ! Standard Plant an n l1 pletion of preoperational* testing, the reactor 3.92.8 Dynamic Systen Analysis et Rasetor vessel head and the shroud head are removed, the asteenal:Under Fanised Comentions vesselis drained, and major components are inspected on a selected basis. The inspections The faulted events that are evaluated are cover the shroud, shroud head, core support defined in Subsection 3.9.5.2.1. The loads structures, recirculation laternal pumps, the that occur as a result of these events and the peripheral control rod drive, and incore guide analysis performed to determine the response tubes. Access is provided to the reactor lower of the reactor laternals are as follows: plenum for these inspections. (1) Reactor Internal Pressures . The reactor The analysis, design and/or equipment that are laternal pressure differentials (Figure to be utilized in a facility will comply with 3.9 la) due to assumed break of main steam , Regulatory Guide 1.20 as explained below, or feedwater line are determined by analysis as described in Subsection Regulatory Guide 1.20 describes a 3.9.5.2.2. In order to assure that no comprehensive vibration assessment program for significant dynamic amplification of load 3 reactor internals during preoperational and occurs as a result of the oscillatory ' initial startup testing. The vibration nature of the blowdown forces during an assessment program meets the requirements of accident, a comparison is made of the Criterion 1. Quality Standards and Record,- periods of the applied forces and the Appendix A to 10CFR50 and Section 50.34, Contents natural periods of the core support of Applications; Technical Information, of structures being acted upon by the applied - 10CFR50. This Regulatory Guide is applicable to ' forces. These periods are determined the core support structures and other reactor from a comprehensive vertical dynamic inte nals, model of the RPV and internals with 12 degrees of freedom. Besides the real l a 3
; ration testing of reactor internals is masses of the RPV and core support performed on all GE BWR plants. At the time of structures, account is made for the water '
original issue of Regulatory Guide 1.20, test inside the RPV, programs for compliance were instituted for the ! then desigt.ed reactors; The first ABWR plant is (2) External ~ Pressure and Forces on the considered a prototype and is instrumented and Reactor Vessel An assumed break of the subjected to preoperation and startup flow main steam line, the feedwater line or the < testing to demonstrate that flow induced RHR line at the reactor vessel nozzle vibrations similar to those expected during results in jet reaction and impingement operation will not cause damage. Subsequent forces on the vessel and asymmetrical-plants which have internals similar to those of pressurization of the annulus between the the prototypes are also tested in compliance with reactor vessel and the shield wall, the requirements of Regulatory Guide 1.20. GE is These time varying pressures are applied committed to confirm satisfactory vibration to the dynamic model of the reactor vessel performance of internals in these plants through system. Except for the nature and preoperational. flow testing followed by locations of the forcing functions, the inspection for evidence of excessive vibration, dynamic model and the dynamic analysis Extensive vibration measurements in prototype method are identical to those for seismic , plants together with satisfactory operating analysis as described below. The experience in all BWR plants have established the resulting loads on the reactor internals, adequacy of reactor internal designs. GE defined as LOCA loads, are considered as continues these test programs for the generic shown in Table 3.9.2. plants to verify structural integrity and to establish the margin of safety. (3) Safety / Relief Valve Loads (SRV Loads) The discharge of the SRVs result in reactor See Subsection 3.9.7.1 for interface building vibration (RBV) due to O r'equirements of the reactor internals vibration testing program. suppression pool dynamics as described in Appendix 3B. The response of the reactor Anwadownt 11 3.9-17 1 5
awa RemmAmed Plant an n (3) Snubbers The operating loads on snubbers i are the loads caused by dynamic events (eg., seismic, RBV due to lhCA and SRV - discharge, discharge through a relief J valve-line or valve closure) during- ! L various operating conditions. Snubbers ; restrain piping against response to the vi-bratory excitation and to the associated j
. differential movement of the piping system '
support anchor points. The criteria for j locating snubbers and ensuris.g adequate 3 load capacity, the structural and me-chanical performance parameters used for - anubbers and the lastallation and inspec- i tion consider ations for the snubbers are I as follows: (a) Required Lead Capacity and Snubber Lo-cation The entire piping system including 3
. valves and support system between an.
. chor points is mathematically modeled j for complete piping structural analysis. In the dynamic analysis, the snubbers are modeled as a spring with a given spring stiffness depending on the snubber size. The .
analysis determines the forces and. .j moments acting.on each piping components and the forces acting on i the snubbers due to all dynamic ; loading and operating conditions ; defined in the piping design specification. The forces on snub-bers are operating loads for various operating conditions. Those loads are assumed no to exceed the snubber design load capacity for various operating conditions, i.e., design, normal, upset, emergency and faulted. 9 Aamodstat 11 3.9-31.1 l
ABM Ettandard Plant is4sioars any n .
.The results of the data analyses, vibration 3.9 2'and are contained in the design-l O
A./ amplitudes, natural frequencies, and mode shapes specifications and/or design reports of the are then compared to those obtained from the respective equipment._ (See Subsection 3.9.7.3-L ' theoretical analysis, for laterface requirements) L Such comparisons provide the analysts with Table 3.9 2 also presents the evaluation added insight into the dynamic behavior of the models and criteria. The predicted loads or reactor internals. The additional knowledge stresses and the design or allowable values - L gained from previous vibration tests has been for the most critical areas of each component utilized in the generation of the dynamic models are compared in accordance with the applicable i for seismic and loss of coolant accident (LOCA) code criteria or other limiting criteria. The - analyses for this plant. The models used for calculated results meet the limits, this plant are similar to those used for the ;' vibration analysis of earlier prototype BWR The design life for the ABWR Standard - , plants. Plant is 60 years. A 60 year design life is a l requirement for all major plant components 3.9.3 ASME Code Class 1,2, and 3 with reasonable expection of meeting this
- Components, Component Supports, and design life. However, all plant operational.
Core Support Structures components and equipment except the reactor vessel are designed to be replaceable, design 3.93.1 I.mading Combinations, Design !!fe aot withstanding.1 The de: Iga 11fe t Transients,and Stress Limits requirement allows for refurbihment and repair, as appropriate, to assure the design l This section delineates the criteria for life of the overall plant is achieved. In selection and definition of design limits and effect, essentially all piping systems,. ; loading combination associated with normal components and equipment are designed for a 60 operation, postulated accidents, and specified ~ year design life. Many of these components O i seismic and other reactor building vibration' are classified as ASME Class 2 or 3 or Quality (RBV) events for the design of safety related' : Group D. Applicants referencing the ABWR ASME Code components (except containment design willidentify these ASME Clast 2,3 amd t components which are discussed in Section 3.8). Quality Group D components and provide the analyses required by the ASME Code, Subsection This section discusses the ASME Class 1,2, NB. These analysis will include the and 3 equipment and associated pressure retaining appropriate operating vibration loads and for
. parts and identifies the applicable loadings, the effects of mixing hot and cold fluids, calculation methods, calculated stresses, and allowable stresses. A discussion of major 3.93.1.1 Plant Conditions eqaipment is included on a component by component basis to provide examples. Design transients and All events that the plant will or might dynamic loading for ASME Class 1,' 2, and 3 credibly experience during a reactor year are equipment are covered in Subsection 3.9.1.1. evaluated to establish design basis for plant Seismic related loads and dynamic analyses are equipment. These events are divided into four discussed in Section 3.7. The suppression plant conditions. The plant conditions pool related RBV loads are described in Appendix described in the following paragraphs are 3B. Table 3.9 2 presents the combinations of based on event probability (i.e., frequency of dynamic events to be considered for the design occurrence as discussed in Subsection and analysis of all ABWR ASME Code Class 1,2, 3.9.3.1.1.5) and correlated to service levels
, and 3 components, component supports, core for design limits defined in the ASME Boiler l support structures and equipment. Specific and Pressure Vessel Code Section III as shown loading combinations considered for evaluation of in Tables 3.91 and 3.9 2. each specific equipment are derived from Table .l lO
' Amendment 11 3.9 18 ;
1
23A6100AE Arandard Plant an n l 3.9.3.4 Component Supports correspond to those und for design of the sup-i ported pipe. The component loading l , The design of bolts for component supports combination are discussed in Subsection is specified in the ASME Code Section III, 3.9.3.1. TFe stress limits are per ASME III, Subsection NF. Stress limits for bolts are given Subsectior NF and Appendix F. Supports are in NF 3225. The rules and stress limits which general'y designed either by load rating must be satisfied are those given in NF 3324.6 ' meths per paragraph NF 3260 or by the stress ; multiplied by the appropriate stress limit factor ilmits for linear supports per paragraph ' l l for the particular service loading level and _ NF 3231. The critical buckling loads for the stress category specified in Table NF 3725.21. Class 1 piping supports subjected to faulted
- loads that are more severe than normal, upset Moreover, on equipment which is to be, or and emergency loads, ard determined by using may be, mounted on a concrete support, sufficient the methods discussed in Appendices F and XVII >
holes for anchor bolts are provided to limit the of the Code. To avoid buckling in the piping
- anchor bolt stress to less than 10.000 psi on the supports, the allowable loads are limited to i l nominal bolt area in shear or tension, two thirds of the determined critical buck!!ng i loads.
! Concrete anchor bolts which are used for pipe support base plates will be designed to the The design of all supports for non. nuclear applicable factors of safety which are defined in - piping satisfies the requirements of ANSI I&E Bulletin 79 02, ' Pipe Support Base Plate B31.1, Paragraphs 120 and 121. Designs Using Concrete Expansion Anchor Bolts," Revision 1 dated June 21,1979. For the major active valves identified in j Subsection 3.9.3.2.4, the valve operators are - ! 3.9.3.4.1 Piping not used as attachment points for piping 3 supports.
- p r
Supports and their attachments for essential 4 ASME Code Section III, Class 1,2, and 3 piping The design criteria and dynamic testing re-late designed in accordance with Subsection NF' up quirements for the ASME III piping supports [ q 'to the interface of the building structure. The are as follows- '
- a building structure component supports are de--
- signed in accordance with the AISC specification (1) Piping Supports All piping supports are .
for the Design, Fabsim!on, and Erection of designed, fabricated, and assembled so l Structural Steel for buildings. The loading com- that they cannot become disengaged by the -
- binations for the various operating conditions movement of the supported pipe of l equipment after they have been installed.
All pip _ing supports are designed in accordance with the rules of Subsection NF of the ASME Code up to the building structure interface as defined in the l
- Augmented by the following: (1) application of project design specifications.
Code Case N 476, Supplement 89.1 which governs i the der.ign of single angle members os ASME Class (2) Spring Hangers Thu operating load on i 1,2,3 and MC linear component supports; and (2) spring bangers is the load caused by dead i when eccentric loads or other torsiont.) loads are weight. The hangers are calibrated to en-
- not accommodated by designing the load to act sure that they support the operating load through the shear center or meet ' Standard for at both their hot and cold load settings.
Steel Support Design", analyses will be performed Spring hangers provide a specified down in accordance with torsional analysis methods travel and up travel in excess of the such as: " Torsional Analysis of Steel Members, specified thermal movement. USS Steel Manual *, Publication T114 2/83 or
' Design of Weld Structures
- by Omar W. Blodgett, etc. r Amendment 11 3.9-31 s
4
-$ - w -
-r-- ---- .
2%6100AB
- ArandardPlant REV.B system anchor displacements, and reaction (P/Perit) + (4/4 crit) + (t/ferit) forces caused by relief valve discharge or valve closure, etc.
< (1/S.F.)
Struts are desiped la accordance with ASME where: Code Section III, Subsection NF 3000 to bc capable of carrying the design loads for q' = longitudinalload { various operating conditions. As in case of P = enternal pressure i snubbers, the forces on struts are obtained r = transverse shear stress from an analysis, which are assured not to S.F. = safety factor exceed the design loads for various = 3.0 for design, testing, service operating conditions, levels A & B
= 2.0 for Service I.4 vel C 3.934.2 Reactor Pressure Vessel Support Sklet = 1.5 for Service level D. !
i The ABWR RPV support skirt is designed as an 3.9J.4.3 Renetor Pressere Vessel Stabiliser ASME Code Class 1 component per the requirements of ASME Code Section III, Subsection NF'. The The RPV stabilizer is designed as a Safety i loading conditions and stress criteria are given Class 1 linear type component support in-la Tables 3.91 and 3.9 2, and the calculated accordance with the requirements of ASME stresses meet the Code allowable stresses in the; Boiler _ and Pressure Vessel Code Section III - j critical support areas for various plant Subsection NF. The stabilizer provides a . operating conditions. The stress level margins reaction point near the upper end of the RPV - assure the adequacy of the ,RPV support skirt. An to resist horizontalloads due to effects such analysis for buckling shows that the support as earthquake, pipe rupture and RBV. -The skirt complies with Subparagraph F 1332.5 of ASME design loading conditions, and stress criteria III, Appendix F, and the loads do not exceed two are given in Tables 3.91 and 3.9 2, and the thirds of the critical buckling strength of the calculated stresses meet the Code allowable skirt. The permissible skirt. loads at any stresses in the critical support areas for elevation, when simultaneously applied, are various plant _ operating conditions. limited by the following interaction equation: 3.9.3AA Floor Mounted Msjor Equipment ' (Pumps, Heat Exchangers, and RCIC Turbine) Since the major active valves are supported 1 by piping and not tied to building structures, valve ' supports
- do not exist (See Subsection 3.9.3.4.1).
- The HPCF, RHR, RCIC, SLC, FPCCU,
' Augmented by the following: (1) application of SPCU, and RWCU pumps; RMC, RHR, Code Case N 476, Supplement 89.1 which governs RWCU, and FPCCU heat exchangers;and RCIC -
the design of single angle members of ASME Class turbine are all instyzed to verify the 1,2,3 and MC linear component supports; and (2) adequacy of their. support structure under j when eccentricloads or other torsionalloads are various plant operating conditions. In all not accommodated by designing the load to act cases, the load stresses in the critical > through the she,ar center or meet " Standard for support areas are within ASME Code allowables. Steel Support Design *, analyses will be performed in accordance with torsional analysis methods Seismic Category I active pump supports are such as: ' Torsional Analysis of Steel Members, qualified for dynamic (seismic and other RBV) USS Steel Manual", Publication T114 2/83 or loads by testing when the pump supports
' Design of Weld Structures
- by Omar W. Blodgett, etc.-
Amendment 11 3.9 34
23A6100AE ' Remaderd Plane wA Snubbers are generally used in agreement, they are brought in . situations where dynamic support is agreement, and the system analysis - required because thermal growth of the ~ is redone to confirm the snubber piping prohibits the use of rigid loads. This iteration is continued supports. The snubber locations and until all snubber load capacities support directions are first decided by and spring constants are ! estimation so that the stresses in the reconciled. i piping system will have acceptable .! values. The snubber locations and (c) Snubber Design and Testing I support directions are refined by 1 performing the dynamic analysis of the To assure that the required ! piping and support system as described atructural aad aeebanical ! above in order that the piping stresses performance characteristics and 4 and support loads meet the Code product quality are' achieved, the requirements, following requirements for design and testing are imposed by the
-The pipe support design specification design specification:
requires that snubbers be provided with position indicators to identify the rod - (i) The snubbers are required by position. This indicator facilitates the pipe support design-the checking of hot and cold settings of
^
specification to be designed the snubber, as specified in the in accordance with all of the~ installation manual, during plant rules'and regulations of the l preoperational and startup testing. ASME' Code Section Ill' , 1 l Subsection NF. This design (b) Inspection, Testing, Repair and/or requirement includes analysis
- Replacement of Snubbers
~
for .the: n or m al,' ups et, *
. emergency,'and faulted The pipe support design specification- loads. 'These calculated requires that the snubber supplier loads are then compared prepare an installation instruction against.the allowable loads - !
manual. This manual is required to to make-.sure that the contain complete instructions for the stresses are below the code testing, maintenance, and repair of the allowable limit, snubber. It also contains inspection points and the period of inspection. (ii) The snubbers are tested to insure that they can perform The pipe support design specification as required during the-requires that hydraulic snubbers be seismic and other RBV events, equipped with a fluid level indicator so and 'under anticipated-that the level of fluid in the snubber operational transient loads can be ascertained easily. or other mechanical loads associated with the design : The spring constant achieved by the requirements for the plant. i snubber supplier for a given load The f oilowIn g. t e s t capacity snubber is compared against the requirements are included: spring constant used in the piping j system model. If the spring constants o Snubbers ere subjected to are the same, then the snubber location { force or displacement versus and support direction become confirmed. time loading at frequencies If the spring constants are not in within the range of O Amendment 1 3,9 32 I
4 .- - . . _ . . _ _ _ _ - . . -- --- MM 2SA6100AB any. n EttanAmed Plane -I
- - i l
sigalficant modes of the piping (i) There are no visible signs ofl { ' system; damage ;or impaired operability n a result of {* l o Displacements are measured to storage, handling, or-l determine the performance lastallation, , characteristics specified; , l (ii) The snubber location,. 4
- o. Tests are conducted at various orientation, position -
d temperatures to ensure operability settles, and configurationi (' !) over the specified range;' (attachments, extensions, _ etc.) are according to design ~ i o Peak test loads in both tension and drawings and specifications. , I compression are required to be equal . _. . < to or higher than the rated load (iii) Snubbers are not seized, ! requirements; and frozen or jammed. . o The snubbers are tested for various (iv) Adequate swing clearance is ! I abnormal environmental conditions, provided to allow snubber ! 'Upon completion of the abnormal movements. ; cnvironmental transient test, the
~
snubber is tested' dynamically at : -(v) If applicable, fluid is to be i frequency. within a specified recommended level and not be 1 l frequency range. The snubber must ., leaking from the snubber ! operate normally during the dynamic system, 3
- test. ;
(vi) : Structural connections such ' (d) Snubber Installation Requirements as pins, fasteners and other
- 1 connecting hardware such 9 l An installation instruction manual is lock nuts, tabs, wire, cotter; required by the pipe support design pins are installed correctly.
! specification. This manual is required j to contain instructions for storage, If the period between the handling, erection, and adjustments (if .Initia1 pre s e rvIce - necessary) of snubbers. Each snubber examination and initial has an installation location drawing system pre operational tests which contains the installation location exceeds 6 months because of-of the snubber on 'the pipe and unexpect e'd sit u ations,. structure, the hot and cold settings, - reexamination of Items 1,4,- and additional information needed to and 5 will be performed. install the particular snubber. Snubbers which are installed incorrectly or otherwise fail. (e) Snubber Pre service Framination to meet the aboye I requirements will be repaired The pre service examination plan of all or replaced and re-examined' snubbers covered by the Chapter 16 tech- in accordance with the above nical specifications will be prepared. criteria.
- This examination will be made after snubber installation but not more than 6 (4) Struts The design load on struts i months prior to initial system pre oper - includes those loads caused by. dead i ational testing. The pre service weight, thermal expansion, seismic forces l examination will verify the following: (i.e., OBE and SSE), other RBV loads,'
- 1 l Amendment 7 3.9-33 i
l
,_u:. _
-.mm -
ABM seandmed Pinne isasi ors nev. n
-Structures, systems, and components are (1) Appropriate consideration of the most severe [
identified in Section 3.2 with respect to their of the natural phenomena that have been his. j location, service and their relationship to the torically reported for the' site and sur. safety.related or non safety related function to rounding area, with sufficient margin for be performed. Recognired codes and standards are - the limited accuracy, quantity, and period applied to the eqdipment per the safety of time in which the historical data have classifications to assure meetlag the required been accumulated; i safety related function. ; (2) Appropriate combinations of the effects of Documents are maintained which demonstrate normal and accident conditions with the i that all the requirements of the quality effects of the natural phenomena; and j assurance program are being satisfied. This documentation shows that appropriate codes, (3) The importance of the safety functions to be standards, and regulatory requirements are performed. observed, specified materials are used, correct procedures are utilized, qualified personnel are 3.1.2.1.2.2 Evaluation Against Critadon 2 provided, and the finished parts and components meet the applicable specifications for safe and Since the ABWR design is designated as a ! reliable operation. These records are available standard plant, the design bases for j so that any-desired item of information is safety related (See Subsection 3.1.2.1.1.2) ' retrievable for reference. These records will be structures, systems, and components, cannot 1 maintained during the life of the operating accurately reflect the most severe of the natu. ; licenses. ral phenomena that have been historically re. ported for each possible site and their sur. The detailed quality assurance program is in rounding areas.- However, the envelope of site-conformance with the requirements of Criterion 1. related parameters which blanket the majority of - potential sites in the conterminous United For further discussion, see the following States is defined in Chapter 2. The design sections: bases for these structures, systems, and compo. nents reflect this envelope of natural phenomena Chapter / including appropriate combinations of the ef.- ; Esslinn lille fects of normal and accident conditions with this envelope. The design bases meet the _j (1) 1.2 GeneralPlant Description - requirements _of Criterion 2. (2) 3.2 Classification of Structures, Detailed discussion of the various phenomena - Components, and Systems considered and design criteria developed are presented in the following sections: 3.1.2.1.2 Criterion 2 Design Bases for Protection Against Natural Phenomena Chapter / Sc&liQn lillt 3J.2.1.2.1 Criterion 2 Statement (1) 2/3 Summary of Site Characteristics Structures, systems, and components important to safety shall be designed to withstand the (2) 3.2 Classification of Structures, effects of natural phenomena such as earthquakes, Components, and Systems tornadoes, hurricanes, floods, tsunami, and , seiches without loss of capability to perform (3) 3.3 Wind and Tornado Loadings their safety functions. The design bases for these structure systems and components shall (4) 3.4 Water level (Flood) Design reflect: O Amendment 7 3.i-2 l
23A6100AE Standmed Plant arv. n - together with the pump meet the folleving test ! conditions: (1) simulate actual mounting conditions; I (2) simulate all static and dynamic loadings on the pump; (3) monitor pump operability during testing; (4) the normal oper6 tion of the pump during and after the test Indicates that the supports are adequate (any deflection or deformation of the pump supports which precludes the operability of the pump is not accepted); and (5) supports are inspected for structuralin-tegrity after the test, Any cracking or i permanent deformation is not accepted, Dynamic qualification of component supports by analysis is generally accomplished as fol. lows: ; i l O i i i i i d Amendment 11 3.9 34.1
ABM . . siasi. ass-atendard Plant arv a (1) Stresses at all support elements and parts
,0- such as pump holddown and baseplate holddown bolts,' pump support pads, pump pedestal, and -
foundation are checked to be within the al. Iowable limits as spoeified in the ASME Code Section III, Subsection NF. (2) For normal and upset conditions, the deflections and deformations of the supports are assured to be within the elastic limits, '; and to not exceed the values permitted by the designer based on design verification tests. This ensures the operability of the pump. ! 1 (3) For emergency and faulted plant conditions, 'I the deformations do not exceed the values permitted by the designer to ensure the E operability of the pump.- Elastic / plastic k j analysis are performed if the deflections are y above the clastic limits. j . 3AJ.5 Other ASME Ill Component Supports The ASME Ill component supports and their at. 4 tachments (other than those discussed in preced. l ing subsection) are designed in accordance with 1 Subsection NF of the ASME Code Section 111' up to O si$ the interface with the building structure. The building structure component supports are de. i i signed in accordance with the AISC Specification for the Design, Fabrication, and Erection of !
- Structural Steel for Buildings.-. The loading
- combinations for the various operating conditions 3.9.4 Control Rod Drive System (CRDS) correspond to those used to design the supported component. The component loading combinations A control rod drive system CRDS)in an ABWR are discussed in Subsection 3.9.3.1. Active plant is equipped with an electro. hydraulic fine component supports are discussed in Subsection motion control rod drive (FMCRD)~ system, which 3.9.3.2. The stress limits are per ASME III, includes the control rod drive (CRD) mechanism, Subsection NF and Appendix F. The supports are the hydraulic control unit (HCU), the condensate evaluated for buckling in accordance with ASME supply system,'and power for FMCRD motor, and III. extends inside RFV to the coupling interface with the control rod blades.
' Augmented by the following: (1) application of l: Code Case N 476, Supplement 89.1 which governs the 3.9A.1 Descriptive Information on CRDS-
!= design of single angle members of ASME Class 1,2,3
- and MC linear component supports; and (2) when Descriptive information on the CRDs as well I l eccentric loads or other torsional loads are not as the entire control and drive system is con- 3 accommodated by designing the load to act through tained in Section 4.6. !
the shear center or meet " Standard for Steel Support Design *, analyses will be performed in 3.9A.2 Applicable CRDS Design Specificatloa , accordance with torsional analysis methods such I as: " Torsional Analysis of Steel Members, USS CRDS is designed to meet the functional de-l Steel Manual", Fublication T114 2/83 or " Design of. sign criteria outlined in Section 4.6 and con. t Weld Structures
- by Omar W. Blodgett, etc.
I Asnemement 11 3.9 35 i i I
ABM 33461oo4e Senna.ed Plant aw n I alsas of the followirg: (2) factory quality controltests; (1) Ane motion control rod drive; (3) Pive year maintenance life tests; i (2) hydraulic control unit; (4) 1.5X designlife testa; j (3) hydraulic power supply (pumps); (5) operationaltests; (4) electric power supply (for PMCRD motors) (6) acceptancetests;and (5) interconnectingpip*mg; (7) surveillancetests. (6) Sow and pressure and isolation valves; and All of the tests except (3) and (4) ere dis. cussed in Section 4.6. . A discussion of tests ; (7) instrumentation and electrical controls. (3) and (4) follows: Those components of the CRDS forming part of (3) Five Year Maintenance Life Tests . Four the primary pressure boundary are designed control rod drives are normally picked at according to ASME Code Section III, Class 1 random from the production stock each year requirements, and subjected to various tests under simu. i lated reactor' conditions and 1/6th of the The quality group classification of the service life cycles. components of the CRDS is outlined in Table 3.21 ) - and they are designed to the codes and standards, Upon completion of the test program, per Table 3.2 2, in accordance with their control rod drives must meet or surpass individual quality groups, the minimum specified performance requirements. Pertinent aspects of the design and qualifica. _ tion of the CRDS components are discussed in the (4) 1.5X Design Life Tests . When a signifi. - following locations: transients in Subsection Icant design change is made to the com-3.9.1.1, faulted conditions in Subsection ponents of the drive, the drive is sub-3.9.1.4, seismic testing in Subsection 3.9.2.2. jected to a series of tests equivalent to j 1.5 times the service life cycles. 1 3.9.4.3 Design Loads, Stress Limits, and Allowable Deformations 3.9.5 ReactorPressureVesselinternals The ASME III Code components of the CRDS have This subsection identifies and discusses the been evaluated analytically and the design load. structural and functionalintegrity of the major ; ing conditions, and stress criteria are as given reactor pressure vessel (RPV) internals, includ. in Tables 3.91 and 3.9 2, and the calculated ing core support structures. stresses meet the Code allowable stresses. For the non Code components, the ASME 111 Code re. 3.9.5.1 Design Arrangements quirements are used as guidelines and experimen-tal testing is used to determine the CRD perfor. The core support structures and reactor mance under all possible conditions as described vessel internals (exclusive of fuel, control in Subsection 3.9.4.4. rods, and incore nuclear instrumentation) are: 3.9AA CRD Performance Assurance Program (1) Core Support Structures The CRD test program consists of these tests: Shroud; (1) development tests; Shroud support (including the internal pump deck)- Amendment 3 3.9 36 i
23A6100AE Ef andard Plant nev. n -l driven from underneath by a pump shaft, with the 3.9J.1.2.4 Feedwater Sparsers O -impeller being encircled by a diffuser shroud assembled into the pump deck opening. These are Safety Class 2 components. They l are discussed here to describe coolant flow I The RM-section of the RIP is located paths la the vessel and their safety i underneath, and at the periphery of, the RPV function. Each of two feedwater lines is l bottom head inside a pressure retaining housing connected to three spargers via three RPV f termed the motor casing. The motor casing itself nozzles. One line is utilized by the RCIC is not part of the RM, but is in:tead a part of system; the other by the RHR shutdown cooling and welded into an RPV nozzle (pump nozzle). The system. During the ECCS mode, the two groups motor casing thus comprises part of the reactor of spargers support diverse type of flooding coolant pressure boundary .nd is a Safety Class 1 of the vessel. The RCIC system side supports component, high pressure flooding and the RHR system side supports low pressure flooding, as required , The principal element of the stretch tube - during the ECCS operation.' . section is a thin walled tube configured as a-hollow bolt fitting around the pump shaft and The feedwater spargers are stainless steel-within the pump nozzle. It has an externallip headers located in the mixing plenum above the - (bolt head) at its upper end and an external downcomer annulus. A separate sparger in two < threaded section at this lower end. The stretch halves is fitted to each feedwater nozzle via tube function is to achieve tight clamping of the a tee and is shaped to conform to the curve of-IP diffuser to the gasketed, internal mount end the vessel wall. The sparger tee inlet is of the RPV pump nozzle, at all extremes of connected to the RPV nozzle safe end by a thermal transients and pump operating conditions, double thermal sleeve arrangement, with all connections made by full penetration welds. 3.93.1.2.3 Steam Dryer Assembly Sparger end brackets are pinned to vessel t' brackets to support the spargers. Feedwater The steam dryer assembly is a non safety class flow enters the center of the spargers and is component. It is discussed here to describe discharged radially inward to mix the cooler coolant flow paths in the vessel. The steam feedwater with the downcomer flow from the dryer removes moisture from the wet steam leaving steam separators and steam dryer before it the steam separators. The extracted moisture contacts the vessel wall. The feed water flows down the dryer vanes to the collecting also serves to condenses steam in the region troughs, then flows through tubes into the above the downcomer annulus and to subcool downcomer annulus. water flowing to the recirculation internal pumps. The steam dryer assembly consists of multiple banks of dryer units mounted on a common 3.9J.1.23 RHR/ECCS Low Pressure Flooder structure which is removable from the reactor Sparsers pressure vessel as an integral unit. The assembly includes the dryer banks, dryer supply These are Safety Class 2 components. The l and discharge ducting, drain collecting trough, design features of these two spargers of the drain piping, and a skirt which forms a water RHR shutdown cooling system are similar to l- seal extending below the separator reference zero' those of the six feedwater spargers, three of elevation. Upward and radial movement of the which belonging to one feedwater line support dryer assembly under the action of blowdown and additionally the same RHR (and ECCS) seismic loads are limited by reactor vessel function. During the ECCS mode, these internal stops which are arranged to permit dif. spargers support low pressure flooding of the ferential expansion growth of the dryer assembly vessel. The feedwater spargers are described , wlth respect to the reactor pressure vessel. The in Subsection 3.9.5.1.2.4. I assembly is arranged for removal from the vessel i as an integral unit on a routine basis. Two lines of RHR shutdown cooling system t enter the reactor vessel through the two diagonally opposite nozzles and connect to the Amendment 1t 3.9-39
ABM ziasioons Standard Plant ny n - spargers. The sparger tee inlet is connected to ~ connection to the steamline is blocked. When the RPV nozzle safe end by a thermal sleeve draining the vessel during abutdown, air enters arrangement with all connections made by full the vessel through the vent. penetration welds. 3.9J.1.2A Core and laternal Pump 3.9J.1.2J ECCS High Pressure Core Flooder DINorential Preasure Lines Spargers and Piping
' These lines comprise the core flow measure-
% The 'core floeder spargers and piping are ment subsystem of the recirculation flow control N Safety Class 2. Tbe spargers and piping are the system (RPCS) and provide two methods of measur-means for directing high pressure ECCS flow to ing the ABWR core flow rates. The core DP lines the upper end of the core during accident (Safety Class 3) and laternal pump DP lines conditions. (non safety class) cater the reactor vessel se-parately through reactor bottom head penetra. Each of two high pressure core flooder (HPCF) tions. Four pairs of the core DP lines enter - system lines enters the reactor vessel through a the head in four quadrants through four penetra-diagonally opposite nozzle in the same manner as tiens and terminate immediately above and below an RHR low pressure flooder line, except that the the core plate to sense the pressure in the re-curved sparger including the connecting tee is gloa outside' the bottom of the fuel assemblics routed around the inside of and is supported by and below the core plate during normal the cylindrical portion of the top guide. A .cperation, flexible coupling is interposed between the _ sparger tee inlet and the sleeved inlet connector Similarly, four pairs of the internal pump DP-inside the nozzle. The two spargers are lines terminate above and below the pump deck supported so as to accommodate thermal expansion, and are used to sense the pressure across the pump during normal pump operation. Each pair is - ~! 3.9.5.1.2.7 RPV Vent and Head Spray Assembly routed concentrically through a penetration and . upward along a shroud support leg in the lower i This is designed as a Safety Class 1 plenum. component. However, only the nozzle portion of the assembly is a reactor coolant pressure 3.9.5.1.2.9 In Core Guide Tubes and boundary, and the assembly function is not a - Stabillaers - safety related operation. The reactor water cleanup return flow to the reactor vessel, via These are Safety Class 3 components. The ; feedwater lines, can be diverted partly to a guide tubes protect the in core instrumentation ~ spray nozzle in the reactor head in preparation from flow of water in the bottom head plenum and for refueling cooldown. The spray maintains provide a means of positioning fixed detectors 3 saturated conditions in the reactor vessel head in the core as well as a path for insertion and 1 volume by condensing stream being generated by withdrawal of the calibration monitors (ATIP,- the hot reactor vessel walls and internals. The automated traversing incore probe subsystem). -i head spray subsystem is designed to rapidly The in core flux monitor guide tubes extend from ! cooldown the reactor vessel head flange region . the top of the in core flux monitor housing to for refueling and to allow installation of steam the top of the core plate. The power range de- l ' line plugs before vessel floodup for refueling. tectors for the power range monitoring units and the dry tubes for the startup range neutron The head vent side of the assembly passes monitoring and average power range monitoring steam and noncondensable gases from the reactor (SRNM/APRM) detectors are inserted through the ! head to the steamlines during startup and opera- . guide tubes. ! tion. During shutdown and filling for hydro-testing, steam and noncondensable gases may be Two levels of stainless steel stabilizer vented to the drywell equipment sump while the latticework of clamps, tie bars, and spacers give lateral support and rigidity to the guide G : Amendment 11 3.9 40 ___--_____x-
ABWR msime Standard Plant RIV. B 3.9.7 Interfaces 3.9.7J Audit of Design Specifleation and Design Reports O 3.9.7.1 Reactor latersals Vibration Analysis, Measurement and laspection Program Applicants referencing the ABWR design will make available to the NRC staff Design The first applicaat tefereacing the ABWR Specifications and Design Reports required by design will provide, at the time of application, the ASME Code for vessels, pumps, valves and the results of the vibration assessment program piping systems for the purposc of audit. for the ABWR prototype internals. These results will include the followleg information toccified 3.9.8 References in Regulatory Guide 1.20.
- 1. Bil'R Fuel Channel Mechanical Design and R. G.1.20 Snhirst Deflection, NEDE 21354 P September 1976.
C.2.1 Vibration Analysis Program 2. BilR/6 A4el Assembly Evaluation of Combined C 2.2 Vibration Measurement Program Safe Shutdown Earthquake (SSE) and C.13 laspection Program Loss of Coolant Accident (LOCA) Loadings, C.2.4 Documentation of Results NEDE 21175.P. November 1976. NRC review and approval of the above 3. NEDE 24057 P (Class 111) and NEDE.24057 Information on the first applicants docket will (Class I) Assessment of Reactor Internals. complet*. ihe vibration assessment program Vibration in BWR/4 and BWR/5 Plants, requirenanti ior prototype reactor internals. Nowmber 1977. Also NF.DO 24057 P. Amendment 1, December 1978, and NEDE 2.P 24057 10 iWdition to the information tabulated Amendment 2 June 1979. above, the first applicant referencing the ABWR design will provide the information on the 4. General Eleuric Company, Analytical Model schedules in accordance with the applicable for Loss of Coolant Analysis in Accordance portions of position C.3 of Regulatory Guide 1.20 with 10CFR$0, Appendix K, NEDE.20566P, for non prototype laternals. ?roprietary Document, November 1975. Subsequent applicants need only psovide the 5. B11'R Feedwater Nozzle and Control Rod Drive information on the schedules in accordance with Retum Line Nozzle Cracking, NUREG 0619. the applicable portions of position C.3 of Regulatory Guide 1.20 for i.on prototype 6. General Electric En vironmental internals. (See Subsection 3.9.2.4 for interface Qualification Program, NEDE 243261 P, requirements). i'roprietary Docume.'.ts January 1983. 3.9.7.2 ASME Class 2 or 3 er Quality Group 7. Functional Capability Criteria for Components with 60 Year Design Life Essenflal Mark Il Piping, NEDO 21985, September 1978, prepared by Battelle Applicants referencing the ABWR design will Columbus Laboratories for General Electric i identify ASME Class 2 er 3 or Quality Group D Company. 3 components that are ejected to loadings which could result 10 thermal or dynamic fatigue and 8. Generic Criteria for High Frequency Cutoff provide the analyses required by the ASME Code, of Bi4R Equipment, NEDO 25250 Proprietary ! Subsectics NB. These analyses will include the Document, January 1980, appropriate operating vibration loads and for the effects of mixing hot and cold fluids. (See Subsection 3.9.3.1 for interface requirements). iO Amendment 11 3 9-43 a
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M~N pi... 214610QAE m. Table 3.9 7 FATIGUE LIMIT FOR SAFETY CIASS REACTOR INTERNAL STRUCTURES ONLY Su=mme6 of fatipe damage usage fouowing Minor bypotheses(8): - Limit for Service kvels A&B (Normal pumulathe Damage in Patipe and Upset Canditlans) l l Design fatigue cycle usage from analysis 11.0 ' using the method of the ASME Code g (1) Miner, M.A., Cumulative Damage in Fatigue, Journal of Annlled Mechanica. Vol. , 12, ASME, Vol. 67, pp A159 A164, September 1945. O 1 1 4 r O ' Amendment 1 3.9-58 f
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j. Toble 3.94 4 REACIOR COOLANT SYSTEM PRESSURE ISOLNI10N VALVES I
, STANDBY UQUID CONTROL BYSTDI 1
- C417006 A,B leiection m )
C41 P008 laboard Check Valw g RESIDUALIEEAT REMOVAL BY5TDI E1;J005 A,B,C lajectica Vahe14 ops A,B&C i E11 P006 A,B.C Testable Check Valw A,5&C E11 P010 A,B,C Shutdowa Cooling Inboard Sucalon Isolation Valw14opsA,B&C E11 P011 A,B,C Shutdown Cooling Outboard Suction leolation . Valw loops A,BAC l e HIGH PRESSURE CORE FLDODER SYSTEM N E22 P003 B,C lajection Vaheloops B&C E22 P004 B,C Testable Check Valve loops BAC REACTOR CORE ISOLATION COOLING SYSTEM i E51 F004 Injecti n Valve E51 F005 Testable Check Valve ! i
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AB%R - Standard Plant _ ^$8 G . l PO!NT P AT INT [R$tCT10N POW (R (AW CLRV(. , OF TH( CURV[ CPYU $Y $ K(" CONN!CTING P0lNT$ Y AND U 1 ELA$11C LINE OP~ $ AND $ FROM ASME 111, STR4SS - 8 ' l AEPINDICf$!.t.2AND13.2 R(SPttilytLY 8v . .- ._. __. 'U,,=*** I l 1 i l i i I 8, . _ . . O 8, I
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ABM swioarB m- = 1 Standard Plan
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O sie r veie.norizetionsi a ror=/> aaccv i X-AXIS (RB 0180DEG) ; ELEY .gcrOR BUIDING RCCV i , 30 SHEAR (TON 11 MOMENT (T M) EHEAR (TONS) MOMENT (T M) 44.7 1.93E+ 04 3.95E + 03 33.2 747E+ 04 j 8.29E + 03 26.7 ~1.46E + 05 5.44E+ 04 . 135E+ 04 $.28E + 03 . 18.5 2.73E + 05 8.70E + 04 1.56E + 04 1.47E + 04 13.1 3.66E + 05 1.08E+ 05 : 2.10E + 04 1.63E+04 7.3 4.84E+ 05 1.78E + 05 2.10E + 04 1.63E + 04 41.2 4.84E + 05 2.13E + 05 2.10E+ 04 1.63E + 04
-6.7 4.84E + 05 232E+05 2.10E + 04 1.63E+ 04 13.2 5.64E + 05 2.95E+ 05 ;
- Y-AXIS (RB 90 270 DEG) l NJ ELEV REAcrOR BUIDING RCCV ,
Qd} MR (TONS) MOMENT (T M) ' SHEAR (TONS) MOMENT (T M) 44.7 9.53E + 03 3.93E+ 03 33.2 9.11E + 04 7.55E+ 03 i 26.7 1.79E+ 05 4.44E+ 04 l 1.28E + 04 3.11E+ 03 18.5 2.94E + 05 1.12E+ 05 1.72E+ 04 7.80E+ 03 13.1 3.79E + 05 1.82E+ 05 2.11E+ 04 9.73E+ 03 l 73 4.82E + 05 2.61E + 05 l 2.11E+ 04 9.73E+ 03
-0.2 5.09E+ 05 3.06E+ 05 2.23E+ 04 1.04E + 04 6.7 5.62E + 05 332E+05 239E+ 04 1.11E + 04 13.2 637E+ 05 3.51E+ 05 ,
~
l Notes: l 1. Elevations are rel ative to the RPV bottom head.
- 2. Forces on the RO between EL 33.2M and 18.5M along the x. axis are the sum of maximum p forces of the tem s.icks representing the walls as shown in Fig. 3G.21.
1 I Anwndment 4 30.42
. _ . - , _.-. _ , , ._. . _ _ _ _ _ _ . . _ , . _ _ . _ _ , 2.
e 33A6100AE M at a, . Table 3G.44 Site Envelope OBE leads for fielected Locations g SPAM MAYIMUM FORCIE MAXIMUM MOMENT IDCATION _ELT. E M M Mxer.M) m Shroud Sup's 28 98.0 263.8 286.0 1830.0 2281.0 RPV Skirt 69 577.0 780.3 614.0 $322.1 3909.9 RSW Base 78 458.0 1044.0 853.0 5064.0 4168.0 Podestal Base 86 2027.0 3343.0 2806.0 72077.0 $8671.0 Notes:
- 1. P is verticalload due to vertical excitation
- 2. VX and MX are shear and moment due to HOR X excitation
- 3. VY and MY are shear and moment due to HOR Y excitation O'
a O Amendment 11 30.4 3 1
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_ _ _ _ _ _ _ ~
ABWR m- , Standard Plant am e i cessive oxidation, hydriding, or crud deposition sure during operation and maintenance of the may lead to a breach of the cladding wall. plant components. Water quality parameters can have an influ. ; Metallic impurities can result in neutron losses and associated economic penalties which in. ence on radiation buildup rates. In laboratory l crease in proportion to the amount being intro- tests, the water conductivity and pH were varied duced into the reactor and deposited on the systematically from a high purity base case. In I fuel. With respect to iron oxide type crud depo. each case, impurities increased the rate of , sits,it can be concluded that operation within cobalt 60 uptake over that of the base case. l l the BWR water chemistry guidelines (specifically The evidence suggests that these impurities the limits on feedwater iron levels) effectively change both the corrosion rate and the oxide precludes th3 buildup of significant deposits on film characteristics to adversely increase the fuel elements, cobalt 60 uptake. Thus, controlling water purity should be beneficial in reducing ra. ] 5.23.2.2.2 Radiation Field buildup diation buildup. l The pr! mary long term source of radiation Prefilming of stainless steel in cobalt.60 fields in most BWRs is cobalt.60, which is formed free water, steam, or water /stete mixtures also ! I by neutron activation of cobalt.59. Corrosion appears to be a promising method to reduce ini. products are released from corroding and wearing tial radiation buildup rates. As an example, i surfaces as soluble, colloidal, and particulate the radiation buildup rates are reduced sig. ! species. The formation of cobalt.60 takes place nificantly when sapples are prefilmed in high after the corrosion products precipitate, adsorb, temperature (288 C), oxygenated (200 ppb or deposit on the fuel rods. Subsequent reen. oxygen) water prior to exposure to cobalt 60 trainment in the coolant and deposition on out. containing water. Mechanical polishing and of. ore stainless steel surfaces leads to buildup electropolishing of piping internal faces should of the activated corrosion products (such as co. also be effective in reducing radiation buildup. f' balt.60) on the out of. core surfaces, The depo. sition may occur either in a loosely adherent 5.2.3.2.23 Sources of impurities layer created by particle deposition, or in a ' tightly adherent corrosion layer incorporating Various pathways exist for impurity Irigress radioisotopes during corrosion and subsequent ion to the primary system. The most common sources exchange. Water chemistry influences all of of impurities that result in increases in reac. . these transport processes. The key variables are for water conductivity are condenser cooling the concentration of soluble cobalt.60 in the re. water inleakage, improper operation of lon ex. actor water and the characteristics of surface change units, alt intenkage, and radwaste re. oxides. Thus, any reduction in the soluble co. cycle. .n addition to situations of relatively balt 60 concentration will have positive continuous ingress, such as from low level con. , benefits, denser cooling water inleakage, transient events can also be significant. The major sources of As a means to reduce cobalt, GE has reduced impurities during such events are resin intru.
- cobalt content in alloys to be used in high slons, organic chemical intrusions, inorganic fluence areas such as fuel assemblies and control chemicalintrusions, and improper rinse of re.
{s rods, in addition, cobalt base alloys used for sins. Chemistry transients resulting from intro. ( pins and rollers in control rods have been duction of organic substances into the radwaste replaced with noncobalt alloys, system comprised a significant fraction of the transients which have occurred. The reactor water cleanup system, which pro.
= cesses reactor water at a rate of 2% of rated The condensate cleanup system has two stages feedwater flow, will remove both dissolved and of water treatment. The first stage, the hollow
{3 undissolved impurities which can become radioac. fiber filters, is eifective in removing e O M g tive deposits. Reduction of these radioactive insoluble solids, such as condensate system deposits will reduce occupational radiation expo. insoluble corrosion products. The second stage, O v $.2 9 Amendment 11
MalM meandard Plant aw e the deer bed demineralizers, is effectivo in removing soluble solids, such as soluble corrosion products and impurities from possible condenser ler.kage. The following factors are measured for control or diagnostic purposes to maintain proper water chemistry la the ABWR. O I
'I I
J Amendment 11 O 5.29.1
susa:an A..B.a..a Mpi... gy; ; (1) GoedW1bity orygen control cannot be achieved through traditional chemistry and operational prac-t incresslag levels of many lonic impurities tices. Oxygen control to low, plant spe-adversely influence both the stress corro- cific levels can be obtained through hydro-alon cracking behavior of RCS materials, the gen injection. Control of reactor water ' rate of radiation field buildup and also can oxygen during startup/ bot standby may be < affect fuel performance. Therefore, conduc. accomplished by utilities the de aeration tivity levels la the reactor water should be capabilities of the condenser. Independent maintained at the lowest levels practically control of control rod drive (CRD) cooling achievable, water oxygen concentration of <$0 ppb during , power operation is desirable to protect
- (2) Chlodds against IOSCC of CRD materials. Carbon j steels exhibit minimal ge* trst corrosion and i
- Chlorides are among the most potent promo- release rates in water with a conductivity
l ters of IGSCC of sensitised stainless steels less than 0.1pS/cm if the concentration of and are also espable of laducing transgran- oxygen is in the range of 20 to 1000 ppb. ular cracking of nonsensitized stainless Regulation of reactor feedwater dissolved steels. Chlorides also promote pitting and oxygen to 20 to 50 ppb during power op. , crevice attack of most RCS materials. Chlo- eration will minimize corrosion of the 3 rides normally are associated with cooling condensate and feedwater system and reduce water inleakage, but inputs via radwaste the possibility of locally increasing processing systems have also occurred, reactor water oxygen concentrations, it is important to note that for oxygen concentra. Because chloride is implicated in several !!ons below 20 ppb, the data indicates an l different corrosion phenomena,its levelin increase in the corrosion and corrosion prod- t reactor water should be kept as low as uct release for carbon steels. practically achievable during power operation. (5) kan , (3) Sulfais High iron inputs into the reactor have been associated with excessive fuel deposit Recently, sulfate has been found to be more buildup. Proper regulation of feedwater sagressive in promoting IGSCC of sensitized purity and dissolved oxygen levels will
' ype 304 stainless steel in BWR type water scinimize Iron transport to the reactor.
(in laboratory tests) than any other lon, This,in turn, should minimize fuel deposits including chloride. Sulfates have also been and may assist in controlling radiation implicated in environment assisted cracking buildup. of high. nickel alloys and carbon and low-alloy steels. Sulfate ingress can result (6) Fluoride i from cooling water inleakage, regenerant chemical inleakage, or resin ingress. Fluoride promotes many of the same corrosion phenomena as chloride, including IGSCC of (4) Qugtn sensitired austenitic stainless steels, and may a. 've the potential to cause corro- i Dissolved oxygen has been Identified as a slon of Zircaloy core components, major contributor to IGSCC of sensitized stainless steels and reduction of oxygen (7) Oreanics content is known to reduce the tendency for pitting and cracks of most plant materlats. Organic compounds can be introduced into the RCS via turbine or pump oilleakage, rad. During power operation, most of the oxygen waste, or makeup water systems. Of par-content of reactor water is due to the radi- ticular concern is the possibility that olysis of water in the core and, therefore,- halogenated organic compounds (e.g., Amendment 3 $.210 l
- _ _ _ _ _ _ _ _ _ __ .1. __
ABM zusmin meandad Plant me eleaning solvents) may pass through the (12) Rectreulation System Water Dissolved radwaste systems and enter the RCS, where Hydtagga they will decompose,' releasing corrosive halogens, e.g., chlorides and fluorides. A direct measurement of the dissolved hydro-gen content in the reactor water serves as a (8) gilisa cross check against the hydrogen gas flow meter in the injection system to confirm the
$ltica, an indicator of general system clean- actual preseace and magnitude of the liness, provides a valuable ladication of hydrogen addition rate, the effectiveness of the reactor water clean.
sp system. Silica inputs are usually assocl- (13) Main Steam line Radiation 12 vel sted with incomplete silica removal in make-up water or radwaste facilities. The major activity in the main steam line is nitrogen.16 produced by an (n, p) reaction (9) g g with oxygen 16 in the reactor water. Under conditions of hydrogen wat:.r chemistry, the There are difficulties of measuring pH la fraction of the nitrogen.16 that volatilizes low conductivity water. Nevertheless, p11 ef with the steam increases with increased dis-
, the liquid environment has been demonstrated solved hydrogen. The main steam line radia.
to have an important influence on IGSCC ini. tion monitor readings increase with the hy. g tlation times for smooth staluless steel drogen addition rate. During Initial plant R specimens in laboratory tests. in addition, testing, the amount of hydrogen addition pH can serve as a useful diagnostic parame- required to reduce the electrochemical cor. ter for laterpreting severe water chemistry rosion potential to the desired range is transients and pH measureraents are recom. determined at various power levels. Chan-mended for this purpose, ges in the main steam line radiation mont. tor readings at the same power level Indl. 1 69) Ft~emA=lemi corroston Potential cate an over addition (high readings) or under. addition (low readings) of hydrogen. The electrochemict.1 corrosion potential i (ECP) of a metalis the potential it attains (14) Constant Extension Rate Test i when immersed in a water etwironment. The i ECP is controlled by various oxidizing Constant extension rate tests (CERTs) are agents including copper and radiolysis pro- accelerated tests that can be completed in a ducts. At low reactor water conductivities, few days, for the determination of the sus-the ECP of st Inless steel should be below ceptibility to IOSCC. It is usefe for 0.23 V to suppress IOSCC, verifyir"t IGSCC suppression during initial SHE imple.wntation of hydrogen water chemistry (11) Feedwater Hydronen Addition Rate (HWC) or following plant outages that could have had an impact on system chemistry A direct measurement of the feedwater hydro- (e.g., condenser repalts during refueling), gen addition rate can be made using the by. drogen t.ddition system flow measurement de. (15) Continuous Crack Growth Monitorinn Test vice and is used to establish the plant spe. , cific lydrogen flow requirements required to This test employs a reversing DC potential satisfy the limit for the ECP of stainless drop technique to detect changes in crack steel (Paragraph 10). Subsequently, the ad. length in IGSCC test specimens. The crack dition rate measurements can be used to help growth test can be used for a vsriety of diagnose the origin of unexpected ECP purposes, including the following: changes. ' (a) Initial verification of IGSCC suppres. O Amendment 3 $.211
"ABWR **"
Reandard Plant arv c aion following HWCimplementation. code will be used to identify areas where design improvements (piping dexign, materials , (b) Quantitative assessment of water chemis. selection, hydrodynamic conditions, oxygen try transients, content, temperature) are required to ensure adequate margin for extended piping performance (c) Long term quantification of the sucess on the ABWR design. of the HWC propam. The major impurities In various parts of a BWR under certain operating conditions are listed in Table 5.2 5. The plant systems have been designed to achieve these limits at least 90% of the time. The plant operators are encouraged to achieve better water quality by using good operating practice. Water quality specifications require that erosion. corrosion resistant low alloy steels are to be used in susceptible steam extraction and drain lines. Stainless steels are considered for 3 Is baffles, shields, or other areas of severe duty. A* Provisions are made to add nitrogen gas to f extraction steamlines, feedwater heater shells, heater drsin tanks, and drsin piping to minimize corrosion during layup. Alternatively, the system may be designed to drain while hot so that dry layup can be achieved. g* Condenser tubes and tubesheet are required to yj be made of titanium alloys. Erosion. corrosion (E/C) of carbon steel . components will be controlled as follows. The mechanism of E/C or, preferably, flow assisted ) corrosion is complex and involves the electrochemical aspects of general corrosion plus j the effects of mass transfer. Under single phase flow conditions, E/C is affected by water i chemistry, temperature, flow path, material composition and geometry. For wet steam (two -
$ phase), the percent moisture has an additional Fi effect on E/C.
The potential deterioration of ABWR carbon , steel piping from flow assisted corrosion due to high velocity single phase water flow and two phase steam water flow will be addressed by using the EPRI developed CHECMATE (Chexal Horowitz Erosion Corrosion Methodology for Analyzing Two phase Environments) computer code. CHECMATE will be used to predict corrosion rates and calculate the time remaining before reaching a defined acceptable wall thickness. Thus, this Amendment 11 5.211a O
ABWR 2mme : Standarti Plant arv. c SECTION 5.3 . t CONTENTS l Secilo's M Ragt
$3.1 Itanctor Vanaal Matartals Sal 53.1.1 Materials Specifications SS1 i 53.1.2 Special Processes Used for Manufacturing $31 ;
and Fabricatfon 53.13 Special Methods for Nondestructive SS1 [ Examination 53.1.4 Special Controls for Feuitic and 53 1 [ i Austenitic Stainless Steels 53.1.4.1 Regulatory Guide 131: Control of Stainless SS1 ! SteelWelding _, 53.1.4.2 Regulatory Guide 134: Control of $3 2 l Electroslag Weld Properties ! O V 33.1.43 Regulatory Guide 1.43: Controlof Stainless SteelWeld Cladding of SS2 Low Alloy Steel Components , t t 53.1.4.4 . Regulatory Guide 1.44: Control of the . Use of Sensitized Stainless Steel 53-2 ; 53.1.4.5 Regulatory Guide 1.50: Control of ' Preheat Temperature for Welding Low AlloyStcel 53-2 53.1.4.6 Regulatory Guide 1.71: Welder , I Oualification for Area of Limited Accessibility '53 2 53.1.4.7 Regulatory Guide 1.99: Effects of Residual Elements on Predicted Radiation Damage to Reactor Pressure ' VesselMatcrials 53 2 ', 53.1.4.8 Regulatory Guide 137: Quality Assurance Requirements for Cleaning of Fluid Systems and Associatrxl Components of Water Cnoted Nuclear Pow er Plants 53-2 g 53.1.5 Fracture Toughness 53 2.1 53 il : A nendment 11
,w ,. , . - - - - . . . - . - ,_ . . . . . . ..i
=
21461 NAB me..Amrd Plant . me- l 1 SECTION 5.3 CONTENTS (Continued) h l sa:nna m em i 53.1J.1 Compliance with 10CFR50, Appendix G $3 2.1 l 53.1.5.2 Methods of Compliance 512.1 i 53.1.6' MaterialSurveillance SS3 l 53.1.6.1 Compliance with Reactor Vessel Material i Surveillance Program Requirements $3 3 ' 53.1.6.2 Neutron Flux and Fluence Calculations SS4 i 53.1.63 Predictedinadiation Effects on $3 4 Beltline Materials ,
$3.1.6.4 Positioning of Surveillance 53 4 -
Capsules and Methods of Attachment (Appendix H.il C(2)) , 53.1.6.5 Time and Number of Dosimetry Measurements - 53 5 53.1.7 Reactor Vessel Fasteners : 53 5-
$3.1.8 Regulatory Guide 1.65 53 5 53.2 Pressure / Temperature IJmits $3 6 ;
53.2.1 Limit Curves 6 l
- i l
53.2.1.1 Temperature Limits for Holtop 53 6 - 53.2.12 Temperature Limits for 151 . Hydrostatic and 1.eak Pressure Tests $3 6 53.2.13 Operating Limits During Heatup, Cooldown, and Core Operatie. 53-6 ! 53.2.1.4 Reactor Vessel Annealing $3-6
'53.2.1.5 Predicted Shift in RT and Drop in T
Upper. Shelf Energy ppendix G IV B) 53 6 ; y
' Amendment 11
=
. - - , - - . , , e . , - - , - , -e-,,.- c - , - - - - - - -
r- -'ld -
2sA6100AB ReanmAaM Plant aw a 8.3 REACIDRVESSEL these vessel compos:sts is la accordance with procedures qualified per ASME Section !!! and 5J.1 ReactorVesselMaterials IX requirements. Weld test aamples are requir. ed for cach procedure for major vessel full.
$J.1.1 Materials Speenacetions penetration welds. Tensile and impact tests are performed to determine the properties of Tbs materials used la the reactor pressure the base metel, heat affected zone, and weld vessel and appurtenances are shown la Table 5.2 4 metal together with the applicable arebiam Submerged are and manual stick electrode The RPV materials shall comply with the welding processes are employed. Electrostag a
provisions of the ASME Code Section !!!, Appendix weldlag is not applied. Preheat and laterpass N I and meet the speelfication requirements of temperatures employed for welding of law. alloy 10CFT50, Appendix G. stee! meet or exceed the values given in ASME, Section !!!, Appendix D. Post weld heat treat. 53.1.2 Sportal Procedures Used for Manutseter. mest at 11000F minimum is applied to all lag and Fabrication low. alloy steelwelds. The reactor pressure vesselis primarily con. Radiographic examination is performed on all structed from low alloy, high strength steel plate pressure containing welds in accordance with re. and forgings. Plates are ordered to ASME SA 533, quirements of ASME, Section 111, Subsection NB TYPE B, Class 1, and forgings to ASME SA 508, 5320. In addition, all welds are given a Class 3. These materials are melted to fine grain supplemental ultrasonic examination. pristice and are supplied in the penched and tem. pered condition. Further restrictions include a The materials, fabrication procedures, and requirement for vacuum degassing to lower the by. testing methods used in the construction of BWR drogen level and improve the cleanliness of the reactor pressure vessels meet or exceed require. O Iow alloy steels. Materials used in the core ments of ASME Section !!!, Class 1 vessels. beltline region also specify limits of 0.05% max. Imum copper and 0.015% :taximum phosphorous con. 5.3.1.3 Special Methods for Nondestructive tent in the base materials and a 0.08% maximum Examlastlos l copper and 0.070% maximum phosphorous content in ' weld materials. The materials and welds on the reactor pres. sure vessel are examined in accordance with Studs, nuts, and washers for the main closure meibods prescribed and meet the acceptance re. flange are ordered to ASME SA.540, Grade B23 or quirements specified by ASME, Section !!!. In Gradc D24. Welding electrodes for low alloy steel addition, the pressure.retalning welds are ut. are low. hydrogen type ordered to ASME SFA 5.5. trasonically enmined using manual techniques. The ultrasonic examination melnod, including All plate, forgings, and botting are 100% ultra. calibration, instrumentation, scanning sensitiv. sonically tested and surface e::amined by magnetic ity, and coverage, is based on the requirements particle methods or liquid penetrant inethods in ac. Imposed by ASME, Section XI, Appendix 1. Accep. l cordance with ASME Section III, Division 1. tance standards are equivalent or more restric. tive than requM by ASME,Section XI. h Frscture toughness properties are also measured and controlled in accordance with Division 1. $J.1A Special Controls for Ferritic and Austealtic Stalaless Steels All fabrication of the teactor presrure vessel , is performed in accordance with GE.spproved draw. 5.3.1.4.1 Regulatory Guide 1.31: Control of ings, fabrication procedures, and test proce. Stainless Steel Welding dures. The shells and vessel heads are made from formed plates or forgings, and the flanges and Controls on stainless steel welding are dis. nozzles from forgings. Welding performed to join cussed in Subsection 5.23.4.2.1. Amendment 2 5.31
ABM i n si m . an e - padard Plaar l
$3.14.2 Regulatory Guide 134: Centrol of oflow hydrogen olectrodes to prevent hydrogeu !
Doctreeles Wold Properties cracklag (provided la subsection 5.23.3A) does : not explicitly meet this requirements the ABWR ! Electroslag welding is not employed for the control will assure that cracking of components : i reactor pressure vessel fabricatloa, made from low alloy stpls does not occur during l l fabrication. . Further, the ASWR eentrol minimi. i l SJ.1 AJ Regulateey guide 1 AS: Centrol of ses the possibility of subsequeat eracking rc. ; Statalens SteelWeld Cineding ef tme. Alley Stest sulting from hydrogen being retained la the Compements weldsent. ! l Reactor pressure vml specifications require All welds are mondestructively esamined by : that all low alloy steel be produced to fine grala radiographic methods. la addition, a supplemen.
+
practice. The requireneats of this regulatory tal sistasoale examination is performed. j guide are not applicable to BWR vessels, 53.144 RegulatoryGolde1.71: Walder l $J.lAA RegulatoryGolde144: Centrolof Quall8caslea ler Areas ofIJmited Acesssibility j the Use of Sensitised Stalstens Steel Qualification for areas of limited accessibil. , Srnsitiration of stainless steel is controlled Ity is discussed in Subsection 5.2.3A.2.3. 4 by ,he use of service proven materials and by use 3 of appropriate design and processing steps in- 53.1A.7 Regulatory Gulde t.99: Easets of [ > .suding solution heat treatment, cortosion resis. Reeldmal Dements ca Predicted Radiatica Damage tant cladding, control of welding heat input, te Reseter Pressero Vessel Materials control of heat treatment during fabrication and control of stresses. Predictions for changes in transition tem. perature and upper shelf energy are made la accor. 83.1A.8 Regulatory Guide 1.80: Centrol of dance with the requirements of Regulatory Guide : Preheat Temperature For Welding law. Alloy Steel 1.99. Regulatory Guide 1.50 delineates preheat tem. 8.3.1AJ Regulatory Guide IJ7: Quality peinture control requirements and welding proce. Atsursace Requirements for Cleaning of Fluid dure qualifications supplementing those in ASME Systems and Asseclated Components of Water. Sections Ill and IX. Conled Nuclear Power Plants The use of low. alloy steel is restricted to the The cleaning of systems and components on the reactor pressure vessel. Other ferritic com. site during and at the completion of construction ponents in the reactor coolant pressure boundary is accomplished to written procedures which assure , i are fabricated from carbon steel materials, both cleanliness and that the components are not l exposed to materials or practices which will , l Preheat temperature employed for welding oflow degrade their performance. For components alloy tteel meet or exceed the recommendations containing stainless steel the procedures will of ASME Code Section III, Appendix D. Components comply with Regulatory Guide IJ7. De procedures 1 0 are either held for an extended time at preheat will prohibit contact with low melting point A temperature to assure removal of hydrogen, or compmds, substances which are known to cause prebest is maintained until post. weld heat stress corrosion cracking or which can relc.ase in treatment. The minimum preheat and maximum any manner substances that can cause such l Interpass temperatures are specified and problems. In addition there are controls placed monitored. on the use of grinding wheels and wire brushes that assure that they cannot introduce degrading Acceptance Criterion II.3.b(1)(a) of SRP materials either through prior usage or through Section 5.2.3 for control of preheat temperature their materiais of construction. In this context requires that minimum and maximum laterpass degradatW includes stress corrosion cracking. temperature be specified. While the ABWP control Controls also controlintroduction of unnecessary 4 Amendment 11 $.3-2 l l
.i i zwima l A.BM...a
. pi... uy. c dirt and require control of dirt producing
/"
processes such as welding or grinding ine'ading i prompt cleamlag. ; SJJJ FractureToughness SJJJ.1 Complianer with 10CFR50, Appendia G f i Appendix 0 of 10CFR$0 is interpreted for Class 1 primary coolant pressure boundary component of the ABWR reactor design and complied with as dis- - l cucsed in Subsections 5.3.1.5.2 and 5.3.2. The specific temperature limits operattom of the i reactor when the core is critical are based on 10CFR50, Append!r G, Paragraph IV, A.3 See ; Subsection 5.3.4.1 for fracture toughness data !
, -laterface requirements.
5J.1J.2 Methods of Compliance l The following items are the laterpretations and methods used to comply with 10CFA50, Appendix G. (1) Material Test Coupons and Test Specimens , (Gill A) Test coupons are removed from the location in each product form as specified ' I I e Y 1 I 1 v Amendment !! SS2.1 t .. . -
t A.B
. .a .M .m . .. . imai ma mc Preparation of impact testlag procedures, safety to those required for shells and O calibration of test equipment, and the re.
testion of the records of these functions and test data comply with the r( p'? heads are demonstrated uslag a 1/4 T postu. lated defect at alllocations, with the ex. ception of the mais closure flange to the meats of the ASME Code, Section III ?3t head and shell discostlaulty locations. Ad-sonnel conductlag impact testing ,,cc ditional lastruction on operating !!alts is quallfled by experience, training or required for.outside surface flaw stres qualification tettlag that demonstrates greater than 0.24 lach at the outside competence to perform tents in accordance surface of the flange to shell joint based with the testing procedure, on analysis made for ABWR reactor vessels using the calculations methods shown in WRCB (4) Charpy V Curves for the RPV Beltline 175. It has been demonstrated, uslag a test (0 IIIA and G.IVA 1) mockup of these areas, that smaller defects can be detected by the ultrasoalc laservice A full transverse Charpy V curve is deter- exami melons procedures required at the aga. mined for all heats of base material and cent weld joint. $ lace the stress latensity weld metal used in the core beltline factor is greate" at the outside surface of region with a minimum of three (3) the flange to.s. ell and head joints, a flaw specimen tested at the actual TNDT. can also bc detected by outside surface ex. The minimum upper shelf energy level for amination techniques, base material and weld metal in the beltline region is 75 ft lbs as required (7) Fracture Toughness Margins in the Control by G.IVA.1. of Reactivity (Appendix G IV A). In regard to G lil A, it is understood ASME Code, Section Ill, Appendix G, was used that separate, unitradiated baseline in determining pressure / temperature limita-0l specimens per ASTM E 185, Paragraph 6.Al will be used to determine the transition temperature curve of the core beltline base material, HAZ and weld metal. tions for all phases of plan operation, 8.3.1.6 !?.aterialSurveillance - 8.3.1.6.1 Compliance with Reactor Vessel (5) BoltingMaterial Material Surveillance Program Requimments All bolting material exceeding one inch The materials surveillance program monitors diameter has a minimum of 45 ft lbs changes in the fracture toughness properties of J Charpy V energy and 25 mils lateral expan. ferritic materials in ihe reactor vessel slon at the minimum bolt preload tem- beltline region resulting from exposure to perature of 70 0F, neutron irradiation and thermal environment. . (6) Alternative Procedures for the Calcula. Reactor vessel materials surveillance tion of Stress latensity Factor (Appendix specimens are provided in accordance with re. 0 IV A) quirements of ASTM E 185 and 10CRF50, Appendix [ H. Materials for the program are selected to rep. Stress intensity factors are calculated resent materials used lu the reactor beltline by the methods of ASME, Section Ill, region. Specimens are manufactured from a plate Appendh G. Discontinuity regions are or forging actually used in the belti_ine region evaluated as shell and head areas, as and a weld typical of those in the beltline part of the detailed thermal and stress region and thus represent base metal, weld mate-analyses in the vessel stress report. rial, and the weld heat affected zone material. Considerations are given to membrane and The plate and weld are heat treated in a manner bending stresses, as outlined in Para- which simulates the actual heat treatment per- i graph G 2222. Equivalent margins of formed on the core region shell plates of the m i
3&As100AB j Etandmed Plant arv.c I completed vessel. Each in. reactor surveillance temperature a'. end.of life is less than 100 'F, capsult contains 36 Charpy V. notch and 6 tensile and the end oflife upper. shelf energy exceeds 50 specimens. The capsule loading consists of 12 ft lb. (See response to Question 251.5 for the l Charpy V Specimens each of base metal, weld metal, calculation and analysis associ. attd with this heat effected rene material, and 3 tensile estimate). 3 specimens each from base metal and weld metal. A i set of out of reactor baseline Charpy V. notch SJ.lM poststening et survelunnes capastaa specimens, tensile specimens, and archive material and Moshoda of A**he (Appendix N.!! B (2)) are provided with the surveillance test specimens. Neutros desleeters and temperature surveillance specimen capsules are located at monitors will be located within the capsules as re. three azimuths at a common elevation in the core quired by ASTM E 185, beltline region. The scaled capsules are not at. tached to the vessel but are is welded capsule Three capsule are provided in accordance with holders. The capsule holders are mechanically re. 1 requirements of 20CpR$0, Appeccix H, since the pre. tained by censule holder brackets welded to the dicted end of the adjusted reference temp,ers. ture vessel clad &g. Sloce reactor vessel spe. effica. of the reactor vessel steel is less than 100 F. tions require that all low. alloy steel pressure j vessel boui,dary materials be produced to j The following proposed withdrawal schedule is fine grain practice, underclad cracking is of no in accorder'ee with ASTM E 185. concern. The capsule holder brackets allow the removal and reinsertion of capsule holders. Al. ! First capsule: After 6 effective full. power though not code parts, these brackets are de. years signed, fabricated, and analyzed to the require. J Second capsule: After 15 effective full power ments of ASME Code Section 111. A positive years spring loaded locking device is provided to re. Third capsule: Schedule determined based on tain the capsules in position throughut any an. l
- results of first two capsules per ASTM E 185, ticipated event during the lifetime of the ,
I paragraph 7.6.2. vessel. See Subsection 5.3.4.2 for interface l requirements pertaining to materials and 1 Fracture toughness testing of irradiated cap. survellance capsules, r sule specimens will be in accordance with require. ; ments of ASTM E 185 as called out for by 10CFR$0, In areas where brackets (such as the survell. lance specimen holder brackets) are located, addi. Appendix H. tional mondestructive examinations are performed 8.3.1.6.2 Neutron Flux and Fluence Calculations on the vessel base metal and stainless steel weld. deposited cladding or weld buildup pads
- A description of the methods of analysis is during vessel manufacture. The base metal is ut.
l contained in Subsections 4.1.4.5 and 4.3.2.8. trasonically esamined by straight. beam techniques to e &pth at least equal to the thickness of the 83.1.6.3 predicted levadiation Effects on bracket being joined. The area examined is the Beltline Materials area of width equal to at least half the thickness of the part joined. The required stainless steel Transition temperature changes and changes in weld. deposited cladding is simi. larly examined. , upper. shelf energy shall be calculated in accor. The full penetration wel6 are liquid penetrant ex. dance with the rules of Regulatory Guide 1.99, amined. Cladding thickness is required to be at Reference temperatures shall be established in ac. least 1/8 inch. These requirements have been suc. cordance with 10CFR50, Appendix 0, and Nm2330 of cessfully applied to a variety of bracket designs the ASME Code. which are attached to weld. deposited s:alaless steel cladding or weld buildups in many operating Since weld material chemistry and fracture BWR reactor pressure vest.els. toughness data are not available at this time, l- the limits in the purchase specification were Inservice inspection examinations of core , used to estimate worst. case irradiation effects. beltline p. essure retaining welds are performed - i from the outside surface of the reactor pressure These estimates show that the adjusted reference Amendment 11 5.34 l
I 3&Asl00AB ! Rf andard Pinar arv. c l vessel. If a bracket for mechanically retalning Code, Section 111, Class I, requirements. The , surveillance specimen capsule holders were r6 actor pressure vessel closure studs ate SA-540 i l located at or adjacent to a vessel shell weld,it Orade B23 or 24 (A1S! 4MO). De maximum allow. would not laterfere with the straight beam or able ultimate tensile strength 6a 170,000 psi. : half. node, angle beam inservice inspection ultra- Also, the Charpy impact test requirements of soalc esaminat ens performed from the outside NB 2333 will he satisfied (the lowest Cy surface of the vessel, energy will be greater than the requirement of 45 ft lb at 70 0F; the lowest reported Cy ex-
$J.1.6J Time and Number of Deelestry Measure- pansloc will exceed the 25 mils required). ;
i monts ! In relationship to regulatory position C.2.b, r GE provides a separate neutron dosimeter so . the bolting materials are ultrasonically exam-that fluence measurements may be made at the ined in accordance with ASME Code Section III, vesselID during the first fuel cycle to verify NB 2580, after final heat treatment and prior to
! the predicted fluence at an early date in plant threading as specified. The requirements for ex- ;
- operation. This measurement is made over this amination according to SA 388 and ASTM A614 were short period to avoid saturation of the dosim- complied with. The procedures approved for eters now available. Once the fluence to thermal in practice are judged to insure cor parable mate- ,
power output is verified, no further dosimetry is rial quality and are considered adepan in the , considered necessary because of the !!near rela. basis of compliance with the applicable require-tionship between fluence and power output. It ments of ASME Code, subsubarticle NB 2580, will be possible, however, to install a new dosim. _ eter,if required, during succeeding fuel cycles. The straight beam examination is performed on 100 percent of cylindrical surfaces and from 5.3.1.7 RanctorVesselFasteners both ends of each stud using a 3/4 inch maximum ; O diameter transducer. Tht. n? : sace standard for The reactor vessel closure head (flange) is the radial scan contains a 1/2 inch diameter fastened to the reactor vessel shell flange by flat bottom hole with a depth of 10 percent of multiple sets of threaded studs and nuts. The the thickness. The end scan standard is per lower end of each stud is installed in a threaded ASTM A614. Surface examinations are performed l bole in the vessel shell flange. A mut and on the studs and nuts after final heat treatment washer are installed on the upper end of each and threaded as specified in the guide, in accor-
- stud. The proper amount of preload can be dance with ASTM A614. Anyindication grer.ter applied to the studs by :.equential tensioning than the indication from the applicable calibr using hydraulic tensioners, tion feature is unacceptable. The distance /ampli-tude correction curve for the straight beam end Hardness tests are performed on all main scan of main closure studs, nuts, and washers '
l closure bolting to demonstrate that beat treat- are established as follows: ment has been properly performed. For cylinder having a length (L) to O.D. 5.3.1.8 Regulatory Guide l.65 ratio of 7 of less, the distance / amplitude curve is established by a minimum of three Regulatory Guide 1.65 defines acceptable mate- test points along the test distance.. For cyl-rials and testing procedures with regard to inders having length to O. D. ratios larger reactor vessel closure stud bolting for than 7, the miniera number of test points is light water cooled reactors. I four. The test points are nearly equally spaced along the test distance. One calibra-The design and analysis of reactor vessel tion hole is located at a test distance equal bolting materials is in full compliance with ASME to L/2. O Ammeassent 'l 5.35 '
__ . - . _ . . . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ - . _ _ ~ i 25As10(AB EtanAard Plant ann c 5.3.2 Pressure /TemperatureIAmits aos. nuclear heatup and cooldown following a anclear shutdown. ] l 83J.1 thalt Carves Reactor Operation The pressure / temperature limit curves la Figure 5.31 are based on the requirements of Curve C la Figure 5.31 specifies limits ap. ; 10CFR50, Appendix 0. The peressure/ temperature plicable for operation whenever the core is ; limits look different than SRP $setion 5.3.2 critical except for low level physics tests, t because the ABWR temperature limits are based on a more recent revlalon of Regulatory Oulde 1.99. 83J.1.4 RaneterVessel Am=aana, ] l 1 All the vessel shell and head areas remote from laplace anneallag of the reactor vessel, > discostlaulties plus the feedwater sosales were because of radiation embrittlement,is not an. , evaluated, and the operstlag limit curves are ticipated to be necessary. - l based on the limiting locatloa. The boltup limits ; for the flange and adjacent shell reglos are based g3J.1.g predicted Shin la RTNDTand on a alaimum metal temperature of RTNOT Drop ia Upper. Shelf Emergy (Appendt G.IV B)
+600F, The maximum throughwall temperature gra. ,
dient from continuous heatlag or cooling at For design purposes the adjusted reference > 1000F per hour was considered. The safety all ductility temperature and drop la the i factors applisal were as sper:ified in ASME Code, Ap. upper. shelf energy for SWR vessels is predicted peridix G, and Reference 2. using the procedures in Regulatory Oside 1.99. The material for the vessel will be provided The calculations (see response to Quertion
,,, with the following requirements of RTNDT as 251.5) are based on the rpecified limits on g determi ed in accordance with Brauch Technical Phosphorous (0.020%), Vanadium (0.05%), Copper Positiot. MTER 5 2: shell ang flanges .20*F; (0.08%) and Nickel (1.2%) in the weld material. Q aortin 20'F and welds 20 P. In plate material, the limits are Copper (0.05%) -
and Nickel (0.73%). Forgings will have the same
$J.2.1.1 Temperature Units for Boltop chemitry as plate but the nickel limit is 1%. )
Minimum closure fitage and fastener tem. The ABWR neutron fluences are top weG peratures of RTNDT r600F are required for compared with the past reactors becaust 4 the ~l J tensioning at preload condition and during fact that the incorporation of interaal pumps i g,, detensioning. Thus, tbe limit is 60'F + increased tlse annulus between the shourd and the 10' F = 70' F. vessel wall. 53.2.1J Tempersture Units for ISI Hydro. A surveillance program in accordance with static and I.mak Pressure Tests - ASTM E 185 will be used. The surveillance n program willinclude samples of base metal, weld g 1 Pressure (measured in the top head) versus tem. metat and heat affeeted aone eatcrial, perature (minimum vessel shell and head metal tem. Subsection 5.3.1.6 provides added detail on the perature) limits to be observed for the test and surveillance program. operating conditions are specified in Figure l 5.31. Temperature limits for preservice and 53.2.2 Operating Procedums inservice tests are shown in Curve A of Figurc 5.3 1. A comparison of the pressure versus tem. perature limit in Subsection 5.3.2.1 with in. 5.3.1.13 Operstlag units During Heatup, tended normal operation procedures of the most Cooldown.and Cort Operation severe upset transient shows that those limits will not be exceeded drring any foreseeable Heatup and Cooldown. upset condition. Reactor operating procedures have been established so that actual transients Curve B in Figure 5.31 specifies limits for will not be more severe than those for which the Anwndnwat 11 SM
ABM isisionin Renadard Plant uvq vessel design adequacy has been demonstrated. Of the design transients, the upset condition producing ' - O' the most adverse temperature and pressure condition anywhere in the vessel head and/ or shgli areas yields a minimum fluid temperature of $28 F and a maximum peak pressure of 1215 psig. Scram auto-matically occurs as a result of this event priorfo a possible reductica in fluid temperature to 250 F at a pressure of 930 psig. Per Figurg 5.31, both the 1215 psig vessel pres F (Curve C) and the 930 psig at 250,sure at $28F (Curve B) are within the calculated margin against nonductile failure. 5.3.3 ReactorVesselIntcgrity The reactor vessel material, equipment, and ser-vices associated with the reactor vessels and appur-tenances would conform to the requirements of the subject purchase documents. Measures to ensure con. formance included provisions for source evaluation and selection, objective evidence of quality fur-nished, inspection at the vendor source and examina. tion of the completed reactor vessels. GE provides inspection surveillance of the O 1 i Amendment 11 SMe I i _ _ _ . a
ABM ~ nasiaarn Standard Plant uve Red::ndant interlocks prevent opening valves discussed in Sections 3.5, 3.6, 3.7 a n d to the low pressure suction piping when the Subsection 9.5.1 . reactor pressure is above the shutdown range. These same interlocks initiate valve closure on $A.7.3 Systems Design increasing reactor pressure. 8.4.7.2.1 System Diagrams In addition, a high pressure check valve will close to prevent reverse flow if the pressure All of the components of the RHR system are should increase. Relief valves in the discharge shown in the P&lD (Figure 5.410). A piping are sired to account for leakage past the description of the controls and instrumentation check valve, is presented in Subsection 7.3.1.1.1 emergency core cooling systems control and instrumen. SA.7.1 A Design Basis With Respect to General tation. Design Criterion 8 Figure 5.411 is the RHR process diagram anJ The RflR system for this unit does not share data. All of the sizing modes of the system are equipment or structures with any other nuclear shown in the process data. The interlock block unit, diagram (IBD) (Logic Diagram) for the RHR system is provided in Section 7.3. SA.7.13 Design Basis for Reliability and Operablilty Interlocks are provided to prevent: (1) drawing vessel water to the suppression pool, The design basis for the shutdown cooling (2) opening vessel suction valves above the mode of the RHR System is that this mode is suction lines or the discharge line design controlled by the operator from the control pressure, (3) inadvertent opening of drywell room. The only operations ; erformed outside of spray valves during RHR operation where the the control room for a normal shutdown is manual injection valve to the reactor is open and when operation of local flushing water admission drywell pressure is not high enough to require
.. ' valves, which are the metns of providing clean the drywell spray for pressure reduction, and water to the shutdown portions of the RHR system. (4) pump start when suction valve (s) are not open. A description of the RHR system logic Three separate shutdown cooling loops are (i.e., interlocks, permissives) is presented in providedt and although the three loops are Table 5.4 3.
required for shutdown under normal circumstgnees, the reactor coolant can be brought to 212 P in SA 7.2.2 Equipment and Component Description less than 36 hours with only two loops in operation. The RHR system is part of the ECCS (1) System Main pumps and therefore is required to be designed with redundancy, piping protection, power separation, The following are system performance require-etc., as required of such systems. (See Section ments the main pumps must satisfy. The pump 6.3 for an explanation of the design bases for equipment performance requirements include ECCS Systems.) additional margins so that the system perfor. mance requirements can be achieved. These Shutdown suction and discharge valves are roargins are standard GE equipment specifica-required to be powered from both offsite and tion practice and are included in procure-standby emergency power for purposes of isolation ment specifications for flow and pressure and shutdown following a loss of offsite power, measuring accuracy and for power source fre-quency variation. SA.7.1.6 Design Basis for Protection from physical Damage Number of Pumps 3 The design basis for protection from physical Pump type Centrifugal damage, such as internally generated missiles, pipe break, seismic effects, and fires, are Drive unit type Motor Amendment 7 5.4-19 l
- m. .. . .
ABM sasioni. me..a..oi... me f Design Sow rate 4200 GPM lato the shutdown cooling mode below a somlaat vessel pressure of 135 psig. Totaldischarge head 410 ft at design Sow rate The RHR beat exchanger capaelty is re-quired to be suffielent to meet each Manimum bypass 8ow 6500PM of these faaetional requirements. The limiting function for the RHR heat as-Minhamn total 623 ft c. hanger capaelty is post LOCA eostala-discharge head at . meat cooling. 'Ibe heat enchanger capac. manimum bypass Bow rote ity, K, is 19$ Stu/sec SP per heat ex- ' changer. Manhaus runout Sow 4975 GPM The performance characteristics of the Maximum puse brake $50 kw heat exchangers are shown in Table - borsepwei $.4 4. Not posiths salon 7.87 ft (3) Valves head (NPSH) at 3.28 ft above the All of the directional valves la the system pump floor setting are conventional gate, globe, and check valves designed for nuclear service. The in. Process fluid $0 to 360 F jection valves are high speed valves, as op. temperature range eration for RHR injection requires. Valve pressure ratings are to provide the control (2) Heat Exchangers or isolation function as necessary; i.e., all vessel isolation valves are treated as The RHR beat exchangers have three major Class 1 nuclear valves at the same pressure functional tequirements imposed upon them, as the primary system. They are as follows: (4) ECCS Portions of the RHR System (a) Post.LOCAPnntalaenant ennu . TheRHR limits the peak bulk suppression pool The ECCS portions of the RHR system include temperature to less than 2070F those sections that inject water into the (970C) by direct pool cooling with two reactor vessel, out of the three dhisions. The route includes suppression pool suction (b) Reactor Shutdown. The RHR removes strainers, suction piping, RHR pumps, dis-enough residual heat (decay and sen. charge piping, RHR heat exchangers,injec. sible) from the reactor vessel water to tion valves, and drywell piping into the cool it to 1400F) (600C) within 24 vessel nortles and core region of the hours after the control rods are in. reactor vessel. serted, This mode shall be manually ac. tivated after a blowdown to the main con. Pool cooling components include pool suction denser reduces the reactor pressure to strainers, piping, pumps, heat exchangers, below 135 psig with all three divisions and pool return lines, in operation. Containment spray components are the same as (c) Safe Shutdown. The RHR brings the pool-cooling components except that the reactor to a cold shutdown condition of spray headers replace the pool return lines, less than 2120F (1000C) within 36 hours of control rod lasertion with two 5.4.7.2.3 Controls and instrumentation out of the three divisions in op-eration. The RHR i manually activated Controls and instrumentation for the RHR system are described in Section 7.3. Amendment 11 $.4 20 1
ABWR =>. m- 2..a pi... l me CHAPTER 6 TABLEOFCONTENTS (Continued) Eastian Ilda East 6.7 NHROGEN GAS SLIPP1XMTIEM 6.71 6.7.1 Punedoes 6.71 6.7.2 System Descripdos 6.71 6.7.3 System Evaluation 6.7 1 6.7.4 Inspection and Testing Requirements 6.72 6.7.5 lastrumentation Requirements 6.72 APPENDIX 6A REGULATORY GUIDE 1.82,SECTION C, . COMPLLANCE ASSESSMENT APPENDIX 6B SRP 6.5.1 TABLE 6.8.11 COMPLtANCE ASSESSMENT O r 6.0lv Ammadment 11 l 1 hv
e 2M6100AB RimmAard Plant arv n SECTION 6.2 CONTENTS (Continued)
- m. m e.S l 614311.1.3 HPCP Line 6.2 28 614311.L4 Standby Uquid ControlSystem Une 6.2 28 614.311.1.5 Reactor Water Cleanup System Une 6.2 28 (Reactor VesselHead Spray) 6.2.4311.1.6 Recirculation Pump Seal Purge Water Supply Line 6.2 28 6.2.4.311.2 Effluent Unos 6.2 28 61431111 Main Steam and Drain unes and RCIC Steam Une 6.2 28 6143.2.112 RHR Shutdown Cooling Line 6.2 29 61431113 Reactor Water Cleanup System Suction une 6.2 29 4 6143.2.13 Conclusion on Criterion $5 6.2 29 6.2.43.2.2 Evaluation Against Criterion 56 6.2 29 6.2.4312.1 lafluent Unes to Suppression Pool 6.2 29 6.2.43.2.2.1.1 HPCF and RHR Test and Pump Minimum Flow Bypass Lines 6.2 29 6.2.43.2.2.1.2 RCIC Turbine Exhaust and Pump Minimum ;
Flow Bypass unes 6.2 30 6143.2.2.13 SPCU Discharge Line 6.2 30 6.2.43.2.2.2 Effluent Lines from Suppression Poo. 6.2 30 l l 6.2.43.2.211 RHR, RClC and HPCF Lines 6.2 30 614312.2.2 SPCU Suction Lines 6.2 30 1 6.2.4.3.2.23 Conclusion on Criterion $6 6.2 30 6.2.43.23 Evaluation Against Criterion 57 6.2 30 6.2.43.2.4 Evaluation Against Regulatory Guide 1.11 6.2 31 61433 Evaluation of Single Failure 6.2 31 O 6.2.vi i Amendment 2
21A610(WB RemmAmed Plame me, e SECTION 6.2 CONTENTS (Continued)
= = n.
614.4 Tests and lampections 6.2 31 618 Camahmatible Gma centent la cantalamment 6.2 31 6.2.5.1 Design Bases 6.2 31 615.2 System Design 6.2 33 6.2.5.2.1 General 6.2 33 61512 leerting Equipment 6.2 34 6.2.5.2.3 Nitrogen Make.Up 6.2 35 6.2.5.2.4 Drywell Bleed 6235 6.2.5.2.5 Pressure Control 6.2 36 6.2.5.2.6 Overpressure Protection 6.2 36 6.2.5.2.7 Recombiner 6.2 36 6.2.5.3 Design Evaluation 6.2 36 6.2.5.4 Tests and Inspections 6.2 36.1 6.2.5.5 Instrumentation Requirements 6.2 37 6.2.5.6 PersonnelSafety 6.2 38 6.2.6 containment Lankage Testing 6.2 39 616.1 Containment lategrated IAakage Rate Test 6.2 39 6.2.6.1.1 laitiallategrated Leak Rate Test 6.2 39 6.2.6.1.1.1 Objectives 6.2 39 6.2.6.1.1.2 Preoperation Test Procedure 6.2 39 6.2.6.1.1.3 Supplement Verification Test 6.2 40 6.2.6.1.1.4 lastrumentation Requirements 6.2 40 6.2.6.1.1.5- Acceptance Criteria 6.2 40 6.2.vil i Aswadment it
l A.BWR
.a.a pi...
m ima am i l l SECTION 6.2 ) O ILLUSTRATIONS (Continue:1)
)
ripre Ihlt East . t 6.2 10 MSLA Area as a Punaion of Tkne 6.240 6.2 11 Foodwater Sped 6e Ratkalpy as a Punaion of
- i. Integrated Feedwater Flow Mass 6.241 l 6.2 12 Presswe Tkne Etery for MS12 with Two Phase i '
Bio down starting when the RPv contapsed 1Avel i Rendes the Main Steam Noarje 6.242 6.2 13 Temperature Time History for MSLB with Two Phase , Blowdown Starting When the RPV Collapsed level i~ Reaches the Main Ste.am Nozzle 6.243 6.2 14 Pressure Time Etory for MSLB with Two Phase Blowdown Starting at One Second 6.244 6.2 15 Temperature Time Etory for MSLB with Two Phase Blowdown Starting at One Second 6.24$ 3 6.2 16 General Pressure Trends in the Containment { During a Post LOCA Depressurization Transient 6.246 6.2 17 Differential Preaures in Wetwell and Drywell Relative to Reactor Building for Vacuum Breaker t Size of 8.3 ft' 6.2-67 I 6.2 18 Differential Pressures in Wetwell and Drywell Relative l l to Reactor Building w'ith Wetwell Spray for Vacuum ! Bre.aker Size of 8.3 ft 6.268 , 6.2 19 Temperature and Pressure Time Etories in the I Containment During Stuck Open Relicf Wive Transient 6.2-69 t 6.2 20 Quencher Bubble Pressure Time Etory 6.2 70 6.2 21 Peak Pressure Pulse Train in Top Vent During Chugging 6.2 71 l 6.2 22 Break Flow Rate and SpeciGc Enthalpy for the Feedwater ' Line Break Flow Coming from the Feedwater System Side ' 6.2 72 6.2 23 Break Flow Rate and Specific Enthalpy for the , l Feedwater Line Break Flow Coming from RPV Side 6.2 73 O > 6.2.x
..,,r. _m,..,-...---..,..-~...r---<-
MSN 2nA610tas ! Etandard Plant mg - l SECTION 6.2 gi j ILLUSTRATIONS (Continued) Eigurr 31tle East 1 i 6224 Break Flow Rate and Specific Enthalpy for the Main i Steam Line Break With Two Phase Blowdown Starting ! When Collapsed Water Level Reaches the Steam Nonle 6274 6.2 25 Break Flow Rate and Specific Enthalpy for Main Steam Line Break with Two Phase Blowdown f Starting at 1 Second 6.2 73 I 6.2 26 ABWR Containment Boundary Nomenclature 6.2 76 ~ i 6.2 27 Three Basic Types of Leakage Paths 6.2 77 , 6.2 28 ! Containment Boundaries in the Rgaetor Building Plan Section A A (0180 ) 6.2 78 6.2 29 Containment Boundaries in the Reactor Building Plan Section B B (90 270 ) 6.2 79 6.2 30 Containment Boundaries in the Reactor ; Building Plan at E1(-)13200 mm 6.2 80 6.2 31 Containment Boundaries in the Reactor
- Building Plan at El(-) 6700 mm 6281 '
i 6.2 32 Containment Boundaries in the Reactor Building Plan at El(-) 200 mm 6.2 82 6.2-33 Containment Boundaries in the Reactor l Building Plan at El 7300 mm 6.2 83 6.2 34 Containment Boundaries in the Reactor Building Plan at El13100 mm 6.2 84 6.2 35 ' Containment Boundaries in the Reactor Building at El 18500 mm 6.2 85 6.2 36 Containment Boundaries in the Reactor Building Plan at El 26700 mm 6.2-86 6.2 37 Secondary Containment Schematic Flow Diagram 6.2 87 l 6.23S Group Classification and Containment Isolation Diagram 6.2 88 6.2 xi j Amendment 11 [
ABWRL uxuma -
$13ndard Plant = aw. c SECTION 6.2 -$ ILLUSTRATIONS (Continued) [
Figure Dile East 6.2 30m Atmospheri: Control System P&lD ' '6.2 90 j i 6.2 40 Flammability Control System P&ID 6.2 92 6.2 41 Hydrogen and Oxygen Concentrations in Containment after Design Basis LOCA - 6.2 93. ' e 1 1 l i' i i 6.2-xii
. Amendment 11 i l 1 1
ABM - =>4siaorn --
> Reandard Plant , ' arv. c Subsection 5.4.7.4 for ihrther discussion of laltlal suppression pool temperature and the RHR.
O preoperational testing.) 42JJ Desiga Evalentles of tbe Contataneemt service water temperature are at their maxhnum values. This assumption maximlacs the heat sink temperature to which the containment heat is Coellag System rejected and thus maximites the containment temperature. In addition, the RHR heat 6223.1. System Operstlos and Segosace of sachanger is ast,umed to be in a fully fouled Events condition at the time the accident occurs. This conservatively ininimizes the heat exchanger heat In the event of the postulated LOCA, the removal capacity. Even with the degraded short term energy release from the reactor primary conditions outlined above, the maximum system will be dumped to the suppression pool. temperature is maintained below the desige limit Subsequent to the accident, fission product decay specified in Subsection 6.2.2.L beat will result in a continuing energy input to . .
. the pool. The RHR auppression pool cooling mode It should be noted that, when evaluating this -
will remove this energy which is released into the long term suppression pool transient, all. heat primary containment system, thus resulting in sources in the containment are considered with . acceptable suppression pool temperatures and no credit 1a. n for any heat losses other than containment pressures. .through the RHR heat exchanger. These heat-l sources are discussed to Subsection 6.2.1.3. ! In order to evaluate the adequacy of the RHR system, the following is assumed: It can be concluded that the conservatue evaluation procedure described above cleerly i (1) With the reactor initially operating at 102% demonstrates that the RHR system in the of rated power, a LOCA occurs. ' suppression peol cooling mode !!mits the post LOCA containment temperature transient. (2) A single failure of a RHR beat exchanger is the most limiting single failure. 6.2.2.4 Test and laspections (3) The ECCS flows assumed available are 2 HPCF, The containment cooling system is required 1 RCIC, and 2 LPFL (RHR). to have scheduled maintenance. The system testing and inspection will be-performed 3 (4) Containment cooling is iniated.after 10 periodically during the plant normal operation C minutes. (See Response to Question 430.26) and after each plant shutdown. Functional testing will be performed on all active p Analysis of the net positive suction head components and controls.' The system reference g (NPSH) available to the RHR and HPCF pumps in characteristics will be established during
- accordance with the recommendations of Re6clatory preoperational testing to be used as base points Guide 1.1is provided in Tables 6.2 2b and 6.2 2c, for checking measurernents obtained from the l respectively, system tests during the plant operation.
General con pliance for Regulatory Guide 1.26 Thu-pre operational test program of the may be, found in Subsection 3.2.2. containment cooling systern is described in Subsection 14.2.12. The following functional tests will be performed. The RHR pump will be tested through the suppression pool cooling loop operation by measuring flow and pressure. Each 6J.2J.2 Summary of Containment Cooling pump will be tested individually. Analysis Containment spray spargers will be tested When calculating t!'c long term, post LOCA pool during reactor shutdown by air, and by visual temperature transient, it is assumed that the inspection to verify that all the aczzles are Ammendment 11 6.2 17
' ~' ' '
, u
ABM nasi ora r nev. e Standard Plant clear. RHR heat enchangers will be checked for source of radioactive release after an ; effectivene'es by measuring talet and outlet accident. During normal plant operation, the temperatures at the tube and shell sides. secondary containment areas are kept at a , negative pressure with respect to the All motor. and alr. operated valves required ' environment and clean zone by the HVAC system.- i for safety are capable of being esercised and Following an accident, the standby gas treatment their operation demonstrated.:The layout and system (STOS) provides this fasetion. These arrangement of critical equipment outside the . systems are described la sobsections 9.4.5 and < drywell is designed to permit access for 6.5.1, respectively.. 'j appropriate equipment osed in testing and inspecting system. integrity. Relief valves on Fission products that may leak from the f the low pressure lines are removable for testing. prhnary to secondary are processed by the SGTS . 4 before being discharged to the environment. The - g Periodic inspection and maintesance ef the HVAC eshaust systsas and SO'Is are located within , main system pumps, pump motors, and heat. the accondary containment to assure collection ' i sachangers are conducted in accordance with the - of any leakage. The secondary containment
- . manufacturer's instructions, provides detection of the level of radioactivity released to the environment during abnormal and ,
During the normal plant operation, the pumps, accident plant conditions.' Personnel or
- heat exchangers, valves, piping, instrumentation, material entrances to the secondary containment wiring and other components outside the. consist of altlocks with laterlocked doors or containment can be inspected visually at any hatches.
time. Testing frequencies are generally-
- correlated with testing frequencies of the There are basically three types of potential t
- associated controls and instrumentation. When a - leekage paths for the release of fission product i pump or valve control is tested, the operability during and following an accident, These leakage l of that pump or valve and its associated' pr.ths are shown in Figure 6.2 27. Potential -
lastrumentation .is tested by the same action. icek paths that can bypass;the secondary When a syrtem is tested, operation of the containment are shown in Table 6.210. 1 ! components is indiented by miled ' l- instrumentr.tlon. Relief valves are removed as 6.13.1 Dades 3 awn L scheduled at refueling outages for bench tests and setting adjustment. (1) Secondary containment is provided to collect fission products which may leak 1- 6.2.2J Instrumentation Regeleveents from the pritury containment following a DBA. This ' collection allows filtration by Details of the in strumentation are provided in the SGTS prior to release to' the , Section 7.3. The suppression pool cooling mode environment. The secondary containment of the RHR system is automatically 1nitiated, region completely surrounds the primary cos.iainment ve~ssel. 6.2J Secondary Containment Functional 'e l Design (2) During a DBA, the secondary contaissent and
- supporting systems such as the SGTS, is ,
o The secondary containment boundary, as shown - designed to limit the thytold sad whole in Figure 6.2 26, completely surrounds the bdy doses to less than 10CFR100 guidelines primary containment vessel (PCV) except for the at the site, boundary and low population l ' basemat and together with clean zone ccmprises . sone an6 to less than 10CER50, Appendix A, l- the reactor building. The secondary containment Genexai Design Criterion 19 doses for the L encloses all penetrations through the primary control room operator.
- ontainment and all those systems emernal to the primary containment that may become a potential. (3) The mechanical, electrical, instrumeny tation, and strnwal componemts'of the ret:ondary containw.nt desi6 n are protected as necessary from laternally and- - i l _
? G ib i Amendment 9 l l , L
- AB M =arsioorn Standard Plant nev. c q Influent and effluent lines of this group are 6.2.4.4 Test and laspections isolated by automatic or remote manualisolation F
O valves located as close as possible to the-contalament boundary,
- The containment isolation system is scheduled to undergo periodic testing during reactor operation. The functional capabilities 6.2.4.3.2.4 Evaluation Against Regulatoty of power operated isolation valves are tested Guldel.11 remote manually from the control room. By observing position indicators and changes in the Instrunient lines that connect to the RCPB and affected system operation, the closing ability 1 penetrated the containment have 1/4 inch orifices of a particular isolation valve is demonstrated. '
and manual isolation valves, la compliance with Regulatory Guide 1.11 requirements. Air testable check valves are provided on
. influent emergency core cooling lines of the i 6.2.4.3.3 Evaluation of Single Fallen HPCF and RHR systems whose operability is relied upon to perform a safety function.
A single failure can be defined as a failure , of a component (e.g., a pump, valve, or a utility A discussion of testing and inspection of '{ such as offsite power) to perform its intended isolation valves is provided in Subsection safety functions as a part of a safety system, 6.2.1.6. Instruments are periodically tested The purpose of the evaluation is to descastrate and inspected. Test and/or calibration points that the safety function of the system will be are supplied with each instrument. . Leakage l completed even with that single failure.. Integrity tests shall be performed on the ' Appendix A to 10CFR50 requires that electrical containment isolation valves with resilient systems be designed specifically against a single material seals at least once every 3 months. passive or active failure. Section 3.1 describes . . the implementation of these standards as well as 6.2.5 Combustible Gas Controlin General Design Criteria 17,21,35,38,41,44, Containment i 54,55 and 56. O The atmospheric control system (ACS T31) is l Electrical as well as mechanical systems are provided to establish and maintain an inert designed to meet the single failure criterion, atmosphere within the primary containment during ' regardless of whether the component is required all plant operating modes except during shutdown to perform a safety action. Even though a com- for refueling or equipment maintenance and ponent, such as an electrically operated valve, during limited periods of time to permit access is not designed to receive a signal to change for inspection at-low reactor power. The state (open or closed) in a safety scheme,it is flammability control system (FCS T49) is i assumed as a single failure if the system compon- provided to control the potential buildup of 3 ent changes state or fails. Electrically oper. oxygen from design basis radiolysis of water.
~
ated valves include valves that are electric- The objective of these systems.is to preclude , ally piloted but air operated, as well as valves combustion of hydrogen and damage to essential . that are directly operated by an electrical de- equipment and structures, i vice. In addition, all electrically operated . i valves that are automatically actuated can also 6.23.1 Design Bases j be manually actuated from the main control room. ' Therefore, a single failure in any electrical . Following are criteria that serve as the system is analyzed, regardless of whether the= bases for designi l loss of a safety function is caused by a component falling to perform a requisite (1) _Since there is no design requirement for mechanical motion or a component performing an the ACS or FCS in the absence of a LOCA and < unnecessary mechanical motion, there is no design basis accident in the ABWR that results in core uncovery or fuel . failures, the following requirements mechanistically assume that a LOCA Ammendment 11 6.2-31
. . ~ . - . - - _ . - . - . . - - - - - .-- . --- .. .
M . 23A6100AB Ef anAard Plani - un c producing the design basis hydrogen and backup purge function need not meet this oxygen has occurred. ' eriterion. (2) _ The hydrogen generation from ractal water- (10) Components of the system are protected from reaction is defined in Regulatory Onide 1.7. postulated missiles and from pipe whip, as required to assure proper action as well as (3) The hydrogen and oxygen generation from ' other dynamic effects such as tornado radiolysis is defined in Regulatory Guide missiles and flooding._ i 1.7. (11) The system has the capability to withstand : (4) The ACS establishes an inert atmosphere the wynamic effects associated with the throughout the primary containment following safe shutdown earthquake without loss of-' , an outage or other occasions when the function. i containment has been purged with air to an i i oxygen concentration greater than~ 3.5 - (12) The system is ~ designed so that all percent. components' subjected to the primary-containment atmosphere (inboard isolation (5) The ACS maintains the primary containment valves) are capable of withstanding the oxygen concentration below the maximum temperature an'd pressure transients permissible limit per Regulatory Guide 1.7 resulting from a LOCA. These components during normal, abnormal, and acident will~ withstand the humidity and r# don l conditions in order to assure az inert . conditions in the wetwell or drywell. atmosphere, following a LOCA. (6) The ACS also maintains a slightly positive (13) The'ACS is nonsafety class except as pressure in the primary containmet.t during necessary to assure primary containment normal, abnormal and accident conditions to integrity (penetrations, isof ation- - prevent air (oxygen) leakage _into the
~
valves). The ACS and FCS are designed and' inerted volumes from the secondary built to the requirements specified in containment, and provides non essential Section 3.2. * , monitoring of the oxygen concentration in l' the primary containment to assure a l breathable mixture for safe personnel access L ! or an inert atmosphere, as required. (14) The ACS includes the nitrogen' storage l Essential monitoring is provided by the tanks,; vaporizers, valves and piping containment atmospheric monitoring system carrying nitrogen.to the containment, (CAMS) as described in Chapter 7. valves and piping from the containment to the SGTS and HVAC (U41) exhaust line, (7) The drywell and the suppression chamber will non safety oxygen monitoring, and all-be mixed uniformly after the design basis related instruments and controisi The ACS IDCA due to natural convection and molecular does not include any structures housing or s diffusion. Mixing will be further promoted supporting the aforementioned equipment or ' by operation of the containment sprays. any ducting in the primary containment. L , (8) The system is' capable of controlling (15) The~ system is designed to facilitate L combustible gas concentrations in the periodic inspections and tests. The ACS l- containment atmosphere for the design baser, can be inspected or tested during normal i LOCA without relying on purging and without plant conditions. l releasing radioactive material to the I environment. (9) The system is designed to anaintain an inert primary containment after the design bases . LOCA assuming a single active failure. The Ammendment 11 6.2-32 i.
E L MM 25A6100AB - Standard Plant an c (16) The primary containment purge system will required, through a pathway from the wetwell 1 aid in the long term post accident cleanup altspace to the stack. The pathway is isolated operation. The primary containment during normal operation with two rupture disks, atmosphere will be purged through the SGTS to the outalde environment. ' Nitrogen makeup The following modes of operation are provided: ; will be available during the purging { operation. (1) . Startup - Inerting.' Liquid nitrogen is 1 vaporized with steam or electric heatere to l (17) The system is also designed to release a temperature greater than 200F and is containment pressure before uncontrolled injected into the wetwell and the drywell.- ! containment failure could occur. The nitrogen will be mixed with the primary l containment atmosphere by the drywell i 6.2.8.2 Systaa Design coolers in the drywell and, if necessary, by the sprays in the wetwell. 6.2.5.2.1 General (2) Normal Maintenance of Inert (Won.' A The ACS provides control over hydrogen end nitrogen makeup system automatically sup- ! oxygen generated following a LOCA. In an inerted plies nitrogen to the wetwell and upper =1 containment, mixing of any hydrogen generated is drywell to maintait, a slightly positive not required. Any oxygen evolution from pressure in the drywell and wetwell to pre-radiolysis is very slow such that natural clude air leakage from the secondary to the convection and molecular diffusion is sufficient primary containment. An increase la con-to provide mixing. Spray operation will provide tainment pressure is controlled by venting further assurance that the drywell or wetwellis through the drywell or wetwell bleed line, uniformly mixed. The system consists of the following features: (3) Shutdown - Deinerting. Air is provided to { the drywell and wetwell by the primary ' (1) Atmospheric mixing is achieved by natural containment HVAC purge supply fan. Exhaust processes. Mixing will be enhanced by is through the drywell exhaust lines and operation of the containment sprays, which wetwell to the plant vent, through the HVAC l are used to control pressure in the primary or SGTS, as required. containment. (4) Overpressure Protection. If the wetwell I (2) The primary containment nitrogen purge pressure inctcases to about 5.6
~
establishes and maintains an oxygen - kg/cm2 g, the rupture disks will open. l deficient atmosphere (.s.3.5 volume percent) The overall containment pressure decreases in the primary containment during normal as venting continues. Later, the operators operation. can close the two 350A air operated butterfly- valves to re establish (3) The redundant oxygen analyzer system (CAMS) contsinment isof ation as required. measures oxygen in the drywell and suppression chamber. Oxygen concentration The following interfaces with other systems i are displayed in the. main control room, are provided: Description of safety related display instrumentation for containment monitoring (1) Residual Heat Removal System (RHR Ell).- l Is provided in Chapter 7. Electrical The RHR provides post accident suppression a requirements for equipment associated with pool cooling as necessary following heat l the combustible gas control system are in dumps to the pool, including the exothermic accordance with the appropriate IEEE heat of reaction released by the design : i stasdards as referenced in Chapter 7. basis metal water reaction. This heat of reaction is very small and has no real In addition, the ACS provides overpressure affect on pool temperature or RHR heat protection to relieve containment pressure, as exchanger sizing. The wetwell spray- i Ammendment 11 6.2-33
MM. 21A6100AB MC liliamAard Plane portion of the RHR may be activated during-- i the inerting and deinerting processes to ~ help mixing by reducing pocketing. Wetwell .1 spray would also serve to accelerate l deseration of the suppression pool water, i though the impact of the dissolved oxygen on wetwell airspace orygen concentration is , very small. Wetwell spray is not required to provide mixing in the wetwell.' The RHR also provides coollag water to the exhaust flow from the FCS. , (2) ' Drywell Cooling System (DWC T41): Drywell cooling system provides enough circulation -! to all portions of the upper and lower drywell, the drywell head area, and the
- vessel support skirt area such' that this mixing occurs and d o e s' n o t t
\,
i I t l l eH Amendment 11 6.2-33.1 g
, _....___.a ,_ , 5
ABWR === Standard Plant nev. c
-limit completion of either the inerting or nitroge.n gas supply system is supplied from 1 deinerting_ process and provides a the ACS nitrogen storage tank.
9 - representative oxygen sample for the ACS oxygen sensors.' Should the arrangement of j the RPV insulation leave a significant gap ! between itself and the RPV, forced circulation will be provided to that area. (5) Standby Gas Treatment System (SGTS T22). Alternatively, the major portion of the ' The SGTS processes any drywell bleedoff, drywell will be inerted to sufficiently inerting, and deinerting flows, cs required' below 3.5% such that the bulk average oxygen by offsite release constraints. concentration does not exceed 3.5% percent r 0' 2 [ (3) HVACSystem(U41).TheHVACaccommodates the drywell and wetwell exhaust flow during (6) Containment Atmosphere Monitoring System ! inerting and deinerting, accommodates (CAMS D23). The CAMS monitors oxygen j drywell bleedoff flows during startup and levels in the wetwell and drywell during - normal op: ration, provides sufficient air accident conditions to confirm the primaty. ; flow to limit the concentration of any. containment is inert. nitrogen leaking from the primary l containment into the secondary containment, - Radiation monitoring in the plant vent, part l l and supplies air for purging the primary of Process Radiation Monitoring (PRM D11), ' containment during deinerting. The real detects high radiation during deinerting. effect of leaking nitrogen from the primary . containment is insignificant and does not There are no potential sources of oxyge'n in 4 impact HVAC design. containment other than that resulting from radiolysis 'of the reactor coolant. Consideration of potential sources of leakage of - oxygen into the containment included'not only normal plant conditions =but also postulated ! loss of coolant accident conditions. Potential sources of leakage are instrument air systems, service air lines, leakage control systems, The intake of the control room portion of purge lines, and inflatable door seals. .3 the HVAC system is located to protect Nitrogen is substituted for air service wherever personnel in the control room in the event leakage into the inerted containment could be of a nitrogen pipe or storage tank rupture, postulated. ; Similarly, suction for all HVAC systems is located to minimize the introduction of 6.2J.2.2 Inerting Equipment nitrogen from a break into occupied areas of the plant. The inerting subsystem is capable of reducing the wetwell and'drywell oxygen - 1 (4) High Pressure Nitrogen Gas Supply System concentrations from atmospheric conditions to (HPIN P54). Because the containment is less than 3.5 percent in less than four hours. i l inerted, all pneumatically-operated The_ inerting vaporizers are sized to provide at components in the primary containment are least 2.5 times the containment-(wetwell and I l normally supplied with nitrogen. The ;drywell) free volume of nitrogen within the pneumatic devices in the primar/ contain nent allotted four hours. The specific.d oxygen limit _ : or those which could lenk ' o the primary of 3.5 volume percent must be adjusted'for l I containment are supplied ,th nitrogen for initial containment conditions, instrumentation 4 [ the purpose of preventing oxygen addition to errors, operator and equipment response time, the inerted volumes. The high pressure an:! equipment performance to ensure that the actual Ammendment 11 6.2.M
. _ _ _ _ . . . _ . . . . - - . . . . . . ~ . -
MM 2SA6100AB Riandard Plant ' nn c l caygen concentration does not exceed 3.5 volume then be isolated. Penetrations through the O percent during normal operatk . The actual nitrogen storage tank insulation are minimized oxygen concentration shall not cxceed five to reduce heat gala. The length of piping volume. percent during an accident when the through the lasulation is manielsed to the hydrogen concentration is greater than tour extent practicable to reduce heat gain, percent. The inert containment can be deinerted to allow safe personnel access without breathing The drywell and wetwell atmospheric oxygen - l apparatus in less than four housu, concentration will be less than 3.5% by volume l ' within 24 hours after thermal power is greater Each penetration and pipe carrying nitrogen is sloped as necessary to prevent condensation than operation 15% of3.5% above plant oxygen rating. by volume Twenty and 15% four hours of l collection and llac blockage and shall be power is allowed before a scheduled shutdown. protected against entry of debris. All piping cutside the outboard primary l containment isolation valves carrying nitrogen l All pipe volumes where liquid or very cold are protected from overpressurization by relief nitrogen could be trapped between closed valves valves ducted to the atmosphere. have relief valves. All relief valves exhaust
- outside the reactor building. Means are provided 6.2.5.2J Nitrogen Makeup to add nitrogen to the nitrogen storage tank vapor space (to decrease tank pressure) and the (1) The nitrogen makeup equipment is sized to liquid volume (to increase tank pressure). Tank maintain a positive pressure in the drywell level and pressure indication are provided at the and wetwell during the maximum drywell cool tank. Means for startup full scale testing of down rate not caused by apray actuation.
, the inerting and makeup portions of the system without nitrogen injection to the containment is (2) Automatic addition of. nitrogen is . provided. During startup, the test discharges physically limited to less than the maximum shall be temporarily piped away from the control dr>well bleed capacity , l panel and storage and vaporization equipment to avoid excessive noise from the open discharge. 6.2.5JA DrywellBleed l Strainers are provided in the liquid portion of the makeup and inerting lines. Means are provi. Primary contaicment bleed capability is ded to feed the makeup circuit from either the provided in accordance with Regulatory Guide 1.7 i liquid or vapor portion of the nitrogen storage. and as an aid in cleanup following an accident.
- Pressure is automatically maintained in the During normal plant operation, the bleed line 4
nitrogen storage tank during nitrogen discharge also functions, in conjunction with the nitrogen by a circuit with another ambient heat exchanger purge line to maintain primary containment i fed by a pressure control valve. The inerting i and makeup portions of the system do not rely on pressure concentrationat about below 0.75 psig 3.5 percent and oxygen by volume. This l
- pumps to perform their function. Means are is accomplished by makeup of the required l provided to manually vent the tank vapor space to quantity of nitrogen into the primary
; control pressure. Means are provided to drain containment through the makeup line or relieving - , the storage tank. The vessel bottom is sloped or pressure through the bleed line. The drywell
, dished to facilitate this draining, bleed line is manually operable from the control room. Flow through the bleed line will be Pressure relief for the nitrogen storage tank directed through either the SGTS or the is provided at 10 percent above the upper limit secondary containment HVAC, and be ruonitored by of the normal range of operating pressures. the SGTS and SCHVAC flow and radiation 3~ Rupture disks, set 00 percent above the upper instrumentation. All ACS primary containment limit but not higher than the design pressure of isolation valves are automatically closed when the vessel, are provided. Redundant pressure high radiation is detected in the exhaust flow. relief valves are provided so that protection is immediately.available should a disk rupture and The drywell bleed line is located above an l Annead<nent 11 6.2-35
ABM meandard Plant siaaicais - anvic - 5 _ elevation which would be covered by post LOCA - the same time and made from the same sheet , flooding for unloading the fuel, to provide uniformity of relief pressure. 1.
\ .
62AJJ Pressarecentrol (6) The rupture disks are capable of. ; withstanding full vacuum in the wetwell (1) In general, during startup, normal, and vapor space without leakage. abnormal operation, ths wetwell and drywell pressures is maintained greater than 0 psig (7) The piping material is carbon steel. The to prevent leakage of air (oxygen) into the design pressure is 10.5 kg/cm8s (150 primary containment from secondary psi), and the design temperature is. , containment but less than the nominal 2 psig 171*C, i scram set point, Sufficient margin is l provided such that normal containment 6JJ.1.7 leaaamid==r ' temperature and pressure fluctustions do not . cause either of the two limits to be reached (1) Two recombiner skids are located in . ; considering varlations in initial secondary containment. Each skid, as shown j containment conditions, instrumentation in Figure 6.2-40, takes suction from the i errors, operator and equipment response drywell, passes the process flow through a , time, and equipment performance, heating section, a reactor chamber,;and a spray cooler. The gas is returned to the ! Nitrogen makeup automatically maintains a wetwell. l(2) 530 kg/m2 (0.75 psig) positive pressure to avoid leakage of air from the secondary (2) The recombiners are normally initiated on into the primary containment. high levels as determined by CAMS (if ' hydrogen is not present, oxygen (3) The drywell bleed sizing is capable of concentrations are controlled by nitrogen '! maintaining the primary containment pressure makeup). L less than 880 kg/m2 (1.25 psig) during the maximum containment atmospheric heating 6.2J.3 Design Evaluation i which could occur during plant startup. '. 1 The- ACS is designed to maintain the 6.2JJ 6 Overpressure Protection containment in an inert condition except for - j nitrogen makeup needed to maintain a positive (1) The system is designed to passively relieve containment pressure and prevent = air (02) the wetwell vapor space pressure at 5.6 leakage from the secondary into the primary. kg/cm2g The system valves are capable containment. of being closed from the main control room . using AC power and pneumatic air. The primary containment atmosphere will be inerted with nitrogen during normal operation of (2) The vent system is sized so that residual y core thermal power in the form of steam can the plant.willOxygen containment be maintained concentration below 3.5 volume in the primary # be passed through the relief piping to the- percent measured on a dry bas's.
~
stack. Following an accident, hydrogen concentration , (3) The initial driving force for pressure will increase due to the addition of hydrogen- - relief is assumed to be the expected from the specified design basis metal water pressure setpoint of the rupture disks. reaction. Hydrogen concentration will also - increae due to radiolysis. Any increase in 1 (4) The rupture disks are constructed of hydrogen concentration is of lesser concern stainless steel or a material of similar because the containment is inerted. Due to corrision resistance. dilution, additional hydrogen moves the operating point of the containment atmosphere -
~
(5) A number of rupture disks are procured at farther from the envelope of flammability. , Anuneedment 11 - 6.2 36
23A6100AB
~ RinnAmed Plant am e '
Contaimment oxygen concentration also increases . temporary plugs. Hydrostatic testing of piping-9 due to radiolysis. During plant operation, there systems will be performed at a pressure 1.5 are no other sources o* oxygen in the times the design pressure, but in no case at containment. less thau 75 psig. The test pressure will be y j held for a minimum of 30 minutes. Pneumatic. In the ABWR, there are no design baals events testing may be substituted for hydrostatic j that result in core uncovery or core heatup testing in accordance with the applicable codes. , sufficient to cause signilicant metal water { reaction. Therefore, per Regulatory Guide 1.7, Preoperational testing will demonstrate the i the design basis metal water reaction is that ability of the ACS to meet design requirements, equivalent to the reaction of the active clad to Each valve will be exercised both opened and a depth of 0.00023 laches. This is equivalent to closed and position indication verified. Trip 0.72% of the active clad. Radiolysis is and alarm logic signals will also be checked. ; , calculated based on Regulatory Guide 1.7 source The tests assure correct functioning of all terms. Hydrogen and oxygen concentration controls, instrumentation, compressors, 1 profiles in containment after the design oasis recombiners, piping and valves. System LOCA are provided as Figure 6.2 41. reference characteristics, such as pressure Overpressure relief is provided to passively v relieve the containment pressure, as required, by vesting the wetwell armosphere to the plant ! stack. Venting the wetwell airspace to the plant stack precludes an uncontrolled containment failure. -Venting from the wetwell, as opposed to the drywell, takes adventage of the decontamination factor provided by the G a monitored, elevated release. suppression Precluding containment failure !!mits the maximum fission _ pool. Venting to the stack provides l product relesse as shown in Figure 19E.3 2. Details of the effect of overpressure relief on ABWR performance goals are found in ; Subsections 19.5.2 and 19.5.3. : i Unintended opening of the overpressure relief rupture is highly unlikely and would be characterized by opening of one of the rupture disks in place. Unintended operation at a lower pressure, such as during a design basis accident, would not significantly affect offrite doses, since no fuel failures would be expected. Failure of both rupture disks would be required for this unintended operation. In addition, the butterfly valves could be closed if in line radiation monitoring indicated unexplained flow in the relief line. 6.2JA Tests and Inspections Complete process systems are pressure tested to the maximum practicable extent. Piping systems will be hydrostatically tested in their entirety, utilizing available valvi,s or A-a - - 11 6.2 36.1
- StandtJd MM plant 2M61MAB REV.C 6.17 Reftsences
- 1. ' .WJ. Bilsnia, The G.E. Mark 111 Pressure Suppression Containment Analytical Model, '
June 1974, (NEDO 20533).
- 2. FJ. Moody, Maximum Discharge Rate of.
Liquid Vapor Mixtures from Vessels, General Electric Company, Report No. NEDO 21052, . September,1975.
- 3. ' W.J. Bilanin, The G.E. Mark 111 Pressure a Suppression Containment Analytical Model, Supplement 1, September 1975 (NEDO 205331).
l
- 4. Idaho National Engineering ~ Laboratory,'
RELAP4/ MODS A Computer Program for Transient Thermal Hydraulle' Analysis of Nuclear Reactors and Related Systems, User's Manual, September,1976 (NUREG 1335). O
/
.I O
Ammesubment 11 6.2 41
.AB M 23A6 o0AB -
Standard Plant we TABLE 6.2 7 (Continued)
- h. CONTAINMENT ISOIATION VALVE INFORMATION AWOSPHERIC CONTROLSWIEM Velve No. D1POO1 733F002 DI MiO3 DI-It04 731PJ05 T31-Ivo6 T3119tri SSAR Fig 6.2 N 6.2 M 6.2 M 6139 6.2 3'r 6.2 39 6139 .
Applicable Beels ODC56 ODC$6 ODC56 ODC56 ODC56 ODC $6 - ODC$6 l 1 Fluid Air Air or N2 ' heer N2 DWAD40S DW ADdOS WW ATMOS WW ATMOS f.IJaeSime 22* ' 22' 22' 22' 2* 22* 2' ESF Yes Yes Yes Yes Yes Yes Yes Imahage Class . @) @' @) @) @) . @) @) Imeestem O I I I I I I j i Type ClankTest Yu Yes Yes Yu Yes Yes Yes Vahe'I)pe Butter 0y Butter 0y Butterfly Gutter 0y - Globe Butter 0y Globo Operater Pneum Pneum Pneum Pneum Pneum Pneum Pneura PrL Artseties Air Air Air Air Air Air - Alt Sec. Aetenties N/A N/A N/A N/A N/A N/A N/A I NorimalPoskien Shut Shut Shut Shut Shat Shut Shut l Shetdowa Feskien Shut Shut Shut Shut Shut- Shut Shut ! Poet Ace Peskion Shut Shut Shut Shut Shut Shut Shut Pwr Fall Fesklos Shut Shut Shut Shut Shut Shut- Shut Cent. Iso. Sig. A,K A.K A,K A.K A,K A,K A.K CnosseeTbne (see) < 30 < 30 < 30 < 30 <13 < 30 < 15 Pwr Searce (Div) 1 II II 11 11 11 11 Amendment 11 6.2 50.25
- ABM 234610aAB Standard Plant we TABLE 6.2 7 (Continued)
CONTAINMENT ISOIATION VALVE INFORMATION ! ATMOSPHERIC CON'IROL SWIEM (Contissed) Wehe No. T31Poot TS1 M00 T31 M25 T31 Ro9 D1 PO40 131 M41 T3177J0AB SSAR Fig 6439 6&W 6139 6 S 39 6.2 39 6SW 6439 AppikeMe Reds ODC$6. ODC$6 GDC$6 ODC$6 ODC$6 GDC$6 GDC$7 i Pheld PCYA'DdOS PCV ATNOS N2 . N2 N2 N2= N2 IJee ties 10' 22' 16' _ 2' 2' 2' 3/4' ESF- Yes Yes Yu Yes . Yes Yes No'
&aelege Claes (b) @) @) (b) ~ @) @) @)
lamelee 0 1 O O O I I I Type ClankTen Yes Yes Yes Yes Yes Yes No l 1 VeheType Betterfly Butterfly ButterGy - Obbe Globe . Olobe Oete Operseer Num Num Pneum - Num - Puum Num Soloniod - PrL Attention Air Air Air Air Air Alt'- Elec.
.(
See. Assostlen . . . . . .. .- Nesmal Peeklen Shut Shut Shut Open Open Open Shut shutdows Peeltlee Shut Shut Shut Shut Shut Shut Shut Post Aec Poskien Shut 1 Shut Shut Shut Shut Shut Shut
-{
-1 Pwr FellPoskien Shut Shut Shut Shut Shut Shot Shut
/.
~ Coet.les.Sig.I') A,K A.K A.K A,K A,X g- AK A,K 1
Cleomellene (see) < 30 < 30 - < 30 <15 < 15 <15 <$ Por Seeree (Div) I I I I 11 II i J[ _ __. _ __ -- i Amendment 9 ! 6.2-50.26
,\,,,, ,, ,
^AB M 21A6100AB
' Shadard Plant may. c YABLE 657 (Continued)-
h- CONTAINMENT ISOIATION VAINE INFORMATION SUPPRESSION POOL CIIANUP SYSTEM ' Vebe No. O$1P001 051 R106 OSI7030 SRAR Fig 951 951 951 Appusebb tenis OOC 57 ODC57 09C $7 - PhW RPV H2O RPV H2O - RPV H2O , Line e m P W r ESF Yes Yes No laelege Class (a) - (s) (a) Location 1 1 0 Type C LeakTest No(d) ' No(d) No(d) Vehe Type Gate Oete Gate Operaser Motor Motor Motor Ec. PrL Aetueeien Elec. Elet. Sec.Aetuation Manual Manual Manual Nennal Positten Open Open Open Sheedowa Posallen Open Open Open Poet AN Pesition Shut Shut Shut Pwe Fall Peekten As is As is As is . Cent. lee. Sig.I') A,K A,K A,K Cheure Thee (sec) <30 <30 < 30 Pwr Seem (Div) 11 11 1
~
\
I
..O 1
Amenshment 9 6.2-50.39
ABWR 25Asi0aAn - hadard Plant m.c TABLE 6.2 7 (Continued) CONTAINMENT ISOIA110N VALVE INFORMATION h REACTOR BUILDING COOLING WA1ER SYSTEM Vehe No. - P21 PUPSA P21 POB1A F2149758 P21 R218
/RMA /PUtQA /MEB / PUB 08 88ARPig 941 941 911. 9.21 Appakeueseein. ODC$1 ODCSS ODC$$ ODCSS Maid Water Weeer Weser Water l IAme 8les r- r F - F-RSF-Yes Yes - Yes Yes lashese Clies (b). (b) (b) (b) >
l im ol: og - O/t o/I q I Type C tashTest Yes(e) Yes(e) Yee(s) . Yes(e) i Velve T)pe Oste / Check Gate /Gete Oste / Check Oste /Gete -- Operseer Motor /NA Motor / Motor Motor /NA Motor / Motor Pri, Arselon Beet. , Bret. Beet. Beet'. q Soc. Adestion HW/N/A HW/N/A HW/N/A ' HW/N/A I
--y Norieel Penielen Open Open Open Open i Sheldews Positten Open Open Open Open Pest Aee Peeltion Shut Shut Shut Shut ;j Fwr Fall Peekten As is Asis As is As is Cent.lae, SIS CX,K CX,K CX,K CX,K i,
Closure'line (see) 00/N/A 00/80- 80/N/A 00/00 Fwr Seeree (Div) I/N/A I/II '1/N/A 1/11
\
l Amad=*et 11 6150A0 ' e1 , l
, , +
e 18
. . . . , , , , , . . . , , , . , ,, . ,. .. . ....I I MM hndard Plant 25A6100AB '
new c TABLE 6.2 7 (Continued) ,
- CONTAINMENT ISOIATION VALVE INFORMATION -
.i
'NSTRUMENT AIR SYSTEM Vehe No. PS3-P216 PS2-P2D l
SEARFt 9M 9M 'I AgylleeMeReale ~ ODCSS ODC$5 Flew Air Air une alan r r i EN No No 1 laelege Class (b) (b) l j lanatien 0 1 I
% C laekTest No No Vehe Type Globe Oieck
.9 Operater PrL!atention
- Motor Elect.
N/A N/A Sec. Aeemation . HW N/A ~i NenealPosklen Open Open Sheedews Puskies Open Open Peet Are Poskien Open Open . Pwr Feu Peaklow Asis N/A I Cent. Isa. Sig.I') RM N/A Closure'Ilaw (see) 20 N/A Pwr Seem (Div) i N/A e W9 63.'.0.43
i MM ^ 23A6100AB: Standard Plant we j 1-l
- TABLE 6.2-7 (Continued) -
CONTAINMENT ISOIXLION VALVE INFORMATION h, .I HIGH PRESSURE NmtOGEN GAS SUPPLY SYSTEM Vale No. PS1407/f4DB PS4-POO7/140B PSP 300/P309 3 i SBARFIg 6.71 6.7-1 6.71 AppliseWe Basis - ODC$5 ODC35 ODC$5 Phu 10 N2 N3 i Y Y Y l hh ! ESF Yes - . Yes Yes - laelageCiess (b) @) -- @) 7 laseties O/I - O/I . O/I : t Type ClankTest No . No No ; Valve Type 'llobe/ Check Olobs/ Check Olobe/ Check - i Operator - Motor /N/A Notor/N/A Motor /N/A PrL Actuation h t/N/A h /N/A Bloct./N/A , See.Aatemenen HW/N/A HW/N/A HW/N/A - l Noneal Postiles Open Open Open
]
Sheldews Position Open Open Open 1 Post Aec Position Shut Shut Shut Fwr Fall Position Asis /N/A Asis /N/A Asis /N/A j Cent.Isa Sig.(' O.O. 0.0. 0.0. l Cleeseellene (see) 30/N/A 30/N/A 30/N/A , Pwr Seem (Div) II/N/A I/N/A I/N/A 1 I
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ABWR- m62mn - Standard Plant arv. c G 1l 5.0% E E
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ABM 2asioorn REV.C Standard Plant
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See Chapter 14 for a thorough discussion of 4
.p-. plant variables in the conservative direction
(- - simultaneously. The results of this calculation preoperational testing for these systems. for the limiting case are given in Figure 63 67 through 63 75 and Table 63-4. The bounding PCT 6.3.4.2 Reliability Tests and laspections .. is greater than the Appendix K value. This is , typical of the first peak PCT values, which are The average reliability of a standby . 2 dominated by the amount of stored energy assumed, (nonoperating) safety system is a function of as documented in Reference 2. The core remains - the duration of the interval between periodic covered throughout the LOCA event, so there is no functional tests. The factors considered in ' I second peak PCT. Since the ABWR results have determining the periodic test interval of the-large margins to the 10CFR50.46 licensing ECCS are: (1).the desired system availability acceptance criteria, the ABWR licensing PCT can (average reliability); (2) the number of be based on the bounding PCT which is well below .' redundant functional system success paths; (3) the 22000F PCT limit. the failure rates of the individual' components in the system; and (4) the schedule of periodic ; 63.3J LOCA Analysis Conclusions tests (simultaneous versus uniformly staggered versus randomly staggered). Having shown compliance with the applicable acceptance criteria of Section 6.3.3.2, it is ' All~of the active components of the HPCF-concluded that the ECCS will perform its function System, ADS, RHR and RCIC Systems are designed - ; in an acceptable manner and meet all of the so that they may be tested during normal plant 10CFR50.46 acceptance criteria, given operation operation. Full flow test capability'is at or below the MAPLHGRs in Table 63 7. provided by a test line back to the suction ' source. The full flow test is~used to verify 6.3.4 Tests and Inspections the capacity of each ECCS pump loop while the plant remains undisturbed in the power i 6.3.4.1 ECCS Performance Tests generation mode, in addition, each individual g' valve may be tested during normal plant , i All systems of the ECCS are tested for their operation, operational ECCS function during the-preoperational and/or startup test program. Each All of the active components of the ADS .i component is tested for power source, range, System, except the safety /r'elief valves and direction of rotation, setpoint, limit switch their associated solenoid valves, are designed setting, torque switch setting, etc. Each pump so that they may be tested.during normal plant 1 is tested for flow capacity for comparison with operation. The SRVs and associated solenoid i vendor data. (This test is also used to verify valves are all tested during plant initial power l flow measuring capability). The flow tests ascension per Appendix A Paragraph D.2.c of. involve the same suction and discharge source Regulatory Guide 1.68. SRVs are bench tested to (i.e., suppression pool), establish lift settings. 4 1 All logic elements are tested individually and Testing of the initiating instrumentation and then as a system to verify complete system controls portion of the ECCS is discussed in response to emergency signals including the Subsection 73.1. The emergency power system, ability of valves to revert to the ECCS alignment which supplies electrical power to the ECCS in from other positions. the event ti at offsite power is unavailable, is tested as described in Subsection 83.1. The
- Finally, the entire system is tested for frequency of testing is specified in the Chapter response time and flow capacity taking suction 16 Technical Specifications. Visual inspections from its normal source and delivering flow into of all the ECCS components located outside the the reactor vessel. This last series of tests is drywell can be made at any time during power performed with power supplied from both offsite operation. Components inside the drywell can be l power and onsite emergency power. visually inspected only during periods of access Amendment 11 6.3-13 i
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- ABWR - m6i=^s Standard Plant REV.C =l
'I to the drywell. When the reactor vesselis open, 63.4.2.3 RHR Testing the spargers and other internals can be .
inspected. The RHR' pump and valves are tested. , periodically during reactor operation. With the l 6.3A.2.1 HPCFTesting injection valves closed and the return line open to the suppression pool, full flowing pump The HPCF can be tested at full flow withD capability is demonstrated.. The injection valve
. and the check valve are tested in a manner l suppression operation exceptpool when water at any a system time signal initiation during plant to that used for_the HPCF valves. The
.similar ~;
j is present.- If an initiation signal occurs while s system test conditions during reactor operation the HPCF is being tested, the system returns are shown on the RHR system process diagram automatically to the operating mode. The (Figure 6.3 3), motor operated valve in the line to the condensate storage system is interlocked closed 6.3A.2A RCIC Testing y I when the suction valve from the suppression pool ' . is open. The RCIC loop can be tested during reactor operation. To test the RCIC pump at rated flow, A design flow functional test of the HPCF over 'the test bypass line valve to the suppression the operating pressure and flow range is pool and the pump suction valve from the performed by pumping water from the suppression suppression pool are opened and the pump is pool back through the full flow test return line started using the turbine controls in the to the suppression pool, control room. Correct operation is determined 'i by observing the instruments in the control The suction valve from the condensate storage room. _, tank and the discharge valve to the reactor . i , remain closed. These two valves are tested If an initiation signal occurs during the separately to ensure their operability. - test, the RCIC system returns to the operating - - mode. The valves in the test bypass lines are The HPCF test conditions are tabulated on the- closed automatically and the RCIC pump discharge _; HPCF process flow diagram (Figure 6.31). valve is opened ~to assure flow is correctly routed to the vessel. ' 6.3A.2.2 ADS Testing 6.3.5 Instrumentation Requirements - l An ADS logic system functional test and simulated automatic operation of all ADS logic Design details including redundancy and logic channels are to be performed at least once pc. of the ECCS instrumentation are discussed in plant operating interval between reactor Section 7.3. refuelings. Instrumentation channels are demonstrated operable by the performance of a All instrumentation required for automatic
-channel functional test and.a trip. unit and manualinitiation of the HPCF, RCIC, RHR and calibration at least once per month and a ADS is discussed.in Subsection 7.3.1 and is transmitter calibtation at least once per designed to meet the ecquirements of IEEE 279L operating interval. and other applicable regulc3ry requirements.
The HPCF, RCIC, RHR and ADS can be manually All SRVs, which include those used for ADS are initiated from the cwrol room. bench tested to establish lift settings in i compliance with ASME Code Section .XI. The RCIC, HPCF, and RHR are automatically ~ initiated on low reactor water level or high-drywell pressure. The ADS is automatically actuated by sensed variables for reactor-vessel ' , O, Amendment 7 6.3-14 i
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23A6]M -' nev. c -l Standard Plant l SECTION 6.5 ; I CONTENTS Sectlon M P.88t 6.51 i 6.5.1 Engineered Safety Features Filter Systems
.i 6.5.1.1 Design Bases 651 6.5.1.1.1 Power Generation Design Bases 6.5 6.5.1.1.2 Safety Design Bases . 6.51 6.5.1.2 System Design 6.51 6.5.1.2.1 Ocneral 651' ,
6.5.1.2.2 Component Description 651 j 6.5.1.23 SGTS Operation 6.51 i 6.5.1.23.1 Automatic 651-6,5.1.23.2 Manual 6.5-2
.i g O 6.5.1.233 Decay Heat Removal . 652' l
'6.5.13 Design Evaluation -6.52 6.5.13.1 General 6.52 6.5.13.2 Sizing Basis 6.5-2 6.5.13 3 Justification for Single SGTS Filter Train . 6.5-3 6.5.13.4 Source Terms for SGTS Design 6.5-4.1 6.5.13.5 Compliance with Regulatory Guide 1.52 6.54.1 6.5.13.6 Primary Containment Purging 654.1 6.5.1.4 Test and Inspection- 6.5-4.1 l
O l l
- Amendment 11
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~1 MM 23A6100AB
' Standard Plant nev. c - .
l SECTION 6.5 g CONTENTS (CONTINUED). . Section M East ; 6.5.1.5 Instrumentation -6.54.1 . 6.5.1.6 Materials - 6.5-4.1 63.2 Containment Spray Systems 6.5-5 6JJ Fission Product Control Systems 6.55 6.53.1 . Primary Containment - 6.5 6.53.2 Secondary Containment 6.54 6.5.4 lee Condenser As A Fission Product
- Control System ' 6.5-6 i
TABLES: Table Iult East 6.5-1 Summary of Major Standby Gas Treatment - . System Components - 6.57-6.52 Source Terms Used For SGTS _ i Charcoal Adsorber Design 5.5-11 i ILLUSTRATIONS Figure . Iult ' Eagt t 6.51 Standby Gas Treatment System PI&D 6.5 13 J i-6.52 Secondary Containment Pressure Transient , After the Design Bases LOCA 6.5-14 { i 6.5 lii Amendment 11
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ABM :sasioars arv. c Standard Plant 6.5 FISSION PRODUCTS REMOVAL AND (4) Remain intact and functional in the event of ( CONTROL S'* STEMS a safe shutdown earthquake (SSE). d l- 6.5.1 Engineered Safety Features Filter (5) Meet environmental qualification l l Systems requirements established for system l l operation.- The _ filter systems required to perform , safety.related functions following a design basis I accident are: -l
- (1) Standby gas treatment system (T22 SGTS). -l l
(2) Control room portion of the HVAC system. : (U41.HVAC) . 63.1.2 System Design The control room portion of the HVAC system is - , discussed in Section 6,4 and Subsection 9.4.1. 6J.1.2.1. General , The SGTS is discussed in this Subsection (6.5.1). . The SGTS P&lD is provided as Figure 6.5-1. I 6.5.1.1 Design Basis 6.5.1.2.2 Component Description 6.5.1.1.1 Power Generation Design Basis > Table 6.5-1 provides a summary of the major j The SGTS has the capability to filter the SGTS components. The SGTS consists of two gaseous effluent from the primary containment or parallel and redundant trains of active from the secondary containment when required to equipment which share a single filter train. limit the discharge of radioactivity to the Suction is taken from above the refueling a.rea Os environment to meet 10CFR100 requirements, or from the primary containment via the atmospheric control system (T31 ACS). The 63.1.1.2 Safety Design Basis discharge goes to the main plant stack. l- The SGTS is designed to accomplish the The SGTS consists of the following principal ! following: - components: (1) Maintain a negative pressure in the (1) Two independent dryer trains consisting of a secondary contairment, relative to the molFlute separator and an electric process outdoor atmosphere, to control the release heater. I l of fission products to the environment.
-(2) Two independent process fans located upstream of the filter train.
l (3) A filter train consisting of a prefilter, a j (2) Filter airborne radioactivity (halogen and high efficiency particulate air (HEPA) , air particulates) in the effluent to reduce filter, a charcoal adsorber, a second HEPA l offsite doses to within the limits specified filter, and space heaters, i in 10CFR100. l 6.5.1.2.3 SGTS Operation 6.5.1.2.3.1 Automatic (3) Ensure that failure of any active component, I assuming loss of offsite power, cannot Upon the receipt of a high primary impair the ability of the system to perform containment signal or a low reactor water level
- , its safety function. signal, or when high radioactivity is detected l '- in the secondary containment or refueling floor l l
Amendment 11 6,5-1 l 4
ABWR - 2-= l l- Standard. Plant _ no e , ventilation exhaust, the SGTS is automatically (2) The SGTS filter particulate and charcoal actuated. If system operation is not confirmed, the ' efficiencies are outlined in Table 6.51. Dose - , redundant process fan and dryer train are analyses of events requiring SGTS operation, - automatically placed into service. In the event a described in Subsections 15.6.5 and 15.7.4, ; , malfunction disables an operating process fan or indicate that offsite doses are within the limits i l dryer train, the standby process fan and dryer train established by10 CFR 100. are manuallyinitiated. ' (3) The SGTS is designated as an engineered 63.1.2.3.2 Manual safety feature since it mitigates the consequences of a postulated accident by Tbc SGTS is on standby during normal plant controlling and reducing the release of ; operation and may be manually initiated before or radioactivity to the environment. The SGTS, during primary containment purging (de inerting) except for the deluge, is designed and built to when required to limit the discharge of contaminants the requirements for Safety Class 3 equipment to the environment. . It may be manually initiated as defined in Section 3.2, and 10 CFR 50, 1 whenever its use may be needed to avoid exceeding Appendix B. radiation monitor setpoints. The SGTS has independent, redundant active. 6.5.1.2.33 Decay Heat Removal components. Should any active component fail, SGTS functions can be performed by the Cooling of the SGTS filters may be required to redundant component. The electrical devices prevent the gradual accumulation of decay heat in ofindependent components are powered from the charcoal. This heat is generated by the decay of separate Class 1E electrical buses, radioactive iodine adsorbed on the SGTS charcoal. The charcoalis typically cooled by the air from the (4) The SGTS is designed to Seismic Category I process fan. requirements as specified in Section 3.2. The SGTS is housed in a Category I structure. All A water deluge capability is also provided, but surrounding equipment, components, and primarily for fire protection since redundant process supports are designed to appropriate safety fans are provided for air cooling. Since the deluge is class and sci *mic requirements. available, it may also be used to remove decay heat for sequences outside the normal design basis. (5) The SGTS design is based on the maximum Temperature instrumentation is provided for control pressure and differential pressure, maximum - of the SGTS process and space electric heaters. This integrated dose rate, maximum relative L instrumentation may also be used by the operator to humidity, and maximum temperature expected j [re-] establish a cooling air flow post accident,if in secondary containment for the LOCA event. I required. 63.1.3.2 Sizing Basis Water is supplied from the fire protection system and is connected to the SGTS via a speol piece. Figure 6.5 2 provides an assessment of the secondary containment pressure after the 6.5.13 Design Evaluation design basis LOCA assuming an SGTS fan capacity _ of 4000 scfm (70 F,1 atmosphere) per fan. C:cdit 6.F.13.1 General for secondary containment as a fission product control system is only taken if the secondary (1) A slight negative pressure is normally containment is actually at a negative pressure by ; maintained in the secondary containment by considering the potential effect of wind on the . the reactor building HVAC system (Subsection ambient pressure in the vicinity of the reactor
~
9.4.5). On SGTS initiation per Subsection building. For the ABWR dose analysis,~ direct 6.5.1.2.3.1, the secondary containment is transport of containment leakage to the environment automatically isolated from the HVAC system. was assumed for the first 20 minutes after LOCA event initiation (in addition to the leakage through , the MSIVs to the main turbine condenser). Each SGTS fan was sized to establish a continuously ' negative differential pressure (considering the Amendment 11 6.5-2 n
ABM 23^6ioors nu c Standard Plant (1) The advanced design of the filter housing and i f^ effect of wind) within 10 minutes of SGTS initiation. The dose analysis therefore assumes direct leakage flow pattern virtually climinates any untreated from the containment to the environs for twiec the. bypass of the filter. In addition, the all welded required period. In addition,it should be recognized ' design is such that degradation of filter housing that fission product release on the order of that integrity is not likely to occur during system !- specified in Regulatory Guide 1.3 and used in the standby or operation. LOCA dose analyses (Subsection 15.6.5) realistically requires significant core damage and most likely (2) Sufficient instances of inadvertent deluge i more than 10 or 20 minutes for transport to and wetting the charcoal and rendering the filter leakage from the containment. train unavaile' - have been observed to warrant an imr" d deluge design concept. The calculation accounted for all expected heat These unintendea deluge operations have been sources in secondary containment after a LOCA. caused by personnel error and by failures in Where appropriately conservative, a realistic basis mechanical or electrical components. In the was used to determine the heat loads. For example, ABWR design, the deluge piping is not no single failure of a diesel was assumed since it is connected permanently from the fire protection . most likely all divisions of power would be available. . system to the filter housing nozzle, Instead, a Failure of one SGTS fan to start was assumed as the normally disconnected hose from the fire single failure. Therefore, heat loads from all protection system is provided to act as a " spool divisions of ECCS motors and piping were used in piece' for connection by operating personnel to the calculation, the filter housing, as required. Per SRP 6.2.3, II.3(b) and SRP 6.5.3, II.2, (3) Decay heat is not sufficient to cause a fire in secondary containment should be held below 0.25 the charcoal adsorber or HEPA filter. Calcula-I inch w.g. under all wind conditions up to the wind tions indicate that air flow from either speed at which diffusion becomes great enough to redundant process fan is more than enough to ; A assure site boundary exposures less than those remove the heat from decay of the radioactive V calculated for design basis accidents, even if ex filtration occurs (i.e., no credit for SGTS is lodine on the charcoal or filters. Heating does not occur sufficient to cause iodine desorption
- l. taken). For the ABWR, dispersion factors were ' or ignition of the charcoal. With the reduced l calculated for each stability class over a range of. . source term expected for most sequences (see wind speeds. Above 8,0 m/s, stability class D _ Subsection 6.5.1.3.3(4)), any heating of the predominates and conservatively bounds observed charcoalis even further reduced. No other meteorological conditions, At 8.9 m/s, above the 8.0 mechanism for starting a fire in the filter .
, m/s stability class D transition, the dispersion from housing during an accident has been identified. I the increased wind speed results in offsite doses Other possible sequences for starting a fire in equal to or lower than the design basis calculation, the filter train could occur during normal plant which assume the most stable, F-class stability and a operation or plant shutdown. These sequences 1 m/s wind speed. Therefore, the ABWR SGTS was would involve an unspecified maintenance or designed to establish and maintain a negative operating personnel activity or an incredible l pressure in secondary containment within 10 minutes malfunction of the space heaters. In this case a i for any wind speed up to and including 8.9 m/s (20 fire in the SGTS charcoal, like in the offgas l mile /hr). system, would be a matter of plant availability . and not of plant safety. The space heaters, 6.5.1.3.3 Justification for Single SGTS Filter Train located inside the SGTS filter housing, are . ,. powered only during SGTS standby and not l- The SGTS filter train, consisting of a pre filter, two during system operation. Therefore, the space HEPA filters, and an iodine adsorber, is considered heaters are not a potential cause of fire (and l passive, and in practice provides the reliability SGTS unavailability) when the SGTS is t associated with a passive component. Furthermore, required to meet the licensing basis release j the ABWR SGTS has incorporated design features limits (and presumably inaccessible for repair). l to eliminate potential failures or improper operation. These features include: Amendment 11 653 l- {
ABM 23A6100AB' Standard Plant am e . Note that the space heaters each have a small nominally requires 1750 lbs of charcoal based fan which better distributes the heat and . on a 4000 scfm fan size, meeting the 0.25 sec minimizes local warming by providing a more per 2 inch of bed depth (40 fpm) requirement uniform temperature throughout the filter of R.G.1.52 (Position Cy.1), and using a - housing. This uniform beating further reduces conservatively high 35 lb/ft charcoal density. the risk of fire by lowering local temperatures The weight of charcoal will be adjusted to be around the space heater and byimproving the consistent with the puregased charcoal density accuracy of the temperature measurements (usually less than 30 lb/ft ) and any dead space 1 (used to detect high temperature) taken at in the adsorber section itself. necessarily discrete points within the filter housing. The effect of suppression pool scrubbing, per , SRP 6.5.5, also serves to reduce the actual (4) Degradation of the charcoal effectiveness source term, providing capacity margin over the between charcoal efficiency surveillance tests is design basis calculation. Reasonable scrubbing not likely to occur During normal operation, factors of just 10 for elemental and particulate the filter is isolated, and valves upstream and lodine results in only 100 lbs of charcoal being . downstream of the filter train are closed. . required versus the nominal 1750 lbs provided. Therefore, during SGTS standby the potential This margin between the charcoal realistically for impurities entering the filter train and required and that needed per the design basis unacceptably reducing charcoal efficiency is provides additional protection against any aging small, or weathering that may occur. The retention of iodine in the suppression poolis discussed in The SGTS may be used either for a NUREG 0772 and NUREG 1169, which design basis accident identified in Chapter 15 established the basis for the ABWR design [ or during de inerting of the primary under Paragraph 8.9 of the Licensing Review containment prior to plant shutdown. The Basis, more likely, though still infrequent, potential . . , use of SGTS is during de inerting. Depending - (5) Because of the high availability of the ABWR, on indications from leak detection and isolation de.inerting, and the potential use of SGTS ' i system (E31 LDS) primary containment during de-nerting will occur primarily at the radiation monitoring before de inerting is end of the fuel cycle. In this way, HEPA filter initiated or from the process radiation and charcoal adsorber effectiveness will be monitoring (D11 PRM) reactor building tested, and the filter and/or charcoal replaced, ventilation exhaust radiation monitors during if necessary, before the plant returns to power + 1 de inerting, SGTS may be placed into service, operation. l The ABWR SGTS charcoal bed thickness has been increased two inches, to six inches, as With these SGTS design features,long term, i compared to the GESSAR 11 design. The undetected, passive failure of the filter train has been l additional two inches of charcoal provide an minimized. Therefore, one filter train will be effective measure of protection against adequate to assure that the SGTS is available to weathering or aging effects when the SGTS is perform its required safety function. It is recognized placed into operation. that 10CFR50, Appendix A, GDC 43, cites " filters
- as ,
an example of an active component. General i In addition to the increased charcoal bed depth, Electric considers an active component to be a i significantly more charcoal is provided than is component in which mechanical movement must-required to meet the 2.5 mg. iodine per gram occur to accomplish the nuclear safety function of . carbon requirement. This added charcoalis the component. Therefore, a filter would be a > used to meet the requirement specifying a passive component and in fact provides the reliability residence time of 0.25 sec per 2 inches of bed ' associated with a passive component. depth. Approximately 732 lbs of charcoal are required based on iodine loading calculated per All active SGTS components are redundant. Two Regulatory Guide 1.3 requirements, a 100% redundant dryer trains, each containing a moisture l cificient charcoal adsorber, and no MSIV separator and process heater, and two exhaust fans leakage. The SGTS charcoal adsorber are provided. The non-safety space heaters are also i Amendment 11 654 4
4 ABWR . nanoorn Standard Plant nvc p provided with spares, with heaters from each division ventilation valves are open (valves expected to be
! both upstream and downstream of the charcoal bed, open only during inerting or de inerting), little
/
Divisional power is used for reliable space heater fission product release to the environment would operation. actually occur. Therefore, the plant design and-analysis in this regard is conservative and bounds 63.13.4 Source Terms for SGTS Design releases actually expected in the event of a LOCA. The basis for calculating the iodine source term for 6J.1.4 Tests and Inspection the SGTS filters is provided in Table 6.5-2. For the purposes of sizing the SGTS charc'oal adsorber, no The SGTS and its components are periodically additional credit for iodine retention or holdup tested during construction and operation. These above that specified in Regulatory Guide 1.3 is tests fallin three categories: assumed. Charcoal sizing is discussed in Subsection ' 6.5.133(4),' Justification for Single SGTS Filter (1) Environmental qualification tests Train? ' (2) Acceptance tests as d: lined in ANSI N509 and 6J.1.33 Compilance with Regulaton Guide 132 N510 An assessment of compliance with Section C of .(3) Periodic surveillance tests. Regulatory Guide 1.52, including testing, is provided . in Appendix 6A. The above tests are performed in accordance with the objectives of Regulatory Guide 1.52 and its 63.1.3.6 Primary Containment Purging teferences. Acceptance tests (including pre operational tests) and periodic surveillance tests If purging (i.e., de inerting) through the HVAC are defined and extensively described in ANSI N509 will[or does) result in a trip from the ventilation and ANSI N510. Testing requirements in ANSI exhaust radiation monitors, then de inerting will be N509 are generally located lin Section 5, ( [re-] initiated at a reduced rate through the SGTS. ' Components? ANSI N510 provides details of each Use of SGTS during de-inerting is expected to be type of functional test. These tests are summarized infrequent. in Table 91 of ANSI N509 and Table 1 of ANSI N510. Specific surveillance testing requirements for The LOCA dose analyses do not assume any SGTS are provided in technical specification 3.6.43 release from open containment isolation valves, (Chapter 16). Environmental qualification testing is + cither through the SGTS or through the normal discussed in Section 3.11 and is applicable to SGTS ventilation system. The design basis condition for components. the relevant dose analyses assume the large ventilation valves are closed, because the probability 63.13 Instrumentation of a LOCA occurring at the same time the ventilation valves are open is very small. The large Appendix 6B provides a discussion of the ventilation valves are in fact closed throughout instrumentation for SGTS. Control and normal plant operation except during inerting and instrumentation for the SGTS is also discussed in de.inerting. Subsections 73.L1.5 and 73.2.5. A realistic assessment of plant capability in 63.1.6 Materials support of the exclusion indicates that the ventilation valves,if open, would be isolated before significant The construction materials used for the SGTS are fission products are transported to the containment. compatible with normal and accident environments ,
"Significant* means fission products above that postulated for the area in which the equipment is normally present in the primary system. A period located. The construction materials used in the much longer than the closing time of the ventilation dryer and filter trains are consistent with the valves would be required to generate conditions recommendations of Regulatory Guide 1.52 and its leading to the release of TID 14844 like source references.
terms. Therefore, should a LOCA occur when the (_ Amendment It 6.541 l l l' .,
23A6100AB l Standard Plant m- c - 63.2 Containment Spray Systems program confirm the integrity of the leakage
=t boundary. The assumed leak rate from primary A Credit is not taken for any fission product containment is 0.5% of the free containment removal provided by the drywel' and wetwell spray volume per day measured at the containment portions of the RHR system. design pressure.
Containment leak rate testing is described in Subsection 6.2.6. The primary containment walls, liner plate, mechanical penetrations, isolation valves, hatches,'and locks function to limit release of radioactive materials, subsequent to postulated accidents, such that the resulting offsite doses are less than the guideline values of10CFR100. The structural design details of the primary containment are discussed in Subsection 3.8.2. Primary containment isolation valves are-discussed in Subsection 6.2.4. The conditions in the containment during and after the <tesign basis events are given in Section 6.2. Layouts of the primary containment 6.5.3 Fission Product Control Systems structure are given in the building arrangement - drawings in Section 1.2. Fission product control systems are provided g l in conjunction with other ESF systems to limit the The primary containment atmosphere is t release of radioactive material from the inerted with nitrogen by the atmospheric control containment to the environment following ' system (ACS), The ACS is described in postulated design basis events. Dose analyses are Subsection 6.2.5. Following the design basis provided in Chapter 15. ' The fission product LOCA, the flammability control system (FCS) control systems consist of the primary containment controts_ the concentratio'n of oxygen in and the secondary containment. The following is a containment. Oxygen is gencrated by the discussion of each fission product control system. radiolytic decomposition of water, 633.1 Primary Containment 'On appropriate signals, containment isolation - valves close as required. The primary The primary containment is a cylindrical containment provides a passive barrier to limit the ' steel. lined reinforced concrete structure forming leakage of airborne radioactive material. Systems a limited leakage boundary for fission products- required to accomplish ECCS or other ESF ' released to the containment atmosphere following. functions are not isolated. See Subsection 6.23 a LOCA or other event. The containment is for further details of isolation valve closure ' divided into the upper and lower drywells and the ' signals. , suppression chamber (wetwell) by the reinforced ! concrete diaphragm floor and the reactor vessel 6.53.2 Secondary Containment pedestal. The diaphragm floor is rigidly attached to the reactor pedestal and the containment wall. The secondary containment is provided so that , A liner is also provided as part of the diaphragm leakage from the primary containment is collected floor to prevent bypass of steam from the upper and treated and monitored by the SGTS prior to drywell to the suppression chamber air space release to the environment. Refer to Subsectian during an accident. The primary containment is 6.2.3 for a description of the secondary i totally within the secondary containment. A test containment boundary and Subsection 6.5.1 for a 4 description of the SGTS. V ' Amendment 11 6.5-5 4
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1 l 6.5A Ice Condenser as a Fission Product. [ L - Control System l The GE ABWR does not utilize any kind of an $ ice condenser feature as a fission product . control system. i O r i i i s O Amendment 11 6.54:
k 23A6100AB E Standard Plant nu e TABLE 6.51. STANDBY GAS TREATMENT SYSTEM COMPONENT DESCRIPTION Drver Train - General Consists of a moisture separator and process heater . Quantity 2 Capacity 4000 scfm each t Dner Train . Motsture Separator - General - Woven wire,' stainless steel mesh pads - Ouantity 1 bank 'of standard size moisture separators - Efficiency per ANSI N509, Section 5,4 Dryer Train Electric Process Heater General Electric, finned tubular - 1 ..
/' Ouantity 1 per dryer train E Rating 53 kW minimum,26.2 kW maximum Relative humidity Inlet 100 %
l Outlet 70% - Air AT 15 F-Process Fan , General - Centrifugal Quantity 2 Capacity 4000 scfm each; l Filter Traig General Consists of pre HEPA filter, HEPA filters, charcoaladsorber , l-l Amendmen: 11 63-7
i
. 23A6100AB '
- Standard Plant new c TABLE 6.51 (Continued) l l 1 STANDBY GAS TREATMENT SYSTEM:
. COMPONENT DESCRIPTION- 1 l.
Quantity 1 Capacity 4000 scfm l.
' Filter Train Prefilter y . General Cartridge type
! Quantity 1 bank of stand size fihers Media Glass fiber l Efficiency per ANSI N509-1980, Section 53 Filter Train HEPA Filters > General Vertically oriented Quardity Bank of standardisize HEPA filters both ' upstream and downstream of charcoal
~
adsorber ' l l- Media Glass fiber l Efficiency 2.99.97% with 03 micron DOP (shop test) ; 2.99.9% with 0.5 micron DOP (surveillance test) , FilterTrain Charcoal Adsorbers General . Vertically oriented deep beds Quantity 1 Efficiency 2,99.825% (laboratory) .
.199.95% (in place bypass test)
Charcoal weight. 1750 lbs (nominalweight) f Depth of Bed 6 inches i Maximum Face velocity 40 fpm ]! Amendment 11 63 8
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b 23A6100AB Standard Plant mvc TABLE 6.5-2 SOURCE TERMS USED FOR SGTS CHARCOAL ADSORBER DESIGN Source term assumed available for leakage from containment (Regulatory Guide 1.3): o 100% of noble gases from fuelinventory o 25% ofiodine from fuelinventory Chemical form of iodine assumed available for leakage from primary containment: o 4% organics o 91% elemental o 5% particulates Suppression pool lodine decontamination factor used in calculation: o 1 for organics o I for elemental
- o. 1 for particulates Containment spray iodine decontamination factor used in calculation:
o 1 for organics o i for elemental o 1 for particulates Leakage Rates assumed for calculation: o 0.50%/ day for primary containment o 50%/ day for secondary containment o 0 scfh through MSIVs Amendment 11 6.5-11
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'ABWR 234sioorn Standard Plant n- c l APPENDIX 6A 1
.O REGULATORY GUIDE 1.52, SECTION C, COMPLIANCE ASSESSMENT
.s This Appendix provides the compliance status of the ABWR SGTS design with each of the regulatory positions specified under Section C of R.G.1.52, and the revision cited in Table 1.8-20. Following each l provision of R.G.1.52 is an evaluation of the ABWR compliance with that position. If the ABWR deviates ~)
from the R.G.1.52 position, justification is provided. Note that the similarly numbered sections from the revisions cited in Table 1.8-21 for ANSI N509 and N510 are used for ABWR SGTS design except as otherwise .j moted; Regulatory Guide 1.52 references older revisions (1976) of these standards. Compliance as described in ! the remainder of this response is mesured against the applicable section of the standards referenced in Table I 1.8-21.. q i In addition, the term "demister,' used in R.G.1.52, is a trademark of Otto H. York Co., Inc. of Parsippany, New Jersey. The ABWR SGTS design includes provision for the use of moisture separators. s ABWR Comnliance with R.G.151 Revisinn 1 Section C
*1.' ENVIRONMENTAL DESIGN CRITERIA i
- a. The design of an engineered safety feature atmosphere cleanup system should be based on the maximum pressure differential, radiation dose rate, relative humidity, maximum and minimum temperature, and
=
other conditions resulting from the postulated DBA and on the duration of such conditions.' j The design is in compliance with this position.
)
*b. The design of each ESF system should be based on the radiation dose to essential services in the vicinity of the adsorber section, integrated over the 30-day period following the postulated DBA~. The radiation j source term should be consistent with the assumptions found in Regulatory Guides 1.3,1.4 and 1.25. l Other engineered safety features, including pertinent components of casential services such as power, air, and control cables, should be adequately shielded from the ESF atmosphere cleanup systems " -
l The design is in ec.mpliance with this position. Table 313 20 provides the radiation environmental conditions , inside secondary containment for plant abnormal and accident conditions. Note that integrated doses for als i months, not 30 days, are provided in Table 313-20. ! t 4
"c. The design of each adsorber should be based on the concentration and relative abundance of the iodine species (elemental, particulate, and organic), which should be consistent with the assumptions found in Regulatory Guides 13,1.4 and 1.25."
i The design is in compliance with this position. A revised Table 6.5 2 is provided. i i i "d. The operation of any ESF atmosphere cleanup system should not deleteriously affect the operation cf other engineered safety features such as a containment spray system, nor should the operation of other i engineered safety features such as a containment spray system deleteriously affect the operation of any l ESF atmosphere cleanup system." i The design is in compliance with this position. Amendment 11 6A 1 l i
_ _ ____ ___ . _ _. ~ _ . M - 23A6100AB Standard Plant we i
'ci Components of systems connected to compartments that are unheated during a postulated accident abould be designed for post accident effects of both the lowest and highest predicted temperatures?
The design is in compliance with this position.
"2.- SYSTEM DESIGN CRITERIA
)
i ESF atmosphere cleanup systems designed and lastalled for the purpose of mitigating accident doses : a. should be redundant. The systems should consist of the following sequential components: (1) demisters, (2) prefilters (demisters may serve this function), (3) HEPA filters before the adsorbers, (4) iodine adsorbers (impregnated activated carbon or equivalent adsorbent such as metal zeolites), (5) HEPA > 1.. -
. filters after the adsorbers, (6) ducts and valves, (7) fans, and (8) related instrumentation. Heaters er
, cooling coils used in conjunction with heaters should be used when the humidity is to be controlled before fikration? The design is in compliance with this position, with the exception of filters. The prefiker, pre HEPA filter, and , iodine adsorber are considered passive for the purposes of design. Therefore, a single filter train is provided to i . perform SGTS functions. Justification for the single filter train is provided in Subsection 6.5.1.3.3,
" Justification for Single SGTS Filter Train? The moisture separators, heaters, and all other active components are redundant.
i
'b. The redundant ESF atmosphere cleanup systems should be physically separated so that damage to one
- system does not also cause damage to the second system. The generation of missiles from high pressure l equipment rupture, rotating machinery failure, or natural phenoeena should be considered in the design
! for separation and protection? .- c l l~ The design is in compliance with $is position, k L { l
'c. All components of an engineered safety feature atmosphere cleanup system should be designated as Seismic Category I (see Regulatory Guide 1.29) if failure of a component would lead to the release of significant quantities of fission products to the working or outdoor environments? ,
The design is in compliance with this position. l l
'd. If the ESF atmosphere cleanup system is subject to pressure surges resulting from the postulated accident, the system should be protected from such surges. Each component should be protected with such devices as pressure relief valves so that the overall system will perform its intended function during and after the passage of the pressure surge?
The design is in compliance with this position. The ABWR SGTS is not subject to pressure ' surges
- from the postulated accident sufficient to cause damage to the filter train. Secondary containment pressure does' ;
increase slightly as part of the post LOCA heatup process. I
'e. In the mechanical design of the ESF system, the high radiation levels that may be associated with buildul- '1 '
of radioactive materials on the ESF system components should be given particular consideration. ESF system construction materials should effectively perform their intended function under the postulated radiation levels. The effects of radiation should be considered not only for the demisters, heaters, HEPA filters, adsorbers, and fans, but also for any electrical insulation, controls, joining compounds, dampers, i.
~
Amendment 11 6A4 I l
ABWR sasioors Standard Plant am c gaskets, and other organic containing materials that are necessary for operating during a postulated l DBA.* Y. %e desip is in compliance with this position. ! 3
*f. i The volumetric alt flow rate of a t ngle cleanup train should be limited to approximately 30,000 ft / min.
If a total system air flow in excess of this rate is required, multiple trains should be used. For case of . maintenance, a filter layout three HEPA filters high and ten wide is preferred.' Tbc design is in compliance with this position.
'g. The ESF atmosphere cleanup system should be instrumented to signal, alarm, and record pertinent pressure drops and flow rates at the control room.'
Tbc design is in compliance with this position. Filter train exhaust flow and reactor building differential pressure are indicated and appropriately annunciated in the main control room. Pertinent pressure drops ! across the individual components of the dryer units and filter train filters are indicated at a local rack and are used during testing. l
'h. The power supply and electrical distribution system for the ESF atmosphere cleanup system described in Section C.2.a above [one that is used to mitigate accident doses) should be designed in accordance with Regulatory Guide 1.32. Allinstrumentation and equipment controls should be designed to IEEE Standard 279. The ESF system should be qualified and tested under Regulatory Guide 1.89.. To the extent applicable, Regulatory Guides 1.30,1.100, and 1.118 and IEEE Standard 334 should be considered in the design."
The design is in compliance with this position. Commitments for all except IEEE 334 are provided in Chapters 7,8 and 11 and Sections 3.10 and 3.11. IEEE 334 is applied to the SGTS per this Regulatory Guide. l l 'i. Unless the applicable engineered safety feature atmosphere cleanup system operates continuously during all times that a DBA can be postulated to occur, the system should be automatically activated upon the occurrence of a DBA by (1) a redundant engineered-safety. feature signal (i.e., temperature, pressure) or - (2) a signal from redundant Seismic CategoryI radiation monitors." The design is in compliance with this position.
'j. To maintain radiation exposures to operating prsonnel as low as is reasonably achievable during plant maintenance, ESF atmosphere cleanup system should be designed to controlleakage and facilitate maintenance in accordance with the guidelines of Regulatory Guide 8.8. The ESF atmosphere cleanup train should be totally enclosed. Each train should be designed and installed in a manner that permits .
replacement of the train as an intact unit or as a minimum number of segmented sections without removal ofindividual components.' The design is in compliance with this position.
*k. Outdoor air intake openings should be equipped with louvers, grills, screens, or similar protective devices to minimize the effects of high winds, rain, snow, ice, trash, and other containments on the operation of \
Amendment 11 6A.3 l l, - ,
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i 1 ABM :sasioox. Standard Plant m.m c
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the system. If the atmosphere surrounding the plant could contain significant environmental eantaminants, such as dusts and residues from smoke cleanup systems from adjacent coat burning power ; plants or industry, the _ design of the system should consider these contaminants and prewat them from i affecting the operation of any ESF atmosphere cleanup system " 3
- The ABWR SOTS has no outdoor alt intakes, taking suction only from within secondary containment.
Secondary containment air is filtered by the HVAC system (U41).
*l. ESF atmosphere cleanup system housings and ductwork should be designed to exhibit on test a maximum total leakage rate as defined in Section 4.12 of ANSI N5091976. Duct and housing leak tests should bc ,
perforood in accordance with the provisions of Section 6 of ANSI N5101975? ' The design is in compliance with this position.
'3. COMPONE!fr DESIGN CRITERIA AND QUALIFICATION TESTING
- a. Demisters should be designed, constructed, and tested in accordance with the requirements of Section 5.4 of ANSI N5091976. Demisters should meet Underwriters' Laboratories (UL) Class 1 requirements?
The design is in compliance with this position.-
'b. Air heaters should be designed, constructed, and tested in accordance with the requirements of Section 5.5 of ANSI N5091976?
The design is in compliance with this position.
'c. Materials used in the prefilters should withstand the radiation levels and environmental conditions . ,
prevalent during the postulated DBA. Prefilters should be designed, constructed, and tested in accordance with the provisions of Section 53 of ANSI N5091976? The design is in compliance with this position.
*d. The HEPA filters should be designed, constructed, and tested in accordance with Section 5.1 of ANSI N509-1976. Each HEPA filter should be tested for penetration of dioctyl phtblate (DOP) in accordance with the provisions of Mil F 51068 and MIL STD 282? }
The design is in compliance with this position. The applicable portion of MIL F 51068 is Section 3.4.1. The applicable portions of Mil-STD 282 are Methods 102.1,102.8 and 102.9.1. [
*c. Filter and adsorber mounting frames should be constructed and designed in accordance with the
. provisions of Section 5.63 of ANSI N5091976? t The design is in compliance whh this position. Amendment 11 6A-4 s. t
I MM 23A6100AB we Standard Plant - q l O "f. Filter and adsorber banks should be arranged in accordance with the recommendations of Section 4.4 of ERDA 76-21.' The design is in compliance with this position. ;
'l
'g. System filter housings, including floors and doors, should be constructed and designed in accordance with l the provisions of Section 5.6 of ANSI N5091976 '
The design is in compliance with this position, i
'h. Water drains should be designed in accordance with the recommendations of Section 4.5.8 of ERDA 76-21."
The design is in compliance with this position.
'i. ' The adsorber section of the ESF atmosphere cleanup system may contain any adsorbent material demonstrated to remove gaseous iodine (elementaliodine and organic lodides) from air at the required efficiency. Since impregnated activated carbon is commonly used, only this adsorbent is discussed in this guide."
Impregnated activated carbon is used in the ABWR SGTS design. >
"Each original or replacement batch of impregnated activated carbon used in the adsorber section should meet the qualification and batch test results summarized in Table 5.1 of ANSI N5091976. In this tabic, a ,
' qualification test'should be interpreted to mean a test that establishes the suitability of a product for a general application, normally a one. time test reflecting historical typical performance of material. In this table, a ' batch test'should be interpreted to mean a test made on a production batch of product to establish suitability for a specific application. A ' batch of activated carbon'should be interpreted to mean a quantity of material of the same grade, type, and series that has been homogenized to exhibit, within reasonable tolerance, the same performance and physical characteristics and for which the manufacturer can demonstrate by acceptable tests and quality control practices such uniformity.'
The test requirements for the adsorber section will comply with this position.
'All material in the same batch should be activated, impregnated, and otherwise treated under the same i process conditions and procedures in the same process equipment and should be produced under the same manufacturing release and instructions. Material produced in the same charge of batch equipment ,
constitutes a batch; material produced in different charges of the same batch equipment should be inclugd in the same batch only ifit can be homogenized as above. The maximum batch size should be 350 ft of active carbon." The test requirements will comply with this position. t "If an adsorbent other than impregnated activated carbon is proposed or if the mesh size distribution is different from the specification in Table 5.1 of ANSI N5091976, the proposed adsorbent should have O demonstrated the capability to perform as well as or better than activated carbon in satisfying the V specifications in Table 5.1 of ANSI N5091976." Amendment 11 6A 5
+v , - w ~ ,
ABWR s wioorn a- c hndard Plant , s . ; imprepated activated carbon is used in the ABWR SGTS design. The performance requirements of Table ! 51 of ANSI N509 will be met.
'If imprepated activated carbon is used as the adsorbent, the adsorber system should be designed for an average atmosphere residence time of 0.25 sec per two inches of adsorbent bed. The adsorption unit should be designed for a maximum loading of 2.5 mg of total lodine (radioactive plus stable) per gram of activated carbon. No more than 5% of imprepant (50 mg of imprepant per gram of carbon) should be ,
used. The radiation stability of the type of carbon specified should be demonstrated and certified (see Section C.1.b of this guide for the design source term)." ne design is in compliance with this position. l
'j. Adsorbir cells should be designed, constructed, and tested in accordance with the requirements of Section 5.2 of ANSI N5091976?
The design is in compliance with this position.
*k. The design of the adsorber section should consider possible iodine desorption and adsorbent auto ignition that may result from radioactivity. induced heat in the adsorbent and concomitant
. temperature rise. Acceptable designs include a low flow air bleed system, cooling coils, water sprays for the adsorber section, or other cooling mechanisms. Any cooling mechanism should satisfy the
- single fallu
- e criterion. A' low flow air bleed system should satisfy the single failure criterion for j providing low. humidity (less than 70% relative humidity) coolir.g air flow.' 1 TLe design is in compliance with this position. The design utilizes redundant process fans for any necessary l cooling of the charcoal. ;
l I
- *l. The system fan,its mounting, and the ductwork connections should be designed, constructed, and tested in accordance with the requirements of Sectious 5.7 and 5.8 of ANSI N5091976."
The design is in compliance with this position. l l
'm. The fan or blower used on the ESF atmosphe're cleanup system should be capable of operating under the environmental conditions postulated, including radiation.'
The design is in compliance with this position.
*n. - Ductwork should be designed, constructed, and tested in accordance with the provisions of Section 5.10 of l ANSI N5091976."
l The design is in compliance with this position.
"o. Ducts and housings should be laid out with a minimum of ledges, protrusions, and crevices that could collect dust and moisture and that could impede personnel or create a hazard to them in the performance l of their work. Straightening vanes should be ins:alled where required to ensure representative air flow measurement and uniform flow distribution through cleanup components.'
'l I Amendment it 6A-6 lt
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l MM 23A610QAB Standard Plant m.< c ne design is in compliance with this position.
" Dampers should be designed, constructed, and tested in accordance with the provisions of Section 5.9 of ANSI N5091976?
The design is in compliance with this position.
)
'4 MAINTENANCE I
- a. Accessibility of components and maintenance should be considered in the design of ESF atmosphere ,
cleanup systems in accordance with the provisions of Section 2.3.8 of ERDA 76 21 and Section 4.7 of ANSI N5091976? The design is in compliance with this position. ;
*b. For case of maintenance, the system design should provide for a minimum of three feet from mounting t frame to mounting frame between banks of components. If components are to be replaced, the dimension to be provided should be the maximum length of the component plus a minimum of three ;
fect? The design is in compliance with this position. s
'c. The system design should provide for permanent test probes with external connections in accordance with l the provisions of Section 4.11 of ANSI N509-1976?
The design is in compliance with this position.
-t
'd. Each ESF atmosphere cleanup train should be operated at least 10 hours per month, with the heaters on (if so equipped),in order to reduce the buildup of moisture on the adsorbers and HEPA filters?
l The surveillance test requirements are provided in technical specification 3.6.43 (Chapter 16). Each space - heater, used only during SGTS standby, contains a small circulating fan which serves to maintain the entire filter train at a uniform temperature. The filter train is physically isolated during system standby Therefore, . SGTS operation other than as required for mitigation of offsite dose is not required. The isolation of the filter train also serves to minimize the degradation of the HEPA filters and charcoal adsorber due to weathering effects between surveillance tests. This is further discussed in Subsection 6.5.1.3.3, " Justification for Single ! SGTS Filter Train? ' \>
'c. The cleanup components (i.e., HEPA filters, prefilters, and adsorbers) should not be installed while [
active construction is stillin progress? Installation of the SGTS will comply with this position.
'5. IN PIACE TESTING CRITERIA
- a. A visualinspection of the ESF atmosphere cleanup system and all associated components should be
( V). Amendment 11 6A 7 i
l ABM insiaan am e i i R#mndard Plant made before each in place air Dow distribution test, DOP test, or activated carbon adsorber section leak test in accordance with the provisions of Sedion $ of ANSI N3101975? . i ne system test proced ires will comply with this position. l 1
%. The air flow distribution to the HEPA filters and lodiac adsorbers should be tested in place for ;
uniformity laitially and after maintenance affecting the flow distribution. The distribution should be j within t20% of the average flow per unit. The testing should be conducted in accordance with the provinnons of Section 9 of
- Industrial Ventilation
- and Section 8 of ANSI N5101975? !
Acceptance tests, performed after completion of initial construction and after any system modifications or repair (per Table 1 of ANSI N$10), will comply with this position. The guidance in ' Testing of Ventilation Systems,' Section 9 of ' Industrial Ventilation,' will be applied to any testing performed.
*c. The in. place DOP test for HEPA filters should conform to Section 10 of ANSI N510-1975. HEPA filter sections should be tested in place (1) initially, (2) at least once per 18 months thereafter, and (3) following palating, fire, or chemical release in any ventilation zone communicating with the system to confirm a penetration of less than 0.0$% at rated flow. An engineered safety feature air filtration system ,
satisfying this condition can be considered to warrant a 99% removal efficiency for particulates in s accident dose evaluations. HEPA filters that fail to satisfy this condition should be replaced with filters qualified pursuant to regulatory position C.3.d of this guide. If the HEPA filter bank is entirely or only partially replaced, an in place DOP test should be conducted? The surveillance test procedure will comply with this position. Technical specification 3.6.4.3 (Chapter 16) complies with this position.
'If any welding repales are necessary on, within, or adjacent to the ducts, housing, or mounting frames, the filters and adsorbers should be removed from the housing during such repairs. The repairs should be >
completed prior to periodic testing, filter inspection, and in place testing. The use of silicone scalants or any other temporary patching material on filters, housing, mounting frames, or ducts should not be allowed? The SGTS maintenance procedures will comply with this position. ,
'd, The activated carbon adsorber section should be leak tested with a gaseous halogenated hydrocarbon refrigerant in accordance with Section 12 of ANSI N5101975 to ensure that bypass leakage through the adsorber section is less than 0.0$%. After the test is completed, air flow through the unit should be maintained until the residual refrigerant gas in the efnuent is less than 0.01 ppm. Adsorber leak testing should be conducted (1) initially, (2) at least once per 18 months thereafter, (3) following removal of an adsorber sample for laboratory testing if the integrity of the adsorber section is affected, and (4) following painting, fire, or chemical telease in any ventilation one communicating with the system?
Surveillance testing is provided to comply with this position.
*6. LABORATOR 'l TESTING CRITERIA FOR ACTIVATED CARBON
- a. The activated carbon adsorber section of the ESF atmosphere cleanup system should be assigned the decontamination efficiencies given in Table 2 for elemental iodine and organic lodides if the following conditions are metf e
Amendment it 6A 8 - _ , .- _ , _ _ ,_ _ . _ _ . _ _ _ _ _ _ _ _ . __ . . ~
ABM susiota. ms Standard Punt The carbon bed is 6 inches deep. Per Table 2, the decontamination emeiency for be.d depths four inches or i greater is 99% The radiological analyses described in Subsections 15.6.5 and 15.7.4 assume a charcoal adsorber emciency ofless than 99%
*(l) The adsorber section meets the conditions given in regulatory Po6hlon C.5.d of this guide?
As atstod previously, the ABWR SGTS complies with Position C.5.d.
*(2) New activated carbon meets the physical property specifications given in Table 5.1 of ANSI !
N5091976, and' i Activated carbon installed in the SGTS will be covered by purchase requirements to meet the physical properties specified in Table 51 of ANSI N509. . t
*(3) Representative samples of used activated carbon pass the laboratory tests given in Table 2?
I Surveillance testing is provided to comply with this position. This position is laterpreted as follows. i , Reptesentative samples of used activated carbon will be laboratory tested with a frequency defined in Footnote c of Table 2 and as reflected in the technical specifications. Also per Footnote c of Table 2, a representative sample is defined lu Position C.6.b. Testing will be performed at a relative humidity of 70% per ASTM D3803. . The test acceptance criterion will be a methyl iodide penetration of less than 0.175% ASTM D3803 is cited in Table 51 of ANSI N5091980 for tests equivalent to those specified in Test 5.b of ANSI N5091976. I O 'If the activated carbon falls to meet any of the above conditions,it should not be used in enginected safety feature adsorbers? The activated carbon for the SGT3 wil; meet the conditions of Position 6.a(1),(2) and (M
'b. The efficiency of the activated carbon adsorber section should be determined by laboratory testing of i
representative samples of the activated carbon exposed simultaneously to the same service conditions as the adsorber section. Each repiesentative sample should be not less than two inches in both length and diameter, and each sample should have the same qualification and batch test characteristics as the system adsorbent. There should be a sufficient number of representative samples located in parallel with the adsorber section to estimate the amount of penetration of the system adsorbent throughout its service life. The design of the samplers should be in accordance with the provisions of Appendix A of ANSI . . N5091976. Where the system actiWed carbon is greater than two inches deep, each representative sampling station should consist of enough two inch samples in series to equal the thickness of the system adsorbent. Once representative samples are removed for laboratory test, their positions in the sampling array should be blocked off? > The detailed design will be in compliance with this position.
' Laboratory tests of representative samples should be conducted, as indicated in Table 2 of this guide, with the test gas flow in the same direction as the flow during service conditions. Simi!ar laboratory tests should be performed on an adsorbent sarnple before loading into the adsorbers to establish an initial- ,
point for comparison of future test results. The activated carbon adsorber section should be replaced with new unused activated carbon meeting the physical property specifications of Table 5.1 of ANSI O Amendment 11 6A.9 ;
, e -- L
t i r NN E*=ndard Plant 23A6tooAs a,.. e i N5091976 if (1) testing la accordance with the frequency specified in Footnote e of Table 2 resuhs in a l representative sample failing to pass the applicable test in Table 2 or (2) no representative sample is j ovellable for testing? ; The SGTS design and testing will comply with this position Physleal property testing is addressed in the ! response to Position C.6.a(2). l I e i i
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i O i O Amendment 11 6A 10 r -enw , - * ~ - .ns---+ ., e.+---- ., m ,- - ,- , ,-,.. --y . w-- .- ,-- -ne.--
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i l 3 APPENDIX 6B : SRP 6.5.1, TABLE 6.5.1-1 COMPLIANCE ASSESSMENT O 5 i i 3 l l 1 e i
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l 214610MB Standarsl Plant m- e i APPENDIX 6B SRP 6.5.1, TABLE 6.5.1 1 COMPLIANCE ASSESSMENT The following provides a comparison betvien the instrumentation specified in SRP 6.5.1, Table 6.5.1 1, and the instrumentation provided in the ABWR SGTS deaign. Justification is provided for those items that deviate from the SRP. The ulection and location of lastrumentation for the ABWR SGTS was re-ezamined during system design to rationaliar the operator interface. Instrumentation strictly required for monitoring the operation of the SGTS to mitigate offsite releases is provided in the main control room (MCR) on panel displays designed for that purpose. Monitoring, of couru, is a fundamental plant requirement specified in GDC 13. Instrumentation used for testing or maintenance is located at the local instrument rack. SRP Table 6.5.11 requires certain instrumentation to be in the main control room that is not necessarily , required for ABWR SGTS operation. To a certain degree, this is because SRP Table 6.5.111s based on a two filter. train design whereas the ABWR SGTS design uns a slagle filter train. Because of this difference, the instrumentation supporting ABWR SGTS operation for example, component pressure drops . is different. Therefore, because of the single filter train design, main control room readout of component differential pressures is not provided as noted in the remainder of the Appendix. I There are two basic parameters that are important to assure SGT3 function, ucondary containment preuure and charcoal adsorber inlet relative humidity. If the secondary containment preuure is less than the ambient pressure, any release from the plant passes through and is treated and monitored by the SGTS. If the inlet relative humidity to the charcoal adsorber is leu than or equal to 70%, then credit for a 99% efficiency may be taken (although charcoal performance at higher humidities provides significant decontamination factors), if an O , operator confirms the secondary containment pressure is negative with respect to ambient on all faces of the building and the relative humidity is less than 70% entering the adsorber, then the system is functioning as intended to mitigate calculated offsite doses. The ABWR SGTS design provides k divisional differential pressure transmitters with high and low alarms monitoring secondary containment pressure with respect to ambient pressure outside cach of the four walls of the reactor building, in addition, h divisions of moisture measurement with high alarms are provided in the filter housing upstream of the charcoal adsorber, providing a direct measurement of relative humidity. As a secondary indication of relative humidity, two divisions of inlet temperature (upstream of the process electric heaters) and two divisions of tempersture indication (upstream of charcoal adsorber) are also provided. The maimum possible relative humidity may be calculated based on the temperature rise acrou the heater. These basic parameters cach have main control room indication and alarm. > Unit inlet or Outlet Local Panel Main Control Room SRP Table 6.5.1 1 Plow rate (indication) Flow rate (recorded indication, high alarm and low alarm signals) ABWR SGTS None Outlet flow rate (recorded indication, low alarm); FRS618AB, F1618AB. Inlet temperature (Indication); T1602AB Amendment 11 6B-1
m MN 2m6tmAs Remndard Plant mm. e Local: The SRP indicates that the local panel should include an indicator for monitc. ring flow rate at the unit inlet or outlet. Flow rate indication at the local panel would presumably be used for testing. Dwing testing, flow rate measurements can be obtained from the main oomtrol room. Operation of the SGTS to mitigate offsitt relemes will not be affected by the abaeace of flow rate indication at the local panel. MCR: SRP Table 6.5.11 int!udes a high alarm signal to detect high ficw rate at the system inlet or outlet. The ABWR SGTS does not have this high alarm. A flow rate higher than the design value may indicate a potential failure in the fan or an increase in secondary containment leakage. However, as long as a negative pressure is maintained in accondary containment, SGT3 function is accomplished. Low negative secondary containment preuure is alarmed in the main control room. Operation of the SGTS to mitigate offsite releues will not be affected by the absence of the high flow alarm at the local panel. In addition, the ABWR SGTS design provides inlet temperature indication which is used in concert with downstream temperature meuvrement as a second means to determine relative humidityin the procons stream to the charcoal adsorber. Direct moisture mesurement is the primary me.ans to determine charcoal adsorber inlet relative humidity and is discussed in a later section of this response. Malature Renarator inealPanel Main Onatrol Ranen SRP Table 6.5.1 1 Pressure drop (Indication) None (optional high alarm signal) ABWR SGTS Pressure drop (indication); None DP1003AB The ABWR design is in compliance with SRP Table 6.5.1 1. Electric Henter Im1 Panel Maln Pantro1 Room SRP Tabic 6.5.11 Status indication None ABWR SGTS None Hand switch, status indication local: Provision of electric heater status at a local panelis not required for system operation given the controls in the main control room. The electric heaters are only used during system operation. Local panel or instrument tack information is used for testing. MCR: The ABWR design exceeds the control room requirements specified in SRP Table 6.5.11. Amendment it 6D-2 O ; I
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MNS 23A6100AB l Standard Plant am e Space Betwaza He.ater and Prefiker 1 um1 Panel Main r'antrol Ranen SRP Table 6.5.1 1 Temperature (indication, high Temperature (indication, high , alarm and low alarm signals) alarm, low alarm, trip alarm ' signals) l ABWR SGTS None , Temperature (high alarm and trip signals);TS005AB 1.ocal: Temperature indication required for testing is available from the control room. Operation of the SGTS to mitigate offsite releases will not be affected by the absence of temperature indication or the high and low alarms at the local panel. . MCR: The 1 Inh alarm and trin in the ABWR SGTS design ispd to electric heater should the heater temperature increase F). This isabove 110 the slightly above C (230, 225 F alert the o referenced in ANSI N509, Subsgetion 5.5.1, but well within the available margin. Per ANSI N509, Section 4.9, higher temperatures (above 300 F) may lead to significant desorgtion of, lodine from the charcoal. Potential . ignition of the charcoal occurs at a much higher temperature (290 C/554 F per ERDA 76 21, Subsection 3.4.2) i and is also not a concern. Note that the ABWR SGTS charcoal will meef the agre stringent physical property specification of ANSI N509, Table 51, for ignition ternperature (330 C/626 F) [see also the response to Position C.3.1, Appendiz 6.5A). Traditionally, relative humidity is maintained by controlling the temperature rise across the heater. A low temperature Alarm indicates a potential heater failure such that the relative humidity la the process stream may
; not be maintained. Additional temperature and relative humidity indication and high alarms are provided between the first HEPA filter and the charcoal adsorber and are described in a later section of this response.
These additional instruments serve the same function as the SRP instruments - assuring a relative humidity in the proecss stream passing through the charcoal of less than 70% A high relative humidity signals the same problem as low temperature. Therefore, the ABWR design meets the intent of SRP Table 6.5.11. See the discussion of basic parameters at the beginning of this response for an understanding of ABWR SGTS instrumentation design. E.LCldfat Local Panel Main Control Roorg,,,, SRP Table 6.5.1 1 Pressure drop (indication, high None alarm signal) , . ABWR SGTS Pressure drop (indication); None '~ DP1007 The SRP includes a high alarm signal for monitoring pressure drop across the prefilter. The ABWR design does not have this alarm. Localinstrumentation for prefiker pressure drop measurement is used for testing , purposes. A high alarm signal would not be appropriate during testing given the direct indication available on l the instrument tack. Low system flow is alarmed in the control room should fan runback occur from any cause. 4 Operation of the SGTS to mitigate any potential offsite release will not be affected by the absence of the alarm on the local panel. O Amendment it 6B 3 , i
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l 21A6100AB Standard Plant a,< c thal 5fEPAfdierlPic_HEPA) I u nt pan.1 unin enneentRnam
$RP Table 6.5.11 Pressure drop (indication, high Pressure drop (tecorded
, alarm signal) indication) l l AEWR SOTS Pressure drop (indication); Nonc j i DP1008 j i i local: The local panel has indication for confirming the proper pressure drop across the HEPA filter during testing. Like the prefilter, a high alarm signal would not be appropriate during testing given the direct indication i available on the instrument rock.14w system flow is alarmed in the control room should fan runback occur. Operation of the SGT3 to mitigate any potential offsite release will not be affected by the abacace of a local high , alarm. MCR: During system operation,it is not expected that the HEPA filter would exhibit an excessively high l pressure drop by virtue of the periodic testing for pressure drop and filter efficiency performed in accordance ! with the schedules specified in the Technical Specifications. J l Knace between First HEPA Fiher and Adtarber )i t uml Panel Maln enneral Rnnm SRP Table 6.5.1 1 None None ABWR SGTS None Moisture (four divi 61ons of indication and high alarm)* MI611 ABCD. Temperature (two divisions of indication, l control and trip, high alarm); ; 1 T1610AB, TS009AB, TS610AB. Space heater hand switch and atstus indication. ) 1 As mentioned previously, direct moisture indication is provided to assure relative humidity is less than 70% in the gases entering the charcoal adsorber. Relative humidity is a fundamental parameter for system function and l has been emphasized in instrumentation design. Space heaters with related temperature and status instrumentation are provided both upstream and downstream of the charcoal. Discussion of this lastrumentation is provided in a later section," Space between Adsorber and Second HEPA Filter." The ABWR SGTS design exceeds the requirements of SRF Table 6.5.11 and ANSI N509, Table 41. Adsorber local Panel Main Control Room SRP Table 6.5.11 None None i ABWR SGTS Pressure drop (indication); Temperature (high alarm); ' DPIO12 TS013AB Amendment il 6B 4 O
ABWR- mama ; Standard Plant we ) The ABWR SGTS design provides two divisions of high temperature alarm both directly upstream and downstream of the charcoal adsorber. The purpose of this alarm is to alert the operator to the potential for ) y desorption of iodine from the charcoal (if the SGTS is operating post. accident) or of a failure in one of the temperature control and high alarm circuits associated with the heaters. The setpoint for this alarm signal is 155 C (311 F). Should temperature reading and alarms indicate a continued and uncontrolled high temperature during SGTS operation, deluge actuation may be warranted. Pressure drop is provided at a local rack for j testing. The ABWR SGTS design exceeds the requirements of SRP Table 6.5.11. Snace between Adsorber and second HEPA Filter (Post HEPA) Local Panel Main Cnntrol Room , SRP Table 6.5.1 1 Temperature (two stage high- Temperature (indication. l alarm signal) two-stage high alarm signal) . l I ABWR SGTS None Temperatore (two divisions of indication, control and trip, high alarm); T1616AB, TS015AB, TS616AB Space heater hand switch and status indication, e Local: Local temperature alarms are not provided since the area is not continuously manned. Appropriate alarms and indication are provided in the control room along with the necessary controls to respond to a high temperature signal, , MCR: The intent of the SRP MCR position, judging from Footnote 2 of Table 41 of ANSI N509,is to provide . an alarm on high temperature and signal for manual deluge actuation on a high.high temperature alarm. In the
- ABWR design, high and low temperature trips based on readings from the filter housing, both upstream and downtream of the charcoal, are used to cycle one of the two heating elements in each space heater on and off.
A low temperature alarm is provided if low temperature coincident with space heater operation (i.e., not out of service)is detected. High temperature between the space heater elements gives an alarm in the MCR and cuts power to the other heating element supplied with each space heater. , Each space heater heating element is prosided with status indication. Each space heater fan is prosided with a hand switch and status indication. The need for deluge actuation is discussed in a later section of this Appendix,
- Deluge Valves," and also in Subsection 6.5.13.3,' Justification for Single SGTS Filter Train."
Second HEPA (Post HEPA) Local Panel Main Control Room SRP Table 6.5.1 1 Pressure drop (indication, high None alarm signal) ABWR SGTS Pressure drop (indication); None DP1017 b)
\
Amendment 11 61k$
ABM swiecan Standard Plant m.v. c Local: The local panel has indication for confirming the proper pressure drop across the HEPA filter during testing. Like the prefdier and first HEPA filter, a high alarm signal would not be appropriate during testing given the direct indication on the rack. Low system flow is alarmed in the control rooms should fan runback occur. Operation of the SGTS to mitigate any potential offsite release will not be affected by the absence of a localhigh alarm. [ Process) Fan t ual Panat Main ControlRoom SRP Table 6.5.11 (Optional hand awitch and Hand switch,statusindication atstusindication) ABWR SGTS None Hand switch, status indication (run/stop) The ABWR SGTS design complies with SRP Table 6.5.1 1. Valve /Dannner Onerator i mm1 Panel Main enntrol Room SRP Table 6.5.11 (Optionalstatus indication) Status indication ABWR SGTS None Hand switch, status indication (open/ closed), position. indication; pol 601AB The ABWR SGTS design exceeds the requirements of SRP Table 6.5.11 Valve position indication (and control) is provided on the inlet dampers, P002AB. Delune Valves Incal Panel Main Cnntrol Room SRP Table 6.5.11 Hand switch, status indication Hand switch, status indication ! ABWR SGTS None None Manual deluge capability is provided on the ABWR SGTS Inadvertent wetting of the charcoal has led to system unavailability in operating plants. Remote deluge control, either from a local panel or the main control room is not provided. As such status indication (open/ closed) is not required. System avaliability is improved without compromising fire protection requirements. The recently issued ASME N5091989 also shows a move away from remote operated valves, since the requirement for the hand switch has been deleted and status indicationis now optional. l Amendment 11 6B 6
_ . _. _. _- _~ _ . _ . . _ . _ __ _ _ _ _ _ _ _ - _ _ _ __ i ABWR zwtms Standard Plant am e > Evatem inlet to Og}gg
.O. I mm1 Panel Main cantrol Raam SRP Tabic 6.5.11 None Summation of pressure drop across total system, high alarm signal ABWR SGT1s None None Pressure drop across the entire system is not required for either operation after an accident or for testing.
Maintenance of a negative pressure and assurance of a relative humidity below 70% in the process alream entering the charcoal adsorber section are the key parameters for system operation. Flow indication is provided ' in the main control room along with a low system flow alarm for an indication of high pressure drop. Other C*eandary Cnntainment Differential Pratence Imcal Panel Main cantrol Room SRP Table 6.5.11 None None -
- ABWR SGTS None Differential pressure (four [
divisions of indication and high and low alarms) The ABWR SGTS design exceeds the requirements of SRP Table 6.5.11. Measurement of secondary I- containment pressure with respect to the emirons is a fundamental system parameter which is specified within, and is under the control of, the ABWR SGTS design. Other 1m00 Seah imcal Panel Main Control Room l SRP Table 6.5.11 None None ABWR SGTS None Level (two divisions of low alarm) Loop seals are provided within the dryer and filter train and in the piping downstream of the filter train discharge block valves. Redundant low level alarms are provided to assure loop seallevel is maintained. The loop seals function to continuously and passively drain any accumulation of water in the SGTS. Accumulation of water in piping to the stack has been a problem in operating plants, s I G Amendment 11 6D-7
1 I l MM Rtendard Plant 2M6100AF REV. B l l 4 CHAPTER 7 TABLE OF CONTENTS (Continued) Section Dtle P.ase .) i
. Neutron Monitoring System 7.6 12 ]
'l
. Process Radiation Monitoring .
System 7.6 14 ]i
. HP/LP System Interlock Function 7.6 15
. Fuel Pool Cooling and Cleanup ;
t System 7.6 17
. Wetwell to Drywell Vacuum Breaker [
System 7.6 18
. Containment Atmosphere Monitoring System 7.6 18
- Suppression PoolTemperature Monitoring System 7.6 20 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 7.71 7.7.1 Description 7.71 7.7.2 Analysis 7.7 36
- Nuclear Boiler System . Reactor VesselInstrumentation 7.7 36 i
. Rod Control and Information System 7.7 37
. Recirculation Flow ControlSystem 7.7 38
. Feedwater ControlSystem 7.7 39
. Process Computer System 7.7 39
. Neutron Monitoring System .
(ATIP Subsystem) 7.7 40 7.8 INTERFACES 7.8-1 q 7.8.1 Effects of Station Blackout on the HVAC 7.8-1 1 7.8.2 Electrostatic Discharge on Exposed Equipment Components 7.81 7iv Amendment 11 l i l
MM 23A6100AF Etandard Plant arv n l l CHAPTER 7 j TABLE OF CONTENTS (Continued) : Mn Dijg - P.ase ,; 7.83 LMH Hist. deat Spots in Semiconductor . l
' Materials for Caputing Devices 7.8-1 7.8.4 Safety Related C&lInterfaces 7.8-1 APPENDIX 7A' Design Response to Appendix B of .
ABWR IJcensing Review Bases; i lastrumentation and Controls 7A.1 1
'i G1
+
l l I 1 I i 7v Amendment 11 ) I 1 l
I , 1sA62 JOAT' Etandard Plant nry a j l O D 7.1.2.1.6 Protection System In. Service Testability (3) The third test is the single rod seram l testwhich verifies the capability of each j rod tc scram. It is accomplished by The RPS and ESF systems can be tested during operating switches for the particular reactor operation by six separate tests. The control rod drive. Timing traces can be first five tests are primarily manual tests and, made for each rod scrammed. Prior to the ' although each individually is a partial test, test, a physics review is conducted to combined with the sixth test they constitute a assure that the rod pattern during scram complete system test. The sixth test is the testing will not create a rod of
- self test of the safety system logic and control unacceptable reactivity worth.
which automatically tests the complete system escluding sensors and actuators. (4) The fourth test checks calibration of analog sensor inputs at the analog inputs of the (1) The first of these is the manual scram remote multiplexing units. . With a test. The manual scram test verifies the division.of. sensors bypass in place, , ability to de energize the scram pilot valve calibrated, variable ramp signals are i solenoids without scram by using the manual injected in place of the sensor signals and scram pushbutton switches. By depressing monitored at the SSLC control room pancis the manual scram button for one trip logic, for linearity, accuracy, fault response, and balf of the seram sof enof ds are downscale and upscale trip response. The de.cnergized. After the first trip logic is test signals are adjustable manually from reset, the second trip logic is tripped the control room and also are capable of manually to complete the test for the two performing an automatic sequence of events. manual scram buttons. In addition to When surveillance testing during plant control room and computer printout shutdown, trip colacidence and actuated indications, scram groups indicator lights device operation can be verified by l O Indicate that the actuator tilp logics have de energized the scram pilot valve solenoids. simultaneous trip tests of coincident channels. Pressure transmitters and level transmitters are located on their respec. tive local panels. The transmitters can be On the back panels, a separate, manual in. dividually valved out of service and pushbutton switch in each of the four subjected to test pressure to verify divisions provides means to manually trip operability of the transmitters as well as all trip actuators in that division. This verification of calibration range. To gain sealed.in division manuat trip is equivalent access to the field controls on each to a sealed.in automatic trip from the same transmitter, a cover plate or sealing device division of trip logic. (Au alternate must be removed. The access to the field manual scram can be accomplished by controls is administratively controlled. depressing any two or more of the four Only qualified personnel are granted ac. divisional manual trip pushbuttons.) cess for the purpose of testing or calibration adjustments. (2) The second test includes calibration of the neutron monitoring system by means of (5) The fifth test is the sensor check. Digital l : simulated inputs from calibration signal inputs are tested by varying the monitored units. Calibration and test controls for variable (e.g., stop valve closure, control the neutron monitoring system are located in valve fast closure, main steamline isolation the control building equipment room. They valve closure) or by disconnecting thc are under the administrative control of the sensor from the process variable and control room operator and can be done either inputting and varying a manually or automatically. Subsection 7.6.1.1, ' Neutron Monitoring System," describes the calibration procedure. Amendment 11 7.15
, - - . . , - , , - + , , . , .
ABWR
- Renndard Plant test source. In those cases where the sensor is disconnected from the process variable, an out of. service alarm will be indicated in the main control room. Analog input is checked by cross comparison of the instrument channels measuring the same variable.
O Amendment 11
l 2M6100AF 1 ReanAard Plant arv. m J [.
- , (6) The sixth test is an lategrated self test self diagnosis includes monitoring of l provision built into the microprocessors overall program flow, reasonableness of ,
within the safety system logic and control process variables, RAM and PROM condition, l (SSLC). It consists of an on line, nr.d verification of 2/4 colacidence logic ' continuously operating, self diagnostic and device laterlock logic. Testing j monitoring network; and an off line includes continuous error checking of all , semi automatic (operator initiated, but transmitted and received data on the serial automatic to completion), end to end data links of each SSLC controller; for surveillance program. Both on line and example, error checking by parity check, off line functions operate independently checksum, or cyclic redundancy checking > within each of the four divisions. There (CRC) techniques. r l are no multi divisional laterconnections associated with self testing. A fault is considered the discrepancy - between an expected output of a permissive l l The primary purpose of the self test is to circuit and the existing present state. I improve the availabilty of the SSLC by optimizing the time to detect and determine Actuation of the trip function is not , the location of a failure in the functional performed during this test. The self. test system, it is not intended that self test function is capable of detecting and logging , eliminate the need for the other five manual intermittent failures without stopping tests. However, most faults are detected system operation. Normal surveillance by more quickly than with manual testing alone, plaat personael will identify Ihese failures, via a diagnostic display, for The self test function is classified as preventive maintemance, safety associated. However, its hardware and soffweie are an integral part of the Self test failures (except latermittent SSLC and, as such, are qualified to Class 1E failures) are annunciated to the operator at O V standards. the main control room console and logged by the process computer. Faults are The hierarchy of test capability is provided indentified to the replacement board or to ensure maximum coverage of all EMS /SSLC module level and positively indicated at the functions, including logic functions and failed unit, data comtounications links. Testing shall include: Tbc continuous surveillance monitoring also , include; power supply voltage levels, (a) On line Continuous Testing. card.out of. file interlocks, and battery voltage levels on battery backed memory A self diagnostic program monitors cach cards (if used). Out of tolerance signal processing module from input to conditions will result in an inopertive , i output. Testing is automatic and is (out of service) condition for that performed periodically during normal particular system function. i operation. Tests will verify the basic integrity of each card or module on the Automatic system self testing occurs during moctoprocessor bus. All operations are part a portion of every periodic transmission of normal data processing intervals and will period of the data communication network, ' not affect system response to incoming trip Since exhaustive tests cannot be performed or initiation signals. Automatic initiation during any one transmission interval, the ; signals from plant sensors will override an test software is written so that sufficient automatic test sequence and perform the overlap coverage is provided to prove system required safety function. Process or logic performance during tests of portions of the signals are not changed as a result of circuitry, as allowed in IEEE 338. self. test functions. Amendment 11 El 6
ABM RimmAard Plant nasion arv. n i The essential multiplexing system (EMS) is roceed automaticsily to conclusion after ,I included in the continuous, automatic fnitiation by the oprator. Surveillance
]
self test function. Faults at the remote testing is performed in one division at a ; multipleting units (RMUs) are alarmed in the time. . ! main control room. Since EMS is dual in ,_ each division, self. test supports automatic The STC injects test patterns through the i i reconfiguration or bypass of portions of EMS essential multiplexing system (EMS) ! after a detected fault, such that the least communications links to the RMUs. It then i effect on system availability occurs, tests the RMUs ability to format and transmit sensor data through and across the (b) Off line Semi automatic End to End EMS /SSLC interface, in the prescribed time, Testing to the load drivers. Under the proper. bypass conditions, or with the reactor shut The more complete, manually initiated, down, the load drivers themselves may be . l Internal self test is available when a unit actuated, i is off line for surveillance or maintenance l 1 testing. This test exercises the trip All testing features adhere to the single outputs of the SSLC logic processors. The failure criterion, as follows: 1) No single channel containing the processors will be failure in the test circuitry shallincapacitate bypassed during testing. an SSLC safety function. 2) No single failure in the test circuitry shall cause an inadvertent-A fault is considered the inability to open scram, MSIV isolation, or actuation of any ' or close any control circuit. safety systems served by the SSLC. Self test failures are displayed on a front panel readout device or other diagnostic unit. To reduce operator burden and decrease l outage time., a surveillance test controller (STC)is provided as a dedicated instrument l in each division of SSLC. The STC performs semi automatic (operator initiated) testing ;
- of SSLC functional logic, including trip,
! initiation, and interlock logic. Test coverage includes verification of correct operation of the following capabilitier, as defined in each system IBD. (i) Each 2/4 coincident logic function. (ii) Serial and parallel I/O, including manual control switches, limit switches, and other contact j closures. l l (iii) The 1/N trip selection function. l 1 (iv) Interlock logic for each valve or pump. A separate test sequence for each safety l system is operator selectable; testing will l Amendment 11 7.16.1 l
2&A6100AF m... . pi... m. :
.. . result of escessive internal pressurs !
O~ (i.e., to prevent nuclear system pres. ; sure from eseeeding the limit allowed by l applicable ladustry codes); l (c) to limit the uncontrolled release of radioactive materials from the fuel - assembly or reactor coolant pressure i
- boundary, by precisely and reliably ;
initistlag a reactor scram on gross ! l failure of either of these barriers; .! l (d) to detect conditions that threates the l fuel assembly or rear, tor coolant j pressure boundary from inputs derived i from variables that are true, direct ; l measures of operational conditional- ;
' (e) to respond correctly to the sensed vari. .
ables over the espected range of magni. l tudes and rates of change; (f) to provide a sufficient number of sen. ' I sors for monitoring essential variables that have spatial dependence;- 5 e The following bases assure that the RPS is i designed with sufficient reliability: , 5 (g) If a single random failure can cause a $ control system action that causes a $ plant condition that requires a reactor ~ scram but also prevents action by some RPS channels, the remain, ing portions i 7.1.2.2 Reactor Protection (Trip) System (RPS). of the RPS shall meet the functional Instrumentation and Control requirements (items a, b and c above), 4 even when degraded by a second random I (1) Safety Design Bases (Conformance to the fol. failure. lowing design bases is discussed in Section l *1.2.2.1). (h) Loss of one power sup/ thall neither l directly cause nor prt o reactor , i The reactor protection (trip) system (RPS) scram. shall meet the following functional require. ments: (i) Once initiated, an RPS action shall go l to completion. Return to normal opera. (a) to initiate a reactor scram with preci. tion shall require deliberate operator sion and reliability to prevent or limit action. :
- i. fuel damage following abnormal opera. i i
tional transients; (i) There shall be sufficient electrical and physical separation between redundant (b) to initiate a scram with precision and instrumentation and control equipment reliability to prevent damage to the monitoring the same variable to prevent O reactor coolant pressure boundary as a environmental factors, electrical tran. Amendment 11 7.17
A.BM. a.. ,i... s w ia m
.rv ,
sients, or physical events from impair. Specine Regulatory Requirements: j les the ability of the system to rupond correctly. The specific requirements applicable to the , RPS instrumentation and somtrol are shown in ; (k) Earthquake ground motions, as amplified Table 7.12. i by building and supporting structures, -i shall themselves initiate reactor scram, (2) Nossafety.Related Deales Bases l and shall not Impelt the ability of the t RPS to otherwise initiate a reactor The RPS is dealgaed with the added objective l scram, with the exception of turblee of plant availability. The setpolets, power . building trips which originate from a sources, and control and lastrumentation j son.scisale building. These shall be shall be arranged is such a manner as to -i backed up by diverse variables such as preclude spurious actans lasofar as .! reactor pressure and power trips. practicable and safe.- [ i , (1) No single failure within the RPS shall 7.1JJ Eagleested safety Features (ESF) ; l prevent proper reactor protection system l action when required to satisfy Safety 7.1JJ.1 6.,ci ceev Coellag Systems . t Design Bucs as described by the first lastrumentation ned Castrels l three bullets under 1(a5 ove. + l (1) Safety Design Bases , (m) Any one latentional - f.un, malatenance operstloa, calibrath speration, or General FunnlonalRequirements: . g test to verify operational availability { shall not prevent the ability of the The ECCS control and lastrumentation shall I- l reactor protection system to respond be designed to meet the following 7 e correctly, requirements: (n) The system shall be designed so that two (a) automatically initiate and control the or more sensors for any monitored emergency core cooling systems to i variable escceding the scram setpoint prevent fuel cladding temperatures from ; will initiate an automati: seram, reaching the limits of 20CFR50.46. The followir.g bases reduce the probabl. (b) respond to a need for emergency core ! lity that RPS operational reliability cooling regardless of the physical and precision will be degraded by location of the malfunction or break - operator error: that causes Ibc need; (o) Access to trip settings, component call. (c) limit dependence on operator judgement l bration controls, test points, and other in times of stress by: ; L terminal points shall be under the con. trol of plant operations supervisory automatic response of the ECCS so that personnel, no action is required of plant operators R within 30 minutes after a loss of. $ (p) Manual bypass of instrumentation and coolant accident; , control equipment components shall be r under the control of the control room indication of performance of the ECCS by [ operator. If the ability to trip some main controt room lastrumentation; and essential part of the system has been , bypassed, this fact shall be continuous. provision for manual control of the ECCS ly annunciated in the main control room, in the main control room. O, Amendment 9 7.18 .
. , . - . . . , . . . , . . _ . _ . _ _ , , _ , ~ , _ . . , , . . , . _ . , . . ~ _ ., , -m. , .-.m-.
1 l 21A610aAF EfamAn d Plant REV. B . I TABLE 7.11 .
>O COMPARISON OF GESSAR 11 AND ABM3 I&C SAFETY SYSTEMS (Continued) l&C System GESSAR !! Design ABWR Design RHR/ SUPPRESSION 2 loops and 2 divisions 3 loops and 3 divisions POOL COOLING MODE: .
FLAMMABILITY CONTROL Part of combustible gas ladependent system SYSTEM: control system. STANDBY GAS TREAT- Redundant active and pas- Redundant active compo-MENT SYSTEM: sive components ments; single filter train.' . l l EMERGENCY DIESEL ESF diesels: Divisions 1 & ESF Diesels: Dhklons I, GENERATOR SYSTEM: 2. HPCS diesel: Div.3. 11 & !!!(HPCFincluded on : Dhhions II & 111). REACTOR BUILDING Open loop to ultimate Closed loop with limited : COOLING WATER: heat sink. System was quantity of water. l- O called
- essential senice water system *,
Hydrogen mixing system Dedicated hydrogen mixing interface, not required for inerted containment. HIGH PRESSURE (Air supply only) Replaces air supply to ADS NITROGEN GAS SUPPLY: and SRV accumulators. Also used for testing MSIVs. . ALTERNATE ROD (Not applicabic) New function provided by INSERTION (ARI) fine motion control rod l FUNCTION: drive (FMCRD) capability of the rod control & information system (RC&lS). t STANDBY LIQUID Squib-type injection valve. Motor operated. type CONTROL SYSTEM injection valve. ; (SLCS): Pump indication 'RUN', Pump indication 'RUN',
'STOP", " TRIPPED * 'STOP" O '
Amendment 11 7.124
?
ABWR awi=^r I Standard Plant REY.B TABLE 7.1 1 COMPARISON OF GESSAR II AND ABWR !&C SAFETY SYSTEMS (Continued) l&C System GLSSAR II Design ABWR Deelen RHR/ SHUTDOWN 2 shutdown coolmg 3 shutdown cooling COOLING MODE: divisions with 1 section divisions with 3 suction-line, lines (1 per division). REMOTE SHUTDOWN RCIC controls available at RCIC controls replaced with SYSTEM (RSS): RSS panel ' HPCF controls at RSS panel. SAFETY RELATED Designed to address Designed to address DISPLAY - Regulatory Oulde 1.97, Regulatory Guide 1.97, INSTRUMENTATION: Revision 2. Revision 3. NEUTRON MONITORING Class 1E subsystems are Class 1E subsystems are SYSTEM (NMS): IRM,LPRM & APRM. SRNM (combines IRM & SRM), i LPRM & APRM. NEUTRON MONITORING Non-Class 1E subsystems are Non Class 1E subsystem is SYSTEM (Continued) SRM &TIP. ATIP. PROCESS RADIATION - -- ~. New system definition and MONITORING SYSTEM organization, i.e., new (PRMS): instrument groupings, - locations and ranges. FUEL POOL COOLING & - - - - - - - Comparable ! & C CLEANUP SYSTEM: DRYWELL VACUUM Electrically operated Mechanically operated - , RELIEF SYSTEM: butterfly valve, relief vake. 1 CONTAINMENT (Not in GESSAR !! scope) New system provided in ABWR ATMOSPHERE scope. MONITORING SYSTEM (CAMS): SUPPRESSION POOL 4 thermocouplesin each of 4 thermocouples in each of TEMPERATURE the 4 containment 2 divisions at each of 6 MONITORING quadrants. locations. SYSTEM: 4 4 X 4 = 16 total T/C's. 4 X 2 X 6 = 48 total T/C's. Added suppression pool level monitoring function. O i
2%6100AF EtanA.ed Pl==* mm a SECTION 7.3 O CONTENTS Section Due Eas 73.1 Daaeriptian 73 1 73.1.1 Systems Description 73 1 73.1.1.1 Emergency Core Cooling Systems lastrumentation
. and Controls 73 1 73.1.1.1.1 High Preuure Core Flooder System lastrumentation and Controis 73 1 73.1.1.1.2 Automatic Depresurization System lastrumentation and Controls 73-4 7.3.1.1.13 R eactor Core Isolation Cooling (RCIC) System .
Instrumentation and Controls 73-9 73.1.1.1.4 RHR/ low Pressure Flooder (LPFL) Instrumentstion i and Controls 73 15 73.1.1.2 leak Detection and Isolation System (LDS) O Instrumentation and Controls 7 3-19 73.1.13 RHR/Wetwell and Dr>well Spray Cooling Mode Instruraentation and Controls 7 3-20 73.1.1.4 RHR/ Suppression Pool Cooling Mode. Instrumentation and Control 7 3-22 73.1.1.5 Standby Gas Treatment System lastrumentation and Controls 73 24 7.3.1.1.6 Emergency Diesel Generator Support Systems 73 26 73.1.1.7 Reactor Building Cooling Water System - lastrumentation and Controls 73 26 73.1.1.8 EuentialHVACSystems Instrumentation and Controls 7.3-28 73.1.1.9 HVAC Emergency Cooling Water System Instrumentation Controls 7 3-28 73.1.1.10 High Pressure Nitrogen Gas Supply System Instrumentation and Controls 73 31 7 3.1.1.11 Flammability Control System (See Section 6.2.5) 73 32. -i 73 11 Amendment 11
MM i w ia m i WPlant arv t; SECTION 7J CONTENTS (Continued)
= = e 7.3.1.2 Design Basis Information 7 S 32 7.3.1.3 System Drawings 7&M 7J.2 Aakhals 7&M 7.3.2.1 Emergency Core Cooling Systems lastrumentation and Controls 7.3.M 7.3.2.1.1 General Functional Requirements Conformance 7&M 7.3.2.1.2 Specmc Regulatory Requirements Conformance 1SM 73.2,2 leak Detection and laolation System lentrumentation and Controns 7S39 7.3.2.2.1 General Functional Requirement Conformance 7 S 39 73.2.2.2 Specific Regulatory Requirements Conformance 7.M9a 7.3.2.3 RHR/Wetwell and Drywell Spray Mode Instrumentation and Contioh 7.3 40 73.2.4 RHR/ Suppression Pool Cooling Mode .
Instrumentation and Controis 7.341 73.2.4.1 General Functional Requirements Conformance 7.3-41 7.3.2.4.2 Specific Regulatory Requirements Conformance 7.3-42 73.2.5 Standby Ga5 Treatment System Instrumentation and Controls 73 43 73.2.5.1 Conformance to General Functional Requirements - 7.3 43 73.2.5.2 Snecific Regulatory Requirements Conformance 7.343 7.3.2.6 Emergency Diese1 Generator Support System lastrumentation and Control 73 45 73111 Amendment 6 O
ABWR msar memmAmed Plans arv.m , i vided with containment isolation valves l l ( 7J.1.1.10 High Proasure Nitrogen Gas Supply System . instrumentation and Controls where the HPIN system lines enter the
]
j containment. l (1) Systemidentification i The valves are manually operated from in. . The high pressure nitrogen gas (HPIN) aupply dividual control switches in the control 1 system provides compressed nitrogen of the room. ! required pressure to the ADS safety / relief i l valves, the mainsteam isolation valves (for (a) lattistir.g Currents testing only), instruments and pneumatically i operated valves in the PCV and other nitro. During normal operation, nitrogen gas gen.using components in the reactor build. pressure is controlled and measured in a j ing. The P&ID is shown in Figure 6.71 sad pressure control valve followed by a : the interconnection block diagram is shown pressure transmitter. The pressure ! In Figure 7.310. control valve setpoint is high enough to i ensure that adequate nitrogen pressure (2) Support Systems (Power Source) Is deliyered to all tbe servcd ; accumulators and valves. ' The safety.related portion of the HPIN system is powered from the onsite Class 1E Automatic closure of the isolation valve AC and DC systems. The safety.related from the normal nitrogen gas supply and portion is switched automatically to the the opening of the isolation valve from , staedby AC power supply during a loss of the emergency nitrogen gas bottle is l normal power. The nonsafety related portion laitiated by low mittogen pressure ' is connected to the normal AC power supply, seaaed in the lines to the ADS accumulators. (3) Equipment Design l O The HPIN system is separated into (b) Logic and Sequencing nonsafety related and safety related The initiation of the flow of nitrogen sections. gas from the high pressure storage bottles is by low pressure in the lines , The nonsafety related portion of the system to the ADS accumulators. Concurrent. i includes an inlet filter, piping, and valves ly,the valves isolating the nonsafety. to all nitrogen users, related portion of the system are < , The safety related portion of the system includes two banks of high pressure nitrogen (c) Bypasses and laterlocks bottles and associated piping, valves, and controls. The isolation valves on HPIN system lines serving systems in the containment When low nitrogen gas pressure is detected have motor operators. The isolation : la the lines to the ADS accumulators, the valves may be closed to prevent any safety related portion of the system is possible leakage from the containment if . I isolated from the nonsafety related portion a leak occurs in the system outside of l by isolation valves which automatically cut the containment. l off the normal nitrogen gas supply and open the emergency nitrogen gas bottle supply to (d) Redundancy und Diversity i the ADS accumulators. The HPIN system is separated into two in addition to valves that isolate non. mechanically and electrically indepen. safety related equipment from safety. dent divisions. Each division has
- p relat-d equipment, the HPIN system is pro. Instrumentation, controls, and power V .
Amendment 11 't.3 31
I l i M M.a.e... matour ma sources which are separated and ledepen. (g) Testability I dent from each other. One division ! supplies emergency sitrogen to four ADS The HPIN system can'be tested at any' ; valve accumulators and the other divl. time by isolating the system from the 1 alon supplies emergency sitrogen to the normal alttogen source and allowing the ] esmalalag four ADS valves. This level nitrogen pressure to decrease. At the ! of redundancy is sufficleat because only proper pressure, valves should open, 1 the leltlal 1.OCA depressuritation requi. . admitting nitrogen from the high pres. ) res more than four ADS valves and the sure storage bottles; other valves ! Class 1E accumulators have sufficleat should close, Isolatlag the sonsafety. J espacity for at least two valve opera. related portions of the system, l tions. ' (b) EnvironmentalConsiderations ] The high pressure nitrogen gas storage : bottles are la two racks separated from The system safety.related equipment is ; each other. Additionally,in each tack selected in consideration of the normal ! there are two banks Of two bottles and accident environments in which it i each. One bank is in service and the must be operated. ] second is in standby. ! (i) OperationalConsiderations :
. (c) Actuated Devices l
'!he HPIN ayatem, when required for emer.
Nitrogen is admitted to the system and gency conditions,is initiated automa. the nonsafety related portion isolated tically with no operator action requit. l by operating valves controlled by pres. ed. i sure switches in the HPIN system. These i valves can also be operated from the Running lights, valve positions, Indi. ' main control room, cating lights, and alarms are available ; in the control room for the operator to ; All isolation valves can be manually accurately assess the HPIN system ope. ; operated from the main control room, ration. Common trouble alarms are Each valve is provided with Indicating available in the main control room for l position lights in the main control room the system. Isolation valves-have ; which verify the open and closed indicating lights for full.open and i positions of the valve, full. closed positions. I (f) Separation 7J.1.1.11 Flammahliity Control System. ! The HPIN system is separated into two divisions, each having storage bottles (See Section 6.2.5) and racks and piping to the ADS accu. mulators. 7.3.1.2 Dealga Basis Inforination Physical separallon of Division I and IEEE Standard 279 defines the requirements Division 11 systems is obtained by for design bases. Using the IEEE.279 ! closing valves which Interconnect the format, the following nine paragraphs i divisions during normal operation, fulfill this requirement for systems and ' equipment described la this section. Electrical separation is maintained by separate sensors and circuits indepen. (1) Conditions l dent of each other. ' The plant conditions which require protec. ! tive action involving the systems of this secilon and other sections are examined and , j presented in Chapter 15. j l \ A m M m m 11 UM l
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i i MM 2 w ton @ am a : l l itandard Plant , l (] SECTION 7.7 CONTENTS - l listuna Ildt East i 7.7.1 gauddha 7.7*1 l 7.7.1.1 Nuclear Boiler System . Reactor Vessel ! l-lastrumentation 7.71 ! 7.7.1.2 Rod Control sad information System + , lastrumentation and Controis 7.75 l 3 7.7.1.2.1 Contiol Rod Drive Contr01Sptem 7.78 ! l 7.7.1.2.2 Other Sptems interfaces 7.7 17 7.7.1.23 Reactor Operator Information 7.7 19 7.7.1.2.4 Test and Malntenance 7.7 21 7.7.1.2.5 Environmental Considerations 7.7 21 7.7.13 Recirculation Flow Control System Instrumentation and Controh 7.7 21 7.7,1.4 Feedwater Control System Instrumentation and Controls 7.7 27 ; 7.7.1.5 Process Computer Sptem (PCS)- Instrumentation and Controls 7.7 31 7.7.1.5.1 Power Generation and Control System. Instrumentation f and Controls 7.7 34 7.7.1.6 Neutron Monitoring System . Non Safety : Related Subsystems 7.7 34 , 7.7.1.6.1 Automatic Traversing In-Core Probe (ATIP) 7,7 34.1 i 7.7.1.6.2 Multi. Channel Rod Block Monitor (MRBM) 7.7 34.1 7.7.1.7 Automatic Power Regulator System. Instrumentation and Controls 7.7 34.2 i 7.7.1.8 Steam Bypass and Pressure Control System. Instrumentation and Controls l 7.7 34.4 l l 7.7.1.9 Non.Essentlal Multiplexing System 7.7 34.7 i k 7.7.1.10 Other Nonsafety-Related Control Systems. , O Instrumentation and Controls 7.7.i1 7.7 35 t Amendnwnt 11
L ABWR name ! l standard Plam' uv. 9 ! SECTION 7.7 g { CONTENTS (Continued) ! EMnta DOS EAR f i 7.7J Anshals 7.7 35 l l 1' l 7.7.2.1 Nuclear Boiler System . Reactor Vessel instrumentation - 7.7 35 j 7.7.2.1.1 General Fuaataamt Requirements Confor aance 7.7 35 f 7.7.2.1.2 SpecHic Repletory Requirements Coniormance 7.7 35 t l i 7.7.2.2 Rod Control and Informa%s System . l Instrumentation and Controls 7.7 36 7.7.2.2.1 General Functional Requirements Conformance 7.7 36 7.7.2.2.2 Specific Regulatory Requirements Conformance 7.7 36 7.7.2.3 Recirculation Flow Control System . Instrumentation and Controis 7.7 37 I 7.7.2.3.1 General Functional Requirements Conformance 7.7 37 7.7.2.3.2 Specific Regulatory Requirements Conformance 7.7 37 7.7.2.4 Feedwater Control System . Instrumentation ' [ and Controls 7.7 37 7.7.2.4.1 General Functions: Requirements Conformance 7.7 37 7.7.2.4.2 Specific Regulatory Requirements Conformance 7.7 38 - [ 7.7.2.5 Frocess Computer System . Instrumentation and Controls 7.7 38 7.7.2.5.1 General Functional Requirements Conformance 7.7 38 h t 7.7.2.5.2 Specific Regulatory Requirements Conformance 7.7 38 7.7.2.6 Neutron Monitoring System . ATIP Subsystem lastrumentation and Controls 7.7 38 7.7.2.6.1 General Functional Requirements Conformance 7.7 38 f 7.7.2.6.2 Specific Regulatory Requirementi Conformance 7.7 38 f i 7 . 7 . i11 Amendment 11
I 2M6100AF ElemmAard Plant miy 3 O sociioN v., i CONTENTS (Continued) ! Amdan Due P.ast 7.7.2.7 Automatic Power Regulator System. Instrumentation and Controis 7.7 38.1 7.7.2.7.1 General Functional Requirements Conformance 7.7 38.1 7.7.2.7.2 Specific Regulatory Reutrements Conformance 7.7 38.1 . 7.7.2.8 Steam Bypass and Pressure Control System. Instrumentation and Controis 7.7. 38.1 7.7.2.8.1 General Functional Requirements Conformance 7.7 38.1 i 7.7.2.8.2 Specific Regulatory Requirements Conformance 7.7 38.1 , I 7.7.2.9 Non. Essential M ultiplexing System. Instrumentation and Centrols 7.7 38.2 7.7.2.9.1 General Requirements Conformance 7.7 38.2 7.7.2.9.2 Specific Regulatory Requirements Conformance 7.7 38.2 7.7.2.10 Other Nonsafety Related Control Systems 7.7 38.2 :
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k i i l l 7.7.lv Amendment 11
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- ABWR nuine standard Plant arv. n i TABLES -
Inble Ilde East l 7.7 1 - Deleted ILLUSTRATIONS Ilde Hein F.ast i i 7.71 Water Level Range Definition . 7.7 41. 7.72 Rod Control and Information System IED 7.7 42 7.73 Rod Control and Information System IBD 7.7 45 s
- 7.74 ControlRod Drive System IBD 7.7 54 1
7.75 Recirculation Flow Control System IED 7.7 61 7.7-6 RFC System Algorithm Structure 7.7-63 7.77 Recirculation Flow Coctrol System IBD - 7.7 64-7.78 Feedwater Control System IED - ~ 7.7 73 - 1 7.79 Feedwater ControlSystem IBD 17.7 75 7.7 10 Ataignment of LPRM Strings to TIP Machines '7.7 87 l 7.7 11 Automatic Power Regulation System Block Diagram 7.7 88 7.7 12 Steam Bypass and Pressure Control System IED '7.7-89 7.7 13 Steam Bypass and Pressure Control System IBD 7.7 92 {; i l l r l i-i l 7.7 v l Amendment 11 L
ABM is46ioorr Standard Plani REV.A O tion calculation is determined and is used processor provides the capability to: ( to update the distribution of cumulative alarm the main control room annunciator fuel exposure. Each fuel bundle is iden- system in the event of abnormal PCS tified by batch and location, and its expo- operation, sure is stored for each of the axial seg-ments used in the power distribution calcu - (b) Trip / Scram Data Recall Logging . lation. These data are printed out on i operator demand. Exposure increments are The processor measures and stores the , determined periodically for each quarter- values of a set of analog variables at , , length section for each control rod. The predefined intervals to provide a . corresponding cumulative exposure totals are history of data. An on demand request periodically updated and print 6d out on permits the operator to_ initiate > operator demand. . printing of this data and to terminate l the log printout when desired. The exposure increment of each local power . range monitor is determined periodically and (c) Trend Logging , is used to update both the cumulative ion chamber exposures and the correction factors An analog trend capability is provided for exposure dependent LPRM sensitivity for logging the values of the operator-loss. These data are printed out on selected analog inputs and calculated ^ operator demand, variables. The periodicity of the log is limited to a nominal selection of l The computer provides online capability to intervals, which'can be adjusted as - determine monthly and on. demand isotopic desired by program control, composition for each fuel bundle in the core. This evaluation consists of computing (d) Status Alarm !' the weight of one neptunium, three uranium, l s and five plutonium isotopes as well as the- Thc~ status alarm of a point shall be - l total uranium and tota'l plutonium content, updated.with a time after occurre' ace - , The isotopic composition is calculated and equal to the processing cycle of the ! l summed accordingly by bundles and batches, point plus two seconds.' A printed l record of system alarms is provided (8) Reactor Operation Information (Monitor, which includes point description and Alarm, and Logging Programs) time of occurrence. (a) General (e) Alarm Logging The processor is capable of checking The alarm logs required by the associ-each analog input variable against two ated process programs are printed, types of limits for alarming purposes: Alarm printouts' inform the operator of computer system malfunctions, system (1) Process alarm limits as determined operation exceeding acceptable limits, C by the computer during computation and unreasonable, off normal, or failed or as preprogrammed at some fixed input sensors. value by the operator; and
- 1. . (9) BOP Performance Calculation Prcy; rams (2) A reasonableness limit of the analog ;
input signal level programmed. These programs perform calculations and log. ging of plant performance data not direct-The alarming sequence consists of an ly related to the nuclear system. The data
- audible alarm, a console alarm, and a stored by the BOP program is printed out on descriptive message for the variables logs. The BOP periodic log gives hourly that exceed process alarm limits. The p)- \_
and daily values for temperatures, power Amendment 2 7.7 33
k j l' [M 23A610W I handard Plant ma t outputs, and flows associated with the main . semi. automatic mode, the PGCS provides -. , generator and turbines and with the feed. guidance messages to the operator to carry , water, recirculation,'and reactor water out the startup, shutdogwn, and power range i k cleanup systems. The BOP monthly log con. operations tains monthly averages and accumulations for plant gross and act power Mtputt !oad dis- (2) Classification tributions, turbine heet %g ad fuel burn . up. BOP performatc& :;siculhions include The PGCS is classified as a power generation
' flow calculation % clMaicd calculations, system and is not required for safety, thermodynamic twiculaNs, nuclear boiler Safety events requiring control rod scram l:
system pe:formance calculations, turbine su censed and controlled by the cycle redormance calculations, condenser safety sciated reactor protection _ system , p calculation, feedwater heaters and moisture (RPS), which is completely independent of t ! separators performance calculations, and unit . the PGCS. The RPS is discussed in Section performance calculations. 7.2 7.7.1J.1 Power Generation Control System. 3) Power Sources , Instrumentation and Controls
- _ .The PGCS.is a separate part of the plant-(1) System identification process computer. The plant ~ process computer is-powered by redundant. _
The power generation control system (PGCS) is uninterruptable non Class 1E power supplies - a top level controller that monitors the and sources. No single power failure will overall plant conditions, issues control result in the loss of any PGCS function. ; commands and adjusts setpoints of lower level controllers to support automation of the (4) NormalOperation normal plant startup, shutdown, and poer range operations. The PGCS is a separate _The PGCS interfaces with the operator's function of the power plant process computer console to perform its' designated, -
-1 system, it contains the algorithms for the . functions. The operator's control console automated control sequences associated with for PGCS consists of a series of breakpoint plant startup, shutdown, and normal power controls for a prescribed plant operation range operations, it issues reactor command sequence. When all the prerequisites are signals to the automatic power regulator satisfied for a prescribed breakpoint in a - 4 (APR) system. The reactor power change control sequence, a permissive is given and algorithms are implemented in the APR system upon verification by the operator, the as discussed in Subsection 7.7.1.7. operator initiates 'het prescribed control sequence. The power generation control '
The PGCS issues command signals for turbine, system then initiates demand signals to feedwater, and related auxiliary systems to various system controllers to carry out the. the turbine master controller which contains predifined control functions. [ Note: for appropriate algorithms for automated non automated operations that are required * - l-sequences of these systems. Command signals during normal startup or shutdown (e.g., for setpoint adjustment of lower level change of reactor mode. switch status), controllers and for startup/ shutdown of other automatic prompts are provided to the systems required for plant operation are operator. Automated operations continue executed by the PGCS. The operator after the operator completes the prompted interfaces with thw PGCS through a series of action manually.) r breakpoint controls to initiate automated sequences from the operator control console. 7.7.1.6 Neutron Monitoring System - For selected operations that are not Non. Safety Related Subsystems - automated, the PGCS prompts the operator to perform such operations, in the , Amendment 11 7.7.M
ABWR mew nry n
~ Standard Plant u
7.7.1.6.1 Automatic Traversing In Core Probe (2) Classification i
.O (ATIP) b This subsection describes the non safety re -
The ATIP is nonsafety related as shown in Table 3.21. The subsystem is an opera. i 1 l lated automatic traversing in core probe (A flP) tional system and has no safety function. l subsystem of the neutron monitoring sysicm (3) Power Supply l discussed in Subsection 7.6.1.1.(NMS). Safety related NMS subsystems are
- The power for the ATIP is supplied from the (1) Description instrument AC power source.
The ATIP is comprised of three TIP machines, (4) Testability each with a neutron sensitive sensor attach-l cd to the machine's flexible cable. Other The ATIP equipment is tested and calibrated s than the sensor itself, each machine has a using heat balance data and procedures drive mechanism, a twenty position index described in the instruction manual, mechanism, associated guide tube, and other parts. While not in use, the sensor is not. (5) EmironmentalConsiderations mally stored and shielded in a storage area inside the TIP room in the reactor build. The equipment and cabling located in the ing. During operation, the ATIP sensors are drywell are designed for continuous duty inserted, either manually or automatically, (see Section 3.11). ~ via guide tubing and through desi+ d index positions to the designated LPRM assembly (6) OperationalConsiderations calibration tube. Each ATIP machine has de. signated number and locations of LPRM assem. The ATIP can be operated during reactor blics to cover, such that the ATIP sensor can operation to calibrate the LPRM channels. j O V travel to all LPRM locations assigned to this - machine via the index mechanism of this The subsystem has no safety setpoints. , machine, The LPRM assignments to the threc 7.7.1.6.2 Multi Channel Rod Block Monitor i machines are shown in Figure 7.710. (MRHM) Flux readings along the axial length of the This subsection describes the non safety core are obtained by first inserting the related multi channel rod block monitor (MRBM) sensor fully to the top of the calibration subsystem of the neutron monitoring system I tube and then taking data as the sensor is (NMS). Safety related NMS subsystems are withdrawn continuously from the top. Sensor discussed in Subsection 7.6.1.1. flux reading, sensor axial positions data in 1 the core, and LPRM location data are all sent (1) System identification to an ATIP control unit located in the control room, where the data can be stored. The multi channel rod block monitor (MRBM) The data are then sent to the process subsystem logic issues a rod block signal computer for calibration and performance that is used'in the rod control & calculations. The whole' ATIP scanning information system (RC&lS) logic to enforce sequence and instructions are fully rod blocks that prevent fuel damage by y automated, with manual control available, assuring that the minimum critical poer ratio (MCPR) and maximum linear heat The index mechanism allows the use of a generation rate do not violate fuel theimal single sensor in any one of twenty different ' safety limits. Once a rod block is LPRM assemblics. There is a common LPRM initiated,' manual action is required by the location that allows all three ATIP operator to reset the system, scanning. This is for ATIP cross machine calibration. Amendment 11 7.7 >t,t
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k N 23A6100AF am n - ReanAard Plant
- . The MRBM micrcenputer based logie receives physically and electrically isolated from .
loput signals from the local power range the rest of the safety NMS subsystems. All ' monitors (LPRMs), and the average power range laterfaces with the safety NMS subsystems r monitors (APRMs) of the NMSc It also are via optical isolation. receives core flow data from the NMS, and control rod status data from the rod action- 7.7.1.7 Automatic Power Regulator 8ystem. , and position information subsystem to Instammentation and Controls i determine when rod withdrawal blocks are l . required. The MRBM averages the LPnM signals (1) Identification - ,
- to detect local power change during the rod withdrawal. If the averaged LPRM signal The Primary objective of the automatic power l exceeds a preset rod block setpoint, a regulator (APR) system is to controi reactor control rod block demand w"I be issued. The ' power during reactor startup, power MRBM monitors many 4-by 4 fuel bundle regions generation, and reactor shutdown, by
- i in the core in which control rods are being appropriate commands to change rod l
withdrawn as a gang. Since it monitors more positions', or to, change reactor than one region, it is called thc . recirculation flow. The secondary objective ' i iruiti channel rod block monitor. The rod of the APR system is to control the pressure block setpoint is a core flow biased variable regulator setpoint (or turbine bypass valve , setpoint. The MRBM is a dual channel, highly position) during reactor heatup and .i reliable system, but not classified as a depressurization (e.g., to control the i safety system. A block diagram showing a reactor cooldown rate). The automatic power typical MRBM system is shown in Figure regulator system consists of redundant < 7.7 10b. process controllers. Automatic power i regulation is achieved by appropriate (2) Classification control algorithms for different phases of the reactor operation which include approach ! The MRBM is non safety related. Its to criticality, heatup, reactor power- , l activating interface is through the rod increase, automatic load following, reactor . < control and information system (RC&lS) which . power decrease, and reactor depressurization is also a non safety related system. and cooldown. The automatic power regulator
- system receives input from the plant process-l (3) Power Supply computer, the power generation control '
L system (described in Subsection 7.7.1.5.1), I The power supply for the MRBM is from the the steam bypass and pressure control system' , non divisional 120 VAC UPS bus. '(described in Subsection 7.7.1.8), and the operator's control console. The output (4) Testability demand signals from automatic power regulation system are to the rod control and ' The MRBM is a dual channel, independent information' system to position the control , i subsystem of the NMS. One of the MRBM rods, to the recirculation flow control 1 (~ channels can be bypassed for testing or system to change reactor coolant i l maintenance without affecting the overail recirculation flow, and to the steam bypass ! MRBM function. Self test features are and pressure control system for automatic i' employed'to monitor failures in the load following operations. The power
- microprocessor system. Test capabilities generation cystem performs the overall plant -
allow for calibration and trip output startup, power operation, and shutdown l testing. functions. The automatic power regulation system performs only those functions (5) Environmental and Operational Considerations associated with reactor power changes and [ with pressure regulator setpcint (or turbine The MRBM is located in the control room bypass valve position) changes during adjacent to the APRM panels. It is reactor heatup or depressurizatien. A Amendment 11 7.7 41.2 {
.a _ _ . _ . - . - - -
ABWR == mandmed Plant an a simplified functional block diagram of the power regulation system contains algorithms 1 automatic power reguh tion system is provided that can change reactor power by control rod in Figure 7.711. motions, or by reactor coolant recirculation-flow changes, but not both at the same time. (2) Classification A prescribed- controf rod sequence is followed when manipulating control rods for The APR is classified as power generation reactor criticality, heatup, power changes, l system and is not required for safety, and automatic load following. Each of these - Safety events requiring control rod scram are functions has its own algorithm to achieve sensed and' controlled by the safety related its designed objective. Tbc control rod reactor protection system (RPS), which is sequence can be updated from the process completely independent of the APR. The RIPS computer based on inputs from the reactor is discussed in Section 7.2. engiacer.' A predifined trajectory of power flow is followed when controlling (3) Power Sources reactor power. The potentially unstable region of the power flow map is avoided The automatic power regulation system digital during plant startup, automatic load , controllers are powered by redundant following, and shutdown. During automatic uninterruptable non Class 1E power supplies load following operation, the automatic - 3 I l and sources. No single power failure shall power regulation system laterfaces with the-result in the loss of any automatic power steam ~ bypass and pressure control system - regulation system function, to coordinate main turbine and reactor power , changes for optimal performance. (4) NormalOperation (5) AbnormalOperation-The automatic power regulation system interfaces with the operator's console to The_ normal mode of operation of the O- perform its designed functions. The automatic power reguletor system is. operator's control panel for automatic plant automatic. .If any system or component startup, power operation, and shutdown conditions are abnormal during execution of functions is part of the power generation the prescribed sequences, the power control system. This control panel consists generation control system will be of a series of breakpoint controls for a automatically switched into the manual mode ! prescribed plant operation sequence. When and any operation in prcgress will be all the prerequisites are satisfied for a - stopped. Alarms will be activated to alert prescribed breakpoint in a control sequence, the operator. With the automatic power a permissive is given and upon verification regulation system in manual . mode, the by the operator, the operator initiates the operator can manipulate control rods and prescribed control sequence. The power recirculation flow through the normal-generation control system then initiates controls. A failure of the automatic power demand signals to various system controllers regulation system will not prevent manual to carry out the predefined control- controls of reactor power, nor will it l functions. [ Note: For non automated prevent safe shutdown of the reactor, operations that are required during normal startup or shutdown (e.g.. change of Reactor (6) Equipment . Mode Switch status), automatic prompts are provided to the operator. Automated The automatic power regulation system i operations continue after the operator control functional logic is performed by completes the prompted action manually.] The redundant, microprocessor based functions associated with reactor power fault tolerant digital controllers (FTDC), control are performed by the automatic power The FTDC performs many functions. It reads regulation system. and validates inputs from the non essential Q multiplexing system (NEMS) interface once Q For reactor power control, the automatic every sampling period. It performs the Amendment 11 7.7.R3
f ABM awionar nev a me=Aard Plant specific ^ power control calculations and will stop automatic reactor power changes, processes the pertinent alarm and laterlock - If any system or component conditions are _ functions, then updates all system outputs to, abnormal during execution of the prescribed the NEMS._ To prevent computational sequences, continued operation is stopped divergence among the redundant processing automatically and alarms will be activated - channels, each channel performs a comparison to alert the operator. With the automatic check;of its calculated results with the - power regulation system in manual mode, the : ;
. other redundant channels. The laternal FTDC operator can manipulate control rods and architecture features redundant multiplexing recirculation flow through the normal laterfacing units for communications between- controls. A failure of the automatic power J the NEMS and the FTDC processing channels.- regulation system will not prevent manual - 1 controls of reactor power, nor will it j (7) Testability prevent safe shutdown of the reactor. 1 The fault tolerant digital controller (FTDC), (1) Setpoints input and output communication laterfaces, i are continuously functioning during normal The automatic power regulation system has no l power operation. Abnormal operation of these safety setpoints.
l components can be detected during operation. _ In addition, the FTDC is equipped with 7.7.1.8 Steam Bypass & Pressure Control System-self test and on.line diagnostic capabilities - Instrumentation and Controls l for identifying and isolating failure of- ) input / output devices, buses, power supplies, (1)ldentification processors, and interprocessor communication j I paths. These on line tests and diagnosis can The primary objective of the steam bypass & be performed without disturbing the normal pressure control (SB&PC) system is to I control functions of the automatic power control reactor system pressure during plant regulation system. startup, power generation and shutdown modes of operation. This is accomplished through (8) EnvironmentalConsiderations control of the turbine control and/or steam
. bypass valves, such that susceptibility to
- The automatic power regulation system is not reactor trip, turbine' generator trip, main-l- required for safety purposes, nor is it steam isolation and safety relief valve l required to operate during or after any opening is' minimized. I design basis accident. The system is I required to operate in the normal plant Command signals for the turbine control environment for power generation purposes valves and the steam bypass valves are ,
only. The automatic power regulation system . generated by a triplicated fault tolerant l equipment is located in the main control room digital controller using feedback signals - I and subject to the normal control room from vessel pressure sensors. For normal environment as listed in Section 3.11. operation, the turbine control valves regulate steam pressure. ' However, whenever (9) Operator Information and Operational the' total steam flow demand from the Considerations pressure controller exceeds the effective turbine control valve steam flow demand, the During operation of the automatic power pressure control system sends the excess . regulation system, the operator observes the steam flow directly to the main condenser, performance of the plant via CRTs on the main - through the steam bypass valves. console or on large screen displays in the main-control room. The automatic power Ability of'the plant to follow grid system regulation system can be switched into the load demands is enabled by adjusting reactor manual mode by the operator and a control power level, by varying reactor sequence, which is in progress, can be recirculation flow (manually or stopped by the operator at any time. This automatically), or by moving control rods Amendment 11 7.7-34.4 1
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ABM 234sioorr ma
' Standard Plant (manually or automatically). In response to The SB&PC also controls pressure during the resulting steam production changes, the ' normal (main steam isolation valves open) ;L 'g') pressure control system adjusts the turbine reactor shutdown to control the reactor l-control valves to accept the steam output cooling rate, change, thereby controlling steam pressure.
, In addition, when the rcactor is (5) AbnormalPlant Operation automatically following grid. system load l demands, the pressure control system permits Events which induce reactor trip present- ( I. an immediate steam flow response to fast _ significant transients during which the ! changes in load demand, thus utilizing part SB&PC must maintain steam pressure. These of the stored energy in the vessel, transients are characterized by large o variations in vessel steam flow, core l (2) Classification thermal power output, and sometimes recirculation flow all of which affect The SB&PC is classified as a primary power vessel water level. The SB&PC is designed generation system. Its function is not to respond quickly to stabilize system required for safety, but its operation is pressure and thus aid in the feedwater/ level essential to the power production cycle. No control in maintaining water level, explicit seismic requirements are imposed for the SB&PC system equipment. -The SB&PC is also designed for operation with other reactor control systems to avoid * (3) Power Sources reactor trip after significant plant disturbances. Examples of such disturbances The SB&PC controls and bypass valves are are loss of one feedwater pump, loss of powered by redundant uninterruptable three recirculation pumps, inadvertent non Class IE power supplies and sources. - No opening of safety relief valves or steam single power failure will result in the loss- bypass valves, main turbine stop/ control () of SB&PC system function. Upon failure of valve surveillance testing, and steam line 1 () two or more channels in the controller, the isolation valve testing. turbine will trip. (6) Equipment ; (4) NormalPlant Operation i The SB&PC system control functional logic is At steady state plant operation, the SB&PC performed by triplicated microprocessor-l system maintains primary system pressure at a based fault tolerant digital controllers L nearly constant value, to ensure optimum (FTDC) similar to those used for the plant performance, feedwater and recirculation flow control . systems. It is therefore possible to lose During normal operational plant maneuvers one complete processing channel without (pressure setpoint changes, level setpoint impacting the system function. This also changes, recirculation flow changes), the facilitates' taking one channel out of SB&PC system provides responsive, stable service for maintenance or repair while the performance to minimize vessel water level system is on line. The IED and IBD are 1 and neutron flux transients, provided as Figures 7.712 and 7.713 respectively.- During plant startup and heatup, the SB&PC provides for automatic control of the reactor Controls and valves are designed such that system pressure. Independent control of steam flow is shut off upon loss of control reactor pressure and power is permitted, system electrical power or hydraulic system j during reactor vessel heatup, by varying pressure. - steam bypass flow as the main turbine is brought up to speed and synchronized. . The pressure control function provides ABWR automatic load following by forcing the O turbine control valves to remain under
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Amendment 11 7.7.M.5 T
__ ~ _ _ . __ . ._ __ -. -__ MM. 23A6100AF-nev.n l Atandan! Plant i pressure control supervision, while enabling (f) Displayed variables and alarms from the fast bypass opening for transient events SBAPC to the main control room panel - requiring fast reduction in turbine steam operator interface. flow. ' (g) Narrow and wide range' pressure signals, _ The steam bypass function controls reactor - MSIV position ~ signals from the nuclear - . pressure by modulating three automatically boiler system to the SBAPC. . operated, regulating bypass valves in , L response to the bypass flow demand signal. (h) Bypass valve position, servo current, i l This control mode is assumed under the ' position error and valve open and following conditions: closed signals from the turbine bypass - system. (a) . During reactor vessel heat up to rated pressure. (i) Emergency bypass valve fast opening > signals and bypass valve flow demand (b) While the turbine is brought up to speed ' signals from the SB&PC to the turbine - and synchronir.ed. bypass system. (c) During power operation when reactor (i) Electric power from the non Class 1E steam generation exceeds the turbine power supply to the SB&PC. q steam flow requirements. (k) Pressure setpoint change
- (d)L During plant load rejections and requests / commands from the turbine turbine generator trips, master controller, for automatic l start up and shut down sequences.
l_ (c) During cool down of the nuclear boiler. _ l' (1) Governor free demand signal to the: (7) C&IInterface reactor power compensator in the - ' t l automatic power regulator (APR) system. i The external signal interface for the SB&PC l are listed as follows: (m) Reactor power compensation signal in- - accordance with speed error from the - (a) Narrow range dome pressure signals from SB&PC to the APR. ! I- -the SB&PC system to the recirculation flow control system. (n) Main condenser vacuum low signal from , the extraction system to the SB&PC. (b) Equivalent load or steam flow feedback i signal from the turbine control system (7) Testability I (which is included in the same i triplicated fault tolerant digital The FTDC input and output communication I controller as the SB&PC). interfaces are continuously functioning during normal power operation. Abnormal (c) Signals to and from the main control operation of these components can be room. - detected during operation. In addition, the FTDC is equipped with self test and on.line i (d) Bypass hydraulic power supply trouble diagnostic capabilities for identifying and I signal from the turbine bypass system to isolating failure of input / output devices, the SBAPC system. buses, power supplies, processors, and interprocessor communication paths. These (c) Output signals from the SB&PC to the on line tests and diagnosis can be performed performance monitoring and control without disturbing the normal control function of the process computer. functions of the SB&PC system. [ Amendment 11 7.74t6
ABM nisioorr Standard Plant arv. n . l f
\
(8)- Environmental Considerations signal by the normal bypass demand. The system automatically returns to the manualL l l The SB&PC system is not required for safety demand signal when pressure transient purposes, nor is it required to operate' causing the increased bypass demand is during or after any design basis accident, relieved. The system is required to operate in the normal plant environment for power generation In order to preserve steam for the main purposes only. The SBAPC system equipment is turbine gland seal functions, the bypass located in the main control room'and subject valves are inhibited from opening when to the normal control room environment as ~ either the inboard or outboard MSIV's close listed in Section 3.11. .to their 90% positions. This bypass inhibit condition is annunciated in the main control (9) Operatorlaformat ion room and must be manually reset by the operator. Any plant or component condition During operation of the SB&PC, the operator' that inhibits bypass valve opening is may observe the performance of the plant via . annunciated. CRTs on the main control console or on large screen displays in the main control room. As (11) Setpoints described in (7) above, the self test . 1 provisioa assurea t h a t' aIi The SB&PC has no safety setpoints because it transducer / controller failures are indicated is not a safety system. Pre operational to the operator and maintenance personnel. . setpoints and design parameters for the , The triplicated logic facilitates on line power generation functions are identified in repair of the controller circuit boards. the system design' specifications (see Subsection 1.1.3). Actual operational (10) Operational Considerations setpoints will be determined for each During abnormal conditions that result in low main condenser vacuum, the steam bypass 7.7.1.9 Non Essential Multiplexing System valves and main steam isolation valves close to prevent positive pressure conditions that The non essential multiplexing system (NEMS) would rupture main condenser diaphragms, is separate and distinct from-the essential Manually operated provisions permit opening multiplexing system (EMS), though both are of the main steam isolation valves (i.e., similar in design and architecture. .Except for . inhibit the closure function) during startup systems interfaces and quality assurance c operation. This vacuum protection function requirements unique to Class IE' systems,
- l. bypass permits heat up of the main steam specific design attributes discussed in Appendix l lines (up to the steam bypass valves and 7A.2 pertain to the NEMS as well. Both systems l
turbine stop valves) before normal condenser are fully described in their subsection design. vacuum is obtained and permits cold shutdown specifications available from the Master Parts testing of the isolation valves. ' List referenced in Subsection 1.1.3. This subsection. describes those features which are The steam bypass system allows remote manual unique to the NEMS, bypass operation in the normal sequenes ~ during plant start up and shut-down.' This (1) System Description I facilitates purge of-the vessel and main steam lines of accumulated non condensable The NEMS provides distributed control and gases early on in the start up process, and instrumentation data communication networks controls the rate of cooling during reactor to support the monitoring and control of. shutdown to atmospheric pressures. Upon interfacing plant power generation ,' increasing pressure transients during such (non-safety related) systems. [The manual operation, the controls provide essential multiplexing system (EMS) performs automatic override of the manual demand the same function for the protection (safety L Amendment 11 7.7-X7 '
MM 23A6100AP - Standmed Plant arv.n'
.I related) systems.] The NEMS provides all the (2) SystemInterface .j electrical devices and circuitry (such as multiplexing units, data transmission line .. The NEMS Interfaces with the following '
and transmission controllers), between systems, which are all non safety related: sensors, display devices, controllers and actuators, which are defined by other plant Reactor . systems.;The NEMS also includes the Nuclear'Soiler (non safety related associated data acquisition and communication portion) ' i software required to support its function of Reactor Recirculation transmitting plant wide data for distributed Rod Control . control and monitoring.- Feedwates Control (including feedwater - pump turbine) The NEMS acquires both analog and digital . Recirculation Flow Control signals from remote process sensors and Steam Bypass and Pressure control . discrete monitors located within a plant, and Performance Monitoring and Control' , multiplexes the signals to a central control Power Generation Control ' room to drive annunciators, saonitors and Process Radiation Monitoring . recorders, and to send signals, and output (non safety related portion) control signals are multiplexed to actuators,- Area Radiation Monitoring-- valves, motor, drives and other control Dust Radiation Monitoring equipment in the plant associated with Refueling and Reactor Servicing non safety related systems.- Reactor Water Cleanup
- Fuel Pool /Wetwell Cleanup (excluding Consistent with fault tolerant (triplicated) cooling) digital control systems utilized in feedwater Control Complex.' J control, reactor recirculation flow control Makeup Water (purified, condensated) :
and steam bypass and pressure regulation, the HVAC NormalCooling Water , NEMS is also triplicated for these systems Ultimate Heat Sink . interfaces, as appropriate, each with its own Turbine Service Water , independent control. Steam and Heated Water Compressed Gas The remaining communication functions of the Sampling i - NEMS provides the following system functions: Condensate Demineralizer/ Filter
. Facility -
(a) Acquires non safety related data, e.g., Radwaste (includes Offgas) sensed input and equipment status Turbine Bypass signals, throughout the plant. Turbine Control Feedwater Condenset: Water L (b) Conditions, formats and transmits Heater Drain signals via fiber optics to displays, Lubricating Oil controllers, and the performance Turbine Gland Steam . monitoring and control system (PMCS). Extraction
. Main Generator (c) Receives signals via fiber optics, then HVAC Reactor Building multiplexes and prepares them for use in HVAC-Other Buildings . )
interfacing non safety related equipment Electrical Power Distribution i as required. (non safety related portion) Annunciator , (d) Formats and transmits processed control signals via fiber optics to actuator (3) Classification
. circuits, and then converts the fiber optic control signals to electrical The NEMs, of itself, is neither a power signals for the actuator circuits, generation system nor a protection system. - i i
Amendment 11 7.7-R8 l _ _ _ _ _ _ _ _ _ . _ _ _ . _ __ . . . , , .. . .- . , ..m. . . - -
23A6100AP Standard Plant REv.n .i e It is a-support system utilized for are required in that the system is capable i' .-- assimilation, transmission and interptetation of seIf starting foilowIng power of data for power generation (non safety interruptions, or any other single failure, l related) systems and their associated including any single processor failure. l sensors, actuators and interconnections, it After repairs or replacements are performed,
.is classified as non safety related. the system automatically re initializes'to ,
normal status when power is restored to any. (4) Power Sources ' unit and automatically resets any alarms. The NEMS receives its. power from non-Class IE (9) OperatorInformation - distribution branches of the same multi divisional 125 VDC power supply which The 'self test provisions are designed to feeds the EMS. Isolation is accomplished - alert the operator to system ' anomalies via through the use of DC to DC convertors (see interfaces with the process computer and the - Figure 8.3 7). This redundancy allows the annunciator. Problems significant'enough to NEMS to supply triplicated logic functions ~ cause system channel failures are such that any single failure in the system annunciated separately from those which power supplies will not cause the loss of the allow continued operation. ,The circuitry is' , validated outputs to the interf acing designed'such that no control output or , actuators and to the' monitors and displays, alarm is inadvertently activated during system initialization or shutdown ;For such 5 (5) Equipment - events, control outputs change'to ' predetermined f ail safe outputs. The hardware and "firmware* architectures for the NEMS are the same as those of the EMS,
- which are described in Appendix 7A. [ See g the response to NRC Requests (10) and (11) of Section 7A.2.]
(6) Testability &
]
The EMS test feature described in Appendix 7A, Section 7A.2, Items (3), (4) and (6) are generally equivalent for the NEMS, except that the NEMS does not interface with, nor
- rely upon, the SSLC [see the response to b7C Request (6) of Section 7A.2). Also, the NEMS self test features include the analog fault tolerant voting system unique to the control systems employing logic, t
(7) EnvironmentalConsiderations The NEMS is not required for safety purposes, nor is it required to operate after the i design basis accident. Its support function serves power generation purposes only and it ; is designed to operate in the normal plant environment.
- (8) Operational Considerations -
Ti.c system automatically initiates for both i cold and warm starts. No operator actions Amendment 11 7.7 M.9 l
4 dEM . 2M6100AF Standard Plant uv. n p 7.7.1.10 Other Nonastety Related Control For consideration of item (2), above, it is e e
.. S7 Las - necessary to refer to the safety evaluations in Chapter 15. In that chapter it is first shown
-The following'nonsafety related control that the subject systems are not utilized to i systems are described in other subsections of the provide any design basis accident safety func. - j l SSAR as indicated. tion. Safety functions, where required, are >
l- provided by other qualified systems. For ex- !
- $gra subsection pected or abnormal transient incidents follow- 1 ing the single operator error (SOE) or single l Fire Protection 9.5.1 component failure (SCF) criteria, protective i functions are also shown to be provided by other l Ofigas/Radwaste 11.2, 113,11.4 systems. - The expected or abnormal transients cited are the limiting events for the subject _ ;
Drywell Coohng 9.4.8 systems. ; Sampling - 93.2 7.7.2.1 Nuclear Boller System ReactorVessel _! Instrumentation !
- lastrument Air 93.6 i 7.7J.1.1 General Functional Requirements ,
Makeup Water 9.23- Conformance i Atmospheric The reactor vesselinstrumentation of the nu- , l Control 6.2.5 clear boiler system (NBS) is designed to provide ! redundant or augmented information to the exist. 7.7.2 Analysis ing information required from the engineered ; safeguards and safety related systems. None of l The purpose of this subsectionis to: this nonsafety related instrumentation is- i regt. ired to initiate or control any engineered , V (1) Demonstrate by direct or referenced analysis safeguard or safety related system function. I that the subject described systems are not required for any plant safety function. _7.7.2.1.2 Specific Regulatory Requirements - Conformance (2) Demonstrate by direct or referenced analysis , that the plant protection systems described - Table 7.12 identifies the nonsafety related elsewhere are capable of coping with all control systems and the associated codes and failure modes of the subject control system. standards applied in accordance with Section 7.7 of the Standard Review Plan for BWRs. The In response to item (1) above, the following' following analysis lists the applicable criteria is cited: upon considering the design basis, de- in order of the listing on the table, and scriptions, and evaluations presented here and discusses the~ degree of conformance for each. elsewhere throughout the document relative to the Any exceptions or clarifications are so noted. 4 subject system, it can be concluded that these < systems do not perform any safety related (1) General Design Criteria (GDC): function. (a) Criteria GDCs 13 and 19. Design Basis: Refer to Subsection 7.1.1. (b) Conformance: The NBS is in compliance
Description:
Refer to Subsection 7.7.1. with these GDCs, in part, or as a whole, as applicable.' The GDCs are gen-The individual system analysis in this section erally addressed in Subsection 3.1.2. concludes that the subject systems are not required for any plant safety action. (2) RegulatoryGuides(RGs): In accordance with the Standard Review Plan Amendment 11 7.7 35
. _ _ _ _ _ - ~ _ _ _ _ . .- -. . - . - - ._-. - - - _ - . . . . - .
ABM nisioore
' Standard Plant arv. n for Section 7.7 and with Table 7.1' 2, only The-NBS lastrument lines are not
- Regulatory Guide 1.151 Instrument Sens- exposed to cold temperatures and are ing Lines need be addressed for the ABWR, designed to meet the ASME code j requirements of Regulatory Guide 1.151 i (a) fahrJia: RG 1.151 Instrument Sens- and ISA $67.02.
Ing Lines The nuclear boiler system is thus in (b)' fJa[grmance: There are four independent full compliance with these criteria. sets of instrument lines which are me-chanically separated into each of the 7.7JJ Rod Control and Information System -
' four lastrument divisions of the nuclear lastrumentation and Controls boiler system (see Figure 5.13, Nuclear .
3 Boiler System P&ID). Each of the four - 7.7.2.2.1 GeneralFunctional Requirements instrument lines interfaces with sensors conformance! l assigned to each of.the four Class 1E . . - . . .. . electrical divisions for safety related The circuitry described for the rod control systems, and information system (RC&lS) is completely independent of the circuitry controlling the- 4 There are also non Class 1E instruments scram valves. This separation of the scram and- , that derive their input for the reactor normal rod control functions prevents failures ; vesselinstrumentation portion of the in the rod control and information circuitry- , NBS from these lines. There is no from affecting the scram circuitry. The scram j safety related controlling function in. circuitry is discussed in Section 7.2. Because j volved in this instrumentation and it is - each control rod is controlled as an individual < entirely separate (including its own MUX unit,'a failure that results inl energizing of. I system) from the safety related instru - any. of the insert or withdraw solenoid valves ments and their associated systems. can affect only one control rod. The effective- ) ness of a reactor scram is not impaired by the .! The safety related instrumentation pro- malfunctioning of any one control rod. It can vides vessel pressure and water level be concluded that no single failure in the rod , sensing for all protection systems, control and information system can result in the These instruments are arranged in two- - prevention of a reactor scram, and that repair, out of four logic combinations and their adjustment, or maintenance of the rod control' l ' signals are shared by both safety relat- and-information system' components does not ed and nonsafety related systems. . All affect the scram circuitry. of these. signals are multiplexed and. _l l passed through fiber optic media before: Chapter 15 examines the various failure mode' =j entering the voting logic of the redun- considerations' for this' system.? The expected l L dant divisions of the safety related and abnormal-transients and accident events l systems; or of nonsafety related sys- analyzed envelope the failure modes associated j tems which make up the various net. -with this system's components. works. Separation and isolation is thus: . . preserved both mechanically and electri- 7.7.2.2.2 Specific Regulatory Requirements cally in accordance with IEEE 279 and . Conformance Regulatory Guide 1.75. Table 7.12 identifies the nonsafety related With four independent sensing lines and control systems and.the associated codes and four independent electrical and mecha- standards applied in accordance with Section 7.7 nical divisions, the two out of four of the Standard Review Plan for BWRs. The l voting logic assures no individual following analysis lists the applicable criteria' sensing line failure could prevent pro- in order of the listing on the table,'and. per action of a protection system. When discusses the degree of conformance for each.' ; a system input channel is by passed, the Any exceptions or clarifications are so noted, logic reverts to two out of three. ' Amendment 11 7.7-36
NMStanalmed Plant 33A6100AF uv. A
- (1) General Design Criteria (GDC): 7.7.2.3.2 Specific Regulatory Requirements d
- - Conformance 1 (a) Qhstia GDCs 13 and 19. !
Table 7.12 identifies the nonsafety related I (b) Conformance: The RC&lS isin compliance - control systems and the associated codes and-with these GDCs, in part, or as a whole, standards applied in accordance with Section 7.7 ,l as applicable. The GDCs are generally of the Standard Review Plan for BWRs.: The . addressed in Subsection 3.1.2.' following analysis lists the applicable criteria in order of the listing on the table, and (2). RegulatoryGuides(RGs): discusses the degree of conformance for each. Any exceptions or clarifications are so noted. -i in accordance with the Standard Review Plan 1 for Section 7.7 and with Table 7.12, only '(1) GeneralDesign Criteria (GDC): 1 Regulatory Guide 1.151 Instrument Sensing Lines need be addressed for the ABWR.. (a) Qitatia GDCs 13 and 19. However, the RC&IS has no direct interface with the instrument lines, so this guide is (b) Conformance: The RFC is in compliance not applicable. The criteria of this guide. with these GDCs, in part, or as a- .; are discussed in relation to the nuclear whole, as applicable. The GDCs are ge- ; boiler system in Subsection 7.7.2.1.2(2). 'nerally addressed in Subsection 3.1.2,
)
7.7.2.3 Recirculation Flow Control System . -(2) Regulatory Guides (RGs): Instrumentation and Controls in accordance with the Standard Review Plan 7.7.2.3.1 General Functional Requirements for Section 7.7 and with Table 7.12, only Conformance Regulatory Guide 1.151 /nstrument Sens-ing Lines need be addressed for the ABWR. The recirculation flow control system (RFC) . The RFC receives signals from sensors on consists of the triplicated-RFC process vessel instrument lines!via the nuclear. I controller, adjustable speed drives, switches, . boiler system. The criteria of this guide j sensors, and alarm devices provided'for are discussed in relation to the nuclear i operational manipulation of the ten reactor boiler system in Subsection 7.7.2.1.2 - (2). l internal pumps (RIPS) and the surveillance of j associated equipment. 7.7.2.4 Feedwater Control System - Instrumentation and Controls Although not required to meet single failure criteria, each processing channel of the triply 7.7.2.4.1 GeneralFunctional Requirements redundant digital processor receives its Conformance. respective power input from an uninterruptible, independent source of the instrument and control The feedwater control system (FDWC) is not a - power supply system. The allocation of the RIP safety related system and is not required for equipment on four power buses is such that on : safe shutdown of the plant. It is a power ge-loss of any single power bus, only a maximum of neration system for purposes of inaintaining pro-three can be affected. per vessel water level. Its operation range is ~ from water level 8 (L8) to water level 2 (L2).- ! System single failure or single operator Should the vessel level rise too high (L8), the errors are evaluated in the transient analysis of feedwater pumps and plant main turbine weald be Chapter 15. It is shown that no malfunction in tripped. This is an equipment protenive action the recirculation flow control system can cause a which would result in reactor rh.down by the transient sufficient to cause significant damage RPS system as outlined in Sc. tion 7.2. Lowering to the fuel barrier or exceed the nuclear system of the vessel level would also result in action pressure limits, of the RPS and ECCS to shut down the reactor. i O Amendment 2 7.7 37
- .- - .- - - . . - - .. - -_ . - ._. -. ~ .
ABM 234sioo4r
. Etandard Plant ' ma n
' The system digital controllers and process -reactor startup and shutdown. The system aug . =} ~
measurement equipment are powered by rednahat ments existing information from other systems uninterruptible power supplies. No single power. such that the operator can start up, operate at supply failure shall result in the loss of any power, and shut down in an efficient manner.
- FDWC system function. The PGCC function provides signals to the automated power regulator (APR) as explained in Chapter 15 examines the various failure modes Subsection 7.7.1.5.1. However, this is a power ;
for this system relative to plant safety and generation function. Neither the PCS, nor its operational effects. , PGCC function initiate or control any engineered safeguard or safety related system. 7.7.2.42 Specific Reguh. tory Requirements . . , Conformance ' 7.7J.8.2 Specific Regulatory Requirements Conformance - .c Table 7.12 identifies the nonsafety related . 1
- control systems and the associated codes and L Table 7.12 identifies the nonsafety related i standards applied in accordance with Section 7.7 control systems and the associated codes and ;'
of the Standard Review Plan for BWRs. The ' standards applied ia a:cordance with Section 7.7 following analysis lists the applicable criteria of the Standud Review Plan for BWRs. However,- in order of the listing on the table, and since Ge computer has no controlling function,; discusses the degree'of conformance for each, none of the listed critcria are applicable. Any exceptions or clarifications are,so noted. Input dau for the PCS are derived from both (1) General Design Criteria (GDC): Class 1E and non Class 1E sources. - All such - L ' interfaces are optically isolated,'where L (a) QittJia GDCs.13 and 19. necessary, to assure the proper separation of ;' ! redundant signals in accordance with Regulatory l (b) Conformance: The FDWC is in compliance Guide 1.75. ' with these GDCs, in part, or as a whole, - as applicable,' The GDCs are generally 7.7.2.6 Neutron Monitoring System ATIP l addressed in Subsection 3.1.2. Subsystem Instrumentation and Controls - l (2) Regulatory Guides (RGs): 7.7.2.6.1 General Functional Requirements . Conformance In accordance with the Standard Review Plan for Section 7.7 and with Table 7.12, only The ATIP subsystem of the neutron monitoring' Regulatory Guide 1.151 Instrument Sensing _ : system is nonsafety related and is situated Lines need be addressed for the ABWR, The separately from' safety related hardware. It is . FDWC receives signals from' sensors on vessel used as a means of calibrating LPRM instrumert instrument lines via the nuclear boiler channels and has no controlling function with
-system. The criteria of this guide are other systems.
discussed in relation to the nuclear boiler system in Subsection 7.7.2.1.2 (2). 7.7.24.2 Specific Regulatory Requirements Conformance 7.7JJ Process Computer System - Instrumentation and Controls Table 7.12 identifies the nonsafety related control systems and the associated codes and L 7.7JJ.1 General Functional Requirements standards applied in accordance with'Section 7.7 1 Conformance of the Standard Review Plan for BWRs. However, since the ATIP has no controlling function, and The process computer system (PCS) is designed is used only for calibration of the LPRMs, none to provide the operhtor with certain categories of the listed criteria are applicable. I. of information and to supplement procedure ' requirements for control rod manipulation during Amendment 11 7.7-38
a QQ Remndard Plant :
. gg .
arv. n , l 7.7JS Automatic Power Regulator System- 7.7J.S.1 General Fanctional Requinments Imatrumentation and Controls Conformance {'N d' , L 7.7.2.7.1 General Fonctional Requinmenta The steam bypass & pressure control (SBAPC) ! !. Conforanance system is a power generation system in that it L inputs information to the _ automatic power < l The automatic power regulator (APR) system is regulator, which in turn controls reactor power a power generation system in that it receives : by manipulating control rods (via the rod command signals from the power generation system, control and information system) or recirculation , and the steam bypass and pressure control system; flow (via the recirc flow control system). The then controls reactor power by manipulating protective scram function is entirely separate control rods (via the rod control and information (via the reactor protection system). , system) or recirculation flow (via the recirc _ flow control system). The protective scram The SB&PC is classified as non safety function is entirely separate (via the reactor related and does not interface with any protection system). engineered safeguard or safety related system.
- The APR is classified as non safety related 7.7.2.8.2 Specine Regulatory Requinments and does not interface with any engineered Conformance safeguard or safety related system. .
Table 7.12 identifies the non safety 7.7.2.7.2 SpeelRc Regulatory Requinments related control systems and the associated codes Conformance and standards applied in accordance with Section ; 7.7 of the Standard Review Plan. The following Table 7.12 identifies the non safety related analysis lists the applicable criteria in order control systems and_the associated codes and of the listing on the table, and discusses the standards applied in accordance with Section 7.7 degree of conformance for each. Any exceptions ("3 of the Standard Review Plan for BWRs. The or clarifications are so noted. V following analysis lists the applicable criteria in order of the listing on the table, and (1) General Design Criteria (GDC) discusses the degree of conformance for each. Any exceptions or clarifications are so noted. (a) Qitnia: GDCs 13 and 19 (1) General Design Critcria (GDC) (b) Conform ance: The SB&PC is in compliance with these GDCs, in part, or-as a (a) Criteria: GDCs 13 and 19. whole, as applicable. The GDCs are generally addressed in Subsection 1 (b) Conformance: The APR is in compliance 3.1.2. with these GDCs, in part, or as a whole, I as applicable. Tbc GDCs are generally (2) Regulatory Guides (RGs) . L addressed in Subsection 3.1.2, l In accordance with the Standard Review Plan j' (2) RegulatoryGuides(RGs) for Section 7.7 and with Table 7.12,' only Regulatory Guide 1.151 Instrument Sensing L In accordance with the Standard Review Plan Lines need be addressed for the ABWR. D for Section 7.7 and with Table 7.12, only l Regulatory Guide 1.151 Instrument Sensing (a) Criteria: Regulatory Guide 1.151
- Lines need be addressed for the ABWR. The instrument Sensing Lines l APR does not have any direct interface with the instrument lines; therefore, this guide (b) Conformance
- The SB&PC interfaces with
, is not applicable, sensors connected to instrument lines ! on both the reactor and the turbine. 7.7.2.8 Steam Bypass and Pressure Control The reactor instrument line interface. l
/7 System. Instrumentation and Controls V Amendment 11 7.7-38.1
- . -. ..- . . - - . -. .- .~ -_-__ --- - - . . - .
I
'AB M isasioasr i Amandard Plant arv. n is vla the nuclear boiler system, which failure will cause its basic function to fail. ;
is in full compliance with this guide as 1 discussed in Subsection 7.7.2.1.2 (2). 7.7.2.9.2 Specific Regulatory Requirements l There are four independent turbine - instrument lines, which contain turbine- . Table 7.12 identifies the non safety- J 1st stage pressure sensors as part of related control systems and the associated codes > the turblue control system,in addition - and standards applied in accordance with Section . to the non Class 1E sensors associated 7.7 of the Standard Review Plan. However, as i with the SB&PC. The 1st stage turbine -. mentioned above, the NEMS is not a separate pressure signals are used as bypass control system subject to separate review, but , interlocks for the turbine control valve is the data communication vehicle for virtually fast closure and turbine stop valve all of the nonsafety related systems. It.
~
closure scram functions [see Subsection- provides specific enhancement for all control 7.2.1.1.4.2 (6) (d)].- No single failure ' systems in their conformance with GDCs 13 and can cause this function to be disabled. 19. In addition, since the turbine building _ itself is a nonseismic structure, these. 7.7.2.10 Other Nonsafety Related Control s scram functions sie backed up by diverse . Systems reactor . variables which will . independently initiate scram, should the The following nonsafety related control sys-turbine signals be lost. These diverse variables to cause scram are reactor. high pressure and high flux (via NMS).
' Therefore, no event associated with
[ turbine instrument lines can cause an action' requiring scram, while at the same time disabling the scram function. The SB&PC fully complies with Regulatory - Guide 1.151.= l 7.7.2.9 Non Essential Multiplexing System- ) Instrumentation and Controls 7.7.2.9.1 General Requirements Conformance The NEMS, of itself, is neither a power generation system nor a protection system. It is a support system utilized for assimilation, transmission and interpretation of data for power generation (non safety related) systems and their associated sensors,: actuators.and interconnections. It is classified as l nonsafety related and does not laterface with any engineered. safeguard or safety related system except for isolated alarms for annunciation. The NEMS-is an integral part of the power generation systems 'vhich it supports. As such, it meets the same functional requirements imposed on those systems. Although not required to meet the single failure criteria, the system is redundant and receives its power from redundant, highly reliable power sources such that no single
- l. Amendment 11 7.7 38.2 4
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,MM 21A6100AF -
Remndard Plant RN B tems are described in other subsections of the O, SSAR as indicated.'
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i GE PROPRIETARY . provided under separate cover . , O't. ; A (
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i Figure 7.7-13 STEAM BYPASS AND PRESSURE CONTROL SYSTEM IBD, SHEET 5 l Amendment 11 7.7 96
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i SECTION 7.8 ' b CONTENTS Astina m P.ast i 7A.1 Effects et klatina Blakant na the HVAC 7.81 7A.2 Electraataile Dischame na Expaaed Emminawat Canipaments 7.81 7AJ 1mentined Mlah Heat Spot, la Smaicanductor Materials far Canipullag Devices 7.81 7A.4 Enfets Related Cal latedaces 7.81 i V l ( I r 7.8il Amendment 11 -
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Mhk 21A6100.V Etandard Plant arv n
- 7AINTERFACES safety related electrical signal laterfaces for any of these spleins which calend beyond the 7A.1 Effects of Station Blackout scope defindtion.
on the HVAC A temperature beat rise at.alysis shall be performed for the station blackout scenario applied to the control room on consideration of the envirummental temperatures unique to the plant location. [ See Chapter 20, NRC Ouestion 420.14 and S'Jbsection 7.1.2.3.9) 7A.2 Electrostatic Discharge on Exposed Equipment Components : The response to NRC Ouestion 420.90 provides recommendations for limiting the effects of electrosatatic discharge (ESD) at keyboards, keyed switches and other exposed equipment. The applicant shall provide assurance that the grounding and shielding techniques are consistent with these recommendations, or provide an acceptable alternative plan for controlling ESD. [See Chapter 20, NRC Ouestion 420.90) 7J.314calized High Heat Spots in - l Semiconductor Materials for Computing Devices The response to NRC Ouestion 420.92 provides recommendations for limiting high cuurent i densities whigh could result in localized heat spots in semiconductor materials used in computing devices. The applicant shall provide
- ^ assurance that these recommendation are followed, or an acceptable alternative is presented, by the selected equipment vendor (s). To ensure that adequate compensation for heat rise is incorporated into the design, a thermal analysis shall be performed at the circuit board, instrument and panel design stages. [See Chapter 20, NRC Question 420.92)
F 7.8A Safety Related C&lInterfaces Each of the systems addressed in Chapter 7 were reviewd for safety related C&l (signal) interfaces which extend outside the scope of the ABWR Standard Plant. Since the scope of the ABWR Standard Plant includes all of the reactor building, the turbine building and the control p building, the study determined there are no Amendment 11 7.81
i ABM :. w io w i : I Standard Plant arv. n . CHAPTER 9 i TABLE OF CONTENTS 1 l Section B11g Eagg i 9 AUXILIARY SYSTEMS j 9.1 FUEL STORAGE AND HANDLING 9.1.1 New FuelStorage 9.11 ; 9.1.2 Speet FuelStorage 9.12 9.1.3 Fuel Pool Cooling and Cleanup Sptem 9.13 9.1.4 Light load Handling Sptem (Related to Refueling) 9.16 i 9.1.5 Overhead Heavy Lead Handling Sptems 9.17 9.1.6 References 9.1 13 9.2 WATER SYSTEMS 9.2.1 Station Service Water System 9.21 O 9.2.2 Closed Cooling Water System 9.21 9.2.3 Demineralized Water Makeup System 9.21 9.2.4 Potable and Sanitary Water Systems 9.21 9.2.5 Ultimate Heat Sink 9.21 9.2.6 Condensate Storage Facilities and Distribution System 9.21 9.2.7 Plant Chilled Water Systems 9.21 [ 9.2.8 Makeup Water Systems (Preparation) 9.21
- 9.2.9 Makeup Water System (Condensate) 9.21 >
4 9.2.10 Makeup Water System (Purified) Distribution System 9.2-2 t I 9 il Amendment 6 l __ _
ABWR muern ! Standard Plant am n
'l CHAPTER 9 I
TABLE OF CONTENTS (Continued) i Sectson lillt PAge ! 9.2.11 Resetor Building Cooling Water System 9.23 l 9.2.12 HVAC Normal Cooling Water System 9.27 I 9.2.13 HVAC Emergency Cooling 9.2-8 9.2.14 Domestle Water Systems 9.2 11 9.2.15 Reactor Service Water System 9.2 11 9.2.16 Turbine Senice Water System 9.2 12.1 Interfaces 9.2 13 9.2.17 f 9.3 PROCESS AUXILIARIES , 9.3.1 Compressed Air Systems 9.31 9.3.2 Process and Post Accident Sampling System 931a 93.3 Equipment and Floor Drainage Systems 93 2 , 93.4 Chemicat and Volume Control System (PWR) 93 2 { 9.3.5 Standby Liquid Control System 93 2 93.6 Instrument Air System 9.37 93.7 Seniec Air System 93 8 ; 93.8 Radioacthe Drain Transfer System 93 11
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93.9 Hydrogen Water Chemistry System 9.3 12 93.10 Oxygen injection System 9.3 13 . 9,4 AIR CONDITIONING. HE ATING. COOLING AND YENTILATION SYSTEMS > l 9.4.1 Control Room Area Ventilation System 9.41 9.4.2 Spent Fuel Pool Area Ventilation System o.42 9.43 Auxiliary Area Ventilation System 9.42 l 9 111 Amendment 11 O ! t
ABWR z w io w : Standard Plant " Rev. B ; I l < CHAPTER 9 l TABLE OF CONTENTS (Continued) l Section Ihlt East
- l 9.4.4 Turbine Ares Ventilation $ptem 9.42 9.4.5 Resetot Building Ventiation System 9.42e 9.46 Radwaste Building HVAC 9.42j ;
j 9.4.7 Diesel Generator Area Ventilation Sptem 9.43 9.4.8 Senice Building Ventilation System 9.43 9.4.9 Drywell Cooling Sptem 9.44 9.5 QTilEk AUXILI ARY SYSTEMS 9.5.1 Fire Protection Sptems 9.51 9.5.2 Communleation Sptems 9.52 9.5.3 Lighting and Senicing Power Supply i Systems 9.53 I 9.5.4 Diesel Generator Fuel Oil Storage and Transfer System 9.5-4 9.5.5 Diesel Generator Cooling Water System 9.55 9.5.6 Diesel Generator Starting Air Splem 9.56 9.5.7 Diesel Generator Lubtleation System 9.57 9.5.8 Diesel Generator Combustion Air intake and Exhaust System 9.58 9.5.9 Suppression Pool Cleanup Sptem 9.58 I 9.510 Motor Generator Set 9.5 10 l 9.5.11 Interfaces 9.5 11 APPENDIX 9A FIRE ilAZARD ANALYSIS 9A 1 ( 9 iv Amendment 11
ABM 2sisicosti Standard Plant , au n 9.1.2 Spent Fuel Storage The spent fuel poo11s a reinforced concrete O 9.1.2.1 Design Bases structure with a stainless steelliner. The bottoms of all pool gates are sufficiently high to maintain the water level over the spent fucl storage racks form 9.1.2.1.1 Nuclear Design adequete shielding and cooling. All pool fill arsd drain linea enter the pool above the safe shielding water (1) A full array in thz loaded spent Nel tack is level. Redundant anti. siphon vacuum breakers are designed to be suberitical, by at least 5% yk. located at the high point of the pool circulation lines Neutron. absorbing material, as an integral part to preclude a pipe break from siphoning the water of the design,is employed to assure that the from the pool and jeopardizir:g the safe water level. calculated k including biases and uncertaintles, wilEn,ot exceed 0.95% under all The racks include individual solid tube storage normaland abnormal conditions. compartments, which provide lateral restraints over the entire length of the fuel assembly or bundle. The (a) Monte Carlo techniques are employed in weight of the fuel assembly or bundle is supported the calculations performed to assure that axially by the rack fuel support. Lead-in guldes at the k does not exceed 0.95 under all normal top of the storage spaces provide guidance of the fuel a$ abnormal conditions. duringinsertion. (b) The assumption is made that the storage The racks are fabricated from materials used for attny is infinite in all directions. Since no construction are specified in accordance with the credit is taken for neutron leakage, the latest issue of applicable ASTM specifications. The values reported as effective neutron racks are constructed in accordance with a quality multiplication factors are, in reality, assurance program that ensures the design, infinite nevtron multiplication factors, construction and testing requirements are met. - (c) The biases between the calculated results The racks arc designed to withstand, while O and experimental results, as well as the uncertainty involved in the calculations, are taken into aecount as part of the maintaining the nuclear safety design basis, the impact force genersted by the vertical free fall drop of a fuel assembly from a height of 6 feet. The rack is calculational procedure to assure that the designed to withstand a pullup force of 4000 pounds specific g k limit is met. and a horizontal force of 1000 pounds. There ate no readily definable horizontal forces la excess of 1000 9.1.2.1.2 Storage Design pounds, and in the event a fJet assembly should jam, the maximum lifting force of the fuelhandling The fuel storage racks provided in the spent fuel platform grapple (assumes limit switches fall) is 3000 storage pool provide storage for 270% of one full pounds. corc fuelload. The fuel storage racks are designed to hndle fr.l.2.1.3 Mechanical and Structural Design irradiated fuel assemblics. The expected radiation levels are well below the design levels. The spent fuel storage racks in the reactor building co;tain storage space for fuel assemblies (with In accordance with Regulatory Guide 1.29, the fuel channd) nr bundles (without channels). They are storage racks are designated Safety class 2 and designed to withsttad all credible static and seismic Seismic Category 1. The structuralintegrity of the loadings. The racks ate designed to protect the fuel rack has been demonstrated for the load assemblics and bundles from excessive physical combinations described below using linear clastic damage which may cause the release of radioactive design methods, materials in excess of 10CFR20 and 10CFR100 requirements, under normal and abnormal The applied loads to the rack are: conditions caused by impacting from either fuel assemblics, bundles or other equipment. (1) dead loads, which are weight of rack and fuel assemblies, and hydrostaticloads; Amendment 11 9,12 l
1 Mkk thA61DQAH Standard Piant am a !
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(2) live loads effect of lifting an empty rock The loads in the three orthogonal directions were j duringinstallation; considered to be acting simultaneously and werc l combined using the $R$$ method suggested in l (3) thermal loads . the uniform thermal expansion Regulatory guide 1.92. The loats due to the OBE due to pool temperature changes; event are approximately 90% of hoset due to an $$E i ennt, and allowable atress levels for OBE are 50% of . ; (4) seismic forces of OBE and $5E; $$E, therefore making the OBE event the limiting : load condition eacept for stability, where $$E (5) accidental drop of fuel assembly from acceptance criteria of 67% of critical buckling maximum possible height 6 feet above rack; strength is limiting. i and Under fuel drop loading conditions, the acceptance i (6) postulated stuck fuel assembly causing an criterion is that, although deformation may occur, upward force of 3000 pounds. K ,arust remain <0.95%. The rack is designed such th*a,T, should the drop of a fuel assembly damage the The load combinations considered lei the rack tubes and dislodge a plate of poison material, the K,gg design are: would still be <0.95 as required. (1) live load 4 The effect of the gap between the fuel and the storage tube has been taken into account on a local (2) deadloads plus OBE effect basis. Dynamic response analysis shows that the fuel contacts the tube over a large pnrtion of its (3) dead loads plus SSE;and length, thus preventing an overloaded condition of both fuel and tube. (4) dead loads plus fuel drop. The vertical impact load of the fuel onto its seat has Thermalloads were not included in the above been considered conservatively as being slowly
- combinations because they were negligible due to the applied without any benefit for strain rate effects.
l design of the rack (i.e., the rack is attached only at its base and is free to expand / contract under pool 9.1.2.1.4 Thermal Hydraulic Design temperature changes). The fuel storage rack is designed to provide , The loa.ls.cxperienced under a stuck fuel assembly sufficicat natural convection coolant flow to remove condition are less than those calculated for the 68,000 Btu /hr/ bundle of decay heat. scismic conditions and, therefore, have not been included as a load combination. The support structure must be designed to provide an adequate flow rate to pbevent water reaching The storage racks are attached to the support excessive temperatures 212 F. The flow rate is l structure by bolting, sufficient to counteract the dependent on the decay heat load, the AP losses I tendency to overturn from horizontal loads and to lift through the structure and the losses through the rack fiom vertical loads. The analysis of the rock and bundle. i assumed an adequate supporting structure, and loads were generated accordingly. In the spent fuel storage pool, the bundle decay heat is removed by recirculation flow to the fuel pool l Siress analyses were performed by classical cooling heat exchanger to maintain the pool tempers methods based upon shears and moments developed ture. Although the design pool exit temperature by the dynamic method. Using the given loads, load within the rack is high depending on the naturally conditions and analytical methods, stresses were induced bundle flow which carries away the decay calculated at critical sections of the rack and , heat generated by the spent fuel. The rate of compared to acceptance criteria referenced in naturally circulated flow and maximum rack exit ASME Section lit subsection NF. Compressive temperature have been evaluated. stability was calculated according to the AISI code for light gage structures. The parameters which will affect the water flow Amendmem 6 9.12a i
J ABM nisiooni Standard Plant am m . 9.!A Light lead Handling System fuel bundle drop. Maaimum deflection limitations are ! (Related to Refueling) imposed on the main structures to maintain relative , l atiffness of the platform. Welding of the platforms is ) 91.4.1 Dealen Bases in accordance with AWS D14-1 or ASME Boiler and 1 Pressure Vessel Code section IX. Ocars and bearing l The fuel handling system is designed to provide a meiet AGMA Gear Classification Manual and ANSI safe and effective means for transporting and B3.5 Materials used in construction of load bearing ; handling fuel from the time it reaches the plant until members are to ASTM specifications. For personnel 1 it leaves the plant after post irradiation cooling. Safe safety, OSHA Part 1910179 is applied. Electrical
- handling of fuelincludes design considerations for equipment and controls meet ANSI CI, National ,
- maintaining occupational radiation exposures as low Electric Code, and NEMA Publication No. ICSi, as practicable during transportation and handling. M01.
Design criteria for major fuel handling system The auxiliary fuel grapple and the main telescoping i equipment are provided in Table 9.12 through 9.14, fuel grapple have redundant lifting features and an which list the safety class, quality group and seismic indicator which confiras positive grapple ' category. Where applicable, the appropriate ASME, engagement. ANSI, lodustrial and Electrical Codes are identified. Additional design criteria are shown below and The fuel grapple is used for lifting and transporting ; expanded further in Subsection 9.1.4.2. fuel bundles, it is de+1gned as a telescoping grapple that can catend to the proper work level and,in its The transfer of new fuel assemblics between the fully retracted state, still maintain adequate water ; uncrating area and the new fuelinspection stand shielding over fuel. , and/or the new fuel storage vault is accomplished . using 5 ton auxiliary hook on the reactor building in addition to redundant electricallaterlocks to crane equipped with a suitable grapple. preclude the possibility of raising radioactive material out of the water, the cables on the auxiliary hoists b d The 1,000 pound auxiliary hoist on the reactor building crane is used with an auxiliary fuel grapple incorporate an adjustable, removal stop that will jam the hoist cable against some part of the platform .' to transfer new fuel from the new fuel vault to the structure to prevent hoisting when the free end of the fuel storage pool. From this point on, the fuel will cable is at a preset distance below water level, either be handled by the telescoping grapples on the handling platform or jib cranes at rechanneling Provision of a separate cask pit, capable of being models, isolated from the fuel storage pool, will climinate the potential accident of dropping the cask and rupturing l The refueling platform is Seismic Category 1 from the fuel storage pool. Purthermore, limitation of the a structural standpoint in accordance with 30CFR50, travel of the crane handling the cask will preclude Appendix A. The refueling platform is constructed transporting the cask over any fuel storage podl. in accordance with a quality assurance program that , ensures the design, construction and testing 9.1.4.2 System Deacription - requirements are met. Allowable stress due to safe shutdown carthquake (SSE) loading is 120% of yield Table 9.15 is a listing of typical tools and servicing l or 70% of ultimate, whichever is least. A dynamic equipment supplied with nucleer system. The analysis is performed on the structures using the following paragraphs describe the use of some of the response spectrum method with load contributions major tools and servicing equipment and address resulting from each of three directions acting safety aspects of the design where applicable, simultaneously being combined by the RMS procedure. Working loads of the platform structure Subsection 9.1.5 provides the data that verifies the
- are in accordance with the AISC Manual of Steel ABWR Standard Plant heavy load handling systems Construction. All parts of the bolst systems are and satisfies the guidelines of NUREG 0612.
designed to have a safety factor of at least ten, based on the ultimate strength of the material. A 9.1.4.2.1 Spent Fuel Cask redundant load path is incorporated in the fuel hoists O so that no single component failure could result in a Out of ABWR Standard Plant scope. 914 Amendment 11
i 23A6100AH ! Sundard Plant am n J 9.1.4.1.2 overhead Bridge Cranes 9.1.4.23J kw Fwllupw& Stand 9.1.4.2.2.1 Reactor Building Crnee Tbc new fuellaspection stand (Figure 9.14) serws ; as a support for the new fuel bundles undergoing re- l The reactor building crane is a seismically analysed ceiving inspection and provides a working platform piece of equipment. The crane consists of two eranc for technicians engaged in performlag the laspection. ; girders and a trolley which carries two hoists. The runway track, which supports the crane girders, is 'Itc new fuelinspection stand consists of a vertical supported from the reactor building walls at eleva- guide column, a lift unit to posittoa the work platform , tion 34,600.~ The trolley travels laterally on the crane at any desired level, bearing seats and upper clamps i girders carrying the main hoist and auxiliary hoist, to hold the fuelbundles in position. The reactor building eranc is used to move all of 9.1.4JJJ Channel Bolt Wrench the major components (reactor vessel head, shtoud ! head and separator, dryer assembly and pool gates) The channel bok wrench (Figure 9.b5) is a manu-l as required by plant operations. The reactor build- ally operated device approximately 3.'16 :neters (12 ft) ing crane is used for handling new fuel from the re- in overall length. The wrench is used for removing actor building entry hatch to new fuel storage, the and installir.g the channel fastener assembly while the new fuel inspection stand and the spent fuel storage fuel assembly is held in the fuel preparation machine. i pool. It also is used for handling spent fuel cask. Tbc channel bolt wrench has a socket which mates l The principal design criteria for the reactor building and captures the channel fastener capscrew. crane are described in Subsection 9.1.5. 9.1.4.23.4 Channel. Handling Tool . 9.1.4.2.3 Fuel Servicing Equipment ! The channel. handling tool (Figure 9.16) is used in The fuel senicing equipment described below has conjunction with the fuel preparation machine to been designed in accordance with the criteria listed remove, install and transport fuel channels in the fuel i in Table 9.12. Items not listed as Seismic Category storage pool. 1, such as hoists, tools and other equipment used for , senicing shall either be removed during operation, The toolis composed of a handling bail, a moved to a location where they are not a potential lock / release knob, extension shaft, angle guides and > hazard to safety related equipment, or seismically re. clamp arms which engage the fuel channel. The strained to prevent them from becoming missiles. clamps are actuated (extended or retracted) by manu- l ally rotating lock / release knob. 9.1.4.2J.1 Fuel Prep Machine The channel. handling tool is suspended by its ball Two fuel preparation machines (Figure 9.13) are from a spring balancer on the channel. handling boom l mounted on the wall of the fuel storage pool and are located on the fuel paol periphery. ! used for stripping reusable channels from the spent fuel and for rechanneling of the new fuel. The ma. 9.1.4.2.3.5 Fuel PoolVacuum Sipper , chines are also used with the fuelinspection fixture to provide an underwater inspection capability. The fuel pool vacuum sipper (Figure 9.17) pro- t vides a means of identifying fuel suspected of having
- Each fuel preparation machine consists of a work cladding failures. The fuel pool Vacuum sipper con- ,
platform, a frame, and a movable carriage. The sists of a fuelisolation container, fluid console, moni- , frame and movable carriage are located below the toring console with program controller and beta de. normal water levelin the fuel storage pool thus pro- tector and the inter connecting tubing and cables, viding a water shield for the fuel assemblies being The suspected fuel assembly is placed in the isolation , handled. The fuel preparation machine carriage has container. A partial Vacuum is established in the gas ' a permanently installed up travel.stop to prevent volume above the fuel assembly. The fission product i raising fuel above the safe water shield level, gas leakage is sensed by the beta detector and moni-toring console, ; I l Ameadment 7 9.16a
ABWR 2mimn Standard Plant nev. n Table 9.1 1 DEFINITION OF TERMS A Flow area through bundles = 15.353 in.2), b Ag Arbitrary area used in bundle friction correlation = 10 in. C Specific heat of water = 1.0 Blu/lb- F. p 2 g Gravitational constant 32.2 ft/sec , H Head loss through bundle (ft H O). b 2 h, Effective depth of cold water over entrance point into bundle = 13.5 ft in this example 3 i Intercept in p versus t correlation = 63.45 lb/ft . 3 M Slope of p versus t correlation = 0.0145 lb/ft . F. p, Densityof water = 62.00lb/ft3(at 100 F). O Heat evolution rate frorn bundle - 68,000/3/00 Btu /sec. t inlet water temperature (100 F). Y Vel city of water through bundle (ft/see). b l i
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ABWR nAMMAH Standard Plant Rev o i Table 9.12 hj FUEL SERVICING EQUIPMENT l Essential Safety Component - Classift. Classifl. Quality Selsmic ! No. Identification cation entlos Group Category i (a) (b) (c) (d) 1 Fuel Prep Machine NE O E- I l r 2 New Fuellaspection Stand NE O E I - 3 Channel Bolt Wrench NE O E NA 4 Channel Handling Tool NE O E NA i 5 Fuel Pool Vacuum Sipper NE O E NA . 6 General Purpose Grapple NE O E. NA i 7 Jib Crane NE O E I l
.t 8 Auton.atic Refueling NE O E I-Machine
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j 9 Channel Handling Machine NE O E NA Eggy (a) NE = Non Essential PE = Passive Essential i (b) O = Other ,
'B = ASME Code Section 111 Class 2 (c)
D = ANSI B31.1 E = ElectricalCodes Apply I = Industrial Code Applies : (d) NA = No Seismic Requirements I = Class I ' ? t 9~ Amendment 11 9.1 13b h k
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ABM swioorn Remndard Plant REV.B , SECTION 9.2 CONTENTS (Continued)
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9.2.13.5 Instrumentation Application .9.29 l 9.2.14 Turbine Building Coolina Water System 9.2 10 9.2.14.1 Design Bases 9.2 10 ! 9.2.14.1.1 Safety Design Bases 9.2 10 ! ! 9.2.14.1.2 Power Generation Design Bases 9.2 10 , 9.2.14.2 System Description 9.2 10 9.2.14.2.1 General Description . 9.2 10 t 9.2.14.2.2 Component Description 9.2 10 - , 9.2.14.23 System Operation 9.2 10 9.2.14 3 Safety Evaluation 9.210a 9.2.14.4 Test and Inspection 9.210m 9.2.14.5 Instrumentation Application 9.210a 9.2.15 Reactor Service Water ag(gm 9.2 11 9.2.15.1 Design Bases 9.2.11 9.2.15.1.1 Safety Design Bases 9.2.11 9.2.15.1.2 Power Generation Design Bases - 9.2 12 1 9.2.15.2 System Description 9.2 12 9.2.15 3 Safety Evaluation 9.2 12 , 9.2.15.4 Testing and Inspection Requirements 9.2 12 9.2.15.5 Instrumentation and Control Requirements 9.2.12.1 i 9.2.16 Turbine Service Water Svstem 9.2 12.1 9.2.16.1 Design Bases 9.2-12.1 9.2.16.1.1 Safety Design Bases 9.2 12.1 , 9.2iv Amendment 11 1 I
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~ 21A6100AH Rf amAard Plant nrv a l SECTION 9.2 g
- CONTENTS (Continued)
ERCLI9B M M 9.2.16.1.2 Power Generation Design Bases 9.2 12.1 9.2.16.2 System Descripion 9.2-12.1 ; 9.2.16.2.1 General Deuription 9.2 12.1 )
- l : 9.2.16.2.2 Component Description - 9.2.12.1 9.2.16.23 . System Operation 9.2 '12.1 1 9.2.16.3 Safety Evaluation 9.2 12.2 9.2.16.4 Tests and Inspections 9.2 12.2 '
1 l 9.2.16.5 lastrumentation Application ' 9.2 12.2
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9J.17 Interfaces 9.2 13 9.2.17.1 Ultimate Heat Sink Capability 9.2 13 ^ 9.2.17.2 Makeup Water System Capability 9.2 13 01 > i 4 4 i 1-i 4 t i l s 9.2-iva Amendment 11 i 4 ~
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l l 1 ABM usamui ! mannlard Plant nyn O section'2 ; TABLES l Tables 21tle East j j, 9.21 Users of Makeup Water Condensate 9.2 14 l [ 9.22 Users of Makeup Water Purified 9.2 15 l L 9.2 Capacity Requirements for Condensate Storage Tank 9.2 16 , 9.2 4a Reactor Building Cooling Water Division A 9.2 17-- 9.2-4b Reactor Building Cooling Water Division B 9.2 18 Resetor Building Cooling Water Division C 9.2 4c 9.2 19 9.2-Sa Reactor Building Cooling Water Active Failure Analysis 9.2 20 9.2 5b Reactor Building Cooling Water System Passive Failure Analysis 9.2 21 9.26 HVAC Normal Cooling Water System Component Description 9.2 22 , tO 9.27 HVAC Normal Cooling Water Loads 9.2 23 9.28 HECW System Component Description 9.2 24 , 9.29 HVAC Emergency Cooling Water System Heat leads 9.2 25 ; 9.2 10 HVAC Emergency Cooling Water System Active Failure - Analysis 9.2 25a 9.2 11 Turbine Building Cooling Wtter System Heat Loads 9.2 25b 1 9.2 12 Summary of Turbine Cooling Water System Heat Exchangers 9.2 25e f ; ILLUSTRATIONS l UgEts Dils Eage 9.2 la Reactor Building Cooling Water System P&lD 9.2 26 9.2 laa Reactor Building Cooling Water System PFD 9.2 34a 9.22 HVAC Normal Cooling Water System P&lD 9.2 35 9.2 v l l Amendment 11
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I 21A610QAH . i EttamAard Plant an a SECTION 9.2 (Continued) ILLUSTRATIONS (Continued) gj ' F1gunts .M Rast [ 9.23 HVAC Emergency Cooling Water System P&ID 9.2 37 9.24 MUWC P&ID 9.2 39 9.25 MUWP P&lD 9.2 40 9.26 Turbine Building Cooling Water System Dirpam 9.2 42 i 9.27 Reactor Scryice Water System 9.2 45 9.2-8 Turbine Service Water System 9.2-46 l l 5 e r 9.2-vi Amendment 11 i
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i l i 21A6100AH Remndard Plant arv. m O 9.2 WATER SYSTEMS 9.2.9 Makeup Water system (Condensate) l 9.2.1 Station Service Water System 9.2.9.1 Dealga Bases ;
)
The functions normally performed by the (1) The makeup water. condensate system (MUWC) i station service watet system are performed by the shall provide condensate quality water for systems discussed in Subsection 9.2.11. both normal and emergency operations when ' required. 9.2.2 Closed CoolingWater System , (2) The MUWC system shall provide a required The functions normally performed by the closed water quality as follows: cooling water system are performed by the systems ' discussed in Subsectloss 9.2.11, 9.2.12, 9.2.13, Conductivity (y S/cm) s 0.5 at 25'C l and 9.2.14. Chlorides, as C1(ppm) 1 0.02 i pH 5.9 to 8.3 at 25'C i < 9.2J DemineralizedWaterMakeup Conductivity and pH limits shall be applied . I System after correction for dissolved Co2. (The l above limits shall be met at least 90% of l The functions normally performed by the demin. the time.)
- eralized water makeup system are performed by the i
systems discussed in Subsections 9.2.8, 9.2.9 and (3) The MUWC system shall supply water for the 9.2.10. uses shown in Table 9.21. c 9.2.4 Potable and SanitaryWater (4) The MUWC system is not safety related. i Systems (5) The condensate storage tank shall have a
- .~ Out of ABWR Standard Plant Scope, capacity of 2,110 m3. This capacity was l determined by the capacity required by the
- 9.2.5 Ultimate Heat Sink uses shown in Table 9.2 2.
Out of ABWR Standard Plant scope. See (6) All tanks, piping and other equipment shall l Subsection 9.2.17.1 for interface requirements, be made of corrosion resistant materials. ! 9.2.6 Condensate Storage Fa:llities (7) The HPCF and RCIC lastrumentation, which and Distribution System initiates the automatic switchover of HPCF and RCIC suction from the CST header to the The functions of the storing and distribution suppression pool, shall be designed to ; of condensate are described in Subsection 9.2.9. safety. grade requirements (including , installation with necessary seismic l 9.2.7 Plant Chilled Water Systems suppor:). l The functions of the plant chilled water 9.2.9.2 System Description . system are performed by the systems described in i Subsections 9.2.12 and 9.2.13. The MUWC P&lD is shown in Figure 9.2 4 This system includes the followng: E 9.2.8 MakeupWater System (Preparation)
< (1) A condensate storage tank (CST) is provid.
Out of ABWR Standaril Plant scope. See ed, it is of concrete construction with a Subsection 9.2.17.2 for interface requirements. stainless steellining. The volume is shown t in Table 9.2 3. l /7 (2) The following pumps take suction from the i l .Q l CST: Amendment 11 9.21
- . . . . . . . - . - . . - . . . . - - . . ~ .. . . . . - . _ . . . . _ - _ - . . _ _ . . . . _ . - - _ _ . - - - _
I i l i l' i Mk i 23A6100AH ) Etandard Pl nr uv. n - ! (a) RCICpump , I. - (b) CRD pump l l (c) HPCFpump
)
(d) SPCU pump l l l 1 t n j O t' i Am ndment 8 9.2 1.a O.
t i ABWR 2-n arv n Standard Plant - tant materials. (3) ability to prevent or mitigate the conse. quences of events which could result in r
-(6) The system shall be designed to prevent any potential offsite exposures. [
radioactive contamination of the purified - water. The MUWP system is not safety related. ! ilowever, the systems incorporate features that j' 9.2.10.2 System Description assure reliable operation over the full range of normal plant operations. ; 5; The MUWP system P&lD is shown in Pigure 9.2 5. 6 This system includes the following: 9.2.10.4 Tests and inspections - (1) A purified water storage tank shall be The makeup water purified distribution system , provided outdoors with adequate freete is proved operable by its use during normal ! protection and adequate diking and other plant operation. Portions of the system ; means to control spill and leakage, normally closed to flow can be tested to ensure operability and integelty of the system. (2) Two MUWP forwarding pumps shall take suction from the purified water storage tanks. They The air. operated isolation valves are capable ' shall have a capacity of 308 gpm and a of bc8ng tested to assure their operating discharge head of 114 psi, integrity by manual actuation of a switch located in the control room and by observation (3) Distribution piping, valves, instruments and of associated position indication lights. controls shall be provided. Flow to the various systems is balanced by (.1) Any outdoor piping shall be protected from means of manual valves at the individual takeoff freezing. points. (3 O (5) All surfaces coming in contact with the 9.2.11 Reactor Building CoolingWater purified water shall be made of corro. System sion resistant materials, 9.2.11.1 Design Bases (6) All pumps shall be located at an elevation such that adequate suction head is present 9.2.11.1.1 Safety Design Bases at all levels in the purified water storage tanks. (1) The reactor building cooling water (RCW) - system shall be designed to remove heat from (7) Instruments shall be provided to indicate plant auxiliaries which are required for a purified water storage tank level in the safe reactor shutdown, as well as those main control room. auxiliaries whose operation is desired l following a LOCA, but not essential to safe 1 9.2.10.3 Saf'ety Daluation shutdown. 1 Operation of the MUWP system is not required (2) The RCW system shall be designed to perform to assure any of the following conditicas: its required cooling functions following a LOCA, assuming a single active or passive (1) integrity of the reactor coolant pressure failure. - boundary; # (3) The safety related portions and valves (2) capability to shut down the reactor and isolating the nonsafety.related portions of maintain it in a safe shutdown condition; or i Amendment 6 9.23 L
=_
. i i 1 ! r , l ABM n isi w i j m.a.a pi.m ma g; ' j the RCW ayste.m shall be designed to seismic shutdown requirements, including a LOCA or a i Category I and the ASME Code, Section III, loss of offsite power, or both. Each RCW i l Clus 3, Quality Assurance B, Quality Group division is supplied electrical power from a l l C, fEEE 279 and IEEE.308 requirements, different division of the ESF power system, ; l : (4) The RCW aystem shall be designed to limit During normal operation, RCW couting water leakage to the environment of radioactive flows through all the equipment shown in Table contamination that may enter the RCW from 9.2 4a, b, and c. the RHR System. , , During all plant operating modes, en RCW (5) Safety related portions of the RCW system water pump and beat exchanger are normally ; shall be protected from flooding, spraying, operating in each division. Therefore, if a steam lapingement, pipe whip, jet forces, LOCA occurs, the RCW aystems required to shut missiles, fire, and the effect of failure of down the plant safely are already in operation. I any non Seismic Category I equipment, as The second pump and heat exchanger in each - required. division are put in service if a LOCA occurs. (6) The safety related portion of the RCW system The mensafety related parts of the RCW system , shall be designed to meet the foregoing de. are not required for safe shutdown and, hence. , sign bases during a loss of preferred power are not safety systems. Isolation valves sepa- . (LOPP). rate the essential subsystems from the consafe. e ty related subsystems during a LOCA,in order to 92.11.12 Power Generstion Dealga Bases assure the lategrity and safety functions of the - safety related parts of tbc system. Some non. The RCW system shall be designed to cool safety related parts of the system are operated various plant auxillaries as required during: during all other modes, including the emergency ( (a) normal operation; (b) emergency shutdown; shutdown following an LOPP, or LOCA as shown in (C) normal shutdown; and (d) testing. Table 9.2 4a, b, and c, i 9J.11J System Description Surge tank water levelis monitored. A level ' switch detects leakage and isolates the non.es. l The RCW system distributes cooling water dur. sential subsystem, thus assuring continued oper. l ing various operating modes, during shutdown, and ability of the safety related services, lastru. I during post LOCA operation. The system removes ments, controls, and isolation valves are locat. beat from plant auxillaries and transfers it to ed in the safety related part of the RCW system the reactor service water system (Subsection and designed to safety grade requirements as 9.2.15). Figures 9.2 la through 9.211 show the stated in design basis (3) of Subsection 9.2.11. piping and lastrumentation diagram. Design 1.1. . characteristics for RCW system components are given in Table 9.2 4d, 9.2.11.3 Safety Evaluation g . The RCW system serves the auxiliary equipment 9.2.11.3.1 Failure Analysis listed in Table 9.2 la, b, and c. A system failure analysis of passive and The RCW system is designed to perform its active components of the RCW system is presented required safe teactor shutdown cooling function in Table 9.2 5. Any of the assumed failures of following a postulated LOCA, assutaing a single the RCW aystem are detected in the control room l active failure in any mechanical or electrical by variations of and/or alarms from the various I I system, la order to meet this requirement, the system lastruments and also from the leak detec. RCW system provides three complete trains, which tion system sensing leakage in the ECCS pump and are mechanically and electrically separated in beat exchanger areas, i case of a failure which disables any of the three divisions, the other two division meet plant safe Amnement 11 9.24
i l 2M6100AH meandard Plant RIV B , l l cvaporator. If the temperature of the chilled 9.2.14.2 System Description j water drops below a specified level, the l I controller automatically adjusts the position of 9.2.14.2.1 General Descriptica j ! the compressor inlet guide vanes. Flow switches j l prohibit the chiller from operating unless there The TCW system is illustrated on Figure i is water flow through both evaporator and 9.2 6. The system is a single loop system and q condesser. consists of one surge tank, one chemical ' I addition tank, three pumps, three heat i 9.2.14TurbineBuildingCoolingWaterSystem exchangers (connected in parallel), and " associated coolers, piping, valves, controls, , 9.2.14.1 Dealga Bases and instrumentation. Heat is removed from the l TCW system and transferred to the non safety ; 9.2.14.1.1 Safety Design Bases related turbine service water system (Subsection 9.2.16). ' l The turbine building cooling water (TCW) system serves no safety function and has no A TCW system sample is periodically taken safety design basis, for analysis to assure that the water quality , ricets the chemical specifications. 9.2.14.1J Power Generstion Dealga Bases ; (1) The TCW system provides corrosion. inhibited, deminerallred cooling water to all turbine Codes and standards applicable to the TCW ; island auxiliary equipment listed in Table system are listed in Table 3.21. The system is , i 9.2 11. designed in accordance with quality group D - (2) During power operation, the TCW system O operates to provide a continuous supply of The chemical addition tank is located in the cooling water, at a maximum temperature of turbine building in close proximity to the TCW 1050F, to the turbine island auxiliary system surge tank. . equipment, with a service water inlet tempcrature not csceeding 950F. The TCW pumps are 50% capacity each and are y constant speed electric motor driven, horizontal (3) The TCW system is designed to permit the centrifugal pumps. The three pumps are ! maintenance of any single active component connected in parallel with common suction and without interruption of the cooling discharge lines. ' function. " The TCW heat exchangers are 50% capacity ; (4) Makeup to the TCW system is designed to each and are designed to have the TCW water ; permit continuous system operation with circulated on the shell side and the power cycle design failure leakage and to permit heat sink water circulated on the tube side, expeditious post maintenance system refill. The surface area is based on normal heat load. ! (5) The TCW system is designed to have an The TCW surge tank is an atmospheric carbon atmowheric surge tank located at the steel tank located at the highest point in the , highest point in the system. TCW system. The surge tank is provided with a 4 level control valve that controls makeup water , (6) The TCW system is designed to have a higher addition. pressure than the power cycle heat sink l water to ensure leakage is from the TCW 9.2.14.2.3 System Operation : ! system to the power cycle heat sink in the event a tube leak occurs in the TCW system During normal power operation, two of the heat exchanger. threc 50% capacity TCW system pumps circulate , Amendment 11 I 9.210 T
ABWR momi Etandard Plan, arv. m inhibited demineralized water through the shell systems are preoperational'y tcsted in
<m side of two of the three 50% capacity TCW heat accordance with the requirements of Chapter 14.
I l exchangers in service. The heat from the TCW V system is rejected to the turbine service water The components of the TCW system and system which circulates water on the tube side of associated instrumentation are accessible during thc TCW aystem heat enchangers, plant operation for visual examination. Periodic inspections during normal operation are The standby TCW system pump is automatically made to ensure operability and integrity of the started on detection of low TCW system pump system. Inspections include measurements of discharge pressure. The standby TCW system heat cooling water flows, temperatures, pressures, , exchanger is placed in service vnanually, water quality, corrosion. erosion rate, control . positions, and set points to verify the system i The tooling water flow ,aie io the condition. clectro.hydraulle control (EHC) coolers, the turbine lube oil coolers and aftercoolers, and 9.2.14.5 Instrumentation Application generator exciter air cooler is regulated by , control valves. Control valves in the cooling Pressure and temperature indlestors are water outlet froni these units are throttled in provided where required for testing and i response to temperature signals from the fluid balancing the system. Flow indicator taps are . being cooled, provided at strategic points in the system for i initial balancing of the flows' and verifying The flow rate of cooling water to all of the flows during plant operation, j other coolers is manually regulated by individual theottling valves located on the cooling water Surge tank high and low level and TCW pump outlet from each unit, discharge pressure alarms are retransmitted to the rnain control room from the TCW local control The minimum system cooling water temperature panels. is maintained by adjusting the TCW system heat , h
%/
exchanger bypass valve. Makeup flow to the TCW system surge tank is initiated automatically by low surge tank water - The surge tank provides a reservoir for level and is continued until the normal level is small amounts of leakage from the system and for reestablished, the expansion and contraction of the cooling fluid with changes in the system temperature and Provisions for taking TCW system water is connected to the pump suction, samples are included. - L Demineralized makeup water to the TCW system j is controlled automatically by a level control 9.2.15 Reactor ServiceWater System , i valve which is actuated by sensing surge tank level. A corrosion inhibiter is manually added 9.2.15.1 Design Bases ; to the system. 9.2.15.1.1 Safety Design Bases 9.2.14.3 Safety Evaluation (1) The reactor service water (RSW) system The TCW system has no safety design bases shall be designed to remove heat from
- and serves no safety femton. the reactor cooling water system which
- is required for safe reactor shutdown, 9.2.11.4 Tests and Inspections and which also cools those auxillaries whose operation is desired following a l All major components are tested and LOCA, but not essential to safe s inspected as separate components prior to shutdown, installation, and as an integrated system after installation to ensure design performance. The (2) The RSW system shall be designed to :
Amendment 11 9.2 11 .
l MM 2aAs1MAH : Eenndard Plant arv s i Seismic Category I and ASME Code, (1) flooding, spraying or steam release due Section Ill, Class 3. Quality Assurance to pipe rupture or equipment failure; }
- B, Quality Group C, IEEE.279 and -i IEEE 308 requirements. (2) pipe whip and jet forces resulting from l postulated pipe rupture. of nearby high l (3) The R$W system shall be protected from energy pipes; ,
l flooding, spraying, steam impingement, ! pipe whip, jet forces, missiles, fire (3 missiles which result from equipment j and the effect of failure of any failure; and i non. Seismic Category I equipment, a : required. (4)Hre. < (4) The RSW aystem shall be designed to meet Liquid radiation monitors are provided in the the foregoing design bases during a loss RCW system. Upon detection of radiation leakage , of preferred power, in a division of the RCW system, that system is i' isolated by operator action from the control 9.2.18.1J Power Generation Dealga Bases room, and the cooling load is met by another ! 4 division of the RCW system. Consequently, ' The RSW system shall be designed to cool the tedioactive contamination released by the RSW reactor building cooling water (RCW) as required system to the environment does not exceed 4 during: (a) normal operation; (b) emergency allowable limits defined by 10CFR100. I shatdown; (c) normal shutdown; and (d) testing. System low point drains and high point vents 9.2.15J System Description are provided as required. The RSW system provides cooling water during System components and piping materials are ! various operating modes, during shutdown and selected to be compatible with the available post.LOCA operations. The system removes heat site cooling water in order to minimize ; l frca the RCW system and transfers it to the corrosion. Adequate corrosion safety factors ultimate heat sink. Figure 9.2 7 shows the RSW are used to assure the integrity of the system ; system diagram, during the life of the plant. The RSW system is able to function during During all plant operating modes each abnormally high or low water levels and steps are division shall have at least one service water ' taken to prevent organic fouling that may degrade pump operating. Therefore,if a LOCA occurs, , system performance. These steps include trash the system is already in operation, if a loss racks and provisions for blocide treatment (where of offsite power occurs during a LOCA, the pumps L discharge is allowed). Where discharge of momentarily stop until transfer to standby ! blocide is not allowed, non blocide treatment diesel. generator power is completed. The pumps i will be provided. Thermal backwashing capability are restarted automatically according to the will be provided at sea water sites where diesel loading sequence. No operator action is infestations of macrobial growth can occur, required, following a LOCA, to start the RSW , system in its LOCA operating mode.
- 9.2.15J Safety Evaluation ,
9.2.15.4 Testing and Inspection Reyletments l The components of the RSW system are ! separated and protected to the extent necessary The RSW system .is designed for periodic to assure that sufficient equipment remains pressure and functional testing to assure: , operating to permit shutdown of the unit in the event of any of the fellowing (Separation is (1) the structural and lenktight integrity applied to electrical equipment and by visible inspection of the components; , instrumentation and controls as well as to i mechanical equipment and piping.): i Amendment 11 9.2 12 l
i ABM maastard Plant swwmi arv s
]
J f- (2) the operability and the performance of component without laterruption of the l
' active the cotoponents of the system; and cooling function.
(3) the operabillty of tbc system as a 9.1.16.2 System Denedption l whole. i 9.1.16.2.1 General Description The tests shall assure, under conditions as ! close to design as practical, the performance of The TSW system is illustrated on Figure : the full operational sequence that brings the 9.28. , system into operation for reactor shutdown and for LOCA, including operating of applicable The TSW pumps take suction from the power ; portions of the reactor protection system and the cycle heat sink and supply coollag water to the , transfer between normal and standby power tube side of the TCW beat exchangers. The heat i sources, rejected to the TSW system is discharged to the + power cycle heat sink. ! 9.2.18.5 Instrumentation and Control , Itequirements Piping and valves in the TSW system are ; carbon or low alloy steel and are protected from l Locally mounted temperature indicators or interior and exterior corrosion with suitable : test wells are furnished on the equipment cooling corrosion resistant material as required by site watcr discharge lines to enable verification of specific soll and water conditions. specified heat removal during plant operation. 9.2.16 Titrbine Service Water System The TSW systern consists of two 100% capacity The turbine service water (TSW) system vertical wet pit pumps located at the intake supplies cooling water to the turbine cooling structure. One pump is in operation during O water (TCW) system heat exchangers to transfer normal operation with one pump on standby. heat from the TCW system to the power cycle heat sink. Two 100% capacity duplex strainers are provided (one for each TSW pump). Each half of 9.2.16.1 Design liases the duplex strainer is designed for the design ; flow of one TSW pump. Only one half of each 9.2.16.1.1 Safety Design Innes duplex strainer is in operation when its associated pump is in operation. The duplex , The TSW system does not serve or support any stralacrs are motor operated and automatically safety function and has no safety design basis. switch from the half in service to the clean half of the duplex strainer on detection of high 9.2.16.1.2 Power Generation Design itases differential pressure. Debris collected in the strainer is automatically sluiced to a disposal - (1) The TSW system is designed to remove collection area, heat from the turbine cooling water (TCW) system heat exchangers and reject The TSW pumps supply cooling water to the two this heat to the power cycle heat sink TCW heat exchangers (one is normally in senice during normal and shutdown conditions. and one is on standby). (2) During normal power operation the TSW A summary of the TCW heat exchangers is , system supplies cooling water to the TCW provided in Table 9.212. system heat exchangers at a temperature not csceeding 1000F, 9.2.16.2.3 System Operation (3) The TSW system is designed to permit the The system normally is started manually from
- - maintenance of any single active the main control room and one pump is operated Amendment 11 9.2 12.1
i I ABWR zwei i RemmAard jblant ILEV B ! continuously during normal power operation ' conditions. The standby pump is started automatically in the event the normally operating pump trips or the discharge bender pressure drops below a ! preset limit. l
)
9116J SafetyEvalention ] 1 l Tbc TSW system does not serve or support any 1 safety function and has no safety dealga bases. 9116.d Testa and Inspoetions l All major cotoponents are tested and inspected
. as separate compontats prior to lastallation, and as an lategrated system after installation to ,
ensure design performance. Tbc systems are preoperationally tested in accordance with the requirements of Chapter 14. The components of the TSW system and associated instrumentation are accessible during plant operation for visual examination. Periodic , nspections during normal operation are made to ensure operability and integrity of the system. laspections include measurement of the TSW system flow, temperatures, pressures, differential pressures and valve positions to verify the ! system condition, e j 9.2.16J lastrumentation Application Pressure and temperature indicators are provided where required for testing the system. TSW system pump status is indicated in the main contrel room. - TSW system trip is alarmed and the automatic - startup of the standby pump is annunciated in the l main control room. High differential pressure across the duplex , filters is alarmed in the main control room. O Amendment 11 9,2 12.2
ABM 2 m ia m i arv a i me a.ed Plant l
-l t.t.17 lateefaces l 9.2.17.1 Utahmate Heat sink Capabluty j t
- The ultimate best sink shall be capable of
- I L dissipating reactor decay best and essential teclieg system heat loads after a normal reactor ,
shutdown er a shutdown following an accident, ' including LOCA. The amount of heat to be dissipated under normal and accident conditions is listed in Table 9.2 4 The ultimate heat sink and any pumps, valves, i structures, or other components that remove heat L l from safety systems shall be designed to Seismic Category I and ASME Code, Section !!!, Class 3, l Quality Assurance B, Quality Group C,IEEE 279, , I and IEEE.308 requirements. The safety related ' i portions shall be protected from flooding, , spraying, steam lapingement, pipe whip, jet .' forces, missiles, fire and the effect of failure l of any non.Scismie 1 equipment. The safety > related portions of these systems shall be , designed to meet the above mentioned design bases during a loss of offsite power. The safety l related portions of these systems shall be designed to perform their required cooling function assuming a single active failure in any i mechanical or electrical system. The divisions : of these systems shall be mechanically and electrically separated. 9.2.17.2 Makeup Water System Capability The raw water treatment and preparation of the demineralized water is sent to the makeup water system (purified) described in Subsection 9.2.10. , O Amendment 11 9213
+
L ! MNS 21A6100MI i [ Etandard Plant ny n l l .( Table 9.212
. SUhntARY OF TURBINE COOLING WATER SYSTEM HEAT EXCHANGERS ,
Egelpment Number f Description /In Use l Plant Chillers $/4 I Gen stators coolers 2/1 Gen H2 coolers 2/2 i Gen }{2 seal oilcooler 4/2 ! Gen exciter cooler ' 1/1 Gen breaker cooler 1/1 Turbine lube oil coolers 2/1 t Mech vacuum pump cooler 1/1 ; Isolated phase bus cooler 1/1 . Air compress & aftercooler 3/3 [ EHC coolers 2/1 RFP variable speed motor coolers 2/2 . RFP motor thyrister coolers 2/2 i Standby RFP motor coolers 1/0 Condensate pump motor coolers 4/3 licater drain pumpo coolers 2/2 i O ! i-1 r O , Atacadment 11 p.2 25c t i
O O O I i le> a
- 135 2 E,$
$ RCW HEAT 4
EXCHANGERS 4 ii" 3
~
i __ . , MX- -M- NVV\r X l FROM Ik Ut.TasATE -->
'~
U HEAT SINK RCW SYSTEM ! TO i e
= ULTas4TE !
.; HEATSBSC , 4 ) - 1 i NX- O M -M i k FROM ULTasATE --> "' STRAINERS y HEATSINK RCW SYSTEM REACTOR SERVICE WATER PUMPS 5 e m= j A sf_. j Figure 9.2-7 REACTOR SERVICE WATER SYSTEM L
, _ . ,_ s _ = , . . . _ - , . _ , ~ - - , .. ._- . . __. -.____ _ ____ _ _.__________._.______ _ __
?
t
? La> i e gg I k m.,M
. TCW HEAT EXCHANGERS
[% ., 4 , ! 3
~
i JL FROM POWER CYCLE + - 1I t HEAT SIIGC TCW SYSTEM : TO 5 POWER CYC1.E HEATSPGC i 2 i 1 Mx- -N- Nvv\r -M-I il 1 FROM STRAINERS I POWER CYCLE + - 1I i ' HEAT SINK TCW SYSTEM i i TURBWEE BUIUMNG
- SERVICE WATER PUEFS N 1-m A
ae
<R e
e yy Figwe9.2-8 TUR91NESERVICEWATERSYSTEM O . O O
y :. e i-M 21A6100AH Standardflant nv n < 9J PROCESS AUXILIARIES 9.3.1 Compressed Air Systems The instrument air system is discussed in Sec-l tion 9.3.6 and the service air system is discus- 4 sed in Section 9.3.7. Neither of these systems are safety related. The atmospheric control sys-tem and the high pressure nitrogen system provide ! nitrogen gas for safety related uses. They are discussed in Subsection 6.2.5 and Section 6.7, _. respectively, i 1
}
\
O ! i i i ( i I i O Amendment 11 911
e 21A6100AH Etandard Plant un and heater system status is provided locally at defined in Subsection 9.3.5.2 and maintained [ the storage tank. Table 9.31 contains the pro- above saturation temperature. cess data for the various modes of operation of the SLCS. Seismic category and quality class are Cooldown of the nuclear system will require a included in Table 3.21.. Principals of system minimum of several hours to remove the thermal testing are discussed in Subsection 9.3.5A. energy stored in the reactor, cooling water, and associated equipment. The controlled limit for 93.53 SafetyEvaluation the reactor . vessel cooldown is 100'F/hr., and normal operating temperature is approximate-The SLCS is a reactivity control system and is , ly 550'F. Use of the main condenser and ; . maintained in an operable status whenever the various shutdown cooling systems requires 10 to r reactor is critical. The system is never 24 hours to lower the reactor vessel to room. ' expected to be needed for safety reasons because temperature (70'F); ~ this is the condition ;-
~
of the large number of independent control rods of maximum reactivity and, therefore, is the available to shut down the reactor, condition that requires the maximum concen-tration of boron. To assure the availability of the SLCS, two ' sets of the components required to actuate the The specified boron injection rate is limited system (pumps and injection valves) are provided to the range of 8 to 20 ppm / min. The lower rate in parallel redundancy, assures that the bwoar ie lejected into the reactor in approximately two and one half ' The system is designed to bring the reactor' hours. This resulting reactivity insertion is from rated power to a cold shutdown at any time considerably quicker than that covered by the - in core life. The reactivity compensation cooldown. The-upper limit injection rate provided will reduce reactor power from rated to - assures that there is sufficient mixing so~ that zero level and allow cooling of the nuclear boron does not_ recirculate through the core in system to room temperature, with the control rods uneven concentrations that could possibly cause 1 ( remaining withdrawn in the rated power pattern. reactor power to rise and fall cyclically. ' It includes the reactivity gains that result from complete decay of the rated power xenon The SLCS equipment essential for injection of i- inventory. It also includes the positive . neutron absorber solution into the reactor is-reactivity effects from eliminating steam voids, designed as Seismic Category I for withstanding ! changing water density from hot to cold, reduced the specified earthquake loadings (Chapter 3). Doppler effect in uranium,' reducing neutron The system piping and equipment are designed, leakage from boiling to cold, and decreasing installed, and tested in accordance with the control rod worth as the moderator cools, requirements stated in Section 3.6. ; The minimum average concentration of natural The SLCS is required to be operable in the boron required in the reactor core to provide event of a plant offsite power failure; there-adequate shutdown margin, after operation of the - fore, the pumps, heater, valves, and controls SLCS, is 800 ppm (parts per million). Calcula; .are powered from the standby AC power supply. tion of the minimum quantity of sodium penta. The pumps and valves are powered and controlled borate to be injected into the reactor is based from separate buses and circuits so'that a sin-on the required 800 ppm average concentration in gle. active failure will not prevent system the reactor coolant at 70'F and reactor operation. normal water level. The result is increased by 25% to allow for imperfect mixing and leakage. The SLCS and pumps have sufficient pressure Additional sodium pentaborate is pron /tl to margin, up to the system relief valve setting of accommodate dilution by the RHR system in the approximately 1560 psig, to assure solution shutdown cooling mode. This concentration will injection into the reactor above the normal be achieved if the solution is prepared as pressure in the bottom of the reactor. The Amendment 2 9.34 l l-L i [
- m . .
.m m _ _ _ . . _ __ -. _ _-.
u
, 23A6100AH mandard Plant ma nuclear system safety / relief valves begin to Crlierlon 26 The SLCS is the second reactivity :
relieve pressure above approximately 1100 psig, control system required by this criterion. ! Therefore, the SLCS positive displacement pumps cannot overpressurize the nuclear system. , Only one of the two standby liquid control criterlon 27: This criterion applies no-
^ pumps is needed for system operation. However, specific requirements onto the SLCS and.
If needed, both pumps can be operated at the same therefore is not applicable._ See the General "
' time. If a redundant component (e.g., one pump) _ Design Criteria Section for discussion of is found to be inoperable, there is no immediate' combined capability, threat to shutdown capability, and reactor .
operation can continue during repairs. The time criterion 29: The SLCS pumps and valves out- ! during which one redundant component upstream of board of the outboard drywell check valve are the injection valves may be out of operation redundant. Two suction valves, two pumps, and = should be consistent with the following: . the two injection valves are arranged and crosstled i probability of failure of both the control rod _ such that operation of any one of each results : shutdown capability and the alternate component . In successful operation of the system. The SLCS , ! in the SLCS; and the. fact that nuclear system also has test capability. A special test tank-cooldown takes several hours while liquid control: is supplied for providing test fluid for the solution injection takes approximately Lycarly injection test. Pumping capability,- two and one half hours. Since this probability l injection valve operability and suction valve
. is small, consi derable time is available for L operability may be tested at any time, repairing and restoring the SLCS to an operable condition while reactor operation continues. . The SLCS is evaluated against the applicable - !
Assurance that the system will still fulfill its regulatory guldes as follows: i function during repairs is obtained by. demonstrating operation of the operabic pump. Reculatory Guide 126: Because the SLCS is a .; reactivity control system, all mechanical The SLCS is evaluated against the applicable components are at least Quality Group B. Those i General Design Criteria as follows: portions which are part of the reactor coolant pressure boundary are Quality Group A. Tbh is , criterion 2: The SLCS is located in the area shown in Table 3.21. I inside the secondary containment, outside drywell and below the refueling floor. In this location,' Reculatory Guide 1.29: - All components of the it is protected by the containment and SLCS which are necessary for injection of compartment barriers from external natural neutron absorber into the reactor are Seismic l phenomena such as carthquakes, tornadoes, Category I. This is shown in Table 3.21. hurricanes and floods and internally from effects of postulated events (e.g., DBA LOCA). ASB 3-1 and MEB 3-1 criterion 4: The SLCS is designed for the Since the SLCS is located within its own expected environment in the secondary conteinment compartment inside the secondary containment, it , and specifically for the area in which it is is adequately protected from flooding, torua. located. In this area, it is not subject to the does, and internally / externally generated more violent conditions postulated in this missiles. SLCS equipment is protected from pipe e criterion such as missiles, whipping pipes, and break by providing adequate distance between the seismic and nonseismic SLCS equipment, where discharging fluids, ~ such protection is necessary. In addition, criterion 21: Criterion 21 is applicable to appropriate distance is provided between the protection systems only. The SLCS is a reactiv- SLCS and other high energy piping systems. ity control system and should be evaluated against Criterion 29. O, Amendment 11 9.35 l
l M 23A6100AH Standard Plant mm s 9.3.9 Hydrogen Water Chemistry System feedwater system, the lower plenum region and the ( RWCU inlet, hydrogen and pH levels in the 9J.9.1 Design Bases feedwater system, the lower plenum region and the
. RWCU inlet, and crack growth of pre cracked 1 93.9.1.2 Safety Design Basis ;
samples in water from the lower plenum region. The hydrogen water chemistry (HWC) system The hydrogen supply system will be site ! is non nuclear, non safety-related and is required to . . dependent. Hydrogen can be supplied either.as a be safe and reliable, consistent with the requirement high pressure gas or as a cryogenic liquid. Hydrogen of using hydrogen gas. The hydrogen piping in the and oxygen can also be generated on site by the - turbine building shall be designed to Seismic dissociation' of water by electrolysis. The HWC Category l requirements to comply with BTP 9.51 - hydrogen supply system is integrated with the + generator hydrogen supply system to save the cost of 93.9.1.2 Power Generation Dealgn Basis ' having separate gas storage facilities for both systems. , r BWR reactor coolat is demineral! zed water, . The oxygen supply system will be site ' 4 typically containing 100 to 200 parts per billion (ppb) dependent. A single oxygen supply system could be dissolved oxygen from the radiolytic decomposition provided to meet the requirements of HWC system - of water. To mitigate the potential for intergranular and the condensate oxygen lujection system described , stress corrosion cracking (IGSCC) of sensitized in Subsection 9.3.10, austenitic stainless steels, the dissolved oxygen in the reactor water can be reduced to less than 20 ppb by - 9.3.9.3 Safety Evaluation the addition of hydrogen to the feedwater.- The ! amount of hydrogen required is in the range of 1.0 to The operation of the HWC system is not { 1.5 ppm. The exact amount required depends on necessary to assure: ; . many factors including incore recirculation rates. '
- The amount required will be determined by tests (1) The integrity of the reactor coolant pressure h
J performed during the initial operation of the plant, boundary,- [ The concentration of hydrogen and oxygen in (2) ' re capability to shut down the reactor; or . the main steam line and eventually in the main condenser is altered in this process. This leaves an (3) The capability to prevent or mitigate the excess of hydrogen in the snain condenser that would consequences of events which could result in ( , not have equivalent oxygen to combine with in the - potential offsite exposures. l offgas system. To maintain the offgas system near its . l normal operating characteristics, a flow rate of The HWC system is used, along with other oxygen equal to approximately one half the injected . measures, to reduce the likelihood of corrosion , hydrogen flow rate is injected in the offgas system failures which would adversely affect plant availability.
- upstream of the recombiner. - The means of storing and handling hydrogen shall utilize the guidelines in EPRI NP 5283 SR A, i The HWC system utilizes the guidelines given 'Ouldelines for Permanent BWR Hydrogen Water in EPRI report NP 5283 SR A,' Guidelines for ChernistryInstallations'. .,
, Permanent BWR Hydrogen Water Chemistry ' l Installation". 93.9A Inspection and Testing Requirements b 93.9.2 System Description The HWC system is proved operable during the initial operation of the plant. During a refueling or The HWC system, illustrated in Figure 93-8, is maintenance outage, hydrogen injection is not composed of hydrogen and oxygen supply systems, required. System maintenance or testing can be systems to inject hydrogen in the feedwater and performed during such periods. . oxygen in the offgas and subsystems to monitor the effectiveness of the HWC system. These systems monitor the oxygen levels in the offgas system, the Amendment 11 9.3-12 My -,ig v. y. - ~m .- -- eys
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Standard Plant - am n - i 93.9.5 Instrumentation and Controls (5) Sensors for measuring dissolved oxygen content.- (6) Sensors for measuring pH and dissolved Automatic control features in the HWC system minimize the need for operator attention and hydrogen. . , improve performance. These are: L . (7) A system for verifying the effectiveness of HWC , I (1) - Automatic variation of hydrogen and oxygen by measuring electrochemical potential (ECP) , flow rates with reactor power level. . and crack growth rate. ; (2) Automatic oxygen injection rate change delay. 9.3.10 Oxygen Injection System - This function is also augmented as a function of reactor power level. 93.10.1 Design Bases (3) Automatic shutdown on several alarms. The oxygen injection system is designed to add suffielent oxygen to the Condensate System to. (4) -Isolation on system power loss, operator suppress corrosion and corrosion product release in restart. the condensate and feedwater systems Experience has shown that the preferred feedwater oxygen - (5) Reptogrammable alarms and controller concentration is 20 to 50 ppb. During shutdown and electronics, startup operation the feedwater oxygen concentration ., is usually much above the 20 to 50 ppb range. ! (6) - Hydrogen and oxygen flow monitor correction However, during power operation, deseration in the function to compensate for nonlinearitics, main condenser may reduce the condensate oxygen concentration below 20 ppb, thus, requiring that some - The recommended trips of the oxygen and oxygen be added. The amount required is up to ' hydrogen injection systems include: approximately 5 cubic feet per hour.- (1) ReactorKeram 93.10.2 System Description (2) Low or high residual oxygen in the off gas 1 The oxygen supply consists of high pressure gas - k-cylinders or a liquid tank, 'A condensate oxygen , (3) High area hydrogen concentration injection module is provided with pressure regulators and associated piping, valves, and controls to t (4) Low oxygen injection system supply pressure depressurize the gaseous oxygen and route it to the condensate injection modules. There are check valves .
. (5) High hydrogen flow and isolation valves between the condensate injection modules and the condensate lines downstream of the The instrumentation provided includes: condensate demineralizers and the optional injection point upstream of the filters.
(1) Flow monitors for measurement of hydrogen and oxygen flow rates. The flow regulating valves in this system are operated from the main control room. The oxygen (2) Hydrogen area monitor sensors to detect any concentration in the condensate /feedwater system is hydrogen to the atmosphere, monitored by analyzers in the sampling system (Subsection 93.2). An' operator will make changes in , (3) Pressure gages for measurement of hydrogen the oxygen injection rate in response to changes in the l and oxygen supply pressures and instrument air condensate /feedwater concentration. An automatic l- pressure, control system is not required because instantaneous changes in oxygen injection rate are not required. . (4) An oxygen analyzer for measuring the percent oxygen leaving the offgas recombiner. Amendment 11 9.3 13 O I
I AMW 234siooxa )
- Standard Plant au n ;
L l l 93.103 SafetyEvaluation in the reactor water is 10 to 15 ppb zine during an ! I initial conditioning period and 5 to 10 ppb over the ! l The oxygen injection system is not required to fuel cycle. assure any of the following conditions. A dilute zinc solution is prepared and injected I (1) integrity of the reactor coolant pressure into a bypass loop around the feedwater pumps. ! boundary; I. 93.11J SafetyEvaluation l- (2) capability to shut down the reacto* and maintain it in a safe shutdown condition, or The injection system is not necessary to ensure: (3) ability to prevent or mitigate the consequences (1) the integrity of the reactor coolant pressure of events which could result in potential offsite boundary; - exposures. (2) the capability io shut down the reactor; or , Consequently, the injection system itself is not l safety related. The high pressure oxygen storage (3) the capability to prevent or mitigate the bottles are located in an area in which large amounts consequences of events which could result in ' of burnable materials are not present. Usual safe potential offsite exposures. practices for handling high pressure gases are followed. The zine injection system will help keep radiation levels as low as possible, thus, reducing 93.10.4 Tests and Inspections personal exposure especially during outages. ! , ' The oxygen injection system is proved operable 93.113 Test and Inspections by its use during normal operation. The system E" valves may be tested to ensure operability from the The zine injection system is proved operable O main control room, 93.103 Instmmentation Application _during initial operation of the plant. Zine injection will not be performed when the plant is in cold shutdown. During these periods, the system can have The oxygen storage bottles have pressure gages which willindicate to the operators when a new 93.11.4 Instmmentation- . l bottle is required. A flow element will indicate the oxygen gas flow rate at all times. The gas flow The injection of zine solution will be stopped regulating valves will have position indication in the automatically if feedwater flow stops. The zinc main controt room. injection rate is manually adjusted based on zine concentration data in the reactor water. The oxygen monitors are discussed in Subsection 93.2. 93.11 Zdne injection System 93.11.1 Design Bases l The continuous presence of small amount of I. dissolved zine in the reactor water has been shown to 4 reduce radiation levels on primary system surfaces. l 3' Zinc injection shall be initiated during reactor 83 startup tests when high temperature operation l' commences. The amount of dissolved zine required Amendment 11 9.3 13.1
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. CHAPTER 9.4 CONTENTS i
Sesilan 1111e' P.ast l l 4 9.4.1 Control Buildina HVAC 9.41 j
'I 9.4.1.1 Control Room Equipment HVAC 9.41 9.4.1.1.1 - Design Basis ' 9.4 1 9.4.1.1.2 Power Generation Design Basis _ 9.41_
9.4.1.13 System Description 9.4 1 9.4.1.1.4 Safety Evaluation ' L 9.4-la 9.4.1.1.5 Inspection and Tasting Requirements 9.4-la 9.4.1.1.6 Instrumentation Application . 9,4 la 9.4.1.1.7 Regulatory Guide 1.52 Compliance Status 9.41b - ' 9.4.1.1.8 Standard Review Plan 6.5.1 Compliance
' Status- 9.41b
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9.4,1,2 Essential Electrical an'd Reactor Building Cooling Water Equipment HVAC - - 9.4-1b ' j 9.4.1.2.1 Design Basis 9.4-1b 9,4.1.2.2 Power Generation Design Basis 9.4-1b 9.4.1.23 System Description 9.41b.1 ' 9.4.1.23.1 Safety Related Subsystem 1 _ 9.4-Ib.1 9.4.1.23.2 Safety Related Subsystem 2 9.4-le 9.4.1.233 Safety Related Subsystem 3 9.4 Ic f 9.4.1.2.4 Safety Evaluation . 9.4-le 9.4.1.2.5 Inspection and Testing Requirements 9.4-Ic 9.4.1.2.6 { Instrumentation Application 9.4-1d - , 9.4.2 Snent Fuel Pool Area Ventilation System 9.42 ( 9 . 4 - 11 Amendment 11 i k 4
. j 23A61DQAH r Standard Plant amn . SECTION 9.4. h CONTENTS (Continued)- i s i Section Ildt East
- , i 4 i 9.43- Auxiliary Area Ventilatlan Svatem 9.42. !
I h 4 9.4.4 Thrbine Area Ventilatlan Svatem ^-9.42 1 j j 9.4.4.1 Design Bases 9.42 l 4'
- Safety Design Bases 9.4-2
[ 9.4.4.1.1 ! 9.4.4.1.2 Power Generation Design Bases 9.42 , l 1 ' 9.4.4.2 Description - 9.4 2' . I Gen'eral Description '9.4 9.4.4.2.1 t 9.4.4.2.1.1 Turbine Building Exhaust System 9.4-2a i- 9.4.4.2.1.2- Turbine Building Battery Room Exhaust Subsystem : ' 9,4 2a . i 9.4.4.2.1.3' Turbine Building Electrical Switchgear and Air Compressor Rooms Ventilation Subsystem 9.4-2a . l 9,4.4.2.1.4 Turbine Building Lube Oil Area Exhaust Subsystem ' 9.4 2a
,- 9.4.4.2.1.5 TBVS Unit Coolers and Electric Unit Heaters 9.4 2a l
- l. 9.4.43 Evaluation 9.4 2a ;
I 9.4.4.4 Tests and Inspections 9.4-2b . 9.4.4.5 Instrumentation Application 9.4-2b 9.4.5 Reactor Building Ventilation System 9.4 2c , 9.4.5.1 Secondary Containment HVAC Sptem 9.4-2c. , 9.4.5.1.1 Design Bases 9.4 2c 9.4.5.1.1.1 Safety Design Bases 9.4 2c 9.4 iii . Amendment 6 O:i
7 ABWR maiowi : is Standard Plant n- n ' O .4 ira coND1TioxiNo
~ HEATING OLING AND (5> The outside design conditions fo, the controi ,oom HVAC system are 115 F during the summer and VENTILA ON SYSTEMS 40 F during the winter. :
L 9.4.1 Control Building HVAC 9.4.1.1J System Description j t The control building HVAC system is divided The control room is heated, cooled and pressurized into two separate systems. A HVAC system for the by a recirculated air system with filtered outdoor air , control room equipment on the top two floors. Plus for ventilation and pressurization purposts. The l a HVAC system for essential electrical and heat ex. recirculated air and the outdoor air will be mixed and . L hanger equipment. drawn through a filter section, a hcating coil section, i and a cooling coil section. Under normal conditions, f 9.4.1.1 Control Room Equipment HVAC sufficient air is supplied to pressurize the control [ room and exfiltrate to pressurize the control building.' i 9,4.1.1.1 Design Basis _ _ . The control building HVAC PFD is shown in (1) The Control Room heating, ventilating and Figure 9.41 The control building recirculation unit m air conditioning (HVAC) system is designed consists of a prefiker section, a high efficient section, . with sufficient redundancy to ensure operation an electric henier, a cooling coil, two 50% capac ity under emergency conditions assuming the supply fans. The supply fans are placed on low speed-single failure of any one active component, when the system is in the smoke removal mode. (2) Provisions are made in the system to detect and Two 50% capacity return exhaust fans draw air limit the introduction of airborne radioactive from the electrical area, corridors, control room, materialin the control room, computer room, office areas, and the HVAC equip-O (3) Provisions is made in the system to detect and remove smoke and radioactive material from ment room. This air is returned to the air condition-ing unit during normal operations. Modulating damp-ers in the return duct work to the fans are controlled the control room, by a pressure controller to maintain the required positive pressure. The controller is located in the (4) The HVAC system is designed to provide a electrical equipment area. During smoke removal , controlled temperature environment to ensure mode, these fans are placed on low speed and the air.
- l. the continred operation of safety-related equip- is discharged to atmosphere, ment under accident conditions.
An emergency recirculation system consisting of an (5) The HVAC system and components are lo- electrical heating coil, a prefilter, HEPA filter, chat-cated in a Seismic Category I structure that is - coal adsorber, and HEPA filter with a booster fan, is tornado missile and flood protected. provided parallel to the normal mixed outdoor and return air path to the supply conditioning units. The - 9.4.1.1.2 Power Generation Design Basis charcoal adsorber will be 6 inches deep as a minimum. The system is normally on standby for use only during - (1) The HVAC system is designed to [ provide an high radiation. A radioactivity monitoring system environment with controlled temperature and monitors the building intakes for radiation. The humidity to ensure both the comfort and safety radiation monitor allows the control room operator to of the operators. The nominal design condi. select the safest intake. The makeup air for tions for the control room environment are pressurization can be diverted through the HEPA and
- 75 F and 50% relative humidity, charcoal adsorbing system before distribution in the control room areas.
(2) The system is designed to permit periodic in-spection of the principal system components. Smoke detectors in the control room and the con-trol equipment room exhaust systems actuate an Amendment 11 9.4-1
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Standard Plant am a '1 I alarm on indication of smoke. Dampers must be po- The safety related isolation valves at the outside air i sitioned through a remote manual switch to allow the intakes are protected from becoming inoperable due j exhaust air fans to exhaust 100% of the conditioned to freezing, icing, or other environmental conditions.
'1 air. _
9.4.1.1J Inspection and Testing Requirsewats 1 The HVAC equipment space is physically sepa- .] rated into divisional rooms. Each divisional room Provisions are made for periodic tests of the out- ' consists of an air intake room and an air exhaust door air cleanup fans and filters. These tests include room. determinations of differential pressure across the; ; filter and of filter efficiency. Connections for testing, -)
' 9.4.1.1.4 Safety Evaluation such as injection, sampling and monitoring are prop- ,
[ erly located so that test results are indicative of per. The control building HVAC System is designed to formance. : maintain a habitable environment and to ensure the. operability of components in the control room. All The high efficiency particulate air (HEPA) filters q control room HVAC equipment and surrounding may be tested periodically with dioctyl phthalate - structures are of Seismic CategoryI design and oper- . smoke (DOP). The charcoal filters may be periodi- 1 able during loss of the offsite power supply. . cally tested with freon for bypasses.
. The ductwork which services these safety functions LThe balance of the system is proven operable by its is termed ESF ductwork, and is of Seismic Category - use during normal plant operation. Portions of the f
I design. ESF ducting is high pressure safety grade system normally closed to flow can be tested to l 51uctwork designed to withstand the maximum posi- ensure operability and integrity of the system, e i tive and/or negative pressure to which it can be sub- . j jected under normal or abnormal conditions. Galva- 9.4.1.1.6 Instrumentation Application - ! nized steel ASTM A526 or ASTM A527 is used for - outdoor air intake and exhaust ducts. All other ducts The area exhaust fan is started manually and the fan ( are welded black steel ASTM A570, Grade A or . discharges the air to atmosphere. Grade D. Ductwork and hangers are Seismic Cate-l_ gory 1. Bolted flange and welded joints are qualified A high radiation signal automatically starts the out-per ERDA 76-21. door air cleanup system, closes the normal air inlet damper and closes the exhaust air dampers.- Redundant components are provided where neces-sary to ensure that a single failure will not preclude A temperature indicating controller senses the tem-adequate control room ventilation. perature of the air leaving the air cleanup system. The controller then modulates an electric heating coil 3 l_ A radiation monitoring system is provided to to maintain the leaving air temperature at a preset i I ' detect high radiation in the outside air intake ducts, limit. A limit switch will cause an alarm to be actu. A radiation monitor is prosided in the control room ated on high air temperature. A moisture sensing j to monitor control room area radiation levels. These element working in conjunction with the temperature j monitors alarm in the control room upon detection controller measures the relative humidity of the air ! of high radiation conditions. Isolation of the control entering the charcoal adsorber. room and initiation of the outdoor air cleanup unit ] fans are accomplished by the following signals: Differential pressure indicators show the pressure j drop across the prefilters and the HEPA filters. A l (1) high radiation in the outside air intake duct, differential pressure indicating switch also measures l and- the pressure drop across the entire filter train. The ' (2) manualisolation. switch causes an alarm to be actuated if the pressure I drop exceeds a preset limit. A flow switch in the out-Under normal conditions, sufficient air is supplied door air cleanup system fan discharge duct automati-to pressurize the control room and exfiltrate to pres- cally starts the standby system and initiates an alarm surize the Control Building. on operating fan failure. O l Amendment 5 9A-la I
i ABWR 234siooxii Standard Plant am n J The electrical equipment area and the control the heater and demisters, The heaters and demisters - l
' O room area return exhaust fans start automatically are put into systems to regulate the relative humidity I when the air conditioning unit is started. Each fan of tbc air as it enters the ESF filter train. Since the -
- inlet damper is'open automatically. The exhaust control room air handling units are designed to l dampers to the conditioning unit are opened maintain the control room emperature and humidity j automatically, within limits, additional enntrols are not necessary l for the ESF filter train.
Differential pressure indicating controllers ; modulatelate dampers in the return air ducts to- 9.4.1.1.8 Standard Review Plan 63.1 Compliance maintain space positive pressure requirements. Status In the event of an alarm from the smoire detection . The control room ESF system complies with SRP system, the exhaust fans are placed in the smoke re- 6.5.1, Table 6.5.11. The only exceptions are for moval mode manually If th electrical equipment heater and moisture separator instrumentation area is to be exhausted, the smoke removal exhaust requirements. Since these components are not
- damper switch is actuated. The air conditioning unit - necessary for the ABWR design, no instrumentation supply fans and the eletrical equipment room ex- has been supplied to monitor their operation, ,
haust fans are automatically put in the low speed po. Relative humidity and temperature of the inlet air is l sition, the exhaust damper is opened and the return maintained by the control room air. handling system. } air damper to air condotioning unit closed. The I area differentail pressure indicating controller goes 9.4.1.2 Essential Electrical and Reactor Building to full open position during the smoke removal Coollag Water Equipment HVAC mode. 9.4.1.2.1 Design Basis The cooling unit starts automatically on a signal from the temperature indicating controller installed (1)The heating, ventilation and air conditioning i 7 in the HVAC room. The controller modulates a (HVAC) system is designed with sufficient ( three-way chilled water valve to maintain the space redundancy to ensure operation under N conditions, emergency conditions assuming the failure of " any one active component. During winter, the electric unit heaters are cycled by temperature indicating controller switches, (2)The HVAC system is designed to provide a l located within the filter rooms and the air handler controlled temperature environment to ensure rooms. the continued operation of safety related equipment under accident conditions. The supply and return air duct work has manual balancing dampers provided in the branch ducts for (3) The HVAC system and components are located balancing purposes. The dampers are locked in in a seismic Category I structure that is place after the system is balanced, tornado-missle and flood protected.~ 9.4.1.1.7 Regulatory Guide 132 Compliance Status 9.4.1.2.2 Power Generation Design Basis The control room ESF filter trains comply with (1)The HVAC system is designed to provide an all applicable provisions of Regulatory Guide 1.52, environment with controlled temperature during Section C except as noted below. ' normal operation to ensure the comfort and safety of plant personnel and the integrity of the The revisions of ANSI N509 and N510 listed in essential electrical and RBCW equipment. Table 1.8 21 are used for ABWR ESF fliter train design; the Regulatory Guide references older (2)The system is designed to facilitate periodic revisions of these standards. inspection of the principal system components. The control room ESF fliter trains are in compliance with the system design criteria except for Amendment 11 9A-lb
ABM Standard Plant 224siaatii. mu n i (3) Design outside alt temperature for the heat exchanger building HVAC system are 115 F during the summer and 40*F during winter. 91 -l l 2 9A.1JJ System Description l The essential electrical HVAC System is divided into 3 subsystems with each subsystem serving a des- < ignated area. Each Subsystem serve as essential elec. , ! trical heat =' ; -r equipment K/AC for divisions : A, B, and C. ' 9A.1JJ.1 Safety Related Subsystem 1 System 1 speciGcally serves: (1) Safety related battery rooms 1 and 4,
. (2) .- Essential chiller room A, (3) RB cooling water pump and heat exchanger J room A,' i
. (4) ' HVAC equipment room, (5) Safety related electrical equipment room, (6) Passages, I
(7) Non essential battery room, t l Amendment 11 9.4-lb.1
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23A6100AH Standard Plant - ma n
/ ' Recirculation unit for subsystem 1 consists of a (4) HVAC equipment room, .* prefilter section, a high efficient filter section, an electric heater, a cooling coil, and two SO% capacity (5) Safety-related clearicalequipment room, supply fans. .The supply fans are placed on low speed when the fans are in the smoke removal '(6) Passages, mode. 4 Two 50% capacity return exhaust fans draw air - -
t from safety related battery rooms. During smoke re. Recirculation unit for subsystems 3 consist of a ; moval mode, the fans are placed on low speed and pretlher acction, a high efficient filter sealon, an elec-the air is discharged to atmosphere, tric heater, a cooling coil, and two 50% capacity aupply fans, ne supply fans are placed on low speed 9A.123J Safety Ralated Subsystem 2 when the fans are in the smoke removal mode. Subsystem 2 specifically serves- Two 50% capacity return exhaust fans draw air from the safety related battery rooms. During smoke ' (1) Safety related battery room 2, temoval mode, the fans are placed on low speed and the air is discharged to atmosphere. (2) Essential chiller room B, L 9A.1JA Safety Evaluatloa (3) RB cooling water pump and heat exchanger room B, The essential electrical HVAC system is designed to l ensure the operability of the casential electrical equipt (4) HVAC equipment room, ment, and to limit the hydrogen concentradon to less 9 than 2% by volume in the battery rooms. A!! $ l (5) Safety-related electrical equipment room, safety related HVAC equipment and surrounding , structures are of seismic category I design and O,, (6) Passages, op:rable during loss of the offsite power supply, (7) Non essential electrical equipment rooms. The ductwork which services these safety functions is termed ESF ductwork, and is of Seismic Category I Recirculation unit for subsystem 2 consist of a design. ESF ducting is high pressure safety grade prefilter section, a high efficient filter section, an ductwork designed to withstand the maximum positive electric heater, a cooling coil, and two 50% capacity and/or negative pressure to which it can be subjected supply fans. The supply fans are placed on under normal'or abnormal conditions. Galvanized low speed when the fans are in the smoke remcr s' steel ASTM A526 or ASTM A527 is used for outdoor 7 mode, air intake and exhaust ducts. All other ducts are welded black steel ASTM A570, Grade A or Grade Two 50% capacity return exhaust fans draw air D. Ductwork and hangers are Seismic Category I. from the safety related battery rooms. During Bolted Flange and welded joints are qualified per smoke removal mode, the fans are placed on ERDA 76-21. Iow speed and the air is discharged to atmosphere. Redundant components are provided where neces-9A.I.2.3.3 Safety Related Subsystem 3 sary to ensure that a single failure will not preclude , Sdequate heat exchanger building ventilation. Subsystem 3 specifically serves: .i 9.f.1J.5 Inspection and Testing Revolvements ! (1) Safety related battery room 3, Provisions are made for periodic tests of the out-(2) Essential diller room C, door air cleanup fans and filters. These tests include detertainations of differential pressure across the (3) RB cooling water pump and heat exchanger filter and of filter efficiency. Connections for testing, room C, such as injection, sampling and monitoring are prop-Amendment 10 9.41c l
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' 23A6100AJ hndard PInnt Rm A material samples using correlation methods which '(1)- Forgings are rough machined with minimum l '. are as conservative, or more so, than those presented ~ stock allowance prior to heat treatment.
In Reference 1. (2) Each finished machined rotor is subjected to l , e Turbine operating procedures are employed to . - 200 percent volumetric (ultrasonic), and surface preclude brittle facture at startup by ensuring that visual examinations, using established metal temperatures are (a) adequately above the acceptance criteria. These criteria are more FA'!T, and (b) as defined above, sufficient to main- restrictive than those specified for Class l' t tain the fracture toughness to tangential stress ratio components la the ASME Boiler and Pressure at or above 2 /28. Vessel Code, Sections III and V, and include the
' requirement that subsurface sonic indications 10.2JJ High Temperature Properties are either removed or evaluated to ensure that they will not grow to a size which will The operating temperatures of the high pres. compromise the integrity of the unit during its sure rotors are below the stress rupture range, service life.
Therefore, creep-rupture is not considered a signifi-cant failure mechanism. (3) . All finished machined surfaces are subjected to I
- a magnetic particle test with no flaw indications Basic stress and creep rupture data are permissible, obtained in standard laboratory tests at appropriate~
temperatures with equipment and procedures (4) - Each fully bucketed turbine rotor assembly is l consistent with ASTM recommendations, spin tested at 20-percent overspeed. 10.23A Turbine Design Additional preservice inspections include air leakage tests performed to determine that the The turbine assembly is designed to withstand hydrogen cooling system is tight before hydrogen is normal conditions and anticipated transients,includ- introduced into the generator casing; The hydrogen ing those resulting in turbine trip, without loss of purity is tested in the generator after hydrogen has - structural integrity. The design of the turbine been introduced. The generator' windings and all assembly meets the following criteria: motors are megger tested, Vibration tests are ;
- performed on all motor driven equipment. Hydro-l (1) Turbine shaft bearings are designed to retain static tests are performed on all coolers. All piping is their structuralintegrity under normal pressure tested for leaks. Motor operated valves are operating loads and anticipated transients, factory leak tested and inplace tested once installed.
including those leading to turbine trips. - j l
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10.23.6 Inservice Inspection l l(2) The multitude of natural critical frequencies of the turbine shaft assemblics existing between The laservice inspection program for the turbine , zero speed and 20 percent overspeed are assembly includes the disassembly of the turbine and controlled in the design and operation so as to complete inspection of all normally inaccessible parts, L cause no distress to the unit during operation, such as couplings, coupling bolts, turbine shafts, l low pressure turbine buckets, low pressure and r p l(3) The maximum tangential stress resulting from high pressure rotors.. During plant shutdown ! ! centrifugal forces, interference fit, and thermal coinciding with the inservice inspection schedule for l gradients does not exceed 0.75 of the yield ASME Section III components, as required by the strength of the materials at 115 percent of rated ASME Boiler and Pressure Vessel Code, Section XI, l speed, turbine inspection is performed in sections during the ;' I refueling outages so that in 10 years totalinspection 10.23J Preservice Inspection has been completed at least once. i - r
- . The preservice procedures and acceptance cri. This inspection consists of visual and surface i teria are as follows
- examinations as indicated below:
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ABM / n46100AI
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' Standard Plant :n.< i i l
(1) Visual examination of all accessible surfaces (c) Deposits on stems and'other valve parts ' j of rotors which could interfere with valve operation i 1 (2) Visual and surface examination of all low. (d) Distortions, misalignment i pressure buckets - . Inspection of all valves of one type will be - (3) 100 percent visual examination of couplings conducted if any unusual condition is d scovered 1 and coupling bolts ..
'. i 10.2A Evaluation The inservice inspection of valves important to overspeed protection includes the following: = ' The turbine generator is not' nuclear safety
.. related and is not needed (o effect or support s' safe (1) All main stop valves, control valves, extrac-
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_ shutdown of the reactor, ; tion nonreturn valves, and CBIVs will be . . tested under load. Test controls installed on . The turbine is designed, constructed, and in. the main control room turbine panel and spected to minimize the possibility of any' major permit full stroking of the stop valve, control component failure, valves, and CBIVs. Valve position indication . - _. . - is provided on the panel. No load reduction . The turbine has a redundant, testable overspeed is necessary before testing main stop and. trip system to minimize the possibility of a turbine . ; control valves, and CBIVs. Extraction conte. overspeed event. turn valves are tested by equalizing air . . pressure across the air cylinder. Movement Unrestrained stored energy in the enraction ' of the valve arm is observed upon action of - steam system has been reduced to an acceptable the spring closure mechanism. _ minimum by the addition of nonreturn valves in selected extraction lines, s (2) Main stop valves, control valves, extraction .
$ nonreturn valves, and CBlVs will be tested at . The turbine generator equipment shielding re . -
1 east moath1y, and'more often if - quirements and the methods of access control for all recommended by the turbine manufacturer, areas of the turbine building ensure that the dose
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Once per month, closure of each valve during criteria specified in 10CFR20 for operating personnel ; test will be verified by observation' of the are not exceeded. valve motion. All areas in proximity to turbine generator j Tightness tests of the main stop and control equipment are zoned according to expected l valves are performed at least once per main- occupancy times and radiation levels anticipated ! tenance cycle by checking the coastdown under normal operating conditions. L characteristics of the turbine from no load - l with each set of four valves closed alternately. Specification of the various radiation zones in accordance with expected occupancy is listed in (3) All main stop valves, main control valves, and Chapter 12.' CBlVs will be inspected once during the first three refueling or extended maintenance -If deemed necessary during' unusual' shutdowns. Subsequent inspections will be occurrences, the occupancy times for certain areas- .i scheduled so that each valve is inspected at 3 will be reduced by administrative controls enacted by l to 5 year interval and at least, one valve of health physics personnel. each type is inspected after each fuel cycle or 31/3 year interval, whichever is less. The The design basis operating concentrations of inspections willbe conducted for: N 16 in the turbine cycle are indicated in Section 12.2. (a) Wear oflinkages and stem packings . The connection between the low pressure H turbine exhaust hood and the condenser is made by (b) Erosion of valve seats and stems means of a stainless steel expansion joint. l l, l Ainendment it 10.2 8 i l 4
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M 23A6100AJ Standard Plant ' na. tystem, the reactor coolant system can still be safely at the resulting static head, inspecting all tube joints, cooled down using only nuclear island systems. . accessible welds, and surfaces for visible leakage and/or excessive deflection. Each condenser water , 10.4.1J Evalention box is to receive a field hydrostatic test with alljoints _ _ and external surfaces inspected for leakage. During operatioc, radioactive steam, gases, and . i condensate are present in the shells of the main 10.4.1J lastrumentation Applientions ; condenser. The anticipated inventory of radioactive . contaminants during operation and shutdown is discussed in Sections 11.1 and 113. 10.4.1.5.1 HotwellWaterlevel 1 ! The condenser hotwell water level is measured - - Necessary shielding and controlled access for by two level transmitters. These transmitters provide ~ # the main condenser are provided (see Sections 12.1 signals to an indicator, annunciator trip units, the < and 12.3). plant computer, andthe hotwell lew1 control system. Level is controlled by two sets of modulating control Hydrogen buildup during operation is not ex- valves; Each set consists of a normal and an emer-pected to occur due to provisions for continuous gency valw. evacuation of the main condenser. During shut- . down, significant hydrogen buildup in the main One set of valves allows water to flow from the ' condenser will not occur as the main condenser will condensate storage tank to the condenser hotwell as ' then be isolated from potential sources of hydrogen, the level drops below setpoint. If the icvelincreases above another set point, the second set of valves Maia condenser tubeside circulating water is . located on the discharge of the condensate pumps, treated to limit algae growth and prevent long-term open to allow condensate to be pumped back to the corrosion of the tubes and other components. Cor- storage tank, rosion of the outside of the condenser tubing is ~ prevented by maintaining strict water quality using 10.4.1.5.2 Pressure ; g the condensate cleanup system described in Subsec-tion 10.4.6. The construction r.iaterials used for the Condenscr pressure is measured by gauges, pres- , main condenser are selected such that the potential sure switches, and electronic pressure transducersi i for corrosion by galvanic and other effects 1s- The pressure switches provide input signals to the i minimized, turbine control system and the annunciater. Two pressure transducers provide input signds to the ' The potential flooding which would result from plant computer, a recorder, and a trip unit. The trip failure of the condenser is discussed in Section 3.4. unit provides input signals to the reactor recircula-
- Section 3.4 shows failure of the condenser will not tion system and steam bypass and pressure regula-adversely affect any equipment required for safe tion systemi-In addition, four independent and re-shutdown of the reactor, dundant safety related pressure transmitters provide input signals to the nuclear steam supply system.
The loss of main condenser vacuum will cause a turbine trip and closure of the main steam isolation As condenser pressure increases above normal valves. The consequences of a turbine trip are dis. levels, an annunciator is activated. A further cussed in Subsection 15.23. Should the turbine stop, increase in pressure results in a turbine trip. As control or bypass valves fail to close on loss of con- pressure increases toward a complete loss of denser vacuum, two rupture diaphrams on each tur- vacuum, the main steam isolation valves and the ' bine exhaust hood protect the condenser and turbine turbine bypass valves are closed to prevent overpres-exhaust hoods against overpressure, surization of the condenser shell. 10.4.1.4 Tests and Inspections The approximate setpoints for these functions are as follows: Each condenser shellis to receive a field hydro-static test before initial operation. This test will (1) High condenser pressure turbine alarm at 24 consist of filling the condenser shcIl with water and, inches Hg. vacuum 3-> , Amendment 3 10.4 3 t
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f L fABWR msm# I Standard Plant hA l (2) High condenser pressure turbine trip at 22 plant power operation, and to the turbine building &
- inches Hg vaccum ventilation system exhaust at the beginni6g of each ,
l : atart up, t , (3) Bypass valve closure at 12 inches Hg vacuum !. 10.4.2.1 Design Bases L (4) Main steam isolation valve closes at,7 to 10 inches Hg vacuum 10.4.2.1.1 Safety Design Bases -
- Condenser pressure is an input to the reactor The MCES does not serve or support any safety recirculation system. Recirculation pump runback is function and has no safety design bases.
initiated upon the trip of a circulating water pump !- when condenser pressure is higher than some site 10.4.2.1.2 Power Generation Design Bases specific preset valve. Runback is automatically initiated when required to avoid a turbine trip on Power Generattan ru t n maale one . The Mcgg high condenser pressure, is designed to remove air and other power cycle non. condensable gases from the condenser during plant 1 t 10.4.1.53 Temperature . startup, cooldown, and power operation and exhaust l them to the offgass system or turbine building - g-
- Temperature is measured in each 1.P turbine ventilatioon system cahaust. p$$. t exhaust hood by pneumatic temperature controllers, The controllers modulate a control vahr in the water Power Generatlan Da* tan Baain Two .'.The MCES -
spray line protecting the exhaust hoods from over. establishes and maintains a vacuum in the condenser i L heating, during power operation by the use of steam jet air a ejectors, and by the mechanical vacuum pump during Circulating water temperatures are monitored early start.up. L upstream and downstream of each condenser tube I bundle and are fed to the plant computer and a main 10.4.2.2 Description contrni room redorder for use during periodic . ( condenser performance evaluations The condenser evacuation system is illustrated in 2
- Figure 10.41. The system consists of two 100% ca- RR' l 10.4.1.5.4 Leakage pacity, double stage, steam jet air ejector (SJAE) $$
i' .
. units (complete with intercondenser) for power plant Leakage of circulating water into the condenser operation, and a mechanical vacuum pump for use shell is monitored by the on.line instrumentation and during startup.. The last stage of the SJAE is a the process sampling system described in Subsection noncondensing stage. One SJAE unit is mormally in RR i 93.2. operation and the other is on standby. . 66 Conductivity of the condensate is continuously During the laitial phase of startup, when the monitored at selected locations in the condenser, desired rate of air and gas removal exceeds the Conduethity and sodium are continuously monitored ' capacity of the steam jet air ejectors, and nuclear at the discharge of the condensate pumps. High , steam pressure is not ailequate to operate the air -
condensate conductivity and sodium content, which ejector units, the mechanical vacuum pump estab-indicate a condenser tube leak, are individually lishes a vacuum in the main condenser and other ! alarmed in the main control room, parts of the power cycle. The discharge from the vacuum pump is then route / to the turbine building 10A.2 Main Condenser Evacuation ventilation system exhaust shee there is then little or RR System no effluent radioactivity proent. Radiation detectors OO , in the the vacuum pump disharge and plant vent Noncondensable gases are removed from the alarm in the main control room if abnormal power cycle by the main condenser evacuation radioacthity is detected (see Section 7.6). Radiation system (MCES). The MCES removes the hydrogen monitors are prodded on the main steam lines which and oxygen produced by radiolysis of water in the trip the vacuum pump if abnormal radioactivity is reactor, and other power cycle noncondensable detected in the steam being supplied to the gases, and exhausts them to the offgas system during condenser. Amendment 11 10.4-4 l f l { i
i i ' ABWR mmw Standard Plant ' am i - The steam jet air ejector is placed in service to formed prior to plant operation in accordance with - j
% remove the gases from the main condenser after a applicable codes and standards. l pressure of about 10 to 15 in Hg obsolute is estab- !
lished in the main condenser by the mechanical Components of the system are continuously mon-vacuum pump and when sufficient nuclear steam itored during operation to ensure natisfactory perfor. pressure is available. mance. Periodic inservice tests and inspections of : the evacuation systern are performed in conjunction l During normal power operation the steam jet with the scheduled maintenance outages. . air injectors are normally driven by condensed steam, with the main steam supply on automatic 10.4.2.5 - Instrumentation Applications standby. The main steam supply, hoqwever,is ag$
$ normally used during startup and low load operation, Local and remote indicating devices for such and auxiliary steam is available for normal use of the parameters as pressure, temperature, and flow .
steam jet air ejectors during early startup, should the indicators are provided as required for monitoring {_ mechanical vacuum pump prove to be smavailable. the system operation. I 10.4.2.3 Evaluation 10.4.2.5.1- Steam Jet Air FJectors The offgas from the main condenser is one Steam pressure and flow is continuously moni. l source of radioactive gasin the station. Normally it tored and controlled in the ejector steam supply -
' includes the activation gases nitrogen-16, oxygen 19, lines Redundant pressure controller sense steam and nitrogen 13, plus the radioactive noble gas pressure at the second stage inlet and modulate the parents of strontium 89, strontium 90, and steam supply control valves. upstream of the air '
ceslum.137. An inventory of radioactive contami. - ejectors. The steam flow transmitters provide inputs pg - 1 nants in the effluent from the steam jet air ejectors is to logic devices. These logic devices provide for - evaluated in Section 11.3. isolating the offgas flow from the air injector unit on - I k a two out of.three logic, should the steam flow drop . Steam supply to the second stage ejector is below acceptable limits for offgas stream dilution. . l maintained at a minimum specified flow to ensure l adequate dilution of hydrogen and prevent the offgas 10.4.2.5.2 MechanicalVacuum Pump from reaching the flammable limit of hydrogen. ? Pressure is measured on the suction lir.c of the h mechanical vacuum pump by a pressure switch. Upon reaching a preset vacuum, the pressure switch The MCES has no safety related function as energizes a solenoid valve which allows additional discussed in Section 3.2. Failure of the system will seal water to be pumped to the vacuum pumps. Seel ' not compromise any safety-related system or compo- pump discharge pressure is locally monitored. Seal nent and will not prevent safe reactor shutdoivn. water cooler discharge temperature is measured by a temperature indicating switch. On high temperature, Should the system fail completely, a gradual the switch activates an annuciator in the main control reduction in condenser vacuum would result from room. The vacuum pump discharge stream is the buildup of noncondensable gases. This reduction monitored for radiation proir to entering the turbine MR in vacuum would first cause a lowering of turbine ventilation system exhaust. The vacuum pump is @$ cycle efficiency due to the increase in turbine tripped and its discharge valve is closed upon exhausts pressurc. If the MCES remained receiving a main steam high high radiation signal. RR inoperable, condenser pressure would then reach the
$$ turbine trip set point and a turbine trip would result. 10.4.3 Turbine Gland Seal System The loss of condenser vacuum incident is discussed in Subsection 15.2.5. The turbine gland seal system (TGSS) prevents )
the escape of radioactive steam from the turbine
- 10.4.2.4 Tests and inspections shaft / casing penetrations and valve stems and prevents air inleakage through subatmospherie O Testing and inspection of the system is per turbine glands.
l Amendment li 10A-5
ABWR muw i Standard Plant Rm A
.10.43.1 Design Bases . supplied from the main steam line or auxiliary steam header, Above approximately 50% load, however, 10.43.1.1 Safety Design Bases scaling steam is normally provided from the heater RR i
drain tankvent header.- At allloads, gland scaling hh The TGS$ does not serve or support any safety can be achieved using auxiliary steam so that plant ; !- function and has no safety design bases. - power operation can be maintained without ' j sppre.clable radioactivity releases even if highly : i 10.4J.1.2 Power Generstloa Design Bases abnormal levels of radioactive contaminants are : [ .
= present in the process steam, due to unanticipated
- Power Generatinn Detten statie One . The TGSS is - fuel failure in the reactor,'
!' designed to prevent atmospheric air leakage into the I turbine casings and to prevent radioactive steam leakage out of the casings of the turbine generator. The outer portion of all glands of the turbine and - main steam valves are connected to the gland steam condenser which is maintained at a slight vacuum by . 3 Power Generatinn Dealen Ba In Two . The TGSS . the exhauster blower.1 During plant . operation, the , returns the condensed steam to the condenser and - gland steam condenser and one of the two installed ! g exhausts the noncondensable gases, via the turbine; 100%' capacity motor driven blowers are in a m' i g building ventilation system, to the plant vent, operation. The exhauster blower discharges gland $$1 -! air inleakage to the turbine building ventilation Power Generation De<tyn Basis Three . The TGSS. . system exhaust.' The gland strc.3 condenser is L l [ has enough capacity to handle steam and air flows . cooled by main condensate flow, resulting itom twice the normal packing clearances. .. 10.433 Evaluation i 10.43.2 Descriptinn .
,. The turbine gland seal system is designed to , bg g..
gg- ! 10.43.2.1 General Descripilon prevent leakage of radioactive steam from the main l turbine shaft glands and the valve stems. The -3 The turbine gland scaling system is illustrated in high pressure turbine shaft seals must accommodate [ Figure 10.4 2. The it:rbine gland seal system consists - a range of turbine shell pressure from full vacuum to;
- of a scaling steam pressure regulator, sealing steam approximately 220 psia. The low. pressure turbine header, a gland steam condenser, with two full. ' shaft seals operate against a vacuum at all times.
capacity exhauster blowers, and the associated The gland seal outer portion steam air mixture is piping, valves and instrumentation. exhausted to the gland steam condenser via the seal vent annulus (i.e., end glands) which is maintained at - 10.43.2.2 System Operation a slight vacuum. The radioactive content of the'- sealing steam which eventually exhausts to the plant ; l The annular space through which the turbine vent and the atmosphere is evaluated in Section 113 pg ; shaft penetrates the casing is scaled by steam and makes a negligible contribution to overall plant supplied to the shaft seals. Where the gland seals - radiation release. In addition, the auxiliary steam lgg , operate against positive pressure, the scaling steam . system is designed to provide a 100% backup to the ; acts as a buffer and flows either inwards for collec. normal gland seal process steam supply. A full ca- t tion at an intermediate leskoff point, or, outwards pacity gland steam condenser is provided, and and into the vent annulus. Where the gland seals equipped with two 100% capacity blowers, operate against vacuum, the scaling steam either is drawn into the casing or leaks outward to a vent Relief valves on the seal steam header prevent t '~ annulus. At all gland seals, the vent annulus is excessive seal steam pressure. The valves discharge maintained at a slight vacuum and also receives air in to the condenser shell. leakage from the outside. From each vent annulus, the air steam mixture is drawn to the gland steam 10.43.4 Tests and Inspections condenser. Testing and inspection of the system will be per- ; The seal steam header pressure is regulated formed prior to plant operation. Components of the l Mm automatically by a pressure controller. During system are continuously monitored during operation
$$ startup and low load operation, the seal steam is Amendment 11 10A 6 i
d ABWR utnooxi . Standard Plant Rm A 10.4.6.1 Design Bases polisher vessel to preclude gross resin leakage into
-( the power cycle in case of vessel underdrain failure,'- :
10,4.6.1.1_ Safety Design Bases and to catch resin fine leakage as much as possibic, j The design bases influent concentrations are pro. The CCS does not serve or support any safety vided in Table 10.4 5 Based on the influent concen- ; function and has no safety design bases, trations the condensate polisher effluent water , quality is as reported in Table 10.4 6._ The CCS i 10.4.6.1.2_ Power Generation Design Bases ' components are located in the turbine building. Power Generation Dealen Bath One . The CCS Provisions are included' to permit mechanical continuously removes dissolved and suspended solids ultrasonic washing and replacement of the ion ex. from the condensate to maintain reactor feedwater change resin. Each of the polisher vessels has fait quality, open inlet and outlet isolation valves which are ; remotely controlled from the local polisher control Power Generation Detten Bs.=h Two . The CCS panel, removes corrosion products from the condensate and from drains returned to the condenser hotwell so as A system flow bypass valve is also provided which to limit any accumulation of corrosion products'in is manually controlled from the main control room. _ the cycle. Pressure downstream of the polisher system is indi- i cated and low pressure is alarmed in the main ' t i Power Generation Decian Banla Three . The CCS control room to alert the operator. The bypass is - removes impurities entering the power cycle due to used only in emergency and for short periods of time condenser circulating water leaks as required to until the polisher system flow is returned to normal permit continued power operation within specified or the plant is brought to an orderly shutdown,-To water quality limits as long as such condenser leaks prevent unpolished condensate from leaking through are too small to be readily located and repaired. the bypass, double isolation valves are provided with
# Power Generation Desien Basis Four . The CCS i
limits the entry of dissolved solids into the feedwater 10.4.6.2.2 Component Description . system in the event oflarge condenser leaks, such as a tube break, to permit a reasonable amount of time Codes and standards applicable to the CCS are i for orderly plant shutdown, listed in Subsection 3.2.2. The system is designed > l and constructed in accordance w'th quality gro'ai, D Power Generation Desien Basis Five . The CCS requirements. Design data for major compor.ents of injects in the condensate such water treatment the CCS are listed in Table 10.4-4 additives as oxygen and hydrogen as required to minimize corrosion /crosion product releases in the . Condensate Polishers Vessels . There are six power cycle. 20-percent capacity polisher vessels (one on standby) - each constructed of carbon steel and lined with Power Generation Deslan Basis Six . The CCS - natural rubber. Normal operation fullload steady maintains the condensate storage tank water quality state design flowrate is 40 gpm per square foot of ns required for condensate makeup and miscella. bed. Mannum flowrates are 50 and 60 r,pm per L neous condensate supply services, square foot for steady state and transient operation I respect!vely. The nominal bed depth is 40 inches. l 10.4.6.2 System Description 10.4.6.2.3 System Operation . 10.4.6.2.1 General Description The CCS is continuously operated, as necessary i The condensate cleanup system is illustrated in to maintain feedwater purity levels. Figure 10.4-4. The CCS consists of six bead resin, i-mixed bed ion exchange polisher vessels arranged in Full condensate flow is passed through five of the g parallel with, normally five in operation and 1 in six polisher vessels, which are piped in parallel. The , standby. A strainer is installed downstream of each sixth polisher is on standby or is in the process of . Amendment 3 10A 11 i
. . _. ~ . _ . _ _ ____ _ .. _ _. _ . _ _ _ ___ _ _ _ ___ __. . _ _ _ _
i 1 1 23A6100AJ
- Etandard Plant ,, Rn A belag cleaned, emptied or refilled. 'The service run 10.443 Evaluation i for each polisher vessel is terminated by either high differential pressure across the vessel or high con. The CCS does not stut at support any safety J ducsivity or sodium content in the polisher effluent function and has no safety ties]p basis. l I. water. Alarms for each of these parameters are '
provided on the localcontrolpanel. ; The condensate cleanut rystem removes some 1
. radioactive material, acpvinc eotrosion products :
The local control panel is equipped with the and fission products that a to:taried over from the' l
- appropriate instruments and controls to allow the reactor, While these rad)ttedve sources do not l operators to perform the following operations
- affect the capacity of the res.ita triconcentration of 4 such radioactive material 3eides shielding (see .
(1) Remove an' exhausted polisher frocq service and Chapter 12), Vent gases and pierwastes from the replace it with a standby unit condensate cleanup systen:w coDected in con-
. trolled areas and sent to the istNtste system for -
(2) Transfer the resin inventory of any polisher treatment and/or. disposal. Ch*pter U describes. . t vesselinto the resin receiver tank for mechmairat - the activity level and removal of radioxtive material cleaning or disposal, from the condensate tystem. i (3) Process the as received resin through the . The condensate cleanup system complies with- t ' t ultrasonic resin cleaner as it is transfered from- . Regulatory Guide 1.56, Maintenance of. Water Purity ~ tae receivet tank to the storage tank. In Bolling WaterReactors.
, (4). Transfer the resin storage tank resins to any . The condensate cleanup system and related - 1
, polisher vessel. support facilities are located in non safety related l' .. buildingsE As a result, potential equipment or piping - (5) Transfer exhausted resin from the receiver tank - failures can not affect plant safety. , to the radwaste system. ,. 10.4.6.4. Tests and Inspections On termination of a service run, the exhausted . . 4 polisher vesselis taken out of service, and the Preoperational tests are performed on the con. standby unit is placed in service by remote manual ~ densate cleanup system to ensure operability, reli- , operation from the local control panel. The resin ability, and integrity of the system. Each polisher: i from the exhausted vesselis transferred to the resin vessel and system support equipment can be isolated receiver tank and replaced by a clean resin bed that during normal plant operation to permit testing and is transferred from the resin storage tank. A final maintenance.- rinse of the new bed is performed in the polisher by condensate. full flow recycle to the condenser before 10.4.6J . Instrumentation Applications - l8N it is placed in service. The rinse is monitored by con- ' ductivity analyzers, and the process is terminated Conductivity elements are provided for the-when the required minimum rinse has been system influent and for each polisher vessel effluent. completed and normal clean bed conductivity is System influent conductivity detects condenser j obtained, leakage; whereas, polisher effluent conductivities provide indication of resin exhaustion. The polisher ! Through periodic condensate makeup and re- effluent conductivity elements also monitor the .
- ject, the condensate storage tank water inventoryis quality of the condensate that is recycled to the 1
processed through the CCS and tank water' quality is condenser after processing through a standby vessel maintained as required for condensate makeup to before it is returned to service. Differential pressure the cycle and miscellaneous condensate supply is monitored across each polisher vessel and each services. The diagram of the condensate storage and vessel discharge resin strainer to detect blockage of ' transfer system is illustrated in Figure 10.4-5. flow. The flow through each polisher is monitored and used as controlinput to assure even distribution l The condensate cleanup and related support of condensate flow through all operating vessels and , systems wastes are processed by the radwaste system by conelation with the vessel pressure drop, to i ! as described in Chapter 11. l i l Amendnwnt !! 10.4-12 b' 1
- l
_ _ . . _ _ . _ _ - _ _ _ _ . _ _ - - - _ _ _ _ _ _ _ ~ ___ - _ -- 4 Mkb 23A6100AJ Standard Plant Rn A permit evaluation of the vessel throughput capacity, designed to permit continuous long term full power Individual vessel effluent conductivity, differential plant operation with the following equipment out of pressure, and flow measurements are recorded at the service: one feedwater pump, one condensate pump, system local control panel. A multipoint annunciator one heater drain pump or, one high pressure heater is included b the local panel to alarm abnormal string with a slightly reduced final feedwater conditions within the system. The local panelis con- temperature, nected to the main control room where local alarms annunicated by a global system alarm but can also be Power Generation Design Basis Four . The CFS is y displayed individually if requested by the operators. designed to permit continuous long term operation g_ with one LP heater string out of service at the Other system instrumentation includes turbidity maximum load permitted by the turbine and other water quality measurements as necessary manufacturer, approximately 85%, value which is set for proper operation of the polisher and miscella. by steam flow limitation on the affected LP turbine, neous support services, and timers for automatic supervision of the resin transfer and cleaning cycles. Power Generation Deslon Basis Five . The CFS is l The control system prevents the initiation of any designed to heat up the reactor feedwater to a operation or sequence of operations which would nominal temperature of 420 F during fullload conflict with any operation or sequence already la operation and to lower temperatures during part progress whether such operation is under automatic load operation, or manual control. Power Generation Detlen Batie Sir . The CFS is designed to minimize the ingress or release of impu-10A,7 Condensate and Feedwater System rities to the reactor feedwater. The function of the conden:. ate and feedwater 10.4.7.t Descriptian system (CFS)is to receive condensate from the con. denser hotwells, supply condensate to the cleanup 10.4.7.2.1 General Description O i i system, and deliver high purity feedwater to the
-V reactor, at the required flow rate, pressure and tem- The condensate and feedwater system is illus-perature, trated in Figure 10.4 6 and 10.4 7. The condensate and feedwater system consists of the piping, valves, 10.4.7.1 Design Bases pumps, heat exchangers, controls and instrumenta-tion, and the associated equipment and subsystems 10.4.7.1.1 Safety Design Bases which supply the reactor with heated feedwater in a closed steam cycle utilizing regenerative feedwater The condensate.feedwater system does not heating. The system described in this subsection serve or support any safety function and has no extends from the main condenser outlet to the safety design bases. second Isolation valve outside of containment. The remainder of the system, extending from the second 10.4.7.1.2 Power Generation Design Bases isolation valve to the reactor,is described in Chapter
- 5. Turbine extraction steam is utilized for a total of Power Generation Desien Basis One - The CFS is six stages of closed feedwater heating. The drains ,
designed to provide a continuous and dependable from each stage of the low pressure feedwater feedwater supply to the reactor at the required flow heaters are cascaded through successively lower rate, pressure, and temperature under all anticipated pressure feedwater heaters to the main condenser, steady state and transient conditions. The high pressure heater drains are pumped back. ward to the reactor feedwater pumps suction. The Power Generation Desien Basis Two The CFS is cycle extraction steam, drains and vents systems are designed to supply up to 115% of the rated feedwa- illustrated in Figures 10.4-8 and 10.4 9. ter flow demand during steady state power operation . and for at least 10 seconds after generator step load - The CFS consists of four 33% capacity reduction or turbine trip, and up to 75% of the rated condensate pumps, three normally operating 33 50% F flow demand thereafter. capacity reactor feedwater pumps (three normally $ operating and one on standby), four stages of O) (, Power Generation Desian Basis Three - The CFS is low pressure feedwater heaters, and two stages of Amendment 11 10A.13 I
i ABWR msiom ; Standard Plant am i of high pressure feedwater heaters, piping, valves, by eith er the condensate pumps or the reactor feed and instrumentation. The condensate pumps take . pumps operating at their minimum fixed speed. E I suction from the condenser hotwell and discharge During power operation, the heater bypass function $ , , the deserated condensate into one common header is to maintain full feedwater flow capability when a ! which feeds the condensate filters and high pressure heater string must be isolated for L demineralizers. Downstream of the condensate maintenance. demineralizers, the condensate is taken by a single header and flows in parallel through five auxiliary Diiring power operation, the conJensate is well
- condenser /coolerr., (one gland steam exhauster con. deaerated in the condenser and continuous oxygen ,
3~ denser and two sets of steam jet air ejector injection is used to maintain the level of oxygen condensers and offges recombiner condenser content in the final feedwater as shown in Subacetion (coolers). The condensate then branches into three 10.4.6.
- parallel strings oflow pressure feedwater heaters. -
Each string contains four stages of low pressure To minimize corrosion product input to the-feedwater heaters. The strings join together at a. reactor during startup, recirculation lines to the-common header which is routed to the suction of the condenser are provided from the reactor feedwater reactor feedwater pumps, pump suction header and from the high pressure. ! feedwater heater outlet header. Another input to the feedwater flow consists of , . the drains which are pumped backward and injected . Prior to plant startup, cleanup is accomplished by . Into the feedwater stream at a point between the allowing the system to recirculate through the con. fourth stage low pressure feedwater heaters and the - densate polishers for treatment prior to feeding any . . suction side of the reactor feed pumps. These water to the reactor during startup. , drains, which originate from the crossaround steam moisture sep.irators and reheaters and from the two j l sets of high pressure feedwater heaters, are directed- ; i to the heater drain tank. The reheater and top .I heater drains are deserated in the'crossaround ,~ i j q. l
=
heaters so that, after mixing with condensate, the drains are compatible with the reactor feedwater - ,
- quality requirements for oxygen content during g
! normal power operation. The heater drain pumps g take suction from their heater drain tank and inject , I - the deaerated drains into the feedwater stream on ; the suction side of the reactor feed pumps. 4 The reactor feedwater pumps discharge the feedwater into two parallel high pressure feedwater , heater strings, each with two stages of high pressure '= feedwater heaters. Downstream of the high pressure 10.4.7.2.2 Component Description
. feedwater heaters, the two strings are then jo'ned into a common header, which divides into two feed- All components of the condensate and feedwater water lines which connect to the reactor, system that contain the system pressure are designed and constructed in accordance with applicable codes A bypass is provided around the reactor feedwa. as referenced in Section 3.2.
l ter pumps to permit supplying feedwater to the reactor during early startup without operating the Condensate Pumns . The four condensate feedwater pumps, using only the condensate pump pumps are identical, fixed speed motor driven g head, pumps, three are normally operated, and the fourth g( is on standby. Valving is provided to allow individual i Another bypass, equiped with a feedwater flow pumps to be removed from service. t control valve,is provided around the high pressure I heaters to perform two independent functions. A minimum flow recirculation line is provided I During startup, the bypass and its flow control valve )
- j. are used to regulate the flow of feedwater supplied Amendment 11 10A 14 I
1
l I ABWR - moom Standard Plant m.v. i l , downstream of the auxiliary condensers for conden- drein line is also used during startup and shutdown j este pump protection and for auxiliary condenser when it is desirable to dump the dra*us S for feedwater , ,_ sninimum flow requlreseats, quality purposes. j Law.nrenure Feedwater Heaters . Thrce The drala tanks and tank drain lines are designed l parallel and independent strings of four. closed to maintain the drato pumps available suction head . feed vater heaters are provided, and one string is la esceu of the pump esquired alaisum under all J installed la cach ecedenser acek. The heaters have asticipaled operating eosditions including, 1 Integral drain coolers, and their (rains are cascaded particularly, lead reduction transients. This is ; to the next lower stage besters of the same string achieved mainly by providing a large elevation except for the lowest pressure heaters which drain to difference between tanks and pumps (approximately 7 l the main condensers. The henter shells are either 30 feet) and optimising the drain lines which would carbon steel or low alloy ferritic aseel, and the tubes affect the drsin system transient response, part. - are stainless steel. Each low preuvre feedwater icularly, the drain pump suction line, heater string has an upstream and downstream isole-I tior, valve which closes on detection of high level in Heater Drain Pumps . Tv tuotor driven heater any one of the low preuure heaters in the string. drsin pumps operate in parallel, each taking suction from a heater drain tank and discharging into the Hinh.ereuure Feedwater Heaters . Two suction side of the teactor feedwater pumps. The l parallel and independent strings of two high. pressure drain system des!gn allows each heater drain pump feedwater heaters are located in the high pressure to be individually temoved from service for end of the turbine building. The No. 6 heaters, maintenance while the balance of the system remains ; which have Integral drain coolers, are drained to the in operation while the affected string drains dump to No.5 heaters. The No.5 heaters, which are con. the condenser. . R densing only, drain to their respective heater drain 8 ,
~
$ tanks. The heate she:Is are carbon steel, and the
, Controlled drain recirculation is provided from O tubes are stainless s'tcl. the discharge side of each heater drain pump to the O Hester string isolation and bypan valves are provided to allow each string of high pressure associated heater drain tank. This ensures that the minimum safe flow through each heater drain pump is maintained during operation.
i R hczters to be removed from service, thus, slightly l $ reducing final feedwater tempersture but requiring 2eactor Feedwater Puts N . Three identical and , l no reduction in plant output. The heater string indepenent,33 60% capMy reactor feed pumps isolation and bypass vahts are actuated on detection (RFPs) are provided. The three pumps normally-of high levelin either of the two high pressure operate in parallcl and discharge to the high pres-heaters in the string, sure feedwater heaters. The pumps take suction downstream of the last stage low preuvre feedwater The startup and operating vents from the steam heaters and discharge through the high. pressure side of tl.c feedwater heaters are piped to the main feedwater heaters. Each pump is driven by an ad. condenser except for the highest preuvre heater justable speed synchronous motor, operating vents which discharge to the cold reheat lines. Discharges from shell relief valves on the Isolation valves are provided which allow each steam side of the feedwater heaters are piped to the reactor feed pump to be individually removed from main condenser, service for maintenance, while the plant continues operation at full power on the two remaining pumps. Heater Drain Tankt . Two heater drain tanks , 3 are provided. Drain tank levelis maintiined by the Controlled feedwater recirculation is provided '
$ heater drain pump and control valves in the drain from the discharge side of each reactor feed pump to pump disharge and redrculation line. the main condenser . This provision ensures that the minimum safe flow through each reactor feed pump Each heater drain tank is provided with an alter. is maintained duringoperation.
l , nate drain line to the main condenser for sutomatic dumping upon detection of high level. The alternate hendment il 10.4 15
. . . - - - -,,s. - - -
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ABWR m sto w l Standard Plant am i i 10.4.7J.3 Syrtem Operation possible by use of the multistring arrangement and l the provisions for isolating and bypassing equipment NORMAL OPE 3tATION . Under normal op. and sections of the system. orating conditions, system operation is automatic. 4 Automatic level control systems control the levels in The majority of the condenaste and feedwater ] < all feedwater heaters, the heater drein tanks, and the piping considered in this section is located within the condenser hotwells. Feedwater heater levels are non safety related turbine building. The portion controlled by modulating drain valves. Contrel which connects to the second valve outside the con. l valves in the discharge and recirculation lines of the tainment is located in the attam tunnel between the i heater drain pumps control the level in the heater turbine and reactor buildings. This portion of the j drein tank 6. Valves in the makeup line to the piping is analysed for dynamic effect from postulated 1 condenser from the condensate storage tank and in seismic events and safety relief valve discharges. The ] the return line to the condensate storage tank control entire condensate and feedwater system piping is f, l the levelin the condenser hotwells, analyzed for water hammet loads that could j potentially result from anticipated flow transients, i
, During power operation feedwater flow is automatically controlled by the reactor feedwater 10.4.7.4 Tests and laspections ,
g ' pump speed that is set by the feedpump speed control system. The control system utilizes 10.4.7.4.1 Preservlee Testing - , measurements of steam flow, feedwater flow, tad reactor level to regulate the feedwater pump speed. Each feedwater heater and condent. ate pump re. ; During startup, feedwate' r flow is outomstically ceives a shop hydrostatic test which is performed in i regulated by the high pressure heater bypass flow accordance with applicable codes. All tube joints of .j control valve. feedwater heaters are shop leak tested. Prior to initial operation, the completed condensate and feed. Ten. percent step load and 5 percent per minute water system receives a fic!d hydrostatic and perfor. i ramp changes can be accommodated without major mance test and inspection in accordance with the i effect in the CFS. The system is capable of accepting applicable code. Periodic tests and inspections of the l a full generator load rejection without reducing system are performed in conjunction with scheduled feedwater flow rate. maintenance outages. 10.4.7.3 Evaluation 10.4.7.4.2 Inservice Inspections The condensate and feedwater system does not The performance status, leaktightness, and , serve or support any safety function. Systems structural leaktight integrity of all system analysis show that failure of this system cannot components are demonstrated by continuous ; compromise any safety.related systems or prevent operation. ; safe shutdown. 1 10.4.7J Instrumentation Applications During operation, radioactive steam and con. densate are present in the feedwater heating portion Fecdwater flow.controllastrumentation men. I of the system, which includes the extraction steam sures the feedwater discharge flow rate from each L piping, feedwater heater shells, heater drain piping, and heater vent piping. Shielding and access control reactor feed pump and the heater bypass startup flow control valve. These feedwater system flow l
)
l are provided as necessary (see Chapter 12). The measurements are used by the feedwater control condensate and feedwater system is designed to min. system to regulate the feedwater flow to the reactor
- imize leakage with welded construction utilized to meet system demands. The feedwater control i where practicable. Relief discharges and operating system is described in Subection 7.7.1.4 l vents are channeled through closed systems. i Pump flow is measured on the pump inlet line l If it is necessary to remove a process nent from and flow controls provide automatic pump recircula.
service such as a feedwater heater, pump, or control tion flow for each reactor feedwater pump. Auto. 4 valve, continued operation of the system is matic controls also regulate the condensate flow i Amendment 11 10.4 16 O1 l 1 i 9
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ABWR mum, , Standard Plant m.m i i rq R through the auxillary condemers (off-gas recombiner ; ('j $ condenser / coolers, gland steam condenser, and > steam jet air ejector condcasers) and maintaink sondensate pump minimum flow. Measurements oi pump suction and discharge pressures are provided for all pumps in the system. The high pressure feedwater beater isolation valves are interlocked such that if a string of henters i were to be removed from senice the extraction non. return valves and/or isolation valves for those p heaters would automatically close and the heater ,
$ - string bypass valve open. The low pressure feedwa.
ter heater isolation valve 4 are interlocked such that. if a string of heaters were removed from setsice, the , extractions to the affected heaters which are equiped 3 with nonreturn vahts would automatically close. Sampling means are presid:d for monitoring the , quality of the condensate and final feedwater, as described in Subsection 9.3.2. Temperature mea. , surements are provided for each stage of feedwater l heating. Steam pressure measurements are prosided at each feedwater heater. Levelinstrumentation and controls are provided .'or automatically regulating the heater draln flow rate to maintain the proper levelin each feedwater heater shell or heater drain O tank. High level control valves provide automatic dump.to. condenser of heater drains on detection of high levelin the heater shell. The total water volume in the condensate and feedwater system is maintained through automatic makeup and rejection of condensate to the conden. sate storage tank. The system rnakeup and rejection are controlled by the condenser hotwelllevel ' l controllers. 10.4.8 Steart Generator Blowdown System (PWR) Not applicable to ABWR, l , 10.4.9. Auxiliary Feedwater System (PWR) Not applicable to ABWR. I O Amendment 11 10.4 17
l ABWR m .io m i Standard Plant m.,_ A O TABLE 10.41
. CONDENSER DESIGN DATA * !
. I ham Condenser Type Transversal,3 shells, Demerating Design duty, Btu /hr. tot:13 shells 8.7 x 10' ,
Shell pressures w/80F Circ. water, inches Hga 2.0/2.7/3.5 Circulating uter flow rate, gpm 600,000 - Tubeside ternp. rise. tota 13 shells. F 30 Shell design pressure range, psla 0 to 30 Hotwell storage capacity total 3 shells, gallon 100,000 R Channel design prc sure range, pia 0 to 85 ' 6 Surface Area, sq. ft. 1 x 10 Number of tube passes per shell 1 Applicable codes and standards ASME Sect. Vill, Div.1, ANSI Standards, HEl l Standards for Steam Surface l Condensers i l
- Condenser surface andperfonnanceparorneters are site dependent.
l'ohnes gouted above arefor reference purposes only, i Amendment il 10418 l
, MN . nA6100AJ l Standard Plant m.< i ! l I &I W l TABLE 10.4 2 j MAIN CONDENSER,WACUATION SYSTEM ! i see = Ja er rJener mAE) spee. 1 i I l - Number of ejector stages 2 J R 1 Number oflatercondenser 2 4 g i Number of ejedor sets and caps:lty ia 100%
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i Required supply steam pressure, psig 120 Normalsteam supply sourcc Crossaround i i i start up Vacuum Pump System Number of pumps and capacity 1 x 100% , i l
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- ster Piping Design (VWO) flowrate,Ib/hr =18,000,000 Number oflines 2 i Nominal size,in. 26 Fluid velocity, fps =20 ,
Fluid temperature, of 435 f Design code ANSI B31.1 [ Seismic design Analyzed for OBE and SSE
- designloads
- Based on SWOfeedwaterflow and one heater drain pump out ofservice.
O Amtndment 11 10.4-24
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COMPONENT FAILORE ANALYSIS f,anggasat Fallum Erect om halm Fallum Effect am Evataan Fallum Effect an ItCS Condensate None. Condenser hot. Operation continues at full None pump t.sils are interconnected. - capacity, using parallel pumps (condensate pump runout capacityis $0%). No. 3,4, $ or One trala of No. 3,4,5 Operation continues at te- None. No. 2 feedwater 6 feedwater and 6 feedwater heaters duced capacity, using parallel heater is designed to main. heater is shut down. Remaining feedwater heaters. Lead tain normal outlet feedwater trains continue to operate, must not exceed $$% to temperature under this protect the turbines from condition, escessh'e exhaust flow. Heater drain Drains from affected 50% of the high pressure Reactor control system , tank feedwater heater are cycle drains are dumped to seduces reactor power to > dumped to condenser, condenser, compensate for reduced feedwater temperature. Heater drain None. Parallel pump with $0% of HP feedwater heater Reactor control system pump condensate pumps have drains are dumped to reduces reactor power to sufficient capacity to cotidenser. compensate for reduced handle fullload, feedwater temperature. Resetor None. Two parallel trains Operations may continue at None feedwater pump are interconnected. full capacity, using 2 parallel pumps. Each reactor feedwater pump runout capacity is 50% No.1 or 2 One train is shut down. CF3 operation continues at Reactor control system feedwater capacity, using parallel reduces rcactor and gener-heater train and bypass line, ator output power to comp-ensate for reduced feedWDier temperature. O Amendment 3 10A 25
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MN 21A6100AM f Standard Plant nvn 13.3 EMERGENCY PIANNING f Emergency planning is not within the scope of , the ABWR design. However, there are design features, facilities, funelloits, and equipment necessary for emergency planning that must be considered in the design bases of a standard plant. Table 13.31 is a summary of the ABWR design ' considerations pertaining to emergency planning. l J F I t O - k e l ! i i O Amendment 11 13.3 1
L l ABWR mumu Standard Plant mm._ n Table 13.31 ARWR DESIGN CONSIDERATIONS FOR EMERGENCY PIANNING REQUIREMEhTS Primary Doeoment/ DnerprocyFlamelag ABWR facility Seglia Reantramienta llanten Canaidamtlan TechnicalSupport NUREG 0696/ The TSCis as cashe The ABWR Standard Center (TSC) 1.3.1 facilitylocated close Plant will comply to the control room with allthe TSC that shall provide design requirements, plant management and Specifically, n'TSC technical support to of sufficient slae to the reactor operating support 20 peopleis personnellocatedin located in the the control room dur. service building Ing emergency cond- a(acent to the ions,it shallhave controlbuilding. technical data displap The necessary fac. and plant records ilities and equip. avaliable to assist ment called for in in the detailed ans. Section 2 of lysis and diagnosis of NUREG 06%. abnormal plant condit-lons and any significant release of radioacthity to the environment. The TSC shallbe the primary communications center for the plant during an emergency. A senior official. designated by the licensee, shall use the resources of the TSC to assist the control toom operators by handl. Ing the administrative items, techal:al evaluat. lons, and contact with offsite activities, reliev-ing them of these functions. The TSC facilities may also be used for performing normal fur.ctions, such as shift iechnical supervisor and plant oiperations/ maintenance analysis functions, as well as for emergencies. Amendment 11 13.3 2 O
ABWR ==w Standard Plani am a Table 13.31 ABWR DESIGN CONSIDERATIONS FOR EMERGENCY PLANNING REQUIREMENTS (Continued) Primary Document / Emergency Pleaning ARWR f.Ellity Endina Requirements Innalen Canaldermilan OperatinalSupport NUREG.0696/ The OSCis an onsite The OSCis not within t Center 13.2 auembly area separate the acope of the ABWR (OSC) from the control toosa Standard Plant. An ,
- and the TSC where applicant referencing licensee operations the ABWR desigis re. ;
support personnel sponsible for ident. ; report in an emergency, ifying the OSC and ! There is direct commun. the communication ', ications between the laterfaces for OSC and the control loclusion in the room and between the detailed design of : OSC and the TSC so the control room and 1 that the personnel TSC. The detalled , reporting to the OSC requirements are - can be assigned to - provided in Section 3 , dutiesin support of of NUREG.0696, emergency operations. ; Emergency Oper. NUREG.0694/ The EOFis a offsite The EOFis not within
,O ations Facihty (EOF) 133 support facility for the management of over the scope of the ABWR Standard Plant, it i
.alllicensee emergency is the responsibility response, coordination of the applicant i of radiological uJ , referencing the AB%%
l emironmental nuess. toidentifyhis EOF j ments and determination and the communication j of recommended public laterfaces for inclu. i protective actions. sionin the detalled The EOF has appropriate design of the TSC and
- technical data displays control room. The and plant records to ' detailed requirements assist in the dagnosis are provided in '
of plant conditions to Section 4 of evaluate the potential NUREG 06%. . or actual release of radioactive materials to . the emirnomnet. A senior i licensee officialin the EOF organizes and mamages licensee offsite resources to support the TSC and the control room operators. assembly area seperate from the 7 V Amendment 11 13,3 3 ! i
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-- . r + ,2a _ - . . 2 - _
ABWR mamu ' Standard Plant _ __ _ am m l Table 13,31 ABWR DESIGN CONSIDERATIONS FOR EMERGENCY P1ANNING REQUIREMENTS (Continued) hf f i Primary i Document / Dnergency Pleanlag ABWR l M Etglga Requimments Danign Canalderatinn control room and the ! TSC, shallbe prmided l for operations support personael to report in l an emergency. '!)ere shallbe dired comm. 4 ucations br* ween the , OSC and contiof ecom l and between the OSC , and the TSC to that ! I ' l the personnel report. ing to the OSC can be asigned to duties in : emergeacy operations. l Emergency Oper. NUREG 0654/ Each licensee shall Not within the scope l ations Center. !!.11.6 make provision to of the ABWR Standard (EOC) acquire data from or Plant. Houver, no i fot emergency access impact on ABWR ! to offsite monitoring design. equipment including l (. geophysical phenomena and radiological
' monitors.
Laboratory Facil. NUREG 0654/ Provisions to acquire Resposibilityof app- 1 itles, Fixed or ll.H.6.c data from or for emer. refering ABWR design. Mobile gency access to off. ABWR design allows site monitoring and applicant to meet - analysis equipment for this requirement, laboratory facilities, ; fixed or mobile. Post Accident NUREG 0737/ Post accident sampl- Post accident sampl. , Sampling System 11.B3 ing capability ing system of Sub- ; i section 93.2 meets ; tequirements except e as descirbedin i Section 1A.2.7 for i l the upper limit of ncthityin the samples at the time they are taken. Amendment 11 13.3 4
ABWR uumu i Standard Plant m,r n l Table 13.31 l ABWR DESIGN CONSIDERATIONS FOR EMERGENCY PLANNING ! REQUIREMENTS (Continued) Primary i Document / Emergency Planelag ABWR i Easilhr indon plenuirementa penipa Canalderation Onsite Decontam. 10CTR$0 Provisions shallbe Decontamination of ; ination Tacility Appendix E/ made and described of onsite indidduals ! IV.E.3 for facilities and will be provided by supplies at the alte the applicant refer. for decontamination - encing the ABWR of onsite it.dividuals. designle the 6erdec ; building adjacent to the main change rooms (See Figure , 1.2 20). ; i O t
?
l l h i O' Amendment 11 13.3 3 ; I t
l ABWR swuom 1 a..a..a pi... mv i l O section 14.2 ! CONTENTS (Continued) ) Section 2 kit East : I 14.2.12.137 Containment Isolation Valve L4akage Rate Tests 14.2 33 ! 14.2.12.138 Containment Penetration L4akage Rate Teats 14.2 33 l 14.2.12.139 Containmeat Airlock 1.cakage Rate Tests 14.2 33 14.2.12.1.40 Contalament Integtsted leakage Rate Tests 14.2 33 i i 14.2.12.1.41 Pressure Suppression Containment Bypass L4akage i Ter,t 14.2 33 14.2.12.1.42 Containment isolation Valve Punction and Closure ') Timing Test 14.2 33 l 14.2.12.1.43 Wetwell to Drywell Vacuum Breaker System j Preoperational Test 14.2 33 14.2.12.1.44 Primary Containment Monitoring Instrumentation PreoperationalTest 14.2 34 O- 14.2.12.1.45 Electrical Systems Preoperational Test 14.2 34 14.2.12.1.45.1 DC Power Supply System Preoperational Test 14.2 35
- 14,2.12.1.45.2 Emergency AC Power D'.stribution System PicoperationalTest 14.2 35 !
- 14.2.12.1.453 Emergeney Diesel Generator
- PrcoperationalTest 14.2 36 l 14.2.12.1.45.4 Normal AC Power Distribution System PrcoperationalTest 14.2 37 14.2.12.1.46 Integrated ECCS loss of Offsite Power (LOP)/LOCA PreoperationalTest 14.2 38 14.2.12.1.47 Plant Communleations System PrcoperationalTest 14.2 39 14.2.12.1.48 Fire Protection System PreoperationalTest 14.2 39 14.2 si Amendment 2 I
ABWR maius Standard Plant am m SECTION 14.2 CONTENTS g sastian Ilth East 14.2.12.1.49 Radionethe Ligald Drainage and Transfer Systems Preoperational Test 14.2 40 14.2.12.1.50 Fuel Handling and Reactor Component Servicing Equipment Preoperational Test 14.2 41 14.2.12.1.51 Expansion, Vibration and Dynamle Effects PreoperationalTest 14.2 42 14.2.12.1.52 Reactor Vessel Flow Induced Vibration Prcoperational Test 14.2 43 14.2.12.1.53 Condensate and Feedwater Systems Pre-operationalTest 14.2 44 14.2.12.1.54 Condensate Cleanup S3 5 tem PrcoperationalTest 14.2-44.1 14.2.12.1.55 Reactor Water Chemistry Contr01 Systems PreoperationalTest 14.2 44.2 14.2.12.1.56 Condenser Air Removal System Preoperational Test 14.2 44.2 , 14.2.12.1.57 Offgas System PreoperationalTest 14.2 44.3 14.2.12.1.58 Hotwell LevelControf System PrcoperationalTest 14.2 44.4 14.2.12.1.$9 Condensate Storage and Transfer System PreoperationalTest 14.2 44.4 14.2.12.1.60 Circulating Water System Preoperational Test 14.2 44.4 14.2.12.1.61 Reactor Senice Water System PreoperationalTest 14.2 44.5 14.2.12,1.62 Turbine Building Cooling Water System PreoperationalTest 14.2 44.6 14.2.12.1.63 Turbine Senice Water System Preoperational Test 14.2 44.7 14.2.12.1.64 Main Turbine Contrel System Preoperational Test 14.2 44.8 14.2.12.1.65 Main Turbine Bypass System PreoperationalTest 14.2 44.8 14.2.sil Amendment 11 Oi I
t ABWR muoom ! Standard Plant m 1 C,m SECTION 14.2 ! CONTENTS (Continued) Arction Int East ! 14.2.12.1.66 Steam Bypass and Pressure Control System Preoperallonal Test 14.2 44.9 14.2.12.1.67 Peedwater Heater and Drsin S) stem ; PreoperationalTest 14.2 44.9 ; 14.2.12.1.68 Extraction Steam System Preoperstional Test 14.2 44.10 f 14.2.12.149 Moisture Separator / Reheater System PreoperationalTest 14.2 44.10 14.2.12.1.70 Main Turbine and Auxiliaries Prcoperational Test 14.2 44.11 i L 14.2.12.1.71 Main Generator and Auxiliary Systems Preoperational Test 14.2 44.12 14.2.12.1.72 riammability Control System Preoperational Test 14.2 44.13 14.2.12.1.73 Loose Parts Monitoring System ,q Prcoperational Test 14.2 44.13 14.2.12,1.74 Seismic Monitoring System Prcoperational Test 14.2 44.14 14.2.12.1.75 Liquid and Solid Radwaste Systems PreoperationalTest 14.2 44.14 . 14.2.12.1.76 Offgas System PreoperationalTest 14.2 44,15 14.2.12.1.77 Ultimate IIcat Sink Preoperational Test 14.2 44.15 14.2 vii.1 Amendment 11 a
ABWR 2amem Standard Plant Rrv. n SECTION 14.2 l CONTENTS (Continued) l l
$ttilDA M EARC i i
. I i j l I 14.2.12.2 General Discussion of Stattup Tests 14.2 46 14.2.12.2.1 Chemical and Radiochemical Measurements 14.2-46 l 14.2.12.2.2 Radiation Measurements 14.2 47 ) l 14.2.12.2.3 Fuel Loading 14.2 48 14.2.12.2.4 Full Core Shutdown Margin Demonstration 14.2 48 14.2.12.2.5 Rod Control System Performance 14.2 49 , l 14.2.12.2.6 Neutron Monitoring System Performance 14.2 50 l 14.2.12.2.7 Proccu Computer System Operation 14.2 50 ! 14.2.12.2.8 Core Performance 14.2 51 14.2.12.2.9 Nuclear Boller Process Monitoring 14.2 51 14.2.12.2.10 System Expansion 14.2 52 i 14.2.12.2.11 S) stem Vibtallon 14.2 53 l l 14.2.12.2.12 Reactor Internals Vibration 14.2 54 14.2.12.2.13 Recirculation Flow Control 14.2 54 l 14.2 vil.2 Amendment 11 I
ABWR mumm Standard Plant nvn Specific testing to be performed and the (c) proper operation of MSIVs and main O applicable acceptance criteria for each preoperational test are in accordance with the detailed system specifications and equipment steamline drain valves, including veri. fication of closure time in the isola. tion mode, and test mode,if applicable; specifications for equipment in those systems. The tests demonstrate that the installed (d) verification of SRV and MSIV secumulator equipment and systems perform within the limits capacity; of these specifications. (c) proper operation of SRV air piston The preoperational tests anticipated for the actuators and discharge line vacuum ABWR Standard Plant are listed and described in breakers; and the following paragraphs. Testing of systems outside the scope of the ABWR Standard Plant, but (f) proper system lastrumentation and that may have related design and therefore equipment operation while powered from testing requirements, are discussed in Subsection primary and alternate sources, including 14.2.12.3, transfers, and in degraded modes for which the system and/or components are 14.2.12.1.1 Nuclear Boller System expected to remain operational, Preoperational Test Other checks should be performed, as appro. (1) Purpc'se priate, to demonstrate that design requirements, such as those for string or installation, are To verify that all pumps, valves, actuators, met via as built calculations, visual inspec. l Instrumentation, trip logic, alarms, annun. tions, review of qualification documentation or clators, and indication associated with the other methods. For instance, SRV setpoints and nuclear boiler system function as specified, capacities should be verified from certification or bench tests to be consistent with applicable (2) Prerequisites requirements. The construction tests have been success. 14.2.12,1.2 Reactor Recirculation System fully completed and the SCG has reviewed the Preoperational Test test procedure and has approved the initla. tion of testing All required interfacing (1) Purpose systems shall be available, as needed, to support the specified testing and the To verify the proper operation of the , appropriate system configurations. reactor recirculation system at conditions ! approaching rated volumetric flow, including (3) GeneralTest Methods and Acceptance Criteria' the reactor internal pumps (RIPS) and motors, and the equipment associated with Performance should be observed and recorded ihe motor cooling, seal purge, and during a series of individual component and inflatable shaft seal subsystems, integrated system tests to demonstrate the following: (2) Prerequisites (a) verification that all sensing devices The construction tests have been success. respond to actual process variables and fully completed and the SCO has reviewed the provide alarms and trips at specified test procedure and has approved the initia. values; tion of testing. Cooling water from the re. actor building cooling water system and seal (b) proper operation of system instrumen. purge flow from the CRD hydraulle system tation and any associated logic, inclu. shall be available. The recirculation flow ding that of the automatic depressuri. control system should be sufficiently tested zation system (ADS); Amendment 11 14.2.'l
21A6100AN Remndard Plant trV. A to support RIP operation. Other interfacing (f) proper system flow rates including indl. systems shall be available, sw needed, to vidual pump capacity and discharge head; support the specified testing and the cor. responding system configurations. Reactor (g) proper manual and automatic system ope-vessellaternals should be capable of being ration and margin to actuation of pro. subjected to rated volumetric core flow. tective devices; (3) GeneralTest Methods and Acceptance Criteria (h) proper operation of inteflocks U.d equipment protective devices in pump and Testing of the recirculation system should be motor controls; coordinated closely with that of the recirco. lation flow control system (Subsection (i) proper operation of permissive, pro. 14.2.12.1.3) in order to adequately demon. hibit and bypass functions; strate proper lategrated system response and operat on. Also, the preoperational phase of (l) proper system operation while powered the reactor laternals vibration assessment from primary and alternate sources, program (Subsection 14.2.12.1.$2) involves including transfers, and in degraded entended operation of the recirculation sys. modes for which the system is expected tem and should be scheduled accordingly so as to remain operational; to optimize overall plant integtsted testing. (k) proper operation of the recirculation The scope and intensity of the preoperational motor seal purge subsystem over the full testing of the recirculation system and range of RPV pressures including the associated support subsystems will be limited proper functioning of the main header by the unavailability of nuclear heating. pressure control valve and proper Comprehensive testing of the system at rated distribution of seal purge flow to temperature and pressure will be performed individualpumps and motors; during the startup phase. (1) proper functioning of the recirculation l performance should be observed and recorded motor cooling subsystem and its ability during a series of individual component and to remove design heat loads from each lategrated system tests to demonstrate the RIP motor via the dedicated heat exchan. following: gers; (a) proper operation of instrumentation and (m) proper functioning of the recirculation equipment in all combinations of logic motor inflatable shaft seal subsystem , and instrument channel trip; and its ability to provide a temporary backup nealing mechanism for each pump (b) proper functioning ofinstrumentation and motor shaft during recirc motor alarms used to monitor system operation maintenance or removal; and availability; (n) acceptable pump / motor vibtation levels - (c) proper operation of system valves under and system piping movements during both expected operating conditions; transient and steady state operation; and (d) proper operation of pumps and motors in all normal design operating modes as well (o) acceptable reactor vesselinternals flow as any specified special testing induced vibration levels per the configurations; requirements of Subsection 14.2.12.1.52. (c) acceptable pump NPSH under the most System operation is considered acceptable limiting design flow conditions; when the observed / measured performance charac. O Amendment 2 14.2-8 i ii
ABM 2asiooxx mandmed Plant ' REV.A tr.rhtics, from the testing described above, meet . (d) proper operation of control systems in the applicable design specifications, all dedg:a operating modes and all le-vens of controls; 14.2.12.lJ Racircok tion Flow Control System Prooperational Test (c) proper operation of the sdjustable speed drives; (1) Pirpose
. . (f) ~ ability of the controi ,ystem to comm-To verify that the operation of the recir- unicate properly with equipment and culation flow control system, including that controllers in other systems;-
of the adjustable speed drives, RIP trip and rouback logic, and the. care flow measurement (g) proper control of pump motor start subsynem,is as spr*n? sequence; (2) Prerequisites (h) proper operation oiinterlocks and i equipment protective devices; The construction tests have been succw.- fully completed and the SCG has reviewed the (1) proper operation of permissive, prohi. _l test procedure and has approved the initia. bit and bypass functions; and - i tion of testing. All required inte. facing j systems shall be available, as needed, to (j) proper system operation while powered support the specified testing and the from primary and ~ alternate sources, corresponding system configurations. Including transfers, and in degraded modes for which the system is expected. (3) General Test Methods and Acceptance Criteria to remain operational. Some portions of the recirculation flow con. System operation is considered acceptable trol system testing should be performed in when the observed / measured performance charac. conjunction with that of the recirculation teristics, from the testing described above, ; system, as described in Subsection meet the applicable design specifications. 14.2.12.L2i Close coordination of the testing specified for the two systems is 14.2.12.1.4 Feedwater Control System j required in order to demonstrate the proper PreoperationalTest ; kegrated system response and operation. (1) Purpose J
. Performance should be cbserved ard recorded during a series of individual component and To verify proper operation of the feedwater integrated system tests to demonstrate the control system, including individual compo-following: nents such as controllers, indicators, and controller software settings such as gains (a) proper operation ofinstrumentation and and function generator curves.
equipment in all combinations of logic and instrument channel trip including (2) Prerequisites i recirculation pump trip (RPT) and runback circuitry; The construction tests have been success-fully completed and the SCG has reviewed the (b) proper functioning of instrumentation test procedures and has approved the initi-and alarms used to monitor system ation of testing. Preoperational tests must operation and availability; be completed on lower level controllers that do not strictly belong to the feedwater con-(c) prope: functioning of the core flow trol system but that may affect system re- , measurement subsystem; sponse. All feedwater control system com. Amendment 2 14.2 9 l
I: 2sA6100AN neandard Plane' arv.n ponents should have an initial calibration'in - . System operation is considered acceptable < accordance with vendor instructions, All when the observed / measured performance required interfacing systems should be avall- characteristics, from.the testing described' able, as needed, to support the specified above, m eet- the applicable design testing a'ad the appropriate system configu- specifications.- rations.
.14J.12.1.5 Standby Ugold Control System (3) - GeneralTest Methods and Acceptance Criteria Pr, operational Test I l
Testing of the feedwater cbatrol system dur. (1) Purpose i ing the preoperational phase may be limited - by the absence of an acceptable feedwater To ve'rlfy that the operation of the standby recirculation flow path. Comprehensive flow ' liquid control (SLC) system, including testing will be conducted during startup pumps, tanks, control, logic, and . Phase.' instrumentation, is as specified. !
-i Performance should be observed and recorded (2) Prerequisites during a series of individual component and overall system response tests to demonstrate - 'The construction tests have been success-the following: fully completed and the SCG has reviewed the test procedure and has approved the-(a) proper operation of instrumentation and initiation of testing.; Valves'should be controls in all combinations of logic previously bench tested and other pre- ;
and instrument dannel trip including . cautions relative to positive displacement i verification of setpoints; pumps taken, The reactor vessel should be 1 available for injecting Deminerallred - (b) proper functioning of instrumentation water. All required interfacing systems ' and alarms used to monitor system opera- ~s hall be available, as needed, to support - ! tion and status; the specified testing and the appropriate j system configurations. - i (c) Proper operat.4n of system valves, in. i cluding timing and stroke,in response (3) GeneralTest Methods and Acceptance Criteria to control demands (including the reactor water cleanup system dump valve Performance should be observed and recorded response to the low flow controller); during a series of individual component and - ) lategrated system tests to demonstrate the (d) proper operation of interlocks and following: eq'uipment protective devices in pump and valve controls; -(a) proper operation of instrumentation and equipment in all combinations of logic ; (e) proper operation of permissive, prohl- and instrument channel trip; bit, and bypass functions; i (b) proper functioning of instrumentation (f) proper system operation while powered and alarms used to monitor system opera-from primary and alternate sources, in- tion and availability; cluding transfers, and in degraded modes foe which the system is expected to re- (c) proper operation of system valves, in-main operational; and cluding timing, under expected operating conditions; (g) proper communication and interface with other control systems and related (d) proper operation of pumps and motors in equipment. all design operating modes; ' O' Amendment 11 14.2 10 l l
ABWR- 2mim
. Standard Plant - aw n (c) proper operation of the tank heaters and should be installed and ready to be stroked proper mixing of the neutron absorber and scrammed, Reactor building cooling solution; water, instrument air,' and other required interfacing systems shall be available, as (f) proper system flow paths and flow rates needed, to support the specified testing and including pump capacity and discharge the corresponding system configurations.
- head (with demineralized water substi-tuted for the neutron absorber mix- Additionally, the rod control-and ture); information system shall be functional when
, needed, with the applicable portion of its (g) proper pump motor start sequence and specified preoperational testing complete, margin to actuation of protective de. vices; (3) GeneralTest Methods and Acceptance Criteria (h) proper operation of interlocks and equip- Performance should be observed and recorded ment protective devices in pump and val. during a series of individual component and' ve controls; integrated system tests to demonstrate the following: (i) proper operation of permissive, pro. l hibit, and bypass functions; (a) proper functioning of instrumentation
^
and alarms used to monitor system opera-(j) proper system operation while powered tion 'and status; j from primary and alternate sources, in. cluding transfers, and in degraded modes (b) proper communication with, and response for which the system is expected to re. to demands from, the rod control and main operational; and information system and the reactor pro-tection system, including those associa. (k) acceptability of pump / motor vibration ted with alternate rod insertion, alter- l levels and system piping movements dur. nate rod run in (post scram), and select - ing both transient ~ and' steady state control rod run in functions; operation, ; (c) proper functioning of system valves, in. System operation is considered acceptable when - cluding purge water pressure control-the observed / measured performance characteris- valves, under expected operating condi. tics, from the testing described above, meet the tions; . applicable design specifications, (d) proper operation of CRD hydraulic sub. 14.2.12.1.6 Control Rod Drive System system pumps and motors in all design :l PreoperationalTest operating modes;
.i (1) Purpose (c) acceptable pump NPSH under the most lim- !
iting design flow conditions; 3 To verify that the control ro'd drive (CRD) -! system, including the CRD hydraulic and fine (f) proper pump motor start sequence and mar-motion control subsystems, functions as de- gin to actuation of protective devices; signed,
~
(g) proper system flow paths and flow rates (2) Prerequisites including sufficient pump capacity and discharge head; The construction tests have been success-fully completed and the SCG has reviewed the (h) proper operation of interlocks end test procedure and has approved the equipment protective devices in pump, initiation of testing. The control blades motor, and valve controls; Amendment 11 14.2 11 I i I
. . . _ . . _ . . . . . ~ . . . . .. .
1 L -
'i i, MN 23A6100AN L Standard Plant ag;u (i) proper operation of permissive, prohl.
bit, and bypass functions; clated alarms and annunciators in all
- combinations of logic and lastrument e;. 'l) channel trip including all positions of -
.(j) proper system operation while powered 'the reactor mode switch;
'from primary and alternate sources, Including transfers, and in degraded (b) proper functioning of lastrumentation l lE modes for which the system is expected! used to monitor CRD system status such ,
to temain operational; . as rod position indication lastrumenta. 1 tion and that used to monitor continuous ' 1
- (k) acceptability of pump / motor vibration full-in and rod / drive separation status; . l levels and system. piping movements during both transient and steady state '
(c) proper operation of RC&lS software in- , cluding verification of gang and group
. operation; assignments and predictor.comparator, .
(1) proper operation of fine motion motors rod worth limiter, and banked position ; and drives and anociated control units, withdrawal sequence functions; and ; including verification of acceptable j normalinsert and withdraw timing;; (d) proper communication with interfacing i systems such as the power generation con-(m) proper operation of hydraulic control trol system, the automatic power regula-units and associated valves including. tor, and the' automatic rod block - monitor. ! CRD scram timing demonstrations against ' atmospheric pressure. System operation is considered acceptable ; Systr.m operation is considered acceptable when when the observed / measured performance charac-the observed / measured performance characteris- teristics :from the testing described'above, tics, from the testing described above, meet the meet the applicable design specificationsi . applicable design specifications. _ . r 14.2.12.1.8 Residual Heat Removal System 14.2.12.1.7 Rod Control and Information System Pnoperational Test - Pnoperational Test (1) Purpose (1) Purpose . To verify the proper operation of the resi-~ To verify that the rod control and informa- dual heat removal (RHR) system under its tion system (RC&lS) functions as designed. various modes of operation: core cooling, L i' shutdown cooling, wetwell and drywell spray, I (2) Prerequisites suppression pol cooling, and supplemental j fuel pool cooling. 1 The construction tests, including initial , check out of RC&lS software, have been suc. (2) Prerequisites .! cessfully completed and the SCG has reviewed. . a the test procedure and has approved the ini. The construction tests have been successful- ) tiation of testing. ly completed and the SCG has reviewed the test procedure and has approved the initia-(3) General Test Methods and Acceptance Criteria tion of testing. The reactor vessel shall-be intact and capable of receiving injection ') Performance should be observed and recorded flow from the various modes of RHR. The I during a series of tests to demonstrate the reactor building cooling water system and following: other required interfacing systems shall be. available, as needed, to' support the speel- 1 (a) proper operation of rod blocks and asso- fied testing and the apprt,priate system i configurations. Amendment 2 14.2-12 1 i j l
4 23A6100AN Standard Plant nev. A and instrument channel trip; System operation is considered acceptable 9 when the observed / measured performance charac-
'(b) proper functioning ofinstrumentation and - teristics, from the testing described above, alarms used to monitor system operation meet the applicable design specifications.
and availability; 14.2.12.1.11 Safety System Imgic and Centrol (c) proper operation of system valves,includ. Preoperational Test
. ing timing, under expected operating conditions; (1) Purpose 1
(d) proper operation of pumps and motors in To verify proper operation of the plant alldesign operating modes; safety system logic and control (SSLC). (c) acceptable pump NPSH under the most (2) Prerequisites limiting design flow conditions; The applicable construction tests have been (f) proper system flow paths and flow rates successfully completed. including pump capacity and discharge . head and time to rated flow; (3) GeneralTest Method and Acceptance Criteria - 1 (g) prcper pump motor start sequence and mar. The SSLC integrates the automatic decision - gin to actuation of protective devices; msking and trip logic functions associated with the safety action of several of the (b) proper operation of interlocks and plants' safety related systems. Such sys-equipment protective devices in pump, tems include the RPS,HPCF RHR,RCIC LDIS, , motor, and valve controls; and ADS. The SSLC is not so much a system itself, but is instead an assembly of the O (i) proper operation of permissive, prohibit, and bypass functions; above mentioned safety related systems signal processors designed and grouped for optimum reliability, availability and oper. (j) proper system operation while powered ability. The SSLC, therefore, should be from primary and alternate sources,in- adecuately tested during the preoperational
'cluding transfers, and in degraded modes phase testing of the associated systems in-for which the system is expected to cluding the integrated LOP /LOCA test Pro-remain operational; vided the construction testing and the assoi clated system preoperational testing has 1 (k) acceptability of pump / motor vibration le. been successfully completed, as it relates !
veis and system piping movements during to proper operation of the SSLC, no specific both traasient and steady state additional testing should be necessary.. operation; SSLC performance would then be considered ac- I (1) the ability of the system to swap pump ceptable provided all _ design and testing speci-suction source from the condensate fications are met, storage pool to the suppression pool without interrupting systern operation; 14.2.12.1.12 Multiplexing System Preoperational Test (m) acceptability of the HPCF sparger flooding pattern; and (1) Purpose * (n) proper operation of the pump discharge To verify proper functioning of the plant line keep fill system and its ability to multiplexing system including both essential prevent damaging water hammer during and nonessential subsystems, system transients. Amendment 2 14.2 15 i
. . . . , , , _ , , , . , , ., , . , * ~
i 4
- ABWR m ioo m l Standard Plant an' n l (2) Prerequisites drywell sumps.
System construction testing has becn Performance should be observed and recorded l successfully completed. during a series of individual component and ' ] integrated system tests to dear.onstrate the 1 (3) GeneralTest Method and Acceptance Criteria- following: ! Since this system is the primary communica. ; (a) proper operation of instrumentation and
- Ltion interface between the various plant' controls in all combinations of logic l systems it should be adequately tested dur- and instrument channel trip;
]
ing the preoperational phase testing per- . formed on those laterconnected systems. (b) proper functioning of indicators, annun. ;
' Provided the construction testing and the ' clators, and alarms used to monitor sys- 2
, associated system testing has been success. tem operation and status; fully completed as it relates to proper ope. . . ration of the. multiplexing system, no spect. . (c) proper operation ofleekoff and drainage l fic additicaal testing should be necessary. . measurement functions such as those asso-- clated with the reactor vessel' head - System performance would then be considered' flange, drywell cooler condensate, and-acceptable provided all design specifications are various primary system valves; I met. (d) proper response of related system val- t 14.2.12.1.13 Isak Detection and isolation ves, including timing, under. expected , System Preoperational Test operating conditions; (1) Purpose (c) proper interface with' related systems in regards to the input and output of leak To verify proper response and operation ofl ' detection indications and isolation ini.' the leak detection and isolation system (LDS) tlation commands; l logic. (f) proper operation of bypass switches and (2) Prcrcquisites related logic; and : The construction tests have been successfully (g) proper system operation while powered completed and the SCG has reviewed the test from primary and alternate sources, in-procedures and has approved the initiation of- cluding transfers, and in degraded modes ,
- testing. The required AC and DC electrical for which the system is expected to re. -
l power sources should be operational and the main operational. s
- l. appropriate interfacing systems shall be l available as required to support the System operation is considered acceptable
- l. specificd testing, when the observed / measured performance charac-teristics, from the testing described above,
! (3) GeneralTest Methods and Acceptance Criteria meet the applicable design specifications.' Since the leak detection and isolation system 14.2.12.1.14 ' Reactor Protection System is comprised mostly of logic, the checks of PreoperationalTest valve response and timing and the testing of i sensors will be performed as part of, or in (1) Purpose conjunction with, the various systems with l which they are associated. These systems in. To verify proper operation of the reactor - l ciudc R11R, RCIC, RWCU, main steam,feedwat- protection system (RPS) including complete l er, recirculation, radiation monitoring, channel logic and response time.
- nuclear boiler, drywell cooling and the Amendment 11 1G16 -
l l ['
u L ABWR mmm , l Standard Plant mn { p_ To verify proper operation of the automatic Verify the feasibility and operability of power regulator (APR) over the range of intended remote shutdown functio::s from the required operating modes.- remote shutdown panel and other local and remote locations outside the main control (2) Prerequisites room which will be utilized during the
.. remote shutdown scenario. ,
The software programming and inith! Alarou}}