ML20082B551
| ML20082B551 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 06/28/1991 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20082B542 | List: |
| References | |
| NUDOCS 9107150349 | |
| Download: ML20082B551 (586) | |
Text
{{#Wiki_filter:, _ _ .. - - - - ABWR SSAR ( Q) Amendment 17. Page change instruction The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back). Hold page numbers represent a page that has been changed by Amendment 17. REMOG ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAQll.Ng, CilAFI'ER 1 3.2 13,14 3.2-13,14 3.2-14a,15 3.2-14 1,15 1 ii,ili 1 il,ill 3.2 16,17 3.2-16,17 3.2 18,19 3.2 18,19 1 , 7 - 11 1.7-11 3.2-19a,20 3.2 19.1.20 1.7-1 1.7 1 3.2 21,21.1 3.2-21,21,1 1.74,5 1.7-4,5 3.2-21.2,22 3.2-21.2.22 1.75.1,5.2 1.7-5.1,5.2 3.2 23,23.1 3.2 23,23.1 3.2-24,24.1 3.2 24,24.1 1.8-42,43 1.8-42,43 3.2-25,26 3.2 25,26 1.8-48,49 1.S-48,49 3.2-27,28 3.2 27,28 1.S-57,5S 1.8 57,58 3.2 29,30 3.2-29,30 1.8-61,62 1.8-61,62 3.2 31,32 3.2 31,32 , 1.8-67,(6 1.8-67,68 3.2 33,34 3.2 33,34 ( 1.S-89,90 1.8-89,90 3.2-34.1 3.2-34.1 l Add 1.8 91 l O 3.4-1,1.1 3.4-1,1.1 h 1A-li,ili 1A-iv,v 1A II,lii 1A-Iv,v 3.5- 1,2 3.51,2 ! 1A.2 li,iii 1AJ-il,ill 3.6-5,6 3.6-5.5.1 l 1A 2 iv,v 1A.2-Iv,v Add 3.6-6 1A.2-8a,9 1A.2-8.1.9 3.6-27,27.1 3.6-27,27.1 1A.217,18 1 A.2-17,18 3fr30,30.1 3.6-30,30.1 Add 1A.2 18.1 3.6-32,33 3.6-32,33 1A.2-19,19a I A.2-19,19.1 3.8 22,23 3.8-22,23 1A.3-1 1A.3-1 3833,34 3.8-33,34 ClIAPTER 3 3.9-43,44 3.9-43,44 3.0-vi 3.0-vi 3.11 1,1.1 3.l bl.l.1 3.11 2,3 3.11 2,3 3.2-6,7 3.2-6,7 3.2-8,8.1_ 3.2-8,8.1 3 G li 3G-il 3.2-9 3.2-9 3G.5-1 3G.5-1 3.2-10,11 3.2-10,11 3.211a,12 3.2 11.1,12 311,4-1 Issued as 311.160 31I.4-1 31.3-ii,iii 3 1.3 -11,111 31.3-1 31.31 25 O 9107150349 910628 PDR ADOCK 05000605 K PDR
ABWR SSAR Amendment 17. Page change instruction (Continued) O l ! Tue following pages have been changed, please make the specilled changes in your SS AR. Pages are listed as page pairs (front & back). Bold page numbers represent a page that has been changed by Amendment 17. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CliAITER 4 6.2-50.23,50.24 6.2 50.23,50.24 6.2-50.25,50.26 6.2-50.25,50.26 f 6.2 50.27,50.28 4.4-3,4 4.4-3,4 6.2 50.27,50.28 l 6.2 50.29,5030 6.2 50.29,5030 CilAITER 5 6.2 5033,5034 6.2 5033,5034 l 6.2-5035,5036 6.2 50.35,5036 l l 5.2-15,15.1 5.2 15,15.1 6.2-5037,5038 6,2-5037,5038 5.2-29,29.1 5.2-29,29.1 6.2-5039,50.40 6.2-5039,50.40 l 5.2 29.2,293 5.2 29.2,293 6.2-50.41,50.42 6.2 50.41,50.42 6.2-50.43,50.44 6.2-50.43,50.44 5.4-iv,v 5.4-iv,v 6.2-50.45,50.46 6.2 50.45,50.46 5.4-18.1 5 4-18.1.19 6.2 50.47,50.48 6.2 50.47,50.48 5.4-19,20 5.4-19.1,20 6.2 50.49 6.2-50.49,50.49a Add 6.2 50.49b CllAITER 6 6.2-50.50,50 51 6.2 50.50,50.51 6.2-50.52,50.53 6.2 50.52,50.53 6.2iv,v 6.2-iv,v 6.2 50.54 6.2-50.54 6.2-vi,di 6.2-vi,vii 6.2 50.57,50.58 6.2 50.57,50.58 6.2-viii,ix 6.2 sill,ix 6.2 50.59 6.2-50.59 6.2-1,2 6.2-1,2 6.2-21,22 6.2-21,22 6.4-3,4 6.43,4 6.2 22a 6.2-22.1 6.2-23,24 6.2 23,24 6.5-1,2 6.5-1,2 6.2-29,30 6.2 29,30 6.5-3,4 6.5-3,4 Add 6.2-30.1 ' 6.2-33,33.1 6.2 33,33.1 ClIAITER 7 6.2-41,42 6.2-41,42 6.2-43,43a 6.2-43,43,1 7.1 26 7.1-26 6.2-44 6.2-44 6.2-503,503a 6.2 503,503a 7.2 31 7.2-31 6.2 503b,50.4 6.2 503b,50.4 6250.5,50.6 6.2 50.5,50.6 7.6-43,44 7.6-43,44 6.2 50.7,50.8 6.2 50.7,50.8 6.2-50.9,50.10 6.2 50.9,50.10 CllAITER 8 6.2-50.11,50.12 6.2-50.11,50.12 6.2 50.13,50.14 6.2-50.13,50.14 8.2 il 8.2 -11 6.2-50.15,50.16 6.2-50.15,50.16 8.2-1 8.21,2 6.2-50.17,50.18 6.2-50.17,50.18 6.2-50.19,50.20 6.2-50.19,50.20 83-6,6.1 83 6,6.1 6.2 50.21,50.22 6.2-50.21,50.22 8 3-23,23.1 8 3-23,23.1 8 3-28,29 83-23,29 O ABWR SSAR l r,,
l Amendment 17. Page change iretruction (Continued)
The following pages base been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back). Bold page numbers represent a page that has been changed by Amendment 17. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. CilAPTER 9 9.5-1 1.7 951 1,7 9.5-10.7 9510.7 9-il,ili 9 1i,111 Add 9510f,10g 9-iv 9-Iv,v Add 9B Cover 9.1 il,iii 9 . 1 - 11,111 Add 9B 11 9,1 vi,vii 9.1-vi,vil Add 98.1 1 9.1 1,1.1 9.11,1.1 Add 9B.2li 9.1-2,2a 9.1 2,2a Add 9B.21 7 9.12d,2e 9.12d,2e Add 9B.2-8,9 9.13,4 9.1 3,4 Add 9B.2 10,11 9.1-5 9.1 5 Add 9B.2 12 9.1 7,8 9.1-7,8 Add 9B3-1 9.1 9,10 9.1 9,10 9.1 11,11a 9.1 11,12 CilAFIT.R 10
^ 9.1 12,13 9.1 13
(,) 9.2 il, iia 9.2 li,ila 10-ii,ili 1 0 11,111 9.2-iii, 9 . 2 -111 10.2 7,8 10.2 7,8 9.2-2,3 9.2 2.3 9.2-3.1 9.2-3.1 10 3-3 103-3 9.2-4,4.1 9.2-4,4,1 9.25,6 9.2-5,6 10.4-ii,iii 10.4.ii,lii 9.2-6.1 9.2-6.1 10.4-iv,v 10.&lv,t 9.2-7,8 9.2-7,8 Add 10.4 v.1 9.2-9,9.1 9.2-9,9.1 10.4 3,4 10.4 3,4 9.2 13 9.2-13 10.4-5,6 10.4-5,$.1 9.2-15.1 9.2-15.1 Add 10.4 6 9.2-24,25 9.2 24,25 10.4-7,8 10.4-7,7.1 Add 10.4-8 9.4-1,1,1 9.41,1,1 10.4 9,10 10.. 1,10 9.4 1.2,13 9.41.2,13 Add 10.& 10.1 9.41.4,1.5 9.4-1.4.1.5 10.417 10.&l7 9.4-2c 2ia 9.4 2c 2ia 10.4 26,27 10.4 26,27 9.4-2j,2k 9.42j,2k CllAFTER 11 9.5-li,ili 9.5 H,li.I Add 9.5 iii 11.4-la,iii 11.4ti,iii 9.5-v.1,v.2 9.5-v.1,v.2 11.41 9 11.4-1 9 9.5-si 9.5-vi O
ABWR SSAR Amendment 17. Page change lustruction (Consinued) The following pages have lieen changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), llold page numbers represent a page that has been changed by Amendment 17. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. 11.5-il,ili 1 1 .5-11,111 15.6-si,vii 15.6-vi,vii 1151,2 11.5-1,2 15.6-1,2 15.6-1,2 Add 1152.1 1153,4 1153,4 15.6-9,10 15.6-9,10 11.5-5,6 1155,6 15.6-11,12 15.6 11,12 11.5-7,8 11.5-7,8 15.6-13,14 15.6-13,14 11.5-9,9.1 11.5-9,9.1 15.6-15,16 15.6-15,16 11.5-10 11510 15.6-17,18 15.6-17,18 11.5-11,12 11.5-11,12 15.6-25,26 15.6-25,26 11 5 13,14 11.5-13,14 15.6-27 15.6-27 11 5 15,16 11.5-15,16 11.5-17,18 11.5-17,18 15.7 ii,iii 1 5 .7 -11,111 11 5 19,20 11.5-19,20 15.7-iv,v 15.7-1v,v 11.5-21,22 11 5 21,22 15.7 1,2 15.7-1,1.1 Add 15.7 2 CilAPTER 12 15.7 5,6 15.7-5,6 15.7-7,8 15.7-7,8 123-20,21 12 3-20,21 15 7-9,10 15.7 9.10 12 3-22,23 12 3-22,23 15.7-11,12 15.7 11,12 12 3-26,27 123-26.2/ 15.7-13,14 15.7-13,14 12 3-36,37 123-26,27 15.7 15,16 15.7 15,16 123-42,43 123-12,43 15.7-17,18 15.7 17,18 12 3-44,45 123-44,45 15.7-19,20 15.7-19,20 12 3-56,57 123 56.57 15.7-21 15.7 21 123-58,59 123-58,59 12 3-60,61 123-60,61 15A.1 1 15A.1-1 123-62 123-62 15A.2 6,7 15A.2-6,7 123-68 123-68 15A.6-ii,ili 15A.6-ii,ill i 15A.6-5,6 15A.6-5,6 CIIAl'TER 15 15A.6-7,8 15A.6-7,8 ! Add 15A.6-8.1,8.2 15/)-10,11 15.0-10,11 Add 15A.6-8.3 15A.6-9,10 15A.6-9,10 15 3-3,4 15 3-3,4 15A.611,12 15A.6-11,12 15A.6-13,14 15A.6-13,14 15.4-5,6 15.4-5,6 15A.6-15,16 15A.6-15,16 Add 15A.6-19.1 15.5-12 15.5-1,2 15A.6-20,21 15A.6-20,21 I I 1
*k-
p ABWR SSAR v Amendment 17. Page change lastruction (Continued)
'The following pages huse been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back), Bold page numbers represent a page that has been changed by Amendment 17, REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No.
15A.6-22,23 15A.6-22,23 19B.2 li,iii 19B.2 II,ili 15A.G24,25 15A.6-24,25 19B.2-1 19B.2-1 62 15A,6-41,42 15A.6-41,42 15A.6-45,46 15A.6-45,46 19B3-1i 1983 11 15A.6-47,48 15A.6 47,48 19B3-1 1983-I.2 15A.6-51,52 15A.6-51,52 15A.6-53,54 15A.6-53,54 Cl(Al"TER 20 15A.6-71,72 15A.6-71,72 15A.6 73,74 15A.6-73,74 20-il,ili 2 0 -i1,111 15A.6-75,76 15A.6-75,76 15A.6-77,78 15A.6-77,78 20.1-1,1.1 20.1 1,1.1 20.1 1.2,13 20.1-1,2,13 CilAfr1TR 16 20.1 1.4,1.5 20.1 1.4,1.5 20.1-1.6,1.7 20.1-1.6,1.7 16.9-29,30 16.9 29,30 20.1-1.8,1.9 20.1 1.8,1.9 C 20.1 1.10.1.11 20.1 1.10,1,11 20.1 1.12,1,13 \ CllAI"TER 17 20.1-1.12.1.13 20.1 1.14.1.15 20.1 1.14,1.15 17,1 1,2 17.1-1,2 20.1-1.16.1.17 20.1-1.16,1,17 20.1-1.18.1.19 20.1 1.18,1,19 CIIAl'TER 18 20.1-1.20.1.21 20.1 1.20,1.21 Add 20.1 1.22 18A.21 18A.2-1 4 18A.41 18A.41 15 20.2-4p,4q 20.2-4p,4q 18A.9-1 18A.91 3 20.2-25,25.1 20.2 25,25.1 18A.121 18A.12-1 7 20.2-25.2,25 3 20.2-25.2,253 20.2-25.4,25.5 20.2-25.4.25.5 ISB 2 ISB-2 23 20.2 25.6,25.7 20.2 25.6,25.7 20.2-25.8,25.9 20.2 25.8.25.9 18D-1 18D.21 6 20.2 25.10,25.11 20.2-25.10,25.11 20.2 25.12,25.13 20.2-25.12,25.13 CilAlrTER 19 20.2-25.14,25.15 20.2 25.14,25.15 20.2-25.16 20.2-25.16,25.17 19A.2-1 19A.21 - 11 Add 20.2 25.18,25.19 Add 20.2-25.20,25.21 19A3-1 19A3-1 Add 20.2 25.22,25.23 Add 20.2-25.24,25.25 19B iv 198-Iv Add 20.2-25.26,25.27 19B.1-1 1M3.1 10 203 ii, iia 203-li, iia 203 iii 20 3 111 20 3-58,59 20 3-58,59 5-
ABWR SSAR Amendment 17. Page change lustruction (Continued) I The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front & back). Iloid page numbers represent a page that has been changed by Amendment 17. REMOVE ADD REMOVE ADD PAGE No. PAGE No. PAGE No. PAGE No. 20 3-82,82.1 20 3-82,82.1 1 20 3-87,87.1 20 3-87,87.1 20 3 201,202 203 201,202 20 3-351.23 - 351.28 203-351.1 351.27 20 3-355 385 20 3-355 386 Add 20 3-387,388 Add 20 3-389,390 Add 20 3-391,392 Add 203-393,394 Add 20 3-395,396 Add 20 3-397,398 Add 203 399,400 Add 20 3-401,402 Add 20 3 403,404 Add 20 3-405,406 Add 20 3-407,408 20,4-1 20.4 1 O
- 23A6100AC Standard Plant nev. c CHAPTER 1 -L TABLE OF CONTENTS Section 'Dtic j'agt 1- INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
1.11 1.1.1 Format & Contents 1.1 1 1.1.2 ABWR Standard Plant Scope 1.1 1 1.13 Enginecting Documentation 1.1.1 1.1.4 Type of License Required 1.11 1.1.5 Number of Plant Units 1.1 1 1.1.6 Description of Location 1.11 1.1.7 Type of Nuclear Steam Supply 1.1 1 1.1.8 Type of Containment 1.1 1 1.1.9 Core Thermal Power Levels 1.1 1 1.2 - GENERAL PLANT DESCRIPTION 1.21 1.2.1 Principal Design Criteria 1.2 1 1.2.2 Plant Description 1.25 1.3 COMPARISON TABLES 13-1 13.1 Nuclear Steam Supply System Design Characteristics 13-1 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 1A IDENTIFICATION OF AGENTS AND CONTRACTORS 1 A-1 0 1.ii Amendment 8
ABWR urame Standard Plant arv c CHAPTER 1 g TABLE OF CONTENTS (Continued) Section Title fact 1.5 RFOUIREMENTS FOR FURTilER TECilNICAI, INFORM ATION 1.51 , i 1.6 M ATERI AL INCORPORHED fly REFERENCE 1.6-1 1.7 DRAWINGS 1.7-1 j 1.7.1 Piping and Instrumentation and Process Flow Drawings 1.7.2 Instrument, Control and Electrical Drawings 1.71 1.73 ASME Standard Units Metric Conversion 1.7 1 Factors i 1.7.4 Metric Conversion to ASME Standard Units 1,7-1 1.7.5 Drawing Standards 1.7 1
'1.7.6 Interfaces 1.7-1 0
1.8 CONFORM ANCE WITil STANDARD REVIEW PI AN AND APPLICAHil_lTY OF CODES AND STANDARDS 1.8 1 1.8.1 Conformance With Standard Review Plan 1.8-1 1.8.2 Applicability of Codes and Standards 1.81 1.83 Interfaces 1.8-1 1.9 INTERFACES 1.91 APPENDIX 1A RESPONSES TO TMI RELATED MATTERS 1-iii Amendment 17
! AilWR. mouc Standard Plant neic SECTION 1.7 CONTENTS Section Illie Page 1.71 - PJpjptandlDitrumentation and Process Flow ,
Drawinns 13-1 l 1
-1.7.2 Instrument. Control and Electrical Drawines 13 1 .
1.7J ASME Standard Units Metric Conversion Factors 1.7 1 13.4 Metric Consersion to ASME Stanard Units 13-1 1.7.5 Drawine Standards 1.71 . 1.76. 1,7-1 Interfaces -
-SECTION 1.7 TAllLES !
h Inhle Title Page 13 1 Piping and Instrumentation and Process Flow
- Diagrams 13-2 1.7-2 Instrument Engineering, Interlock Block and Sing!c Line Diagrams 13 5 13-3 ASME Standard Units Metric Conversion Factors 1.7-5.2-13-4 Conversion Tables Metric to ASME Stanard Units 1.7-5,5 1.7-5 Drawing Standards 1.75.7 ILLUSTRATIONS Figure Illic Eage 1.71 Piping and Instrumentation Diagram Symbols 1.7 6 1.72 GraphicalSymbols for Use in IBDs 1.78 1 . 7 11 Amendment 17
ABWR 2 mmc S.tandard Plant nev. c
,o 1.7 DRAWINGS t )
'~#
1.7.1 Piping and Instrunentation and Process Flow Drawings Table 1.7.1 contains a list of system Piping and Instrumentation diagramr (P&lD) and process flow daigrams (PFD) provis ed in the ABWR SSAR. Figure 1.71 oefine, the sysmbols used on these drawings. 1.7.2 Instrument, Control and Electrical Drawings Interlocking block diagrams (IBD), instrument engineering diagrams (IED) and single line diagrams (SLD) are listed in Table 1.7-2. Figure 1.7-2 defines the graphic symbols used in the IBDs. 1.7.3 ASME Standard Units Metric l Conversion Factors The ASME Standard units are applied with the numerical values converted to the metric system as listed in Table 1.7-3.
/^\ 1.7.6 Metric Conversion to ASME V Standard Units Selected flow, pressure, temperature and lemgth mett:c units are converted to ASME satodard units as tabulated in Table 1.7-4.
1.7.5 Drawing Standards Guidelines for identifying systems, facilities, equipment types and numbers and for drawing P&ID's and FPD's are treated in Table 1.7 5. 1.7.6 Interfaces Applicants references the ABWR shall complete P&ID pipe schedules indicated as: Interface. i l l 1 i (nv) Amendment 17 1.71 L
ABWR usam^c nev c Standard Plant Table 1.71 PIPING AND INSTRUMENTATION AND PROCESS FLOW DIAGRAMS (Continued) SSAR Fig. No. Page No. Title Tygx 11.2 1 11.2-12 Radwaste System PFD 11.2 2 11.2-14 Radwaste System P&lD 11 3-1 11 3-21 Offgas System PFD 11 3-2 113-23 Offgas System P&ID O Amendment l'.1 17-4
ABWR nume Standard Plant nev c
- Table 1.7 2 INSTRU31ENT ENGINEERING, INTERLOCK BLOCK AND SINGLE LINE DIAGRAMS SSAR Fig. No. Page No. Title Tylw 5.2-8 5.2-44 leak Detection and Isolation System ILD 7.2-9 7.2-39 Reactor Protection System IED 7.2 10 7.2-47 Reactor Protect;on System IBD 73-1 73-52 High Pressure Core Flooder System IBD 73-2 73-59 Nuclear Boiler System IBD 73-3 73-74 Reactor Core Isolation Cooling System IBD 73-4 73 90 Residual Heat Removal System IBD 73-5 73-104 Izak Detection and Isolation System IBD 73-6 7 3-105 Standby Gas Treatment System IBD 73-7 731'6 Reactor Building Cooling Water System IBD 73-8 73-132 Essential HVAC System !BD O 73-9 7 3-133 HVAC Emergency Cooling Water System F1D 7 3-10 7 3-141 High Pressure Nitrogen Gas System IBD 7.4-1 7.4-19 Standby Liquid Control System IBD 7.4-2 7.4-24 Remote Shutdown System IED 7.4-3 7.4-26 Remote Shutdown System IBD 7.6-1 7.6-24 Neutron hionitoring System IED
! 7.6-2 7.6-28 Neutton hionitoring System IBD l 7.6-5 7.6-46 Process Radiation hionitoring System IED 7.6-6 7.6-59 Fuel Pool Cooling and Cleanup System IBD 7.6-7 7.6-64 Containment Atmosphere hionitoring System IED i 7.6-8 7.6-69 Containment Atmosphere hionitoring System IBD 1 l 7.6-11 7.6-80 Suppression Pool Temperature Monitoring System iED Amendment 17 1.75
- ABWR- MMMAC
. Standard Plant nev c ' I
- Table 1.7 2 INSTRUMENT ENGINEERING, INTERLOCK BLOCK AND SINGLE LINE DIAGRAMS (Continued)
SSAR l'ig. No. Page No. Title Type 7.6 12 7.6-81 Suppression Pool Temperature Monitoring Sptem IBD 1 7.7-2 7.7 42 Rod Control and Information System IED 7.73 7.7 45 Rod Control and Information System IBD 7.74 7.7 54 Control Rod Drive System IBD 7.75 7.7-61 Recirculation Flow Control System IED 7.7-7 7.7-64 Recirculation Flow Control System - IBD 7.7-8 7.7 73 Feedwater Control System IED 7.7-9 7.7 75 Feedwater Control System - IBD 83 8 3-30 Electrical Power Distribution System SLD 83-2 83-31 6.9KV System SLD S3-3 83-32 480 V System SLD 834 8 3-33 Instrument Power Supply SLD 83 5 3^ Process Computer Constant Voltage Constant Frequency Power Supply SLD 83 6 83 35 Safety System Imgic and Control Power Supply SLD 83-7 - 8 3-36 125 VDC Power System SLD 83-8 8 3-37 250 VDC Power System SLD t A i O Amendment 17 17-51
ABWR nui=^c Standard Plant ned Table 1.7 3 l ASME STANDARD UNITS METRIC CONVERSION FACTORS (1) Pressure / Stress l
\'
1 Pound / Square Inch 0.0007031 Kilograms / Square MM l 1 Pound / Square Inch 0.07031 Kilograms / Square CM 1 Atmospheres (STD) 1.0332 Kilograms / Square CM 1 Feet of Water (68 F) 0.03042 Kilograms / Square CM 0.002535 Kilograms / Square CM 1 inches of Water,C) 1 Inches of HG (0 0.03453 Kilograms /Squaic CM l (2) Force /Weichi 1 Pounds 453.59 Grams 1 Pounds 0.45359 Kilograms 1 Ton (Short) 907.2 Kilograms 1 Tons (Short) 0.9072 Tonnes (3) Heat /Enerev 1 BTU 1055,056 Joules 1 BTU 0.252 Kilocalories 1 BTU 0.000293 kilowatt Hours 1 Horsepower 0.7457 Kilowatts 1 Horsepower Hour 0.7457 Kilowatt Hours 1 BTU / Minute 15.12 Kiiocalorie/ Hour 1 BTU / Pound 0.555568 Kilocalorie/ Kilogram (4) Length 1 Inches 25.4 Millimeters 1 Inches 2.54 Centimeters 1 inches 0.0254 Meters 1 Feet 03043 Meters 1 Feet 30.48 Centimeters 1 Miles 1609.3 Meters 1 Miles 1.W)3 Kilometers (5) Echtmt 1 Cubic Inches 0.016387 Liters 1 Cubic Inches 16387 Cubic Centimeters 1 Cubic Feet 0.02832 Cubic Meters 1 Cubic Feet 28320 Cubic Centimenters 1 Cubic Feet 2832 Liters 1 Cubic Yard 0.7M6 Cubic Meters 1 Gallons (US) 3.7854 Liters O' Amendment 8 1.75.2
ABWR 2 mime Standard Plant PD' G
,m
, TAllLE 1.8 20 t
)
RGs Applicable to AflWR (Continued) AllWR Appl. Issued Appli-FG No, Reculatory Golde Title EtL llait tahlr1 Lhm.;nen13 1.60 I Design Response Spectra for Seismic Design 1 12/73 Yes ' of Nuclear Power Plants. 1,61 Damping Values for Geismic Design of Nu- 0 10/73 Yes clear Power Plants. 1.62 ManualInitiation of Protective Actions. 0 Yes 10/73 1.63 Electric Penetration Assemblies in Contain- 3 Yes 2/87 i ment Structures of Nuclear Power Plants.
)
1.64 Ouality Assurance Requirements for the De- Superceded See Table i sign of Nuclear Power Plants. 17.0-1 1 1.65 Materials and Inspections for Reactor Ves- 0 Yes 10/73 sel Closure Studs. 1,68 Initial Test Programs for Water-Cooled 2 8/78 Yes
]v 1.68.1 Reactor Power Plants, Preoperational and Initial Startup Testing 1 1/77 Yes of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants.
1.68.2 Initial Startup Test Program to Demonstrate Yes 1 7/78 Remote Shutdown Capability for Water-Cooled Nuclear Power Plants. 1.68.3 Preoperational Testing of Instrument and 1 7/78 Yes Control Air Systems. 1.69 Concrete Radiation Shicids for Nuclear Po- 0 Yes 12/73 wer Plants. 1.70 Standard Format and Content of Safety Ana- 3 Yes 11/78 lysis Renorts for Nuclear Power Plants. 1.71 Welder Qualifications for Areas of Limited 0 12/73 -- Interface Accessibility. 1.72 Spray Pond Piping Made From Fiberglass- 2 Yes 11/78 Reinforced Thermosetting Resin. (h V Amendment 14 IM1
ABWR am-c Standard Plant nry c TAllLE 1.8 20 RGs Applicable to AllWR (Contintied) AllWR Appl. Issued Appli. BjMg, Reculatory Guide Tith BE DMt CAbhI C.gmments 1.73 Qualification Tests of Electric Valve Ope. 0 1/74 Yes rators Installed inside the Containment of Nuclear Power Plants. 1.74 O'iality Assurance Terms and Definitions. Superceded See Table 17.0-1 1.75 PhysicalIndependence of Electric Systems, 2 9/78 Yes 1.76 Design 11 asis Tornado for Nuclear Power 0 4/74 Yes Plants 1.77 Assumptic,ns Used for Evaluatiag a Control 0 5/74 No PWR only Rod Ejection Accident for Pressurized Water Reactors. 1.78 Assumptions for Evaluating the liabitability 0 6/74 Yes of a Nuclear Power Plant Control Room Dur-Ing a Postulated Hazardous Chemical Re-lease. 1,79 PreoperationalTesting of Energency Core 1 9/75 No PWR only Cooling Systems for Pressurized Water Reac-tors. 1.81 Shared Emergency and Shutdown Electric Sys- 1 1/75 Yes tems for Multi-Unit Power Plants. 1.82 Water Sources for Long-Term Recirculation 1 11/85 Yes Cooling Following Loss-of-Coolant Accident, 1.83 In-Senice Inspection of Pressurized Water 1 7/75 No PWR only Reactor Steam Generator Tubes. 1.84 Design and Fabrication Code Case Acceptabi- 27 11/90 Yes lity, ASME Section Ill, Division 1. 1.85 Materials Code Case Acceptability, ASME 27 11/90 Yes Section III, Division 1. 1.86 Termination of Operating Licenses for Nu- 0 6/74 ---- Interface clear Reactors. O Amendment 17 1543
1 ABWR mac Sinndard Plant nry c TAllLE 1.8 20
< T V RGs Applicable to ABWR (Continued)
AllWR Appl. Issued Appil. RG No, Regulatory culde 11tle h I!Mit alm fnmments 1.137 Fuel-Oil Systems for Standby Diesel Genera- 1 10/79 Yes tors. 1.13S Laboratory Investigations of Soils for En. 0 4/78 Yes gineering Analysis and Design of Nuclear Power Plants. 1.139 Guidance for Residual Heat Removal. 0 5/78 Yes 1.140 Design, Testing, and Maintenance Criteria 1 10/79 Yes For Normal Ventilation Exhaust System Air Filtration and Absorption Units of Light. Water-Cooled Nuclear Power Plants. 1.141 Containment Isolation Provisions for Fluid 0 4/78 Yes Systems. 1.142 Safet) Related Concrete Structures for Nu- 1 11/81 Yes q clear Power Plants (Other Than Reactor Ves-Q sets and Containments). 1.143 Guidance for Radioactive Waste Management 1 10/79 Yes Systems, Structures, and Components Instal-led in Light Water-Cooled Nuclear Power Plants. 1.144 Auditing of Quality Assurance Programs Nu- Superceded See Table clear Power Plants. 17.0-1 1.145 Atmospheric Dispersion Models for Potential 1 12/82 Yes Accident Consequences Assessments at Nucle-at Power Plants. 1.146 Qualification of Quality Assurance Program Superceded See Table Audit Personnel for Nuclear Power Plants. 17.0-1 1.147 Inservice Inspection Code Case Acceptabi- 8 11/90 Yes lity-ASME Section XI, Division 1. 1.14S Functional Specification for Active Valve 0 4/81 Yes Assemblics in Systems important to Safety in Nuclear Power Plants. O l l ! Amendment 17 ILIR i
ABM uxsioarc Standard Plant nrv. c TAllLE 1.8 20 g RGs Applicable to AllWR (Continued) ABWit Appl. Issued Appil- I RG No. Reculatory colde Title Et1, Dair tat,te? comments l 1.149 Nuclear Power Plant Simulation Facilities 1 5/87 - Interface j for Use in Operator License Examinations. 1.150 Ultrasonic Testing of Reactor Vessel Welds 1 2/83 Yes During Presenice and Insenice Examina-tions. i 1.151 Instrument Sensing Lines. 0 8/83 Yes 1.152 Criteria for Programmable Digital Computer 0 11/85 Yes System Software in Safety Related Systems of Nuclear Power Plants. 1.153 Criteria for Power Instrumentation, and 0 12/85 Yes Control Portions of Safety Systems. 1.154 Format and Contents of Plant Specific 0 3/87 No PWR only Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors. 1.155 Station Blackout 0 8/88 Yes l 5.1 Serial Numbering of Fuel Assemblies for 0 12/72 Yes Light-Water Cooled Nuclear Power Plants. 5.7 Entry / Exit Control for Protected Areas, 1 5/80 Yes Vital Areas, and h1aterial Access. 5.12 General use of Locks in the Protection 0 11/73 Yes and Control of Facilities and Special Nuclear hieterials. 5.44 Perimeterintrusion Alarm Systems. 2 6/80 Yes l l Vital Area Acess Controls, Protection of 5.65 0 9/86 Yes Physical Security Equipment, and Key and Lock Controls. L 8.5 Criticality and Other Interior Evacuation 0 2/73 Yes Signals. 8.8 information Relevant to Ensuring That Occu- 3 6/78 Yes pational Radiation Exposures at Nuclear Power Stations Will Be As Low As is Reason. l ably Achievable Amendment 8 1kl9
u ABWR mswc
- Standard Plant ntw. c -
-l TABLE 1.8 21 (Continued) i' INDUSTRIAL CODES AND STANDARDS APPI.lCABLE TO ABWR -
Code or Standard .- Number Year Title l ASME B30.2 1983 Overhead and Gantry Crancs
' B30.9 -1984 Slings B30.10 1982 Ilooks B30.11 .1980 Monorails and Underhung Cranes B30.16 1931 Overhead floists B31,1 -1986 Power Piping B%.1 1986 Specificatior , w.avd Aluminum Alloy Storage Tanks 1
f u/~1 Amendment 17 1.8-57 I l-I~ l
ABWR ma c Standard Plant niv. c TABLE 1.8 21 (Continued) INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title AShfE N45.4 1972 Leakage-Rate Testing of Containment Structures for Nuclear Reactors NOA1 1983 Quality Assurance Program Requirements for Nuclear Facilities NOA-1A 1983 Addenda to ANSI / ash 1E NOA 1-1983 l l
^
l O Amendment 17 1 & 58 l
ABWR
^
useimac - Standard Plant RIV e - A. TABLE 1.8 21'(Continued) INDUSTRIAL CODES AND STANDARDS
=
APPLICABLETO ABWR
- Code or
' Standard Number Year Title AMWA D100. 1984 . Welded SteelTanks for Water Storage .I l
CMAA70 1983 Specification for Electric Overhead Traveling Cranes l l ICEA i P-46-426/IEEE' 1982 Ampacities Including Effect of Shield Losses for Single S-135 Conductor Solid. Dielectric Power Cable 15kV through
- 69kV
- P 54-440/ NEMA - 1987 Ampacities of Cables in Open-Top Cable Trays W 31 S-66-524/ NEMA 1982 Cross Linked Thermosetting Polyethylene Insulated Wire W7 and Cable for Transmission and Distributor of-Electrical Energy O
Amendment 12 1.8-61 l
ABWR mame Standard Plant REV,C TAHLE 1.8 21 (Continued) INDUSTRIAL CODES AND STANDARDS O APPLICABLE TO ABWR Code or Stan(ard Number Year Utle IEEE 279 1971 Criteria for Protection Systems for NGPS 30S 1980 Criteria for Class 1E Power Systems for NPGS
~
317 1983 Electrical Penetration Assemblics in Containtnent Structures for NPGS 323 1974 Qualifying Class 1E Equipment for NPGS 334 1974 Motors for NPGS, Type Tests of Continuous Duty citas 1E 338 1977 Criteria for the Periodic Testing of NPGS Safety Systems 344 1987 Recommended Practices for Seistnic Qualifications of Class 1E Equipment for NPGS 379 1977 Standard Application of the Single Failure Criterion to NPGS Safety Systems 382 1985 Qualification of Actuators for Power Operated Valve Assemblies with Safety-Related Functions for NPP 383 1974 Type Test of Class 1E Cables; Field Splices and Connections for NPGS 384 1981 Criteria for Independence of Class IE Equipment and Circuits 387 1984 Criteria for Diesel Generator Units Applied as Standby Power Supplies for NPGS 450 1987 Practice for Maintenance, Testing, and Replacement of Large lead Storage Batteries for Generating Stations and Substations 484 1987 Recommended Practice for the Installation Design and Installation of Large Lead Storage Batteries for NPGS 1 l Amendment 17 l842 I I
1 I l
-ABWR maue l Sandard Plant urva l n Table 1.8-22 EXPEllIENCE INFOllMATION API'LICAllLETO AllWit 1
'IYPE: GENERIC LE'ITERS j issue-M liate Illig Comment 80-06 4/25/80 Clarification of NRC Requirement for Emergency Response Facilities at Each Site 80 30 12/15/80 Periodic Updating of Final Safety Analysis Interface Reports (FSARs) 80-31 12/22/S0 Control of llemy Loads 81 03 2/26/81 Implementation of NUREG-0313m ? lev.1 81-(M 2/25/81 Emergency Procedures and Training for Station Interface Blackout Events 81 07 2/3/S1 Control of 11 cavy Loads 81 10 2/IS/81 Post TMI Requirements for the Emergency Operations faciSty (3
Ij s 81-11 2/22/S1 Error in NUREG4bl9 81 20 4/1/81 Safety Concerns Associated with Pipe Breaks in the BWR Scram System 81 37 12/29/81 ODYN Code Reanalysis Requirements 81 38 11/10/81 Storage of Low-Level Radioactive Wastes Interface at Power Reactor Sites 824D 4/20/82 Environmental Qualification of Safety-Relatt d Electrical Equipment 82-21 10/6/82 Technical Specifications for Fire Protection Audits Interface 82 22 10/30/82 Inconsistency between Requirements of 10 CFR 73,40(d) and Standard Technical Specifications for Performing Audits of Safeguard Contingency Plans 82 27 11/15/82 TransmitIal of NUREG-0763," Guidelines for Confirmatory in Plant Tests of Safety Relief Valve Discharges for BWR Piants," and NUREG.0783," Suppression Pool Temperature Limits for BWR Containments." 52 33 12/17/82 Supplement I to NUREG-0737 I '%J 1 8-67 Amendment 12
ABWR mac Standard Plant . REY. C Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) O TYPE: GENERIC LETTERS issue h Dalt. Et!t Comment 82-39 12/22/82 Problems with the Su5mittals of 10 CFR 73.21 Interface Safeguards Informatin Licensing Review 83-O' 2/83 Safety Evaluation of' Emergency Procedure Guidelines, Interface Revision 2, NEDO-24934, June 1982 l 83-07 2/16/83 The Nuclear Waste Policy Act of 1982 Interface l 83-13 3/2/83 Clarification of Survei!!ance Requirements for HEPA Filters and Charcoal Absorber Units in Stendard Technical Specifications on ESF Cleanup Systems 83-28 7/8/83 Required Actions Based on Generic implications of Salem ATWS Events 83-33 10/19/83 NRC Positions on Certain Requirements of Appendix Interface R to 10 CFR 50 8&l5 7/2/84 Proposed Staff Actions to improse and Maintain Diesel Generator Reliability 84 23 10/26/84 Reactor Vessel / Water LevelInsrumentation in BWRs l 8541 1/9/85 Fire Protection Policy Steering Committee Report 86-02 1/23/86 Technical Resolution of Generic Issue B-19 Ther 'M Hyoraulic Stability 86-10 4/24/86 Implementation of Fire Pre: a eluirements 87-06 'l/13/87 Periodic Verification of Leak Tight integ-ity of Interface Pressure Isolation Valves 87-09 6/&87 Sections 3.0 and 4.0 of the Standard TechC al-Specifications (STS) on the Applicability of Limiting Conditions for Operations and Surveillance Requirements 88-01 1/25/88 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping 88-02 1/20/88 Integrated Safety Assessment Program 11 (ISAP II) O Amendment 17 l848
ABWR DM MAC Standard Plant RN C p v - Table 1,8 22 - EXPERIENCE INFORMATION APPLICABLE TO AllWR (Continued)
- TYPE: NUREG
-Issue b'E 11111. Etig Comment 03d 6/88- Technical Report on hiaterial Selection and Processing l Rev. 2 Guidelines for BWR Coolant Pressure Boundary Piping l l
0371 10/78- Task Action Plans for Generic Activities Category A 0471 6/78 Generic Task Problem
Description:
Category B, C & D Tasks 0578 9/80 Performance Testing of BWR and PWR Relief and Safety Valves. 0588 12/79 Interim Staff Position On Environmental Oualification of Safety-Related Electrical Equipment 0619 4/80 BWR Feedwater Noz2le and Control Rod Drive Return Line Nozzle Cracking 0626 1/80 Generic Evaluation of Feedwater Transients and Small Break LOCA in GE Designed Operating Plants and Near-Term Operating License Applicat ions 0660 5/80 NRC Action Plan Developed as a Result of the Thil 2 Accident 0661 8/82 Safety Evaluation Report-hlarx ! Containment Supp.1 Long Term Program. Resolution of Generi: Technical Actisity A 7 0110 6/81 Licensing Requirements for Pending Applications for Rev.1 Construction Permits and hianufacturing License.
'0737 12/82 Clarification of Thil Action Plan Requirements Supp.1 0744 -10/82 Resolution of the Task A-11 Reactor Vessel htaterials Rev.1 Toughness Safety issue 0808 8/81 hlark 11 Containment Program Load Evaluation and Acceptance Criteria 0813 9/81 Draft Environmental Statement Related to the Operation of Calloway Plant, Unit No.1 0977 3/83 NRC Fact-Finding Task Force Report on the ATWS Events at the Salem Nucicar Generating Station, Unit 1, on February 22 and 25,1983 Amendment 17 I 8~89
. . . _ _ _ . - - . _ - . . . - _ . _ _ _ - . - _ . _ - ~ , _ . _ - _ . _ . . . _ _ _ , - . - . - . _ - - _ . _ _ . _ . - , . _ . ~ _
ABWR uume Standard Plant RIN. C Table 1.8 22 EXPERIENCE INFORMATION APPLICABLE TO ABWR (Continued) O TYPE: NUREG issue F9. Half. Elle Comment 1150 6/89 Sewre Accident Risks: An assessment for Five U.S. Nuclear Power Plants, Vol.1 & 2. 1161 5/80 Recommended Resisions to USNRC-Seismic Design Subsection Criteria 19B.2.27 1174 5/89 Evaluation of Systems Interactions in Nuclear Subsection Power Platas 19B.23 1212 6/86 Status of Maintenance in the US Nuclear Power Industry 19S5 Vol.1,2 1216 8/86 Safety Evaluation PP2 Related to Operability and Reliability of Emergency Diesel Generators 1217 4/88 Evaluation of Safety implications of Control Systems Subsection in LWR Nuclear Power Plants-Technical Findings 19B.2.5 Related to USI A-47 1218 4/S8 Regulatory Analysis for Resolution of USI A-47 Subsection 19B.2.5 1229 8/89 Regulatory Analysis for Resolution of USl A-17 Subsection 19B.23 & 19B2.27 1232 9/89 Regulatory Analysis for USl A-40 Subsection 19B.2.27 l 1273 4/88 Containment Integrity Check-Technical Finds Regulatory Ana'ysis 1289 11/88 Regulatory and Backfit Analysis: Unresolved Safety Subsection issue A-45, Shutdown Decay Heat Removal Requirements 19B.2.29 1296 2/88 Pier Review of High Level Nuclear Waste 1341 5/89 Regulatory Analysis for Resolution of Generic Issue 115, Enhancement 1353 4/89 Regulatory Analysis for the Resolution of Generic Subsection Issue 82,"Beyond Design Basis Accidents in Spent 19B.2.14 Fuel Pools" 1370 9/89 Resolution of USl A-48 Subsection 198.2.6 1.8-90 Amendment 17
23A6100AC Standard Plani RM C ,
, Table 1.b 22 i 3
EXPERIENCEINFORMATrON APPLICAHLETO AHWR (Continued) TYPE:NUREG
- lune No. Dalt. 3 111 Comment CR 3922 1/85 Survey and Evaluation of System Interaction Events Subsection and Sources Vol.1, 19D.23 CR.4261 3/86 Assessment of Systems Interactions in Nuclear Power Subsection Plants 19B.2.3 CP. 4262 $/85 Effects o' *cattol System Failures on Transients, i Accidents at a GE BWR Vol.1 at.. 2 CR-4387 12/85 Effects of Control System Failures on Transient and Accidents and Core-Melt Frequencies at a GE BWR CR-4470 5/86 Survey and Evaluation of Vital Instrumentation and Control Power Supply Events CR.5055 $/88 Atmospheric Diffusion for Control"oom liabitability Subsection
. Assessment 198.231 CR 5088 1/89 Fire Risk Scoping Study: Investigation of Nuclear O. Power Plant Fire Risk, including Predously Unaddressed issues.
CR 5112 3/89 Evel::tien of Balling Water Reactor Weer - Level Subsection Sensing line Break and Single Failure 198.2.16 CR 5230 4/89 Shutdown Decay 11 eat Removal Analysis: Plant Case Subsection - Studies and SpecialIssues 19B.2.29 CR SM7 6/89 Recommendations for Resolution of Public Comments Subsection on USl A 40 19B.2.27 CR 5458 12/89 Value Impact Assess for Candidate Operating Procedure Upgrade Program O
' Amendment 17 1891 - ~ . - _ _ - . - - , . . - - .- .-_ ,._ _ ;_. ,_ ._ . . -.- _ ... - -..._ _ ._. _ _ , _ -- . _ __ ._ _ _ _ . ~ _- _ ~ . _. _ . _ .~. ,
_. ~ _ _ . _ __.. -.. _ _ _ _ _ _ _ _ _ _ . _ _ _ ABWR msimac ' Standard Plant niv.c APPENDlX 1A TAHLE OF CONTENTS Section Htle Pagg 1A RESPONSE TO TM1 RELATED MA'1TERS 1A.1 INTMODUCTION 1A11 1 A.2 NHC POSITIONS / RESPONSES 1A.21 1A.2.1 Short Term Accident Analysis Procedure Revision [1.C.1(3)] 1A.21 1A.2.2 Control Room Desigt. Revisions / Guidelines and Requirements (l.D.1(1)) 1A.22 1A.23 Control Room Design Plant Safety Parameter Display Console (l.D.2) 1A.22 1A.2.4 Scope of Test Program Preoperational 0 1A.2.5 and lower Power Testing [1.G.1] Reactor Coolant System Vents [!I.B.1] u.2 2 1A.2 3 1A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post Accident Operation [11.11.2) 1A.25 l 1A.2.7 Post Accident samplinj;[!!.B3] 1A.27 1A.2.8 ' Rule Makin3 Proceeding or Degrading Core Accidents l11.11.8) 1A.2-8 1A.2.9 Coolant System Valves Testing Requirements [lI.D 1] 1A.2-8 1A.2.10 Relief and Safel) Valve Position Indication [lI.DJ) 1A.24 1A.2.11 Systems Reliability lll.ES.2) 1A,2 8 1A.2.12 Coordinated Study of Shutdown IIcat Removal Requirements ((I.E33] 1A.2-8 1A.2.13 Containment Design-Dedicated Penetration (II.E.4.1] 1A.2 8.1 l 1A li Amendment 17 L
.-m.- e- .>. ,,% ,. , . . - - ,--w..v-, -%.-_.----%-,e,-.v,, - ,.--cy.. .--,---,.,-w. _ , , . . .- - . , + -. _%- , % , , .,,,r-+: -,-r.wre,, . . --g.---i-m
ABWR mome S1111dard Plant RIY C APPENDIX 1A g. TAllLE OF CONTENTS (Continued) Eccitan 11 tic Page 1A.2.14 Containment Design Isolation l Dependability [lI.E.4.2) 1A.2 8.1 1 1A.2.15 Additional Accident htonitoring instrumentation [II.F.1(1)] 1A.210 1A.2.16 Identification of and Recovery From Coriditions leading to inadequate Core Cooling [II.F.2] 1A.210 1A.2.17 Instruments for Monk tring Accident Conditions lll.F3] 1A.211 1A.2.18 Safety Related Vahc Position Indication [II.K.1($)) 1A.2 11 1A.2.19 Review and hiodify Procedres for Removing Safety Related Systems From Service [lI.K.1(10)] 1AT 11 1A.2.20 Describe Automatic and h1anual Actions 0 for Proper Functioning of Auxillary 11 eat Removal Systems When FW System not Operable [lI.K.1(22)] 1A.211 1A.2.21 Describe Uses and Types of RV Level Indication for Automatic and hianual laitiation of Safety Systems III.K.1(23)) 1^.2-12 1A.2.21.1 Faibic of PORV or Safety Valve to Close jl!.K3(3)] 1A.213 1 1A.2.22 Separation ofIIPCI and RCIC System istlation 14vels [lI.K3(13)] 1A.213 1A.2.23 hiodify Break Detection Logic to Prevent Spurious isolation of IlPCI and RCIC Systems [lI.K3(15)] 1A.2 13 1/ 1.24 Reduction of Challenges and Failures of Relief Valves Feasibility Study and System hkx!ification [ll.K3(16)] 1A.214 1A iii Amendment 17
46 W Mf9EIM EDN ABWR we Sinndard I'lant . m ._ , , L' APPENDIX 1A TAllLE OF CONTENTS (Continued) Section llllt Eagt 1A.2.25 Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes [lI.K3(17)] 1A.2 15 1A.2.26 hiodification of Automatic Depressuri-ration System legic Feasibility for increared Diversity for Some Event Sequences [lI.K3(18)] 1A.2 15 1 A.2.27 Restart of Core Spray and LPCI Systems on Low L.cret Design and hiodification [lI.K3(21)] 1 A.2 16 1 A.2.28 Automatic Switch Over of Reactor Core Imlation Cmling System Suction - Verify Procedures and biodify Det.ign i [lI.K3(22)] 1A.2-16 l 1A.2.29 Confirm Adequacy of Space Cooling for liigh Picssure Coolant injection and Reactor Core Isolation Cooling Systems i [lI.K3(24)] 1A.2 17 l 1A.230 Effect of less of Alternating-Current i Power on Pt.np Seals [lI.K3(25)) 1A.2 18 1 1A.231 Verify Oualification of Accumulators on Autorr.atic Depressuri7ation System Valves i L" K3(28)) 1A.2 18.1 l 1A.232 Revised Small-Dreak less Of Coolant-Accident hiethods To Show Compliance With 10 CFR PART 50, Appendiz K ((I.K3(30)] 1A.2-19 l 1A.233 Plant Specihc Calculations to Show Compliance With 10 CFR Part 50 46 Ill.K3(31)) 1^.2 19 1A.233.1 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failute llI.K3 (44)) 1A.219 O 1 A iV Amendment 17
ABWR mome Standard Plant nuv. c APPENDIX 1A h TABLE OF CONTENTS (Continued) Srction Elle Ease 1A.233.2 Evaluate Depressuriation otber than Full ADS [lI.lL3 (45)] 1A.219a 1A.2333 Responding to Michelson Concerns a [!!.K3 (46)] 1A.219a j 1A.234 Primary Coolant Sources Outside Contain-ment Structure [Ill.D.1.1(1)] 1A.219a 1A.235 In Plant Radiation Monitoring [Ill.D33(3)) 1A.2 21 1A.236 Control Room liabitability [III.D3.4(1)] 1A.2 21 1 A.3 INTERFACFS 1A3.1 Emergency Procedures and Emergency Procedures Training Program 1A31 1A 3.2 Review and Modify Procedures for Removing Safety Related Systems From Sevice 1A 3-1 1A33 Inplant Radiation Monitoring 1A3-1 1 A.4 REFERENCES 1A.41 1AA ATTACIIMENT A TO APPENDIX 1 A: PLANT SillELDING TO [ PROVIDE ACCESS TO VITAL AREAS AhD PROTECT SAFE'1T EQUIPMEhT FOR POST. ACCIDEST OPERATION (11.11.21 1AA.1 INTRODtTTION 1 AA.1 1 1 AA.2
SUMMARY
OF SillEl. DING DESIGN 1 AA.2 1 REVIEW 1AA3 CONTAINh1ENT DESCRilTION AND POST ACC[ DENT OPER ATIONS 1 AAJ 1 1AA4 DESIGN REVIEW IMSES 1 AA.4 1 O i 1A v Amendment 17 l
ABWR mame S1Pndard Plant nix e i . \ + k_) SECTION 1A.2 CONTENTS Scotiott Title l'11gg 1A.2.1 Short Term Accident Analysis Procedure Revision [1.C.1(3)) 1A.h1 1A.2.2 Control Room Design Revisions / Guidelines and Requirements [1.D.1(1)] 1A.2 2 1A.23 Control Room Design Plant Safety Parameter Display Console [1.D.2) 1A.2-2 1A.2.4 Scope of1..t Progrt 1 Picoperational and Lowei Power Testing [1.G.1) 1A.2 2 1A.2.5 Reactor Coolant System Vents [lI.B.1] 1A.2 3 1A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment fo' Post Accident Operation [lI.fl.2] 1A.25 !q 1A.2.7 Post Accident Sampling [11.!!3] 1A.2 7 1A.2.8 Rule Making Proceeding or Degrading Core Accidents [lI.B.8) 1A.2 8 1A.2,9 Coolant System Valves Testing Requirements [II.D.11 1A.2-8 1A.2.10 Relief and Safety Valve Position Indication llI.D3) 1A.2 8 1A.2.11 Systems Reliability ((I.E3.2l 1A.2 8 1 A.2.12 Coordinated Study of Shutdown lleat Removal Requirements [lI.E33] 1A.2-8 1A.2.13 Containment Design Dedicated Pcnetration [lI.E.4.1) 1A.2 8.1 1A.2.14 Containment Design-Isolation Dependability [lI.E.4.2] 1A.2 8.1 l n
'N 1A.2 ii Amendment 17
ABWR 2nf.100AC Standard PlanL. '*" O SECTION 1A.2 CONTENTS (Continued) Section 11tle PJigt 1A.2.15 Additional Accident Monitoring Instrumentation [II.f.1(1)] 1A.2-10 1A.2.16 Identification of and Recovery from Condillora leading to inadequate Core Cooling [ll.F.2) 1A.210 1A.2.1"f Instruments for monitoring Accident Conditions [lI.F3) 1A.2 11 1A.2.' 3 Safety Related Valve Position indication [lI.K.1(5)) 1A.211 1A.2.19 Review and Modify Procedures for Removing Safety-Related Systems From 1A.2 11 Service [lI.K.1(1)] 1A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems When FW System not Operable [II.K.1(22)] 1A.2-11 1A.2.21 Describe Uses and Types of RPV1.crel Indication for Automatic and Manual Initiation of Safety Systems [lI.K.1(23)] 1A.212 1A.2.21.1 Failure of PORV or Safety to Close [II.K3(3)) 1A.213 1A.2.22 Separation of HPCI and RCIC system Initiation Levels [II.K3(13)) 1A.2-13 1A.2.23 Modify Break Detection Logic to Prevent Spurious isolation of HPCI and RCIC Systems [II.K.2(13)) 1A.2 13 1A.2.24 Reduction of Challenges and Failures of Relief Valves Feasibility Study and System Modifications [lI.K3(16)) 1A.214 O 1A 2.iti Amendment 17 \.
ABWR nume Standard Plant nrv c n 1.J' SECTION 1A.2 CONTENTS (Continued) Section Jhlt hige 1A.2.25 Report on Outages of Emergency Core-Cooling Systems Licensec Report and Proposed Technical Specification Changen [lI.K3(17)) 1A.2 15 1A.2.26 Modification of Automatic Depressuri-ration System Logic Peasibility for increased Diversity for Zone Event Sequences [lI.K3(18)] 1A.215 1A.2.27 Rests 1 Core Spray and LPCI Systems on Low Level Design and Modifications [lI.K3(21)] 1A.2 16 1A.2.28 Automatic Switch Over of Reactor Core Isolation Cooling System Suction-e Verify Procedures and Modify Design ( [!!.K3(22)] 1A.2 16 1A.2 29 Confirm Adequacy of Space Coolant for liigh Pressure Coolant injection and Reactor Core Isolation Cooling Systems llI.K3(24)) 1A.2 17 1A.230 Effect of less of Alternating Current Power on Pump Seals [lI.K3(25)) 1 A.218 { 1A.231 Verify Qualification of Accumulators on Automatic Depressurization System Valves [lI.K3(23)] 1 A.2 18.1 1A.232 Revised Small Break Las of Coolant-Accident Methods To Show Compliance With 10 CFR PART 50, Appendix K [lI.K3(30)) 1A.2 19 l 1A.233 Plant Specific Calculations to Show , Compliance With 10 CFR Part 50.46 [lI.K3(31)] 1A.219 l l 1A.233.1 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure (II.K3(44)) 1A.2 19 1 (% l 1 A.2-tv j Arnendment 17
ABWR mmme iuv. c Standard _ Plant . O SECTION 1A.2 CONTENTS (Continued) Settlan Iltle lhige 1A.233.2 Evaluate Depressuriza:!on other than Full ADS [ll.K3(45)) 1A.219a 1A.2333 Responding to hilchelson Concerns lll.K3(46)) 1A.219a 1A.234 Primary Coolant Sources Outside Contain-hirut Structure [Ill.D.1.1(1)) 1A.219a 235 in Plant Radision hionitoring [Ill.D33(3)) 1A.2 21 236 Control Room liabitability [Ill.D3.4(1)] 1A.2 21 TAllLES Inhlt 11tle lhlec h 1A.2 1 Response to Questions Posed by hir. hiichelson 1A.2 22 [lI.K3(46)) l l l l l O 1A2v Amendment 17
ABWR name ' Standard Plant nov c 1A e .2.13 'r ii Containment tii s 4 ii Design. Dedicated O ! NRC Position For plant designs with external hydrogen recombiners, prmide redundant dedicated contain-ment penetrations so that, assuming a single failure, 4 the recombiner systems can be connected to the con-tainment atmosphere. Respons4 A flammabili:y control system (FCS T49) is provided to controiInc concentration of oxygen in the primary containment. The FCS utilites two permanently insta' led recombiness located in secondary mntainment. The FCS is operable in the event of a single active failure. The FCS is described in Subsection 6.2.5. I A.2.14 Containment Design. Isolation Dependability [lI.E.4.2) -i NRC Position O (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-verslty in the parameters sensed for the in). > tlation of ccatainment isolation). , (2) All plant personnel shall give careful consid-eration to the definition of essential and non- - essential systems, identify each system l determined to be non essential, describe the basis for selection of each essential system, modify their containment isolation designs , accordiagly, and report the results of the reevaluation to the NRC. (3) All noncssential systems shall be automatically
- isolated by the containment isolation signal.
'l (4) The design of control systems for automatic
- containment isolation valves shall be such that reseting the isolation signal will not result in the l
~
. Amendment 17 1 A 2-81
.y,..m_ .. _ , . _ , . . - - . , _ , , , . , , , , , . .,w , . , . . , , , _ _ ,,,,y ..,, , , _ , ..y.
1 ABWR nwmac Standard Plant nrv c automatic reopening of containment isolation (5) The ABWR Standard Plant design is consistant valves. Reopening of containment isolation with this position. vahes shall requhe deliberate operator action. (6) All ABWR containment purge valves meet the (5) The containment setpoint pressure that initiates criteria provided in llTP CS116-4. The main 22-containment isolation for non essen'.ial peectra- purge vahes are fail-closed and are maintained l closed through power operation as defined in the tions must be reduced to the min' mum ccmpat-ible with normal oper ating condit,ons. plant technical specifications. All purge and vent valves are remote pneumatically operated, fail (6) Containment purge vahrs that do not satisfy the closed and receive containment isolation signals, operability criteria set forth in Branch Technical Certain vent valves can be opened manually in l , Position CSB 6-4 or the Staff !nterim Position of the presents of an isolation signal, to permit l-October 23,1979 must be scaled closed as de- venting through the SGTS. fined in SRP 6.2.4, Item II.6.f during operational l conditions 1,2,3, and 4. Furthermore, these (7)In the ABWR design, the containment purge and valves must be verified to be closed at least vent isolation valves will be automatically isolated Ji ! every 31 days. on high radiation levels detected in the reactor { building !!VAC alt exhaust or in the fuel (7) Containment purge and vent isolation valves handling area air exhaust. must close on a high radiation signal.
Response
(1) The isolation provisions described in the Stan-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. Ilowever, 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. (4) Control systems for automatic containment iso-nation valves are designed in accordance with this position for the ABWR Standard Plant Design. O 1A.2-9 Ame ndment 17
ABWR nasimxe Standard Plant nvc switchover is implemented, licensees should verify O that c' car and cogent procedures exist for the manual switchover of the RCIC system suction from the con-densate storage tank to the suppression pool. Respomme
. The RCIC system prosided in the ABWR Stan-dard Plant includes an automat's switchover feature which will change the pump suction source from the l condensate storage pool to the suppression pool.
The safety grade switchover will automatically occur upon receipt of a low level signal from the condensate storage pool or a high level signal from the suppression pool. l See Subsection 7.3.1.1.1.3 for additior.nl infor-I mation. 1A.2.29 Cordirm Adequacyof Space Cooling for Hiph Pressure Coolant ] Injection and F.eactor Core Isolation Cooling Systehts [II.K.3(24)] ' I NRC Position Long term operation of the reactor core isola-O tion cooling (RCIC) and high pressure coolant injec-tion (IIPCI) systems may require space cooling to i maintain the pump-room temperatures within allow-able limits. Licensees should verify the acceptability-l= of the consequeraes of a complete loss of alternat. ing current power. The RCIC and itPCI systems should be designed to withstand a complete lots of t offsite alternating-current power to their support sys-l tems, including coolers, for at least 2 hours. [. Response - The ABWR high pressure core flooder (llPCF) and the reactor core isolation cooling (RCIC)- systems are provided space cooling via individual room safety grade air coditioning systems (See l~ Subsection 9.4.5). If all offsite power is lost, space l cooling for the llPCF and RCIC system equipment l- would not be lost because the motor power supply far each system is from seperate essential power supplies. l O Amendment 15 ~ 1 A 217 i- _._.,1._...._....-,_--......_.____._.c._._.,_..__._..___._.._._.__.
ABWR micoac uvc Standard Plant _ 1AJM Effect of loss of Alternating- The RIP motors are designed and will be plant Current Power on Pump Seals [II,K.3(25)) tested to not be damaged in the stopped hot 6tandby condition indefinitely with RCW cooling NRC Posillon available. 1 ne licensees should determine, on a plant (b) RhtP Faihttg - Subsection 5.4.1.3.2 describes the specific basis, by analysis or experiment, the RhiP operation. Since the RIP and rootor have consequences of a loss of cooling water to the no seals, the water in the RIP motor reactor recirculation pump seal coolers. The pump communicates directly with the reactor water at seals should be designed to withstand a complete loss the aame pre 6sure but at mueh Iower of ahernating-current (ac) power for at least 2 hours. temperature. There is no possibility of this Adequacy of the seal design should be demonstrated. water escaping from the coolaut pressure boundary such as in conventional pumps which Response include seals. The ABWR design features internal recircula. The RhtP water is supplied from the CRD tion pumps (RIP) which do not require shaft seals, system. The CRD pumps will stop temporarily During a LOPP, the RIPS shutdown automatically during a LOPP, which will cause the normal 15 but there are no shaft seals which require cooling gram /sec. (1/8 ppm) RhtP flow to each RIP to water restoration. temporarily stop. The CRD pumps are subsequently restarted automatically, after A plant AC power failure would temporarily several minutes time delay, powered by onsite disrupt the operation of the reactor recirculation power sources and the Rh1P water will restart. subsystems but their failure would not generate a This temporary interruption of RhlP flow will : LOCA as the following describes. not initiate a LOCA. The only effect of loosing k (a) IB1C Failure Subsection 5.4.1.3.1 descriSes the the Rh1P flow temporarily to the RIPS, from a Rh1C operation during normal running or LOPP, is that it will allow reactor stopped condition. This operation assumes the contamination, by diffusion, to enter the RIP reactor building cooling water (RCW) is in motor during the RhlP flow interruption. operation continuously during these operating or stopped conditions. Normal LOCA and (c) RN1ISS Failutt Subsection 5.4.1.3.3 describes LOPP operation of the RCW are described in the operation of the Rh11SE. which is used only Subsection 9.2.11.2. during plant shutdown and RIP maintenance. The power source for the inflatable sealis a e A loss of AC power or loss of Preferred power portable air operated water pump which is (LOPP) will stop all RIPS. The LOPP will also moved from RIP to RIP. A LOPP would g temporarily stop the RCW and RSW pumps. tnerefore not cause a direct loss of Rh11SS The onsite emergency power sources will pressure since (1) the plant air system tas a automatically restart the RCW/RSW pumps finite passive storage capacity in the air receivers which will restore cooling for the stopped RIP and (2) the RhtlSS air operated pump only motors. The RCW primary contalnment operates when the RhilSS pressure drops below isolation vahes will not close on LOPP (only on a preset value. LOCA). The Rh11SS is a secondary seal. Even with a , The RhfC system for each RIP includes a motor long time Rh11SS failure RIP maintenance, the cooling outlet temperature detector TE 301 passive backscat seal of the RIP shaft on the which will automatically runback indhidual RIPS stretch tube will preclude draining the reactor, to minimum speed on high coolant temperature and prevent motor damage which could initiate aLOCA. O 1 A.2.lB r .nendment 17 1__ _ _ _ ___ _
ABWR 2mmac Sinndard Pllint ntx c 1 A.2.31 Study and Verify Qualification of The accumulators are designed to Provide two (m") Accurnulators on Autornatic Depressuri-ration Systern Valtes [lI.K.3(28)) ADS actuations at 70% of dr>well pressure, which is equivalent to 4 to 5 actuations at atmospherie pressure. The ADS valves are designed to operate at NRC Position 709 of drywell design pressure because that is the maximum pressure for which reactor depressuri-Safety analysis reports claim that air or nitrogen ration through the ADS vahes is required. The accumulators for the automatic depressuritation greater dr>well design pressures are associated only system (ADS) valves are presided with sufficient with short duration primary system blowdown in the capacity to cycle the valves open five times at design drywellimmediately following a large rupture for pressures. GE has also stated that the emergency which ADS operation is not required. For large core cooling (ECC) systems are designed to with- breaks which result in higher drywell pressure, stand a hostile environment and still perform their sufficient reactor depressurlution occurs due to the function for 1(U days following an accident. Licensee break to preclude the need for ADS. One ADS should verify that the accumulators on the ADS actuction at 70% of drywell pressure is sufficient to valves meet these requirements, even considering depressatire the reactor and allow inventory makeup normal leakage. If this cannot be demov,trated, the by the low pressure ECC systems. Iloweser, for Licensee must show that the accumulator design is conservathm, the accumulators are sired to allow 2 still acceptable, actuators at 70% of drywell pressure, See Sub-section (n8.1 for a description of the ADS N 7 Response pneumatic supply. The accumulators for the ADS valves are siicd to provide two operating cycles at 709 of drywell design pressure. This cyclic capability is validated during preoperational testing at the rtation. The accumulators are safety grade comp (ments. The 100-day, post accident functional openbility requirement is met through consenative design and redundancy; eight ADS valves are provided with code-qualified accumulators and seismic Category I piping within primary containment. Two redundant 7 day supplies of bottled air are available to compensate for leakage during long term usage, with replacement capebility being provided for the remainder of the postulated accident to assure system functional operability. A maximum of Ihree of eight ADS valves need function to meet short-term demands (see Subsection 19.3.1.3.1) and the functional operability of only one ADS vahe will fulfilllonger actm needs. Loss of pneumatic supply pressure to the ADS SRV accumulator is alarmed to proside the reactor operator with indication of the failure of any of the redundant systems under hostle j emironmental condition. l The llWR Owners' Gror ponsored an l evaluation of the adequacy of the ADS configurations. Evaluation results are summarized in ( i the following paragraph.
/~'N U
Amendment 17 1A 2B t
--v v ,,- n
ABWR n m mAc Standard Plant nrv c O V 1A.2.32 Revised Small.Ilreak less of Coolant Accident Methods to Show 1A.2.33 Plant.Specille Calculations to Show Compliance with 10 CFR Patt 50.46 l Compliance with 10 CFR PART 50, [lI,K.3(31)) Appendix K [lI.K.3(30)) NRC Position NRC Position Plant specific calculations using NRC approved The analysis methods used by nuclear steam models for small break loss of collant accidents supply system (NSSS) vendors and/or fuel suppliers (LOCAs) as described in item II.K.3.30 to show for small break loss of coolant accident (LOCA) compliance with 10 CFR 50.46 should be submitted analysis for compliance with Appendix K to 10 CFR for NRC approval by alllicensees. Part 50 r.hould be revised, documented, and submit-ted for NRC approval. The revisions should account Response for comparisons with experimental data, including data from the LOFT Tc6t and Semir.cale Test facili. The AllWR standard safety small break LOCA ties. calculations are discussed in Subsection 633.7. Response The references listed in Subsection 63.6 desciibe the currently approved Appendix K methodology GE has evaluated the NRC request requiring used to perform these calculatioro Compliance with that the BWR small break LOCA analysis methods 10CFR$0.46 bas previously been established for that are to be demonstrated to be in compliance with Ap- methodology. pendix K to 10 CFR Su or that they be brought into compliance by analysis methods changes. The spe- Since, as noted in the previous item (IA.232), no cific NRC concerns are contained in NUREG426, model changes are needed to satisfy NUREG 0737, q Appendix F. The specific NRC concerns identified item II.K3(30), there is no need to revise the calcu. Q in Subsection 4.2.10 of NUREG-0626 (Appendix F) relate to the following: counter current flow limiting lations presented in Subsection 633.7. (CCFL) effects, core bypass modeling pressure vari- 1 A.2.33.1 Evaluation oI Antieipated l ation in the reactor pressure vessel, integral ex. Transients with Single Failure to Verify No perimental verification, quantification of unceitain- Fuel Failure [lI,K.3 (44)] ties in predictions, the recirculation line inventory modeling, and the homogeneous / equilibrium model. NRC Position The response to the NRC small break model For anticipated transients combined with the concerns was provided at a meeting between the worst single failure and assurning proper operator NRC and GE on June 18,1981. Information pro- actions, licensees should demonstrate that the core vided at this meeting showed that, based on the remains covered or provide analysis to show that no TLTA small break test results and sensitivity studies, significant fuel damage results from core uncovery. the existing GE small break LOCA model already Transients which in a stuck-open relief valve should satisfies the concerns of NUREG 0626 and is in be included in this category. The results of the evah compliance with 10 CFR 50, Appendix K. There- uation are duc January 1,1981. fore, the GE modelis acceptable relative to the con-cerns of item II.K.3(30), and no model changes need Response be made to satisfy this item. GE and the BWR Ownerr.' Group have ron-Documentation of the information provided at cluded, based on a representatise llWR/6 plant the June 18,1981 meeting was provided sia the letter study, that all anticipated transients in Regulatory from R.11. Buchholz (GE) to D. G. Eisenhut Guide 1.70, Revision 3, combined with the worst (NRC), dated June 26,1981. O Amendmen 13 tA L19
I ABWR mmec Standard Plant niw c single f ailure, the reactor core remains covered with Response water until stable conditions are achieved. Further-more, even with more degraded conditions invohing All of the generic l ebruary 21,1980 GE re-a stuck-open selief valve in addition to the worst sponses were reviewed and updated for the AllWR transient (loss of feedwater) and worst single failure Standard Plant. The specific responses are provided (of high pressure core spray), studies show in Table 1A.2-1. (NEDO 24708, htarch 31,1980) that the core re-mains covered and adequate core cooling is available IA.2.34 l'rlmary Coolant Sourecs Outside during the whole course of the transient. The con- Containtnent Structure [Ill.D.I.l(l)) clusion is applicable to the ADWR. Since the ABWR has more high pressure male up systems NRC Position (211PCFs and 1 RCIC), the core covering is further assured. Applicants shallimplement a program to reduce leakage from systems outside containment that Other discussions of transients with single fail- would or could contain highly radioacthe fluids dur-ure is presented in the response to NRC Question ing a serious transient or accident to as low.as prac-440.111. tical levels. This program shall include the following: 1A.2.33.2 Evaluate Depressurization other (1) Immediate leat reduction than Full ADS [lI.K.3 (45)) (a) Implement all practicalleak reduction mea-NRC position sures for all systems that could carry radioac-tive fluid outside of containment. Provide an evaluation of depressuriz.ation meth-ods other tlan by full actuation of the automatic de- (b) hicature actualleakage rates with sptems pressurization system, that would reduce the possi- in operation and report them to the NRC. bility of exceeding vessel integrity limits during rapid cooldown. (Applicable to BWR*$ only) (2) Continuing Leal Reduction- establish and im. plement a program of preventive maintenance to Response reduce leakage to as low a&practicallevels. This This response is provided in Subsection 19A.2.11 1A.2.33.3 R e s p o n d I n g io M ie h eI s o n Concerns [I1.ls.3 (46)] NRC Position General Electric should provide a response to the hilchelson concerns as they relate to boiling water reactors. Garification General Electric provided a response to the hiichelson concerns as they relate to boiling water reactors by letter dated February 21,1980, Licens-eca and applicants should assess applicability and ad-equacy of this response to their plants. O Aniendment 17 1A 2191 l
ABWR mumic Standard Plant avc 1A.3 -INTERFACES appropriate, to improve the availability of the emer-Os gency core coolir.g equipment. IA.3.1 Emergency Paveedures and Emergency Procedures Training Program Emergency procedures, developed from the emergency procedures guidelines, shall be provided and implemented prior to fuelloading. (See Subsec-tion 1A.2.1). 1A.3.2 Review and Modify Procedures for Removing Safety Related Systems From SetTice Procedures shall be reviewed and modified (as required) for removing safety-related systems from senice (and restoring to senice) to ar.sure operabil-ity stalus is known. (See Subsections 1A.2.18 and 19) 1A.3.3 In Plant Radiation Monitoring Equipment and training procedures shall be. _ provided for accurately determining the altimrne io-dine concentration in areas within the facility where plant personnel may le present during the accident. (See Subsection 1A.235) 1A.3A Reporting Fallures of Reactor System Relief Valves Failures of reactor system relief valves shall be reported in the annual report to the NRC. (See Sub-section 1A.23.21.1). IA.3.5 Report on ECCS Outages Starting from the date of commercialopera-tions, an annual report should be submitted which in-cludes instance of emergency core cooling system u.1-availability because of component failure, mainte-nance outage (both forced or planned), or testing, the following information shall be collected: (1) Outage date (2) Duration of outage (3) Cause of outage (4) Emergency core cooling system or component involv-d ! (5) Corrective a . tion talen The above information shall be assembled into a O. report, which will also include a discussion of any changes, proposed or implemented, deemed l-Amendment 17 tAlt
-- -i
AHWR mamn Standard Plani nrv. n CilAPTER3 O raiitc or coxreurs <cDDiiDDPa) Sic 11on Utic APPENDIX 3A SEIShllC S0ll STRUCTURE IN1 TRACTION ANALYSIS APPENDIXJJ CONTAINhlEhT llYDRODYNAh!!C LOADS APPf:NDIXJf COhlPUTTR PROGRAhtS USED IN Till: DESIGN AND ANALYSIS Of SElShllC CATEGORY I STRUCTURI: AITISilX 3D COhlPLTITR PROGRAhtS USED IN Tilt DESIGN Of COhlPONENTS, EQUIPhlENT AND STRUCTURl;S All'LFDIX 31; I RACTURE hlECilANICS. LI:AK RAlt CALCULATION AND trak Dell:CTION hil'UlODS APPENDIX 30 APPLICATION Or LEAK.HEI'ORI:. BREAK APPROACll 10 AinVR PIPING SYSTEhtS APPENDIX 3G SEINh!!C ANALYSIS Hl:SULTS APPI:NDIX 311 DESIGN DI TAILS AND 1:YALUATION RI:SULTS Of SEIShllC CATEGORY I STRUCTURES ! APPENDIX 31 EQUIPhil:NT QUAliflCATION ENVIRONh!!:NTAL
- l. DESIGN CRI1ERIA i
i l' I 3.0ai Amerkiment 17
.....~._._.,._...._,......____a_,......,__.___._...~-.___.._.._.. _. . . _ _ _ _ . . _ . _ , . _ _ -.._.;,-- . _ .,
k 21At>10uAll mndard Plana arv n ; TAllLE 3.21 O- CLASSIFICATION
SUMMARY
he classification information is prnented System wise"* In the following order ! Table Table [ 3J 1 MPL 3J.1 MPL i Item No, Ihlt
~
liem No. Naniber** 2hl Number ** N- Nuclear Steam Supply Systems E2 E22 liigh Pressure Core lhxler l Sptem' !
-B1 B11/J10 Reactor Pressurc Vessel 1 J11/J12: Sptem'/ Fuel' E3 E31 Leak Detection and Isolation Sptem' B2 B21 Nuclear Boiler Sptem' ,
E4 E51 RCIC Sptem' B3 B31 Reactor Recirculation Sptem F Brac1gr Servicina i C Control and inntnament Sntems F1 Fil Fuel Servicing Equipment C1 C11/C12 CRD Splem' F2 F13 RPV Servicing Equipment
-C2 C31 Feedwater Control $ptem F3 F14 RPV Internal Servicing C3 C41 Standby Liquid Control Equipment ;
Sptem O C4 C51' Neutron Monitoring aptem' F4 F5 F15 F16 Refueling Equipment Fuel Storage Equipment
- C5 C61 Remote Shutdown Sptem G Reactor Atalliart Systemn [
C6 C71 Reactor Protection Sptem' G1 G31 Reactor Water Cleanup Spiem
- 1) Badiatlan Monitorinn sniems 02 G41 Fuel Pool Cooling and Cleanup D1 D11 Process Radiation Monitoring
- Sptem i Sptem .
I G3 G51 Suppression Pool Cleanup D2 D23 Containment Atmospheric. Sptem Monitoring Sptem* i Il Control Panels E f,,gre Cooline Systems : 111 1111/1112 Main Control Room Panel
- El Ell' RilR Sptem' ?
112 ll21- LocalControf Panels' l i
- These systerns or subsystems thereof, have o primaryfunction 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 designateifor the system
*** Only those systems that are in the ABliR Standard Plant scope are included in this table.
}
' Amendment 14 3.24 i
..__-.,,._.___.._--.._.____._.._,-._.-,_.._,-.-.~._,__~._,._-.,..,-_.,a
ABWR nuumi St andard l'iniit utv ti TAllLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Table Table 3.21 MPL 3.21 MPL litm b'o. humber" 11 Lit litm No. Etuubst" llllt J Epclear Fuel P S.igionAnillary Sutels See Item 111 P1 P11 Makeup Water System (Purified) K End10EllVe Waste SuttIn P2 P13 Makeup Water System (Condensate) K1 K11 Radioactive Drain P3 P21 Reactor Iluilding Cooling Transfer System Water System' P4 P22 Turbine lluilding Coolice Water System K2 K17 Radwaste System PS P24/P25 llVAC Cooling Water Systems' P6 P41 Reactor Senice Water System P7 P42 Turbine Seniec Water Syatem N Power Cvele Sntems P8 P51/PS2 Instrument / Service Alt Systems N1 N11/N21 Power Conversion System N22/N25 N26/N27 P9 P54 liigh Pressure Nitrogen Systems N31/N32 N33/N34 R Station Electrient SnirIn N35/N36 N37/N38 R1 R10/R11/ Auxiliary AC Power System' N39/N41 R22 N42/N43 N44/N51 R2 R42 DC Power Supply' N61/N71 N72 R3 R43 Emergency Diesel Generator System' N2 N62 Offgas System R4 R52 Lighting and Senicing Power Supply
- These systems or subsystems thereof. have a primarv function that is safety related, As shonn
(!Lthe baluner of this Table, some of these systems contain non safety related components a% converselv. some systemt whose primarv funettons are non-saferv related contain components that have been designated saferv-relatet l " Master Pans List Number desienated for the system l e l Amendment 17 32-7
ABWR m om4u Standard Plant nity. n TAHI.E 3.2 1 CLASSIFICATION
SUMMARY
(Continued) - I Quality Grvup Quality Safet lua. Classi. Assurance Selsmic Prindnal Comnonenta Gau[" lin' RWlad gggge gg f blu Bt Reactor Pressurt Yessel System / l I Fuel Assesabiles
- 1. Reactor vessel 1 C' A 11 1 i ,
- 2. . Reactor vessel support skirt . 1 C A 11 I and stabillier
- 3. Reactor vest.cl appurtenances 1 C A 11 1 (g)
. pressure retalning portions
- 4. Supports for CRD bousing. 1 C A 11 I in-core housing and redrcu- ;
lation internal pump-
$. Reactor internal structures. 2 C 11 11 i feedwater, RilR/ECCS high pressure core flooder spargers 64 Reactor internal structures. 3 C . Il i safety related components including core support structures (See subsection 3.9.5) ,
- 7. Reactor internal structures . N C - E -
non. safety related components 5 (See Subsection 3.9.$)
- 8. - Control rods 3 C - Il !
- 9. -- Power range detector hardware 3 C .. Il I including startup range detector ;
- 10. Fuelassemblies 3 C -- Il i
- 11. Reactor Internal Pump 1 C A Il i Motor Casing it2 Nuclear lloller System
- 1. Vessels . levelinstrumenta. 2 C 11 11 I
. tion condensing chambers O
Arr.cndment 17 318
. - . , . - _ - _ . u .. _ _ _ . _..__ _ - _ _ _ _ _ _ _ _ ._, _ _ _.,... ,-. ,, , _ _ _ ..~. _ ,,, _ _ ,- - -, _
ABWR mame Standard Plant RTT. H TABLE 3.21 h CLASSIFICATION
SUMMARY
(Continued) Quality Group Quallt) Safety Loca. Classi. Assurance Selsmic Prineinal Comt,yga gg *b ge ficationd Etnuiremente catecorrf d'etr3
- 2. Venel . air aconmulators 3 C C D I (for ADS and SRVs)
- 3. Piping including supports . 2/3 C D/C D 1 (b) g safety / relief valve dacharge ;;
l l O O Amendment 11 3241
ABWR maan Standard Plant Rim n (3 TAlli E 3.21
~'
CIASSIFICATION SUMMAlW (Continued) Quality Group Quality Safety laa. Classi. Assurance Selsmic hittripal Componen1 8 flaub ((ge g}g3t[g d MuhM' l'Mm#h R2 Nuclear floller System (Continued)
- 4. Piping induding supports 1 C,SC A 11 I rnain steamline (MSL) and feed-water (ITV)line up to and in.
duding the outerrnost isolation valve g i Piping including supports. 2 SC 11 11 I g htSL from outermost
" isolation valve to and including seismic interface restraint and ITV from outermost isolation valve to the seismic interfm restraint including the Gutoff valve and the restraint r3 .
Piping including supports MSL N E (r) (N 6. SC.T g
- 11 -
'~ [
from the seismic interface restraint to the turbine stop vnbe
- 7. Deleted Piping ITV beyond the seismie N SC,T D E 8.
{ interface restraint
- 9. Deleted
- 10. Pipe whip restraints MSL/ITY 3 SC,C - 11 -
$ 11. Pipinginduding supports other A within outermost isolation valves
- a. RPV head vent 1 C A 11 1 (g)
- b. RPV head spray 1 C A 11 1 (g)
- c. Main steam drains 1 C.SC A 11 1 (g) 8 12. Piping including supports-other S beyond outermost isolation valves
- a. RPV head vent N C D E -
- i
(]' (j b. c. RPV head spray Main steam drains N N SC SC D D E E I I I Amendment 17 12 9
ABWR uuima Standard Plant am n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safet*y loca. Classi. Assuranct Selsmic
- Principal Componenta flan b ge Dentiond Reaulrrment' Catemoni &ln B2 Nuclear Boller System (Continued) ,
13 Pipingincludingsupports. C,SC 1/- (g) 9~j 2/N B/D B/E instrumentation beyc.nd outermost isolation valves
- 14. Safety / relief vahrs 1 C A B 1 ,
A-11 Valves . MSL and FW 1 C,SC B 1 isolation valves, and other , FW valves within containment ,
- 16. Vahrs . FW, other beyond 2 SC B B 1 outermost isolation valves up to l and locluding shutoff valves ,
g 17. Valves . within outermost isolation vahes
- a. RPV head vent 1 C A B I (g)
- b. RPV head spray 1 C,SC A B -1 (g)
- c. Main steam drains 1 C,SC A B 1 (g)
- 18. Valves, other
- a. RPV head vent 3 C C B 1
- b. RPV head spray 2 SC B B 1
- c. Mai:: Steam Drain N SC D E .--
- 19. Valves. instrumentation beyond 2/N SC B/D B/E I/- (g) outermost isolation valves l . .. .
i 20. Mechanicalmodules-instrumen. 3 C,5C - B ! tation with safety related function
- 21. Electrical modules with safety. 3 C,SC,X -- B- 1 (i) -
related function
- 22. Cable with safety-related 3 C,SC,X - - - B 1 function l'
-- Amendmeat 17 3.2 10
.,_4..... .. _._w.__.,,,,_._. _ _ _ , _ . _ u.,_-,......,,._. _ . . _ _ . - . . _ . _ , _ _ _ -
ABWR aumw nw n Siandard Plant __. TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Ol Quality Group Quality Safet loca. Classi- Assurance Selsmic Princinal Componenta g{gg ggggi e gigt[p3 d Recuirernent' CattE2DI Enlu B3 Reactor Recirculation Sptem j
- 1. Piping Primary side, motor 3 C C D I (s) @h cooling system 2 Pipe Supports 3 C .C B 1
- 3. Pump motor cover 2 C D 11 1
- 4. Pump non pressure retain. N C -- E --
ing parts including motor, $ instruments, electrical cables and seals
- 5. Valves 3 C C B i (g)
I
- 6. ATWS equipment associated (cc) g with the pump trip function C1 CRD Sptem
- 1. Valves with no safety related 2 SC B B 1 (g) 3 M
function (not part ofIICU) y 2. Pipingincluding supports. 2 C,5C B B i (j)
- r. insert line
- 3. Piping other (pump suction, N SC D E - - - (g) pump discharge, drive header)
- 4. liydraulic control unit 2 SC -- B 1 (k)
- 5. Fine motion drive motor N C -- E - - -
{
- 6. CRD Drive water pumps N SC D E ---
Control Rod Drive C A /--- B 1 { 7. 1/3
- 8. Electrical modules with 3 C,SC --- B I safety fuaction
- 9. Cable with safety-related 3 C,5C,X B I function Amendment 17 3211
ABWR muun 1 Standard Plant myn i I TAllLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca. Classi. Assurance Selsmle Principal Componenta flaub ge ggd . Etaulmmenit Catemoni h j C1 CRD System (Contd)
- 10. ATWS Equipment associated k ,
with the Alternate Rod Insert (ARI) functions (cc) C2 Fredwater Control System 1.~ Electrical modules N C T,5C. - E . X ,
- 2. Cable - N C.T.SC, ... E -
X O i 1 I i e l v : l Amendment 17 V i 4
ABWR zwimu Standard Plant niv n TAllLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca. Classi. Anurance Selsmic l rincloal Comoonent" Claub ucD' DeaHon d Erguirement' Cnkten f Netts C3 Standby Liquid Control Sptem
- 1. Standby liquid control 2 SC 11 B 1 (u) tank including supports
- 2. Pump induding supports 2 SC 11 B 1 (u)
- 3. Pump motor 2 SC - Il 1 (u)
- 4. Valves . Injection 1 SC A 11 1 (u)
- 5. Valves within injection 1 C.SC A Il I (g.u) valves
- 6. Valves beyond injection 2 SC B 11 1 (g,u) valves
- 7. Piping including supports C,SC A B ! (g,u)
$I 1 within injection valves Nl
- 8. Piping including supports 2 SC 11 11 1 (g.u) beyond injection valves y 9. Electrical equipment 3/N SC,X ---
B/E 1/- (cc) h and devices
- 10. Cable 3/N SC,X --
B/E 1/ - (cc) C/ Neutron Monitoring Sptem
- 1. Electrical modules - 3 SC,X, - It i SRNM, LPRM and APRM
- 2. Cable SRNM and LPRM 3 C,SC,X, --- B I RZ
- 3. Detector and tube 2/3 C D/C B 1 assembly 1 0 1
Ah:endment 17 12-12 l l l
ABWR zw=n Standard Plant nuv n - TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) l Quility Group Quality Safety loca. Classl. Assurance Seismic n' hineinal Connonenta - fJau b 1}2BC ficationd Rtquirement' Calttun I h i C5 Remote Shutdown System Components of this system arc included under 111, El, E4, G3,112, and P2.
- 1. Electrical modules with 3 C,5C,RZ, - - 11 I safety related function X
- 2. Cable with se.p ::ated 3 RZ - 13 I function C6 Reactor Protection System
- 1. Electrical modules with 3 SC,X,T. - 11 1 safety-related function RZ O' 2. Cable with safety 3 SC,X,T, - II -1
-. related functions RZ ,
- 3. Electrical Modules, other N T,X - E -
(u)
- 4. Cable, other N T,X - E -
(u) t N C7 Process Computer (includes N- X - E - PMCS and PGCS) DI Process Radiation Monitoring System (includes gaseous and liquid amuent monitoring)
- 1. Electrical modules with 3 SC,X,RZ ~ 11 I with safety-related functions (includes monitors) - ,
- 2. Cable with safty related - 3 SC,X,RZ - II - I functions
- 3. Electrical Modules, other ' -N T,SC,RZ, - E -
(u) X,W 5 - 4. Cable, other N T,5C RZ, - U --- (u)- I " O e Amendment 17 31 13 +
ABWR naaoose Standard Plant RN n TABLE.. 1 CLASSIFICATION
SUMMARY
(Continued) O Quality Group Quality Safety loca- Classi- Assurance Seismic Princloal Comnonenta ggg b ge ggd Reautrement' Categor1I Notes D2 Containment Atmosphele Monitoring System
- 1. Component with safety related 3 C,5C - - - B I El RIIR System
- 1. Heat exchangers-primary side 2 SC B B !
- 2. Heat exchangers including 3 SC C B 1 supports secondary side
- 3. Piping including supports
- 1/2 C,SC A/B B 1 (g) within outermost isolation valves
- 4. Containment spray piping 2 C B B I iricluding supports and h spargers, wsthin e.nd A including the outer-most isolation valves 4a. Piping including supports 2/3 SC B/C B 1 (g) l- beyond outermost isolation vrives
- 5. Main Pumps including supports 2 SC B B I
- 6. Main Pump rnotors 3 SC --- B 1
! 7. Valves . isoli !c, (LPFL 1 C,5C A B 1 (g) i line) including shutdown suction Le isolation valves
- 8. Valves - isolation, 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
- The RHR/ECCS lowpressureflooder spargers are part of the reactorpressure vessel system, see Item Bl.5.
Amendment 11 3.2-14
1 i i ABM 23462naan -! , Standard Plant - REY. H l i TABLE 3.21 !
.G. -k/ CIASSIFICATION
SUMMARY
(Continued)
]
i Quality ' I Group Quality Safety loca- Classi. Assurance . Seismic Principal Componenta gg,3 b- ge ficationd ' Reaulremtal' Catenord b'010 El RHR System (Continued)
- 10. Mechanical modules with 3 SC C B I safety-related f2nctions
- 11. Electricalmoduleswith 3 C,SC,X - B I safety-related function
- 12. . Cable with safety-re' ted 3 C,SC,X -- -B 1 function
=j' 13. Other mechanical aad .N C,5C,X --- E -
ely:trical modules .
$ . 14. Jockey pumps including 2 SC B B I
-A supports 11 Jockey pump motor N SC -- E ---
i a Amendmeitt 17 ' 3.2 14.1
ABWR zwuu Standard Plant nrv.n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety LAica- Classi- Assurance Seismic Et[nsipal Componenta cl333b gc ficatiend E,tgulrement' Catecorvf Notes E2 liigh Pressure Core Flooder
! 3. Reactor pressure vessel 1/2 C,SC A/B B I (g) injection line and connected s I
pipingincluding supports with-g in outermost isolation valve
- M
- 2. All other piping including 2/3 SC,0 B/C B I (g) supports **
- 3. Main Pump 2 SC B B 1
- 4. Main Pump motor 3 SC -- B 1
- 5. Valves - outer isolation 1 C,SC A B I (g) .-
and within the reactor pressure E vesselinjection line and connected lines
- 6. All other valves 2/.' SC B/C B 1 (g)
- 7. Electrical modules with safety- 3 C SC,X --- B I relat:d function
- 8. Cable with safety-related 3 C,5C,X --- B I function E3 Leak Detection and Isolation System
- 1. Temperature sensors 3/N C,SC --
B/E I/ - (7)
- 2. Temperature switches 3/N X - - -
B/E 1/-- (z) j 3. Pressure transmitters 3/N C,SC --- B/E I/--- (z)
- 4. Pressure switches 3/N X --
B/E I/-- (z)
- 5. Differential pressure 3/N C,SC ---
B/E I/--- (z) transmitters (flow) The ECCS high pressure core flooder spargers are part of the Reactor Pressure Vessel System, see . Item Bl.S.
" Pool suction piping, suction pipingfrom condensate storage tank, test line to pool, pump discharge
. piping and retum line to pool.
Amendmen-
' 3 2.t5
ABWR mamn Standard Plent suv ti TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safet"y loca- Classt. Assurance Selsmic Prineinal Com%nenta gb ge Dentiond Renutrement' Catecord Entn E3 Leak Detection and Isolation System (Continued)
- 6. Differential pressure 3/N X -- B 1/--- (z) switches
- 7. Square root converters 3/N X --- B 1/ -- (z)
- 8. Differential flow summers ~ J X -- B 1/--- (z)
- 9. Differential now switches 3 X - B I
- 10. Timer switches 3 X -- B 1
- 11. Power supplies 3 C,SC --- B 1
- 12. Radiation monitor N SC -- E I
- 13. Instrumentlines 3 C,5C D B 1
- 14. Stimpk lines
- 2/N C,SC C/D/--- B/E 1/---
- 15. Flow transmitters N SC -- E ---
g
- 16. Cables 3/N SC,RZ,X --- B/E 1/---
FA RCIC System
- 1. Piping including supports with. 1/2/3 C,SC A/B/C B 1 (g) 8 in outermost isolation valves
- 2. Piping including supports - N O,5C D E --- (g) ;
A Suction line from condensate storage tank beyond second shut-off valve and vacuum pump discharge line from vacuum pump to containment isolation valves
- These sample lines are totally within containment and radiation monitoringprovides no isolationfunction. l 3.2-16 Amendment 17
ABWR a4620are nrv. n Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca- Classi- Assurance Seismic Princinal Componenta gggg b gc Deation d Reaulrrment' Cateconi Notes E4 RCIC System (Continued) Piping including supports be- B/C B I (g)
- 3. 2/3 C.SC 0 yond outermost isolation valves
- 4. RCIC Pump including supports 2 SC D B 1
, 5. Pump motors N C,SC --- E I 2
- 6. Valver outer isola. 1/2 C,SC A/B B 1 (g) tion and within
- 7. Valves . shut off N O,SC D E --- (g) g' line from condensate storage beyond second shut c,ff valve.
- 8. Valves - other' 2/3 SC D/C B 1 (g) ,
- 9. Turbine including supports 2 SC -- B 1 (m)
- 10. Electricalmoduleswith 3 C,SC,X - - - B I safety-related function
- 11. Cable with safety 3 C,SC,X -- B 1 function
- 12. Other mechanical and N C,SC,X --- E -
$ electrical modules F1 Fuel Servicing Equipment
- 1. Fuel preparations machine N SC -- E ---
c
^ 2. General purpose grapple N SC - E ---
- RCIC turbine steam admission valve, pump suction valvefrom condensate storage tank, and turbine inlet and exhaust drain valves.
Amendment 17 12-17 l l r
ABM 234t.iman
-- Standard Plant nix ii ll 6
-f TABLE 3.21
~
CLASSIFICATION
SUMMARY
(Continued) Quality . Group Quality loca- Classi- Assurance Seismic Safetg Princinal Comnonent# Ogn gon' ficationd Reautremente Catenord N_nts.3 F2 RPV Servidag Equipmen!
- 1. -Steamline plugs N SC --- E ---
- 2. Dryer and separator N SC --- E -
strongback and head strongback F3 .. RPV Internal Servicing Equipment
- 1. Control rod grapple - N SC --- E -- +
F4 Refueling Equipment
) = 1. Refueling equipment N SC --- E I (bb) h
$ platform assembly
- 2. Refueling bellows N SC --- E --
F5 Fuel Storage Equipment L1, Fuel storage racks - N SC -- E I (bb) new and sixnt . 3-n
- 2. Defective fuel storage N SC -- E --
(bb) container
- G1 Reactor Water Cleanup System
- 1. Vessels including supports N SC C E ---
(filter /demineralizer) E 2. Regenerative heat exchangers N SC C E -- including supports carrying reactor water
- 3. Cleanup recirculation N SC C E --
pump, motors Amendment 17 3 2-18 _ , . . _ . ~ . . _ _ . . . _ . - _ _ . . . . . . . . ~ . - - , - _ , ~ , . . _ _ - - . . _ _ . . . . _ , , _ . . . . . . . . _ _ . _ . . _ . _ . . . _ . . _ ~ ,.
l ABWR maman riy.J1 l Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca- Classi. Assurance Seismic Princloal Componenta gb ge ggd Recuirement' fatecord Ntin G1 Reactor Water Cleanup System (Continued) Pipinginduding supports and C,SC A B 1 (g)
- 4. 1 valves within and including s
outermost containment isolation valves on pump suction f 5. Pump suction and discharge piping including supports and N SC C E -- (g) valves from containment isola-tion valves back to shut-off vales at feedwater line connections
- 6. Pipinginduding supports and 2 SC B B 1 (g) valves from feedwater lines to and including shut-off valves e
- 7. Pipingincluding supports and N SC,T C E --- (g) y valves to main condenser
- 8. Non-regenerative heat exchanger N SC C E --- (g) ~
tube inside and piping including supports and valves carrying process water n 2 SC D E --
- 9. Non-regenerative heat exchanger N shell and pipingincluding supports carrying closed cooling water
- 10. Filter /demineralizer N SC D E --
precoat subsystem
- 11. Filter demin holding pumps N SC C E --
including supports - valves and piping including supports O 3.2-19 l Amendment 17
. . . . _. _ , . . . .- . . . . - . . - . - , . , - - . ~ . . - . - - . -
ABWR
~ '
- 2mionin .
- Standard Plant RN D TAHLE 3.2 1 -
CLASSIFICATION
SUMMARY
(Continued) . Quality Group Quality Safety Loca. Classt. Assurance Selsmic Principal Comnonents . ggg b MC Rcationd Reautrement' Catenord - N.utu G1 Reactor Water Cleanup System (Continued) .
- 12. Sample station N SC D E --
4 $
;; 13 Electrical modules and cable N SC,X D E --- -
A with no safety related function
'14'. Electrical modules and cable '3 SC +-- B 1 ;O[
4 ; O Amendment 17 3.2 19.1-l l
. , _ . . , . ~ , . _ . - _ _ _ . , . . . , , , _ . . . _ . . . . _ . , , , - . . , , , . . . . , . . . . . _ _ _ , . . . . , . . .
ABWR 23^ mort RIW D i Standard Plant TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) 0 j 1 Quality Group Quality Safety 14ca. Classi- Assurance Selsmic Princloal Comnonenta gggg b ge Destion d Recutrement' Catecord Notes G2 Fut3 Pool Cooling and Cleanup System 1 .tmls including supports - N SC C E --- filter /aeincralizers 1
- 2. Vessels including supports - N SC C E ---
l drain tanks 8 A 3. Heat exchangers including N SC C E -- supports , a
- 4. Pumpsincluding support and N SC C E ---
pump motors
- 5. Pipingincluding supports, N SC C E --
valves
- 6. Normal makeup system N SC,0,T C E --
components t
- 7. RHR connections for 3 SC C B I $
safety-related makeup P, 8. Electrical modules and N SC,X -- E - - - 5 O cables with no safety-related function G3 Suppression Pool Cleanup System
- 1. Isolation valves and piping 2 C B B 1
$ including supports within M outermost isolation valves
- 2. Pumps N SC D E --
- 3. Pump motors N- SC - E ---
n
- 4. Other piping including supports N SC D E --- $
g o " E
- 5. Electrical modules and Cables N SC,X - - - ---
O Amendment 17 3.2-20
6ABWR~ 2mmu Standard Plant nrv. n TAllLE 3.21 1 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety loca- Classi- Assurance Seismic Princinal Comnonenta ci,33b ge g3 d Reuulrement' ' Catruorvf Notes
. til Main Control Room Panel I Panels ' X B/E
- 1. 3/N ---
1/- - (aa) {.
. 2, Electrical Modules with 3 X -- B I safety related function
- 3. Cable with safety-related 3 X -- 11 1-function E *
. 4. Other mechanical and - -N X -- --
electrical modules k 112 local Control Pancis 5 1. Panels or Racks 3/N C,SC,X --- B/E 1/- - (aa) *
- 2. Electrical modules with' 3 C,SC,X -- e B 1 safety-related function
- 3. Cable with safety-related 3 - C,SC,X, -- B I function ,
- 4. Other mechanical and N C,SC,X ---
E -- electrical modules K1 Radioactive Drain Transfer System
- 1. ' Drain piping including supports N ALL D E. -- -(p) $
- and valves - radioactive (except) 4 RZ,X) 2 Drain pipingincluding supports N .ALL D E -- (p) h and valves nonradioactive
~ ;;
3.- P; ping and valves contain- 2 C,SC B- B I ment isolation 4 Other mechanical and N ALL - E -- (p);
~') electrical modules Amendment 17 3.2-21
+
ABWR momn Standard Plant nrv n TAllLE 3 2-1 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety ina- Classi- Assurance Seismic Princloal Componenta pygg b gg e gd Reautremente Cateconi Notes K2 Radwaste Sptem
- 1. Piping including supports N C,SC 13 B 1 and valves forming part of containment boundary
- 2. Pressure vessels including N W - E -- (p)
- 3. Atmospheric tanks including N C,SC,II, -- E -- (p) supports T,W
- 4. 0-15 PSIG Tanks and supports N W --
E -- (p) k
- 5. Heat exchangers and supports N C,SC,W -- E -- (p)
- 6. Piping including supports N C,SC,11 -- E -- (p) and vahts T,W T1 Power Conversion System
- 1. (Deleted)
- 2. Branch line of MSLincluding N SC,T B B -
(r) supports between the second isolation valve and the turbine stop valve from branch point at MSL to and including the first - valve in the branch line
- 3. Main feedwater line (MFL) N SC B B 1 including supports from second isolation valve branch lines and components beyond up to outboard shutoff valves O
3 2-21.1 Amendment 17
_. ~ . _ _ _ _ _ _ m.-_
-1 JAB M ur6ioaan l
Standard Plant - nw n
- - TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca- . Classi- Assurance Seismic Pdadpal Componenta flaub ht ' neationd Reaulrement' Catenoni Notes N T D E Turbine bypass piping
- 4. ~
including supports
- 5. Turbine stop valve, turbine N T D E --
(1)(n)(o)
- bypass valves, and the main steam leads from the turbine controlvalve to the turbine casing =
- 6. Feedwater system components N T D E -
beyond outboard shutoff valve
- 7. Turbine generator' N T -- E -
- 8. Condenser N T -- E --
(
' 9. Air ejector equipment N T - E -
10.- Turbine gland scaling N T D E -
- system components R
' 11. Circulating water system N T D E --
N2 Offgas System
- 1. Pressure vessels including - N T -- E -- (p)(q) :
supports
- 2. : Atmospheric tanks includhg N T -- E -- (p)(q)
! . supports
- 3. 0-15 psig tanks including - N T -- E - (p)(q) supports
- 4. Heat exchangers including N T -- E - (p)(q) supports
- 5. Pipingincluding supports . N T -- E - (p)(q) and valves
- 6. Pumps including supports . N T - E - (p)(q) l Amendment 17 32-212
ABWR nuinoru Standard Plant nry n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca. Classi. Assurance Seismic frincloal Componenta Classb ge ggd Etautreraente Entecorv f Notes P1 Makeup Water System (Purified) i
- 1. Pipinginduding supports and 2 C B B i ;
valves forming part of the con- ! tainment boundary 1
- 2. Demineralizer water storage tank including supports N O D E - - -
l
- 3. Demineralizer vcater header - 2 SC B B I piping induding supports and valves
- 4. Pipinginduding supports and N O D E ---
valves
- 5. Other components N O D E ---
2 P2 - Makeup Water System (Condensate)
- 1. Condensate storage tank N O D E ---
(w) induding supports
@ 2. - Condensate header piping 2 SC B B 1 A including supports and valves
- 3. Piping including supports and N O D --- ---
valves l l 4. Other components N O D E --- T l 5 l j P3 Reactor Building Cooling Water System
- 1. Piping and valves forming part 2 SC,C B B I (g) l of primary containment boundary 6
A 2. Other safety.related piping, 3 SC,C C B I . including supports pumps and valves l 1 O l Amendment 17 3.2-22 l
?ABWR mamu ?
' Standard Plant nrrv. n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca. Classi- Assurance Selsmic Princinal Comnonenta . Classb gc gd Etauiremente Cateepn f Notes -
- 3. - Electrical modules with .
3 SC,C,X -- B 1 with safety related function
- 4. Cable with safety related 3 SC,C,X -- B 1 function
- 5. Other mechanical and N SC,C,X,M -- E ---
electrical modules P4 Turbine Building Cooling N T D E -- Water System P5 ; HVAC Cooling Water Systems 1,. Chillers, pumps, valves, and 3 SC,X C B 1
. piping including supports-
' Safety related HVAC support ,
;; 2. Chillers, pumps, vahts, and N C,SC,RZ, --- E - M
. piping including support: - 5 non safety related HVAC support
- 3. Pipingincluding supports aEd 2 C,SC B B 1 valves forming part of contain-ment boundary l
l; 4J Electrical modules and cable- 3 SC,X -- B l with safety-related function ! 5. Other mechanical and .N C,SC,RZ, --- E -- M electrical modules T,X $ e i'
- P6 - Reactor Service Water System
- 1. Safety-related piping - 3 U,0,X C B 1 including supports, piping b
and valves '
. 2. ' Electrical modules and cables 3 U,0,X --- B I with safety-related function OL :
Amendment 17 3.2-2.)
. . . - - . = - . - - , - , - . - - - . - . - . . . , , . - - - - - . - . . - - - , - . .- - ,,
ABWR amime Standard Plant Riv n TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) 1 Quality Group Quality Safety loca- Classi- Assurance Scismic Princion! Component
- f.lau b dec R@d Reautremenf cateconf Notes i
- 3. Other non safety reisted N U,0,X -- E ---
mechanical and electrical modules l P7 Turbine Senice Water System g
- 1. Non-safety related piping N P,0,T --- E ---
including supports, piping and valves
- 2. Ele'rical modules and cables N P,0,T --- E --- i with non-safety rclated functioa P8 Instrument / Service Air Systems g 1. Containment bolation includ- 2 C B B I l
g ing supports valves and piping
- 2. Other non-tafety related N SC,RZ, --- E ---
3 X,T,11,
^
mechanical and electrical components W,C P9 liigh Pnssure Nitrogen Fystems
- 1. Containment isolsion includ- 2 C B B 1 l ing supports valvr2 and piping
- 2. Pipingincluding supports with 3 SC,C C B 1 safety-related function l
- 3. Electric modules with 3 RZ,X --- B 1 safety-related functions l
- 4. Cable with safety related 3 SC,RZ, --- B 1 l
l function X
- 5. Other non-safety related N SC,RZ, -- E ---
5 mechanical and electrical X components O Amendment 17 3223.1
ABM usaaasu Standard Plant- nrv. n - h TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group .. Quality Safet,t 14ca. Classi. Assursace Selsmic Princinal ComKtatat" Clau' ilant Acationd itequiremehd Caltgen# Entta
- 2. Control power cabl s (includ. 3- SC,C,X, -- . B 1 Ing under ground cable system, R2 cable splices, connectors and terminal blocks)
- 3. Ccaduit and cable trays and 3 SC,C,X, -. B 1 their supports RZ
- 4. Protective relays and control 3 SC,X,RZ - B !
panck
- 5. Cont:Jnment clectrical pene- 3 SC,C --
B I trations a;.sembhes
- 6. Motors 3 SC,C,X, - B 1 .
RZ ;
- I R2 Auxillary AC Power System .
, : 1. 5900 volt switch gear 3 SC,X,RZ - B 1
- 2. 480 volt load centers 3- SC,X,RZ --- B 1
- 3. 480 volt anotor control
. - 3 SC,X,RZ --. B I
- centers
- 4. 120 VAC safety related dis- 3 SC,X,RZ -- B 1 tribution equipment includ-ing inverters -
- 5. Control and power cables ~ 3 SC,C,X -- B 1 (including underground RZ cable systems, cable splices, connectors and terminal blocks)
- 6. _ Conduit and cable trays and 3 SC,C,X --- B I
, their supports RZ
- 7. Ccatainment electrical 3 SC,C,X - B 1 penetration assemblies RZ b 8. Transiormers 3 SC,C,X --- B 1 ,
RZ Amendment 10 3.2 24
ABWR mamu Standard Plant nrv. n TABLE 3.21 CIASSIFICATION
SUMMARY
(Continued) O Quality Group Quality Loca. Classi- Assurance Seismic Safetg- d Princloal Consonenta Class 1120' Etallna Reautrement' Cattgat2I Notes
- 9. Motors 3 SC,C,X,RZ - - B 1
- 10. Load sequencers 3 SC,X,RZ -- B 1 E 11. Protective relays and control 3 SC,X,RZ -- B 1 panels
- 12. Valve operators 3 SC,C,X,RZ --- B I R3 Emergency Diesel Generator System
- 1. Starting air receiver tanks 3 RZ C B i (y) pipingincluding supports from and including check valve and downstream piping including supports and valves N RZ E
- 2. Starting air compressor and motors
{
- 3. Combustion air intake and 3 RZ,0 C B 1 exhaust system
- 4. Safety-related piping in- 3 RZ,0 C B 1 cluding supports valves - fuel oil system, diesel cooling water system, and tube oil system
- 5. Pump motors - fuel oil 3 RZ,0 -- B I system, dicsci cooling water system and lube oil system
- 6. Diesel generators 3 RZ --- B I (y)
- 7. Mechanical and electrical 3 RZ,0,X --- B I modules with safety-related functions 8 Cable with safety-related 3 RZ,C,X -- B I functions
- 9. Other mechanical and N RZ,0 --- E ---
g
~
electrical modules Amendment 17 32241
4 l l l
?ABWR - 2m uern Standard Plant - myn TABLE 3.21 -
CLASSIFICATION
SUMMARY
(Continued) Quality. Group Quality - Safety Loca. Classi- Assurance Seismic ge EEiDelnal Comnone.nla pjggg b ficationd Reautrement' CatenonI Notes R4 Lighting and Servicing Power Supply
- 1. Emergencylighting N SC,C,X RZ
-- E --
j T1. Primary Containment System
- 1. Primary containment vessel 2 C D D 1 (PVC) . reinforced concrete containment vessel (RCCV) 2, Vent system (verical flow 2 C B B 1 channels and horizontal discharges -
- 3. Suppression chamber.drywell 2 C B B I vacuum breakers
- 4. PCV penetrations . 2 -C B B I and drywell steel head Upper and lower drywell airlocks 2 C,SC - B 1
- 6. Upper and lower drywell: 2 C,SC -- B I equipment hatches
- 7. Iower'drywell access tunnels 2 C -- B 1 I- 8. Suppression chamber . 2 C,SC -- B 1 access hatch
- 9.- Safty related instrumentation 2 C,SC -- B 1 T2 Containment Internal Structures
- 1. Reactor vessel stabilizer 3 C -
B I truss
-l g
Amendment 17 3.2-25
- . - - - - . _ _ , - , _ _ - . . . - . , - , _ . , , _ . ~ . _ . - . _ . _ _ . . - - - _ --
ABWR m6iman Standard Plant iutv. n TAllLE 3.21 CIASSIFICATION SUhlhiARY (Continued) O Quality Group Quality laa. Classi. Assurance Seismic Safetg Princloal Componenta Gan licB' ficallou d Reaulremente Catecord Notes T2 Containment Internal Structures (Continued)
- 2. Support structures for safety- 3 C -- B 1 related piping including supports and equipment T3 RPV Pedestal and Shield Wall
- 1. RPV pedestal and shield wall 3 C -- B 1
- 2. Diaphregm floor 3 C --- B I T4 Standby Gas Treatment System
- 1. All equipment except 3 SC,RZ --- b I deluge piping and valves
- 2. Deluge piping and valves N SC ---
E - - - j O Amendment 17 3.2 26 1
1 i ABM Standard Plant - 2346imin RIN. B TABLE 3.21 ! L CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Loca- Classi. Assurance Scismle Safetg d Prineinal Comnonenta Class 112nC (ication Reautrement' CatenonI Notes T5 Atmospheric Control System
- 1. Nitrogen Storage Tanks N .O --- E --
R
- 2. Vaporizers and controls N O --- E --
- 3. Pipingincluding supports 3 C,SC -- B l and valycs forming part-i of containment boundary
- 4. Electrical modules with' 3 . C,SC -
B I safety related function 5 - Cables with safety-related ' 3 C,SC,X -- B 1 function
- 6. Other piping and valves N SC,RZ,0 -- E --
L 7, Oxygen monitoring component N C,SC -- E --- (normal operation) T6 : Drywell Cooling System
- 1. . M otors N C -- E ---
j
- 2. Fans. N C --
E. ---
- 3. Coils, cooling N C --- E --
l 4. - Other mechanical and - N C,X - E --
- l. electrical modules
- 17. F!ammability Control System 2 SC B B 1 T8 Suppression PoolTemperature Monitoring System i
I 1. Electrical modules with 3 C,X,5C, --- B 1 > l- safety-related function RZ i - 2. Cable with safety-related 3 C,X,SC, -- B I function RZ Amendment 17 3.2-27
, - _ . - _ . . - - , . ..u-___. . , _ , __,,u.... . _ _ . _ _ _ _ . . . _ . _ _ . . _ _ . . _ _ . . _ _ _ _ _ . _ _ _ , _ - _ _ _ _ _ . . . - - -
I ABM 2346ioore nrw. n Standard Plant TAllLE 3.21 CLASSIFICATION
SUMMARY
(Continued) O Quality Group Quality Loca. Classt. Assurance Seismic Safetg d Princloal Componenta ggg gjpg c (1 cation Reoulrement' Cateconi Notes U1 Cranes and floists
- 1. Reactor Building ciane N SC -- E --
(x)
- 2. Refueling Bridge crane N SC -- E -
(x) 3. 4. Fuel handling jib crane Upper Drywell Senicing N N SC C E E I (x) {
- 5. Lower Drywell Servicing N C --- E I
- 6. Main Steam Tunnel Senicing N M -. E --
- 7. Special Service Rooms N SC,RZ, --- E ---
T,W,X U2 Ileating, Ventilating,and Air Conditioning Systems *
- 1. Safetyactated equipment"
- a. Fan-coil cooling units 3 SC,X --- B 1
- b. Heating units - steam 3 SC,X -- B !
or water
- c. Blowers - Air supply or 3 SC,RZ,X --- B I exhaust
- d. Ductwork 3 SC,RZ,X --- B 1
- c. Filters - Equipment areas 3 SC,RZ --- B I
- f. HEPA Filters, Charcoal 3 SC,RZ,X --- B 1 Absorbers ControlRooms and Primary Containment
- g. Valves and Dampers - 2 SC,RZ --- B I primary containment isolation
- Includes Reactor Building, Control Building, and Service Building thermal and radiological environmental control functions within the Nuclear Island scope.
" Controls environment in Main and Local control rooms, diesel-generator rooms, battery rooms, ECCS.,
RCIC , pump rooms within the Nuclear island, 3.2 28 Amendment 17 i
MM 23A6100AE Standard Plant At P TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca- Classi- . Assurance Seismic b Princinal Componenta fJau hC ficationd Reautremente catenorvf Notes U2 Heating,Yentilating,and Air Conditioning Systems" (Continued)
- h. Other safecy-related 3 SC,RZ, - B I valves and dampers X g
- i. Electrical modules with #
3 SC,R7, -- B 1
- ufety related function X
- j. Cable with safety related 3 SC,RZ, -- B- I function X -
- 2. Non-safety related equipment"
- a. 'HVAC mechanical or electrial components N SC,RZ,H, --
X,5C, E -- l
-%- with non safety related WT functions U3 Fire Protection System
- 1. Pipingincluding supports an'd 2 C B B I valves forming part of the :
l primary containment boundary
-2. Other pipingincluding supports N SC,C,X, D E --
(t) (u) and vahrs - RZ,II .
- 3. Pumps . -N F D E -
(t) -(u) *
- 5
- 4. ~ Pump motors N F --- E ---
(t) (u)
- 5. Electrical modules N C,SC,X, E (t) (u)
RZ,li, T,W ' Includes thennat and radiological emironmental controlfunctions within the ABHR Standard Plant scope.
- " Controls environment in rooms or areas containing non. safety related equipment within the .Q ABHR Standard Pla'st.
Amendment 17 3 2-29
. _ _ . . _ . . _ . , _ . _ . _ _ . . _ _ . _ . . _ . _ ~ . _ . . . . _ _ . . . . _ . .. --._ _ _ _ _ - - . . . . _ . . _ , _ , . . , _ _, . . . _ . . _
ABWR maw Standard Plant nrv n TABLE 3.2-1 CLASSIFICATION
SUMMARY
(Continued) Quality Group Quality Safety Loca- Classi. Assurance Seismic Mncloal Comnonenta gggb ge Ilcatien d }<equiremente Cateconf h U3 Fire Protection System (Continued)
- 6. CO2actuation modules N RZ,T --- --- ---
(t) (u)
- 7. Cables N SC,C,X, --- --- ---
(t) (u) y RZ,T,W $
- 8. Sprinklers N SC,X D --- ---
(t) (u) U4 CivilStructures 1, Reactor Building (Secondary ? SC,RZ --- B I Containment and Clean Zone)
- 2. Control Building 3 X --- B 1
- 3. Service Building N 11 --- - ---
4 Radwaste Building substructure 3 W --- B 1 W E T 4a. Radwaste Building N -- ---
- 5. Turbine Building N T -- -- ---
(v) O 12-30 Amendment 17
ABWR u-n Standard Plant Riw.15 (3 NOTE S
\_]
- a. A module is an assembly of interconnected components which constitute an identifiable device or piece of equipment. For example, electrical modules include sensors, power supplies, and signal processors and mechanical modules include turbines, strainers, and orifices.
- b. 1,2,3, N = Nuclear safety-related function designation dermed in Subsections 3.2.3 and 3.2.5.
- c. C = Primary Containment 11 - Service Building M = any other location O = Outdoors onsite RZ = Reactor Building Clean Zone (balance portion of the reactor building outside the Secondary Containment Zone)
SC = Secondary Containment portion of the reactor building T = Turbine Building W = Radwaste Btalding X = Control Building F = Firewater Pump liouse U = Ultimate lleat Sink Pump flouse P = Power Cycle ficat Sink Pump Ilouse
- d. A,B,C,0 = 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.
O v
= Ouality Group Classification not applicable to this equipment.
- e. B = the quality assurance requirements of 10CFR50, Appendix B are applied in accordance with the quality assurance program described in Chapter 17.
g E = Elements of 10CFR50, Appendix B are generally applied, commensurate with the importance of the equipment's function.
- 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) at: not applicable to the equipment. liowever, the equipment that is not safety related but which could damage Siesmic Category I equipment if its structural integrity failed is checked analytically and designed to assure its integrity under seismic loading l resulting from the SSE.
l
- g. 1. Lines one inch and smaller which are part of the reactor coolant pressure boundary shall be ASME Code Section lit, Class 2 and Seismic Category I.
- 2. All instrument lines which are connected to the reactor coolant pressure boundary and are l
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. Allinstrument lines which are connected to the reactor coolant pressure boundary and are not p utilized to actuate and monitor safety systems shall be Code Group D from the outer isolation V valve or the process shutoff valve (root valve) to the sensing instrumentation.
Amendment 17 3.2-31 l-
l AMM- 23xsiocre REV B i Standard Plant NOTES (Continued)
- 4. All other instrument lines:
i Through the root valve the lines shall be of the same classification as the system to which they are attached, il Beyond the root valve,if used to auuate a safety system, the lines shall be of the same classification as the system to which they are attached, iii 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 the process root valve through the remainder of the sampling system may be Code Group D.
n 6. All safety relaed instrument sensing lines shall be in conformance with the criteria of f Regulatory Guide 1.151.
- h. Relief valve discharge piping shall be Quality Group B and Seismic 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 analyzed 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.
- i. 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. Fail safe devices are non-Seismic Category 1.
- j. The control rod drive insert lines from the drive 11ange 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 insertion 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 int _egrity 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 band valves, etc.), it is considered that they do not apply to the specialty parts (e.g., solenoid valves, pneumatic components, and instru nents). O Amendmera 6 3.2-32
ABWR uumn Standard Plant niv n NOTES (Continued) (e) v The design and construction specifications for the liCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but of the remaining parts and details. For example: (1) all welds are LP inspected; (2) all socket welds are inspected for gap between pipe and socket bottom; (3) all welding is performed by qualified welders; and (4) all work is done per written procedures. Quality Group D is generally applicable because the codes and stardards invoked by that group contain clauses which permit the use of manufacturer standards and proven design techniques which are not explicitly defined within the codes for Quality Groups A, B, or C. This is supplemented by the OC Nchniqu described.
- 1. The tmbine stop valve is designed to withstand the SSE and maintain its integrity.
- m. The RCIC turbine is not included in the scope of standard codes. The assure that the turbine is fabricated to the standards commensurate with safety and performance requirements, General Electrie has established specific design requirements for this component which are as follows:
- 1. All welding shall be qualified in accordance with Section IX, ash 1E 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 eastings shall be radiographed according to:
l l ASTM E-94 l
.O E-141 l
b E-142 maximum feasible volume E-71, IS6 or 280 Severity level 3
- 4. As cast surfaces shall be magnetic particle or liquid.penetraat tested according to ash 1E Code, Section Ill, Paragraphs NB-2575, NC-2576, or NB-2576, and NC-2576.
- 5. Wheel and shaft forgings sha'l be ultrasonically tested according to ASThi A-388.
- 6. Butt welds shall be radiographed and magnetic particle or liquid penetrant tested according to the ash 1E Boiler and Pressure Vessel Code. Acceptance standards shall be in accordance with ash!E Boiler and Pressure Vessel Code Section 111, Paragraph NB-5340, NC-5340, NB-5350, or NC-5350, respectively.
- 7. Notification shall be made on major repairs and records maintained thereof.
- 8. Record system and traceability shall be according to AShf E Section Ill, NCA-4000.
- 9. Control and identification shall be according to ash 1E Section 11!, NCA-4(X10.
- 10. Procedures shfl conform to ASN1E Section III, NB-5100 and NC-5100.
- 11. Inspection personnel shall be qualified according to ash!E Section III, NB-5500 and NC-55(X).
l l l Amendment 10 3 2-33 l l
ABWR mme Standard Pinnt RIN. Il NOTES (Continued)
- n. All cast pressure retaining parts of a size and configuration for which volumetric methods are effective are examined by radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards is used as an alternate to radiographic methods. Examination procedures and acceptance standards are at least equivalent to those defined it. Paragraph 136.4, Nonboiler External Piping, ANSI B31.1.
- o. The following qualifications are met with respect to the certi6 cation requirements:
- 1. The manufacturer of the turbine stop valves. rbine control valves, turbine bypass valves, and main steam leads from turbine con':.it valve to turbine casing utilizes quality control procedures equivalent to those defines in GE Publication GE7 4982A, General Electric 'Large Steam Turbine Generator Quality Control Program.
- 2. A certification obtained from a n anufacturer of these valves and steam loads demonstrates that the quality control program as 0 fined has been accomplished.
The following requirements shall be met in addition to the Quality Group D requitetnents: 1 All longitudinal and circumferential butt weld joints shall be radiographed (or ultrasonically tested to equivalent standards). Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrate examination may be substituted. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI D31.1, 2 All fillet and socket welds shall be examined by either magnetic particle or liquid penetrate methods. All structural attachment welds to pressure retaining materials shall be examined by either magnetic particle or liquid penetrate methods. Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1. 3 Allinspection records shall be maintained for the lik et the plant. These records shall include data pertaining tt. qualification of inspection. personnel, examination procedures, and examination results.
- p. A quality assurance program meeting the guidance of Regdatory Guide 1.143 will be applied g during design and construction,
- q. Detailed seismic design criteria for the offgas system are provided in Section 11.3.
- r. The main steam lines from the containment outboard isolation valves and all branch lines 2-1/2 inches in diameter and larger, up to and including the first valve (including lines and valve supports) are designed by the use of an appropriate dynamic seismic system analysis to withstand the operating bases carthquake (OBE) and safe shutdown earthquake (SSE) design loads in combination with other appropriate loads, within the limits specified for Class 2 pipe in the ASME Section Ill. The mathematical model for the dynamic seismic analyses of the main steam lines and branch line piping includes the turbine stop valves and piping to the turbine casing.
The dynamic input loads for design of the main steam lines are derived from a time history model analysis or an equivalent method as described in Section 3.7. O Amendment 17 3.244
ABWR mten nry n Standard Plant (~] NOTQi (Continued) V s. The recirculation motor control syc_ < asyfied Quality Group C and Safety Class 3 widch is in accordance with the requiremenu . d50.55a. The RhtCS, which is part of the reactor coolant pressure boundary (RCPB), meets luGR50.55a(c)(2). Postulated failure of the RMCS piping cannot causc a loss of reactor coolant in excess of normal makeup (CRD return or RCIC flow), and the RMCS is not an engineered safety feature. Thus,in tne event of a postulated failure of the RMCS piping during normal operation, the reactor can be shutdown and cooled down in an orderly manner, and reactor coolant makeup can be provided by a normal make up system (e.g., CRD return or RCIC system). Thus, per 10CFR50.55a(c)(2), the RMCS need not be classified Quality Group A or Safety Class 1. Since the RMCS is not an engineered safety feature (e.g., it does not provide emergency reactivity control, emergency core coolant, or primary reactor containment), the system need not be classified Quality Group B or Safety Class 2. The RMCS is classified Quality Group C and Safety Class 3. however, the system is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Class I criteria as specified in Subv. ion 3.93.1.4 and Figure 5.4-14.
, t. A quality assurance program for the Fire Protection System meeting the guidance of Branch Technical
$ Position CMEB 9.51 (NUREG-0800), is applied.
- u. Special seismie qualification and quality assurance requirements are applied.
- v. See Subsection 113.4.6 for the offgas vault scismic requitrements.
- w. The condensate storage tank will be designed, fabricated, and tested to meet the intent of API Standard API 650. In addition, the specification for this tank will require: (1) 100% surface examination of the side wall to bottom joint and (2) 100% volumetric exat ination of the side wall l ) weld jo nts.
v
- x. The cranes are designed to hold up their loads under conditions of OBE and to maintain their positions over the units under conditions of SSE.
- y. All off engine components are constructed to the extent possible to the ASME Code, Section III, Class 3.
l z. Components associated with a safety-related Emetion (e.g., isolation) are safety-related. aa. Structures which support or house safety-related mechanical or electrical components are safety-related. l l bb. A quality assurance requirements shall be applied to ensure that the design, construction and testing requirements are met. l , cc. A quality assurance program, which meets or exceeds the guidnace of Generic Letter 85 06, is applied
% to all non-safety rela'ed ATWS equipment.
l l l (3 V I Amendment 17 12411 l l
ABWR mww SIRudutMRDI = ItllD
, 3.4 WATER LEVEL (FLOOD) DESIGN 3.4.1.1.1 11ood Protet tlon ' rom laternal t Sourecs The types and methods used for protecting the f
=
AllWit safety related structures, systems and components from external flooding shall conform Seismic Category I structures that may bc affected by design basis floods are dnigned to to the guidelines defined in itG 1.102. withstand the floods patulated in Table 2 01 h using the hardened protection approach with j Criteria for the design basis for protection structural prosisions with incorporated in the against external flooding shall conform to the plant design to protect saf ety related I requirements of itG 1.$9. The design criteria for structures, systen.s, and cornponents itom ! protection against the effects of compartment postulated flooding. Scir.mic Catt gor y I flooding shall anform to the requirements of strudures required for safe shutdown remain ANSI /ANS $6.11. The design basis flood levi's accessible during all flood conditions, are specified in Table 3.41. Safety related systems and components are 3.4.1 Flood Protecflori flood protected either because of their location abose the design flood lesel or because they are This section discusses the flood protection enclosed in reinfoned conorte $cistnic rategory measures that are applicable to the standard AllWit I structures u hich have Ihe foilowing plant Scistnic Category I structures, sptcms, and requirements; b components for both esternal flooding and
& postulated flooding from plant component (1) w all thit L nessn below flood lesel of not failures. These protection rneasures alto apply less than two Icet; to other structures that house systems and components importan; to safety which fall within (2) water stops provided in all conur uction the scope of plant specific. joints below flood incl;
( 3.4.1.1 Ilood Protection Measurn for Scismic (3) watertight doors and equiptwnt hatches Category I Structurn installed below design flood inel; and o The safety related systems and components of (4) waterproof coating of nternal surface ( D the AllWR Standard Plant are located in the 6 q reactor, control, and radwaste buildings which Waterproofing of foundations and walls of g are scismic category 1 str uctures. These Seismic Category I structures below grade is structures together with those identified in accomplished principally by the use of water Table 3.41 are protected against external flood stops at expansion and construction joints. In damage. Flood protection of safety related addition to water stops. waterproofing of the systems and components is provided for all plant structures that house safety selated H systems and components is prmided u, ;o 8 cn' (3 l s hl described postulated daign in Table flood lesels 2.01. Postulated flooding and conditions in) abmc the plant ground level to protni the O from component failures in the building compart- external surfaces from exposure to water. ments does not adversely affect plant safety nor does it represent any harard to the public. The flood protection measures that are described above also guard against flooding from Structures which horse the safety related on. site storage tanks that may r upture. The equipment and offer flood protection are largest is the condensate storage tent that has
' l identified in Table 3.41. Descriptions of these a capacity of 2,110 cub?c meters. This tant is structures are provided in Subsection 3L$ and constructed from stainless steel and is located 3.8.5. Exterior or access openingt and between the turbine building and the radwaste penetrations that are below the design flood building where there are no direct entries to level are identified in Table 6.2 9 these buildings. All plant entries start one foot abmc grah Any flash flooding that may Amendment 17 tr t
l l ABWR man ! Standard Plant _ iuv n l i result from tank rupture will drain away from the 1 site and cause w damage to site equipment. ! Additional specific provisions for flood protection inclu6e administrative procedures to assure that all watertight doors and hatch covers are locked in the event of a flood warning. If local scepage occurs through the walls, it is controlled by sumps and sump pumps. In the event of a flood, flood levels take a relatively long time to develop. This , allowsample lead timc to perform necessary i en.ergency actions for all acce.ses which need to be protected. The safety related components located below the design flood level inside a Seismic Category I structure are shown in the Section 1.2 plant layout drawings. All safety related components located blow the oesign flood level are protected using the hardened protection approach described above. O O Amendment 17 3411
AllWR mmm Sandard I'lant iot n (mv) 3.5 MISSILEl'ROTECrlON Criterlon 4 of 10cPit50 Appendix A, General Design Criteria for Nuclear Power Plants. The missile protection design basis for Seismic Category I structures, systems and Potential missiles that have been identified components is described in this section. A are listed and discussed in inter subsections, tabulation of safety related structures, systems, and components (both inside and outside Af ter a potential missile has been containment), their location, seismic category, identified, its statistient significance is and quality group classification is given in determined. A statistically significant raissile Table 3.21. General arrangement drawings is defined as a missile which could cause showing locations of the structures, systems, and unacceptable plant consequentes or violation of components are presented in Section 1.2. the guidelines of 10Cril100. Missiles considened are those that could The examination of potential missiles and i result from a plant.relatea failure or incident their consequences is done in the following i including failures within and outside of manner to determine statistically significant 1 containment, environeental generated missiles and missilen site proximity missi!cs. The structures, shields, and barriers that have been designed to (1) If the probability of occurrence of the withstand missile effects, the possible missile miss/le (P,) is determined to be less than loadings, and the procedures to which each 10' per irar, the missile is dismissed barrier has been designed to resist missile from further consideration because it is impact are described in detail. considered not to be statistically significant. 3.5.1 Missile Selection and (2) If (P 3) is found to be greater than 10' (q "j Descripilon per year, it is examined for its pro,, ability Cornponents and equipment are designed to have of impacting a design target (P,). a low potential for generation of rnissiles as n basic safety gircenution, in general, the design (3) If the projuct of (P j ) and (P,) is less that results in reduction cf missile generation ihan 10 p e r y e a r , t h e m'i s s il e i s potential promotes the long life and tsability of dismissed from further consideration, a component and is well within permissible limits of accepted codes and standards. (4) I f t h e p r o d u c t.7of (P j) and (P,) is greater than 10 per year, the miss'ile is Seismic Category I structures have been examined for its damage probability (P ). analyzed and designed to be protected against a If the combined probability (i.e.,7 P[3 x wide spectrum of missiles. For example, failure P.,x P3 - P )4 is less than 10' per of certain rotating or pressurized components of year, the missile is dismissed. equipment is considered to be cf sufficiently high probability and to presumab!y lead to (5) Finally, measures are taken to design generation of missiles. Ilowever, the generation acceptable protection again$t missiles with of missiles from other equipment is considered to (P 3 ) greater than 10 per year io be of low enough probability and is dismissed reduce (P,), (P,), and/3 r (P ), so from further consideration. Tornado generated tha; (P 4) is less'than 10 per'3year, missiles and missilt s resulting from activiti>:s particular to the site are also discussed in this Protection of essential structures, systems section. The missile protection criteria to and components is afforded by one or more of the which the plant has been analyzed comply with following practices: i r
/
( ) i %) l Amendment 3 m
ABWR wwm Sandard Plant stiv [! (1) Location of the system or component in an (6) Automatic depressuriration system relief individual missile proof structure; valves; (2) Physical separation of redundant systems or (7) Standby diesel generator system; components of the system for the missile trajectory path or calculated range; (8) CRD scrarn system (hydraulic and electrical); (3) Prevision of localired protection shields or barriers for systerns or components; (9) Fuel goo! cooling and cleanup system; (4) Design of the particular structure or (10) Remote e.hutdown panel; component to withstand the impact of the most damaging missile; (11) Reactor protection system; (5) Provision of design features on the (12) All containment isolation vahrs; potential missile source to prevent mir.sile generation; and/or (13) IIVAC emergency chilled water system; (6) Orientation of the potential missile source (14) IIVAC systems required during operation of to prevent ut. acceptable consequences due to items (1) through (12); and missile generation. (15) Electrical and control systems and wiring l The following criteria have been adopted to required for operation of items (1) through provide an acceptable design basis for the (14). l plant's capability to withstand the statistically significant missiles postulated inside the The following general criteria are used in reactor building. the design, manufacture, and inspection of equipment: (1) No loss of containment function as a result of missiles generated internal to (1) All pressurized equipment and sections of containment. piping that may periodically become isolated under pressure are provided with (2) Reasonabic assurance that a safe plant shut- pressure-relief valves acceptable under down condition can be achieved and ASME Code Section 111. The valves ensure rn ai n t a i n e d. that no pressure buildup in equipment or piping sections exceeds the design limits (3) Offsite exposure within the 10CFR100 of the materials invnived. guidelines for those potential missile damage events resulting in radiation (2) Compenents and equipment of the various activity release. systems are designed and built to the stan. dards established by the ASME Code or other The systems requiring protection are: equivalent industrial .'andard. A strin-gent quality rar' trol program is also en-(1) Reactor coolant pressure boundary; forced during manufacture, testing, and in-stallation. (2) Residual heat removal system; (3) VolumGnc and ultrasonic test ing chs te re-(3) High pressure corc flooder system; quired by code coupied with periodic inser-vice inspections of ntterials usev m com-(4) Reactor core isolation cooling system; ponents and equipment add furth:r assurance that any material flaws that could permit (5) Reactor building cooling water system; the generation of mailes aie dete:ted. 3.5-2 Amendment 17
ABWR me SlandaBlitint _, P!v_D requirement for redundant separation is protected against the ef fects of these (n)
~
met. Other redundant divisions are postulated pipe failures will be provided by the available for safe shutdown of the plant and applicant referencing the AllWR design (see no further evaluation is performed. Subsection 3.6.4.1, item 4 and 6). (4) If damage could occur to more than one liarricts or shicids that are identified as division of a redundant essential s) stem necessary by the llELSA evaluation (i.e., based within 30 f t of any high energy piping, on no specific break locations), are designed other protection in the form of barriers, for worst. case loads. The closest high energy shields, or enclosures is used. These pipe location and resultant loads are used to methods of protection are discussed in Sub. sire the barricts. section 3.6.1.3.2.3. Pipe whip restraints as discussed in Subsection 3.6.1.3.2.4 a r e 3.6.1J.2.4 Pipe Whip Reatndnin used if protection from whipping pipe is not possible by barriers and shields. Pipe whip restraints are used where pipe break protection requirements could not be 3.6.13.23 Itarrirrs, Shields, and 1:nclosur es satisfied using spatial separation, barriers, shields, or enclosures alone. Restraints are Protection requirements are met through the located based on the specific break locations protection afforded by the walls, floors, determined in accordance with Subsections columns, abutments, and foundations in many 3.6. 2.1. 4.3 a n d 3. 6. 2.1. 4. 4. Af ter the cases. Where adequate protection is not already restraints are located, the piping and essential present due to spatial separation or existing systems are evaluated for jet imp.ngement and plant features, additional barricts, deflectors, pipe whip. For those cases where jet or shields are identified as necessary to meet impingement damage could still occur, barriers, the functional protection requirementri, shields, or enclosures are utilized.
\
[V ' Ilarriers or shields that are identified as The design criteria for restraints is given in nete :ity by abe use of specific break locations Subsection 3.6.2.3.3. in the darwel: are designed for the specific lo ds .insociales with the particular break 3.6.133 Speelfic Protection Measures locatioc. (1) Nonessen'ial systems and system components The steam tunnel is rnade of reinforced are not required for the safe shutdown of concrete 2m thick. A steam tunnel subcompartment the reactor, not are they required for the analysis was performed for the postulated rupture limitation of the offsite release in the of a mainsteam line and for a feedwater line (see event of a pipe rupture. Ilowever, w hile Subsection 6.233.1). The peak pressure from a none of this equipment is needed during or mainsteam line break was found to be 11 psig, following a pipe break event, pipe whip The peak pressure from a feedwater line break was protection is considered where a resulting found to be 3.0 pdg. The steam tunnel is failure of a nonessential system or designed for the effects of an SSE coincident component could iaitiate or escalate the with high energy line break inside the steam pipe break event in an essential system or tunnel, Under this conservativ- ad component, er in another nonessential system combination, no failure in any portio the whose failure could affect an essential steam tunnel was found to occur; therefore, a system. high energy line break inride the steam tunnel will not effect control room habitability. (2) For high el.cegy piping systems penetrating through the containment, isolation vals es The MSIVs and the feedwater isolation and check are located as close to the containment as valves located inside the tunnel shall be possible, p designed for the effects of a line break. The Q details of how the MSIV and feedwater isolation and check valves functional capabilities are (3) The preuure, water level, and flow sensor instrumentation for those essential systems, Ame ndment 17 3 r,3
ABWR mme i Standard Plant 1a1.1! j which are required to function following a pipe rupture, are protected. (4) liigh. energy fluid system pipe whip restraints and protective measures are designed so that a postulated break in one pipe could not, in turn, lead to a rupture of other nearby piper, or components if the secondary rupture could result in consequences that would be considered unacceptable for the initial postulated break. (5) For any postulated pipe rupture, the structural integrity of the containment Etructure is maintained, in addition, for those postulated ruptures classified as a loss of reactor coolant, the design leak tightness of the containment fission product barrier is maintained. (6) Safety / relief valves (SRV) and :he reactor core isolation cooling (RCIC) system stearn-line are located and restralned so that a pipe failure would not prevent depressuri. ration. O O Amendment 17 3 ("5 I
f ABWR mmo sandanuhnt mn g (7) Separation is provided to preserve the those systems or portions of splems that, independence of the low pressure flooder during normal plant conditions (as defined in (v) (LPI L) sptems. Subscetion 3 6.1.1.3(1)),4rc cither in operation or are maintained pressurized under conditions (8) Protection ior the I MCl(D scram insert lines where either or both of the following are met: is not required since the motor operation of the FMCRD can adequately insert the control (1) masimum operating temperature excccds rods even with a complete loss of insert 200* F, or lines (See Subscetion 3.6.2.1.6.1). (2) ruasimum operating prest.ure esteeds 275 psig. (9) The a wpe of steam, water, combustible or cor e fluids, gases, and heat in the 3.6.2.1.2 Ikunition of Moderate.1:nerva l'lutd even a pipe rupture do not preclude: Spierns. (a) Acussibility to any areas required to Moderate energy fluid splems are defined to cope with the postulated pipe rupture; be those r.ptems or portions of sprems that, during normal plant conditions (as defined in (b) liabitability of the controt room; or Subsection 3. 6.1.1. 3. ( 1 ) ) , are either in operation or sie maintained pressurited (abose (c) T h e a biiit y of e s s e n tiaI atmospheric pressure) under conditions where instrumentation, elect ric powe r both of the following are met: supplies, components, and controls to perform their safety related function. (1) maximum oprinting temperature in 200*l or less, and 3.6.2 Deterrnination ofIltcak locations and Dynatnic Effects (21 maximum operating pressure is 275 psig or p () Associated ullh the l'ostulated Rupture ofI'iping less. Piping systems are classified as information concerning break and crack moderate energy sptems when they operate as location criteria and n.ethods of analpis for high energy piping for only short operational dynamic cifects is presented in this Subsection, periods in performing their system function but, The location criteria and methods of analph are for the major operational period, qualify as needed to esaluate the dynamic effects av.ociated moderate energy fluid splems. An operational with postulated breaks and cracks in '4igh and period is considered short if the total fraction moderate energy fluid system p!;.ing inside and of tirne that the sptem operates within the outside of primary containmet.. This infortnation pressure ternperature conditions specified for provides the basis for the requirements for the high energy fluid systems it. Ictr. than two protection of essential structures, sptems, and percent of the total time that the spiem components defincd in introduction of Section operates as a moderate energy fluid sptem. 3.6. 3.6.2.1,3 Postulated Pipe lirraks and Cracks l 3.6.2.1 Criteria Used to ikfine lireak and l Crack location and Configuration A postulated pipe break is defined as a i r,udden gross f ailure of the pressure boundary The following subsections establish the either in the forrn of a complete circumferential criteria for the location and configuration of severance (guillotine break) or a sudden postulated breaks and cracks. longitudinal split without pipe severance, and is postulated for high energy fluid sptems 3.6.2.1.1 Definition of liigh 1:ncryy I'luid only. For moderate energy fluid r.ptem, pipe Splems failures are limited to postulation of cracks in piping and branch runs. These cracks affect the liigh energy fluid sprems are defined to be surrounding environmental conditions only and do Amendment 7 W l
ABWR mam sty n Standard Plant including deadweight and SSE (inertial) (1) A summary of the dyr.amic analpes (n) components, applicable to high energy piping splems in accordance with Subsection 3.6.2.5 of Shielded Mctal Are (SMAW) and Submerced An Regulatory Guide 1.70. This shall diAW) Welk include: The flow stress used to construct the master (a) curve is $1 ksi Sketches of applicable piping systems showing
- he location, sire and orientation of postulated The value of SI used to enter the master pipe becaks and the location of pipe whip curve for SMAW and SAW is restraints and jet impingement barriers.
Si = M (Pm + Pb + Pe) Z (8) (b) A summary of the data developed to select where postulated break locations including calculated stress intensities, cumulatise usage factors and Ib stress ranges as delineated in llTP MEll 31.
- the combined primary b,nding stress, including deadweight and seismic components. (2) for failu e in Ihe moderatc energy piping systems listed in Table 3.6 6, Pe descriptions showing how safety related 5
= combined expansion stress at normal sptems are protected from the resulting E operation, jets, flooding and other adverse environmental ef fects.
Z = 1.15 [1.0 + 0.013 (OD-4)] for SMAW, (9) (3) Identification of protective measures provided against the effects of q (d,) Z = 1.30 [1.0 + 0.010 (OD 4)] for SAW, postulated pipe failures in each of the $ (10) splems listed in Tables 3.61,3.62 and 3.6 4. and (4) The details of how the MSIV functional OD- capability is protected against the h pipe outer diameter in inches. cffeets of pontulated pipe failures. 7 When the allowable flaw length is determined (5) Typical e xa m ple 5, if any, where from the master curve at the appropriate Si protection for safety related systems value, it can be used to determine if the and components against the dynamic required margins on load and flaw size are effects of pipe failures include their 6 met using the following procedure. enclosur e in suitably designed 5 structures or compartments (including for the rnethod of load combination described any additional drainage system or in item (5), let M = 1.4, and if the equipment environmental qualification allowable flaw length from the master curve needs). is at least equal to the leakage size flaw, then the margin on load is met. (6) The details of how the feedwater line check and feedwater isolation valves 3.6.4 Interfaces functional capabilities are protected against the effects of postulated pipe 3.6.4.1 Details of Pipe fireak Analpis itesults failures. and Protection Methods l (q
/ The following shall be provided by the 3.6.4 2 leak liefore Itreak Analph iteport l
applicant referencing the A13WR design (See As required by Reference 1, an Lilll analpis
- Subsection 3.6.2.5)
Amendment 17 3 fr27 i
ABWR 2mmu sov n Slandard Plant report shall be prepared for the piping systems proposed for inclusion from the analyses for the dynamic effects due to their failure. The report shall include only the piping stress analysis results for the piping systems analyzed and reported for LBB in Appendix 3F in order to show that the piping stresses are within the stress ; levels assumed in Appendia 3F (See Subsce- I tion 3.6.3). 3.6.5 References
- 1. Modification of General Design Criterion 4 Requirernents for Protection Against Dynamic Effects of Postulated Pipe Rupture, Federal Recistet Volume $2, No. 207 Rules and Regulations, Pages 41288 to 41295, October 27,1987
- 2. RELAP 3, A Computer Program for Reactor Blowdown Analysis, IN 1321, issued June 1970, Reactor Technolocv TID 4500.
- 3. Moody, F. J., Fluid Reactor and impingement loads, Vol.1, ASCE Specialty Conference on Structural Design of Nuclear Flant Facilities pp. 219 262, December 1973.
- 4. Standard Review Plan,' Public Comments Solicited, Federal Recister. Volume $2 No.
167 Notices, Pages 32626 to 32633, August 28,1987. O 3427.1 Amendment 17
ABWR mmau ! Standard Plant _ RI'Y H
- Table 3.6 2 ESSENTIAL SYSTEMS, COMPONENTS, AND EQUIPMENT
- FOR l POSTULATED PIPE FAILURES OUTSIDE CON'I AINMENT !
- 1. Containment Isolation System and containtnent boundary.
- 2. Reactor Protection System (SCRAh! signals)
- 3. Core Cooling systems (a) IIPCF(B or C) or RCIC !
. (b) RilR.LPil.(A or B or C) + ADS (c) R11R shutdown cooling mode (two loops)
(d) RilR suppression pool cooling mode (two loops)
- 4. Flow restrictors 4
- 5. Controf rcom habitability
- 6. Spent fuel pool cooling
- 7. Standby gas treatment
- 8. - The following equipment / systems or portions thereof required to assure the proper operation of those essential items listed in items 1 through 7.
(a) Class 1E electrical systems, ac and de (including diesel generator syst em, 6900,480 and 120V ac, and 125V de emergency buses, motor control centers, switchgear, batteries, auxiliary shutdown control panel, and distribution systems). (b) Reactor fluilding Cooling water to the following: .
- (1) Room coolers (2) Pump coolers (motors and seals)
(3) Diesel generator auxiliary system coolers l
-(4) Electrical switchgear coolers (5) RilR heat exchangers l
- The essential items listed in this table are protected in accordance with Subsection 3.6.1 consistent with the particular pipe break evaluated.
Amendment 17 3 G30
ABWR auimat Riv li Standard Plant Table 3.6 2 ESSENTIAL SYSTEMS, COMPONENTS, AND EQ'9PMENT* FOR POSTULATED PIPE FAILURES OLTI' SIDE CONTAIN ENT (Continued) (6) I'PC heat exchangers (7) IIECW refrigerators (c) IIVAC (d) Irtstrumentation (including post accident tnonitoring) (c) Fire Water System (f) IIVAC Emergency Cooling Water System (g) Process Sampling System O O Amendment 17 3 &NJ 1
ABWR unmn Standard Plant wv. ti Table 3.6 4 I!!Gli ENERGY l'It'ING OUTSIDE CONTAINMENT Piping System
- Main Steam Main Steam Drains Steam supply to RCIC Turbine CRD(to and from IICU)
RilR(injection to feedwater fiom nearest check vahes in the RilR lines) Reactor Water Cleanup (to Feedwater via RilR and to first inlet vahe to RPV head spray) Reactor Water Cleanup (pumps suction and discharge) Fluid systenss operating at high energy levels less than 2 percent of the total time are not included. These systerss are classified rnodcrate energy systems, (l.c., HPCF, RCIC, SAh! and SLCS). O Amendrnent 17 3 432
ABWR mme klY A Standard Plant Table 3.6 5 COMPARISON OF PDA AND NSC CODE O lirtak Restraint Restraint % of Design l'ipe ladertl. Indenti. No. Denection Restraint Denection neation neation of liars lead (Klps) (In.) Denect!on (In.) (Fle.3.621 fric. 3.6 21 EM MC EIM b'.SC flM b'EC ELM MC Elh hSC RC1j RCR1 5 5 803.2 7833 6.57 7.926 79.93 96.4 17.72 15.58 RC2LL RCH1 5 5 766.4 458.4 14.99 7.495 125 62.6 35.83 24.52 RC31,L RCR2 6 6 747.0 639.7 2.27 3.73 27.65 4535 17.16 20.11 RC3LL RCR2 6 6 796.6 780 3 10.22 1034 57.0 59.6 41.4S 43.0 RCR3 5 5 846.0 838.4 7.64 8.05 92.95 97.98 18.87 16 43 RC4LL RCR3 8 8 1019.0 1073.9 5.43 4.21 99.23 76.85 2338 17.25 RC4LL RC4CV RCR3 8 8 1260.7 1275.0 4.49 5.58 8037 99.89 22.56 18.7.1 RC6AV RCR3 8 8 928.5 722.5 1.22 1.77 22.46 31.7 2M8 9539 RC7j RCR7 6 6 9533 801.6 6.20 5.75 76.4 79.12 16 46 21 63 RCR6 4 4 599.0 0 8.28 0 112.46 0 26.75 839 RC8LL RCR7 6 6 895.0 0 S.16 0 110.76 0 29316 839 RCSLL RCR6 4 4 575.8 520.16 4.16 5.53 50 63 6733 13.2 14.56 RC9CV RCR8 6 6 830.2 546.8 11.400 6.815 95.29 56.9 .w 612 2624 RC9LL RCR8 6 6 818 3 493 6 10.98 5.99 91.72 $0.07 31.404 23.71 RC11A RC13 RCR10 4 4 668.4 478.0 5.87 3.66 93.5 5839 1337 10.44 RC16 RCR11 4 4 687.4 518.4 6.59 438 105 69.86 1537 10.22 RCR20 S 8 285.0 309.6 2.83 5.58 46 3 95.92 15.45 13.46 RC14CV RC14LL RCR20 8 8 116 3 129.9 0.% 336 10.5 37.1 22.13 23.56 9 Amendment 1 3431
ABWR mum Stan_dard Plant Riv n (] (d) Regulatory Guide 1.29, Seismic Design V Classification; (c) Regulatory Guide 1.31, Control of (c) Deleted Stainless Steel Welding; (f) Regulatory Guide 1.44, Control of the Use of Scru,itized Stainless Stect; (f) ANSI N45.4,1 eakage Rate Testing of Containment Structures for Nuclear (g) Regulatory Guide 1.$$, Concrete Reactors: Placement in Category 1 Structures; (g) ANSI N101.2, Protective Contings (h) Regulatory Guide 1.60, Design Response (Paints) for Light Water Nuclear Spectra for Seismic Design of Nuclear Reactor Containment Facilities; and Power Plants; (h) Ar'S! N101.4, Quality Assurance for (i) Regulatory Guide 1.61, Quality Assurance Protective Coatings Applied to Nuclear Requirements for the Design of Nuclear Patilities; Power Plants; (10) Steel Structures Paintiag Council Standards (',) R c ,ulat or y C uid e 1.69, Con c r e t e Radiation-Sbicids for Nuclear Power (a) SSPC PA 1, Shop, Field and hiaintenance Plants. Painting; (1) Regulatory Guide 1,76, Design !! asis (b) SSPC-PA 2, hicasurernent of Paint Pilm Tornado. Thickness with hiagnetic Gages; n (m) Regulatory Guide 1.142, Safety Related (c) SSPC SP-1, Solvent Cleaning; Concrete Structures for Nuclear Power Plants (Othen than Reactor Vessels and (d) SSPC SP 5, White hittallilast Cleaning; Containment); and (c) SSPC SP-6, Cominercial lilast Cleanirg; (n) Regulatory Guide 1.94, Quality As.,urance and Rr quircruents for Installation, inspec-tion, and Testing of Structural Concrete (f) SSPC SP 10, Near White Illast Cleaning; and Structural Steel During the Con. struction Phase of Nuclear Power Plants. (11) ACI ASCE Committec 326, Shear and Diagonal Tenslore, ACI h1anual of Concrete Practice, (9) ANSl: Part 2; (a) ANSI A58.1,13uilding Code Requirements (12) Applicable AST A1 Specifications for for hiinimum Design leads in Iluilding and h1aterials and Standards; and Other Structures; (13) AASilTO Standard Specifications for liighway (ti) ANSI N5.12, Protective Coatings (Paint) liridges for truck loading area. for the Nuclear Industry; 314.2.2 ControlIlullding (c) NOA 1, Quality Assurance Program Requirements for Nuclean Pacilities and Refer to Subsection 3A4.2.1. NQA 1A, Adenda to ANSl/AShiH NOA 1; Add NRC Rules and Regulations Title 10, Chap. (d) Deleted ter 1, Code of Federal Regulations, Part 73.2 ( and 73.55. Amendment 17 4 22
ABWR msima SU)ndard Plant __ nv n 3.N.J.2.3 Radust llullding Substructure Concrete fIoors and sIabs (including roofs) 200 psf. The radwaste building substructure shall be d: signed using the same codes and standards as Stairs, stair platforrns, grating floors, and platforms .100 psf. h the reactor building. Refer to Subsection Concrete roofs, live or snow load 3.8.4.2.1 for a complete list. (not concurre nt) . 50 psf. Construction live load on floor In addition, the non Seismic Category I framing in addition to dead reinf orced concrete portion of the weight of floor . 50 psf'. superstructure is designed according to the seismic provisions of Section 2314 of the Ro = pipe reactions during normal uniform building code. operating or shutdown conditions based on the most critical 3J.4.2.4 Seismte Category i Cable Tray and t r aosie at or steady. state Condult Supports condition (1) All codes, standards, and specifications ap- R, = pipe reactions under thermal plicabic to the building structures shall conditions generated by the also apply to cabic tray and conduit postulated break and including supports. Ro (2) AISI 50 673, Specification for the Design of Yr = equivalent static load on a Cold. formed Steel Structural Members. structure generated by Ihe r e action on the broken (3) NEMA, Fittings and Supports for Conduit and high cnergy pipe during ihe Cable Assemblics. postulated break and including a calculated dynamic factor to 3.8.4J Loads and Load Combinations account for the dynamic nature of 3J.4.3.1 Reactor Dullding the load h Yj = jet impingement equivalent static The temperature and pressure loads caused by load on a structure generated by a LOCA do not occur on the reactor building. The the postuIaied break aad reactor building ventilation system is designed including a calculated dynamic to keep the building within operating design factor to account for the dynamic l conditions, nature of the load. 3.8.4.3.1.1 Loads and Notations Ym = missile impact equivalent static load on a structure generated by Loads and notations are as follows or during the postulated break, like pipe whipping, and including D = dead load of structur; plus any a calculated dynamic factor to other perr.anent load account for the dynamic nature of the load. L = conventional floor or roof live loads, movable equipment loads, and W = wind force (Subsection 33.1.) other variable loads such as construction loads. The following _ live Ioads ate used:
- If the actual constmetion live load is grater than this vahue a desigs check of the simctures will be made.
Amendment 8 3 8-23 O
i ABWR m im^n :
$landard Plani RI;V 11 l i
O T sie 3.. 4 i CODES, STANDARDS, SPECIFICATIONS, AND REGUlXfl0NS i USED IN THE DESIGN AND CONSTRUCr10N OF SElSMIC CATEGORY l INTERNAL STRUCfURES OF THE CONTAINMENT (Continued) SPEClflCATION SPEClflCATION l REFERENCE OR STANDARD ; NUMBER DESIGNATION TI11.11 e 15 ANSI /AISCN690 Specification for the Design, Fabrication, and Election of Steel Salcty Related Structures for y Nuclear Facilities ! 16 AWS DI.1 Structural Welding Code , 17 NCIG-02 Visual Weld Acceptance Criteria for Structural ! Welding at Nuclear Power Plantr, 18 . ANSI /ASME Ouality Assurance Program Requirements for ' i NQA.11986 Nuclear Facilities 19 (Deleted) i 20 NRC Regulatory Ouality Assurance Hequirements for Installation,
-- Guide 1.94 Inspection, and Testing of Structural Concretc [
and Structural Stect During the Construction Phase of , Nuclear Power Plants , i 21 NRC Regulatory Materials for Concrete Containments ; Guide 1.136 (Artide CC 2000 of the Code for Concrete Reactor Vessels and Containments) . 22 NRC Regulatory Safety Related Concrete Structures for Nuclear Power l
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Guide 1.142 Plants (Other than Reactor Vessels and Containments) Eyplanation of Abbretlations . ACI American Concrete Institute . ; AISC American inr.titute of Steel Construction l AISI American iron and SteelInstitutc ANSI American National Standards Institute ASME American Society for Mechanical Engineers , AWS American Welding Society ; NCIG Nuclear Construction issues Grcup , NRC Nuclear Regulatory Commission NOTES:
- 1. Unters specified, the Edition of the Specification or Standard shall be the latest issurd-for industry use.
Amendment 17 3M3 r
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d
- TAftt.E3OS
_y LOAD COMftfMAT10N,1,OAD FACTORS AND ACCEPTANCE CRITERIA 3 i-s FOR REINFORCED CONCRETE STRUCTURES INSIDE Tite CONTAINMENT (1),(2) jw a as as J
,f LOAD COMBINATsOM LnAn coMDfTIOM ACCEMAtoCE CalTFatA W Drwitrtit1M Me D L P. P. P. FR F. T, 7 7 F E W W 9, R, @ 39488I SECA t' Ars O ALL Tese i t9 99 99 99 3 88amel 1 89 99 I9 89 89 89 99 5 Se 54 9.7 89 87 8.7 8.7 17 M 8 SS 3.3 89 81 f.3 8.1 8.3 U l Sm de f4 f7 89 89 8.7 8.7 f .7 U P,.tr.=ensest 40 t GS t.1 89 11 8e a3 S.3 9.3 U Se 14 8.7 89 17 S.7 8.7 9.7 U 36 8 91 9.3 89 81 t_3 f.3 8.3 1.3 U I
r- 9 19 39 89 st 89 to 19 99 U
- P,.6 east ? $9 99 99 89 89 99 89 8.9 E 1
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- Ab -sey 11 88 89 8 21 19 8 21 39 99 89 *
- U Se,w, Ise t9 99 1.25 89 8 21 89 89 99 89 *
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- U r,.6 net t56 39 9 9 *1 9 e9 e9 19 89 *
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- 1. The 4.4e see em,h8 in 5.4 eem 3 8 3 3_
- 2. Ges=e se Nie 2.TeMe 3 e np y
- 3. S = W www.rb en e-wen e +-i e 8 de e., Awr r-4e seo.- m tw. 2 3g J
K U e e 9 J e. - ge e.w e fe e-,4 I le g-, AG 14e [ Y 4 Ga.neesf+=e3,leMe18 1) j 3. Os.neseN e6teMe3aI) at 6 =e ee Name 7.T.Me 3 8 8) 220.13 220.14
ABWR 2mme SatidantPlet en O Subsection 3.9.2.5. Dynamic analysis is pet. formed by coupling the lumped man model of the 3.9.53.3 Design leading Categories reactor vessel and internals with the building The basis for determining faulted dynamis model to determine the system natural frequencies event loads on the reador internals is sbown in and node shapes. The relative displacement, Sections 3.7, 3.8 and Subsections 3.9.2.5, acctleration, and load response is then deter- 3.9.5.2.3 a nd 3.9.5.2.4. Table 3.9 2 shows the mined by either the time. history rnethod or the load combinations used in the analysis, response spectrum method. Core support structures and safety class 3.9.53 Design liases internals stress limits are consistent with ASht!! Code Section 111, Subsection NO. For 3.9.53.1 Safety Dcsign Itases these components. Level A, D, C, and D senice limits are applied to the normal, upset, The reactor internals including core support emergency, and f aulted loading conditions, structures shall meet the following safety design respectively, as defined in the design bases: specification. Stress intensity and other design limits are discussed in Subsections (1) The reactor vessel norries and internals 3.9.5.3.5 a nd 3.9.5.3.6 shall be so arranged as to provide a floodable volume in which the core can be 3.9.5.3.4 Response of internals Due to Steam adequately cooled in the event of a breach the lirtak Accident in the nuclear system process barrier external to the reactor vessel; As described in Subsection 3.9.5.2.3.2, t h e maximum pressure loads acting on the reactor (2) Deformation of internals shall be limited to internal components result from steam line break assure that the control rods and core upstream of the main steam isolation valve and, standby cooling systems can perform their on some components, the loads are greatest with safety related functions; and operation at the minimum power associated with the masimum core flow (Table 3.9 3, Case 2), (3) hicchanical design of applicable strectures This has been substantiated by the analytical shall assure that safety design bases (1) compa:ison of liquid versus steam line breaks and (2) are satisfied 50 that the safe and by the investigation of the effects of core shutdown of the plant and removal of decay power and core flow, heat are not impaired, it has also been pointed out that, although 3.9.5.3.2 Power Generation Design liases possible, it is not probable that the reactor would be operating at the rather abnormal The reactor internals including core support condition of minimum power and maximum core structures shall be designed to the following flow, htore realistically, the reactor would be power generation design bases: at or near a full power condition and thus the maximum pressure loads acting on the internal (1) The internals shall provide the proper components would be as listed under Case 1 in coolant distribution during all anticipated Table 3.9 3. normal operating conditions to full power operation of the core without fuel damaget 3.9.53.5 Stress and Fatigue Limits for Core Support Structures (2) The internals shall be arranged to facilitate refueling operations; and The design and construction of the core support structures are in accordance with ash 1E (3) The internals shall be designed to Code Section Ill, Subsection NO. f acilitate inspection. O O Amendment 7 3941
ABWR m=w niv 9 Santard l I'lant 3.9.5.3.6 Stress, Deformation, and fatigue 3,9.6 Insenice Testing ofl'urnps and Yahes 1.imits for Safety Class and Other Reactor Internals (racept Core Support Structurcu inser ice testing of safety related pumps and ; h vahen will be performed in accordance with the G for safety class reactor internals, the stress requirements of Section XI, Subsection lWP and deformation and fatigue criteria listed in Tables IWV, of the AshtE Code. Table 3.9 8 lists the 3.9 4 through 3.9 7 are based on the critcria insenice testing parameters and frequencies for established in applicable codes and standards for the safety related pumps and vahes. Valves similar equipment, by manufacturers standards, or having a containment isolation function are also by empirical methodt based on field experience noted in the listing. Code testing flexibility and testing. For the quantity SFmin (minimum in the ash 1E/ ANSI Okht part 6 for pumps and part safety factor) appearing in those tables, the 10 for valsen produced no need for relief following values are used: reo .ests. A review of field experience for typical !!WR testing problems also showed the SenIce Senice gr Code encompassed common relief requests.
],nt] Condition ,,_mLt3 Inservice inspection is discussed in Subse(tion 5.2.4 and Section 6.6.
A Normal 2.25 11 Upset 2.25 Details of the insenice testing program, C Emergency 1.5 including test schedules and frequencies will be D Faulted 1.125 reported in the inservice inspection and testing plan which will be prmided by the applicant Components inside the reactor pressure sessel referencing the AllWR design. The plan will such as control rods which must move during integrate the applicable test requirements for accident condition have been examined to safety related pumps and valves including those determine if adequate clearances exist during listed in the tuhnical specifications, Chapict emergency and faulted conditions. No mechanical 16, and the containment isolation vah es, clearance problems hase been identified. The forcing functions applicable to the reactor S ubsection 6.2.4. An example is the periodic leak testing of the reactor coolant pressure h internals are discussed in Subsection 3.9.2.5. isolation valves in Table 3.9 9 will be 3@ performed in accordance with Chapter 16 RA The design criteria, loading conditions, and Surveillance Requirement 3.6.1.5.10. This plan analyses that provide the basis for the design of will include bawline pre service testing to the safety cless reactor internals other than the support the periodic in service testing of the core support structures meet the guidelines of components. Depending on the test results, the NG 3000 and are constructed so as not to plan will provide a commitment to disassemble adversely affect the integrity of the core and inspect the safety related pumps and vahes support structures (NG il22). when limits of Subsection IWP or IWV are exceeded,as described in the following The design requirements for equipment paragraphs. The primary elements of this plan, classified as non safety (other) class internals including :he requirements of Generic Letter (e.g., steam dryers and shroud heads) are 8910 for motor operated salves, are delineated specified with appropriate consideration of the in the subsections to follow. (See Subsection intended service of the equipment and expected 3.9,7.3 for interface requirements). plant and environmental conditions under which it will operate. Where Code design requirements are 3.9.6.1 Insenice Testing of Safet).Related not applicable, accepted industry or engineering Putrps practices are used. The ABWR safety related pumps and piping ;; configurations accommodate inservice testing at 8 a flow rate at least as large as the maximum ! design flow for the pump. In addition, the O Amendment 17 3944 l _
ABWR mme Standard Ihtnt kiv H f) 3.11 ENVil(ONMENTALQUALIFICATION 3.11.1 Equiptuent identl0 cation and v OF SAFETY l(ELATED MECilANICAL Environniental Canditions AND ELECTitlCAL EQUll' MENT Safety related electrical equipment within i This section delines the environmental the scope of this $cction includes all three conditions with respect to limiting design categories of 10CTR50.49(b) (Reference 1), conditions for all the safety related mechanical Safety related mechanical equipment (e.g., and electrical equipment, and docurnents the pumps, motor operated valves, safety relief qualification snethods and procedures employed to valves, and check valves) are as defined and demonstrate the capability of this equipment to identified in Section 3.2. perform safety related functions when exposed to the environmental conditions in their respective A list of all safety related electrical and locations. The safety related equipment within mechanical equipment that is located in a harsh the scope of this section are defined in environrnent area will be included in the Subsection 3.11.1. Dynamic qualification is Emironmental Qualification Document (EQD) to be addressed in Sections 3.9 and 3.10 for Seismic prepared as rnentioned in Subsection 3.11 A Category I mechanical and clcctrical equipment, respectively. Environtnental conditions for the rones where safety related equipment is located are Limiting design conditions include the calculated for normal, abnormal, test, accident following: and post accident conditions and are documented in Appendix 31. Equipment Qualification (1) Normal Operating Conditions planned, Environmental Design Criteria (EOEDC). purposeful, unrestricted reactor operating Ensironmental conditions are tabulated by tones, modes including startup, power range, hot each rone defining a specific area in the standby (condenser available), shutdown, aad plant. Typical equipment in the noted zones is O r <n >ine -edes she*n in the fi 8n es in se<tien < 2. >2 3 nd Appendix 9A. (2) Abnormal Opetuing Conditions any deviation from normal conditions antierpated Environmental parameters include temperature, to occur often enough that the design should pressure, relative humidity, and neutron dose include a capability to withstand the rate and integraied dose. Radiation dose for conditions without operational impairment; gamma and beta data for both normal and accident conditions will be provided by applicant refer-(3) Test Conditions - planned testing including encing the AllWR design in accordance with the pre operational tests; interface requirctnents in Subsection 12.2.3.1. The radiation requirements are site specific (4) Accident Conditions a single esent not documentation owing to the need to model reasonably expected during the course of specific equipment which is applicant deter-plant operation that has been hypothesiicd mined, the llVAC detailed modeling, and the for analysis purposes or postulated from evolving considerations in the area of accident unlikely but possible situations or that has source terms are expected to generate signi. the potential to cause a release of ficantly differing radiation requirements. radioactive material (a reactor ecolant Where applicable, these parameters are given in pressure boundary rupture may qualify as an terms of time based profiles. accident; a fuel cladding defect does not); and The magnitude and 60 year frequency of occurrence of signifi: ant deviations from normal (5) Post Accident Conditions - during the length plant environments in the zones have of time the equipment rnust perform its insignificant effects on equipment total thermal safety related function and must remain in a normal aging or accident aging. Abnormal safe mode after the safety related function conditions are mershadowed by the normal or p\ is performed. accident conditions in the Appendix 31 tables. Amendment 17 3114
1 ABWR :wme HIv H l Silndard I'llint j Margin is defined as the difference bctween the I most severe specified service conditions of the h, ' plant and the conditions used for qualification. Margins shall be included in the qualification parameters to account for normal variations in commercial proJuction of equipment and reasonable errors in d-fining satisfactory performance. The environmental parameters s.hown in the Appendix 31 tables do not include margins. Some mechanical and clectrical equipment may be required to perform an intended safety J function between sninutes of the occurrence of the j event but less than 10 hours into the event. Such equipment shall be shown to remain functional in the accident environment for period of at least I hour in excess of the time assumed I in the accident analysis unless a time margin of i less than one hout can be justified. Such justification will include for each piece of equipment: (1) consideration of a spectrum of brenhs; (2) the potential need for the equipment later in the event or during recovery operations; (3) a determination that failure of the equipment after performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) deterrnination that the margin applied to the minimum operability, time, when combined with other test margins, will account for the uncertainties associated with the use of analytical techniques in the derivation of environmental parameters, the number of units tested, production tolerances, and test equipment inaccurancies. The environmental conditions shown in the Appendix 31 tables are upper-bound envelopes used to establish the environmental design and qualification bases of safety related equipment. The upper bound envelopes indicate that the zone data reflects the worse case expected environment produced by a compendium of accidet; conditions. Estimated chemical environmental conditions are also reported is Appendix 31. 3.11.2 Qualification Tests and Analyses Safety related electrical equipment that is located in a harsh environment is qualified by test or other methods as described in IEEE 323 O Amendment 17 311-11
AIMR m ai S11tildard PlaRt f(LLD and permitted by 10CFR50.49(f) (Refercnce 1). 3.ll.3 Qualification Test 1(esults Equipment type test is the preferted method of qualification. The r e sult s, of q ualific a tion t e s t $ fo Safety.related equipnunt will be documented Safety related mechanical equipment that is maintained, and reported as mentioned in located in a harr.h environment is qualified by Subsection 3.11.6. analysis of materials data which are generally based on test and operating experience. 3.ll A less ofllealing, Ventilating, and Alt Conditioning The qualification methodology is described in detail in the NRC approved licensing Topical To ensure that loss of heating, ventilating, Report on GE's environmental qualification and air conditioning (llVAC) system does not program (Reference 2). This report also adversely affect the operability of safety. addresses compliance with the applicable portions related controls and electrical equipment in of the General Design Criteria of 10CTR$0, buildings and areas sersed by safety.related Appendix A, and the Quality Assurance Criteria of IIVAC systems, the llVAC systems serving these 10CFR$0, Appendix B. Additionally, the repott areas meet the single.f ailure criterion. describes conformance to NUREG 05ss (Reference Section 9.4 describes the safetyociated IIVAC 3), and Regulatory Guides and IEEE Standards systems including the detailed safe t y referenced in Section 3.11 of NUREG.0800 cvaluations. The loss of ventilation (Standard Review Plan). calculations are based on maximum bcat loads and consider operation of all operable equipment Mild environment equipmerit is that equipment regardless of safety classification. which, during or after a design basis esent (DHE, as defined in Reference 2), does not experience 3.11.5 Estintated Chemical and l{adiation an environment that is significantly more severe Environment than that existing during normal and abnormal events. Additionally, equipment that experiences 3.11.5.1 Chemical 1:nstronment the environment of a D!!E can be treated as if it were in a mild environment if the equipment falls Equipment located in the containrr nt drywell into either of the following categories: and wetwell is potentially subjec t to w ater spray modes of the RilR system. In addition, (1) The equipment accomplishes its safety func- equipment in the lower por tions of the tion prior to experiencing the environment containment is potentially subject tol of the DBE and the equipment will not fail submergence. The chemical composition and in a manner detrimental to plant safety, or resulting pil to which safetprelated equipment is exposed during normal operation and design (2) The equipment is not needed to mitigate the basis accident conditions is reported in DHE and the equipment will not fail in a Appe ndix 31. manner detrimental to plant safety. Sampling stations are provided for periodic The sendors of mild environment equipment are analysis of renctor water, refueling and fuel required to submit a certificate of compliance storage pool water, and suppression pool water certifying that the equipment has been qualified to assure compliance with operational limits of for the requirements specified to assure its the plant technical specifications. required safety related function in its ap. plicable ensitonmen: This equipment is qual- 3.11.5.2 Radiation Ensironment ified for dynamic loads as addressed in Sections 3.9 a n d 3.10. Further, a surveillance and Safetv related systems and components are maintenance program will be deseloped to ensure designed to perform their safety related l equipment operability during its designed hfe. function when exposed to the normal operational radia.mn levels and accident radiation les elt l Amendme nt 17 s i10
AllWR 23^mean St;mdard I'lant tu v .11 The normal operational exposure is based on the (3) Interim Staff Position on Ensironmental f radiation sources presided in Chapter 12. Qualification of Safety-Related filectrical Equipment, NUREG 0588. Radiation sources associated with the DBA and developed in accor dance with NUREG 0588 (Reference 3) are provided in Chapter 15. Integrated doseu associated with normal plant operation and the design barils accident condition for various plant compartments are described in Appendix 31, 3.11.6 Interfaces 3.11.6.1 1:nstronmental Qualification Document The EOD shall be prepared summarizing the qualification results for all safety related equipment. The EQD shallinclude the following: (1) The test environmental parnmeters and the methodology used to qualify the equipment located in harsh as well as mild e nvironm e nts shall be ide ntified. (2) A st.mmary of environmental conditions and qualified conditions for the safety related equipment located in a harsh environment zone shall be presented in the system component evaluation work (SCEW) sheets as desetibed in Table I 1 of GE's ensironmental qualification program (Reference 2). The SCEW sheets shall be compiled in the EOD. 3.11.6.2 Ensironmental Qualification Records The results of the qualification tests shall be recorded and maintained in an auditable file. 3.11.7 References (1) Code of Federal Regulations, Title 10, Chapter 1. Part 50, Paragraph 50.49, Environmental Qualification of Electric Equipment important to Safety for Nuclear Power Plant. (2) General Electric Environmental Qualifica00n Program,NEDE.243261-P, Propriet try Document, January 1983. g Amendment 17 3i13
ABWR naam^n-Rev il Standard Plant Q APPENDIX 3G TAHLE OF CONTENTS Sudan lide l' age 3G SEISMIC ANALYSIS RESULTS 3G.! Ih]yd)DUC110N 3G.1 1 3G.2 RI: ACTOR Hull DING SSI A_NAIXSIS RESULTS 30.2 1 l 3G.2.1 Effect of Soil Stiffness 3G.21 , l 30.2.2 Effect of Soil Depth 30.2M 30.23 Effect of Groundwater Table 30.2 2 30.2.4 Effect of Adjacent fluildings 30.2 2 30.2.5 Effect of 3 D 30.2 3 30.2.6 Effect of Alternate Approach 3G.2-3 30.2.7 Summary 3 0.2 4 j 3G3 ADDITIONALl'ARAMETERS ANAINSES 30 3.1 Effect of RSW Stabilizer 3G31 3G3.2 Effect of Revised ControlIluilding 3G31 3G.4 !ilIE.ENVEl OPE SElSMIC LOADS FOR RE ACTOR llUll. DING COMI'I EX 30.4.1 Maximum Structural Loads 30.4 1 30.4.2 Floor Response Spectra 3 G.4 1 3G.$ CONTROL llUILDING ANALYSIS 30.5.1 SSI Analytical Results 3G.51 3G.5.2 Site Envelope Seismic Loads 3 G.5 1 LO w.ii A.r.icndment 17
- . . - _ _ . . . _ . . . ~ . _ _ _ . . _ _ _ . _ _ _ . _ . . - _ . _ _ _ ~ . . . _ _ _ _ . . ~ . _ _ _ _ . . . . _ . . _ . . _ . _ _ _ _ _ _ _ . . _ _ _
l J AIMR mamai l Sinndant l'hu11 %a p V 3G.5 CONTilOL llUILDING ANALYSIS ihe o.15g onE are obtained according to the following steps: 3G.5.1 SSI Analyticalllesults (1) The floor response spectra at the various elevations for the various soil cases are The maximum accelerations for the 3 cases are enveloped. Further more, the two horizontal shown in Table 30.51. The maximum forces in response spectra at any clevation are terms of shear, moment and asial forces for the 3 enveloped to form the bounding horizontal cases are shown in Table 30.5 2, spectra. To further assess the effect of the surrounding (2) A generie scale factor is applied to all floor soils, floor response spectra at the basemat and at terponse spectra envelopes to account for the control room are compared. Figures 30.51 and missing soil profiles for all dampings of 30.5 2 are the basemat response spectra in the X interest. The scaled SASSI envelope spectra dnd vertical dkrectkons rCspectivCly. Figure 30.5 3 are smoothed and peak broadened by + 10% and 3G.5 4 are the control room response spectra in as a minimum , the X and vertical directions respectively. The site envelope OBE floor response spectra of Note that the effect of toil stilfr.ess im the critical damping ratios of 2,3, and 5% are shown in control building follows the same trend as the soil Figure 30.5 5 through Figure 3G.5 22 for the stiffness effect for the reactor building. horizontal and settical directions. 'Ihe hotirontal response spectra are applicable to each c,f the two 3G.5.2 Site Enulope Scismle imads For Control horirontal principal directions. The vertical wall fluilding Comptes spectra are applicable to equipment located within 3m (10 ft) from the face of the exterior walls. For The site.cnvelope seismic loads are established equipment located elsewhere, the appror6.ie slab
- ] from the ernelopes of all SAssi analysis results for a spectra shall be used.
wide range of sites. The site envelope seismic loads obtained are applicable to the control building and other equipment housed inside the control building at any site whose site characteristics meet the conditions defined in Section 3A.1 of Appendix 3A. The site envelope seismic loads in terms of maximum structural responses and floor response spectra due to the 0.15g OBE condition in the control building are presented. The 0.3g SSE loads can be conservatisely taken as two times the 0.15g OBE loads. 3G.5.2.1 Musimum Structural Imads The maximum accelerations, shears, and moments along the control building walls due to the X, Y, and Vertical excitations are shown in Table 3G.5 3. These loads represe t the enveloped responses of the 3 soil cases plus a scale factor to acsount for the other soils cases that were analyzed in the reactor building but not analyzed herein. 3 G.5.2.2 1loor Response Spectra The site-envelope floor response spectra due to Amendment 17 Kall
ABV/R 2muum Standard I'lant FIV H
/7 311.4 References
\b
- 1. c.chtel Topical Repost DC-TOP-4, August 1980, Scisinic Analysis of Structures and Equipment for Nuclear Powcr Plants.
- 2. Tseng, W.S., D.D., Sirnplified Method of Predicting Scistnic Basemat Uplift of Nuclear Powcr Plant Structurcs, Transactions of the 6th International Conference on SMIRT, Paris France, August 1981.
l I V i f l l l O v Amendment 17 31101
' - = * - m - , , , ,. -
ABWR m== Standard Plant unn SECTION 31.3 l CONTENTS J Section Title fligt j 313.1 Plant Normal Oneratinn Conditions 313 1 31 3.1.1 Pressure, Temperature and Relative ilumidity 31,3 1-31 3.1.2 Radiation 31 3-1 313.2 Plant Accident Conditions 313 1 , 313.2.1 ~. Pressure, Temperature and Relative ilumidity 313 1 31 3.2.2 Radiation 313 1 313.23 Water Quality and Submergence 31 3-1 SECTION 31.3 l
'O TABLES Table lith Eage 313 1 Thermodynamic Emironment Conditions inside Primary Containment Vessel, Plant Normal Operating Conditions 313 2 31 3-2 Thermodynamle Emironment Conditions inside Reactor Building (Secondary Containment), Plant Normal Operating Conditions 313 3 31 3-3 Thermodynamic Emironment Conditions inside Reactor Building (Outside Secondary Containment), Plant Normal Operating Conditions 31 3-4 31 3-4 Thermodynamic Emironment Conditions inside Control Building, l Plant Normal Operating Conditions 313 5 l
f 313 5 Thermodynamic Emironment Conditions inside Turbine Building,
- Plant Normal Operating Conditions 313 6 31 3-6 Deleted 31 3-7 313 7 Deleted 31 3-8 313 il !
Amendment 17 ' _,_.,_..a,:,;--_._..,_.---..u._._,_... - . . - _ _ _ _ _ _ _ _ ._ _ _.
..m_. . _ . - - . _ . _ . - . . _ _ _ _ _ . _ - _ _
, ABWR m aman Shuldard Plant mvn SECTION 31.3 TAllLES (Continued)
Iable Htle hire 31 3-8 Delcted 313 9 313 9 Radiation Emironment Conditions inside Primary Containment Vessel, Plant Normal Operating Conditions 313-10 313 10 Radiation Environment Conditions inside Reactor Building (Secondary Containment), Plant Normal Operating Conditions 313 11 313-11 Radiation Emironment Conditions inside Reactor Building (Outside S'condary Containment), Plant Normal Operating Conditions 313 12 313 12 Radiation Emironment Conditions inside Control Building, Plant Normal Operating Conditions 313-13 313-13 Radiation Emironment Conditions inside Turbine Building, Plant Normal Operating Conditions 313 14 h
?!3 14 Thermodynamic Emironment Conditions Inside Primary l Containment Vessel, Plant Accident Conditions 313 15 313-15 Thermodynamic Emironment Conditions inside Reactor Building (Secondary Containment), Plant Accident l Conditions 313 16 313-16 Thermodynamic Emironment Conditions inside Reactor Building (Outside Secondary Containment), Plant Accident l 313 19 Conditions 313-17 Deleted 313 20 313-18 Thermodynamic E.aironn ent Conditions Inside ControlIluilding, Plant Accident Conditions 313-21 l
313-19 Radiation Emironment Conditions inside Primary Containment Vessel, Plant Accident Conditions 313-22 0 313 iii Amendment 17
,y , v ,
~ABWRL mame Standard Plant new n .-
31.3 ENVIRONMENTAL CONDITIONS PARAMETERS - {- e e GE PROPRIETARY . provided under separate cover - t
- bgt Amendmenj hgg AmendmeD1
-31 3-2 13 __ 313 17 313 17 313 15 17
- 31 3-4 17 313 16 17 313 5'- 313-17 17 31 3-6 17 ' 313 14
- 31 3-7 ' 13 313 17 31 3-8 13 313 20 17 -
31 3-9 13 313-21 17
-313 17 313-22 17 313 11 17 313-23 17 313-12 17 313-24 17 L 313 17 313 25 17 1 '.
i O
- Amendment 17 313-1 25 i,---. . _ - . - ~ _ , ,_ _. _ - - . _ .
ABWR men ! Standard Plant Riv c ; () 4.4.2.33 Regions of the Power Flow Map 4.4.2.3.5 Flow Cantrol Region I This region defines the system The normal plant startup procedure requires operational capability with the the startup of all RIPS first and maintain at reactor internal pumps running at their minimum pump speed (30% of rated), at their minimum speed (30%). Power which point reactor heatup and pressurization changes, during normal startup and can commence. When operating pressure has been shutdown, w ll be in this region. The established, reactor power can be increased, normal operrting procedure is to start This power flow increase w;ll follow a line up along curn 1. within Region 1 of the f'ow control map shown in Figure 4.41. The system is then brought to the Region 11 This is the low power area of the desired power-flow level within the normal oper-operating map where the carryover ating area of the map (Region IV) by increasing through steam separators is expected the RIP speeds and by withdrawing control rods. to exceed the acceptable value. Operation within this region is Control rod withdrawal with constant pump precluded by system interlocks, speed will result in power / flow changes along lines of constant pump speed (Curves 1 through Region 111 This is the high power / low flow area 8). Change of pump speeds with constant control of the operating map which the system rod position will result in pawer / flow changes is the least damped. Operation within along, or nearly parallel to, the rated flow this region is precluded by SCRRI control line (curves A through F). (Selected Control Rods Run In).- RegionIV This represents the normal operating p v zone of the map where power changes can be made, by either control rod movement or by core flow changes, through the change of the pump speeds. 4.4.23.4 Design Features for Power Flow Control 4.4.2.4 Thermal and liydraulic Characteristics Tbc following limits and design features are Summary Table employed to maintain power-flow conditions shown in Figure 4.4-1: The thermal hydraulic characteristics are provided in Table 4.41 for the core and tables (1) Minimum Power Limits at Intermediate and of Section 5.4 for other portions of the reactor High Core Flows: To prevent unacceptable coolant system. separator performance, the recirculation system is provided with an interlock to reduce the RIP speed. (2) . Pump Minimum Speed Limit: The Reactor Internal Pumps (RIPS) are equipped with Anti-Rotation Devices (ARD) which prevent a tripped RIP from rotating backwards. The ARD begins operating at 300 rpm decreasing speed In order to prevent mechanical wear in the ARD, minimum speed is specified at 300 rpm. However, to provide a stable (,) operation, the minimum pump speed is set at 450 rpm (30% of required). Amendment 15 4.4-3
. . _ . _ _ _ . . _ _ . . _ . . . . - _ . _ _ _ _ _ m- _ . ..m _ - -
ABWR tw=n - i Standard Plant RFY. C 4.4.3 Imose Parts Monitoring System -g Th'e applicant referencing the ABWR design shall provide a loose parts monitoring system on
- the reactor pressure vessel, and_impicment a-
-loose parts detection program which conforms to the guidelines of the regulatory position .
contained in Regulatory Guide 1.133. The design of the loose parts monitoring system is deferred so that it may be defined utilizing commercially. 1 available components,' at the time of construc-tion, integrated with other lastrumentation systems-and it can reflect-the applicants
. preference and experience.' See Subsection 4.4.4.3 for interface requirements.-
_4.4.4 Interfaces - 4.4.4.1 Power Flow Operating Map The specific power flow operating map to be used at the plant will be provided by the utility to the USNRC for information._
. ) 4.4.4.2 ' thermal Limits? :
g* The thermal limits for the core loading at the plant will be provided by the utility to the' USNRC for information. 4.4.4.3 Loose parts Monitoring System . The applicant referencing the ABWR design will provide _ a loose-parts monitoring system and - implement a loose parts detection progras (See L Subsection d 4.3). 4 O Amendment 17 4.44
ABWR 2 m man RIV C l Standard Plant j 5.23.4.13 Cold Worked Austenttic Stainless high alloy steels or other materials such as
~) Steels static and centrifugal castings and bimetallic joints should comply with fabrication require.
Cold work control; are applied for components ments of Sections til and IX of the ASME Boiler made of sustenitic stainless steel. During and Pressure Vessel Code, it also requires fabrication cold work is controlled by applying additional performance qualifications for limits in hardness, bend radii and surface finis', welding in areas of limited access. on ground surfaces. All ASME Section til welds are fabricated in 5.23.4.2 Control of Welding accordance with the requirements of Sections !!! and 1% of the ASME Boiler and Pressure Vessel 5.23.4.2.1 Avoidance of Ilot Cracking Code. There are few restrictive welds involved in the fabrication of BWR components. Welder Regulatory Guide 131 describes the acceptable qualification for welds with the most restric-method of implementing requirements with regard live access is accomplished by mockup welding. to the control of welding when fabricating and Mock up i examined by sectioning and radiography 5 joining austenitic stainless steel components and (or UT). n systems. The Acceptauce Criterion ll.3.b.(3) of SRP Written welding procedures which are approved Section 5.2. is based on Regulatory Guide by GE are required for all primary pressure boun- 1.71. The ABWR design meets the intent of this dary welds. These procedures comply with the regulatory guide by utilizing the alternate requirements of Sections ill and IX of the ASME approach as follows. Boiler Pressure Vessel Code and applicable NRC Regulatory Guides. When access to a non volumetrically examined ASME Section 111 production weld (1) is less U,, All austenitic stainless steel weld filler than 305 mm in any direction and (2) allows l l materials were required by specification to have welding from one access direction only, such a minimum delta ferrite content of 8 FN (ferrite weld and repairs to welds in wrought and cast number) determined on undiluted weld pads by low alloy steels, austenitic stainless steels magnetic measuring instruments calibrated in and high nickel alloys and in any combination of accordance with AWS specification A4.2 74. these materials shall comply with the fabrica-tion requirements specified in ASME Boiler and Delta ferrite measurements are not made on Pressure Vessel Code Section ill and with the qualification welds. Both the ASME Boiler and requirements of Section IX invoked by Section Pressure Vessel Code and Regulatory Guide 131 111, supplemented by the following requirements: specify that ferrite measurements be performed on N undiluted weld filler material pads when magnetic (1) The welder performance qualification test instrumen:s are used. There are no requirements assembly required by ASME Section IX shall for ferrite measurement on qualification welds. be welded under simulated access condi-tions. An acceptable test assembly will 5.23.4.2.2 Regulatory Guide 134: Electrostag provide both a Section IX weider Welds performance qualification required by this Regulatory guide. l See Subsection 5.233.2.2. If the test assembly weld is to be judged by bend tests, a test specimen shall be 5.23.4.23 Regulatory Guide 1,71: Weldcr removed from the location least favorable Qua'lilcation or Areas of Limited Accessibility for the welder. If this test specimen cannot be removed from a location Regulatory Guide 1.71 requires that weld prescribed by Section IX, an additional ! ~N fabrication and repair for wrought low alloy and bend test specimen will be required. If (V the test assembly weld is to be judged by 5.2 15 Amendment 15 i 1
ABWR 2miman nw c
; Standard Plant radiography or UT, the length of the weld to of the ASME B&PV Code Section XI. '
- be examined shall include the location least ._ _
favorable for the welder. 5.2.4.1 Class 1 System Boundary Records of the results obtained in welder 5.2.4.1.1 Definition ac_cessibility qualification shall be as _ certified by the manufacturer or installer, The class I system boundary for both shall be maintained and shall be made preservice and inservice inspection programs and _i access!ble to authorized personnel, the system pre:,sure test program includes all ] those items within the Class 1 and Quality Group Socket weld with a 50A nominal pipe size and A boundary on the piping and instrumentation under are excluded from the above drawings (P&lDs). That boundary includes the requirements, following: (2) _ (a) For accessibility, when more restricted (1) Reactor pressure vessel N -access conditions than qualified will (2) Portions of the main steam system N obscure the welder's line of sight to (3) Portions of the feedwater system
- the extent that production welding will (4) Portions of the standby liquid control '
require the use of visual aids such as system mirrors. The qualification test as- (5) Portions of reactor water cleanup system sembly shall be welded under the more . (6)_ Ponions of the residualh:at removalsystem ~ restricted _ access conditions using the (7) Portions of the reactor core isolation
- visual aid required for production cooling system welding. (8)' Portions of the high pressure core flooder system (b) GE complies with ASME Section IX.
Those portions of the above systems within 3 (3)L Surveillance of accessibility qualification the Class 1 boundary are those items which are W ' requirements will be performed along with part of the reactor coolant system up to and normal surveillance of ASME Section IX including any and all of the following:
- performance qualification requirements. -3 (1) the outermost containment isolation valve in 5.23A.3 Regulatory Guide 1.66: the system piping which penetrates primary Nondestructive Examination of Tubular Products reactor containment.
For discussion of compliance with Regulatory - (2) the second of two valves normally closed Guide 1.66, see Subsection 5.2.3.3.3. during normal reactor operation in system
- piping which does not penetrate primary
- 5.2.4 Preservice and Inservice reactor containment.
Inspection and Testing of Reactor Coolant Pressure Boundary (3) the reactor coolant system safety and relief valves,
-This subsection describes the preservice and inservice inspection and system pressure test (4) the main steam and feedwater system up to programs for NRC Quality Group A, ASME Boiler and including the outermost conteinment
- Class l',-items.* : It describes those programs isolation valve.
implementing the requirements of Subsectioa IWB Pressure Vessel (B&PV) Code Section 111 and XI 5.2A.t.2 Exclusions Portions of systems within the reactor
- Items as used in this subsection are products coolant pressure boundary,-as defined in constructed under a Certificate of Authorization 5.2.4.1.1, that are excluded from the Class 1 (NCA4120) and material (NCA 1220). See Section boundary are as iollows:
111, NCA 1000, footnote 2. Amendment 17 ' 53 til w . :..
!BWR.
A = = n. nrw. c Standard Plant - Table 5.21
' REACI'OR COOLANT PRESSURE IlOUNDAR) COMPONEN b APPLICAllLE CODE CASES Number 31t!t Applicable Eauloment Remarks N 7115 (1) Component Support Accepted per RG 1.85 N 122 -(2) Piping Accepted per RG 1.84 N 247 (3) Component Support Accepted per RG 1.84 N 249-9 (4) Component Support Conditionally Accepted per RG 1.85 N 309 1 (5) Component Support Accepted per RG 1.84 N 313 (6) Piping Accepted per RG 1.84 N 316- (7) Piping Accepted per RG 1.84 N 318-3 (8) Piping Conditionally Accepted $
per RG 1.84 N-319 (9) Piping Accepted per RG 1.84 N 391 (10) Piping Accepted per RG 1.84 N 392 (11) Piping Accepted per RG 1.84
'N-393 (12) Piping Accepted per RG 1.84
- N-411 1. .(13) Piping Conditionally Accepted per RG 1.84 N 414 (14) Component Support Accepted per RG 1.84 N-430 (15) Component Support Accepted per RG 1.84 N-2.El (16) Containment Conditionally Accepted Per RG 1.147 N-3071 (17) RPV StuJs Accepted per RG 1.147 LO Amendment 17 12 29
. TABWR 2mimn nity, c Standard Plant Table .4.2 1 -g '
REACTOR COOLANT PRESSURE HOUNDARY COMPONENTS APPLICABLE CODE CASES (Continued) Numbet' Iltls Annilentdc EuulpJntal Remarks 1 (Deleted) (18) )
)
-- (Deleted) (19)
N 416 (20) Piping Accepted Per R G 1.147 N 432 .(21) CL . - i Accepted Per Conu anents RG 1.147 N-4351 (22) Class 2 Accepted Per Ve.ssels RG 1.147 N-457 (23) 130its and - Accepted Per Studs RG 1.147 N 463 (24) Piping . Accepted Per RG 1.147 N 460- (25) Class 1 & 2 Accepted Per Components and RG 1.147 Piping N 472 (26) Pumps Accepted Per RG 1.147 1 N-479 (27) Main Steam Not Listed System in RG 1.147 O 5.2 29J
. Amendment 17
ABWR mamu Standard Plant REV.C Table 5.21 REACTOR COOLAN'I' PRESSURE BOUNDARY COMPONENTS APPLICAHLE CODE CASES (Continued) (1) Additional Materials for Subsection NF, (12) Repair Wcill tg Structural Steel Rolled Classes 1, 2, 3 and MC Component Suppons Shapes and Plates for Component Supports, Fabricated by Welding, Section 111, Division Section HI, Division 1. L (13) Alternative Damping Values for Seismic (2) Stress indices for Structure Attachments, Analysis of Classes 1, 2, 3 Piping Class 1, Section 111, Division L Sections, Section 111, Division 1. (3) Certified Design Report Summary for Com- (14) Tack Welds for Class 1, 2, 3 and MC ponent Standard Supports, Section 111, Components and Piping Supports. Division 1, Class 1, 2, 3 and MC. (15) Requirements for Welding Workmanship and (4) Additional Material for Subsection NF, Visual Acceptance Criteria for Class 1, 2, ~ Classes 1, 2, 3 and MC Component Supports 3 and MC Linear-7)pe and Standard Supports. Fabricated Without Welding, Section 111,
. Division 1. (16) Repair and Replacement of Class MC Vessels (3) ^ Identification of Materials for Component (17) Revised Examination Volume for Class 1 Supports, Section IH, Division L Bolting, Table IWB 2SOO 1, Examination Category B-G-1, When the Examinations Are
.(6) Alternate Rules for Half Coupling Branch Conductedfrom the Drilled Hole Connections, Section 111, Division 1. ,
(18) (Deleted) (7)-' Alternate Rules for Fillet Weld Dimensions for Socket Welded Fittings, Section JH, Division 1, Class. ), 2, 3. (19) (Deleted) (8) Procedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping, Section Hi, Division L (20) Alternative Rules for Hydrostatic Testing of Repair or Replacement of Class 2 Piping (9) Alternate Procedure for Evaluation of Stress in Butt Weld Elbows in Class 1 Piping, (21) Repair Welding Using Automatic Or Machine Sectson I(1, Division 1. Gas Tungsten-Arc Welding (GTA W) Temperbead Technique (10) Procedure for Evaluation of the Design of Hollow Circular Cross Section Welded Attach- (22) Alternative Examination Requirements for ments on Class 1 Piping. Section Hi, Vessels Hith Wall Thick sesses 2 in. or Less Division 1. (23) Qualification Specimen Notch Location for (11) Procedure for Evaluation of the Design of Ultrasonic Examination of Bolts and Studs Hollow Circular Cross Section Welded Attachments on Classes 2 and 3 Piping, (24) Evaluation Procedures and Atceptance Section 1H, Division 1. Criteria for Flaws in Class 1 Ferritic Piping That Exceed the Acceptance Standards ofIWB-3514-2 Amendment 17 1 2-29.2
, ~ . . _ . . . _ _ __ -. - _ ~ _ _.
m6mn ABWR nrw c Shmdard Plant Table 5.2 1 REACTOR COOLANT PRESSURE IlOUNDARY COh1PONENTS APPLICAllLE CODE CASES (Continued) (23) Alternative Examination Coverage for Class I and 2 If *cids (26) Use of Digital Readout and Digital bicasurement Devices for Performing Pump Vibration Testing (27) Boiling il'ater Reactor (BlVR) blain Steam Hydrostatic Test O O 5.2-29 3 Amendmcat 13
4 ABWR uumn Sinndard Plant RIV C O SECTION 5.4 v CONTENTS (Continued) Section Illic Eagt 5.4.6.2.5.1 Standby Mode 5.4 15 5.4.6.2.5.2 Emergeney Mode 5.4-15 5.4.6.2.53 Test Mode 5.4 16 5.4.6.2.5.4 Limiting Single Failure 5.4-16 5.4.63 Performance Evaluation 5.4-16 5.4.6.4 Preoperational Testing 5.4 16 5.4.7 Residual IIcat Removal Ssstem 5.4-16 5.4.7.1 Design Bases 5.4-16 5.4.7.1.1~ Functional Design Basis 5.4-16a 5,4.7.L1.1 Low Pressure Flooder (LPFL) Mode 5.4-17 ry O' 5.4.7.1.1.2 Test Mode 5.4 17 5.4,7.1.13 Minimum Flow Mode 5.4 17 5.4.7.1.1.4 Standby Mode '5.4-17 5.4.7.1.1.5 Suppression Pool Cooling 5.4-17 5.4.7.1.1.6 Wetwell and Dr>well Spray Cooling 5.4-18 5.4.7.1.1.7 Shutdown Cooling 5.4-18 5.4.7.1.1.8 Fuel Pool Cooling 5.4 18 5.4.7.1.1.0 Reactor Well and Equipment Pool Drain 5.4-18 5.4.7.1.1.10 AC Independent Water Addition 5.4-18 5.4.7.1.2 Design Basis for Isolation of RHR System from Reactor Coolant Systerr 5.4 18a - 5.4.7.13 Design Basis for Pressure Relief Capacity 5.4-18a 5.4.7.1.4 Design Basis with Respect to General Design Criterion 5 5.4 19 \",1 5.4iv Amendment 10
,, . . ~ _ - - ~.- _- -. ~ _ - . - . . . - - . _ . - .
TABWR maman - Standard Plant nrrv. c SECTION 5.4 CONTENTS (Continued)- O Section Title Eagt ~
-5.4.7.1.5~ Design Basis for Reliability and Operability _ 5.4 19 5.4.7.1.6 Design Basis for Protection from Physical Damage 5.4-19
'.5.4.7.2 Systems Design 5.4-19 5.4.7.2.1 System Diagrams 5.4-19 5.4.7.2.2 ' Equipment and Component Description 5.4-19.1
]_
5.4.7.23 Controls and Instrumentation 5.4-20
'5.4.7.23.1 Interlocks 5,4-21
-5.4.7.2.4 Applicable Codes and Classification 5.4-21
.5.4.7.2.5- Reliability Considerations 5.4 21 5.4.7.2.6 Manual Action 5.4-21.1 l _
5.4.73 Performance Evaluation ' 5.4-23'
- 5.4.73.1 Shutdown with All Components Available 5.4-23 5.4.73.2 Worst Case Transient 5.4-23 5.4.733 Emergency Shutdown Cooling : 5.4-24 . .
5.4.73.4' -- Normal Shutdown Coolini,- 5.4-24= 5.4.7.4 Pre-operational Testing 5.4-24 14.8 5 Ecactor water cleanun system 5.4-24
.5.4.8.1' Design Bases - 5.4-25 5.4.8.2 . System Description 5.4 25 - 5.4.83- System Evaluation 5.4-26 5.4.9 Main Steamilnes and Fredwater Pining 5.4-27 '
5.4.9.1 Safety Design Bases 5.4-27 5.4-v g Amendment 17
ABWR noin Standard Plant uvc (o) directly into the reactor pressure vessel to signal closes the RHR containment isolation the drywell spray header degraded plant valves that are prosided for the shutdown cooling conditions when AC power is not available from suction. Subsection 5.2.5 provides an explanation either onsite or offsite sources. The RHR of the leak detection system and the isolation provides the piping and valves which connect the sigt ilst see Subsection 5.2.5.2.1 (12) and Table FPS piping with the RHR loop C pump discharge 5.2-t. piping. The manual valves in this line permit adding water from the FPS to the RHR system if Th : RHR pumps are protected against damage the RHR is not operable. The primary means for from a closed discharge valve by means of supplying water through this connection is by use autonsatic minimum flow valves which open on low of the diesel-driven pump in the FPS. A backup mainlige flow and close on high mainline flow, to this pump is provided by a connection on the outside of the reactor building which allows 5.4.7.1.3 Design Ilasts for Pressure Relief hookup of the FPS to a fire truck pump. Capacity The vessel injection mode is intended to The relief valves in the RHR system are sired prevent core damage during station blackout after on the basis of thermal relief and valve bypass RCIC has stopped operating, and to provide an leakage only, in. vessel core melt prevention mechanism during a severe accident con ('ition. If the AC-independent water addition mode is not actuated in time to prevent core damage, core melting and vessel failure, then it covers the corium in the lower drywell when initiated and adds water to containment, thereby slowing the pressure rise. (A) The drywell spray mode prevents high gas temperatures in the upper drywell and adds additional water to the containment, which increases the containment thermal mass and slows the pressurization rate. Additionally, the drywell spray provides fission product scrubbing to reduce fission product release in the event of failure of the drywell head. Operation of the AC-independent water addition mode is entirely manual. All of the valves which must be opened or closed during fire water addition are located within the same ECCS valve room. The connection to add water using a fire truck pump is located outside the reactor building at grade level, 5.4.7.1.2 Design Basis for Isolation of RilR System from Reactor Coolant System The low pressure portions of the RHR system are isolated from full reactor pressure whenever the primary system pressure is above the RHR system design pressure. (See Subsection 5.4.7.1.3 for further details.) In addition, (V3 automatic Isolation occurs for reasons of maintaining water inventory which are unrelated to line pressure rating. A low water level Amendment 17 5.4 18.t
ABWR mmo S11Lrldard Plant suv c Redundant interlocks prevent opening v:lves controlled by the operator from the control & to the low pressure suction piping when the rocm. The only operations performed outside of W reactor pressure is above the shutdown range. the contrcl room for a normal shutdown are These sarne interlocks initiate valve closure on manual operation of local flushing water increasing reactor pressure. admission valves, which are the means of providing clean water to the shutdown portions Overpressure protection is achieved during of the RilR system. systers operation when the system is not isolated from the reactor coolant pressure. The RilR Three separate shutdown cooling loops are system is operational and not isolated from the provided; and although the three loops are reactor coolant system only when the reactor is required for shutdown under normal depressurized. Two modes of operation are circumstances, the reactor coolant can be applicable; the flooder mode and the shutdown brought to 1000C in less than 36 hours with cooling mode. For the flooder mode, the only two loops in operation. The Rif R system is injection valve opens through interlocks only for part of the ECCS and therefore is required to be reactor pressure less than approximately 500 designed with redundancy, piping protection, psig. For the shutdown cooling mode, the suction power separation, etc., as rdquired of such valves can be opened through interlocks only for systems. (See Section 6.3 for an explanation of reactor pressures less than approximately 135 the design bases for ECCS Systems.) psig. Once the system is cperating in these lower pressure modes, events are not expected Shutdown suction and discharge valves are that would cause the pressure to increase. If required to be powered from both offsite and for some unlikely event the pressare would standby emergency power for purposes of increase, the pressure interlocks that allowed isolation and shutdown following a loss of the valves to initially open would cause the offsite power. valves to close on increasing pressure. Tbc RHR system piping would then be protected from overpressure, The valycs close at low pressure, 5.4.7.1.6 Design liasis for Protection from Physical Damage g and the rate of pressure increase would be low. During the time period while the valves are The design basis for protection from closing at these low pressure conditions, the RHR physical damage, such as internally generated system design and margins that satisfy the missiles, pipe break, seismic effects, and interf acing system LOCA provide ample fires, are discussed in Sectior- 3.5, 3.6, 3.7, overpressure protection. and Subsection 9.5.1 In addition, a high pressure check valve will 5.4.7.2 Systems Design close to prevent reverse flow if the pressure should increase. Relief valves in the discharge 5.4.7.2.1 System Diagrams piping are sized to account for leakage past the check valve. All of the components of the RilR system are shown in the P&ID (Figure 5.4-10). A 5.4.7.1.4 Design Basis With Respect to General description of the controls and instrumentation Design Criterion 5 is presented in Subsection 7.3.1.1.1 emergency core cooling systems control and instrumen-The RHR system for this unit does not share tation. equipment or structures with any other nuclear l unit. Figure 5.4-11 is the RHR process diagram and I data. All of the sizing modes of the system are 5.4.7.1.5 Design Basis for Reliability and shown in the process data. The interlock block Operability diagram (IBD) for the RHR system is provided in Section 7.3. The design basis for the shutdown cooling mode of the RHR System is that this mode is h Amendment 17 5419
ABWR = =^n Sinndard Plant Ril,_C /^') Interlocks are provided to prevent: (1) v drawing vessel water to the suppression pool, (2) opening vessel suction valves above the suction lines or the discharge line design pressure, (3) inadvertent opening of drywell spray valves during RilR ope:ation where the injection valve to the reactor is open and when drywell pressure is not high enough to require the drywell spray for pressure reduction, and (4) pump start when suction vslve(s) are not open. A description of the RHR system logic (i.e., interlocks, permissives) is presented in Table 5.4 3.
.%7.2 , Qulpment and Component Description (1) Sptem Main pumps The following are system performance require-ments the main pumps must satisfy. The pump equipment performance requirements include additional margins so that the system perfor-mance requirements can be achieved. These margins are standard GE equipment specifica-tion practice and are included in procure-ment specifications for flow and pressure measuring accuracy and for power source fre-(, quency variation.
x Number of Pumps 3 Pump type Centrifugal Drive unit t)pe Motor
,~
t_) Amendment 17 5+191
ABWR nu m Standard Plant ntw_c Design flow rate 3 954 m /h into the shutdown cooling mode Ic,s a nominal vessel pressure ( . U g w Total discharge head 125 m kg/cm 2g, l at design flow rate The RilR heat exchanger capacity i, o hiaximum bypass flow 3 147.6 m /h quired to be sufficient to meet e ch of these functional requirements. The hiinimum total 220 m max limiting function for the RiiR heat ex. discharge bead at 195 m min changer capacity is st 1 OCA contain-maximum bypass flow rate ment cooling. Thi uhanger espac-ity, K, is 88.4: C-sec per hiaximum runout flow 3 1130 m /h heat exchangcr hiaximum pump brake 550 kw The performance characteristics of the horsepower heat exchangers are shown in Table 5.4-4. I Net positive suction 2.4 m head (NPSii) (3) Valves ! at 3.28 ft above the pump floor setting All of the directional valves in the system j e conventional gate, globe, and check Process fluid 10 to 182oC valves designed for nuclear service. The in-temperature range jection valves are high speed valves, as op-eration for RiiR injection requires. Valve (2) licat Exchangers pressure ratings are to provide the control or isolation function as necessary; i.e, The RilR heat exchangers have three major all vessel isolation valves are treated as functional requirements imposed upon them. They are as follows: Class 1 nuclear valves at the same pressure as the primary system. g (a) Post-LOCA Containment Cooline. The RilR (4) ECCS Portions of the RilR System limits the peak bulk suppression pool l temperature to less th n 970C by The ECCS portions of the RIIR system include direct pool cooling with two out of the those sections that inject water into the three divisions. reactor vessel. l (b) Reactor Shutdown. The RiiR removes The route includes suppression pool suctiot: enough residual heat (decay and sen- strainers, suction piping, RIIR pumps, dis-sible) from the reactor vessel water to charge piping, RIIR heat exchangers, injec-l cool it to 600C within 24 hours after tion valves, and drywell piping into the the control rods are inserted. This vessel nonles and core region of the mode shall be manually activated after a reactor vessel, blowdown to the main condenser reduces the reactor pressure to below 9.5 Pool-cooling components include pool suction kg/cm2 g w th all three divisions in strainers, piping, pumps, heat exchangers, operation. and pool return lines. l (c) Safe Shutdown. The RHR brings the Containment spray components are the same as reactor to a cold shutdown condition of pool cooling components except that the l less than 1000C within 36 hours of spray headers replace the pool return lines. control rod insertion with two out of the three divisions in operation. The 5.4.7.2.3 Controls and instrumentation RHR is m a n u ally activated Controls and instrumentation for the RilR system are described in Section 7.3. h l Amendment 15 5 4-20 t
l ABWR :waan l tu v n Standard Elant SECTION 6.2 O) L CONTENTS (Continuet!) Section Iille Page 6.2.1.63 Design Provisions for Periodic Pressuization 6.2-14 6.2.1.7 Instrumentation Requirements 6.2 15 6.2.2 Containment llent Remmal Sistem 6.2 16 6.2.2.1 Design Bases 6.2 16 6.2.2.2 Containment Cooling System Design 6.2 16 6.2.23 Design Evaluation of the Containment Cooling System 6.2 17 6.2.23.1 System Operation and Sequence of Events 6.2-17 6.2.23.2 Summary of Containment Cooling Analysis 6.2-17 6.2.2.4 Test and Inspections 6.2-17 10 ( ) 6.2.2.5 Instrumentation Requirements 6.2 18 6.23 Secondary Con.tninnlent l'unctional llnign 6.2 18 6.23.1 Design Bases 6.2 18 6.23.2 System Design 6.2 19 6.233 Design Evaluation 6.2-22 6.233.1 Compartment Pressurization 6.2-22 6.233.1.1 Design Bases 6.2-22 6.23 3.1.2 Design Features 6.2 22 6.233.1.2.1 Reactor Core isolation Cooling (RCIC) Compartment 6.2-22 6.233.1.2.2 Reactor Water Cleanup (RWCU) Equipment and Valve Rooms 6.2 22 6.233.1.23 Main Steam Tunnel 6.2 22.1 l (j 6.2.IV Amendment 17
ABWR maun Standatd Plant uvn SECTION 6.2 g CONTENTS (Continued) Section Title Eage 6.233.13 Design Evaluation 6.2 22.1 6.23.4 Tests and Inspections 6.2 21 6.23.5 Instrumentation Requirements 6.2 23 6.2.4 Containment isolation Sntem 6.2 23 6.2.4.1 Design Bases 6.2 23 6.2.4.1.1 Safety Design Bases 6.2-23 6.2.4.1.2 Design Requirements 6.2 24 6.2.4.2 System Design 6.2 24 6.2.4.2.1 Containment Isolation Valve Closure Times 6.2-25 6.2.4.2.2 Instrument Lines Penetrating Containment 6.2 25 6.2.4.23 Compliance with General Design Criteria and Regulatory Guides 6.2-25 6.2.4.2.4 Operability Assurance, Codes and Standards, and Valve Qualification and Testing 6.2 25 6.2.4.2.5 Valve Operability and Leakage Control 6.2 26 6.2A.2.6 Redundancy and Modes of Wlve Actuations 6.2-26 6.2.43 Design Evaluation 6.2 26 6.2.43.1 Introduction 6.2-26 6.2.43.2 Evaluation Against General Design Criteria 6.2 27 6.2.43.2.1 Evaluation Against Criterion 55 6,2-27 6.2.43.2.1.1 Influent Lines 6.2-27 6.2.43.2.1.1.1 Feedwater Line 6.2 27 6.2.43.2.1.1.2 RHR Injection Line 6.2 27 6.2-v h Amendment 17
ABWR m aman Standard Plant ruv c (~') SECTION 6.2 v CONTENTS (Continued) section lille Ihige 6.2.43.2.1.13 11PCF Line 6.2-2R 6.2.43.2.1.1.4 Standby Liquid Control System Line 6.2-2R 6.2.43.2.1.1.5 Reactor Water Cleanup System Line 6.2-28 (Reactor Vesseillead Spray) 6.2.43.2.1.1.6 Recirculation Pump Seal Purge Water Supply Line 6.2 28 6.2.43.2.1.2 Efiluent Lines 6.2-28 6.2.43.2.1.2.1 Main Steam and Drain Lines and RCIC Steam Line 6.2 23 6.2.43.2.1.2.2 RilR Shutdown Cooling Line 6.2 29 6.2.43.2.1.23 Reactor Water Cleanup System Suetion Line 6.2 29 6.2A3.2.13 Conclusion on Criterion 55 6.2 29
, 6.2.43.2.2 Evaluation Against Criterion 56 6.2 29 l ,
6 i l 6.2.43.2.2.1 Influent Lines to Suppression Pool t.2-3 6.2.43.2.2.1.1 IIPCF and R11R 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 Lines 6.2-30 6.2.43.2.2.13 SPCU Discharge Line 6.2 30 6.2.43.2.2.2 Effluent Lines from Suppression Pool 6.2-30 6.2.43.2.2.2.1 RilR, RCIC and l{PCF Lines 6.2-30 6.2.43.2.2.2.2 SPCU Suction Lines 6.2-30 6.2.43.2.23 ACS Lines to Containment 6.2-30 6.2.43.2.2.4 Conclusion on Criterion 56 6.2 30 6.2.43.23 Evaluation Against Criterion 57 6.2-30.1 6.2.43.2.4 Evaluation Against Regulatory Guide 1.11 6.2 31 6.2.433 Evaluation of Single Fallare 6.2-31 (~ 3 V 6.2-vi Amendment 17
ABWR uwwan Standard Plant niv c SECTION 6.2 CONTENTS (Continued) O section Etic Eage 6.2.4.4 Tests and Inspections 6.2-31 6.2.5 C_embustible Gas Controlin Conta:nment 6.2-31 6.2.5.1 Design Bases 6.2 31 6.2.5.2 System Design 6.2-33 6.2.5.2.1 General 6.2-33 6.2.5.2.2 Inerting Equipment 6.2 34 6.2.5.23 Nitrogen Make-Up 6.2 35 6.2.5.2.4 Drywell Bleed 6.2 35 6.2.5.2.5 Pressure Control 6.2-36 6.2.5.2.6 Overpressure Protcetion 6.2-36 6.2.5.2.7 Recombiner 6.2-36 h 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 Personnel 5afety 6.2-38 6.2.6 fontpi, ment 14'akace Testine 6.2-39 l 6.2.6.1 Containment integrated Leakage Rate Test 6.2-39 6.2.6.1.1 InitialIntegrated Leak Rate Test 6.2-39 6.2.6.1.1.1 Objectives 6.2-39 6.2.6.1.1.2 Preoperation Test Trocedure 6 2-39 6.2.6.1.1.3 Supplement Verification Test 6.2-40 6.2.6.1.1.4 1nstrumentation Requirements 6.2 40 6.2.6.1.1.5 Acceptance Criteria 6.2-40 6.2-vii Amendment 11
ABWR mamn Standard Plant nw c s
'~', SECTION 6.2 '.J CONTENTS (Continued)
Sectlon Title Eage 6.2.6.1.2 Periodic If akage Rate Tests 6.2 40 6.2.6.1.2.1 Integrated Leakage Rate Tests (ILRT, Type A) 6.2-40 6.2.6.1.2.2 Acceptance Criteria 6.2-41 6.2.6.1.2.3 T'st Frequency 6.2 41 6.2.6.1.3 Additional Criteria for Integrated Leakage Rate Test 6.2-41 6.2.6.2 Containment Penetration Leakage Rate Test (Type 13) 6.2 42 6.2.6.2.1 Geoeral 6.2-42 6.2.6.2.2 Acceptance Criteria 6.2 42 6.2.6.2.3 Retest Frequeney 6.2-42 m I i 6.2.6.2.4 Lesign Provisions for Periodic Pressurization 6.2-42 6.2.6.3 Containment isolation Valve Leakage Rate Test (Type C) 6.2-43 6.2.6.3.1 General 6.2 43 6.2.6.3.2 Acceptance Criteria 6.2-43 6.2.6.4 Scheduling and Reporting of Periodic Tests 6.2 43 6.2.6.5 Special Testing Requirements 6.2-43.1 6.2.7 Refereign 0.2-44 r~g (,) 6.2-viii Amendment 17
. . .-- . _- . . . _ ~ - _ . . . -- - . . - . - . - .
LABWR m6 man nrv c Standard Plant SECTION 6.2 _
-TABLES -l Table Title Page 6.2 1 ~ Containment Parameters 6.2-45 6.22 Containment Design Parameters 6.2 46 [
- p
.g 6.2 2a Engineered Safety Systems information
' 6.2 46a for Contal:. ment Response Analyses h
6.2 2b . Net Positive Suction l{ead (NPSH) Available to RHR Pumps 6.2 46c 6'.2 2c -: Net Positive Suction Head (NPSH) Available to HPCF Pumps 6.2 46d 6.23. Compartment Nodal Description 6.2 47-6.2-4 . Compartment Vent Path Description 6.2 48-6.25 Reactor Coolant Pressure Boundary (RCPB) Influent Lines Penetrating Drywell 6.2-49 <
' 6.2 6 - Reactor Coolant Pressure Boundary (RCPB) .
Effluent Lines Penetrating Drywell 6.2 50 6.2-7 Containment isolation Valve Information 6.2 50.1-
- 6.2 8 . Primary Containment Penetration List 6.2 50.50
"~
6.2 9 Secondary Containment Penetration List 6.2 50.55 6.2 10- Potential Bypass Leakage Paths 6.2 50.57 ILLUSTRATIONS Figure II.tle Page 6.2-1 ~ A Break in a Feedwater Line - 6.2 51
- 6.22i Feedwater Line Break- RPV Side Break' Area 6.2-52
~
6.2-3 Feedwater Line Break Flow--Feedwater System Side of Break '6.2-53 6.24 Feedwater Line Break Flow Enthalpy-- Feedwater System Side of Break 6.2-54 6.2ix '- Amendment 16
- . ~ _ . . - ~ . - . . - . - - , . . . . _ . _ __ -
ABWR m6ima Standard Plant RIN. C 6.2 CONTAINMENTSYSTEMS
]x-- (4) The containment structure can withstand 6.2.1 Containment Functional Design coincident fluid jet forces associated with the flow from the postulated rupture of any 6.2.1.1 ' Pressurt Suppression Containment pipe within the containment.
6.2.1.1.1 Design Bases (5) The containment structure is desi;,ned to accommodate flooding to a suffic'.ent depth The ABWR pressure suppression primary above the active fuel to permit safe removal containment system, which comprises of drywell of the fuel assemblies from the reactor core and wetwell and supporting systems, is designed after the postulated DBA. to have the following functional capabilities: (6) The containment structure is protected from (1) The containment structure has the capability or designed to withstand hypothetical to maintain its functional integrity during missiles from internal sources and and following the peak transient pressures uncontrolled motion of broken pipes which and temperatures which would occur following could endanger the integrity of the
. any postulated loss of coolant accident containment, (LOCA). A design basis accident (DBA) is defined as the worst LOCA pipe break (which (7) The containment structure provides means to leads to maximum containment and drywell channel the flow from postulated pipe pressure and/or temperature) and is further ruptures in the drywell to the pressure postulated to occur simultaneously with loss suppression pool, of offsite power and a safe shutdown
- carthquake (SSE). (8) The containment system is de_ signed to allow for periodic tests at the calculated peak or O Th= e t i - i ir
- the full range-of loading conditions
' is 8 >is a rer >=a a i >i rre r ie - <= th i
- 8-from individual penetrations, isolation consistent with normal plant operation and salves and the integrated leakage rate from accident conditions including the LOCA the structure to confirm the leaktight ;
related design loads in and above the . integrity of the containment, suppressi- pool; (9) The atmospheric control system (ACS) The containment structure is designed to establishes and maintains the containment accommodate the negative pressure difference atmosphere to less than 3.5 volume To oxygen between drywell and wetwell and relative to during normal operating conditions to assure the reactor building surrounding. inert atmosphere operation of two permenately installed recombiners can be (2) . The _ containment structure and isolation initiated on high levels as determined by
- system, with concurrent operation of other CAMS.
accident mitigation systems, is designed to limit fission product' leakage during and following the. postulated DBA to values less than leakage rates which would result in 6.2.1.1.2 Design Features offsite doses greater than those set forth in 10CFR100. The containment structure consists of the following major components (Figures 1,2-2 (3) Capability for rapid closure or isolation of through 1.212); all pipes or ducts which penetrate the containment boundary is provided to maintait (1) A drywell (DW) comprised of two volumes;(a) leakage within acceptable limits, an upper drywell (UD) volume surrounding the O Amendment 17 6.2-1 l l:
, - - - - . + . - .,,,,y, ---c, ,wy- ,,
ABWR mun Slandard Plant umA reactor pressure vessel (RPV) ar o housing The design parameters of the major components the steam and feedwater lines and other of the containment system are given in Table g connections of the reactor primary coolant 6.2 1. A detailed discussion of their W system, safety / relief valves (SRVs) and the structural design bases is given in Section 3.8. drywell llVAC coolers, and (b) a lower drywell (LD) volume housing the reactor 6.2,1.1.2.1 Drpell internal pumps, fine motion control rod drives and under vessel components and The drywell is designed to withstand the senicing equipment. The upper drywell is a pressure and temperature transients associated cylindrical, reinforced concrete structure with the rupture of any primary system pipe with a removr.ble steel head and a .einforced inside the drywell and also the rapid reversal concrete diaphragm floor. Th: cylindrical in pressure when the steam in the drywell is RPV pedestal which is connected rigidly to cor 3cnsed by the emergency core cooling system the diaphragm floor sep trates the lower (ECCS) flow following past LOCA flooding of the drywell from the wetwell. It is a RPV. pref abricated steel structure filled v.ith concrete after erection. Ten drywell A vacuum breaker system has been provided connecting vents (DCVs), approximately IMx2M between the drywell and wetwell. The purpose of in cross section, are built into the RPV tne drywell-to-wetwell vacuum relief system is pedestal and connect the UD and LD. The to prevent backflooding of the suppression pool DCVs are extended downward via 1.2M inside water into the lower drywell and to protect the diameter steel pipes, each of which has integrity of the diaphragm floor (DIF) slab three horizontal 0.7M diameter vent outlets between the drywell and the wetwell, and the into the suppression pool, drywell structure and liner. Redundant vacuum relief systems are provided to protect against (2) A wetwell comprised of an air volume and failure of a single system. The design suppression pool filled with water to drywell to-wetwell pressure difference is ( + )25 rapidly condense stea'n from a reactor vessel psid and (-) 2 psid. The vacuum breaker system g blowdown via the SRVs or from a break in a is also designed to withstand the high W major pipe inside the drywell through the temperature associated with the break of a small vent system. The wetwell boundary is a line in the drywell which does not result in cylindrical reinforced concrete wall which rapid depressurization of the RPV. is continuous with the upper drywell boundary. A reinforced concrete mat The maximum drywell temperature occurs in foundation supports the entire containment case of a steam line break j337.5 F) and is system, and enclosed structures, systems and below the design value (340 F). components, and extends to support the reactor building surrounding the The maximum drywell pressure occurs in case containment. of a feedwater line break (39 psig). The design pressure for the drywell (45 psig) includes 16% (3) The containment structure includes a steel margin. line.r to reduce fission product leakage to allowable levels. All normally wetted No vacuum breaker system is required for the surfaces of the liner in the suppression primary containment-to-reactor building negative pool are stainless steel, Penetrations pressure which is predicted to be maximum 1.8 I through the liner for the drywell head, psi, between the wetwell and the reactor equipment hatches, personnel locks, piping, building, compared to the design negative electrical and instrumentation lines are pressure of 2 psi. provided with seals and leaktight connections. The allowable leakage is 0.5% A heating and cooling system is provided to per day from all sources. maintain drywell temperatures during normal O 6.22
ABWR mswm Sigtidard Phmt RIV C O (3) HPCF System Rooms secondary containment integrity (i.e., do not open the blowout panels to the steam tunnel), (4) Fuel Pool Cooling and Cleanup System Rooms the normal room ventilation subsystems are isolated and the SGTS begins to exhaust the air (5) RHR System Rooms (through its filter) from the rooms, maintaining the pressure at -1/4-in, water gage or less with (6) Suppression Pool Cleanup System Rooms respect to the environs. No mixing of fission products with the room volumes is assumed. (7) SGTS Filter Compartments liigh energy pipe breaks in the secondary (8) Spent Fuel Storagc Pool containment cor etments do not require secondary conta%nint integrity. Following The liquid leakage from secondary containment breaks of this type, pressure will be relieved to the clean zone or the environment is by blowout vent openings and panels within the controlled, as required, by means of water loop secondary containment or to the steam tunnel and seals, automatic shutoff valves in series, or turbine building. piping upgrade to safety class. All system operations that transport liquid from the The fuel storage and handling areas are part secondary containment to the clean rone or to the of secondary containment, where throughline environment will be automatically shut off during leakage of fission products are collected. an accident and not be automatically initiated These areas are constructed of reinforced fcilowing an ace;.ici.t. concrete. (See Subsection 3.8.4 for design details.) The secondary containment boundaries A postulated high energy pipe break in the are the concrete walls and ceiling of the secondary containment is accommodated so as not refueling floor and the stainless steel lined to exceed the environmental qualification limits upper pool. (g
~) of the equipment required for plant shutdown.
Blowout pa.vls are installed as required in rooms Following accidents requiring secondary where high energy pipe breaks are postulated and containment integrity, the normal reactor the panels reti:ve the thermal temperature and building ventilation system is isolated, and the pressure build ip in the room. SGTS begin s txhausting the secc.sdary containment air. The SGTS thus maint .ns the pressure at High energy pipe breaks in the secondary 1/4 ir.., ;; less with respect to the environs containment will cause a failure to maintain when the exterior wind speed is 14 se than or negative pressure in the secondary contaiament, equal to 10 mph. (Above that wind speco, M en This is acceptable since there is no significant exfiltration does occur,10CFR100 guidelina release of radioactivity from this accident event will not be exceeded because of the increasco because fuel is not damaged and the plant is shut atmospheric dispersion which may be assumed.) down promptly. No mixing is assumed for fission products within the secondary containment volume. There are no All effluents processed by the STGS from the high energy lines in the fuel handling and secondary containment areas are monitored for storage areas, whose failure would result in gamma radiation level prior to their release to pressurization or loss of secondary containment the environment. integrity. The ECCS, RCIC, RWCU, FPCCU and SPCU system Penetrations between secondary containment l rooms of the secondary containment collect and the environs are of four different types: throughline leakage of fission products. The (1) piping penetrations; (2) architectural l pump rooms are of reinforced concrete openings (doors, hatches, and blowout pancis); l construction. (See Subsection 3.8.4 for design (3) HVAC duct penetrations; and (4) clectrical d e t ails.) Following accidents which require penetrations. Each of these categories is p) L discussed below separately. Most piping which Amendment 7 6.2-21
ABWR u-n nrv c Standard Plant forms a part of the secondary containment conjunction with a worst case single active boundary is designed to at least Seismic Category component failure is selected for the analysis I and ASME Section III, Class 3 requirements. of each compartment. For this analysis, a worst Some lines have no special isolation provisions case single active component failure is defined and are not ASME Section til or Seismic Category as the failure to close of an isolation valve I if an analysis shows that exfiltration would which separates the reactor pressure vessel from not occur in the event of failure of that pipe the high or moderate energy pipe break in the [i.e., the -1/4 in, water gage pressure secondary containment. For the room design, the differential would be maintained). peak differential pressures are not to exceed the design differential pressure. For architectural open:ngs the inleakage is based on -1/4 n. water gage pressure 6.23.3.1.2 Design Features differential, All doors have a vestibule with a second (outer) door. HVAC and electrical The following paragraphs are brief penetrations are designed to minimize leaks, and descriptions of the compartments analyzed for HVAC system is designed and tested for isolation pressurization, A more detailed description under accident conditions will be found in Subsection 3.S.1. Figures 6.2 28 through 6.2-36 are the plan and elevation Table 6.2-9 provides a listing of secondary drawings showing component and equipment containment openings. All piping and cabletray locations and configurations. Figurc 6.2 37 penetrations will be scaled with a scaling shows the schematic layout of the secondary l compound for leakage and fire protection. All containment compartments with the interconnected doors are vestibule type with card reader access vent paths and blowout panels. Table 6.2-3 g security systems that are monitored (See tabulates the free volumes, initial environment
- Subsection 13.63.4). The HVAC penetrations are conditions and DBA break characteristics for the designed to close on a design basis accident (See compartments which are analyzed. Table 6.2-4 Subtection 9.4.3 on reactor building HVAC). enumerates the flow path and blowout panel g Testing procedure and frequency can be found in characteristics. I T tha plant tednical specifications.
6.2.33.1.2.1 Reactor Core Isolation Cooling 6.233 Design Evaluation (RCIC) Compartment l The design of the secondary containment The RCIC compartment is located in the bounda;ies is described in the preceding secondary containment at El(-)13200 mm. The subsection. Evaluation of this design, such that design basis break for the RCIC compartment is all regulatory requirements are met, are given in the double ended break of the 6-in. RCIC steam the following subsections: supply line. This line is a high energy line out to the normally closed isolation valve (1) 6.5.1 Standby Gas Treatment System inside the RCIC compartment, it supplies high I energy steam to the RCIC turbine in the event of I (2) 9.4.5 Reactor Building HVAC System reactor vessel isolation. In the event of a postulated design basis high energy line break, l 6.2.33.1 Compartment Pressurization the steam / air mixture is directed into adjoining l compartments and is eventually purged into the 6.233.1.1 Design Bases steam tunnel. The design of the secondary containment 6.2.33.1.2.2 Reactor Water Cleanup (RWCU) compartments with respect to pressurization is Equipment and Valses Rooms l based upon the worst-case DBA rupture of a high or moderate energy line postulated to occur in The RWCU equipment (pump, heat exchanger, each compartment (see Subsection 3.6.2 for and filter /demineralizer) and valves rooms are rupture details). The rupture producing the lo ated in the 00-2700 quadrant of the greatest blowdown mass and enthalpy, in rector building. The floor elevations t.re from g 6.2 22 Amendmerit 17 l l l
ABWR mc=n Sinndard Plant RIV c ( ) (-)13200 mm to (-)200 mm with separate rooms for the equipment and valves. The design basis break for the RWCU system compartment ne' work is either on 8 in or 6-in double-ended break of the water supply line. This depends upon which break diameter produces the maximum pressurization in the break compartment and any adjoining compartments. This high' energy piping, which connects the RWCU equipment, originates at the reactor pressure vessel. After being routed through the RWCU system the high energy line is directed back to the reactor pressure vessel through special pipe chases and the steam tunnel, In the event of a postulated design basis high energy line break, the steam / air mixture is directed into adjoining compar.ments and eventually purged into the steam tunnel. 6.233.1.23 Main Steam Tunnel The reactor building main steam tunnel is located between the primary containment wssel l and the turbine building. The D13A for the steam tunnel is the double ended break of one of the 28 in main' steam lines. These lines originate at 7 the reactor pressure vessel and are routed (3) through the mainsteam tunnel to the turbine building, in the event of a postulated design basis high energy line break, the pressurized steam / air mixture is held up in the main steam tunnel and purged into the turbine building through blowout pancis, as required. 6.233.13 Design Evt.luation The blowdown mass and enthalpy release rates for the high energy line breaks are determined using Moody's homogeneous equilibrium model. A discussion of the methodology and assumptions used in this model can be found in Reference 2. The resulting compartment pressures and temperatures are calculated by the engineering computer program SCAM. A detailed discussion of the methodology and assumptions used in this program can be found in Reference 4. (D
'n/
Amendment 17 62221
ABWR nwwan Standard Plant ru v c The initial conditions for the analysis include accomplishee $ M.' ..m of lines or ducts that the assumption of 102% rated reactor power and penetrate the containment vessel. Actuation of the compartment pressurcs, temperatures and the containment isof ation syste m is relative humidity to maximize the mass and energy automatically initiated at specific limits j release rates, defined for reactor plant operation. After the L isolation function is initiated, it goes through to completion. . 6.2.4.1 Design liases 6.2.4.1.1 Safety Design liases Blowout panels are used in place of open vent (1) Containment isolation valves provide the { p;ahways when the environmental conditions of one necessary isolation of the containment compartment must be isolated from the environment in the event of accidents or ather in another compartment. The panels are designed conditions and prevent the unfiltered to open upon a differential pressure of 0.50 release of containment contents that psid. cannot be permitted by 10CFR50 or 10CFR100 limits. Leaktightness of the 1 6.23.4 Tests and Inspections valves shall be verified by Type C test. L Testing and inspection of the integrity of (2) Capability for rapid closure or isolation of . secondary containment will be made as part of the all pipes or ducts that penetrate t he testing of the STGS (Subsection 6.5.1). containment is provided by means that provide a containment barrier in such pipes Status lights and alarms for door opening of or ducts sufficient to maintain leakage secondary containment will be tested periodically within permissible limits, G by their operation, with observation of lights and alarms. Leakage testing and inspection of (3) The design of isolation valving for lines all other architectural openings will be made as penetrating the containment follows the they are utilized periodically, requirements of General Design Criteria 54 through 57 to the greatest extent 6.23.5 Instrumentation llequirements practicable consistent with safety and reliability. By their nature, electrical penetrations of secondary containment do not have any (4) Isolation valves for int.trument lines t$ at instrumentation requirements. Piping and llVAC penetrate the drywell/ containment conforms penetrations instrumentation requirements are to the requirements of Regulatory Guide discussed as part of each system's description in 1.14. this ' AR. Details of the initiating signals for isolation are given in Subsection 7.3.1.1.10. (5) Isolation valves, actuators and controls are protected against loss of their safety Certain doors are fitted with status function from missiles and postulated indication lights. effects of high- and moderate energy line ruptures. 6.2.4 Containment Isolation System (6) Design of the containment isolation valves The primary objective of the containment and associated piping and penetrations meets isolation system is to provide protection against the requirements for Seismic Category I releases of radioactive materials to the envir- components. onment as a result of accidents occurring in the systems inside the containment. The objective is (7) Containment isolation valves and associated Amendment 17 (t2 2 )
ABM 23461oorn Standard Plant REV C piping and penetration meet the requirements of the AShiE Boiler and Pressure Vessel Code, listed in Subhiot. 7.1.2. The bases for and reactor vesselisolation control system are h Section III, Class 1, 2, or blC, in assigning certain signals for containment accordahce with their quality group isolation are listed and explained in classification. S u bse ctio ns 7.3.1 a n d 7.6.1. (8) The design of the Control Systems for On signals of high drywell pressure or low automatic containment isolation valves water level in the reactor vessel, all isolation ensures th:t resetting the isolation signal valves that are part of systems not required for shall not result in the automatic reopenly emergency shutdown of the plant are closed. The of containment isolation valves. same signals initiate the operation of systems associated with the emergency core cooling 6.2A.1.2 Design Requirements system. The isolation valves that are part of the ECCS can be closed remote-manually from the The containment isolation system, in general, control room or closed automatically, as closes fluid penetrations that support systems appropriate. not required for emergency operation. Fluid penetrations supporting engineered safety feature 6.2A.2 System Design sptems have remote manual isolation valves which can be closed from the control room, if required. The containment isolation system consiats of the valves and controls required for the The isolation criteria for the determination isolation of lines penetrating the containment. of the quantity and respective locations of Figure 6.2 38 identifies the containment isolation valves for a particular system conform isolation provisions. Table 6.2-7 shows the q to the General Design Criteria 54,55,56,57, pertinent data for the containment isolation g and Regulatory Guide 1.11. Redundancy and valves. A detailed discussion of the controls g physical separation are required in the associated with the containment isolation system W clectrical and mechanical design to ensure that is included in Subsections 7.3.1.1.2 and no single failure in the containment isolation 7.3.1.1.11. system prevents the system from performing its intended functions. Power operated containment isolation valves have indicating switches in the control room to Protection of containment isolation system show whether the valve is open or closui. Loss components from missiles is considered in the of power to each motor-operated valves is design, as well as the inte;rity of the detected and annunciated. Air operated components to withstand seismic occurrences containment isolation valves are designed to without loss of operability. For power-operated fail in a safe position upon loss of air or valves used in series, no single event can power to the solenoid pilot valve. Power for interrupt motive power to both closure devices, valves used in series originates from physically Air-operated containment isolation valves are independent soutces without cross ties to assure designed to fail to the required position for that ne single event can interrupt motive power containment isolation upon loss of the instrument to both closure devices. air supply or electrical power. Tue htSIVs are spring loaded, steam medium, The containment isolation system is designed piston-operated globe valves designed to close to Seismic Category I. Classification of on loss of steam pressure or loss of power to equipment and systems is found in Table 3.21. the solenoid operated pilot valves. Each valve Figure 6.2-38 identifies the quality group has two pilot valves supplied from independent classifications and containment isolation power sources, both of which must be de-ener provisions, gized to close the htSIV. Twu htSIVs are used ie t.eries to assure isolation when needed. Each The criteria for the design of the containment htSIV uses reactor steam pressure for closure upon interruption of electrical power to the pi-g lot valves. A spring closes the valve when Amendment 9 6.2-24
m eiua e l ABWR v n ,y n Standard Plant U reactor pressure vessel to the main turbine and minimite the probability or consequences of so condenser system, penetrate the primary accidental pipe rupture. T he qualit) containment. The main steam drain lines connect requirements for these components ensure that the low points of the steam lines, penetrate the they are designed, fabricated, and tested to tbc primary containment and are routed to the bigbest quality standards of all reactor plant condenser botwell. The RCIC turbine steamlins components. The classification of coroponents connects to the main steam lier in the upper whict comprise the reactor coolant pressure drywell and pen:trates the primary containment, boundary are designed in accordance with the For these lines isolation is provided by ASME Boiler a.nd Pressure Vessel Code, Sect!on automatically actuated block valves, one inside Ill, Citts 1. and one just outside (be containment. It is therefore concluded that the design of 6143.2.12.2 Rif R Shotdown Cooling Une piping system mbich comprise the reactor conlant pressure boundary and penetrate containment Three RilR shutdown cooling lines connect to satisfies Criterion 55.
.be reactor vessel and penetrate the primary containment. Isolation is provided by two 6.2.43.2.2 Evaluation Against Criterion 56 automatically actuated block valves, one inside Criterion 56 requires that lines which sad other outside ibe containment.
penetrate the containment and communicate with 6.2.43.2.1.23 Reactor Water Cleanup System the containment interior must have two isolation valves; one inside the containtnent, and one Suetion une outside, unless it can be demonstrated that tbc containment isolation provisions for a specific The RWCU inkes its suction from the bottom clats of lines are acceptable on some other head of Ibe RPV and from tbc R11R *B' shutdown O b cooling suction line. The RWCU suction line is basis, isolated by two autocantic motor. operated gate valves on the inside and outside of in e Although a word for. word comparison with Criterion 56 in some cases is not practical, it containment. Should a break occur in the RWCU system, tbc check valves would prevent backflow is possitic to demonstrate adequate isolation from the RPV and the isnlation valves would provisions on some other defined basis. prevent forward flow from ibe RPV. 6J.43.2.2.1 influent unes to Suppression RWCU pun.ps, best eacbangers and filter Pool demineralizers are located outside the drywell. Figure 6.2 38 identifies the isolation 62.43.2.13 Conclustnn on Criterion !$ provisions in the influent lines to the suppression pool. In order to asture protection against the 6.2.43.2.2.1.1 IIPCr and RHR Test and Pump l consequences of accidents involving tbc release of radioactive material, pipes which form the Minimum Mcm Bnats unes reactor coolant pressure boundary have been shows Tbc liPCF and RHR test and pump minimum flow l to provide tdequate isolation capabilities on a case by. case basis, in all cases, a minimum of bypass lines have isolation capabiliiies two barr2:rs were shown to wotect against the commensurete with the is portance to safety of reltate of radioactive materials. isolating these lines. Each line has a rootor-operated valve located outside the containment. In addition to meeting the isolation Contain:nent isolation requirements are met on ibe basis that the lines are low pressure lines requirements stated in Criterion 55, the constructed to the same quality standards pressure.retainit.g components which comprise the reactor cooltet pressure boundary are designed to commensurate with their importance to safety. meet other appropriate requirements which Furthermore, the consequences of a break in
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ABWR .w =n Slapdard1%fil Frv c these lines result ir. no significant safety con. 6.2.43.2.2.2.1 RilR. HCIC and IIPCF Llues sideration. All of the "~ es terminate below the minimum drawdown level in the suppression pool. The RilR, RCIC, and IIPCF suction lines contain motor. operated, remote-manually actuated The test return lines are also used for gate salves which provide assurance of itolating suppression pool return flow during other modes these lines in the event of a break. These of operation, in this manner, the number of valves also provide long. term leakage control, penetrations are reduced, thus minimiring the in addition, the suction piping f rom the potential pathwaya foi radioactive material suppression pool rnust be available for longactm release. Typically, pump minimum flow bypass usage following a design basis LOCA. and, as lines join the resl.cctive test return lines such, is designed to the quality standards downstream of the test return isolation valve, commensurate with its importance to safety. The The bypass lines are isolated by motor. operated RilR discharge line fill system suction lines valves in series with a restricting orifice, have manual valves for operational purposes. These systems are isolated from the containment 6.2.43.2.2.1.2 RCIC Turbine 1:shaust and Pump by the respective RilR pump suction salves from hilnimum flow Ilypass Lines suppression pool. The RCIC turbine exhaust line which 6.2.43.2.2.2.2 SPCU Suetion line penetrates the containment and discharges to the suppression pool is equipped with a normally The SPCU system suction line has two open, motor. operated, remote manually actuated isolation valves, flowever, because the gate valve located as close to the containment as penetration is under water, both the isolation possible, in addition, there is a simple check valves are located outside containment. The valve upstream of the gate valve which provides first valve is located as close as possible to positive actuation for immediate isolation in the the containment, and the second is located to event of a break upstream of this valve. The provide adequate separation from the first. & gate valve in the RCIC turbine exhaust is W designed to be locked open in the control roem 6.2.43.2.23 ACS Lines To Containment and is interlocked to preclude opening of the inlet steam valve to the turbine until the The Atmospheric Control Sple (ACS) has both turbine exhaust valve is in its full open influent and effluent lines which per.ctrate the position. The RCIC pump minimum flow bypass line containment. Iloth isolation valves on these is isolated by a normally closed, remote manually lines are outside of the containment vessel to , actuated valve outside containment. provide accessibility Io the valves. The 5 inboard valve is located as close as practical 0 6.2.4J.2.2.1J SPCU Dischar1;e Line to the containment vessel. The piping to both valves is an extension of the containment The suppression pool cleanup (SPCU) system boundary. discharge line to the suppression pool (i.e., containment penetration, piping and isolation 6.t4J.2.2.4 Conclusion on Criterion 56 valves) is designed to Seismic Category 1, ash 1E Section Ill, Class 2 requirements. In order to assure protection against the consequences of accidents involving release of 6.2.43.2.2.2 Lfiluent Lines from Supprnsion significant amounts of radioactive materials, Pool pipes that penetrate the containment hase been demonstrated to provide isolation capabilities Figure 6.2 38 identifies the isolation on a case by. case basis in accordance with provisions in the effluent lines from the Criterion 56. suppression pool. O Amendmem 17 62N
ABWR muun i l Slandard Plut niv. c ( ') in addition to meeting isolation requirements,
"' the pressure retaining components of these systems are designed to the quality standards commensurate with their importance to safety.
6.2.43.2.3 Evaluation Against Criterion 57 1.ines penetrating the primary containment, , which are governed by neither Criterion $$ not ' Criterion $6, comprise the closed system isolation valve group. ['T
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l l I
/^\
O Amendment 17 6.2-301 l
ABWR m-n Slatlilatd. Plant fuv. c / (16) The primary containment purge system will acquired, through a pathway from the wetwell (3) sid in the long term post accident cleanup altspace to the stack. The pathway is isolated ope r ation. The primary containment during normal operation with two rupture dids. atmosphere will be purged through the SG15 to the outside environment. Nitrogen makeup The following modes of operation are provided; will be available durlug the purging operation. (1) Startup Inerting. Liquid nitrogen is vaporired with steam or electric heaters to (17) The system is also designed to re: case a temperature greater than 200 F and is containment pressure before uncontrolled injected into the wetwell and the drywell. containment failure could occur. The nitrogen will be mixed wi.h the frimary containment atmosphere by the drywell 6.2.5.2 System Design coolers in the drywell and, if necessary, by the sprays in the wetwell. 6.2J.2.1 General (2) Normal Maintenance of inert Condition. A The ACS provides control over hydrogen and nitrogen makeup system automatically sup-oxygen generated following a LOCA. In an inerted plies nitrogen to the wetwell and upper containment, mixing of any hydrogen generated is drywell to maintain a slightly positive not sequired. Any oxygen evolution f rom 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 containtnent. An increase in con-to provide mixing. Spray operation will provide further assurance that the drywell or wetwell is ta!nment through pressure the drywell bleedis controlled line. by venting l unifovmly mixed. The system consists of the following features: (3) Shutdown Deinerting. Air is provided to f \ the drywell and wetwell by the primary C (1) Atmospheric mixing is achieved by natural containment ilVAC purge supply fan. Ihhaust 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 llVAC are used to control pressure in the primary or SGTS, as required, containment. (4) Overpressure Protection, if the wetwell (2) The primary containment nitrogen purge pressure inct cases to about 5.6 establishes and maintains an oxygen - Lg/cm2 g, the rupture disks will open. deficient atmosphere (13.5 volume percent) The overall containment pressure decreases in the primary containment Juring normal as venti:g continues. Later, the operators operation. can close the two 3DA air. operated b u t ' s t f l y v a l v e s ' c, r c.e s t a blis h (3) The redundant oxygen anal >ver sy. tem (CAMS) eouuinment is01ation as tequired. measures oxygen in the drywell and suppression chamber. Oxygen concentration The following interfaces with other systems are displayed in the main control room. are provided: Description of safety related display instrumentation for containment monitoring (1) Residualllent Removal System (RllR Ell). is provided in Chapter 7, Electrical The RilR provides post accident suppression requirements for equipment associated with pool cooling as necessary following heat 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 standards 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 RilR heat 7] V protection to relieve containment pressure, as exchanger sizing. The wetwell spray Ammendment 17 62 M
AllWR nuia. SAndard Plant RtYl portion of the RilR may be activated during the inerting and deinerting processes to help rutting by reducing poeleting. Wetwell spray would also serve to accelerate deacration of the suppression pool water, though the impact of the dissolved oxygen on wetwell altspace oxygen concentration in very Srnali. Wetwell spray is not required to provide mixing in the wetwell. Tbc RilR also provides cooling water to the exhaust flow from tbc FCS. .' ' 2) Dryv !! Cooling System (DWC T41): Drywell j vjong system provides enough elreulation so all portions of the upper and lower drywell, the drywell head area, and the vessel support skirt area such that this mixing occurs and does not O O Amendment 11 62331
)
ABWR unun l Standard Plant nvc n Except for inspections and actions taken After initial ILRT, a set of three Type A l above, no preliminary leak detection surveys tests shall be performed at approximately equal (v) and repairs shall be performed prior to the intervals during each 10 yr service period, with
)
i conduct of the Type A test. the third test of each set coinciding with the er d of each 10 yr major ic. service inspection (2) Closure of containment isolation valves shall shutdown. In addition, any major modification ' be accomplished by normal rnode of actuation or replacement of components of the primary and without preliminary exercises or adjust. reactor containment performed after the initial ments. All malfunctions and subsequent cor. ILRT shall be followed by either a Type A or a rective actions shall be reported to the NRC. Type !! test of the area affected by the modifi-cation, with the affected area to meet the appli-6.2.6.1.2.2 Acceptance Criteria cable acceptance criteria. The basis for the frequency of testing is established in accor. shall not dance with 10CFR$0, Appendix J The measured leakage rate L"'lhe initial cacced 0.7$ L as established b) t ILR T. 6.2.6.1.3 Additional Criteria for Integrated Rate Test (1) If during a Type A test including the supple-mental test, potentially excessive leakage (1) Those portions of fluids systerr s that are paths are identified which will interfere part of the reactor coolant pressure bound-with satisfactory completion of the test, or ary, that are open directly to the primary which result in the Type A test not rnecting reactor containment atmosphere under post. the acceptance criteria, the Type A test accident conditions and become an extension shall be terminated and the leakage through of the boundary of the primary reactor con-such paths shall be rneasured using local leak. tainment, shall be opened or vented to the ( age testing methods. Repairs and/or adjust- containment atmosphere prior to or during p) $ ments to equipment shall be made and a Type A the Type A test. Portions of closed sys. ( test performed. The corrective action taken tems inside containment that penetrate and the change in leakage rate determined primary containment and are not relied upon from the tests and overall integrated leakage for containment isolation purposes follow-determined from the local leak and Type A ing a LOCA shall be vented to the contain-tests shall be included in the report submit. ment atmosphere, ted to the NRC. (2) All vented systems shall be drained of wa-(2) If any Type A test fails to meet the accep- ter to the extent necessary to ensure expo-tance criteria, prior to corrective action, sure of the system primary containment iso-the test schedule applicable to subsequent lation valves to the containment air test Type A test shall be subject to review and pressure. approval by the NRC, (3) Those portions of fluid systems that pene-(3) If two consecutive periodic Type A tests fall trate primary containment, that are exter. to meet the applicable acceptance criteria, nel to containment and are not designed to prior to corrective action, notwithstanding provide a containment isolation barrier, the established periodic retest schedule, a shall be vented to the outside atmosphere Type A test shall be performed at each plant as applicable, to assure that full post.ac. f shutdown for major nfueling, or approximate-ly every 18 months, whichever occurs first, cident differential pressure is maintained across the cor tainment isolation barrier, until two consecutive Type A tests n. O the acceptance criteria, after which time the p.e (4) Systems that are required to maintain the viously established periodic retest schedule plant in a safe condition during the Type A may be resumed, test shall be operable in their normal mode and are not vented. f\ 6.2.6.1.2.3 Test frequency Amendment 3 6.2 41
ABWR momn Standard Pltint r n .r (5) Systems that are normally filled with water 6.2.6.2.3 Retest Trrquency and operating under post LOCA conditions need not be vented. In compliance with the requirement of Sec-tion Ill.D.2(a) of Appendix J fp 10CTR Part 50, l(6) LLRT results from items 4 and 5 above shall type B tests (except for air locks) are per-be added to the ILRT results. formed during each reactor shutdown for major fuel reloading, or other convenient intervals, 6.2.6.2 Containment Penetration tsalage Rate but in no case at intervals greater than two Test (Type 11) years.' Air locks opened when containment integrity is required will be tested in manual 6.2.6.2.1 General mode within 3 days of being opened, if the air , lock is to be opened more frequently than once , Containment penetrations whose designs incor- every 3 days, the air lock will be tested at f potate resilient seals, bellows, gaskets, or least once every 3 days during the period of A scalant compounds, airlocks and lock door seals, frequent openings. Air locks will be tested at equipment and acce.,s hatch seals, and electrical initial fuel loading, and at least once every 6 canisters, and other such penetrations are leak months thereafter. Airlocks may be tested at tested during preoperational testing and at peri- full power so as to avoid shutting oown. These odic intervals thereafter in conformance to Type airlocks contain no inflatable seals. 4 D leakage rate tests defined in Appendix J of g 10CFR50. A list of all containment penetrations p is provided in Table 6.2 8. The leak tests Main control room readout of time to next
- ensure the continuing structural and leak test, test completion and test results is integrity of the penetrations, provided. An alarm sounds if the specified interval passes without a test being effected.
To facilitate local leak testing, a perma- No direct, safety.related function is served by nently installed system may be provided, consist- the seal test instrumentation system. ing of a pressurized gas source (nitrogen or air) and the manifolding and valving necessary to 6.2.6.2.4 Design Prosisjons for Periodic subdivide the testable penetrations into groups Pressurization of two to five. Each group is then pressurized, and if any leakage is detected (by pressure decay in order to assure the capability of the or flow meter), individual penetrations can be containment to withstand the application of peak isolated and tested until the source and nature accident pressure at any time during plant life of the leak is determined. All Type B tests are for the purpose of performing ILRTs, close atten-performed at containment peak accident pressure, tion is given to certain design and maintenance Pa. The localleak detection tests of Type B and provisions. Specifically, the effects of corro-Type C (Subsection 6.2.6.3) must be completed sion on the structural integrity of the contain-prior to the preoperational or periodic Type A ment are compensated for by the inclusion of a tests. 60 yr service life corrosion allowance, where applicable. Other design features that have the 6.2.6.2.2 Acceptance Criteria potential to deteriorate with age, such as flexible seals, are carefully inspected and The combined leakage rate of all components tested as outlined in Subsection 6.2.6.2.2. In subject to Type B and Type C tests shall not ex- this manner, the structural and leakage integ-cced 60% of La (cfm). If repairs are required rity of the containment remains essentially the to meet this limit, the results shall be reported same as originally accepted. in a separate summary to the NRC. The summary shall include the structural conditions of the components which contributed to failure.
*In compliance with the requirement of Section
{ All Typc B tests are pcriormed at Ill D.2(b)(iii) of Appendix J ro 10CFR Part 50 $ containment peak accident pressure PA. The acceptance criteria is given in Chapter 16. Amendment 17 6242
ABWR msun Standard Plant nry c O 6.2.6.3 Contalument isolation Valte leakage a flange), that are connected between isolation b Rate Test (1)pe C) valves and form a part of the primary containment boundary need not be Type.C tested 6.2.6.3.1 General due to their infrequent use and multiple barriers as long as the barrier configurations g Type C tests will be performed on all are maintained using an administrative control g containment isolation valves requ'. red to be program, tested per 10CFR50 Appendix J. All testing is performed pneumatically, except hydraulic testing For Type C testing of containment penetra-may be performed on isolation valve Type C tests tions, all testing will be done in the correct using water as a sealant provided that the valves direction unless it can be shown that testing in will be demonstrated to exhibit leakage rates the reverse direction is equivalent, or more con. that do not exceed those in the ABWR standard servative. The correct direction for this technical specifications. design is defined as flow from inside the containment to outside the containment. Type C tests (like Type B test) are performed by local pressuriration using either pressure 6.2.6.3.2 Acceptar.ec Criteria decay or flowmeter method. The test pressure is applied in the same direction as when the valve The combined leakage rate of all components is required to perform its safety function, un- subj e ct to Type B and Type C (Subsection less it can be shown that results from tests with 6.2.6.3) tests shall not exceed 60% of 3L . If pressure applied in a different direction are repairs are required to meet this limit, the re-equivalent or conservative. For the pressure de- sults shall be reported in a separate summary to cay method, test volume is pressurized with air the NRC, to include the structural conditions of or nitrogen to at least Pa. The rate of decay the components which contributed to the failure. of pressure of the known test volume is monitored ! b to calculate leakage rate. For the flowmeter method, required pressure is maintained in the Tests 6.2.6.1 Schedulimit and Reporting or Periodic l ( test volume by making up air, nitrogen or water (if applicable) through a calibrated flowmeter. The periodic leakage rate test schedules for The flowmeter fluid flow rate is the isolation Type A, B and C tests are described in Chapter valve (or Type B test solume) leakage rate. 16. All iriolation valve seats which are exposed to Type B and C tests may be conducted at any containment atmosphere subsequent to a LOCA are time during normal plant operations or during tested with air or nitrogen at containment peak shutdown periods, as long as the time interval accident pressure, Pa. betwnn tests for any individual Type B or C tests does not exceed 2 years. Each time a Type ( MSIVs and isolation valves isolated from a B or C test is completed, the overall total leak-age rate for all required Type B and C tests is f
=
sealing system will use a test pressure of at least Pa. updated to reflect the most recent test re-suits. In addition to the periodic tests, any ! Those valves which are in lines designed to major modification, replacement of component be, or remain, filled with a liquid for at least which is part of the primary reactor containment 30 days subsequent to a loss-of coolant accident boundary, or rescaling a seal welded door, per-i are leakage rate tested with that liquid. The formed after the preopertional leakage rate test . liquid leakage measured is not converted to will be followed by either a Type A, Type B, or y l " l equivalent air leakage not added to the Type B Type C test as applicable for the area effected l and C test total. by the modification. Type A, B and C test results shall be submitted to the NRC in the All test connections, vent lines, or drain summary report approximately three months after lines consisting of double barrier (e.g. 2 valves each test. s in series, one valve and a cap, or one valve and Amendment 17 6 2-13
ABWR msen Standard Plant Riw c locluded in the leak rate test summary report will be, a report detailing the containment in-spection, a report detailing any repairs neces-sary to pass the tests, and the leak rate test results. 6.2.6.5 SpecialTesting Requirements The maximum allowable leakage rate into the secondary containment and the means to serify that the inleakage rate has not been exceeded, as well as the containment leakage rate to the environment, are discussed in Subsections 6.2.3 and 6.5.1.3. O O I Ammendment 17 6.2-431
i I ABWR msman Standard Plant now c 6.2.7 References (^]\
'w
- 1. W.J. Bilanin, The G.E. Afark til Pressure Suppression Containtnent Analytical blodel, June 1974, (NEDO 20533).
i
- 2. F.J. Moody, Afaxirnurn Discharge 1: ate of j Llquid 1'apor Afi.ttures frorn l'essels, General 'l Electric Company, Report No. NEDO.21052, September,1975. .
- 3. W.J. Dilanin, The G.E. A!,rk 111 Pressure Suppression Containtnent Analytical Afodel, Supplement 1, September 1975 (NEDO 20533-1).
- 4. J.P. Dougherty, SCAM Subcornpartrnent Analysis Alethod, January 1977, (NEDE-21526).
i l L
~ % /'
Amendment 17 6244 l
ABWR HAM 00AD Standard Plant ney. c TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION ) STANDHY LIQUID CONTROL SYSTEM Vahe No. C41.P007/F008 C41 l'OO6A C41.f E D SMR 11g 951 911 911 Applicable Basis GDC5$ ODC$$ ODC$$ 11uld Doron/ Water Doron/ Water Doron/ Water IJne Sise 40A 40A 40A l ESF No No No l l leakape Class (a) (a) (a) Imation O/l O O Type C leak Test No (w) No (w) No (w) l VaheType- Swing Check Globe Olot* Operator N/A Motor Motor PrL Actuacion Self Dect. Dect. Sec. Actuation N/A Man. Man. Normal Position Shut Shut Shut Shutdown Position Shut Shut Shut Post Ace Posi lon Shut $ hut Shut PwT Fall Position As is As is Asis Cont. Isa, Sig. I') N/A N/A N/A Closure Time pec) Inst, 24 24 Pwr Source (Div)- N/A I- 11 O Amendment 17 6.2 50,3
. . . . - . - . , - . - . , , , - . - . _ - - - . . . ~ . - . - , . . - - - . - - . - , -
ABWR 23^6100^n SlanIIard Plani Rev C TAULE 6.2 7 (Continutd) g CONTAINMENT ISOIATION VALVE INFORMATION CONTAIN'elENT ATh!OSPilERE h!ONITORING Valve No. D2SRIO1A/D D21ROIA/B D211005A/D D2kMKEA/D D21TUD7A/D D21RORA/D SMR l'ig 7.47c 7.47c 7.47c 7.47c 7.47c 7/r7e Applicable Basis GDC $6 GDC $6 ODCS6 ODC$6 ODC56 ODC56 l RO 1.11 Fluid DW Atmcs DW Atmos DW AtmN WW Atmos WW Atmos WW Atmcs line She 20A 20A 20A 20A 20A 20A FSF No No No No No No 14akage Class (a) (a) (a) (a) (a) (a) Imation O O O O O O Type C leak Test No(m) No(0 No(f) No(0 l No(r) No(0 VaheType Gate Gkte Gkte Globe Gkte G k+c Operator Solenoid Manual Manus! Manual Manual Manual Pit Actuadon Dec. Dec. Dec. Dec. Dec. Dec. Sec. Actuation N/A N/A N/A N/A N/A N/A Nonnal Position Open Shut Shut Shut Shut Shut Shutdown Position Shut Shut Shut Shut Shut Shut Post Ace Position Open Open Open Open Open Open Pwr Fall Position As is As is As is As is As b As is Cont, Iso. Sig.(# N/A N/A N/A N/A N/A N/A Closure Time (sec) N/A N/A N/A N/A N/A N/A Pur Source (Dii) I/II 1 /11 1/II 1/11 1/11 1/11 l Amendment 17 6.2 50 3a
ABWR 2auwAD Standard Plant nev c TAHLE 6.2 7 (Continued) CONTAINMENT ISOIATION VAIXE INFORMATION CONTAINMENT ATMOSPHERE MONITORING (Continued) Vahe No. D3It09A/II D211910A/II D211011A/B D211V11A/D . D21191M/D D211984A/11 1
$MR 11g 747c 747c 767c 741c 747c 747c C S O Basis GDC56 GDC56 GDC.% GDC56 ODC $6 GDC56 l 1
11uld DW Atmos DW Atm(s DW Atmts WW Atmos WW Atmos WW Atmos ; 1 Joe slee - 20A 20A 20A 2GA 204 20A 1:SF No No No No No No l 14akage Clau (a) (a) (a) (a) (a) (a) Imelloa O O O O O O Type C 14mk Ted No(O No(O No(0 No(0 No(0 No(0 Valve Type Globe Globe Globe Globe Globe Ok&c Operator N/A N/A N/A N/A N/A N/A PrL Actuation Manual Manual Manual Manual Manual Manual Sec. Actuation N/A N/A N/A N/A N/A N/A I Normal Poshn Open Open Open Open Open Onen Shwedown Position Open . Open Open Open Oper i cn Pod Ace Posen Open Open Open Open Open Open Pur Fall Position .N/A N/A N/A N/A N/A N/A Cont. Isa. Sig. } N/A N/A N/A N/A N/A N/A Closen Tbne (sec) N/A N/A N/A N/A N/A N/A Pwr Source (Div) N/A N/A N/A N/A N/A N/A l 0
- Amendment 17 6 2-503b
ABWR 23^uc3Ali Standard Plant ner c TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFOlulATION g RESIDUAL llEAT REh! OVAL SYS1Ihl WE1Wl:LL SI' RAY Yalve No. I:ll IVl9D IIll.1919C
$54R lig $ 410e $4-10g Appucable Basis GDC56 GDC 56 Ilund Water Water line Stre 100A 100A l'Sr Yes Yes 14akar Class (a) (a) tornilon O O Tnw C teak Test No(g) No(g)
Valve Type Gate Gate Operator Mosor Motor O PrL Actuation Dec. Dec. Sec. Actuation Manual Manual Normal Postalon O med Cksed Shutdown Position Oosed Gused Post Act Position Ocsed Cksed Pwr Fall Position As is As is Cont. Im Sig.( } N/A N/A Closure Time (sec) 20 20 Pwr Source (Div) !! Ill O Amendment 17 6.2 50.4
ABWR - zwmAn , Standard Plant Rev. C TAHLE 6.2 7 (Continued) CGNTAINMEST ISOLATION VAINE INFORMATION RESIDUAL llEAT REMOVAL SYSTEM (Continued) DRYWELL SPRAY Yalee No. till P01711 I!!! 191til 12111917C !!111918C SMR Ils $.4-10e $ 410e $ 4-10g $.41Dg ApplAcable Besh ODC $6 ODC$6 GDC56 ODC $6 , fluid Water Wati t Water Water IJae$he 15M 230A 25M LHIA ESI' Yes .Yes Yes Yes 14akage Class (a) (a) (a) (a) i Imrellom O O O O Type C lek Tese No(g) No(g) No(g) No(g) l Valve Type Globe Osts Globe Gate Operator Motor Motor Motor Motor Pit A.1mation 1'lec. 13cc. [3ec. 12ec. Sec.Actuallee Manual Manual Manual Manual Norn44 Postilon Shut Shut $ hut Shut 6se'tdown Position $ hut Shut Shut Shut l ! Post Ace Position Shut Shut Shut Shut l- Pwr l'all Position As is As is As is As is Cent. Isa. Sig.I' N/A N/A N/A N/A Closure Tinw (ser) 50 50 50 50 Pwr Soorte (Div) 11 11 Ill til - oV l Amendrnent 17 6250.5
. . _ _ _ - , - ~ _ _ . . _ , _ . _ - _ , . - , . . . _ _ _ _ . _ _ . , . _ _ . .
ABWR 236100^n Standitrd Plant Ret c TAllLE 6.2 7 (Continued) h CONTAINMENT ISOLATION VALVE INFORMATION RESIDUAL llEAT REhlOVAL SYSTD1 (Continued) h11NihlOh! FLOW LINE Valve No. El1IV2tA E11192111 Ell IV21C SMR lig 3 4-10c $ 410d $.4-10f AgyliraMe Basle ODC $6 ODC $6 ODC56 11uld Water Water Water LJne Stre 100A 100A ITIA 131' Yes Yes Yes laakage Clau (a) (a) (a) laation O O O l 'I)pe C leak Test No(h) No(h) No(h) Valve Type Gate Gate Gate Operator Motor Motor M< ant PrL Actuation Dec. Dec. Dec. Sec. Actuatloa Manual Manual Manual Normal Position Open Open ( ,, Shutdown Position Shut shut St ut Post Are Fosition $ hut $hus Shu* Pur l'all Potition As is As is As is Cont. lso. Sig. # N/A N/A N/A Closure Time (sec) 20 20 20 Pwe Source (Die) I Il 111 O Amc.edmat 17 6.2-50 6
- - _ - .a., - . _ - - . - -
t ABWR nA6ima ! Staridard Plant nev. c ! l TAllLE 6.2 7 (Continued) ; i CONTAINMENT ISOLATION VALVE INFORMATION l RESIDUAL HEAT REMOYAL SYSTEM (Continued) : S/P COOLING , Yalve No. ElllusA E11 It31A EllIV$11 E11103111 Ell IVEC Ell It31C l SMR Tig $.4-lDe $.410e . $.4-10d 5.410d $ 4-10f $.4-10f i. Appikable leasts ODC$6 ODC56 ODC 56 ODC $6 ODC56 ODC$6 nund Water Watet Water Walet - Water Watet E (Jae Slee 200A 100A 200A 100A 200A 100A ESF Yes Yes - Ves Yes Yes Yes 14akaye Clau (a) (a) (a) (a) (e) (a) Imation O O O O O O Type C 14ek Test . No0) No(J) No(j) No0) No0) No0) Valve Type Globs Okee Globe Olote 01ot4 Olobe Operator- Motor Motor Motor Motor Motor Motor ; L PrL Actuallon . Illce. Dec. Dec. Dec. Dec. Dec. , Sec. Actuation Manust ' Manual Manual Manual Manual Manual i Normal Position Shut Shut Shut Shut Shut Shut Shutdown Position Shut Shut $ hut Shut Shut Shut Pont Att Position Shut Shut Shut Shut Shut Shut Pwr Fan Position As is As is ~ As is As is Asis As is ; Cont. loo. Sis.I' 'N/A N/A N/A N/A N/A N/A Closur,11me (sec) $0 - 20 50 20 50 20 i Per Suerce (Div) i I 11 II Ill !!! l l l i l l I Amendment 17 6.2 50,7
ABWR nuieDan Standard Plant nev c TAllLE 6.2 7 (Continued) l CONTAINMENT ISOLATION VALVE INFollMATION RESIDUAL IIEAT REMOVAL SYS1UI (Continued) S/I' SUCTION (LIT 1.) I Valo No. Ell RotA Ell l'001D Ell RO1C l 1 EMR Fig 5 410c $ 4103 $ 410f I AppliceMe Basis GDC 56 GDC 56 GDC 56 Field Water Water Watet IJne She 45M 450A 450A 1:SF Yes Yes Yes LasLage Class (a) (a) (a) locailon O O O Type C 14ak Teit No(i) No(s) No(l) VaheTme Gate Gate Gate Operator Motor Moint Motor O PrL Actuation Dec. Dec. Dec. Sec. Actuation Manual Manual Manual Normal Position Ogn Open Open shutdown Position $ hut Shut Shui Post Act Position Shut Shut Shut Pwr Fall Position As is As is Asis Cont. Isa. Sig. , N/A N/A N/A Closun Time (sec) 90 90 90 Pwr Source (Dh) I !! !!! l l I O l l Amendment 17 6.2508 l l
ABWR muooAn nu e Standard Plant ~. TAHLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION l RESIDUAL llEAT REMOVAL SYSTEM (Contland) INilOAMil SHUTIK)WN COOLING l i Valo No. !!!!1910A Ell.IVIOD !!16fWC
$141t Tis - 3.410b 5.410b $.410h -
l Applicable Basis ODC$3 GDC53 ODCS$ , 11uld - Water Water Weier IJae Elw 350A 330A 130A 13F Yes Yes Yes ! Isakape Class (a) (a) (a) (Aestion l 3 I - Type C imak Test .No(n) No(n) No(n) l Valw Type Gate Gate Unte - Operator Motor Motor Motor Prl. Actuation 13ec. IUce, 13ce, t Sec. Actuation Manual Manual - Menwal - NormalPointion - - Shut Shut Shut 4 Sheldown Postelen Shut - $ hut Shut i Post Ace Position Shst Shut Shut I Pwr Fall Position Asis As is As is Cont. Inc. Sig.(t) A. M, U RM Z A. M, U, RM,Z A. M, U, RM,Z Closure 'I1nw (ut) 70 70 t Pwr Source (Uli) I 11 Ill : t I e,. 1 i Amendment 17 62509 _.A-,-;..-. -._~._2._,.._._._.,a._...--,. - . . . . , - ~ . _ - - _ - - - . . ~ . _.. , . , . _ - , _ - _ , . . , . . _ . _ . _ _ - - . . . . . _ . . . _ . . . . _ . - _ _ . . . . . , _ -
ABWR naam^u Standard Plant _ Rev c TAllLE 6.2 7 (Continued) h CONTAlNMENT ISotATION VAINE INFORMATION RESIDUAL llEAT R1310 VAL SYSTEM (Continut d) OUTilOARD SilUTD0%W COOLING Vaht No. Ell.PollA 1:111911D Ell-1911C 554R ilg 5.410te 5.410b $ 4105 AppikaMe Basis ODC SS ODC 55 ODC 35 11old Water Water Water line She 350A 350A 350A Est Yt4 Yes Yes Laakage Class (a) (a) (a) location O O O Type C 14ak Test l No(n) No(n) No(n) Valve bre Gate Gate Gate Operator Motor Maior Motor PrL Actuation illec. Elec. Elec. See, Actuation Manual Manual Mar.ual Nonnal Position Shut Shut Shut Shutdown Position Shut Shut Shut Post Ace Position Shut Shut Shut Per Fall Position As is As is As is , Cont. Iso. Sig. A, M, U, RM,Z A, M. U, RM,Z A. M, U, RM,2 Closure Time (set) 70 70 70 Pwr Source (Div) Il ill i O Amendment 17 6.25010
ABWR nan *An Standard Plant nev c ! 4 TABLE 6.2 7 (Continued) l CONTAINMENT ISOLATION VALVE INFORMATION RESIDUAL llEAT NEMOYAL SYSTEM (Continued) INJECTION AND'IISTABLE CilECK Vain No. 011.Iw3H Ell lull Ell.I'ou$C E11 IV4C , l
$1AR 51g 3 410e 3.410e $ 4-10g 3.410g AppliceMe Basis ODC$$ ODC$$ ODC$$ ODC$$
11 eld Water Water Water Water llse Sise 250A 230A 230A 1%A ESF Yes Yes Yes Yes
!aakase Class (a) (a) (a) (a)
I location 1 O I O l Type C taak Tese No(k) No(k) No(k) No(k) l
. Valve Type Gate Check Gate Check Operator Motor N/A Motor N/A Pet. Actuation Dec. Self Dec. $cif '
Sec. Actontion Manual N/A Manual N/A Normal Position Shut Shut Shut Shut Shutdown Position $ hut Shut Shut Shut Post Act Position Shut Shut Shut Shut Pwi rail Position Asis N/A . As is N/A Cont. Isa. Sig,I'} N/A N/A N/A N/A Closure Tinw (set) 10 Inst. 10, last. P*r Soutre (Div) !! N/A til N/A T Amendment 17 6.2 50.11
. _ _ _ - - , __.,_.---.,_.;,.-..._-.....__._..--.-.m. -.--.-.--_,..a.-_-._a-.-,--_, _-
ABWR 2aaman Standard Plant nev c TAllLE 6.2 7 (Continued) g CONTAINMENT ISOLATION VALVE INFORMATION tilGil PRESSURE CORE FLOODER SYSTEM S/P SUCTION Valve No. L'22-LWD l'221006C SMR Ilg - 6 17b 63.7b Applicable llasts GDC 56 ODC S6 Ilund Water Water IJne She 400A 4%4 L5F Yes Ye4 leakage Class (e) (a) laation O O Type C 14ak Test No(i) No(i) Valve 'Iype Gale Gate Operator Motor Motor PrL Actuation Dec. Dec. Sec. Actuation Manual Manual Nonnal Position Shut Shut i Shutdan Position Shut Shut Post Act Position Shut Shut I%t l'all Position As is As is Cont.160. Sig. ' N/A N/A , Closure T6me (set) 80 80 l 1 Pwr soorce (Div) II III l 1 O
.__t,, m.se l,
ABWR 21A6100AD Standard PISHf Rev.C TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION , lilGH PRESSURE CORE FLOODER SYSTEM (Continued)
'lEST AND MINIMUM FIX)W Valve No. - !!22 !UO9D 122 l'U10D 122It09C 122-IVibC 617b 617h
$14R Ilg 6S7b 63 7b Applicable Basis ODC$6 ODCS6 ODC56 ODCS6 (
llend Water Water Water Watet llae Sise 10M 73A 100A 75A ESF Yes Yes Yes Yes 14akage Class (a) (a) (a) (a) Imation O O O O Type C 14ak Test No(h) No(h) No(h) No(h) Valve Type ~ Olobe Gate Globe Oste O Operatot Motor Motor . Motor Motor PrL Actuation Dec.- - Dec. Dec. Dec. Sec. Actuation Manual Manual Manual Manual Noratal Postalon Shut Shut Shut Shut Shutdown Position Shut Shut Shut Shut
- Post Acc Position Shut Shut Shut Shut Pwr Fall Position As is . As is As is As is Cent. Iso. Sig. # N/A N/A N/A N/A Closure Mme (sec) 20 20 20 20 Pwr Source (Div) Il !! 111 - III k
O Amendment 17 6.2-50.13
, ._: _ a :. ._ _ . . - . . . _ _ , , . .___,.._;_;_a...,_.,___._a...,.._. ,__...-._.;_,.-.:.._.,.._..
ABWR n^umAn n r., c Standard Plant TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFOIntATION lilGli PRESSURE CORE 1100 DER SYSIDI (Continued) INJECTION Valve No. 122.MC33 E22-MolB C.2-It03C 122ITO4C SSAR Ils 617a 6S7a 617a 6S7a 1 Applicable Basis ODC$$ ODC $5 GDC$$ GDC$$ ) i 11uld Water Water Water Water IJne Size 200A 200A 200A 200A ESF Yes Yes Yes Yes leakage Class (a) (a) (a) (a) Incation O I O I Type C lank Test No(k) No(k) No(k) No(k) Valve Type Gate Check Gate Check Operator Motor N/A Motor N/A O PrL Actuation Dec. Self Dec. Self l Sec. Actuation Maisual N/A Manual N/A Normaf rosition Shut Shut Shut Shut Shutdown Position Shut Shut Shut Shut Post Ace Position Shut Shut Shut Shut Pwr Fall Position As is N/A As is N/A Cont. Iso. Sig.(#) N/A N/A N/A N/A Closure Time (sec) 36 Inst. 36 Inst. Pwr Source (Div) !! N/A Ill N/A l Amendment 17 6.2-50.14 l
ABWR 2 mi*^n n~ c l Standard Plant TAllLE 6.2-7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION NUCLEAR BOILER SYSTEM MAIN STF.AM LINES A, B, C AND D Vales No, it21 lusA/11 Il21.IW9A/Il C/D C/D SMR Fig 5.1 k 5.1 k Applicable Itasls ODC 55 ODC55 11 id Steam Steam IJne Sire 'fJOA 'K0A ESF Yes Yes leakage Class (t*) (b) 1mcation 1 0
'lype C laok Test Yea (c)(t) Yes(c)(t)
Valve Type Globe Gkbe Operator Pneum Pneum I% Actuation N2 to open N2 to open N2 and/or Spnrig to (kme N2 and/or Spnog to ckse Sec. Actuation N/A N/A Nonnal Position Open Open Shutdown Position Shut Shut Post Acc Position Shut Shut Pwr l'all Position Shut Shut Cont.150. Sig. ' C, D, E. P, II, N, Bil, RM C, D, E, F,1I, N, Im, ItNI Closure Time (sec) M3 M3 Pwr Source (Dh) 1 /11 1/11 O l 6 2-50,15 Amendment 17
ABWR nAcixAD nn. c Standard Plant TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION NUCU:AH llOILER SYSTEh! (Continued) MAIN STEAM LINE DRAINS Valve No. 1121 IVil 1121 1 012
$$AR lig $.1 k $1k Appikable Buts GDC$$ ODC 55 Iloid Sicam/ Water Steam / Water lineSlae 80A IGA ESF Yes Yes lamLage Class (b) (b) 1 mention 1 0 Type C isak Test Yes(e)(t) Yes(t)
Valet pe 3 Oste Gate Operator Motor Motor PrL Actuation Dec. Dec. See, Artvation Manual Manual Normal Position Open Open Shutdown Position Shut Shut Poht Act Position Shut $ hut Pwr Fall Position As is As is Cont. Iso. Sig. ' C, D, F, F, C,D,F,F, 11,N,Bil, 11,N.IID, RM RM Closure Time (wc) 15 15 Pwr Source (Div) Il I O Amendment 17 6.23016
ABWR 8 8^'8 "^n Siendard Plant Ret C TABLE 6.2 7 (Continued) CONTAINhtENT ISOLATION VALVE INFORhtATION DELETED l P 1 L i O i i 2 e i t i O. . 5 Amendment 14 6.2 50.17
. . , , - _ .-,-.. _ ,..,_..,_.._ ....- - . .. _ _ _ _ .. _-_.- _ a ,.-..-.__ _ u . _ _ _ _ _ _ _ ._ ~ -,..- - -._-.- ._.._.,_ . . _ ._ __ _
ABWR au6mo Standard Plant nev c TABLE 6.2 7 (Continued) CONTAINh1ENT ISOLATION VALVE INFORh1ATION NUCLEAR llOILER SYSTEM (Continued) FEEDWATER LINE A AND B Yalie No. D21.IW4A/B B21-lWM/D SSAR rig 5.1 M 5.1 M Applicable flasts GDC 55 GDC$5 Iluid Water Water
!Jne She 550A 550A 1:5r Yes Yes 14akage Class (b) (b) is ation 1 0 Type C !sak Test Yes(t) Yes(t)
Valve Type Check Sprmg Check Operator N/A Pne um. PrL Actuation Self Air to open Sec. Actuation N/A N/A Normal Position Open Open Shutdeva Position Shut Shut Post Ace Position Shut Shut Pwr Fall Position N/A N/A Cont Iso. Sig. N/A N/A Closure Time (ser) inst Inst Pwr Source (Div) N/A N/A O Amendment 17 6.2 50.18
2 ABWR nA61 MAD
- Standard Plant Rev.C t
('y TABLE 6.2 7 (Continued) , v-CONTAINMEF l'1%'LATION VALVE INFORMATION .- NUCLER SOILER SYSTEM (Continued) INSTRUMENT LINES Valve No. Various SSAR Fig 5.13b e,f4,h
. Applicable Basis RG 1.11 Cold Alt / Water IJae Size 20A ESF No(m) leakage Class (a)
Imcation .O Type C leak Teu No(m) - l Vahe Type Excess its Check I.O
~
Operator N/A-Prt Actuation Self Sec. Actuation - N/A Normal Position Open Shutdown Postalon Open Post Ace Position Open Pwr Fall Pos!alon Open . Cont. Iso. Sig. # N/A Closun Time (sec) N/A Pwr Source (Div) N/A Amendment 17 6.2-50.19
ABWR 23A61 MAD - Standard Plant nev e TABLE 6.2 7 (Continued) g CONTAINMENT ISOIATION VALVE INFORMATION REACIOR CORE ISOLATION COOLING SYSTLnl S'IT.AM SUPPLY Valve No. D l F035 E51-lbtB E51 IR'46 SSAR Fig 5.44b 5.4-8b 5.4-8b Applicable Basis ODC 55 GDC 55 ODC55 11uld Steam Steam Steam 1.ina She 150A 25A 150A FSF Yes Yes "es leakageclass (a) (a) (=? Iscation I I O _ Type C leak Test Yes(c)(t) Yes(c)(ts Yes(t) Valve Type Gate Globe Gate Operator Motot Motor Motor 1% Actuation Dec, Dec. Dec-Sec. Actuation Remote Manual Remote Manual Remote Manual Normal Position Open Shut Open Shutdown Position Shut Shut Shut Post Ace Position Shut Shut Shut Pwr Fall Position As is As is As is Cont. Iso. Sig. # S, T. RM.Z S. T, RM.Z S,T RM,Z Closure Time (see) <30 < 30 < 30 Pwr Source (Div) 1 I !! O Amendment 17 6.2-50.20
i 23A6100AD Standard Plant Rev C
- TABLE 6.2 7 (Continued)
CONTAINMENT ISOLATION VALVE INFORMATION - REACTOR CORE ISOIA110N COOLING SYSTEh! (Continued) hilN. FLOW AND TEST RETURN Valve No. P31.F011 E511009 SSAR flg . 3.4-Ba _ 5.44a Applicable Itasis ODC 56 ODC56 Iluid Water Water une She SOA 100A ESF . Yes _ Yes
- Isakage Class - (a) (a)
Iscation O O Type C Leak Test - No(h) No(h) Vale Type . Olot>e Glot< l- .O -
, Operator Motor Motor PrL Actuation Elec. Dec, Sec. Actuation Remote Manual Remote Manual NonnalPostilon - Shut Shut I
Shutdown Position . Shut Shut I Post Ace Position Shut Shut Pws Fall Position As is As is Cont. Iso. Sig. *) RM RM Closure Time (sec) <5 < 60 i l Pwr Source (Div) I ,. I I l-l l -..
~.
l .. Amcodment 17 6.2-50.21
.-m.. ,, , _ . _ . _ _ .._..
ABWR nA6iman Standard Plant nev c TAllLE 6.2 7 (Continued) - 1 CONTAINM ENT ISOIATION VALVE INFORMATION REACTOR CORE ISOLATION COOLING SYSTEM (Continued) l S/P SUCTION ! 1 Valve No, IU1-IV)6 1 SMR Fig 5.4-8a , 1 Applicable Itasis GDC 56 Fluid Water I 1.ine Size 200A j ESF Yes I leakage Class (a) location O Type C leak Test No(i) l Valve Type Gate Operator Motor Prt Actuation l'Jec. Sec. Actuation Remote Manual Normal Position Shut Shutdown Position Shut Post Ace Position Shut Pwr Fall Position As is Cont. Isa. Sig. RM Closure Time (sec) < 30 Pwr Source (Div) 1 O Amendment 17 6.2 50.22
e
- ABWR- nA6iwAn Standard. Plant nev.c
-O V=
TABLE 6.2 7 (Continued)
~ CONTAINMENT ISOIATION VALVE INFORMATION-
- REACTOR CORE ISO!ATION COOLING SYSTEh! (Continued)
'IURHINE EXHAUST
. Valve No. 1!$11939 0511938
-SSARflg 5.44a $A4s AppliceMe Basti - ODC 56 -- GDC56 <
Fluid Steam Steam IJae $13e ' 3504 350A - ESF Yes Yes
.14akage Class . (a) - (a)
- 1 mention O O
. - Type C leak Test Yes(c)(t) Yes(t)
Velve Type _ Gate . Check Operator Motor . Self Actuating PrL Actuation Elec, N/A-
-- Sec. Actuation Macual - N/A
; Normal Position - Im ked 0;wn Shut Shutdown Position Open . Open Post Act Postalon Shut Shut
' Pwr Fall Position As is N/A Cont. Iso. Sig. #} RM N/A Closure Time (see) < 70 Inst.
Pwr Soune (Div) - I N/A O Amendmen* 1/ 6.2-50.23
ABWR 23AciwAn ney. c Standard Plant TABLE 6.2 7 (Continued) g CONTAINMENT ISOLATION VALVE INFORMATION REACTOR CORE ISOLATION COOLING SYSTEh! (Continut d) VACUUh1 PUhlP DISCllARGE Valve No. E51-IN7 E51 IN6 SSAR Fig 5.4-8a 5.44a Apptkable Basis GDC 56 GDC 56 11uld Steam Steam IJne Size 50A 50A ESF Yes Yes 14akage Class (a) (a) location O O Type C laakTest No(l) No(l) Valve Type Gate Check Operator Motor Self. Actuating PrL Actuation Dec. N/A Sec. Actuation Manual N/A Nortna4 Position Locked Open Shut Shutdown Position Open Open Post Act Position Shut Shut Pwr Fall Position As is N/A Cont.150. Sl:-(# RM N/A Closure Time (sec) <10 Inst. Pwr Source (Div) I N/A O 6.2-50.24 Amendment 17
.. 23A6100AH L Standard Plant - new c - ,
l
- y. 3 TABLE 6.2-7 (Continued)
-( ; i A CONTAINMENT ISOLATION VALVE INFORMATION '
ATMOSPHERIC CONTROL SYSTEM Valn No. - DIfM1 T31-IW2 T31 IV03 DIlWI T31@05 DlIt06 SSAR flg 6.2-39a 6139a 6.2 39a 6.2-39a 6.2-39a 6.2 39a Applicable Basis ' ODC56 ODC 56 ODC56 ODC 56 ODC56 ODC56 l 11ald Air Airor N2 Air or N2 DW AntOS DW AntOS WW ATMOS j Ilne She (nun). 550A 550A $50A 550A SOA 550A j ESF Yes Yes Yes Yes Yes Yes leakage Class (b) (b) (b) (b) (b) (b) laation 0 O O O O O
' Type C lenk Test. Yes Yes(e) Yes(e) Yes(e) Yes(e) Yes(e)
' Valve Type Dutterfly Dutterfly Dutterfly Dutterfly Globe Butterfly
- Operator Pneum Pneum Pneum Pneum Pneum Pneum
.\
PrL Actuation Air Air Air Air Air Air See. Actuation N/A N/A N/A N/A N/A N/A ., Norinal Position Shut Shut Shut Shut Shut Shut Shutdown Position - Shut Shut. Shut Shut Shut Shut Post Ace Position ' Shut Shut Shut Shut Shut- Shut Pwr Fall Poslalon Shut Shut Shut Shut Shut Shut Cont. Iso. Sig.(* A,K ' A,K AK A,K A,K A,K Closurv Time (uc) .< 30 < 30 < 30 < 30 < 15 < 30 Pwr Sou'rce (Div) I II 11 !! 11 11 O Amendment 17 6.2-50.25
ABWR 23A64 *AB Standard Plant new c TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION O ATMOSPIIERIC CONTROL SYSTEM (Continued) Valve No. D1-F008 T31 F0W T31-P025 D1-F039 T31-IY>to T31-IT>il D1-F720A/B SSAR Fig 6.2-39a 6.2-39a 6.2-39a 6.2-39a 6.2-39a 6.2 39a 6.2 -39b Appliceble Basis ODC56 GDC 56 ODC 56 ODC56 ODC56 GDC56 GDC 57 Fluid PCV Ah!OS PCV Ah!OS N2 N2 N2 N2 N2 IJne Size 250A 550A 400A SOA 50A SOA 20A ESF Yes Yes Yes Yes Yes Yes No Leakage Class (b) (b) (b) (b) (b) (b) (b) 14 cation O O O O O O O T}pe C 14ak Test Yes Yes Yes Yes Yes(e) Yes(e) Yes Valve Type Butter 0y Butterfly Butter 0y Globe Globe Globe Gate Operator Pneum Pneum Pneum Pneum Pneum Pneum Soleniod PrL Actuation Air Air Air . Air Air Air Elec. Sec. Actuation - - . - - - - Normal Position Shut Shut Shut Open Open Open Shut Shutdown Position Shut Shut Shut Shut Shut Shut Shut Post Ace Position Shut Shut Shut Shut Shut Shut Shut Par Fall Position Shut Shut Shut Shut Shut Shut Shut Cont. Iso. Sig. A,K A.K A.K A.K A,K A.K A,K Closure Time (see) <30 < 30 < 30 < 15 < 15 < 15 <5 Pwr Source (Div) 1 I I I 11 11 II O Amendment 17 6.2-50.26
TABWR - 23A6mo
- Standard Plant am c TABLE 6.2-7 (Continued) -Q -
CONTAINMENT ISOLATION VALVE INFORMATION ATMOSPHERIC CONTROL SYSTEM (Continued) Valve No. T311730 T13-I732A/B T13 F734A-D SSAR Fig 6.2 39b 6.2-39b 6.2-39b
.A ODC$6 ODC56 ODC 56 l pplicable Basis RO 1.11 RO 1.11 RO 1.11 neld DW A1MOS DW ATMOS DW A1NOS tlne Size 20A 20A 20A ESF No No No
' 14akase Class (a) (a) (a) location O O O Type C leak Test No(m) No(m) No(m)
Valve Type Globe Globe Globe Operator N/A N/A N/A
%.))
Prl. Actuation Manual Manual Manual Sec. Actuation N/A N/A N/A-Normal Position Open Open Open
- Shutdown Postalon Open - Open - Open Post Acc Position Ogn Open Open Pwr Fall Position N/A N/A N/A Cont. lso. Sig. N/A N/A N/A Closure Time (sec) N/A N/A N/A Pwr Source (Div) N/A N/A N/A LO i
l Amendment 17 6.2-50 27
b\ 23A61C0AF) Standard Plant nev C TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION ATMOSPIIERIC CONTROL SYSTEM (Continued) Valve No. T311736A/B T31 F738A-D T31-F740A D T31 F742A/B ssut Fig 6 2 39b 6.2-39b 6.2-39b 6.2-39b Applicable Basis GDC 56 GDC 56 GDC 56 GDC 56 RO 1.11 RO !.11 RO 1.11 RO 1.11 Fluid WW A1NOS WW A1NOS SP11 O WW ATMOS Ilne She 20A 20A 20A 20A ESF No No No No 14akage Class (a) (a) (a) (a) tecation O O O O Type C leak Test No(m) No(m) No(m) No(m) Valve Type Globe Globe Globe Globe Operator Manual Manual Manual Manual PrL Attuation N/A N/A N/A N/A Sec. Actuation N/A N/A N/A N/A Normal Position - Opn Open Open Open Shutdown Position Open Oper Ogn Open Post Ace Position Open Or< 4 Open Open Pwr Fall Position N/A .N., A N/A N/A Cont.150. Sig- N/A N/A N/A N/A Closure Time (see) N/A N/A N/A N/A Pwr Source (Div) N/A N/A N/A N/A O Amendment 17 6.2-50.::8
_ _ _ _ _ _ _ . _ . . _ ._ .. ._ _ _.___ m -
~
23A6100All Standard Plant ney. c 1 es TABLE 6.2 7 (Continued)
--U CONTAINMENT ISOLATION VALVE INFORMATION ATMOSPIIERIC CONTROL SYSTEM (Continued) -
.i Velve No. - D11744A/B D11300A/B TJ1-IS02A/B i_
SSAR Fig 6.2 39b 6.2 39b 6.2-39b Applicable Basis ODC57 ODC56 ODC56 RO 1.11 RG 1.11. RO 1.11 fluid SP11 O DWAlhiOS DW ATMOS ' Ilne Size 20A 20A 20A ESF No No No 1 l: Isakage Class (b) (b) (b) l Imation O O O i Type C 1mak Test No(m) . No(m) No(m) Valve Type Globe Globe Globe O-
. Operator Manual Manual Manual Prt Actuation .N/A N/A N/A l Sec. Actuation N/A N/A N/A Normal Position Open Open Open Shutdown Position . Open Open Open Post Ace Position Open Open Open Pwr Fall Position - N/A N/A N/A Cont. Iso. Sig.(#} N/A N/A N/A Closure Time (see) N/A N/A N/A l
Pwr Source (Div) N/A N/A N/A O Amendment 17 6.2-50.29 l
ABWR uA6 MAD Standard Plant _ Rev C TABLE 6.2 7 (Continued) g CONTA.INMENT ISOIATION VALVE INFORMATION ATMOSPilERIC CONTROL SYSTEM (Continued) . Yalve No. T314WIA/D D1.D001 Dl-IE S M R Fig 6.2-39b 6.2-39 6.2-39a Applicable Basis GDC56 GDC 56 GDC 56 RG 1.11 nuld WW A*D.tOS WW ATMOS DW ATMOS IJae She 20A 350A 350A ESF No Yes Yes Isakage Class (a) N/A N/A Location O O O Type C leak Test No(m) No(o) No(o) Vahe Type Globe Rupture Duk Rupture Duk Operator Manual Self Self Prt. Actuation N/A N/A N/A Sec. Actuation N/A N/A N/A Normal Position Open Shut Shut Shutdown Position Oper Shut Shut Post Ace Position Open Open Open Pwr Fail Poshion N/A N/A N/A
- Cont. lso. Sig. N/A N/A N/A Closure Time (sec) N/A N/A N/A Pwr Source (Div) N/A N/A N/A i
i e Amendment 17 6.2 50.30
. 23A6100AD Standard Plant nev.c f'\ .
Q TABLE 6.2 7 (Continued) . CONTAINMENT ISOLATION VALVE INFORMATION
.. FLAMMAHILi1T CONTROL SYS'Iul Valve No. T40IV01A T49-1001D T49-IV02A T49-1002D SSAR Hg 6.240 6.2-40 6.240 6.240 Applicable Basis ODC56 GDC56 GDC56 GDC%
Fluid DW NIMOS DW ATMOS DWATMOS DW NIMOS Line $be 100A 100A 100A 100A FSF Yes Yes Yes Yes Leakage Class (a) (a) (a) (a)
-p R 1mcation O O O O w
Type C 14ak Test No(u) No(u) No(u) No(u).
.. Valve Type Gate Gate Gate Gate Operator Motor Motor Motor Motor l PrL Actuation Dec. Dec. Sec. Dec,
{: Sec. Actuation Manual Manual Manual - Manual Normal Position Shut Shut Shut . Shut Shutdown Position Shut Shut Shut Shut Post Acc Position Open Open Open Open Ivr Fall Position As is As is As is As is l L Cont, Iso. Sig (#} A,K A,K A.K A,K II
. Closure Time (sec) < 30 < 30 < 30 < 30 Fwr Source (Div) 1 11 I !!
oO
- Amendment 17 6.2 50.33
ABWR 2usiwAD Standard Plant nev c TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION FIAMMAlllLTIT CONTROL SYSTD1 (Continued) Valve No. T49 IUMA T49-IUtil T49-IW7A T49IW7D SMR Fig 6.240 6.240 6.240 6.240 Applicable Basis GDC 56 GDC 56 GDC 56 GDC 56 Fluid WW ATMOS WW NINOS %T ATMOS WW ATMOS IJne Size 150A 150A 150A 150A e ESF Yes Yes Yes Yes 12akage Class (a) (a) (a) (a) g location O O O O w Type C 1.eak Test No(u) No(u) No(u) No(u) ; l Valve Type Gate Gate Gate Gate Operator Motor Motor Motor Motor PrL Actuation Dec. Dec. Dec. Dec. Sec. Actuation Manual Manual Manual Manual Nonnal Position Shut Shut shat Shut Shutdown Position Shut Shut Shut Shut Post Ace Position Open Open Open Open IMr Fall Position As is As is As is As is Cont. lso. Sig. A,K A.K A,K A,K Closure Time (sec) < 30 < 30 < 30 < 30 Pwr Source (Div) I II I !! O Amendment 17 6.25034
AllWR u w ,0an ! StIl11dIttd I'lant _ __ _ ac i n t l TAllt.E 6.2 7 (Continued) CONTAINMENT ISOLATION VAINE IN170ltMATION ItEACTOlt WATI:lt CLEANUP SYSTEM Valve No. 031.im2 O tl.Iw3 G11.I'017 SSAR l'ig 5 4-12a 3412a 5 4 12a l Appfwable Dasts GDC 55 GDC$$ GDC 55 11uid RPV 1120 RPV l(20 RPV l120 IJne Sue 2fXiA 200A 150A IWF Yes(e)(t) Yes(t) Yestt) l 1:akage Class (a) (a) (a) location 1 O O Type C teak Test Yes(e)(t) Yes(t) Yes(t) l Valve Type Gate Gate Gaie
/] %)
Operator Motor Motor Motor Pri. Actuation lilec. Elec. F.lec. See, Actuation Manual Manual Manual Normal Position Open Open Shut Shutdown Position Open Open Shut Post Acc Position Shut Shut Shut Pwrl7 ail Position Asis As is As is Cont. Iso. Sig! A,",V,Z,AA A.F,VJ,CC,AA A.F,V2,CC.AA Closure Time (sec) < 30 < 30 < 30 Iw Source (Div) 11 ! I (\ r Amendment 17 (12 A015
ABWR 23A61coAn Standard Plant Rev C TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION REACTOR WATER CLEANUP SYSTEM (Continued) Valve No. G31 f018 G31 IV71 G311972 SMR Fig 5.412a 5.412a 5.4-12a Applicable Basis GDC55 GDC55 GDC55 Fluid RPV !!20 RPV I(20 RPV !!20 1.ine Sise 150A 20A 20A ESF Yes No No L4akage Class (a) (a) (a) Location i I O l Type C leak Test Yes(t) Yes(c)(t) Yes(t)
)
Valve Type Chect Globe Globe Operator Self No AO PrL Actuation N/A Dee Dec Sec. Actuation N/A N/A N/A Normal Position Shut Open Open Shutdown Position - Open Shut Shut Post Ace Position Shut Open Open l Pwr Fall Positio . Shut
-N/A Shut Cont. Isa. Sig (# N/A N/A N/A l Closure Time (see) Inst N/A N/A Pwr Source (Div) N/A N/A N/A O
Amendmeat 17 6.25036
~ - - . -. .. .- _ . . . . - .
ABWR . 2auwAn-Standard Plant nev. C O V TABLE 6.2 7 (Continued)
. CONTAINMENT ISOLATION VALVE INFORMATION REACTOR WATER CLEANUP SYSTEM (Continued)
- Valve No. 031l%QA/B 031ITIA/D 031IT2A/11031ITM/I)
SSARI1g 5.4-12e 5.4-12a 5.412a 5.4-12e
.. Applicalile Basis ODC55 ODC 55 ODC55 ODC 55 RO 1.11 RO 1.11 - RO 1.11 RG 1.11 Hund RTV i120 RTV i120 RPV 1820 RPV 1120 line Stre 20A 204 2M 2M ESF No No No No tanka p Class (s) (a) (a) (s) 14eation O O O O Type C leak Test No(m) No(m) No(m) No(m)
Valve Type Globe ' Olobe XS Check XS Check Operator Manual Manual Self Scif PrL Actuation N/A N/A N/A N/A Sec. Actuation .N/A N/A N/A- N/A Normal Position Open Open - Open - Open Shutdown Position - Open . Open Open Open Post Arc Position Shut ' Shut ' Shut Shut Pwr Fall Position N/A N/A N/A N/A Cont. lso, Sig. NAf N/A N/A N/A Closure Time (m) N/A 'N/A N/A N/A rwr Source (Div' N/A N/A N/A N/A O 6.2-50 37 Amendment 17
. , . . ~
AllWR 3346an ' Standard Plant nev e TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION DELETED O O Amendment 14 6.2 50.38
LABWR naaman Standard Plant nevs l') TABLE 6.2 7 (Continued)
'v CONTAINMENT ISOLATION VALVE INFORMATION SUPPRESSION POOL CLEANUP SYSTEM Vahe No. - G511001 G51-F002 G511006 G51 IV07 SSAR Fig 951 951 951 751 Apputable Basis ODC 57 GDC57 GDC57 ODC57 Fluid Water Water Water Wales line Size 200A 200A 250A 250A ESF Yes Yes Yes Yes Lankage Class (a) (a) (a) (a) location o- o o o Type C leak Test No(p) No(p) No(q) No(q)
Vahe Type Gate Gate Check Gate
, i \j Operator Motor Motor N/A Motor PrL Actuation Dec. Dec. N/A Dec.
Sec. Actuation Manual Manual N/A Manual Nonnal Position Shut Shut Open Shut
- Shutdown Position Open Open Oren Open
. Post Acc Position Shut Shut N/A Shut Pwr Fau Podtnon As is As is N/A As is Cont. lso. Sig.
- A.K A.K N/A A,K Closure Ti.ne (sec) < 30 < 30 Inst. < 30 Pwr Source (Div) 11 I N/A 11 v
6.2 50.39 Amendment 17
ABWR meiwAn Standard Plant nev c TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION REACTOR BUILDING COOLING WATER SYSTEM Valve No. P21lW5A P21 lM1A P21 IW511 P21-l%1D
/lW6A /IV80A /FD76B /luoll SSAR Fig 9.2-le 9.2-1c 9.2-1f 9.2-1f Appikable Basis GDC55 GDC 55 GDC 55 GDC55 Iluid Water Water Water Water ilne She (nun) 200A 200A 200A 200A ESF Yes Yes Yes Yes leakage Class (b) (b) (b) (b) tecation O/1 O/: O/1 O/l Type C leak Test No(s) No(s) No(s) No(s)
Valve Type Gate / Check Gate / Gate Gate / Check Gate / Gate Operator Motor /NA Motor / Motor Motor /NA Motor / Motor Pri. Actuation Ecct. Dect. Dect. Dect. Sec. Actuation IIW/N/A IIW/N/A IIW/N/A IIW/N/A Normal Position Open Open Open Open Shutdown Position Open Open Open Open Post Acc Positha Shut Shut Shut Shut Pwr Fall Position As is As is *o is As is Cont. Iso. Sig. CX,K CX,K CX,K CX,K Closure Time (see) 80/N/A 80/80 80/N/A 80/Su Pwr Source (Div) 1/N/A I/II I/N/A I/11 O Amendment 17 6.2-50 40
ABWR 2 miman Standard Plant- new c TV
& TABLE 6.2 7 (Continued)
' CONTAINMENT ISO 1ATION VALVE INFORMATION HVAC NORMAL LI>JLING WATER SYS1Di Valve No.- P24-1013 P24.IV54 P24.IV142 P2110141
$$AR 11e 9.2-2b 912b 9.2 2b 9.2 2b Applicable Bants ODC$5 ODC $3 ODC $$ ODC SS
]
}1mid Water Water Water Water IJne She 100A 1004 100A 100A ESP No No No No
)
14aksee Clau (b)_ (b) (b) (b) l Iscation O 1 O I Type C leaklest Yes(i) Yes(t) Yes(t) Yes(t) i Valve Type Gate Check Oate Gate
- f. Operator . Motor N/A. Motor Motor PrL Actuation Decti N/A Ecc, Dec Sec. Actuation llW N/A IIW - N/A Normal Position Open Open ' Open Open Shutdown Position Open Open Open Open Post Act Position Shut .. N/A Shut Shut Pwr Fall Pontalon Asis N/A Asis As is Cont, isn. Sig.(#- CX,K N/A CX,K CX,K Clo66te l .me (sec) 25 ' Inst 25 25 Pwr Source (Div) i N/A -1 !!
g Amendment 17 6.2 50AI I L _ , . - _ . .
ABWR 23A6imAn Standard Plant new c TABLE 6.2 7 (Continued) h CONTAINMENT ISOLATION VAI VE INFORMATION SERVICE AIR SYSTEM Valve No. P511731 P51-F132 SMR 113 917 9S7 Applicable 11 ails - 35 GDC-55 Iloid Alt Air IJne Size 25A 25A l'Sr No No leakage Class (b) (b) location O I Type C leak Test Yes Yes Valve Type Globe Check Operator Manual N/A Prl. Artuation Manual N/A Sec. Actuation N/A N/A Normal Position Shut N/A Shutdows Position Open N/A Post Acc Postalon Shut N/A Pwr Fau Pusieion N/A N/A Cont. Iso. Sig. N/A N/A Closure Time (sec) N/A N/A Pwr Source (Div) N/A N/A O 6.2-50.42 Amendment 17
ABWRs - uAnooAn - Standard Plant nm c { TABLE 6.2-7 (Continued) - CONTAINMENT ISOLATION VALVE INFORMATION INSTRUMENT AlR SYSTEM valve No. - PS24276 PS2-1777 SSAR11g 9M 9.M Applicable Basis GDC55 ODC 55 j l fluid Air Air une size soA SOA ESP - Na No l. tankage Class (b) (b) location - -O I l Type C leak Test Yes Yes f Valvelype Globe Check n
- - Operator Motor N/A Pat Actuation PJcct. N/A Sec. Actuation - IIW N/A-Normal Position Open - Opco Shutdown Position Open Open Post he Position Open Open Pwr Fall Position - As is N/A Cont. Iso. Sig.(# RM N/A Closure Time (sec) 20 -N/A Pwr Source (Div) i N/A L) 6.2-50.43 Amendment 17
ABWR DA6100AD Standard Plant Rev. C TABLE 6.2 7 (Continued) g CONTAINMENT ISOLATION VALVE INFORMATION lilGli PRESSURE NITROGEN GAS SUPPLY SYSTEM Valve No. P54-IT07A/F008A P54-IV07D/1906B P54.f700/F209 SSAR Fig 6.7-1 6.71 6.71 Applicable Basis GDC55 GDC 55 GDC55 Huld N2 N2 N2 line Size (mm) SOA SOA 50A ESF Yes Yes Yes leakage Class (b) (b) (b) Imcation O/l O/l O/I Type C Leak Test No(r) No(r) Yes Valve Type Globe / Check Globe /Cheet Globc/ Check Operator Motor /N/A Motor /N/A Motor /N/A PrL Actuation Elect /N/A Elect./N/A bett!N/A Sec. Actuation !!W/N/A liW/N/A IIW/N/A Normal Position Open Open Open Shutdown Position Open Open Open Post Acc Position Shnt Shut Shut Pwr Fall Posit n As is/N/A As Is/N/A As Is/N/A Cont. Isa. Sig. G.G. G.G. G.G. Closure Time (sec) 30/N/A 30/N/A 30/N/A Pwr Source (Div) II/N/A I/N/A I/N/A O Amendment 17 6.2 50.4%
ABWR u^am^u ne< C Standard Plant TAllLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION MAKEUP WATER SYSTEM (PURIFIED) Valve No. P11-F141 P11J'142 SSAR 1% 9.2 5b 9.2-5b Applicable Basis GDC 55 ODC 55 IW4 Water Wate r line She 504 50A ESF No No Iestage Class (b) (b) Imcation O I Type C taskTest Yes Yes Valve Type Globe Check Operator Manual Scil Prl. Actuation N/A N/A Sec. Actuation N/A N/A Normal Position Shut Shut Shutden Position Open Open Post Acc Position Shut Shut Pwr Fall Position N/A N/A Cont. Iso. Sig. N/A N/A Closure Time (sec) N/A N/A Pwr Source (Div) N/A N/A O 6.2 50A5 Amendaient 17
ABWR 23Aucun Standard Plant an. C TABLE 6.2 7 (Continued) CONTAINMENT ISOIATION VALVE INFORMATION IIAK DETECTION & ISOLATION SYSTEM Valve No. E31-IO02 E31 TOO3 E31 R04 E311905 E31-IV)9/ E31 I'701/ E31-I%2/ I'010 I"703 IN A/B/C/D A/D/C/D SMR Fig 5.241 5.24: 5.24i 5.24i 5.24h 5.24r 5.24f Applicable Basis GDC55 GDC 55 GDC55 GDC55 GDC 55 RG 1.11 RG 1.11 Fluid Air Air Air Air Water Steam Steam Line She 32A 32A 32A 32A 20A 20A 20A ESF Yes Yes Yes Yes Yes No No leakage Class (a) (a) (a) (a) (a) (a) (a) Imation O O O O O O O Type C teak Test Yes(e) Yes(c) Yes(c) Yes(c) Yes(e)(t) No(m) No(m) Valve Type Globe Globe Globe Globe Globe Globc/Dellows Exceu Flow Seal Check Operator Pneum Pneum Pneum Pneum N/A Man. N/A PrL Actuation Air Air Air Air Manual N/A Self Sec. Actuation N/A N/A N/A N/A N/A N/A N/A Normal Position Open Open Open Open Shut Open Open Shutdown Position Shut Shut Shut Shut Shut Open Open Post Ace Position Shut Shut Shut Shut Shut Open Open Pwr Fa11 Position Shut Shut Shut Shut N/A N/A Open Cont. Isa. Sig. B.K B.K B,K B,K N/A N/A N/A Closure Time (see) < 15 < 15 < 15 < 15 N/A N/A N/A Pwr Source (Div) I !! !! I N/A N/A N/A i l O l l 1 A'nendment 17 6.2-50A6 I l
-ABWR 2a iOOAn Standard Plant ___
ney. c TABLE 6.2 7 (Continued) CONTAINMENT ISOLATION VALVE INFORMATION RADWASTE SYSTEM Valve No. K17.!W3 Kl7.lW4 K17-F103 K17-F104 S M R Fig - 1112cc 11.2 2cc 11.2-2ee 11.2-2ee i i Applicable Basis ' ODC 57 GDC57 GDC $7 GDCS7 nund IIW 1120 LCW II20 LI W 1120 LCW 1120 IJne Sise 65A - 65A - 65A 65A ESF No No No No I leakage CNs (b) (b) (b) (b) l
- location 1 O I O i I
*I)pe C Leak Test No(v) No(v) No(v) No(vi
- Valve Type Gate Gate Gate Gate Motor - Motor Motor Motor
_. Operator
- PrL Actuation Dec. Gee. Dec. Dec.
Sec. Actuation Manual Manual Manual Manual Nonnal14sition - Open . Open - Open Open
- Shutdown Position Open Open Open Open Post Ace Position Shur Shui Shut Shut Pwr Fall Position AS is Asis Asis Asis.
Cont. Iso. Sig. A/IT FF A/FP IT' Closure Thne (sec) <30 < 30 < 30 < 30
. Ivr Source (Div) 11 1 11 I s
l l Amendment 17 6.2-50A7 l l w gy- ---y- g-e rr e y ,pn...w,,.e.w,ew w,y-w ,-er'-r -+ m-Wfey ' - -M-g m,-t-- - m4 t'&-= r-e-+ " -* r -M
ABWR 2mimAn Standard Plant Rev C TABLE 6.2 7 (Continued) NOTES (a) Termination Region: Secondary Containment (b) Termination Regica: Main Condenser, Turbine Bldg., Bypass leakage Harrier: Redundant Primary Cont. Iso. Vues. (c) Isolation Signal Codes Signal Description A Reactor wael low water level . level 3 B Reactor wssel kw water level - level 2. C Reactor wuct low nier level level 1.5. CX Reanor vessel km uter level - lewi 1. D liigh radiation - main steamline. E Line break main steamline (steamline high steam Dom). F Une break main steamhne (steamhne high tunnel temperatuie). 11 Line break - main steamhne (steamhne high tuttiine building temperature. l J Turbine building high temperature. I K !!igh drywell pressure. L RilR injection valve kw pressure. M Line break in R11R shutdows N tow main condenar vacuum. i ! 5 Line break in RCIC system steamline to turbine (Iow steamhne pressure). l T liign pressure RCIC turbine exhaust diaphragm. V Clue through electncalinterlocks with other n!ves or pump motors. l l W Condensate storage tank low level. 1 X Suppression pool low level. Y RCIC. l Z Fquipment area temp high Amendment 9 6.2-50 48
ABWR n m man Rn c Standarmant TAllLE 6.2 7 (Cttntinued) NOTT.S (Continued) (c) Isolation Signal Codes (continued) Signal Ikscription AA Dtflertntitl mass ikw high
!!D tow rnain steamhne preuvre 6t inlet to tutt>ine (RtJN mWe only).
CC t ine break in reactor water cicanup system (high space tempersture). DD Containment pressure. IIE liigh diffe rential firm in the reactot water cleanup sptem. IT Itigh Radiation l'rocess Line RM Remote manual switch from control room (All automatic iniunted isolanon valves are capable of temote manual operation from the control room), 00 tow Nitrogen pressure. (d) This line is filled with water and pressurind higher than 1107o of the post accident peak containment pressurc, l_ ins is small and postulated failure is considered less ievere than instr ument line. (c) Leakage testing may be performed in the reverse direction in the absence of test connections and/or isolable :est boundaries in the upsticam side of the valve relative to the leakage flow direction (i e. from inside to outside primary containment). The results are conservative or equivalent to the normal direction as described below: (1) For globe valves including MSIV's, testing in the reverse direction is conservative since the test pressure tends to lift the plug from the seat. (2) For gate valves and butterfly valves, leakage characteristics for these types of vahrs are similar in both diredions provided seat construction is designed for scaling on either side. J (f) These sinct are CAM system sample lines that continuously monitor (sample) post accident containment atmosphere. These lines are safety grade closed loop extension of the primary containment. Sampled gases (or leakage if any) are returned to the primary containment. In addition, these lines are subject to periodie Type-A test whose leaktight integrity can be verified. (g) Tbc RilR drywell and wetwell spray lines are always filled with water in the outboard side thereby prodding water seal. The sealis maintained at a pressure higher than 1107o of the post accident peak containment pressure by jockey pumps and/or hydroctatic head; thus precluding leakage.
\.
6MD *19 Amendment 17
j ABWR 2mma hc i 81HHd3Ed.PhlDt TAllLE 6.2 7 (Continued) g NO'ITS (Continued) Fur 6ermore, these valves are required to open post LOCA to provide containment cooling function. When this function is activated, now direction is towards the containment. (h) The ECCS (RllR,llPCF and RCIC) test return and minimum flow lines *ctminate below the suppression pool water level and are sealed from the containment atmosphere by the suppression pool water. The outboard side of the valve (away from the containment)is always filled with water and pressurized higher than 110% of the post accident peak containment prettsure as in (g) above. (i) 'The ECCS (RilR,llPCF and RCIC) suction lines are always filled with water since the suction lines are located below the suppression pool water level and are sealed from the containment atmosphere. (j) The RilR suppression pool cooling discharge line is the same line used for system flow or pump ' low testing. See (h) above. (L) The ECCS (RilR,llPCF and RCIC) injection lines are always filled with water up to the outboard isolation valves thereby forming a water seal. These water seals are kept pressurized higher than 110% of the post accident containment pressure as in (g) above. Furthermore, these valves are subject to ASME Section XI IWX leak raic tests. (1) RCIC vacuum pump discharge line terminates below the suppression pool water level and is sealed from the containment atmosphere. (m) Instrument lines that penetrate the primary containment conform to Regulatory Guide 1.11. The lines that connect to the reactor pressure boundary include a restricting orifice inside containment, are Seismic Category 1, and terminate in Seismic Category 1 instruments. The instrument lines also include manual isolation valves and excess flow check valves or equivalent. These lines are normally open, and are considered an extension of the primar'; containment whose integrity is continuously demonstrated during normal operation. In addition, these lines are subject to periodic Type A tests. Leaktight integrity is also verified during functional and surveillance actisities as well as visual observations during operator tours. (n) The outboard side of the R11R shutdown cooling suction valves are sealed with water since RilR pump and suction lines are located below the suppression pool water level. This is a closed loop water seal since RilR is a closed lop ystem always filled with water. L'W Furthertnore, these valves are subject to ASME Section XIIMX leak rate tests. (o) Rupture discs are cormally closed and scaled from leakage. The opening setpoint of these rupture discs is higher than primary containment test pressures. Additionally, these rupture discs are subject to the Type-A test., (p) SPCU suction line is always filled with water since it is located below the suppression pool water level and is sealed from the containment atmosphere. (q) SPCU retut o line terminates below the suppression pool water level and is scaled from the containment atmosphere. O j l Arnendment 17 62 50 49a
AllWR nui=^n iter c Silitidtird Pittnt TAllLE 6.2 7 (Continued) NOTUS (Continued) (r) The outboard side of these valves is always pressurind with nitrogen gas at a pressure higher than the post accident peak containment pressure. The nitrogen supply in these lines is required for post accident mitigating function. (s) The outboard side of these valves is always filled with water and pressurind above 110% post accident peak containment pressure. These lines are kept charged with cooling water for cooling emergency equipment necessary for post accident mitigation. (t) Line will be drained and tested with air. (u) 11ammability Controlis a dosed loop, safety grade system required to be functional post accident. Whatever is leaking (if any)is returned to the primary containment, in addition, during !LitT, these valves are opened and the lines are subjected to Type A test. (v) These lines terminate below the drywell sumps water level and are scaled from the containment atmosphere. (w) The outboard side of these valves are provide with a water leg. in addition, these valves are sutsject to ASME Section XI IWV leak tests. O O 6 230 47b Amendment 17
ABWR ==u Standard Plant ill1_c Table 6.2 8 Primary Containment Penetration List 2 O Penetratics Name Elesattoa Aalmuth Offset Diameter Harrier Testingl3 l Numter (mm) (den) (mm) (mm) lype X1 U/D Equipment llatch 19170 130 0 2600 Door 11 X2 U/D Personnelllatch 19170 230 0 2400 Door 11 X3 - ISI llatch 1M100 221 0 200 Door 11 X-4 Wetwell Access llatch 6400 45 0 2000 Door 11 X5 L/D Personnelllatch -650 0 0 2400 Door B X-6 L/D Equipment llatch 900 180 0 2400 Door 11 X-10A Mainsteam Line 16300 0 1400 1100 A X 10B Mainsteam Line 16300 0 4200 1100 A X 10C Mainsteam Line 16300 0 4200- 1100 A X 10D Mainsteam Line 16300 0 1400 1100 A X 11 Mainsteam Drain 13650 0 5200 000 A X 12A Feedwater Line 13810 0 2800 950 A X 12B Feedwater Line 13810 0 2800 950 A X 22 Borated Water injection 15250 275 0 40 A X 30A Drywe!! Spray 14500 260 3400 200 A X-300 Drywell Spray 14500 100 3400 -200 A X 31A 11PCF (B) 14500 260 0 60'1 A X 31B llPCP (C) 14500 100 0 600 A X 32A LPFL (B) 14500 260 2000 650 A f X 32B LPFL (C) - 14500 100 1800 650 A g' X 13A RHR Suction (A) 14550 80 800 750 A X T3B R11R Suction (B) 14550 260 1800 750 A X 31C RilR Suction (C) 14550 100 2000 750 A X 37 RCIC Turbine Steam 14450 80 1200 550 A X 38 RPV licad Spray 14450 310 1500 550-- A X 50 - CUW Pump Feed 14480 310 0 600 A X-60 MUWP Suttion 13550 275 0 50 A X-61 RCW Suction (A) 13550 45 2000 200 A X-62 RCW Return (A) 13550 45 3000 200 A X RCW Suction (B) 13550 225 2400 200 .A
. X-64 RCW Return (B) 13550 225 M00 200 A X-65 IINCW Suction 13550 225 400 150 A X 66 IINCW Return 13550 225 1400 150 A X 69 SA 19000 42 0 25 A X 70- IA 9(00 46 0 50 A
. X 71A - ADS Accumulator (A) 15000- 50 0 50 A X 71B ADS Accumulator (B) 19000 296.5 1000 50 A X 72 Relief Valve Accumulator 19000 296.5 2000 50 A X 80 Drywell Purge Suction 13700 68 0 550 A X 81 Drywell Purge Exhaust 19000 216_ 0 550 A X 82 FCS Suction 20100 221 0 100 A X-90_ Spare 20100 46 0 400 A X 91 Spare 20100 296.5 1000 400 A X 92 Spare 16400 45 12700 400 A A
O- X 93 Spare 16400 300 0 400 Amendmeat 17 6.2 50.50-
ABWR nuioorn Sandard Plaul Riv c Table 6 2 8 Primary Containment Penetration List 2 (Continued) Penetration Name Flevation Ar.imuth Offset Dlataeter llarrier Testing1 3 O Number (mm) (deg) (mm) (mm) Type X 100A IP Power 13500 55 1100 450 0 ring B X 1008 IP Power 13500 180 2650 450 0 ring 11 X 100C IP Powcr 13500 180 -6550 450 0 ring Il X 100D IP Power 13500 280 0 450 0 ring B X 100E IP Power 13500 180 2650 450 0 ring Il X 101A LP Power 1900 45 0 300 0 ring Il X 10111 LP Power IM00 180 50 300 0 ring Il X 101C 1 P Power 1M00 180 1350 300 0-ring B X 101D FMCRD Power 19W0 235 1350 300 0 ring B X 101E FMCRD Power 19000 75 1350 300 0 ring Il X 101F FMCRD Power 19000 257 1350 300 0 ring Il X 101G FMCRD Power 19000 103 1350 300 0 ring B X 102A C&I 1M00 45 1350 300 O riag Il X 10211 C&I 16400 180 1350 300 0 ring 11 X 102C C&1 1M00 180 2650 300 0 ring Il X 102D C&1 16400 280 0 300 0 ring 11 X-102E C&1 16400 45 2700 300 0 ring Il X 102P C&I 16400 180 2653 300 O-ring B X 102G C&I 13500 180 1350 300 O ring B X-10211 FMCRD Control 19000 75 2700 300 0 ring B X 102J FMCRD Control 19000 257 2700 300 0 ring B X 103A C&1 1900 45 1350 300 0 ring B X-103B C&1 13500 180 50 300 O ring B X 103C C&1 16400 180 5250 300 O ring B X 104A FMCRD Position Indicator 19000 75 0 300 0-ring B X 104Il FMCRD Position Indicator 19000 257 0 300 0-ring Il X 104C FMCRD Position Indicator 19000 103 0 300 0 ring B X 104D FMCRD Position Indicator 19000 285 0 300 0-ring Il X 104E FMCRD Position Indicator 19000 75 1350 300 0-ring B X 104P FMCRD Position Indicator 19000 257 1350 300 0 ring Il X 104G FMCRD Position Indicator 19W0 103 1350 300 0 sing Il X 104H FMCRD Position Indicator 19000 285 1350 300 0 ring 11 X 105A Neutron Detection 13500 55 1000 300 0 ring B X 105B Neutron Detection 13500 180 1350 300 0 ring D X 105C Neutron Detection 13500 180 5250 300 0 ring Il X 105D Neutron Detection 13500 280 1350 300 0 ring B X 110 Spare 20100 103 2700 300 0 ring Il X-111 Spare 20100 285 2700 300 0 ring B X 112 Spare 1900 45 4050 300 0 ring 11 X-113 Spare 14500 220 0 300 0-ring B O Amendment 17 6.2 30.51
l ABWR 2 mima ev c i Sinndnrd I'latit Table 6 2 8 Primary Containment Penetration List 2 (Continued) Elevation Attmuth O!htt Diameter llarrier Testingl 3 l'enttration Name Numlier (mm) (deg) (mm) (mm) Type X 130A C&I 13500 45 0 300 0 ring B X 130Il C&1 13500 124 0 300 0 ring !! X-130C C&1 13500 212 0 3w 0. ring B X 130D C&1 13500 295 0 300 0 ring 11 X 140A C&1 13500 45 1000 300 0 ring 11 X 140B C&1 13500 300 0 XO O ring 11 X 141A C&1 13500 63 5 0 300 0 ring 11 X-141Il C&1 13500 208 0 300 0 ring B X 142A C&1 20100 38 0 100 0 ring 11 X 14211 C&1 20100 116 0 100 0 ring B X 142C C&I 20100 244 0 100 0 ting Il X 142D C&1 20100 296.5 XX)0 100 0-ring Il X 143A C&1 14700 45 0 100 0 ring 11 X 14311 C&1 14700 124 0 100 0 ring 11 X 143C C&1 14700 212 0 100 0 ring 11 X 143D Cki 14700 300 0 100 0 ring 11 X 144A C&1 12650 45 0 1rx) 0. ring 11 X-144B C&1 12650 124 0 100 0 ring 11 X 144C C&1 12650 212 0 100 0 ring 11 X 144D C&1 12650 300 0 100 0 ring 11 X 146A C&1 19000 38 0 X)0 0 ring 11 X 14611 C&1 19000 112 0 300 0 ring Il X 146C C&1 19000 248 0 XO O ring 11 X 146D C&1 19000 2965 0 300 0 ring 11 X 147 C&I 20100 248 0 100 0 ring 11 X 160 LDS hionitor 20100 42 0 300 0 ring Il X 161A CAhiS C & 1 14700 45 1000 250 0 ring 11 X 16111 CAhtS C & 1 14700 290 0 250 0 ring B X 162A CAhtS C & 1 19000 116 0 250 0 ring 11 X-162B CAh1S C & 1 19000 244 0 250 0 ring 11 X 170 C&1 13500 310 0 100 0-ring 11 X 171 C&1 20100 50 0 300 0 ring Il X 177 C&1 15900 135 1200 X0. 0 ring B X 200A Wetwc3 Spray 8900 260 0 100 A ( 2001) Wetwell Spray 8900 100 0 100 A X 201 RilR Pump Suction (A) 7085 36 0 450 A X 202 RllR Pump Suction (D) 7085 216 0 450 A X 203 RiiR Pump Suction (C) 7085 144 0 45C A X 204 RHR Pump Test (A) 1200 85 0 250 A X 205 RilR Pump Test (11) 1200 265 0 250 A X 206 RilR Pump Test (C) 1200 95 0 250 A 0 400 A O X-210 X 211 1iPCF Pump Suction (11) 11PCF Pump Suction (C) 70S5
-7035 252 108 0 4fD A Amendment 17 6.25052
ABWR mamn Standard I'lant _ niv c Table 6.2 8 l'rlmary Containment l'enetration List 2 (Continued) Penetration Name Elriatlun Atimuth Othet Diameter Barrier Testing 13 l Nu mt.cr (mm) (deg) (mm) (mm) Type X 213 RCIC Turbine Exhaust SMO 60 0 '350 A X 214 RCIC Pump suction 7050 72 0 200 A X 215 RCIC Pump Exhaust 50 A X 216 SPCU Pump Suction 7050 306 1XO 2(O A X 217 SPCU Return 1550 340 0 250 A X 230 tow Conducthity Drain X-231 liigh Conducthity Drain X 240 Wetwell Purge Suction 85(0 45 500 550 A X 241 Wetwell Purge Exhaust 9(KD 230 0 550 A X 242 FCS Return 800 220 0 150 A X 243 VGL Exhaust 8850 32 0 50 A X 250 Spare 8500 45 0 400 A X 251 Spare 9000 213 0 4(0 A X 300A CkI m]0 132 0 300 0 ring Il X 300ll C&1 6000 213 0 300 0 ring Il X 320 C&1 8850 70 0 100 0 ring Il X 321A C&I 6000 32 0 300 0 ring Il X 32111 C&I 6000 220 0 300 0 ring Il X 322A C&1 400 75 0 100 0 ring 11 X 322Il C&1 400 285 0 100 0 ring B X 322C C&I 400 80 0 100 0-ring !! X 322D C&1 400 105 0 100 0 ring Il X 322E C&1 4C0 255 0 100 0-ring 11 X 322I C&I 400 280 0 100 0-ring Il X 323A C&1 4700 75 0 100 0 ring B X 323B C&I -4700 285 0 100 0-ring D X 323C C&I -6700 80 0 H0 O ring H X 323D C&I 6700 115 0 100 0 ring 11 X 323E C&I -6700 245 0 100 O ring B X 323F C&I -6700 230 0 100 0 ring B X 324A Spare X 324B Sparc X 324C Sparc X 324D Sparc l X 330 LDS Monitor Return 8850 78 0 100 0 ring D l X 331A CAMS Gamma Det. 6000 24 0 250 O ring !! l X 331B CAMS Gamma Det. 6000 230 0 250 0 ring B X 332A CAMS Sampling Ret. 6000 128 0 300 0-ring Il X 332B CAMS Sampling Ret. 6000 224 0 300 0 ring 11 X-334 C&I 8850 74 0 100 0-ring 11 1 X-341 C&I IIold 100 0 ring Il X 342 C&1 lloid 100 0 ring Il l Amendment 17 6.2 M53
ABWR 2mian Standard Plant nry c (~') %J Table 6,2 8 Primary Containment Penetration List 2 (Continued) Penetration Name Elevation Ar.imuth Ofhet Diameter llarrier Testing1 3 Numter (mm) (deg) (mm) (mm) 'lype X 400A TIP Drive 95 A X-40011 TIP Drive 95 A X-400C TIP Drive 95 A X 401 TIP Drive Purge 95 A Notes:
- 1. All penetrations will be subject to the Type A test. Those penetrations subject to Type 11 testing are also tested in the Type A test.
- 2. This table provided in response to Questions 430.49d & e.
- 3. All penetrations excluded from Type 11 testing are welded penetrations and do not include resilient seals in their design.
,.C\t G [~'\ V Amendment 17 6.2 50 $/
MM 23A6100AD Standard Plant ntv. c Table 6.2 10 Polential Bypass leakage Paths 1 Termluation Leakage Potential item Name Diameter RegionW Barrierg2) Hypass Path (mm) 4 X1 U/D Equipment ilatch 2600 S No X2 U/D PersonnelIIstch 2600 S No X3 ISiliatch 200 S No X4 Wetwell Access liatch 2000 S No X 5- L/D Personnel 11atch 2400 S No X6 L/D Equipment 11atch 2400 S No j X 10A - Mainsteam Line: 1100 E C No
-X 10B Mainsteam Line 1100 E C No
-X 10C Mainstcam Line 1100 E C No X 10D Mainsteam Line 1100 E C No X 11 Mainsteam Drain - 600 -E C No
.X 12A Feedwater Line 950 E C No X 12B Feedwater Line- 950 E C No
' X 22 Borated Water Injection 40 S No X 30A - Drywell Spray 200 S No X 30B Drywell Spary 200 S No X 31A IIPCF (B)- 600 S No X 31B llPCF (C) 600 S No i X 324 LPFL (B) 650 S No i X 32B LPFL(C) 650 S No X 33A RIIR Suction (A) 750 S No X 33B Rif P. Suction (B) - 750 S No-X 33C - Rif R Suction (C) 750 S No X 37. RCIC Turbine Steam 550 S No X 38E RPV llead Spray 550 S No X 50 CUW Purnp Feed 600 S No X 60 MUWP Suction 50 S No X-61 RCW Suction (A) 200 E C No X 62 ' RCW Return (A) 200 E C No X 63- RCW Suction (B) 200 -E C No X 64 RCW Return (B) . 200 E C No X-65 llNCW Suction 150 E C No X-66 IINCW Return 150 E C No X-69 SA 25 E C No X 70 .1A _
50 E C. No X 71A.' ADS Accumulator (A) 50 S No X 71B ' ADS Accumulator (B) . 50 S No X 72 Relief Yahc Accum. 150 S No X 80 Drywell Purge Suction 550 E C No
. X-81 Drywc!! Purge Exhaust $50 E C No-X 82 FCS Suction - 100 S No X 90 Spare 400 P A No ,
X . Spare 400 P A - No X 92 . Spare 400 P A No i X 93- Spare . 400 P A No X 100A. IP Power 450 S No X 100B . IP Power 450 S No X 100C IP Power 450 $ No l 6.2 50.57 Amendment 17
-r ne r- . -, , - - - ., - ----,- -----wn- =s-w . N -- e ,r+- m e-- -r-w-r+ %+ywev e e - -s- e-y ee-,*m--o--, ' + +
- ABWR u-n Standard Plant nuv c Table 6 210 Potential Ilypass Leakage Paths 1 (Continued) g Termination leakage Potential item Name Diameter RegionP) BarriersG) Ilypass Path (mm) ,
X-100D IP Power 450 S No X 100E IP Power 450 S No X 101A LP Power 300 S No X 101B LP Power 300 S No X-101C LP Power 300 S No X-101D 131CRD Power 300 $ No X 101E Fh!CRD Power 300 S No X 101F PhiCRD Power 300 S No X 101G FhiCRD Power 300 S No X 102A C&1 300 S No X 102B C&1 300 S No l X 102C C&1 300 S No l X 102D C&I 300 S No X 102E C&I 300 S No X 102F C&1 300 S No X-102G C&I 300 S No X-102H I%iCRD Control 300 S No X 102J FhiCRD Controf 300 S No X 103A C&I 300 S No X 103B C&I 300 S No X 103C C&I 300 S No X 104A Fh1CRD Pos. Indicator 300 S No X 104B Fh1CRD Pos. Indicator 300 S No X 104C FhtCRD Pos. Indicator 300 S No X 104D Fh!CRD Pos. Indicator 300 S No X 104E FhiCRD Pos. Indicator 300 S No X-104F Fh1CRD Pos. Indicator 300 S No X 104G Ph1CRD Pos. Indicator 300 S No X 10411 Th1CRD Pos. Indicator 300 S No r X-105A Neutron Detection 300 S No l X 105B Neutron Detection 300 S No x 105C Neutron Indicator 300 S No X 105D Neutron Indicator 300 S No X-110 Spare 300 P A No X-111 Spare 300 P A No X 112 Spare 300 P A No X 113 Spare 300 P A No x-130A C&I 300 S No X-130B C&I 300 S No X-130C C&I 300 S No X 130D C&I 300 S No X 140A C&I 300 S No j X-140B C&I 300 S No X 141A C&1 300 S No X-141B C&I 300 S No X-142A C&1 100 S No X 1428 C&1 100 S No X-142C C&1 100 S No X-142D C&1 100 S No Amendment 17 6.2 50,$8 l
ABWR 2mmn MaBdard Plant niv c Table 6.2 10 l'otential lhpass Leakage l'aths1 (Continued) (v) Termination leakt 1;e Potential item Name Diameter Region (3) llarriers(2) Ilypass Path (mm) X 143A C&I 100 S No X 143B C&1 100 S No X 143C C&1 100 S No X-143D C&I 100 S No X 144A C&1 100 S No X 144B C&I 100 S No X 144C C&1 100 S No j X-144D C&1 100 S No l X 146A C&I 300 S No ' Notes:
- 1. This Table prodded in resportse to Question 430.52b. l l
- 2. A) Penetration is capped l
- 11) Terminates at Primary Containment Wall 'l I C) Terminates inside Secondary Containment l
D) Terminate.5 outside Secondary Containment ! E) Redundant Containment Isolation Valves (] 3. E Emironment V P Primary containment S Secondary containment 1 l l t l
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1 ABWR mamn ni v c Standard Plant F' system it iesigned to provide and No occupation is expected in spaces numbered maintain au environment with controlled 4 and 12 during an emergency mode of opera-temperature and humidity to ensure both Lion. Of these spaces, all but the mechanical comfort and safety of the operators and equipment room are maintained at a positive the integrity of the control room pressure of + 0.125 to + 0.5 in, of water gage components, pressure at all times. The mechanical equipment room is maintained at 0.0 to + 0.5 in. (2) Provisions for periodle inspection, of water gage. Pressure control dampers at the testing and maintenance of the principal inlet of the ventilation system maintain these components shall be a part of the design pressures. These spaces constitute the requirements. operatiou, living and environmental control aress and can be isolated for an extended 6.4.2 System Design period if such is required by the existence of a LOCA or high radiation condition. Figure 9.41 provides the flow diagrams describing the control building IIVAC system. IIcating, cooling and pressurizing t!.e control building, and filtering the air therein, is fully described in Section 9.4.1, wherein function is discussed and equipment is listed. 6.4.2.2 Ventilation System Design 6.4.2.1 Control hullding Fnvelope The design, construction and operation of The control building spaces within the the control building IIVAC system are described envelope supplied by the llVAC habitability in detail in Subsection 9.4.1. Figure 9.41 is systems includes: a diagram of the control building ilVAC system, q seowing major components, scismic (1) control room proper including the e'assifications and instrumentation. critical document file; Description of the charcoal filters is given (2) computer room; in Subsection 9.4.1. (3) control equipment room; Description of control room instrumentation for monitoring of radioinctivity is given in (4) upper and lower corridors; Subsections 11.5.2 and 12.3.4. (5) elevator shaft and stair wells; A description of the smoke detectors is in Subsection 9.5.1. (6) office and chart room; 6A.2.2.1 Control Room Drawings (7) kitchen ed lur ch rooms; Layout drawings of the control room and the (8) instrument repair toom; remainder of the control building are given in Section 1.2. (9) sleeping area 6A.2.2.2 Release Points (10) men's lavatory; Release points (SGTS vent) are shown in (11) women's lavatory and lounge; Figure 6.4-1 (plan view). The air intakes are E cell above grade. Elevation of other & (12) IIVAC mechanical equipment rooms. structures is seen in Figures 1.2 9 and 1.210. t,4 3 Anwadment 17
ABWR mman Standard Plant wvc 6.4.2J Leahtightness & Leakage through the various paths is W The control building boundary walls are negligible except for that through the doors. designed with low leakage construction. All boundary penetrations are scaled. The access 6.4.2.4 Interaction With Other Zones and doors are designed with self closing devices Pressur Containing Equipment which close and latch the doors automatically following the passage of personnel The control building is heated, cooled, ventilated and pressurized by a recirculating All potential leak paths in and from the air system using filtered outdoor air for control building boundary are tabulated and shown ventilation and pressuriration purposes, on Figures 9.41. Recirculated air and outdoor air are mixed and drawn through filters, a cooling coil and rone The control room in leakage analysis was electric reheating coils. performed using the methods and assumptions given in NAA SR 10100 (Conventional Building For Th:re are two intakes on the top floor side Reactor Containment), Atomics International, and walls of the control building, one on each Regulatory Guide 1.78. The leakage rate is end. Radiation monitoring sensors located in calculated using the following: each duct warn the operating personnel (by means of readouts and alarms in the main (1) a 1/2 in. WG differential across control room) of the presence of airborne surfaces and cornponents exposed to or contamination. Also, the signal automatically protected from effects of winds; closes down, the contaminated air intale valves (2) maximuru design differential for closed and normal vent dampers, opens the emergency , dampers on the suction side of the vent dampers, and turns on the primary emergency filter unit fans on reduced flow. If ( supply f ans; and both air intakes are contaminated the control (3) Equation: room operator can manually override the system to open either air intake to draw makeup air q = AP + BPI /2 (M1) when necessary. This makeup air is routed through IIEPA and charcoal filtering system for where cleanup before being used for pressurization. q = leakage rate per unit leak The control room is maintained at positive path (cfm); pressure with respect to atmosphere. In an emergency the pressure differential will o A = empirical constant (cfm per climinate infiltration of airborne contamina- ? unit leak path per inch of tion. The doors are of the double vestibule 0 water pressure); type to increase pressure differential between roomst thereby eliminating infiltration when B = cmpirical constant (cfm per the doors are opened. unit leak path per 1/2 inch of water pressure); and The control room must remain habitable dur-ing emergency conditions. To make this possi-P = differential pressure (in. ble, potential sources of danger such as steam w.g.). lines, pressure vessels, CO2 fire fighting containers, etc. are located outside of the The leak paths considered were concrete walls control room and the compartments containing and slabs, wall and slab joints, door frames, control building life support systems. doors, electric cable penetrations, duct penetrations and pipe penetrations. The A tabulation of moving components in the empirical constants A and B for each leak path control building HVAC system, along with the are taken from NAA-SR 10100. Amendment 7 64-4
- ABWR m aman Standard Plant .. mvc O 6.5 FISSION PRODUCTS REMOVAL AND CONTROL SYSTEMS (4) Remain intact and functional in the event of a safe shutdown carthquake (SSE).
6.5.1 Engineettd Safety Features Filter (5) Meet eaviroamentai qualificatton Systems requirements established for system operation. The filter systems required to perform i safety related functions following a design basis accident are: (1) Standby gas treatment system (T22 SGTS). (2) Control room portion of the liVAC system. (U41 l!VAC) 63.1.2 System Design The control room portion of the llVAC system is discussed in Section 6.4 and Subsection 9.4.1. 6.5.1.2.1. General ( Tbc SGTS is discussed in this Subsection (6.5.1). 6.5.1.1 Iksign liasis 6.5.1.2.2 Component Description 6.5.1.1.1 Power Generatlon Design Basis Table 6.51 provides a summary of the major The SGTS has the capability to filter the SGTS components. The SGTS consists of two gaseous effluent from the primary containment or parallel and iedundant trains of active from the secondary containment when required to equipment which share a single filter train, O limit the discharge of radioactivity to the Suction is taken from above the refueling area environment to meet 10CFR100 requirements. or frora the primary containment via the
. atmospheric control system (T31 ACS). The 6.5.1.1.2 Safety Design Basis discharge goes to the main plant stack.
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 containment, relative to the moisture separator and an electric p*ncess outdoor atmosphere, to control the release heater. of fission products to the environment. (2) Two independent process fans located upstream of the filter train. (3) A filter train consisting of a prefilter, a I (2) Filter airborne radioactivity (halogen and high efficiency particulate air (llEPA) air particulates) in the effluent to reduce. filter, a charcoal adsorber, a second 11 epa offsite doses to within the limits specified filter, and space heaters. l in 10CFR100. 6.5.1.2.3 SGTS Operation 63.1.2.3.1 Automatic (3) Ensure that failure of any active component, assuming loss of offsite power, cannot Upon the receipt of a high primary O 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 in the secondary containment or refueling floor Amendment It 63 1 h .- - - - - _ - - _ - . .--_ _ _ - - _ _ - -__ - -_-
ABWR :wmn Shtljdard Pl11nt nue ventilation exhaust, the SOTS is automatically efficiencies are outlined in Table 6.5-1. Dose actuated. If system operation is not confirmed, the analyses of events requiring SGTS operation, redundant process fan and dryer train are described in Subsecticos 15.6.5 and 15.7.4, automatically placed into service, in the event a indicate that offsite doses are within the limits malfunction disables an operating process f an er established by 10 CFR 10d. dryer train, the standby process fan and dryer train are manually initiated. (3) The SGTS is designated as an enginected safety feature since it mitigates the 6.5.1.2J.2 Manual consequences of a postulated accident by controlling and reducing the release of The SGTS is on standby during normal plant radioactivity to the environtnent. The SGTS, operation and may be manually initiated before or except for the deluge,is designed and built to during primary containment purging (de inerting) the requirements for Safety Clas.3 equipment when required to limit the discharge of contaminants as defined in Section 3.2, and 10 CFR 50, to the environment. It may be manually initiated Appendix It when:ver its use may be needed to avoid exceeding radiation monitor setpoints. The SGTS has independent, redundant active l components. Should any active component fail, 6.5.1.23.3 Decay llent kemoval SGTS functions can be performed by the redundant component. The electrical devices , Cooling of the SGTS filters may be required to of independent components are powered from ) pre,*ent the gradual accumulation of decay heat in separate Class 1E electrical buses. i the charcoal. This beat is generated by the decay of radioactive iodine adsorbed on the SGTS charcoal. (1) The SGT3 is designed to Seismic Category i The charcoal is typically cooled by the air from the requirements as specified in Section 3.2. The process fan. SGTS is housed in a Category I structure. All surrounding equipment, components, and A water deluge capability is also prosided, but supports are designed to appropriate safety primarily for fire protection since redundant process class and scismle irquirements, fans are provided for air cooling. Since the deluge is asailable,it may also be used to remove decay heat (5) The SGTS design is based on the maximum for sequences outside the normal design basis. pressure and differential pressure, maximum Temperature instrumentation is provided for control integrated dose rate, maximum relative of the SGTS process and space electric heaters. This humidity, and maximum temperature expected instrumentation may also be used by the operator to in secondary containment for the LOCA esent. [re ] establish a cooling air flow post accident, if required. 6.5.13.2 String liasis Water is supplied from the fire protection system Figure 6.5 2 provides an assessment of the and is connected to the SGTS via a spool piece. secondary containment pressure after the derign basis LOCA assuming an SGTS fan capacity 6.5.13 Design I: valuation of 4000 scfm (70 F, I atmosphere) per fan and the leakage rates shown in Table 6.5 2. Credit for 6.5.1.3.1 General secondary containment as a fission product control system is only taken if the secondary containment is (1) A slight negative pressure is normally actually at a negative pressure by considering the maintained in the secondary containment by potential effect of wind on the ambient pressure in the reactor building ilVAC system (Subsection the vicinity of the reactor building. Foi the AllWR 9 4.5). On SGTS initiation per Subsection dose analysis, direct transport of containment 6.5.1.2.3.1, the secondary containment is leakage to the emironment was assumed for the first automatically isolated from the IIVAC system. 20 minutes after LOCA event initiation (in addition to the leakage through the MSIVs to the main (2) The SGTS filter particulate and charcoal turbine condenser). Each SGTS fan was sized to Amendment 17 G2
ABWR men Standard _ Plant we (' establish a continuously negative differential These features include: ( pressure, ss measured across the leeward building face, within 10 minutes of SGTS initiation. The dose (1) The advanced design of the filter housing and analysis therefore assumes direct leakage from the flow pattern virtually climinates any untreated containment to the environs for twice the required bypass of the filter, in addition, the all welded period. In addition,it should be recognized that desig.n is such that degradation of filter housing fission product release on the order of that specified integrity is not likely to occur during system in Regulatory Guide 1.3 and used in the LOCA dose standby or operation. analyses (Subsection 15.6.5) realistically requires significant core damu,e and most likely more than 10 (2) Sufficient instances of inadvertent deluge or 20 minutes for transport to and leakage from the wetting the charcoal and rendering the filter containment. train unavailable have been observed to warrant an improved deluge design concept. The calculation accounted for all expected heat These unintended del ge 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, AllWR design, the deluge piping it, not no single failure of a diesel was assumed since it is connected permanently from the flue protection most likely all divisions of power would be available. system to the filter housing nonje. Instead, a Failure of one SGTS fan to start wx assumed as the normally disconnected hose from the fire single failure. Therefore, heat loads from all protection system is provided to act as a 4 pool 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, ll.3(b) and SRP 6.5.3, 11.2, (3) Decay heat is not sufficient to cause a fire in secondary containment should be held below 0.25 the charcoal adsorber or llEPA filter. Calcula-inch w.g. under all wind conditions up to the wind tions indicate that air flow from eithet speed at which diffusion becomes great enough to redundant process fan is more than enough to assure site boundary exposures less than those remove the heat from decay of the radioactise calculated for design basis accidents, esen if iodine on the charcoal or filters. Heating does ex filtration occurs (i.e., no credit for SGTS is not occur sufficient to cause iodine desorption taken). For the ABWR, dispersion factors were or ignition of the charcoal. With the reduced calculated for each stability class over a range of source term espected 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. 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 opcration or plant shutdown. These sequences 1 m/s wind speed. Therefore, the ABWP.SGTS was would involve an unspecified maintenance or designed to establish and maintain a negative operating personnel activity or an incredible pressure in secondary containment within 10 minutes malfunction of the space heaters. In this case a for any wind speed up to and including 8.9 m/s (20 fire in the SGTS charcoal, like in the offgas 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 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 passive, and in practice provides the reliability SGTS unavailability) when the SGTS is O associated with a passive component. Furthermore, required to meet the licensing basis release b the ABWR SGTS has incorporated design features to eliminate potential failures or improper operation. limits (and presumably inaccessible for repair). Amendment 17 M-3
ABWR &wimAu E111fhlId.Plttnf umc 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 sire, meeting the 0.25 sec minimites 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 CJ.i), and using a l housing. This uniform heating further reduces conservative!" 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 beater and by improving the consistent with toe purctgased charcoal density accuracy of the temperature measurements (usually less than 30 lb/ft ) and any dead space (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 actuai (4) Degradation of the charcoal effectiveness source term, prosiding 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 iodine 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 pool is 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 AllWR design or during de inerting of the primary under Paragraph 8.9 of the Licensing Review containment prior to plant shutdown. The 11 asis, more likely, though still infrequent, potential use of SGTS is during de inerting. Depending (5) Ilecause of the high availability of the AllWR, on indications from leak detection and isolation de inerting, and the potential use of SGTS 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,ilEPA litter initiated or from the process radiation and charcoal adsorber effectiser. css will be monitoring (D11 PRM) reactor building tested, and the filter and/or charcoal replaced, ventilation exhaast radiation monitors during if necessary, before the plant returns to power de inerting, SGTS may be placed into service. operation. The ABWR SGTS charcoal bed thickness has been increased two inches, to six inches, as With these SGTS design features,long term, compared to the GESSAR 11 design. The undetected, passive failure of the filter train has been 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 wl.en 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 in addition to the increased charcoal bed depth, Electric considers an active component to be a significantly rnore 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 requirernents, a 100% redundant dryer trains, each containing a moisture efficient 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 Amendment 11 6M
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SECTION 8.2 CONTENTS i Section Illle Eagt j 8.2.1 Ikscription 8.2-1 8.2.2 Anah31s 8.2 1 8.2.2.1 Generic Desig,n Criteria 8.21 8.23 Interfaces 8.2-1 8.23.1 Class IE Feeder Circuits 8.2-1 8.23.2 Non Class IE Feeders 8.2-1 l 8.233 Specific Offsite Power Sptems Interfaces 8.21
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- ABWR ==40 l Standarti Plant REY. D 8.2 OFFSITE POWER SYSTEMS provided by routing each train on a dif.
ferent floor in each building. Although 8.2.1 Description these trains are non. Class 1E, such separ-ation assures the physicalindependence As indicated in Section 8.1, the utility grid requirements of GDC 17 are preserved. and the mala power transformer are not within the ABWR Standard Plant scope. The interface (3) GDC 18 Inspection and Testing of requirements at the main power transformer are Electrical Power Systems; given in Subsection 8.2.3. - All other equipment downstream of the main power transformer is All equipment esa be inspected and testes in included in the ABWR Standard Plant scope. This accordance with this GDC. g includes the auxiliary transformers, switchyard p
. components, the main generator, etc., which are (4) RG's 1.32,1.47, and BTP ICSB 21; l assigned by SRP Section 8.2 as being part of the i
- preferred power system", also called the These distribution trains ate non Class 1E )
*offsite power system." Since GE considers these and non. safety related. Therefore, this I components to be *onsite", their description is criteria is not applicable.'
provided in Subsection 8.1.2.1. ' (5) BTP ICSB 11 (PSB) Stability of Offsite 8.2.2 Arialysis ' Power Systems; In accordance with the NRC Standard Review See Subsection 8.1.4.1 for interface Plan (NUREG 0800), Table 81 and Section 8.2, Ihe equirement. power distribution system between the main transformer and the Class 1E distribution system 8.2.3 Interf' aces q interfaces is designed consistent with the following criteria, so far as it applies to the The interface point between the ABWR design g non. Class IE equipment. Any exceptions or and the utility design for the main generator clarifications are so noted. output is at the connection of the isolated phase bus to the main power transformer low 8.2.2.1 General Design Criteria . voltage terminals. The rated conditions for this interface is 1500 MVA and 26.325 KV. It is (1) GDC 5 and RG 1.81 Sharing of Structures, a requirement that the utility provide Systems and Components; sufficient impedance in the main power transformer and the high voltage circuit to The ALWR is a single unit plant design. limit the primary side maximum available fault Therefore, these critcria ate not current contribution from the system to no more e, applicable. than 275 KA symmetrical and 340 KA asymmetrical at 5 cycles from inception of the fault These 6 (2) GDC 17. Electric Power Systems; values should be acceptable to most utilities. When all equipment and system parameters are As shown in Figure 8.31, each of the Class known, a refined calculation ba ed on the known IE divisional 6.9 KV M/C buses can receive values with fault located at the generator side power from multiple sources. There are of the generator breaker may be made. This may separate utility feeds from the station Crid allow a lower impedance for the main power
-(via the main transformer), and the offsite transformer, if desired.
line (via the reserve auxiliary trans. former). . The two emergency unit auxiliary The second power interface occurs at the high transformer output power feeds are routed by voltage terminals of the reserve auxiliary two completely separate paths through the transformer. The rated load is 30 KVA at a 0.9 turbine btilld'ng, control building and power factor. The voltage and frequency will be reactor building to their destinations in the utilities standard with the actual values to pd the emergency electric rooms. Separation is be determined at contract award. B 2-1 Amendment 17 -
ABWR 2mimo Standard Plant nrv. n Protective relaying interfaces for the two power syatem interfaces are to be defined during the detail design phase following contract award. O O Amendment 17 g22
ABWR mmo Slandard Plant _ iuv n
- 1. OPP dtLt]ne diesel citterator twrallellitg 83.1.1.fL1 lledundant Standby AC Pour
[V (6) itil: If the normal preferred power supply Supplies is lost during the diesel generator paral-leling test, the diesel generator circuit Each standby power system division, including breaker is automatically tripped. Transfer the diesel generator, its auxiliary systemt and to the diesel generator then proceeds as the distribution of power to various Class 1E described in (1), loads through the 6.9kV and 480V systems, is se-gregated and separated from the other divi-If the alternate preferred source is used sions. No automatic interconnection is provided for load testing the diesel generator, and between the Class 1E divisions. Each diesel.ge-the alternate preferred source is lost (and nerator set is operated independently of the no LOCA signal exists), the diesel. generator other sets and is connected to the utility power breaker will trip on overcurrent, and LOPP system by manual control only during testing or cor.dition will exist. Load shedding and bus fo. but transfer, transfer will proceed as described in (1). 83.1.1.8.2 Itatings and Capability (7) fle s t o ra t to n o f o f f sit e p ow er: Upon restoration of offsite power, the Class IE The size of each of the dicsci generators bus (es) can be transferred back to the serving Divisions 1,11 and 111 sathfies the re-offsite source by manual operation only. quirements of NitC itegulatory Guide l.9 and lEEE Std 3S7 and conforms to the following criteria: (8) l'InittihLILDERIRALi[tgradr d 5 011agt: For protection of the Division 1,11 and !!! (1) Each diesel generator is capable of start-electrical equipment against the effests of ing, accelerating and supplying its loads in a sustained degraded voltage, the 6.9 kV ESF the scquence shown in Table 83 4. q bus voltages are monitored. When the bus h voltage degrades to 90% or belov> of its rated value and af ter a time delay (to (2) Each diesel generator is capable of start-ing, accelerating and supplying its loads in n prevent triggering by transients), their proper sequence without exceeding a
$ undervoltage will be annunciated in the 25% voltage drop at its terminals.
control room. Simultaneously a 5 minute timer is started, to allow the operator to (3) Each diesel generator is capable of start. take corrective action. After 5 minutes, ing, accelerating and running its largest the respective feeder breaker with the motor at any time afte' the automatic load-undervoltage is tripped. Should a LOCA ing sequence is completed, assuming that the occur during the 5 minute time delay, the motor had failed to start initially. feeder breaker with the undervoltage will be tripped instantly. Subsequent bus transfer will be as destibed above, (4)full Each speed diesel generator and voltage within 20 ir capable seconds of reaching after receiving a signal to start, and cap. 83.1.1.8 Standby AC Power Sptem able of being fully loaded within the next 30 seconds. (The actual losd sequence has The diesel generators comprising the Divi- additional margin as shown in Table 8.3 4.) sions I,11 and ill standby AC power supplies are designed to quickly restore power to their respec- See Subsection 8.3.4.2 and 8.3.4.8 for l E,", {," tive Class 1E distribution system divisions as interfcce requirements. required to achieve safe shutdown of the plant and/or to mitigate the consequences of a LOCA in 83.1.1.83 Starting Circuits and Sptrms the event of a coincident LOPP. Figure 8.31 shows the interconnections between the preferred Diesel generators I, il and ill stact automa-power supplies and the Divisions I,11 and 111 tically on loss of bus voltage. Under voltage g relays are used to start each diesel engine in ( diesel generator standby power supplies. k. Amendment 17 834
ABWR mama Standard Plant niv. n the event of a drop in bus voltage below preset values for a predetermined period of time. Low water level switches and drywell high pres-l sure switches in each division are used to ini-tinte diesci start under accident conditions. Manual start capability (without need of D.C. power) is also provided. The transfer of the Class 1E buses to standby power supply is automatic should this become accessary on loss of all preferred power. After the breakers connecting the buses to the preferred power supplies are open the diesel generate. breaker is closed when required generator voltage and frequency are established. Diesel generators 1,11 and 111 are designed to start and attain rated voltage and frequency
, l within 20 seconds. The generator, and voltap,e regulator are designed to permit the set to accept the load and to accelerate the motors in the sequence within the time requirements. The voltage drop caused by starting the large motors does not exceed the requirements set forth in Regulatory Guide 1.9, and proper acceleration of these motors is ensured. Control and timing O
l l l O I l Amendment 10 BM1 1. l
ABWR m=m Slflildllid_l'Illill RIV is
,Q facilitate adequate separation of cabling. cur'.es of the electrical penetrations
- primary V 8333 lirr Detection and Protection Systems and secondery current interrupting devices plotted agninst the thermal capability (12 t) curve of the penetration (to maintain mechanical All areas except the diesel generator room are integrity). Also, provide a simplified one line protected by product of combustion detectors. The diagram showing the location of the protective diesel generator rooms are protected by carbon devices in the penetration circuit, and indicate dioxide suppression, which is actuated by com- the masimum available fault current o.' the pensated rate of heat rise and ultraviolet flame circuit.
detectors. Provide specific identification of power p Automatie wet standpipe, sprinklers, hose supplies used to provide external control power R reels, and manual pull boxes for the operator's for tripping primary and backup electrical $ initiation of fire signals are providrd in r ens penetration breakers (if utilized). as described in subsection 9.5.1, which inc a'es areas where cables and cable trays are to au. 83.4.5 Analpls Testing for Spallal
&i arathu per IEEE 384 8.3.4 Interfbees items (4) and (5) in Subsection 8.3.1.4.2.2.2 83.4.1 Interrupting Cupacity of Electrical state that spatial tieparation in general plant q Distribution Equipment areas and in cable spreading areas equal or m y exceed the minimum allowed by IEEE 384.
? The interrupting capacity of the switchgear identify any specific instances where this and circuit interrupting devices ruust be shown to requirement is met by testing and analysis (as be compatible with the magnitude of the available opposed to actual distance).
f ault current based on final selection of the y transformer impedence, etc. (See Subsection 83.4.6 DC Voltage Analpls
\
8.3.1.1.5.2(4)) . Provide a DC voltage analysis showing g 83.4.2 Diercl Generator Design Details battery terminal voltage and worst case DC load p terminal voltage at each step of the Class IE g Subsection 8.3.1.1.8.2 (4 ) requires the battery loading profile. (See Subsection g diesel generators be (apable of reaching full 8.3. 2.1 ) lg speed and voltage within 20 seconds after the signal to start. Demonstrate the reliability of Proside the manufactor's ampere. hour rating the diesel generator start up circuitry designed of the batteries at the two hour rate and at the E to accomplish this, eight hour rate, and provide the one minute [ ampere rating of the batteries (see Subsection 83.43 Certified Proof Tests on 83.2.1.3.2). Cable Samples 83.4.7 Seismic Qualification of E,sewash g Subsection 8.3.1.2.4 requires certified proof Equipment j tests on cables to demonstrate 60 year life, and
- resistance to radiation, flame and the S ubse ction 8.3.2.1.3 spe cifie s t hat an environment. Demonstrate the testing methodology emergency eyewash shall be located in each W to assure such attributes are acceptable for the battery room. Provide assur ance that the 6 60 year life. cyewash and associated piping are seismically qualified, and that the eyewash is located sucn 83.4.4 Electrical Penetration Assemblics that water cannot splash on the battery.
O si Subsection 8.3.1.4.1.2 (7) specifies design 8.3.4.8 Dienl Generator Load
$ requirements for electrical penetration Table Changea 4 l, y/ assemblics. Provide fault current clearing time p Table s 8.3 1 a nd 8.3 3 are ge n e ric.
l Ilowever, changes may be needed for specific Amendment 17 a123 [
ABWR 2minao Standard Plant Rrm ri plant applications. Such changes, if any, shall be identified and addressed. (See Subsection 8.3.1.1.8.2) 8.3.4.9 Offsite Power Supply Arrangement Operating procedures shall require one of the y three divisional buses of Figure 8.31 be fed by t:i the altersate power source during normal operation; in order to prevent simultaneous deenergiration of all divisional buses on the loss of only one of the offsite power supplies. 8J.4.10 Diesel Generator Qualification Tests q P The schedule for qualification testing of the . i diesel generators, and the subsequent results of ) those tests, must be provided. Tbc tests shall l be in accordance with IEF.E 387 and Regulatory l Guide 1.9. (See Susbsection 8.3.1.1.8.9) 83.4.11 Defective Refurbished Circuit Breakers I NRC Bulliten No. 8810 and NRC Information Notice No. 88 46 idtstify problems with defective ; O refurbished circuit breakers. To ensure that k refurbished circuit b>:akers shall not be used in 0 safety related or nove safety related circuitry of the ABWR design, it is an interface requirement that new breakers be specified in the purchase specifications. 8.3.4.12 Minimum Starting Vlotages for Class IE Motors Provide the minimum required starting g voltages for Class 1E motors. Compare these g minimum required voltages to the voltages that will be supplied at the motor terminals during the starting transient when operating on offsite power and when operating on the edicsci generators. ; O Amendment 10 BSD 1
i 23A6100A0 Standard Plant Rev.n f Q-y( ; TABLE 83 3 l NOTES FOR TABLES 831 AND 83 2 (1) -: shows that the load is not connected to the switchgear of this disision. X: shows that the load is not counted for D/G continuous output claculation by the reasons shown on other notes. . 1 (2)
- Motor operated valves
- are operated only 30 60 seconds. Therefore they are not counted for the DG continuous output calculation.
l (3) LOADS are shed with LOCA signal. l (4) FM 'D operating time (about 2 minutes) is not counted for the DG cant .iuous i output I calcwation. l l 4%
$$ (5) ' Deleted (6) CUW pump will not operate under LOCA condition. CUW pump may operate under LOPP condition, but will not operate with SLC pump. On this calculation, the CUW pump is considered and the SLC pump is not since the CUW motor is the larger of the two.
-$ $ (7) Deleted O (8) FCS aill not operate under LOPP condition.
O
- l. gg (9)~ Deleted i AA (10) Deleted' (11) Div.1V battery charger is fed from Div. I motor control center.
(12) Load description aeronyms are interpeted as follws: Control Building HX Heat Exchanger C/B COMP. - Computer IA - Instrument Air CRD - Control Rod Drive MCR - Mair. Control Room CUW. - Clean Up Water MUWC Make Up Water System (condensed) CVCF - Constant Voltage Constant Frequency NPSS - Nuclear Protection S:.fety System gg Reactor Building
.nn. - DG - DieselGenerator R/B
- Reactor Cooling Water (buildmg)
FCS Flammability Control System RCW !- FPC - Fuel Pool Cooling RHR - Residual Heat Removal FMCRD Fine Motion Cotrol Rod Drive RSW Reactor Sea Water HECW - Emergency Cooling Water SBGT Standby Gas Treatment HPCF - High Pressure Core Hooder SLC - Standby Liquid Control l' O 8.3-28 Amendmcn to
h a m e - mu - Table 8.3-4 Cf1 B E-g D/G LOAD SEQUENCE DIAGRAM 15 MA,10R LOADS # 2 4 (Response to Question s 435.14 & 435.15) C3. i BtDCK9 ] 4 BIDCK 6 Bl#KK 7 BIDCE 8 ElDCK 3 BfDCK4 BIDCK$ y Blwk IllDCK 1 IllfKK 2 (55 SEC) (48 SFC) AFTTR 65 SEC (48 SEC) (45 SEC) ('eSEC) - Time (20 SFC) QSSEC) 05 SITC) ..t. M _. Mode Div. Chargers StrPump RIIR Pump RSW Pump RSW Pump TNCRD D/W Coohng Pan RCW1%mp RCW Pump IIECW Refrig Ct.PJ Pamp MOV MtJWC rump CVCFs Inst. Tr DG
- iVAC fifiCW i%mp MCR ifVAC R/D Fmer. IIVAC lir./A Erner. IfVAC 1A Compees. ITC Pump IOPP I SG'IS C/Il Omer. IIVAC SPC11 Pump - l Iighting Omrgers Sir Pomp RilR Peep RSW Pwmp RSW Pump 53dCRD D/W Cochng Fan RCW Pump RCW Pump IfECW Refrig CITN Pomp MOV MUWC Pump CVCFs 1IPCPPump* DGllVAC IICCW Pump MCR IfVAC R/fl Emer. liVAC Ilr/A liner. ITVAC IA C P ITCPemp VTIS C/II Emer. IIVAC InPP II inst Tr lighting Osvgers RilR Pomp RSW Pomp ItSW Pump PMCRD RCW Pump RCW Pump MOV SGTS CVCTs IIPCP romp
- DG IIVAC R/D Fws. IIVAC If r/A limer. IIVAC C/D Pmer. IIVAC
!DPP fil Inst. Tr 1ightir:g Oargers SifPump PCS RCW Pump RSWPump RSW Pump FMCRD MOV RitR Pump PCW Pump ITECW refrig ITC Pump MUWC Pump CVCFs IIFCW Pump MCR IIVAC R/71 Twr. 'fVAC lir/A F.mer. ITVAC SPCU Pome IDCA Inst. Tr DG IIVAC SG15 C/D Emer liVAC
& I lighting iOPP Chargers SLC Pump FCS RSWPump RSW Pump FMCRD RI!R Pwmp RCW Pump RCW Pump ItECW Referg FPC Pemp MOV MUWC rump CVCFs
!!PCF Pump
- DG IIVAC IIECW Pump MCR ilVAC R/fl Emer. !!VAC lis/A Fmer. lfVAC IDCA SGTS C/D fimer. IIVAC
& ff f ast. Tr .
IOPP 1 ightmg RSW Pump INCRD Oargers RCW Pump RCW Pump RSW Pump MOV RIIR Pump SGTS CVCFs 1tPCF Pump
- DG IiVAC R/D Pmer. IIVAC lin/A Tmer IIVAC 10C4 Inst. Tr C/IIl'wr fiVAC y
& til IAPPP lightmg M$
w
'r* Ibec* In case of the fa:Ture of RCIC pump startup Th 9 O O
ABWR m6 min niw. ti Simidard Plant CllAPTER 9 TAllLE OF CONTENTS Secilon Title IMes 9 AUXILIARY SYSTEMS 9.1 FUEL $TORAGE AND II ANI& LEG 9.1.1 New Fuct Storage 9.1 1 9.1.2 Spent Fuel Storage 9.12 9.13 Fuel Pool Cooling and Cleanup System 9.1 3 9.1 A Light Load llandli.ig System (Related to Refueling) 9.1-6 9.1.5 Overhead Ileavy lead llandling Systems 9.1 7 9.1.6 References 9.1 13 9.2 WATER SYSTEMS 9.2.1 Station Service Water System 9.2 1 9.2.2 Closed Cooling Water System 9.2 1 9.23 Demineralized Water Makeup System 9.2 1 9.2.4 Potable and Sanitary Water Systems 9.2- 1 9.2.5 Ultimate llcat Sink 9.2-1 9.2.6 Condensate Storage Facilitics and Distribution System 9.2-1 9.2.7 Plant Chilled Water Systems 9.2-1 9.2.8 hiakeup Water Systems (Preparation) 9.21 9.2.9 Makeup Water System (Condensate) 9.2-1 9.2.10 Makeup Water System (Purified) Distribution System 9.2-2 O 9 ..11 Amendment 6
ABWR mamn l Standard Plant wn CHAPTER 9 g TABLE OF CONTENTS (Continued) Section Thlt Eagt 9.2.11 Reactor Building Cooling Water System 9.2-3 9.2.12 HVAC Normal Cooling Water System 9.2-7 9.2.13 HVAC Emergency Cooling 9.2-8 9.2.14 Domestic Water Systems 9.2-11 9.2.15 Reactor Senice Water System 9.2-11 9.2.16 Turbine Senice Water System 9.2-12.1 9.2.17 Interfaces 9.2-13 93 PROCESS AUXIIIARIES 93.1 Compressed Air Systems 93-1 93.2 Process and Post Accident Sampling System 93-la 933 Equipment and Floor Drainage Systems 93-2 93.4 Chemical and Volume Control System (PWR) 93-2 93.5 Standby Liquid Control System 93-2 93.6 Instro nent Air System 93 7 93.7 Senice Air System 93 8 93.8 Radioactive Drain Transfer System 9 3-11 93.9 Hydrogen Water Chemistry System 9 3-12 9 3.10 Oxygen injection System 9 3-13 9.4 AIR CONDITIONING.IIEATING. COOL.ING AND VENTILATION SYSTEMS 9.4.1 Control Building HVAC 9.41 9.4.2 Spent Fuel Pool Area Ventilation System 9.4-2 9.43 Auxiliary Area Ventilation System 9.4-2 9.iii I Amendment 17
ABWR 2a nooxii:
=
Standard Plant - nev. n ( CHAPTER 9 v TABLE OF CONTENTS (Continued)
. Section Illlt East 9.4.4 Turbine Area Ventilation System 9.42 9.4.5 Reactor Building Ventiation System 9.4-2e 9,4-6 Radwaste Building IIVAC System 9.42j 9.4.7 Diesel Generator Area Ventilation System 9.4-3 9.4.8 Service Building Ver.tilation System 9.4-3 9.4.9 Drywell Cooling System 9.44 9.5 OTIIER AUXILI ARY SYSTEMS 9.5.1 Fire Protection Systems 9.5-1 9.5.2_ Communication Systems 952 9.5.3 Lighting and Servicing Power Supply Systems 9.53 l 9,5.4 Diesel Generator Fuel Oil Storage and Transfer System 954 l 4
9.5.5 Diesel Generator Cooling Water System 955 9.5.6 Diesel Generator Starting Air System 956 9.5.7 - Diesel Generator Lubrication System 9.5N 9.5.8 Diesel Generator Combustion Air intake and Exhaust System - 958 l 9.5.9 Suppression Pool Cleanup System 958 9.5.10 Motor-Generator Set 9.5-10 9.5.11 Combustion Turbine / Generator 9.5-10.1 9.5.12 Lower Drywell Flooder 95103 9.5.131 Interfaces 9510.5 9.5.14 References 9510.7 9-iv Amendment 17
ABWR m aman Standard Plant nev n ClihPTER 9 TABLE OF CONTENTS (Continued) h amo m . APPENDIX 9A FIRE IIAZARD ANALYSIS 9A.1 APPENDIX 9B
SUMMARY
OF ANALYSIS SUPPORTING FIRE PROTECTION DESIGN REQUIREMENTS 9B.1 O Amendment 17
ABWR mmo^n nev n Standard Plant SECTION 9.1 Q CONTENTS SectIon Title Eatt 9.1.1 New Fuel Storace 9.11 9.1.1.1 Design Bases 9.11 9.1.1.1.1 Nucleat Design 9.11 9.1.1.1.2 Storage Design 9.11 9.1.1.13 Mechanical and Structual Design 9.1 1 9.1.1.1.4 Thermalliydraulic Design 9.11 9.1.1.1.5 MaterialConsiderations 9.11 9.1.1.1.6 Dynamic Pnd impact Analysis 9.11 g E! 9.1.1.1.7 Deleted 9.1-1.1 9.1.1.2 Facilities D scription (New Fuel Storage) 9.1-1.1 9.1.13 Safety Evaluation 9.11.1 Criticality Control 9.1 1,1 9.1.13.1 9.1.13.2 Structural Design 9.11.1 9.1.133 Protection Features of the New Fuel Storage Facilities 9.11.2 9.1.2 Soent Fuel Stcrace 9.1-2 9.1.2.1 Design Bases 9.12 9.1.2.1.1 Nuclear Design 9.1-2 9.1.2.1.2 Storage Design 9.1 2 9.1.2.13 Mechanical and Structural Design 9.1-2 9.1.2.1.4 Thermal Hydraulic Design 9.1-2a O 9 .1 11 Amendment 16
ABWR- u.uiman Standard Plant nn n SECTION 9,1 CONTENTS (Continued) O Sectlon Title Page 9.1.2.1.5 Material Considerations 9.1 2c 9.1.2.2 Facilities Description (Spent Fuel Storage) 9.1-2d 9.1.23 Safety Evaluation 9.1 2d 9.1.23.1 Criticality Control 9.2 2d 9.1.23.2 Structural Design and Material Compatibility Requirements 9.2 2d 9.1.2.4 Summary of Radiological Considerations 9.12e 9.13 Fuel Pool Cooline and Cleanun S$ stem 9.13 9.13.1 Design Bases 9.13 9.13.2 System Description 9.13 9.133 Safety Evaluation 9.1 4 9.13.4 Inspection and Testing Requirements 9.1-5 9.13.5 Radiological Considerations 9.15 9.1.4 1icht lead llandline S5 stem (Hdated to Refuelingl 9.1-6 9.1.4.1 Design Bases 9.1-6 9.1.4.2 System Description 9.1-6 9.1.4.2.1 Spent Fuel Cask 9.1 6 9.1.4.2.2 Overhead Bridge Cranes 9.1-6a 9.1.4.2.2.1 Reactor Building Crane 9.1 6a 9.1.4.23 Fuel Servicing Equipment 9.1 6a 9.1-iii Amendment 17
A.WR B w,a m n nev n Standard Plant SECTION 9.1 LJ CONTENTS (Continued) Section Title Ike 9.1.4.2.10.2.5 Vessel Closure 9.1 6i 9.1.42.103 Departure of Fuel from Site 9.1-6j 9.1.43 Safety Evaluation of Fuelllandling System 9.1-6j 1 l 9.1.4.4 - Inspection and Testing Requirements 9.16k 1 9.1.4.4.1 Inspection 9.16k 9.1.4.4.2 Testing 9.16m 9.1.4.5 Instrumentation Requirements 9.1-6m 9.1.4.5.1 Automatie Refueling Machine 9.16m 9.1.4.5.2 Fuel Support Grapple 9.16m 9.1.4 L3 Other 9.16m 79
')
~
9.1.4.5.4 Radiation Monitoring 9.16m 9.1.5 Overhead lleavy load llandhne Systems 9.1-7 9.1.5.1 Design Bases 9.1-7 9.1.5.2 System Description 9.1-7 9.1.5.2.1 Reactor Building Crane 9.1-7 9.1.5.2.2 Other Overhead Load Handling System 9.1-7 9.1.5.2.2.1 Upper Drywell Senicing Equipment 9.1-7 9.1.5.2.2.2 Lower Drywell Senicing Equipment 9.1-8 9.1.5.2.23 Mainsteam Tunnel Senicing Equipment 9.1-9 9.1.5.2.2.4 Other Senicing Equipment 9.1-9 9.1.5.3 Applicable Design Criteria For All OHLH Equipment 9.1-10
/~ / I V 9.1-vi Amendment 17
ABWR 2miaan Standard Plant Rev. Il SECTION 9.1 CONTENTS (Continued) Section Title Page 9.1.5.4 Equipment Operating Procedures Maintenance and Service 9.1-10 9.1.5.5 Safety Evaluations 9.1-11 9.1.5.6 Inspections and Testing 9.1-11 9.1.5.7 Instrumentation Requirements 9.1-11 9.1.5.8 Operational Responsibilities 9.1-11 9.1.6 Interfaces 9.1-13 9.1.6.1 New Fuel Storage Racks Criticality Analysis 9.1-13 9.1.6.2 Dynamic and Impact Analyses of New Fuel Storage Racks 9.1-13 9.1.63 Spent Fuel Storage Racks Criticality Analysis 9.1 13 9.1.6.4 Spent Fuel Racks Load Drop Analysis 9.1-13 9.1.7 E.ererences 9.1-13 9.1-vil Amendment 17
ABWR n^M*^11 Standard Phud Rev it f l 9.1 FUEL STORAGE AND llANDLING (3) The biases between the calculated results and
\ experimental results, as well as the uncertainty 1 O[s The new-fuel storage vault stores a 40% core load involved in the calculations, are taken into j of new fuel assemblies. The fuelis stored in the new account as part of the calculational procedure to j fuel storage racks in the vault which are located as assure thet the specific k g limit is met.
close as practicable to the spent fuel storage pool work area to facilitate handling during fuel The new fuel storage racks are purchased preparation. The new-fuel inspection stand is close equipment. The purchase specification for these to the new-fuel storage vault to minimize fuel racks will require the vendor to provide the g transport distance. information requested in Question 430.180 on g criticality analysis and the inadver'ent placement of a
- Spent fuel removed from the reactor vessel must fuel assembly in other than prescribed locations. See be stored underwater while awaiting off site transfer. Subsection 9.1.6.1 for interface requirements.
Spent fuel storage racks, which are used for this purpose, are touted at the bottom of the fuel storage 9.1.1.1.2 Storage Design pool under sufficient water to provide radiological shielding. This pool water is processed through the The new fuel storage racks provided in the new fuel l fuel pool cooling and fuel pool and cleanup FPC storage vault provide storage for 40% of one full core system to provide cooling to the spent fuel in storage fuelload. and foi maintenance of fuel pool water quality. The spent fuel pool s:orage capacity is 270% of the 9.1.1.1.3 Mechanical and Structural Design reactor core. The new fuel storage racks contain storage space The new fuel and spent fuel storage racks are the for fuel assemblies (with channels) or bundles same high density design. The new fuel racks can be (without channels). They are designed to withstand used for either dry or submerged storage of fuel. all credible static and scismic loadings. The racks are p The design of the spent fuel racks will be described. designed to protect the fuel assemblies and bundles Q Information on the new fuel racks will only be presented when the design is different. The detailed from excessive physical damage which may cause the release of radioactive materials in excess of 10CFR20 analysis of the rack design is contained in Subsection and 10CFR100 requirements, under normal and 9.1.2. abnormal conditions caused by impacting from either fuel assemblies, bundles or other equipment. 9.1.1 New Fuel Storage The racks are categorized as Seismic Category 1. v.1.1.1 Design llases See Subsection 9.1.2.1.3 for additional discussion of design bases and analysis. 9.1.1.1.1 Nuclear Design 9.1.1.1.4 Thermal.llydraulic Design A full array of loaded new fuel racks is designed to be suberitical, by at least 5% t.k. See Subsection 9.1.2.1.4. (1) Monte Carlo techniques are employed in the 9.1.1.1.5 Material Considerations calculations performed to assure that gk does not exceed 0.95 under all normal and abnormal See Subsection 9.1.2.1.5. conditions. 9.1.1.1.6 Dynamic and Impact Analysis (2) The assumption is made that the storage array is infinite in all directions. Since no credit is The new fuel storage racks are purchased taken for neutron leakage, the values reported equipment. The purchase specification for the new , as effective neutron multiplication factors are, fuel storage racks will require the vendor to perform n in reality, infinite neutron multiplication confirmatory dynamic analyses. The input excitation i factors. for these analyses will utilize the horizontal and (n) w vertical response spectra provided in new Figures , Amendment 17 9 l-1
l ABWR m aman mn Sfandard Plant 9.115 and 9.126. (The SSE response is two times normal and abnortnal storage conditions equal to or E the OBE response). less than 0.95 in the new fuel storage racks. To { ensure design criteria are met, the following normal Verticalimpact analysis is required because the and abnormal new fuel storage conditions were fuel assembly is held in the storage rack by its own analyzed: e
$ weight without any mechanical holddown devices. $ Therefore, when the downward acceleration of the (1) normal positioning in the new fuel array, and storage rack exceeds ig, contact between the fuel assembly and the storage rack is lost. Horizontal (2) eccentric positioning in the new fuel array impact analysis is required because a clearance exists between the fuel assembly and the storage rack walls. The new fuel sprage area will accommodate fuel (kg < 135 at 20 C in standard core geometry) with See Subsection 9.1.6.2 for interface requirements. no safety implications.
9.1.13.2 Structural Design l 9.1.1.1.7 (Deleted) 9.1.1.2 Facilities Description (New Fuel (1) The new fuel vault contains one or more fuel Storage) storage racks which provides storage for fuel a maximum of 40% of one full core fuelload. (i) The location of the new fuel storage vault in the reactor building as shown in Section 1.2. (2) The new fuel storage racks are designed to be freestanding (i.e., no supports above the base). (2) The new fuel storage racks are top entry racks designed to preclude the possibility of criticality (3) The racks include individual solid tube storage under normal and abnormal conditions. The compartments which provide lateral restraints upper tieplate of the fuel element rests against over the entire length of the fuel assembly. the module to provide lateral support. The lower tieplate sits in the bottom of the rack, (4) The weight of the fuel assembly or bundle is which supports the weight of the fuel, supported axially by the rack lower support. (3) The rack arrangement is designed to prevent (5) The racks are fabricated from materials used for accidental insertion of fuel assernblies or construction are specified in accordance with the bundles between adjacent racks. The storage latest issue of applicable ASTM specifications. rack is designed to provide accessibility to the fuel bail for grappling purposes. (6) Lead in guides at the top of the storage spaces provide guidance of the fuel during insertion. (4) The floor of new fuel storage vault is sloped to a drain located at the low point. This drain (7) The racks are designed to withstand, while removes any water that may be accidentally and maintaining the nuclear safety design basis, the unknowingly introduced into the vault. The impact force generated by the vertical free-fall drain is part of the floor drain subsystem of the drop of a fuel assembly from a height of 6 feet. liquid radwaste system. (8) The rack is designed to withstand a pailup force (5) The radiation rnonitoring equiptnent for the of 1717 kg (4000 lb) and a horizontal force of new fuel storage areas is described in Section 454 kg (1000 lb). There are no readily definable 7.1. horizontal forces in excess of 1000 lb and, in the event a fuel assembly should jam, the maximum 9.1.13 Safety Evaluation lifting force of the fuel handling platform grapple (assumes limit switches fail) is 3000 lb. 9.1.13.1 Criticality Control (9) The new fuel storage racks require no periodic The design of the new fuel storage racks provides special testing or inspection for nuclear safety for an effective m . tiplication factor (kg) for both purposes. Amendment 16 9.1 1 t !
ABWR m-n sandard.nmt __ wn 9.L2 Spent Fuel Storage The spent fuel poolis a reinforced concrete / ) structure with a stainless steel liner. Tbc bottorn of V 9.1.2.1 Design Bases all pool gates are sufficiently high to maintain the water level over the spent fuel storage racks form 9.1.2.1.1 Nuclear Design adequate shielding and cooling. All pool fill and drain lines enter the pool above the safe shielding water (1) A full array in the loaded spent fuel rack is level. Redundant anti-siphon vacuum breakers are designed to be suberitical, by at least 5% ok. located at the high point of the pool circulation lines Neuron 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 jeopardizing the safe water leve!. calculated L including biases and l uncertainties, wilbn,ot exceed 0.95% under all The racks include individual solid tube storage normal and abnormal conditions. compartments, which provide lateral restraints over the entire length of the fuel assembly or bundle. The (a) Monte Carlo techciques are employed in weight of the fuel assembly or bundle is supported ) the calculations periemed to assure that axially by the rack fuel support. Lead-in guides at the k does not exceed 0.95 under all normal wp of the storage spaces provide guidance of the fuel ababnormal conditions. during insertion. (b) The assumption is made that the storage The racks are fabricated from materials used for array is infinite in all directions. Since no construction are specified in accordance with the credit is taken for neutron leakage, tiie latest issue of applicable ASTM specifications. The values reported as effective neutron racks are constructed in accordaace with a quality rnultiplication factors are, in reality, assurance program that ensures the design, inf'mite neutron multiplication factors. construction and testing requirements are met. n (c) The biases between the calculated results The racks are designed to withstand, while maintaining the nuclear safety design basis, the (*) and experimental results, as well as the uncertainty involved in the calculations, are impact force generated by the vertical free-fall drop taken into account as part of the of a fuel assembly from a height of 6 feet. The rack is calculational procedure to assure that the designed to withstarci a pullup force of 4WO pounds specific kg. limit is met, and a horizontal force of 1000 pounds. There are no readily definable horizontal forces in excess of 1000 9.1.2.1.2 Storage Design pounds, and in the event a fuel assembly should jam, the maximum lifting force of the fuelhandling The fuel storage racks provided in the spent fuel platform grapple (assumes limit switches fail) is 3000 storage pool provide storage for 270% of one full pounds. core fuelload. The fuel storage racks are designed to handle 9.1.2.1.3 Mechanical and Structural Design irradiated fuel assemblies. The expected radiation
!cvels are well below the design levels.
The spent fuel storage racks in the reactor building contain storage space for fuel assemblies (with in accordance with Regulatory Guide 1.29, the fuel channels) or bundles (without channels). They are storage racks are designated Safety Class 2 and l designed to withstand all credible static and seismic Seismic Category 1. The structuralintegrity of the loadings. The racks are designed to protect the fuel rack has becn 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 normat 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 hydrostatic loads; lQ,} Amendment 17 9.1-2
ABWR 23A6100All i Standard Plant nm n (2) live loads effect of lifting an empty rack The loads in the three orthogonal directions were dering installation; considered to be acting simultaneously and were combined using the SRSS rnethod suggested in (3) thermai loads - the uniform thermal expansion Regulatory Guide 1.92. The loads due to the OBE due to pool temperature changes; event are approximately 90% of those due to an SSE event, and allowable stress levels for OBE are 50% of (4) seismic forces of OBE and SSE; SSE, therefore making the OBE event the limiting load condition except for stability, where SSE (5) accidental drop of fuel assembly from acceptance criteria of 67% of critical buckling maximum possible height 6 feet above rack; .trength is limiting. and Under fuel drop loading conditions, the acceptance (6) postulated stuck fuel assembly causing an criterion is that, although deformation may occur, upward force of 3000 pounds. K must g remain <0.95%. The rack is designed such th*aT, should the drop of a fuel assembly damage the The loed combinations considered in the rack tubes and dislodge a plate of poison material, the Kg design are: would still be <0.95 as required. (1) live loads The effect of the gap between the fuel and the storage tube has been taken into account on a local (2) dead loads plus OBE cffect b.ais. Dynamic response analysis shows that the fuel cordacts the tube over a large portion 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 verticalimpact imd of the fuel onto its seat has Thermal loads were not included in the above been considered conservatively as being slowly combinatiota because they were negligible due to the applied without any benefit ter strain rate effects. 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-llydraulle Desin temperature changes). The fuel storage rack is designed to provide The loads experienced under a stuck fuel assembly sufficient natural convection coolant flow to remove condition are less than those calculated for the 68,000 Btu /br/ bundle of decay heat. seismic conditions and, therefore, have not been included as a load combination. The support structure must be designed to provide an adequate flow rate to pgevent water reaching The storage racks are attached to the support excessive temperatures 2?2 F. The flow rate is structure by bolting, sufficient to counteract the dependent on the decay heat load, the AP losses tendency to overturn from horizontal loads and to lift tirough the structure and the losses through the rack from vertical loadr.. The analysis of the rack and bundle. assumed an adequate supporting structure, and loads were generated accordingly. In the spent fuel storage pool, the bundle decay beat is removed by recirculation flow to the fuel pool Stress analyses were performed by classical cooling heat exchanger to maintain the pool tempera-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 bundic flow which carries away the decay calculated at critical sections of th:: 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 III subsection NF. Compressive temperature have been evaluated. stability v'as calculated according to the AISI code for light gage structures. The parameters which will affect the water flow Amendment 17 9.1-2a
ABWR mcun nn n Standard Plant compatible with the environment of treated water above the base), the support structure also O and prosides a design life of 60 years. provides the required dynamic stability. 9.1.2.2 Facilities Description (Spent Fuel (3) The racks include individual solid tube storage Storage) compartments, which provide lateral restraints over the entire length of the fuel assembly or (1) The spent fuel storage racks provide storage in bundle. the reactor building spent fuel pool for spent fuel received from the reactor vessel during the (4) The racks are fabricated from materials used for refueling operation. The spent fuel storage construction and are specified in accordance racks are top entry tacks designed to preclude with the latest issue of applicable ASTM the possibility of criticality under normal and specifications at the time of equipment order. abnormal conditions. The upper tieplate of the fuel elements rests against the rack to provide (5) Lead.in guides at the top of the storage spaces lateral support. Tne lower tieplate sits in the provide guidance of the fuel during insertion. bottom of the rack, which supports the weight of the fuel. (6) The racks are designed to withstand, while maintaining the nuclear safety design basis, the (2) The rack arrangement is designed to prevent impact force generated by the vertical free-fall accidental insertion of fuel assemblic; or drop of a fuel assembly from a height of 6 feet. bundles between adjacent modules. The storage rack is designed to provide accessibility (7) The rack is designed to withstand a pullup force to the fuel bail for grappling purposes. of 4000 lb and a horizontal force of 1000 lb. There are no readily definable horizontal forces (3) The location of the spent fuel pool is shown in in excess of 1000 lb and in the event a fuel Section 1.2 assembly should jam, the maximum lifting force of the fuel handling platform grapple (assumes O 9.1.23 Safety Evaluation limit switches fail) is 3000 lb. (8) The fuel storage racks are designed to handle 9.1.23.1 Criticality Control iriadiated fuel assemblies. The expected radiation levels are well below the design levels. The spent fuel storage racks are purchased equipment. The purchase specification for the spent The fuel storage facilities will be designed to fuel storage racks will require the vendor to provide Seismic Category I requirements to prevent g the information requested in Question 430.100 om earthquake damage to the stored fuel. g criticality analysis of the spent fuel storage including the uncertainity value and associated probability and The fuel storage pools have adequate water confidence level for the Kgvalue. See Subsection shielding for the stored spent fuel. Adequate 9.1.63 for interface requirements. shielding for transporting the fuel is also provided. Liquid level sensors are installed to detect a low pool 9.1.23.2 Structural Design and Material water level, and adequate makeup water is available Compatibility Requirements to assure that the fuel will not be uncovered should a leak occur. (1) The spent fuel pool racks provide storage for 270% of the reactor core. Since the fuel storage racks are made of noncombustible material and are stored under water, (2) The fuel storage racks are designed to be there is no potential fire hazard. The large water supported above the pool floor by a support volume also protects the spent fuel storage racks from structure. The support structure allows potential pipe breaks and associated jet impingement sufficient pool water flow for natural con. loads. vection cooling of the stored fuel. Since the O modules are freestanding (i.e., no supports Fuel storage racks are made in accordance with the latest issue of the applicable ASTM specification at Amendinent 16 912d
ABWR uumn Standani Plant am n the time of equipment order. The storage tubes are permanently marked with identification traceable to the material certifications. The fuel storage tube assembly is compatible with the environment of treated water and provides a design life of 60 years, including allowances for corrosion. Regulatory Guide 1.13 is applicable to spent fuel l storage facilities. The reactor building contains the fuel storage facilities, including the storage racks and pool, is designed to protect the fuel from damage caused by: (1) natural events such as earthquake, high winds and flooding, and (2) mechanical damage caused by dropping of fuel assemblies bundles, or other objects onto stored fuel. 9.1.2.4 Summary of Radiological Considerations By adequate design and careful operational procedures, the safety design bases of the spent fuel storage arrangement are satisfied. Thus, the exposure of plant personnel to radiation is maintained well below published guideline values. Further details of radiological considerations, including those for the spent fuel storage arrangement, are presented in Chapter 12. The poolliner leakage detection system and water _ level monitoring system are discussed in Subsection 2 9 h c.13. The apability is incorrective Subsection 9.1.3. action for loss of heat removal The radiation monitoring system and the corrective action for excessive radiation levels are discussed in Subsections 11.5.2.1.2.1 and 11.5.2.13. O Amendment 17 9,12e
ABWR m am^n nry n Standard Plant 9.1.3 Fuel Pool Cooling and Cleanup The FPC system cools the fuel storage pool by transferring the spent fuel decay heat through f) v System two 6.55 x 106 Blu/hr heat exchangers to the 9.1.3.1 Design Bases reactor building closed cooling water system (RCW). Each of the two heat exchangers is de-signed to transfer one half the system design heat load. The system utilizes two parallel 250 3 The fuel pool cocling and cleanup (FPC) system m /hr gumps to provide a system design flow of shall be designed to remove the decay heat from 500 m- /hr. Each pump is suitable for the fuel pool, maintain pool water level and continuous duty operation. The equipment is quality and remove radioactive materials from the located in the reactor building. pool to minimize the release of radioactivity to the environs. The system pool water temperature is main-tained at or below 125 F. The decay heat The FPC system shalh released from the stored fuel is transferred to the RCW system. The residual heat removal (RilR) (1) minimize corrosion product buildup and shall system can supplement the FPC system to remove control water clarity, so that the fuel the additional heat generated should the reacto* assemblies can be efficiently handled under- be defueled beyond the design basis 35% batch. water; Fuel storage pool water is circulated by (2) minimize fission product concentration in means of overflow through skimmers around the the water which could be released from the periphery of the pool and a scupper at the end pool to the reactor building environment; of the transfer pool. The overflow is collected in the fuel pool drain tanks and the flow passes (3) monitor fuel pool water level and maintain a through the heat exchangers and filter deminera-p water level above the fuel sufficient to lizers and back to the pool through the y proside shielding for normal building occu- dif f use r s. pancy; Clarity and purity of the pool water are (4) maintain the pool water temperature below maintained by a combination of filtering and ion 125 F under normal operating condi- exchange. The filter-demineralizers maintain tions. The temperature limit of 125 F is set to establish an acceptable environ- totalwith pilcorrosion range of 5.6product metals to 8.6 at 250F at 30 ppb for compat- 7 N or less l,, ment for personnel working in the vicinity ibility with fuel storage racks and other equip-of the fuel pool. The design basis normal ments. Conductivity is maintained at less than heat load from spent fuel stored in the pool 1.2 pE/cm at 25"C and chlorides less than :: is the sum of decay heat of the most recent 20 ppb. Each filter unit in the filter-demi- ,g 35% batch plus the heat from the previous 4 neralizer subsystem has adequate capacity to fuel batches. The RHR system will be used maintain the desired purity level of the pools to supplement the FPC system under the under normal operating conditions. The flow maximum load condition as defined in rate is designed to be approximately that Subsection 9,1.3.2. required for two complete water changes per day for the fuel transfer and storage pools. The 9.13.2 System Description maximum system flow rate is twice that needed to maintain the specified water quality. The FPC system (Figures 9.1-1 a and b, and 9.1-2) maintains the spent fuel storage pool The FPC system is designed to remove below the desired temperature at an acceptable suspended or dissolved impuritics from the radiation level and at a degree of clarity following sources: necessary to transfer and service the fuel bundles. (1) dust or other airborne particles; Amendment 17 9.1 -3
l ABWR mmm Standard Plant nix. n (2) surface dirt dislodged from equipment control room and a local panel. Pump low suc-Immersed in the pool; tion pressure automatically turns off the pumps. A pump low discharge pressure alarm is (3) crud and fission products emanating from the indicated in the control room and on the local reactor or fuel bundles during refueling; panel. The circulating pump motors can be powered from the diesel generators if normal (4) debris from inspection or iisposal opera- power is not available. Circulating pump motor tions; and loads are considered nonessentialloads and will be operated as required under accident (5) residual cleaning chemicals or thsh water, conditions. A post-strainer in the effluent stream of the The water level in the spent fuel storage filter demineralizer limits the migration of pool is maintained at a height which is suffi-filter material. The filter holding element can cient to provide shielding for normal building withstand a differential pressure greater than occupancy. Radioactive particulates removed the developed pump head for the system. from the fuel pool are collected in filter de-mineralizer units which are located in shielded The filter-demineralizer units are located cells. For these reasons, the exposure of plant separately in shielded cells with enough clear- personnel to radiation from the FPC system is ance to permit removing filter elements from the minimal. Further details of radiological vessels. considerations for this system are described in Chapter 12. Each cell contains only the filter-deminera-lizer and piping. All valves (inlet, outlet, The circulation patterns within the reactor recycle, vent, drain, etc.) are located on the well and spent fuel storage pool are established outside of one shielding wall of the room, by placing the diffusers and skimmers so that together with necessary piping and headers, particles dislodged during refueling operations instrument elements and controls. Penetrations are swept away from the work area and out of the through shielding walls are located so as not to pools. compromise radiation shielding requirements. Check valves prevent the pool from siphoning The filter demineralizers are controlled from in the event of a pipe rupture. a local panel. A differential pressure and conductivity instruments provided for each Heat from pool evaporation is handled by the filter demineralizer unit indicate when backwash building ventilation systerr- Makeup water is is required. Suitable alarms, differential provided through a remou-operated valve. pressure indi:ators and flow indicators monitor the condition of the filter demineralizers. 9.133 Safety Evaluation System instrumentation is provided for both The maximum possible heat load is the decay automatic and remote-manual operations. A low- heat of the full core load of fuel at the end of low level switch stops the circulating pumps when the fuel cycle plus the remaining decay heat of the fuel pool drain tank reserve capacity is the spent fuel discharged at previous refuel-reduced to the volume that can be pumped in ings; the maximum capacity of the speni fuel l approximately one minute with one pump at rated storage pool is 270% of a core, The temperature capacity (250 m 3/hr). A level switch is of the fuel pool water may be permitted to rise provided in the fuel pool to alarm on high and to approximately 140 F under these condi I low level. A temperature element is provided to tions. During cold shutdown conditions, if if display room. pool temperature in the main control exceed appears that 125 F, the the fuel operator canpool temperature connect the will l FPC system to the RHR system. Combining the ca. The circulating pumps are controlled from the pacities enables the two systems to keep the O Amendment 16 9.14
ABWR m-n Standard Plant wvn water temperature below 125 F. The RllR system will be used only to supplement the fuel A makeup water system and pool water level pool cooling when the reactor is shut down. The instrumentation are provided to replace reactor will not be started up whenever portions evaporative and leakage losses. Makeup water l of the RHR system are needed to cool the fuel during normal operation will be supplied from pool. The connecting piping from the fuel condensate. The suppression pool cleanup system storage pool to the RilR system is designed can be used as a source cf makeup water in case Seismic Category I and can be isolated, assuming of failure of the normal makeup water system, a single active failure, from the remainder of the fuel pool system. Connections from the RilR system to the FPC system provide a Seismic Category I , These connections may also be utilized during safety related makeup capability to the spent emergency conditions to assure cooling of the fuel pool. The FPC system from the RilR spent fuel regardless of the availability of the connections to the spent fuel pool are Seismic fuel pool cooling system, The volume of water in Category 1, safety-related, the storage pool is such that there is enough heat absorption capability to allow sufficient From the foregoing analysis, it is concluded time for switening over to the RilR system for that the FPC system meets its design bases. emergency cooling. The 140 F temperature limit is set to assure that the fuel building environment does No special tests are required because, not exceed equipment environmental limits, normally, one pump, one heat exchanger and one filter-demineralizer are operating while fuel is The spent fuel storage pool is designed so stored in the pool. The spare unit is operated that no single failure of structures or equipment periodically to handle abnormal heat loads or to will cause inability to: (1) maintain irradiated replace a unit for servicing. Routine visual fuel submerged in water; (2) re establish normal inspection of the system components,instrumen-fuel pool water level; or (3) remove decay heat tation and trouble alarms is adequate to verify from the pool. In order to limit the possibility system operability, of pool leakage arouai pool penetrations, the pool is lined with stainless steel, in addition 9.t.3.5 Radlological Considerntkas to providing a high degree of integrity, the lining is designed to withstand abuse that might The water level in the spent fuel storage occur when equipment is moved about. No inlets, pool is maintained at a height which is suffi-outlets or drains are provided that might permit cient to provide shielding for normal building the pool to be drained Eclow a safe shielding occupancy. Radioactive part iculates removed level. Lines extending below this level are from the fuel pool are cellected in filter-equipped with siphon breakers, check valves, or demineralizer units which are located in other suitable devices to prevent inadvertent shicided cells. For these reasons, the exposure pool drainage. Interconnected drainage paths are of plant personnel to radiation from the FPC provided behind the liner welds. These paths are system is minima 1. Further details of designed to: (1) prevent pressure buildup behind radiological consi- derations for this and other the liner plate; (2) prevent the uncontrolled systems are described in Chapters 11,12, and less of contaminated pool water to other rela- 15. tively cleaner locations within the containment or fuel-handling area; and (3) provide liner leak detection and measurement. These drainage paths are designed to permit free gravity drainage or pumping to the equipment drain tank. Amendment 17 915
AB R nasim4>i Standard Plant nev n 9.1.5 Overhead Heavy Imad Handling 9.13.2 System Description Os Systems (OHLH)
- 9.1.5.2.1 Reactor Building Crant 9.13.1 Design Bases The reactor building (RB) is a reinforecd The equipment covered by this subsection concrete structure which encloses the reinforced handle items considered as heavy loads that are concrete containment vessel, the refueling floor, new handled under conditions that mandate critical fuel storage vault, the storage pools for spent fuel and handling compliance, the dryer and separator and other equipment.' The -
reactor building crane provides heavy load lifting Critical load handling conditions include loads, capability for the refueling floor. The main book (150 equipment, and operations, which if inadvertent ton capacity) will be used to lift the concrete shield operations or equipment malfunctions either blocks, drywell head, reactor pressure vessel (RPV) separately or in combination, could cause; (1) a head insulation, RPV head, dryer, separator strong release of radioactivity, (2) a criticality accident, (3) back, RPV head strongback carousel, new fuel the inability to cool fuel within reactor vess :1 or shipping containers, and spent fuel shipping cask, spent fuel pool or (4) prevent safe shutdon of the The orderly placement and movement paths of these reactor. This includes risk assessments to spent fuel components by the reactor building crane precludes and storage pool water levels, cooling of fuel pool transport of these heavy loads over the spent fuel water, new fuel criticality,: This includes all storage pool or over the new fuel storage vault. components and equipment used in moving any load weighing more than one fuel assembly including the The RB cranc wilI be used during weight of its associated handling devices (i.e., one refueling / servicing as well as when the plant is online. ton). During refueling / servicing, the crane handles the shield plugs, drywell and reactor vessel heads, steam The reactor building crane as designed shall dryer and separators, etc. (see Table 9.1-7). provide a safe and effective means for transporting Minimum crane coverage include RB refueling floor heavy loads including the handling of new and spent laydown areas, and RB equipment storage ' pit. , fuel, plant equipment and service tools. Safe During normal plant operation the crane will be used handling includes design considerations for to handle new fuel shipping containers and the spent maintaining occupational radiation exposure as low fuel shipping casks. Minimum crane coverage must as practicable during transportation and handling. include the new fuel vault, the RB equipment hatches, and the spent fuel cask loading and washdown pits, A Where applicable, the appropriate seismic description of the refueling procedure can be found in category, safety class quality group,' ASME, ANSI,- Section 9.1.4. industrial and electrical codes have been identified (see Tables 3.21 and 9.16). The designs will The RB crane will be interlocked to prevent conform to the relevant requirements of General movement of heavy loads over the spent fuel storage Design Criterion 2,4 and 61 of 10CFR Part 50, portion of the spent fue etorage pool. Since the Appendix A. crane'is used for handling large heavy objects over the open reactor the crane is of type 1 design. The The lifting capacity of each crane or hoist is reactor buHding crane shall be designed to meet the designed to at least the maximum actual or single.faliure-proof requirements of NUREG-0054.
- anticipated weight of equipment and handling devices in a given area serviced. The hoists, cranes, 9.1.5.2.2 Other Overhead Load Ilandling System
~or other lifting devices shall comply with the requirements of ANSI N14.6, ANSI B30.9, ANSI 9.1.5.2.2.1 Upper Drywell Servicing Equipment B30.10 and NUREG-0612 Subsection 5.1.1(4) or 5.1.1(5). Cranes and hoists are also designed to The upper drywell arrangement provides criteria and guidelines of NUREG 0612 Subsection servicing access for the main steam isolation valves 5.1.1(7), ANSI B30.2 and CMAA-70 specifications (MSIVs), feedwater isolation valves, safety relief for electrical overhead traveling cranes, including valves (SRVs), emergency core cooling systems ANSI B30.11, ANSI B30.16, and NUREG-0554 as (ECCSs) isolation valves, and drywell cooling coils, applicable.
Amendment 17 91-7
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ABWR 3ami Standard Plant mm fans and motors. Access to the space is sia the RB drywell. Special hoists are provided in the lower through either the upper drywell personnel lock or drywell and reactor building to facilitate handling of equipment hatch. All equipment is removed through these loads, the upper drywell equipment hatch. Platforms are provided for servicing the feedwater and mainsteam (1) Reactor Internal Pump Senicing isolation valves, safety relief salves, and drywell cooling equipment with the object of reducing There are 10 RIPS and their supporting maintenance time and operator exposure. The instrumentation and heat exchangers in the L/D MSIVs, SRVs, and feedwater isolation valves all that require senicing. The facilities provided weigh in excess of 200 kg. Thus they are considered for senicing the RIPS include: heavy loads. (a) L/D equipment platform with facuities to With maintenance activity only being rotate the motor from vertical to horizontal performed during a refueling outage, only safe e.nd place it on a cart for direct pull out to shutdown ECCS piping and valves need be protected the RB. The equipment platform rotates to from any inadvertent load drops. Since only one facilitate alignment with the installed pump division of ECCS is required to maintain the safe locations. shutdown condition and the ECCS divisions are spatially separated, an inadvertent load drop that (b) Attachment points for rigging the RIP heat breaks more than one division of ECCS is not exchanger into place. The RIP heat credible. In addition, two levels of piping support exchanger can be lowered straight down to structures and equipment platforms separate and the equipment platform, shield the ECCS piping from heavy loads transport path. (c) Access to the RIP equipment platform is via stairs. There is a ladder access to the This protection is adequate such that no RIP heat exchanger maintenance platform. credible load drop can cause either (1) a release of radioactivity, (2) a criticality accident, or (3) the (d) The L/D equipment tunnel and hatch are inability to cool fuel within reactor vessel or spent utilized to remove the RIP motors from the fuel pool; therefore, the upper drywell servicing lower drywell, equipment is not subj-ct to the requirements of Subsection 9.1.5. (c) The RIP motor servicing aien is directly outside the L/D equipment batch. 9.1.5.2.2.2 Lower Drywell Servicing Equipment The 10 RIPS have wet induction motors in The lower drywell (L/D) arrangement provides housings which protrude into the lower drywell for senicing, handling and transportation operations from the RPV bottom head. These are in a for RIP, and FMCRD. The lower dryweli OHLHS circle at a radius of 3162.5 mm from the RPV consists of a rotating equipment service platform, centerline. For senice, the motor is removed chain hoists, FMCRD removal machine, a RIP from below and outside, whereas the diffuser, removal machine, and other special purpose tools. impeller and shaft are removed from above and inside the RPV. The rotating equipment platform provides a work surface under the reactor vessel to support the The motor, with its lower flange attached, weight of personnel, tools, and equipment and to weighs approximately 3300 kg, is 830 mm in facilitate transportation moves and heavy load diameter and 1925 mm high. The flange has handling operations. The platform rotates 360 in
- ears" that extend from two sides,180 apart.
cither direction from its stored or " idle" position. These ears, which are used to handle the motor, The platform is designed to accommodate the increase the flange diameter to 1200 mm for a maximum weight of the accumulation of tools and width of 270 mm. equipment plus a maximum sized crew. Weights of . tools and equipment are specified in the interface The motor, suspended from jack screws, is control drawings for the equipment used in the lower lowered straight down out of its housing onto Amendment 17 9. t.8
ABWR umman Standard Plant wn (a) The PhtCRD drive servicing machine has p) ( the equipment platform. The motor is then moved, circumferentially and lifted onto a rail its on mechanisms for rotating and raising FhtCRD drive assemblics from a carrier on mounted transport cart for direct removal through the equipment rem 'alL/D the equipment platform to their installed equipment tunnel and hatch. T. e motor is position. This senicing machine interfaces transported horizontally out of the containment with the L/D equipment platform, which and into the motor service shop immediately permits positioning the servicing machine adjacent to the L/D equipment hatch, under any of the 205 PhtCRDs. The RIP senicing equipment includes the cart (b) A separate machine and cart are provided to transport the motor from the senice area for servicing FhiCRD motors and seal through the equipment hatch to the L/D assemblics and transporting them to the equipment platform. The interface for this service shop located immediately outside equipment is the rails on the equipment the L/D equipment hatch, platform that permit locating the motor below its nozzle on the RPV. The servicing There is no safety related equipment below equipment includes a chain hoist for rotating either component. Inadvertent load drops by the RIP motor from horizontal to vertical and a the FhiCRD senicing equipment can not cause hydraulic lift to raise it from the equipment either (1) a release of radioactivity, (2) a platform to its installed position below the criticality accident, or (3) the inability to cool RPV. Facilities are provided for handling stud fuel within the reactor vessel or spent fuel pool; tensioners, blind flanges, other tools, drains and therefore, the Ph1CRD servicing equipment is vents used in RIP senicing. not subject to the requirements of Subsection 9.1.5. Senicing of the RIP heat exchanger, such as removal of the tube bundie, will be 9.1.5.2.2.3 hlainsteam Tunnel Senicing Equipment O V accomplished by rigging to attachment points on the RPV pedestal and structural steel in the The mainsteam tunnelis a reinforced concrete area. A direct vertical removal path is provided structure that surrounds the mainsteam lines and from the heat exchanger installed position to feedwater lines. The safety-related valve area of the the equipment platform. The operation is mainsteam tunnel is located inside the reactor performed by a chain hoist. This is considered building. Access to the mainsteam tunnelis during a to be a nonroutine senicing operation. refueling / servicing outage. At this time htSIVs or feedwater isolation valves and/or feedwater check These RIPS are seniced only when the reactor valves may be removed using permanent overhead is in a safe shutdown mode. In addition, there monorail type hoists. They are transported by is no safety related equipment below either the monorail out of the steam tunnel and placed on the RIPS or the RIP heat exchangers. Inadvertent floor below a ceiling removal hatch. Valves are thei. load drops of either component can not cause lifted through the ceiling hatch by valve senice shop cither (1) a release of radioactivity,(2) a monorail. During shutdown, all of the piping and criticality accident, or (3) the inability to cool valves are not required to operate. Any load drop can fuel within reactor vessel or spent fuel pool; only damage the other valves or piping within the therefore, the RIP servicing equipment is not mainsteam tunnel. Inadvertent load drops by the subject to the requirements of Subsection 9.1.5. mainsteam tunnel senicing equipment can not cause either (1) a release of radioactivity, (2) a criticality (2) Fine hiotion Control Rod Drive accident, or (3) the inability to cool fuel within reactor vessel or spent fuel pool; therefore, the mainsteam There are 205 Fh1CRDs in the L/D that tunnel servicing equipment is not subject to the require servicing. There are two types of requirements of Subsection 9.1.5. servicing operations: replacement of the FN1CRD drive mechanism and motor and seal 9.1.5.2.2.4 Other Senicing Equipment replacement. Separate servicing equipment is Outside the containment, the mainsteam tunnel, (A) provided for each of these operations. and the refueling floor no safety-related component 9.1 -9 Amendment 17
b\ 23A6100All Stan Jard Plant _ mn of one division shall be routes any portion of a 11eavy load equipment is also used to handle safety-related portion of anot .ivision. A load light loads and related fuel handling tasks. Therefore, drop accident in one division causing the complete much of the handling systems and related design, loss of a second division is not credible. Hence, descriptions, operations, and senice task information inadvertent load drops can not cause either (1) a of Subsection 9.1.4 is applicable here. The cross release of radioactivity, (2) a criticality accident, (3) referenee between the handling operat-the inability to cool fuel within reactor vessel or ions / equipment and Subsection 9.1.4 is prosided in spent fuel pool, or (4) prevent the safe shutdown of Table 9.17. See Table 9.1-8 for a summary of heavy the reactor; therefore, all servicing equipment load operation. located outside the containment, the mainsteam tunnel, or the refueling floor are not subject to the Transportation routing drawings will be made requirements of Subsection 9.1.1 covering the transportation route of every piece of heavy load removable equipment from its installed location to tht appropriate service shop or building 9.1.5.3 Applicable Design Criteria For All OllLil exit. Routes will be arranged to prevent congestion Equipment and to assure safety while permitting a free flow of equipment being seniced. The frequency of trans. All handling equipment subject to heavy loads portation and usage of route will be documented handling criteria will have ratings consistent with lifts based on the predicted number of times usage either required and the design loading will be visibly per year and/or per refueling or service outage, marked. Cranes / hoists or monorail hoists will pass over the centers of gravity of heavy equipment that is Safe load paths / routing will comply with the to be lifted. In locations where a single monorail or requirernents of NUREG-0612, Subsection 11.1(1), crane handles several pieces of equipment, the routing shall be such that each transported piece will 9.1.5A Equipment Operating Procedures pass clear of other parts if, however, due to Maintenance and Service restricted overhead space the transported load cannot clear the installed equipment, then the Each item of equipment requiring servicing will monorail may be offset to provide transport be described on an interface control diagram (ICD) clearance. A lifting eye offset in the ceiling over delineating the space around the equipment required each piece of equipment can be used to provide a for servicing. This willinclude pull space far internal Y-lift so that the load can be lifted upward until free parts, access for tools, handling equipment, and and then swur.g to position under the monorail for alignment requirements. The ICD will specify the transport, weights of large removable parts, show the location of their centers of gravity, and describe installed lifting Pendant controlis required for the bridge, accommodations such as eyes and trunions. An trolley and the auxiliary hoist to provide efficient instruction manual will describe maintenance handling of fuel shipping containers during receipt procedures for each piece of equipment to be handled and also to handle fuel during new fuel inspection. for servicing. Eash manual will contain suggestions The crane control system will be selected considering for rigging and lifting of heavy parts and identify any the long lift required through the equipment hatch as special lifting or handling tools required. well as the precise positioning requirements when handling the RPV and drywell heads, RPV internals, All major handling equipment components: and the RPV bead stud tensioner assembly. The cranes, hoist, etc., will be provided with an operating centrol system will provide stepless regulated instruction and maintenance manual for reference variable speed capability with high empty-hook and utilization by operations personnel.11andling speeds. Efficient handlings of the drywell and RPV equipment operating procedure will comply with the heads and stud tensioner assembly require that the requirements of NUREG-0612, Subsection 5.1.1(2). control system provide spotting control. Since fuel shipping cask handling involves a long duration lift, The operational programs for maintenance and low speed and spotting control, thermal protection senicing are described in Subsec ion 9.1.16. features will be incos porated. Amendment 17 9.1 10
rABWR uumn
~ Standard Plant ,
nev. n ly] 9.1JJ Safety Evaluations
. The cranes, hoists, and related lifting devices tested again to ensure the electrical and/or mechanical functions are operational including visual and,if required, NDE inspection.
l l used for handling heavy loads either satisfy the single failure guidelines of NUREG 0612, Crane inspections and testing will cornply with Subsection 5.1.6,' including NUREG 0554 or the requirements of ANSI D30.2 and NUREG-0612, evaluations are made to demonstrate compliance Subsection 5.1.1(6). with the recommended guidelines of Section 5.1, including Subsection 5.1A and 5.1.5. 9.1.5.7 Instrumentation Requirements i The equipment handling components over the The majority of the heavy load handling ' fuel pool are designed to meet the single failure equipment is manually operated and controlled by the - proof criteria to satisfy NUREG-0554. Redundant. operator's visual observations. This type of operation . ; safety interlocks and limit switches are provided to does not necessitate the need for a dynamic - prevent transporting heavy loads other than spent instrumentation system. fuel by the refueling bridge crane over any spent fuel that is stored in the spent fuel storage pool. Load cells may be installed to provide automatic
- shutdown whenever threshold limits are exceeded for L A transportation routing study will be made of critical load handling operations to prevent
!' all planned heavy load handling moves to evaluate overloading. and minimize safety risks. 9.1.5.8 Operational Responsibilities
. Safety evaluation of related light loads and refueling handling tasks in which heavy load Critical heavy load handling in operation of the equipment is also used are covered in Subsection - plant shallinclude the following documented program 9.1.4.3. foe safe administration and safe implementation of
'- operations and control of heavy load handling N - 9.1.5.6 Inspection at.d Testing systems: L licavy load handling equipment is subject to (1) IIcavy Load Ilandling System and Equipment l the strict controls of Quality Assurance (OA), Operating Procedures ! -incorporating the requirements of Federal Regulation 10CFR$0, Appendix B.- Components (2) Heavy Load Handling Equipment Maintenance , defined as essential to safety have an additional set Procedures and/or Manuals l of engineering specified " Quality Requirements" that
- identify safety-related features which require specific (3) lleavy Load 11andling Equipment inspection j OA verification of compliance to drawing /specifi- and Test Plans; NDE, Visual, etc.
t cation requirements. L . (4) Heavy Load liandling Safe Load Paths and h ' Prior to shipment, every lifting equipment Routing Plans component requiring inspection will be reviewed by OA for compliance and that the required records are (5) OA Program to Monitor and Assure l- available. Qualification load and performance Implementation and Compliance of IIcavy Load l testing, including nondestructive examination (NDE) - Handling Operations and Controls I and dimensional inspection on heavy load handling egripmeat will be performed prior to OA- (6) Operator Qualifications, Training and Control acceptance. Tests may include load capacity, safety Program overloads, life cycle, sequence of operations and functional areas. When equipment is received at the site it will be inspected to ensure no damage has occurred O during transit or storabe, Prior to use and at periodic intervals each piece of squipment will be Amendment 17 9.1 J
- , . .~. , -,
ABWR :u61*an Standard Plant nu n G. (This pge has intentionally been left blank) l 9 L l O I Amendment 17 9 l'I2
-xABWR --
23463oorn Standard Plant nu n g. 1 9.1.6 Interfaces
- 9.1.6.1 New Fuel Storage Racks Criticality Analysis
. The applicant referencing the ABWR design shall provide the NRC confirmatory criticality
- analysis as required by Subsection 9.1.1.1.1.
. 9.1.6.2 Dynamic and Impact Analyses of New Fuel Storage Racks The applicant referencing the ABWR design shall provide the NRC confirmatory dynamic and in"uct analyses of the new fuel storage racks. See Subsection 9.1.1.1.6.-
9.1.6.3 Spent Fuel Storage Racks Criticality Analysis
=
The applicant referencing the ABWR design shall provide the NRC confirmatory critically e analysis as required by Subsection 9.1.23.1. 9.1.6.4 Spent Fuel Racks Load Drop Analysis - !
,The applicant referencing the ABWR design '
. shall provide the NRC confirmatory load drop analysis as required by Subsection 9.1.43, 9.13 References -
1.~ . General Electric Standard Application for. Reactor Fuel, (NEDE 24011-P-A, latest approved revision). O
. Amendment 27 9 1-13
. . . ~ . . .-. -._ _ __- --- ~ _ _, .
LABWR maan Signdard Plant - ww. n - 7 f SECTION 9.2 U ~ CONTENTS Sts1]p.11 Title Page 9.2.1 Station Service Water System 9.21 9.2.2 ' Closed Cooline Water Svstem 9.21 9.23- Demineralized Water Makeun System 9.21 9.2.4 Potable and Snaltary Water Systems 9.2-1 9.2.5 - Ultimate IIcat Sink 9.2-1 9.2.5.1 Safety Design Dases (Interface Requirements) 9.2-1
-9.2.5.2 Power Generation Design Bases (Interface Requirements) 9.2-1.1.
9.2.53 System Description (Conceptual Design) 9.2-1.1 9.2.53.1 General Description 9.2-1.1 9.2.53.2 Spray Pond Description 9.2-1.1 '-- 9.2.533 Spray Pond Pump Structure 9.2-1.1 9.2.53.4- System Components 93 1.1 9.2.5,4 System Operation (Conceptual Design) 9.21.2
.9.2.5.4.1 Normal Operation 9.2-1.2 9.2.5.4.2. Cold Weather Operation 9.21.2-
~9.2.5.5: Spray Pond Thermal Performance (Conceptual Design) 9.2-1.2 9.2.5.5.1 Design Meteorology 9.2-1.2
-9.2.5.5.2 Spray Pond Water Requirements 9.2-1.2
- 9.2.5.6 Evahntion of UHS Performance (Interface Requirement) 9.2-1.2 9.2.5.7 - Safety Evaluation (Interface Requirement) 9.2-1.2 9.2.5.7.1 Thermal Performance 9.2-1.2 9.2.5.7.2 Effects of Severe Natural Events or Site-Related Events 9.2-1.2 9.2il Amendment 16 a-,.. .....u..._...,_,_,. _ . . . _ . _ _ _ . , . . , . . , . . . . , . _ . . _ . . _ ___ . . . _ . . . . _ _ _ . . . , , _ _ _ _ . _ _ , _ . _ _ , . . . _ , .
ABWR mamm Standard Plant imv. n SECTION 9.2 CONTENTS (Continued) O Section Illlt Eagt 9.2.5.73 Freezing Considerations 9.2-13 9.2.5.8 Conformance to Regulatory Guide 1.27 9.2 13 9.2.5.9 Instrumentation and Alarms 9.2-13 9.2.5.10 Tests and Inspections 9.2-13 9.2.6 Condensate Sterace Facilities and Distribution System 9.2-23 9.2.7 Plant Chilled Water Systertta 9.2 13 9.2.8 Makeuo Water System (Prenaration) 9.2-13 92.9 Makeun Water System (Condesate) 9.2-13 9.2.9.1 Design Bases 9.2-13 9.2.9.2 System Description 9,2-1.4 9.2.93 Safety Evaluation 9.2-2 9.2.9.4 Tests and Inspections 9.2-2 9.2.10 Makeun Water System (Pgr&dl , pistribution System 9.2-2 - 9.2.10.1 Design Bases 9.2-2 " 9.2.10.2 System Description 9.2-3 9.2.10 3 Safety Evaluation 9.2-3 9.2.10.4 Tests and Inspections 9.2-3.1 9.2.11 Beactor Buildine Cooline Water System 9.2-3.1 9.2.5.11.1 Design Bases 9.2-3.1 9.2.5.11.1.1 Safety Design Bases 9.2-3.1 9.2-lia Amendment 17
i ABWR men Standard Plant Rim. ti l SECTION 9.2 CONTENTS (Continued) Section lille Page l 9.2.11.1.2 Power Genesation Design liases 9.24 9.2.11.2 System Description 9.2-4 9.2.11 3 Safety Evaluation 9.24 9.2.113.1 Pailuse Analysis 9.24 9.2.11 3.2 Safety Evaluation of Equipment 9.25
- 9.2.11.4 Testing and inspection Requirements 9.26 l l
.- 9.2.11.5 Instrumentation and Control Requirements 9.2-6.1
- 9.2.12 lIVAC Norraal Cooline Water System 9.27
- 9.2.12.1 Design !!ases 9.27 9.2.12.1.1 Power Generation Design Bases . 9.27 ,
9.2.12.1.2 Safety Design Bases 9.25 l 9.2.12.2 System Description 9.27 9.2.123 Safety Evaluation 9.27 9.2.12.4. Tests and Inspeettons 9.28
=
9.2.12.5 Instrumentation Application 9.2-8 9.2.13 IIVAC Ememency Coollne Water System 9.28 9.2.13.1 - Desigr Basis 9.28 9.2.13.1.1 Power Generation Design Dases 9.28 9.2.13.1.2 - Safety Design Bases 9.28 9.2.13.2 System Description 9.29-9.2.13.3 Safety Evaluation 9.29 9.2.13.4 Tests and Inspection 9.29 O t 9 . 2 - 111 Amendment 17 i
.;..,-..._._,a....... .._.,__...__.._,._...__.s..~~.... . . . , , , . _ , , . . - . , . . ~ . - , . . . . . . . . , . . , , - . . . . . , _ _ _ - . , . . - , . . ~
ABWR men Standard Plant _ tinD 1 ('T V (e) MUWC transfer pumps (see Table 9.2 3) (three $50 Fpm at 141 psi head.) boundary; (2) capability to shut down the reactor and (3) Water enn be sent to the CST from the maintain it in a safe shutdown condition; or following sources: (3) ability to prevent or mitigate the conse-(a) MUWP pumps quences of events that could result in po-tential offsite exposures. (b) CRD system The MUWC systern is not safety related. (c) radwaste disposal system However, the systems incorporate features that assure reliable operation over the full range of (d) condensate demineralizer system effluent normal plant operations. (main condenser high level relief) 9.2.9A Tests and Inspettions (4) Associated receiving and distribution piping valves, instruments, and <.ontrols shall be The MUWC system is proved operable by its use provided. during normal plant operation. Portions of the system normally closed to flow can be tested to ensure operability and the integrity of the system. (5) Overflow and drain from the CST shall be sent to the sadwaste system for treatment. The alt operated isolation valves are capable of being tested to assure their operating (6) Any outdoor piping shall be protected from integrity by manual actuation of a switch freezing. Ic,tated in the control room and by observation of associated position indication lights. Plow to the variour. systems is balanced by (7) All surf aces coming in contact with the means of manoal valves at the individual takeoff condensate shall be made of corrosion- points. Divisional isolation valves are instal-resistant materiels. led at the primary containment boundaries. (8) All of the pumps mentioned in (2) above 9.2.10 Makeup Water Sptem (Purilled) shall be located at an elevation such that Distribution System adequate suction head is present at all water levels in the CST. 9.2.10.1 Design liases (9) Instrumentation shall be provided to indi- (1) The makeup water purified (MUWP) cate CST water level in the main control distribution system shall provide makeup room. water purified for makeup to the reactor coolant system and plant auxiliary systems. (10) Potential flooding is discussed in Subsection 3.4. Potential flooding from (2) The MUWP system shall provide purified water lines within the reactor building and the to the uses shown in Table 9.2 2. control building are evaluated in Subsection , 3.4.1.1.1. (3) The MUWP system shall provide water of the 0 quality shown in Table 9 2 2a. If these 9.2.9.3 Safety baluation water quality requirements are not met, the water shall not be used in any safety-Operation of the MUWC system is not required related system. The out of spec water shall to assure any of the following conditions: be reprocessed or oncharged. U (1) integrity of the reactor coolant pressure (4) The MUWP system is not safety-related. Amtndment 17 9M
ABWR mamn RIV II Standard Plant (5) All tanks, giumps, piping, and other equip- purified water storage tank level in the ment shall be made of corrosion resistant main control room. materials. (8) Continuous analyzers are located at the (6) The system shall be designed to prevent any demineralized water makeup system and at any l radioactive contamination of the purified dcmineralized water storage tank. These are water, supplemented as needed by grab samples. Allowance is made in the water quality (7) The interfaces between the hiUWP system and specifications for some pickup of carbon j all safety related systems are located in dioxide and air in any demineralired water l the control building or reator building storage tank. The pickup of corrision which are Seismic Category 1, tornado- products should be minimal because the hiUWT missile resistant and flood protected piping is stainless steel. structures. The interfaces with safety-related systems are safety related valves (9) Intrusions of radioactivity into the hiUWP which are part of the safety related system from other potentially radioactive systems. systems are p;evented by one or more of the following: (8) Safey related equipment located by portions of the h1UWP system are b Seismic Category I (a) check ulves in the hiUWP lines structures and protected from all system impact. (b) air (o, syphon) breaks in the hiUWP lines 9.2.10.2 System Description (c) the hiUWP system lines are pressurized while the receiving system is at g The hiUWP system P&lD i< shown in Figure 9.2 5. essentially atmospheric pressurc. 4 This system includes the (c,llowbg: (d) piping to the user is dead ended. l(1) Any purified water storage tank shall be provided outdoors with adequate frecre (10) There are no autor..atic valves in the protection and adequate diking and other h1UWP system. During a LOCA, the means to control spill and leakage. safety.related systems are isolated from the h1UWP system by automatic valics in (2) Two hiUWP forwarding pumps shall take suction the safety-related system. l from any purified water storage tanks. They shall have a capacity of 308 gpm and a 9.2.10.3 Safety Evaluation discharge head of 114 psi. Operation of the h1UWP system is not required (3) Distribution piping, valvee, instrument and to assure any of the following conditions: controls shall be provided. (1) integrity of the reactor coolant pressure (4) Any outdoor piping shall be protected from boundary; freer.ing. (2) capability to shut down the reactor and (5) All surfaces coming in contact with the maintain it in a safe shutdown condition; or purified water shall be made of corro-sion resistant materials. (3) ability to prevent or mitigate the conse-quences of events which could result in (6) All pumps shall be located at an elevation potential offsite exposures, such that adequate suction head is present l at all levels in a purified water storage The h1UWP system is not safety-related. tanks. However, the systems incorporate features that assure reliable operation over the full range of (7) Instruments shall be provided to indicate normal plant operations. Amendment 17 9.23 l 1
ABWR momu Starttf ard Plarli RIV 11 9.2.10.4 Tests and Inspections r~') , 1 The makeup water purified distribution system is proved operable by its use during normal plant operation. Fortions of the systemnormally closed to flow can be tested to ensure operability and integrity of the system. Flow to the various systems is balanced by rneans of manual valves at the individual takeoff points. 9.2.11 ReactorIlullding CoolingWater System 9.2.!!.1 Design 11ases 9.2.11.1.1 Safety Design Itases
,a
( ) (1) The reactor building cooling water (RCW) 'v_/' system shall be designed to remove heat from plant auxiliaries which are required for a safe reactor shutdown, as well as those auxiliaries whose operation is desired following a LOCA, but not essential to safe shutdown. The heat removal capacity is based on the heat removal requirement during LOCA with the ma'.imum ultimate heat sink temperature, 950F. As shown in Table 9.2 4, the heat removal requirement is higher during other plant operation modes, such as shutdown at 4 hours, llowever, the RCW system is not designed to remove this larger amount of heat when the ultimate heat sink is at the maximum temperature. (2) The RCW system shall be designed to perform its required cooling functions following a LOCA, assuming a single active or passive failure. (3) The safety related portions and valves (7 isolating the nonsafety related portions of Amendment 17 9M1
ABWR m-Sinndardiant nm a m the RCW system shall be designed to Seismic characteristics for RCW sptem components are g
") Category I and the AShfE Code, Section lit, given in Table 9.2 4d.
Class 3, Quality Assurance II, Quality Group y, E C,IEEE 279 and IEEE 308 requirements. The RCW tystem serves the auxiliary equipment listed in Table 9.2 la, b, and c. (4) The RCW system shall be designed to limit leakage to the environment of radioactive The RCW system is designed to perform its contamination that may enter the RCW from required safe reactor shutdown cooling function the RelR System. following a postulated LOCA, assuming a single active failure in any mechanical or cicettical (5) Safety related portions of the RCW system system. In order to meet this requirement, the shall be protected from flooding, spraying, RCW system provides three complete trains, which steam impingement, pipe whip, jet forces, are mechanically and electrically separated. In missiles, fire, and the effect of failure of case of a failure which disables any of the any non Seismic Category I equipment, as three divisions, the other two division meet required, plant safe shutdown requirements, including a LOCA 9: a loss of offsite power, or both. Each (6) The safety related portion of the RCW system RCW division is supplied electrical power from a shall be designed to meet the foregoing de- different division of the ESF power system. sign bases during a loss of preferred power (LOPP). During normal operation, RCW cooling water flows, through all the equipment shown in Table (7) The safety related electric modules and 9.2-4a, b, and c. safety related cables for the RCW system are in the control building which are Seismic During all plant operating modes, an RCW Category 1, tornado missile resistant and water pump and two heat exchangers are normally l flood protected structures, operating in each division. Therefore, II a V LOCA occurs, the RCW systems required to shut (8) Protection from being impacted adversely by down the plant safely are already in operation. missiles generated by any nonsafety related The second pump and the third heat exchanpr in l components shall be provided as discussed in each division are put in service if a LOCA Subsection 3.5.1. occurs. (9) Pruteetion against high energy and The nonsafety-related parts of the RCW system moderate energy line failures will be are not required for safe shutdown and, hence, provided in accordance with Section 3.6. are not safety sptems. Isolation valves sepa-rate the essential subsystems from the novsafe. 9.2.11.1.2 Power Gcr.crution Design llases ty related subsystems during a LOCA, in order to assure the integrity and safety functions of the The RCW system shah be designed to cool safety related parts of the system. Some non-various plant auxiliaries as required during: safety related parts of the system are operated (a) normal operation; (b) emergency shutdown; during all other modes, including the emergency (C) normal shutdown; and (d) testing. shutdown following an LOpP, or LOCA as shown in Table 9.2 4a, b, and c. 9.2.11.2 System Description Surge tank water level is monitored. A level The RCW system distributes cooling water dur- switch detects leakage and isolater, the non es-ing various operating modes, during shutdown, and sentia! subsystem, thus assuring continued oper-l during post LOCA operation. The system removes ability of the safety related services. Instru. heat from plant auxiliaries and transfers it to ments, controls, and isolation valves are locat-l the reactor service water system (Subsection cd in the safety related part of the RCW system p 9.2.15). Figures 9.2-la through 9.211 show the and designed to safety. grade requirements as ( piping and instrumentation diagram. Design stated in design basis (3) of Subsection 9.2.11. 1.1. I Amer.dment 17 924 l
ABWR naaman atandard Plant nrv n A dedicated sump and surep pump are prosided for each RCW division. Any system leakage or drainage may be collected, sampled and analyzed, and either returned to the RCW system or sent to the liquid radwaste system for treatment or to the llSD sample tank for discharge depending upon the radioactivity and impurities in the water. 9.2.113 Safety Evaluation 9.2.11 3.1 l' allure Analysis A system failure analysis of passive and active components of the RCW system is presented in Tabic 9.2 5. Any of the assumed failures of the RCW system are detected in the control room by variations of and/or alarms from the various system instruments and also from the leak detee-tion system sensing leakage in the ECCS pump and heat exchanger areas. O O l Amendment 17 9.2-41
ABWR mimu Irv n i Mandardjiant . 9.2.11.3.2 Safety Evaluation of Pquipment electrical equipment and instrumentation and controls as well as to mechanical eqtipment and O. Equipment served by the RCW system is listed piping): in Tables 9.2 4a, b, and c. The tables contain five operating modes: (1) flooding, spraying, or steam release due to pipe rupture or equipment failure: (1) normal operation; (2) pipe whip and jet forces resulting from pos. (2) s.hutdown at 4 hr; tulated pipe rupture of nearby high energy pipes; (3) shutdown at 20 hr.; (3) missiles which may result from equipment (4) hot 5tandby (No LOPP); Iallute; (5) hot standby (LOPP); and (4) fire; and (6) post.LOCA. (5) failures of any non Category I equipment (pettains to Scismic Category I equipment). The flow rates and heat loads are given for each equipment in each operating mode. Radiation monitors are provided to sample the RCW cooling water. Upen detection of radiation in the event of a LOCA, most of the noness.en. leakage in one of the systems, that system is tial cooling water uses are isolated by proper isolated by operator action from the control isolation valves. The instrument air system room, and the total cooling load can be rnet by service air system, control rod drive pump oil the other two systems. Consequently, radio-cooler and the reactor water cleanup system pump active contamination released by the RCW syrtem O coolers remain in service until the operator removes them from service. The nonsafety.related portion of the system is automatically isolated in the environment does not exceed allowable
' fined by 10CPR100.
in the event of a rupture in the nonsafety re- Th, safety related parts of the RCW system lated subsystem. The surge tank water level is are de ,ned to Seismic Category I and AShtE monitored. A level switch is activated by a Code, ocction 111, Class 3, Quality Assurunee 11 significant leak, sending an isolatian signal to and Quality Group C requirements. The design close two valves. One valve on the supply line an,o meets IEEE 279 and IEEE-308 requirements. and one valve on the discharge line are used, Isolation valves for nonsafety related service with suitable power and controls from divisional water systems also meet the above requirements. sources to assure isolation in the event of any single active failure. Single isolatiou valves The nonessential portion of the RCW system is are used on the basis that an active f ailure of designed to the ANSI B31.1 Power Piping Code and one it,olation valve disables only that system of the requirements of Quahty Group D, which it was a part. The design pressure and temperature of the 2 The RCW system is designed to withstand a RCW system and piping are 14 kg/cm g (200 single active failure without losing its capabi- psig) and 70"C (158 F) maximum. lity to participate in the safe shutdown of the reactor following a LOCA or DBA. Table 9.2 5 System low point drains and high point vents gives the result of a system failure analysis of are provided as required. active and passive components. All divisions are maintained full of water Redundant trains of the RCW system are separa- when not in service except when undergoing main-ted and protected to the extent necessary to tenance. assure that sufficient equipment remains oper O ating to permit shutdown of the unit in the event of any of the following (separation is applied to Amendment 14 9M
ABWR mom Sandard i Innt tw n but not to a temperature that would damage System components and piping materials are equipment or require an immedia'e shutdown, selected where require (' to be compatible with the available site cooling water in order to minimize 9.2.11.4 Testing and Inspection Requirrments corrosion. Cathodic protection of the tubing side of the heat exchanger shall be provided. The RCW system is designed to permit periodic Adequate corrosion safety factors are used to in. service inspection of all system components assure the integrity of the system during the to assure the integrity and capability of the life of 'he plant. system. 1 During all plant operating modes, all The RCW system is designed for periodic pres-divisions have at least one RCW cooling water sure and functional testing to assure.: (1) the pump operating. Therefore, if a LOCA occurs, the structural and lenktight integrity by visible RCW cooling water system required to shut down inspection of the componentst (2) the j the plant safely is already in operation. If a operability and the performance of the active loss of offsite power occurs doiing a LOCA, the components of the system; and (3) the pumps momentarily stop until transfer to standby operability of the system as a whole. diesel generator power is completed. The pumps are restarted automatically according to the The tests shall assure, under conditions as diesel loading sequence. If a LOCA occurs, most close to design as practical, the performance of nonsafety related components are automatically the full operational sequence that brings the isolated from the RCW system. Consequently, no system into operation for reactor shutdown and operator action is required, following a LOCA, to for LOCA, including operating of applicable start the RCW system in its LOCA operating mode, portions of the Reactor Protection System and the transfer between normal and standby power All heat exchangers and pumps will be required sources. during the following plant operating conditions, in addition to LOCA: shutdown at 4 hours, These tests shall include periodic testing of shutdown at 20 hours and hot standby with loss of the heat removal capability of each RCW beat AC power. exchanger. Each of these heat exchangers has been designed to provide 20% margin above the Loss of one RCW division will result in loss beat removal capability required for LOCA in of RCW cooling to every other RIP (five total) as Tables 92-4a, b and c. This margin is provided shown on RRS P&lD (Figure 5.4-4) and will cause to compensate for the combined effects of those five RIPS to runback to minimum speed. The fouling and tube plugging. When this margin is RIP M G set in the same electrical division, no longer present, the heat exchanger heat which is cooled by the same RCW division which removal capacity will be increased by either failed and powers two more RIPS, would stop by cleaning or retubing. l M G set cooling water protection. This would completely shutdown three RIPS and would have the The RCW system is supplied with a chemical resulting total of sesen RIPS cither at minimum addition tank to add chemicais to each speed or stopped. Assuming the event began at division. Tbc RCW system is initially filled full power on the 100G Control Rod Line, the with demineralized water. A corrosion inhibitor l resulting temporary reactor power would be can be added if desired. These measures are ad-approximately 60% power. The operator would then equate to protect the RCW system from the ill correct the RCW problem or initiate a normal effects of corrosion or organic fouling. l plant shutdown. The RCW system is designed to conform with ! The drywell cooling system can perform its the foregoing requirements. Initial tests shall I function after the loss of any RCW division, be made as described in Subsection 14.2.12. With only one RCW division and one drywell cooler operating, the drywell temperature will increase Amen:! ment 17 924
ABWR man Sinndard Plant rti v n i 1 (N 9.2.11.5 Instrumentation and Control l () Requirements l All equipment is provided with either globe or I butterfly valves to give the capability for I manual control. These valves are acrt vible downstream of the equipment for regulation of flow through the equipment or for balancing the circuits. The isolation valves to the neuessen. tial RCW system are automatically and remote manually oper atod. Pressure laps or indicators at equipment are provided to enable the operator to adjust the differential pressure across cach heat exchanger or cooler and also to allow leak checking. Locally mounted temperature indicators or test wells are furnished on the equipment cool-ing water discharge lines to enable verification of specified heat removal during plant opera-tion. The required heat removal and flow rates are shown in Tab!cs 9.2 4a, b, end c. The combination of pressure taps (or indica. _ tors) and temperature indicators allow correct system balancing with or without a system heat [V) load. For purposes of system balancing, provi-sions for flow measurement are provided as re-quired. Connections to a radiation monitor are pro-vided in each division to detect radioactive contamination resulting from a tube leak in one of the RilR exchangers, fuel pool exchangers, or other exchangers. Isolation valves for RilR heat exchangers and nonessential cooling water subsystems are pro-vided with remote manual switches and indication on the remote shutdown panel. l
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Amendment 17 9.241
ABWR m aman Rtv n Sandard Plant () 9.2.12 IIVAC Normal Cooling Water Division 1 isolation valve outside containment and Class 2 piping into the drywell. The return System l line penetration has divisional isolation valves 9.2.12.1 Design flases inside and outside containment. These valves are motor operated. 9.2.12.1.1 Power Generation Des'gn Itases The liVAC normal cooling water system (nonsafe-ty related) shall provide chilled water to the cooling coils of the drywell coolers, of each building supply unit and of local air condition- , ers to maintain design thermal environments dur- ( ing normal and upset conditions. The supply tem-perature is 44.6 F. The return temperature is 53.6 F. 9.2.12.1.2 Safety Design Itases No diesel generator power is available to The llVAC normal cooling water system does not this system during a LOPP or a LOCA. perform any safety functions, except for the containment penetration and isolation valves. 9.2.12.3 Safcty Evaluation 9.2.12.2 System Description Operation of the llVAC normal cooling water system is not required to assure the following The llVAC uormal cooling water system compo- conditions: Iqi nents are listed in Table 9.2 6 and shown in V Figure 9.2 2. (1) integrity of the reactor coolant pressure boundary; System components consist of five 25% chil-1ers, each with pumps, serving a common chilled (2) capability to shut down the reactor and main. water distribution system connected to the chil- tain it in a safe shutdown condition; and led water cooling coils in the drywell coolers, the cooling coils of each building supply unit (3) a bility t o pr eve n t or mitigate the and cooling coils of local air conditioners. consequences of events which could result in Condenser cooling is from the turbine building potential offsite radiological exposures. cooling water system. Each chiller and pump set has either a three way mixing valve for automati- The llVAC normal cooling water system is not cally controlling the temperature of the chilled safety related, liowever, it does incorporate water delivered or a flow control valve to main- features that assume reliable operation over the tain the desired temperature. Each chiller eva- full range of normal plant operations. porator is designed, fabricated and certified in accordance with the AShf E Code Section Vill, Portions of the chilled water system which Division 1. A chemical feed tank is provided. penetrate the primary containment are provided hinkeup water is from the turbine building cooling with isolation valves and penetrations which are water system surge tank which receives water from Seismic Category I, Safety class 2. The valves the hfUWP system. Isolation valves and piping for may be manually operated from the control room, primary containment penetrations are designed to except when a LOCA signal assumes control. seismic Category 1, ash 1E code, Section lit, class 2, Quality Group 13, Quality Assurance B requirements. The supply line penetration has a Amendment 17 937
ABWR -man Standard Plant nir n 9.2.12.4 Tests and inspections signal. Condenser water is provided from the turbine building cooling water system. The Initial testing of the system includes perfor- three way valve on the chilled water circuit mane:: testing of the chillers, pumps and coils controls the temperature of the chilled water to for conformance with design heat loads, water the cooling coils from the areas thermocouple flows, and heat transfer capabilities. An inte- controller. The thermocouples are located in grity test is performed on the system upon each area being cooled. The control roorn completion, operator can adjust the three.way valve position l during startup and whenever high chilled water Provision is made for periodic inspection of return temperatures are indicated and alarmed. major components to ensure the capability and Alternately, instead of the three way valves, a integrity of the system. Local display devices flow control valve may be used. are provided to indicate all vital parametert. required in testing and inspections. Remote controlled valves permit isolation of any drywell cooling coil in the event of the i I The chillers are tested in accordance with coil developing a detectable leak. ASif RAE Standard 30 (Methods of Te . ting for Rating Liquid Chilling Packages). The pumps are tested 9.2.13 IIVAC Emergency CoolingWaler in accordance with standards of the 11ydraulic System Institute. ASME Section Vill and TEMA C stan-dards apply to the ASilRAE Standard 33 (Methods of 9.2.13.1 Dealgn Itasis Testing for Rating Forced Circulation Air-Cooling and licating Coils). 9.2.13.1.1 Power Generation Design llases Samples of chilled water may be obta:ned for The llVAC emergency cooling water system chemical analyses. Radioactivity is not expected (llECW) (safety related) shall provide chilled to be in the chilled water. water under normal plant operating conditiens to the cooling coils of the main control room air 9.2.12.5 lustrumentation Application conditioning units, to the diesel generator zone l coolers, and to the control building essential A regulated supply of demineralized makeup electrical equipment room cooling coils. See water adds water to the turbine building cooling Table 9.2 9. The supply temperature is water TCW expansion tank by water level controls, 44.6'F the return temperature is 62.6 F. and the chiller units are con trolled indi-vidually by remote manual switches. 9.2.13.1.2 Safety Dealgn liases A temperature controller and flow switch The 11ECW system performs a safety design continuously monitor the discharge of the evapo- function, rator. If the temperature of the chilled water drops below a specified level, the control auto- (1) The llECW system shall deliver chilled water matically adjusts the temperature control inlet to the control building essential electrical guide vanes of the chiller compressor. Flow equipment room coolers, the diesel generator switches prohibit the chiller from operating un- zone coolers, and the main control room less there is water flow through both evaporator coolers during shutdown of the reactor, and condenser. See Section 3.11 for temperature operating modes and abnormal reactor requirements. in case of a chiller or pump trip, conditions including LOCA, the standby units are automatically started. (2) Sufficient redundancy and electrical and Chilled water flow into and out of the mechanical separation shall be provided to cor.tainment is controlled by isolation valves ensure proper operations under all con 6;- which shall be automatically closed after a LOCA tions. O Amendment 14 918
ABWR man niv n Standard Plant controlled by a temperature control valve. The o (3) The system shall be designed and constructed in accordance with Seismic Category l, ASME code, Section 111, Class 3 requirements. subsystems are designated Division I and Division 11 power, respectively. One compressor is the operating unit, while the other is on (4) a ne spm shall be powered from Class 1E standby. Condenser cooling is from the buses. correspondlag division of RCW. (5) The llECW system shall be protected from Piping and valves for the llECW system, as mi6siles in accordance with Subsection well as the cooling water lines from the RCW 3.5.1. system, designed entirely to ASME Code, Section 111, Clus 3, Quality Group C, Quality Assurance (6) Design features to preclude the adverse B requireme nts. The extent of this effectr. of water hammer are in accordance classification is up to and including drainage with the SRP section addressing the block valves. There are not primary or resolution of USl A 1 discussed in secondary containment penetrations within the NUR EG 0927. system. The llECW system is not expected to contain radioactivity. These features shall include liigh temperature of the returned cooling (a) an elevated surge tank to keep the water causes the standby chiller unit to start system filled; automatically, Makeup water is supplied from the MUWP system, at the surge tank. Each surge (b) vents provided at all high points in the tank has the capacity to replace system water system; losses for more than 100 days during an e m e r ge n cy. (c) after any system drainage, venting is essured by personnel training and 9.2.t3.3 Safety Esaluation O procedures; and The llECW system is a Seismic Category 1 (d) system vahes are slow acting. system, protected from flooding and tornado missiles. All components of the system are (7) The llECW system r. hall be protected from designed to be operable during a loss of normal failures of high and medium energy lines as power by connection to the ESF buses. Redundant discussed in Section 3.6. components are provided to ensure that any single component failure does not preclude sys. 9.2.13.2 System Description tem operation. The system is designed to meet the requirements of Criterion 19 of 10CFR50. The llVAC emergency cooling water system Each chiller is isolated in a separate room. consists of redundant subsystems in three divisions. Each division consists of two 50% 9.2.13.4 Tests and inspection chiller units, two 50% pumps, instrumentation and distribution piping and valves to corresponding Initial testing of the system includes per-cooling coils. A chemical addition tank is formance testing of the chillers, pumps and shared by all llECW divisions. Each IIECW division coils for conformance with design capacity water shares a surge tank with the corresponding flows and heat transfer capabilities. An inte-division of the RCW system. The chiller capacity grity test is performed on the system upon is designed to cool all heat loads with two completion, chillers, but also cool the heat load of the main control room with one chiller. The llECW system is designed to permit periodic in service inspection of all system Equipment is listed in Table 9.2-9. Each components to assure the integrity and cooling coil has a three-w<.y valve controlled by capability of the system. O a room thermostat. Alternately, flow may be 9.2-9 Amendment 17
ABWR momn Standard Plant nix. n The llECW system is designed for periodic l pressure the structuraland and leaktight functional testing integrity to assure: (1) by visual inspection of the co nponents; (2) the operability and the performance of the active components of the system; and (3) the operability of the system ss a whole. Local display devices are provided to indicate all vital parameters required in testing and inspections. Standby features are periodically tested by initiating the transfer sequence during norrnal operation. The chillers are tested in accordance with ASIIRAE Standard 30. The pumps are tested in acco'i!ance with standards of the llydraulic Institute. ASME Section Vill and TEMA C standards apply to the heat exchangers. The cooling coils are tested in accordance with AS11R AE Standard 33, 9.2.13.5 Instrumentation Application A regulated supply of makeup water is provided to add purified water to the surge tanks by water level controls. The chilled water pumps are controlled from the main control panel. The standby chiller is equipped with an interlock which automatically starts the standby chiller and pump upon failure of the operating unit. The chiller units can be controlled indivi. dually from the main control room by a remote manual switch. Chilled water temperature is controlled by inlet guide vanes on each chiller refrigerant circuit. Condenser water flow is controlled by a three way valve to provide constant inlet condensate water temperature. A temperature controller and flow switch continuously monitor the discharge of each O Amendment 17 9,2-9.1 i .
AllWR mwan S11111dallElullt kiTJJ 9.2.17 Interfaces be designed with no laterconnections with O 9.2.17.1 Ultimate llent Sink Capability systems having the potential for containing radioactive materials. Prateetion shall be f
=
provided through the une of air gapt, where Interface requirements pertainini' to ultimate necessary. (See Subscetion 9.2.4), heat sink capability are delineated in Subsection 9.2.5 as follows: Smhttt!tti 11115 9.2.5.1 Safety Design liases 9.2.5.2 Power Generation Det.lgn 11ases D 9.2.5.6 livaluation of Ulls Performance 9.2.5.7 Safety livabation 9.2.5.8 Conformance to llegulatory Guide 1.27 9.2.5.9 Instrumentation and Alarms 9.2.5.10 Tests and Inspections 9.2.17.2 Makeup Water Splem Capability The raw water treatment and preparation of the drmineralired water is sent to the makeup water system (purified) described in Subsection 9.2.10. The deminerallred water preparation system 6 hall consist of at least two divisions capable of producing at least 200 rpm of demineralired water each. Storage of demineralized water is needed during peak usage periods, rented portable dernineralirers shall be used as required. The makeup water preparation system shall be located in a building which does not contain any safety related structures, systems or g components. If the system is not available, g demineralized water can be obtained from mobile equipment. The system shall be designed so that any failure in the systern, including any that cause flooding, shall not result in the failure of any safety related structure, system or component. 9.2.17.3 Potable and SanitaryWater System l The potable and sanitary water system shall Amendment 17 92 D
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ABWR numn
$18AdArd Plant RFV f)
O TABLE 9.2 2a WATER QUALITY CHARACTERISTICS FOR THE MAKEUP WATER SYSTEM (PURIFIED) WATER QUALITY Operating System Maximum
.fARAMETER Tarcet - Design Value Chloride ppb) 10.0 20.0 100.0 Sulfate (ppb) 10.0 20.0 100.0 l Conducthity* at 250C (uS/cm) 0.2 03 1.0 Silica (ppb as SiO2)- 10.0 20.0 100.0 l plI at 250C min 6.4 6.2 5.6 l max 7.8 8.0 8.6 Corrodon Product Metals (ppb)
Fe insoluble soluble Cu total 10.0 20.0 100.0 all other metals , a sum 10.0 20.0 100.0 E Organic Impuritics" Equivalent K (uS/cm) 0.2 0.4 2.0
- Does not include an incremental conductivity value of 0.8 uS/cm at 250C due to carbon dioxide from air in watet stored in tanks open to the atmosphere.
*'- Organic impurity values apply to fresh makeup water stored in any Demineralized l Water Storage Tank.
1 l 1-O Amendment 17 92 111 l
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ABWR zwiooxn nir si Standard Plant TAllLE 9.2 8 (m) IIECW SYSTEM COMPONENT DESCRIPTION llECW Chillers Type Centrifugal hermetic Quantity 6 (50% capacity units) four 6 Capacity (refrigeration) 2.03 x 10 BTU /h each two 1.23 x 1(l' BTU /h each Chilled water pump flow four 220 gpm each two 120 gpm each Supply temperature 44.6 F Condenser water now four SM gpm each two 341 gpm each Supply temperature (max.) 95'P Condenser Shell and tube Evapore'or - Shell and tube f') U lilX3V Water Pitmid Quantity 4 - 220 rpm each 2 - 120 ppm each Type Centrifugal, horizontal N._/ Arnendment 14 9224
ABWR mamn Standard Plant ._ nov.n TAllLE 9.2 9 IIVAC EMERGENCY COOLING WATER SYSTEM llEAT LOAI)S O NORhtAL EhtEl(GENCY IIcat Chilled lirat Chilled ; Load Water 1xiad Water l (s106 g3ow (ggp6 gjo , DIVISION SYSTEh! IITU/h) (gpm) Il1U/h (gpm) A main controf room 1.35 125 1.25 116 diesel generator 0.83 62 0.83 62 rene (A) control building $ .19 88 1.19 88 elect, eq. toorn (A) Tott.1 3.37 275 3.27 266 11 diesel generator 0.86 63 0.86 63 rone (II) control b!dg. 1.19 88 1.19 88 elect, eq. room (B) Total 2.05 151 2.05 151 C main control 1.35 125 1.25 116 room diesel generator 0.86 63 0.86 63 rene (c) control bldg. 1.19 88 1.19 88 elect. eq. toom (C) Total 3.40 276 3.30 267 l O l _ _ , , - l I
ABWR maan Standard Plant mn 9.4 AIR CONDITIONING (3) The outside design conditions for the control liFATING COOLING AND toom HVAC splem are 115 P during the VENTILAtlON SYSTEMS summer and 40 e during the winter. 9A.1 ControlIlulldingIIVAC 9.4.1.1.3 System ikseription The control building heating,ventilatlog and The control room is heated, cooled and pressurized air conditioning (llVAC) system is divided into two by a recirculated air system with filtered outdoor air separate systems. A IIVAC system for the control for ventilation and pressuritation purposes. The room equipment on the top two floors. Plus a recirculated air and the outdoor alt will be mixed and IIVAC system for essential electrical and heat ex- drawn through a filter section, a heating coil section, , hanger equipment. and a cooling coil section. Under normal conditions, sufficient air is supplied to pressurire the control room 9.4.1.1 Control Itoom Equipment ilVAC and esfiltrate to preuurire the control building. 9.4.1.1.1 Ikslan Itasis The control building IIVAC P&lD is shown in Figure 9.41. The control room now rate is given in (1) The control room (llVAC) system is dc61gned Table 9.4 3, and the system component descriptions with sufficient redundancy to ensure operation are given in Table 9.4 4. The control building ; under emergency conditions assuming the recirculation unit consists of a medium grade bag g single failure of any one active component. filter, a heating coil, cooling coil, two 50% capacity g supply fans. (2) Provisions are made in the system to detect and limit the introduction of airborne radioactive materialin the control room. Two 50% capacity return exhaust fans draw air from the electrical area, corridors, control room, (3) Provisions is made in the system to detect and computer room, office areas, and the llVAC equip-remove smoke and radioactive material from ment room. This air is returned to the air condition. the control room. ing unit during normal operations. Modulating damp-ers in the return duct work to the fans are controlled (4) The HVAC system is designed to provide a by a pressure controller to maintain the required controlled temperature environment to ensure positive pressure. The controller is located in the the continued operation of safety related equip- electrical equipment erca. During smoke removal ment under accident conditions, mode, both fans are turned on and the air is discharged to atmosphere. (5) The llVAC system and components are 10-cated in a Seismic Category I structure that is An emergency recirculation system consisting of an tornado missile and flood protected. electrical heating coil, a prefilter, ilEPA filter, char. coal adsorber, and llEPA filter with a booster fan,is E (6) Tornado missle barriers are provided for intake provided parallel to the :ormal mixed outdoor and and exhaust structures, return air path to the supply conditioning units. The charcoal adsorber w'll be 2 inches deep as a minimum. The system is normally on standby for us 9.4.1.1.2 Power Generation Design liasis high radiation. A radioactivity monitoring system (1) The IIVAC system is designed to provide an monitors the building intakes for radiation. The-emironment with controlled temperature and radiation monitor allows the control room operator to humidity to ensure both the comfort and safety select Ihe safest intake. The makeup air for of the operators. The nominal design condi. preuurization can be diverted through the llEPA and tion s for the control room environment are charcoal adsorbing system before distribution in the 75 F and 50% relative humidity, control room areas. (2) The system is design,d to permit periodic in- Smoke detectors in the control room and the con-O spection of the principal system components. trof equipment rcom exhaust systems actuate au Amendment 17 94t
ABWR mamn Slituditrd PlDi _. Rn n alarm on indication of smoke. Dampers must be po- The safety.related isolation valves at the outside air sitioned through a remote manual switch to allow the intakes are protected from becoming inoperable due exhaust air fans to exhaust 10f 4 of the conditioned to freering, icing, or other emironmental conditions. air. For information on fire protection and smoke Tbc IIVAC equipment space is physleally sepa- removal methods for the control building IIVAC rated into divisional rooms. Each divisional room system see Subsections 9.5.1.1.5 and 9.5.1.1.6. consists of an ait intake room and an air exhaust room. 9.4.1.1.5 Inspection and Testing Requirements 9.4.1.1.4 Safety Evaluation Provisions are made for periodic tests of the out-door air cleanup fans and filters. These tests include The control building IIVAC r,ystem is designed to determinations of differenitial pressure across the rnaintain a habitable emironment and to ensure the fiher and of filter efficiency. Connections for testing, operability of components in the cont.ol room, All such as injection, sempling and monitoring are prop-control room IIVAC equipment and surrounding erly located so that tcst results are indicative of per-structures are of Scistnic Category I design and oper- formance. able during loss of the offsite power supply. The high efficiency particulate air (llEPA) filters The ductwork which senices these safety functions may be tes:cd periodically with dioctyl phthalate is termed ESF ductwork, and is of Seismic Category smoke (DOP). The charcoal filters may be periodi. I design. ESF ducting is high pressure safety grade cally tested with freon for bypasses. ductwork designed to withstand the maximum posi-tive and/or negative pressure to which it can be sub- The balance of the system is proven operable by its jected under normal or abnormal conditions. Galva- use during normal plant operation. Portions of the nbed steel ASTM A526 or ASTM A527 is used for system normally closed to flow can be tested to outdoor air latake and exhaust duca. All other ducts ensure operability and integrity of the system. are welded black steel ASTM A570, Grade A or Grade D. Ductwork and hangers ar Seismic Cate- 9.4.1.1.6 Instrumentation Application gory 1, Bolted flange and welded jointa are qualified per ERDA 76-21. The area exhaust fan is started manually and the fan g discharges the air to atmosphere. g Redundant and independent components are pro-
' sided where necessary to ensure that a single failure A high radiation signal automatically starts the out-will not preclude adequate control room ventilation, door air cleanup system, closes the normal air inlet damper and closes the exhaust air dampers.
A radiation monitoring system is provided to detect high radiation in the outside air intake ducts. A temperature indicating controller senses the tem-A radiation monitor is provided in the control room perature of the air leaving the air cleanup system. to monitor control room area radiation levels. These The controller then modulates an electric heating coil monitors alarm in the control room upon detection to maintain the leaving air temperature at a preset of high nadiation conditions. Isolation of the control limit. A limit switch will cause an alarm to be actu. toom and initiation of the outdoor air cleanup unit ated on high air temperature. A moisture sensing fans are accomplished by the following signals: element working in conjunction with the temperature controller measures the relative humidity of the air (1) high radiation in the inside air intake duct, entering the charcoal absorber. and Differential pressure indicators show the pressure (2) manual isolation. drop across the prefilters and the llEPA filters. A differential pressure indicating switch also measures Under normal conditions, sufficient air is supp!ied the pressure drop across the entire filter train. The to pressurize the centrol room and exfiltrate to pres- switch causes an alarm to be actuated if the pressure surire the control building. drop exceeds a preset limit. A flow switch m the out-Amendment 17 9411
ABWR newi nun hmdard Plant i door air leanup sptem fan discharge duct automati- the heater and demist:rs. The heaters and demisters cally starts the standby system and initiates an alarm are put in'.o systems to regulate the relative 'umidity on operating fan failurr. of the air as i' enters tne ESF filter train, since the control room air handling units are designed to The electrical equipment area and the control maintain the control room temperature and humidity room area return exhaust fans start automatically within limits, additional controls are not necessary for when the air conditio sing unit it started. Each fan the ESF filter train, intel damper is open Lutomatically. The exhaust dampers to the conditioning unit are opened 9 . 4 . 1 . 1 .11 Standard Resit* Plan 6.5.1 Compliance automatically. Status Differential pressure.tudicating controllers The control room ESF sptem complies with SitP modulate dampers in the return air ducts to maintain 6.5.1. Table 6.5.11. The only exceptions are for heater space positive pressure requirements. and moisture separator instrumentation requirements. Since ther.e components are not necessary for the AllWR design, no in trumentation has been supplied to monitor their operation. Relative humidity and temperature of the inlet air is maintained by the control room air handling sptem. 9.4.1.2 Essential 1:lectelcal and Reactor llullding Cooling Water 1:quipment ilVAC 9.4.1.2.1 Design llasts The cooling unit r. tarts automatically on a signal from the temperature indicating controller installed (1) The llVAC sptem is designed with sufficient in the llVAC toom. The controller modulates a redundancy to ensure operation under three-way chilled water vahe to maintain the space emergency conditions assuming the failure of conditions. any one active component. Dt; ring winter, the electric t. nit heatern are cycled (2) The llVAC system is designed to provide a by temperature indicating controller switches, controlled temperature environment to ensure located within the filter rooms and the air-handler the continued operation of safety related rooms. equipment under accident conditions. The supply and return air duct work has manual (3) The llVAC system and components are located balancing dampers provided in the branch ducts for in a Seismic Category I structure that is balancing purposes. The dampers are locked in tornado-missle and flood protected. place after the system is balanced. , , _ (4) Tornado missle barriers provided for intake and A 9.4.1.1.7 Regulatory Guide 1.52 Compliame Status exhaust i.tructures. $ The control room ESF filter trains comply with 9.4.1.2.2 Power Generation Design llasts all applicable provisions of Regulatory Guide 1.52, Section C except as noted below. (1) The llVAC system is designed to prmide an emironment with controlled temperature during The revisions of ANSI N509 and N510 listed in normal operation to ensure the comfort and Table 1.8-21 are used for AllWR ESF fliter train safety of plant personnel and the integrity of the design; the Regulatory Guide references older essential electrical and RCW equipment. revisions of these standards. (2) The sptem is designed to facilitate periodic The control room ESF fliter trains are in inspection of the principal system components. compliante with the sptem design criteria except for Amenkent U 9012
ABWR nwmin Standard Plant nn m (3) Design outside air temperature for the heat exchanger building IIVAC system are 115*F during the summer and -40 F during winter. 9.4.1.2.3 System Description The essential electrical llVAC system is divided gl g into 3 independent subsystems with each subsystem
' serving a designated area. Each Subsystem serve as essential electrical heat exchanger equipment ilVAC for divisions A, B, C, and D.
The control building essential eletricalllVAC system flow rates are given in Table 9.4 3, and system component descriptions are given in Table 9.4-4. 9.4.1.2.3.1 Safety Related Subsystem 1 Subsystem 1 specifically s,erves: (1) Safety related battery room 1, (2) E=sential chiller room A, (3) Ri1 cooling water pump and heat exchanger room A, (4) liVAC equipment room, (5) Safety-related electrical equipment room, (6) Passages, (7) Non-essential battery room, (8) Non-essential electrical equipment rooms. O Amendment 17 9.4-1.3
ABWR muun Standard Plant nedl Recirculation unit for subsystem 1 consists of a (4) IIVAC equipment room, prefilter section, a high efficient filter section, an electric heater, a cooling coil, and two 50% capacity (5) Safety.related electrical equipment room, supply fans. (6) Panages, (7) SGTS equipment at EL 7200 in Cil. Two 50% capacity return exhaust fans discharge to the atmosphere. Recirculation unit for Subsystem s 3 cornists of a prefilter section, a high efficient filter section, an electric heater, a coolirig coil, and two 50% capacity 9.4.1.2.3.2 Safety Related Subsystem 2 supply fans. Subsystem 2 specifically serves: Two 50% capacity return exhaust fans discharge (1) Safety.related battery rooms 2 and 4, to the atmosphere. (2) Essential chiller room II, (3) RCW pump and heat exchanger room II, 9.4.1.2.4 Safety thuluation (4) IIVAC equipment room, The cw.ntial electrical llVAC system is designed to ensure the operability of the essential electrical equip-(5) Safety related.clectrical equipment room, ment, and to limit the hydrogen concentration to less ; than 2% by volume in the battery rooms. All $ (6) Passages, safety related ilVAC equipment and surrounding structures are of scismic category I design and g Remote Shutdown Panel Room, operable during loss of the offsite power supply, g (7) Recirculation unit for Subsystem 2 consists of a The ductwork which senices these safety functions prefilter section, a high efficient filter section, an is termed ESF ductwork, and is of Seismic Category I c!cctric heater, a cooling coil, and two 50% capacity design. ESF ducting is high pressure safety grade supply fans. ductwork designed to withstand the maximum positive and/or negative pressure to which it can be subjected under normal or abnormal conditlons. Galvanized _ steel ASTM A526 or ASThi A527 is used for outdoor Two 50% capacity return exhaust fans discharge to air intake and exhaust ducts. All other ducts are the atmosphere, welded black steel AS*IM A570, Grade A or Grade D. Ductwork and hangers are Seismic Category 1. Ilotted Flange and welded joints are qualified per ERDA 76-21. 9.4.1.2.3.3 Safety Related Subsystem 3 Redundant components are provided where neces-Sub:,ystem 3 specifically serves: sary to ensure that a single failure will not preclude adequate heat-exchanger building ventilation. (1) Safety related battery room 3, 9.4.1.2.5 Inspection and Testing Requirements (2) Essential chiller room C, Provisions are made for periodic tests of the out-(3) RCW water pump and heat-exchanger room C, door air cleanup fans and filters. These tests include determinations of differential pressure across the filter and of filter efficiency. Connections for testing, O' such as injection, sarnpling and monitoring are prop-94-14 Amendment 17
ABWR m6imin Standard Plant nu n erly located so that test results are indicative of per-formance. The balance of the system is proven operable by its use during normal plant operation. Portions of the system normally clor.ed to flow can be tested to ensure operability and integrity of the system. 9.4.1.2.6 Instrumentation Application The area exhaust fans are started manually and the fans discharge the air to atmosphere. A temperature indicating controller senses the temperature of the alt leaving the air cleanup system. The controller then modulates an electric heating coil to maintain the leaving air temperature at a preset limit. A limit switch will cause an alarm to be actuated on high air temperature. l The est.cntial electrical return exhaust fans start automatically when the air conditioning unit is started. Each fan inlet dampet is open automatically. The exhaust dampers are closed automatically and the return air dampers to the conditioning unit are opened automatically. For information on fire protection and smoke removal methods for the heat exchanger building ilVAC systems, see Subsections 9.5.1.1.5 and 9.5.1.1.6. The chiller room cooling unit starts automatically on a signal from the temperature indicating control-let installed in the chiller room. The controller mod-ilates a three way chilled water valve to maintain the space conditions. During winter, the electric unit heaters are cycled by temperature. indicating controller switches, lo-cated within the filter rooms and the air handler rooms. l O
.__ s - i
ABWR 2w.i.n Standard Plant un.a 9A.2 Reactor Building Ventilation System-i GE PROPRIETARY provided under separate cover t l'agt AmendmtEl 9.4 2c 16 , 9.4 2d 16 9.4 2c 16 9.4 2f 16 9.4 2g . 17 9.4 2h 17 9.42i 16 i 9,4 2ia 16 l AmendmW 17 9.*2e - 2ia
~S-+,m.. ...-.,,-,...m,-w, - -
.....mm,v,~w-,r.,.,my-m.-__ __~.m.ww,-w,wwm,
AlnVR u-u Sinndard 1%nt am n 9.4.6 Radwaste Ilullding IIVAC Sptetu system. The air conditioning system is a unit ( air conditioner consisting of a water cooled C J 9.4.6.1 Design llases condenser, compressor, cooling coil, heating mil, filters and fan. Outdoor air and recirculating air are 9.4.6.1.1 Safety Design liases mised and drawn through a prefilier, a heating coil, a cooling coil, and two 100% supply fans. One fan is The radwaste buitding IIVAC system has no normally operating and the other fan is onstandhy. A safety-related functio, os defined in Section 3.2. pressure differential controller regulates the Failure of the system cors not compromise any exfiltration from the control room to maintain it at a safety related systern or component and does not positive static pressure, preventing airborne prevent safe reactor shutdown. Provisions are contamination from entering. incorporated to minimite release of radioactive substances to atmosphere and to prevent operator The exhaust air system consists of two 100% exposure. exhaust fans. One fan is normally operating and the other is on standby. Exhaust air from the control 9.4.6.1.2 Power Generation Design flases room is monitored for airborne radioactivity before i exhausting to the atmosphere. l The radwaste building ventilation sys:cm is l designed to provide an environment with controlled 9.4.6.2.2 Itadwaste llullding ilVAC Contrni Splem temperature and airflow patterns to insure both the comfort and safety of plant personnel and the The llVAC control system for the remainder of the integrity of equipment and components. The radwaste building is a once through type. Outdoor air l tadwaste building is divided into two rones for air is filtered, tempcred and delivered to the conditioning and ventilation purposes. These zones noncontaminated areas of the buildir.g. The supply , are the radwaste control room and the balance of the air system consists of a prefilter, heating coil, cooling I radwaste building, coil, and two 1(XG supply fans. One fan is normally o operating and the other fan is on standby. The supply (j A positive static pressure with respect to the fan furnishes conditioned air through ductwork and balance of the building and to atmosphere is diffusers, or registers to the work areas of the maintained in the radwaste control room. The building. Zone preheat coils installed in the supply balance of the radwaste building is maintained at a air ductwork provide ternperature control. Air from l negative static pressure with respect to atinosphere, the work areas is exhausted through the tant and pump rooms. Thus, the overall airflow pattern is R from the least potentially contaminated areas to the
$ most contaminated areas.
l The system design is based on outdoor summer The exhaust air system consists of two 100% matimum of 115"F. Summer indoor terrperatures exhaust fans, one normally operating and one on include 75"F in the radwaste control station, w"F in standby. Eshaust air from the silo, waste fdter rooms, operating atcas and corridors, a maximum oil separator room and the mixing and filling station is temperature of 104"F in areas that may be occupied monitored for airborne radioactivity. Under normal and 110"F in the equipment cells. Winter indoor conditions with no contamination, normal ventilation l design temperatures include 60"F in occup"ied areas, in the same circuit as the other spaces in the building I 70"F in the radwaste control room and 60 F in the is maintained. Each of the above.noted spaces is ! b separately monitored. A high level of radioactivity equipment cellig ased on an outdoor design temperature of 40 F. activates an alarm in the main control room, simultaneously isolating the effected space. The 9.4.6.2 Splem Description exhaust air is exhausted through the main plant stack. 9.4.6.2.1 Radwaste llullding Control Honm g R llcating, cooling and pressuri7ation of the control e room are accomplished by an air conditioning Amendment 17 944
ABWR m ean Standard Plant mn 9.4.6J Safety Daluation building are started manually. The fan inlet dampers open when the fan is started. A flow switch installed Although the 11VAC system is not t.afety related as in the exhaust fan discharge duct actuates an alarm on defined in Section 3.2, several features are provided indication of fan failure in toe main and radwaste e to insure safe operation. A completely separate control rooms and automatically starts standby fan. l h IIVAC System is provided for the control room. The exhaust fan is inteilocked with the supply fan to
- Pressure control fans for radwaste areas are prevent the supply fan from operating if the exhaust redundant, with provision for automatic start of the fan is shutdown.
standby unit. Radiation detectors and isolation dampers are provided to permit isolation and Two pressure. indicating controllers modulates containment of any radioactive leakage. variable inlet damper vanes in the supply fan to tuaintain the area at a negative static pressure with 9.4.6.4 Tests and Inspections respect to atmosphere. The switch causes an alarm to be actuated if the negative pressure falls below the The system is designed to permit periodic preset limit. inspection of important components, such as fans, motors, belts, coils, filters, ductwork, piping and Differential pressure indicators measure the valves, to assure the integrity and capability of the pressure drop across the filter section. The switch l system. Local display and/or indicating devices are causes an alarm to be actuated if the pressure drop provided for periodic inspection of vital parameters exceeds the preset limit. such as room temperature, and test connections are provided in exhaust filter trains and piping for periodic checking of air and water flows for conformance to the design requirements. Portable Radiation monitors are installed in the radwaste test and monitoring equipment is available to building vent. A high radiation signalin the vent balance the system when required. causes both a n.umming alarm and an audible alarm to annunciate in the main control room with an audible alarm sounding and a display light showing on the radwaste building IIVAC control panel. In addition, the branch high radiation signal automatically closes the branch isolating damper so that air conditioning is 9.4.6.5 Instrumentation Application continued in the balance of the building. 9.4.6.5.1 Radwaste llullding Control Hoom if the vent high radiation alarm continues to annunciate, the work area branch ducts are manually The air conditioning unit for the control room is isolated selectisely to locate the affected building started manually. A temperature indicating area. Should this technique fail, because the airborne controller modulates the air conditioning system via radiation has generally spread throughout the a three way hot water valve to rnaintain space building, control room air conditioning continues conditions. A differential pressure indicating operating. Ilowever, the air conditioning for the for controller modulates dampers in the return air the balance of the buildingis shut down. The opera-ductwork and the room damper to maintain the tors, using approved plant health physics procedures, positive static room pressure. Differential pressure then enter the work areas to locate and isolate the indicators measure the pressure drop across the filter leakage source. bank. The supply and exhaust air ductwork have manually balancing dampers prosided in the branch ducts for i balancing purposes. The dampers are locked in place I after the system is balanced. 9.4.6.5.2 Radwaste flullding Work Areas The air exhaust and supply fans for the radwaste Amendment 17 9 02L
ABWR nwm nty ri Stand.ard Plant
,, SECTION 9.5
( ;
'd CONTENTS Src11on lille hice 9.5.1 Err ProteetlenSnitnu 9.51 9.5.1.0 Plant Features Enhancing Fire Toleranec 9.5 1 9.5.1.0.1 Plant Arrengement 9.5-1.0 9.5.1.0.2 Divisional Separation 9.5-1.0 9.5.1.03 Fire Containment System 9.51.0.1 9.5.1.0.4 Cornbustable leading 9.5-1.0.2 9.5.1.0.5 1iVAC Systems 9.5 1.03 9.5.1.06 Smoke Control Systems 9.5 1.03 9.5.1.0.7 Spurious Control Actions 9.51.0.5 9.5.1.0.8 Support Systems 9.5 1.0.6 Fire Alarm Sptems
(]
'v' 9.5.1.0.9 9.5-1.0.6 9.5.1.0.10 Fire Suppression System 9.5 1.0.6 9.5.1.0.11 Personnel Access Routes 9.51.0.6 l
9.5.1.0.12 Manual Fire Suppression Activities 9.5 1,0.6 9.5.1.1 Design 11 asis 9.51.0.6 9.5.1.2 Systems Descriptions 9.5-1,1 9.5.1.2.1 General Description 9.51.1 9.5.1.2.2 Fire Suppression System Requirements 9.51.1 9.5.1.23 Codes and Standards 9.51.1 9.5.1.2.4 Protection of Operating Units 9.5-13 9.5.1.2.5 General Description of Fire Protection System 9.5 13 9.5.1.2.6 Protection and Extinguishing Equipment for Safety-Related Equipment 9.5 13 (q
/ % s) 9.5-li Amendment 17
i ABWR mamn Standard Plant nty. 9 SECTION 9.5 g CONTENTS (Continued) Ecction 11 tic Eage 9.5.1.2.7 Design Features of Fire Detection and Suppression Systems 9.5-1.4 9.5.1.2.8 Smoke Control 9.51.4 9.5.1.2.9 Fire Alarm System 9.51.4 9.5.1.2.10 Electrical Cable Fire Protection 9.51.5 9.5.1.2.11 Fire Separation for Safe Shutdown 9.51.5 9.5.13 Safety Evaluation (Fire llazard Analysis Append!: 9A) 9.51.6 9.5.1.4 Inspection and Testing Requirements 9.51.7 9.5.1.5 Personnel Qualifica* ions and Training 9.51.7 9.5.1.5.1 Fire Protection Engineer 9.5-1.7 9.5.1.5.2 Ouality Assurance (OA) Program 9.51.7 9.5.1.53 Emergency Response Plan 9.5-1.7 9.5-ii.1 Amendment 17
ABWR unam^n Rt v ti Standard Plant Q SECTION 9.5 CONTENTS (Continued) Sectlon Iltle l' age 9.5.2 Communication Systeln3 93-2 9.91 Design Bases 9.5-2 i.2.1.1 Power Actuated Paging System 9.5-2 9.5.2.1.2 Sound Powered Telephone System 9.52 9.5.2.2 Description 93 2 9.5.2.2.1 Paging Facilities 93-2 9.5.2.2.2 Sound Powered Telephone System for Plant Ma'atenance and Repair 9.5-2a 93.23 System Operation 9.5-2b 95.2,4 Safety Evaluation 9.5-2b 93.2.5 Inspection and Testing Requirements 9.5-2b 9.5.2.6 Portable and Fixed Emergency Communication 9.5 2b 9.53 1lehtine and Senicine Poveer Surmly 9.5-3 Systems 9.53.1 Design Bases 9.5-3 9.53.1.1 General Design Bases 9.53 9.53.1.2 Safety-Related Design Bases 9.5-33 9.53.2 System Description 9.5-33 9.53.2.1 Normal (Non Essential) Lighting 9.5-33 9.53.2.2 Standby (Essential) Lighting 9.5 33 9.53.23 Emergency (DC Essential) Lighting 9.5 3 4 9.53.2.4 Guide Lamps With Self Contained Battery Packs 9.5-3.4 O 9.5-lii Amendment 16
ABM 2mioo^n
- Standard Plant - Rirv. n SECTION 9.5
\
CONTENTS (Continued) Section Tillt East 9.5.11 Combustion Turbine / Generator 9.5 10.1 9.5.11.1 Design Basis 9.5-10.1 9.5.1'1.2 . System Description 9510.2 9.5.11 3 Safety Evaluations - 9.5-10.2 , 9.5.11.4 Tests and Inspections 9.5-103 9.5.11.5 Instrumentation Requirements 9.5-103
- 9.5.12 lower 17rmell Flooder 9.5 103' 9.5.12.1 - Design Basis 95103 9.5.12.2 System Description 95103 1 9.5.123 Safety Evaluation 9.5 10.4 -O -9.5.12.4 Testing and Inspection Requirements 9510.5 V
9,5.12.5 Intrumentation Requirements 9.5-10.5 l p l 9.5.13 . Inittfaces 9.5-10,5 I 9.5.13.1 ' Contamination of the DG Combustion Air Intake 9.5-10.5 9.5.13.2 Use of Communication System in Emergencies ' 9.5 10.5 9.5.13 3 Maintenance and Testing Procedure for l Communi:ation Equipment' 9510.5 i 9.5.13.4 - Use of Portable Hand Light in Emergency 9510.5 L 9.5.13.5 . Vendor Specific Design of Diese! Generator Auxiliaries 9.2-10.6 j 9.5.13.6 Diesel Generator Cooling Water System Design Flow and Heat Removal Requirements 9.5-10.6 9.5.13.7 Fire Rating for Penetration Seals 9.5-10.6 9.5.13.8 Diesel Gr.nerator Requirements 9.5-10.6 l 9.5-v.1 Amendment 16 I-
ABWR 23xciooxii nev. n Standard Plant l l SECTION 9.5 g CONTENTS (Continued) 11 tic Page Srdion 9.5.13.9 Applicant Fire Protection Program 9.5 10.6 9.5.13.10 HVAC Pressure Calculations 9.5-10.7 9.5.13.11 Plant Security Systems Criteria 9.5 10.7 9.5.13.12 Fire Hazard Analysis 9.5-10.7 9.5.14 References 9.5-10.7 0 9.5-v.2 9 Amendment 17
-__.--..._ _ _ . .....-..m..-. .___. _, .. _. _ _ - . . .
.ABMR 2 W 100All Standard Plant - ntw. n Q SECTION 9.5 TABLES Table lidt East 951. Illuminating levels 9.5-10a 952 ' Lighting and Power Sources 9510c
- 953- - Standby Lighting 9.510d ;
I
.954 Emergency ? 'p! ** 3 9510e 955 Summary of nutomatic Fire Suppression
- Systems _ -
9.510f 3 ILLUSTRATIONS Figure Ihlt Eagt 951- Suppression Pool Cleanup System P&lD = 9511 9.5-2 Outline Telephone Communication Systems 9.5-12 953- ' Lower Drywell Flooder System Arrangement / Configuration 9.5-13 , 9.5-4 Fire Protection Water Supply System 9514
. 955 Fire Protection Yard Main Piping 9516 956 Standby Diesel Generator Fuel Oil and -
, intake and Exhaust System 9.5 18 l, J 957 Standby Diesel Generator Jacket Cooling Water System - 9519 958 Standby Diesel Generator Starting Air System 9520 959 ~ ' Standby Diesel Generator Lubricating Oil System 9521' l. L
- 9.5-v.i Amendment 17 m..f.g, g. -,a3-,--.4 ..,.-.-y..,, _ .._ _, ,Ly. em e ,,e.W e e m *Trw W ' D'-- %.- ' - --'----- - - - - - -- -
- ABWR urcman-Standard Plant nuv. n
= 9.51OTHER AUXILIARY SYSTEMS
= 9.5.1--' Fire Protection System -
t GE PROPRIETARY ' provided under separate cover Eagg ' Amendment
-9.5-1 -17
.5 . 1 17 9.5-1.0.2 17
-9.5-1.03 17 9.51.0.4 17 9.5-1.0.5 17 9.5-1.0.6 17 9.5 1.0,7 17 9.5-1.1 -
- - 9.5-1.2' '
9.5-13 16 9.51,4 17 l 9.5-1.5 16 9.5-1.6 17
-9.51.7 17
( I
- Amendment 17 9.51-17
'We- + ~ '
ABWR uwmru l Standard Plant REV 11 1 (4) Ultimate heat sink be conducted. Any non compliance shall be
, documented as being required and acceptable on 6 The applicant's fire protection program shall the basis of the Fire lla7ard Analysis, Appendix $ comply with the SRP Section 9.5.1, with ability 9A, and the Fire llazard Probabilistic Risk to bring the plant to safe shutdown condition Assessment, Appendix 19M.
following a complete fire burnout without a need for recovery, 9.5.14 References 9.5.13.10 !WAC Pressure Calculations 1. Stello, Victor, Jr., Design Requirements Related To The Evolutionary Advanced Light F3 The applicant referencing the ABWR design iPater Reactors (AllPRS), Policy Issue,
$ shall provide pressure calculations and confirm SECY-89 013, The Commissioners, United capability during pre.. operational testing of the States Nuclear Regulatory Commission, smoke control mode of tiie llVAC systems as January 19, 1989.
l described in Subsection 9.5.1.0.6.
- 2. Cote, Authur E., NFPA Fire Protection 9.5.13.11 Plant Security Systems Criteria Handbook, National Fire Protection Association, Sixteenth Edition.
The design of the security system shall include an evaluation of its impact on plant 3. Design of Smoke Control Systems for operation, testing, and maintenance. This Buildings, American Society of IIcating, evaluation shall assure that the security Refrigerating, and Air Conditioning restrictions for access to equipment and plant Engineers, Inc., September 1983. regions is con.patible with required operator actions during all operating and emergency modes 4. Recommended Practice for Smoke Control of operation (i.e., loss of offsite power, access Systems, NFPA 92A, National Fire Protection for fire protection, health physics, maintenance, Association,1988. d testing and local operator). In addition, this evaluation shall assure that-h* (a) There are no areas within the Neclear Island where communication with central and secondary alarm stations is not possible; (b) Portable security radios will not interfere with plant monitoring equipment; (c) Minimum isolation zone and protected area illumination capabilities cannot be defeated by sabotage actions outside of the protected area; and, (d) Electromagnetic interference from plant equipment startups or power transfers will not create nuisance alarms or trip security access control systems. 9.5.13.12 Fire llazard Analysis A compliance review of the as built design against the assumptions and requirements stated in the fire hazard analysis (Appendix 9A) shall O I Amendment 17 9510.7 l
Nd e i e e j Table 9.5-5 m
> E H
y
SUMMARY
OF AUTOMATIC FIRE SUPPRESSION SYSTEMS 1.h "1 E ARFA NAME DIV COMBUSTIBII SPRINKIIR SYS*EM TYPE I BLDG E11V RM. SO. 11RE AREA
*3 IC1110 Division *1* IIVAC Chase D1 Cables Wet pipe spnnkler [
GI -1450 319 D2 CaMes Wet pipe synnkler & CII -1450 324 101210 Division *2" IIVAC Chase Cables Wet ppe sprinkler Division *3* IIVAC Chase D3 CD -1450 335 FC1310 Wet pipe sprinkler IIVAC *A* Supply Roarn DI Bag Miter Cables CH 7350 513 FC1110 Wet pepe sprinkler D1,2,3,4 CaMes CH 7350 522 1 0 4910 CR IIVAC *H" thhaust Doct Oase Bag Etter. Canes Wet pipe sprinkler IIVAC *C" Supply Room D3 CII 7350 532 FC1310 Wet pipe *prinkler CR IIVAC *C" thhaust Doct case DI.2.3.4 None CH 7350 595 IC4910 Wet pipe sprinkler CR IIVAC Supply "C" D3 Bag filter, Canes CU 122m 615 IC4910 Wet pipe sprinkler CR IIVAC Supply *IP D2 Bag Riter, Cables Cil 12200 621 FC4910 Wet pipe sprinkler IIVAC *Ir suppfy Room D2 thg niter.CaMes Cil 12200 624 1C1210 Deluge water Unit AuxiliaryTraasformer ND Od PY 7350 N/A N/A Deluge water ND Od PY 7350 N/A N/A Main Transformer Area Ort Deluge water Reserve Transformer ND PY 7350 N/A N/A Dry pipe. Comed head ND Oass 111 B lobe oi1& canes RH -8200 133 F1300 CRD Pump Roorn Fuel cat. IAc oil, & cables Preaction Ibem-water Diesel Generator A Rrom D1 RB 123no 412 F4100 Preaction Ibem-water D2 Fact cit, Lube oil, & canes RH 12300 423 F4200 Diesel Generator il Room Lubricant, Fuel oil (tansient) Deluge water Ilatch ND Ril 123n0 430 F4301 Fuct m1,1Ae oil, & cables Preaction Fuam-water Diesel Generator C Pam D3 RH 12300 432 F4300 Deluge s;winkkr ND I#ricant,1%el oil (transient) RB 18100 $30 F430' Itatch DI Diesel fuel Deluge Foamwer RB 23500 610 F6101 Diesel Generator FuelTank A Room DI Dag Hiter Wet pipe sprinkkr RB 23500 611 F4100 D/G (A) Emergency Supply Fan Diesel fuel Defuge Ibum-water Diesel Generator Fuel Tank B Roorn D2 Ril 2350P 620 F6201 Bag nher Wet pepe sprinkler D/G (B) Emergency Supply Fan D2 Ril 23500 629 F4200 Deluge Foam-==ter l D3 Diesel fuel RH 23500 630 F6301 Diesel Generator Fuel Tank C Room I Dag Hker Wet pire sprmkler D/G(C)Ilmergency Supply Fan D3 RU 235M 633 F4300 Wet pipe sprinkler ND Bag Dher RB 23500 645 F4400 Supply Fan for Monitoring Room nag Rher Wet pipe spnnkler D/G (A)fZ Supply Fan D1 RB 27200 653 F4100 Wet pipe spnnkler D3 Bag *?er RB 27200 M3 F4200 D/G (C)/Z Supply Fan Dag nher Wet pipe spnnkler D/G (B)/Z Supply Fan D2 Ril 27200 673 F4100 Wet pipe sprinkler DI Bag Diter RB 31700 710 F4IW Rip (B) Supply Far. D2 Dag Hfrer Wet ppe sprinkler RB 31700 740 I%200 Rip (A) Supply Fan Radioactive material Wet ppe sprinkler y Drv Radioactive Waste Storage Area ND RW 1600 N/A N/A ND Rad oective material Wet pipe spnnkler Qg N/A Dry Radioactive Was;e Storage Area y RW '7300 N/A ND Radioactne material Wet ppe spnnkler .:: { E RW 200 *:/A N/A Dry Radioactive Waste Storage Area a
Table 9.5-5 (Continued) ro g EI a " y
SUMMARY
OF AUTOMATIC FIRE SUPPRESSION SYSTEMS @'s
,s e
AREA NAME DIV COMBtJSTT1 fill SPRINKIER5WIDf 7TPE I G llIDG EIEV RM. NO. 11RE AREA Wet pipe sprinkler [ Dry Radioactive Waste Storage Area ND Radioactive matenal RW -690 N/A N/A 3 Sli 7100/ Wet pipe sprir.kler j "Ihroughout the Service Dailding ND Paper, furniture. & etc. 11750 All ND I.mbricants. charcost.& cables Wet pipe sprmkler 350 120 IT1500 Ikneath the Turbine sorroundings .
'111 DI.2.3,4 Cables Wet pipe sprinkler 1 11 7350 219 FI2505 Steam Tunnel ND I.ubncanes & cables Wet gwpe sprinkler Til 7150 222 FI 1500 llencath she Turbine surroundmgs ND Cass III B tube oil Deluge Foam-water
'Ill 73s0 230 FI2500 1.ube oil conditioning area ND Imbricants.Feeloil & cables Preaction sprinkler 111 7350 247 FF2503 Auxiliary Iknier area ND Dreselfuel Deluge Foam +ater 1 11 15150 317 l'I35to GasTus1*ine Generator ND I fydrogen seal col Deluge Ibem+ater
'l il 15150 320 FI150) 'I CW Pumps area ND Imbricants. A rables Wet pipe sprinkler 1 11 15350 321 F11540 llencath the Turbine surroendmgs ND Oass I!! D lobe oil Deluge Foameter lit is350 330 FI1501 Imbe od Rer.ervoir area l
l l 5-E a$ 9 O O t
O APPENDIX 9B
SUMMARY
OF ANALYSIS SUPPORTING FIRE PROTECTION DESIGN REQUIREMENTS O O
ABWR uxwan Standard Plant un_.a APPENDIX 9B ] TABLE OF CONTENTS Section 31tle Eagt 9B
SUMMARY
OF ANALYSIS SUPPORTING FIRE PROTECTION DESIGN REQUIREMENTS
98.1 INTRODUCTION
9B.1 1 98.2 FIRE CONTAINMENT SYSTEM 98.2.1 Fire Types 9B.2-1 9B.2.2 Fire Barriers 9B.2-1 9B.23 Allowable Combustible Loading 9B.'2-2 9B3 REFERENCES 983-1 O O Amendment 17
ABWR . 2w.iman fittindard Plant xm n 9
11.1 INTRODUCTION
This appendix is included to discuss in detall some of the analysis associated with the design decisions and requirements stated in Subsection 9.5.1. O O Amendment 17 9n11
ABWR m aman Standard Plant nean SECTION 9B.2 CONTENTS S.REll0_R Illlt Eagt 98.2.1 Fire Tmes 9B.2-1 98.2.2 Fire Barriers 9B.2-1 9B.23 Allowable Combustibic loading 9B.2 2 9B.2.3.1 Permanent Loading 9B.2-2 9B.23.2 Transient Combustibles 9B.2-4 9B.2.3.3 Cable Trays 9B.2 4 TAHLES Table liite Eage 9B.2-1 Estimated Fire Severity for Offices . and Light Commercial Occupancies 98.2 8 9B.2-2 Fire Severity Expected By Occupancy 9B.2 9 98.2-3 Cable Type And Configuration For UL Tests 9B.2-10 9B.24 Summary Of Burning Rate Calculations 9B.211 ILLUSTRATIONS Elgure Iltle Eage 9B.2-1 Possible Classification of Building Contents for Fire Severity and Duration 9B.212 O - Amendment 17
ABWR usamn iutv. n Standard Plant 911,2 FIRE CONTAINMENT SYSTEM GE PilOPitlETAltY provided under separate cover lhtgt Amendment
!'11.2-1 17 9B.2 2 17 9 11.2 - 3 17 O 9B.2-4 90.2 5 17 17 9 11.2 - 6 17 9B.2 7 17 O Ame n iment 17 9112 1 7
ABWR maan net n Standard Plant Table 9B 21 ESTIMATED FIRE SEVERITY FOR OFFICES AND L1 Gilt COMMERCIAL OCCUPANCIES
- Data applying'to fire-resistive bullidings with combustible furniture and shelvinf, Equivalent Fire Severity Approximately Combustible Content IIent Potential equivalent to that of Total, including finish, Assumrd ** test under standard cune floor, and trim, psf Btu per sq ftt for the following periods:
5 40,000 30 min 10 80,000 1 hr 15 120,000 11/2 hss 20 160,000 2 hrs 30 240,000 3 hrs 40 320,000 41/2 hrs 50 330,000 7 hrs 60 432,000 8 hrs 70 500,000 9 hrs O Reproduce <1 from Table 7 9B, NFPA Fire Protection Handbook, Reference 1.
" Heat of combustion of contents taken at 8,000 Btu per ib up to 40 psf; 7,600 Blu per Ib for 50 lb, and 7,200 Btu for 60 lb and more to allow for relatively greater proportion of paper. Th: weights contemplated by the tables are those of ordinary combustible materials such as wood, paper, or textiles.
2 2 2 tSI units:1 psf = 4.9 kg/m ;1 Blu/ft = 1.14 J/m O Amendment 17 98.2 8
ABWR 2-man Standard Plant nev n Table 911.2 2 g FIRE SEVERITY EXPECTED IlY OCCUPANCY Temperature Curie A (Silght) Well arranged office, metal furniture, noncombustible building. Welding areas containing slight combustibles. Noncombustible power house. Noncombustible buildings, slight amount of combustible occupancy. Temperature Curve B (Moderate) Cotton and waste paper storage (baled) and well arranged, noncombustible building. Paper making processes, noncombustible building. Noncombustible institutio.:al buildings with combustible occupancy. Temperature Curve C (Moderately Senere) Well arranged combustible storage, e.g., wooden patterns, noncombustible buildings. Machine r, hop having noncombustible floors. Temperature Cune D (Severe) Manufacturing areas, combustible products, noncombustible building. Congested combustible storage areas, noncombustible building. Temperature Cune E (Standard Fire Exposurt -Sesere) Flammable liquids. Woodworking areas. Office, combustible furniture and buildings. Paper working, printing, etc. Furniture manufacturing and finishing. Machine shop having combustible floors, a I
- 1. Reproduction of Table 7-9E, NFPA Fire Protection Handbook (Reference 1).
- 2. See Figure 98.21 for the temperature curves identified in this table.
O Amendmer.t 17 9D29
ABWR- mamn Standard Plant am n - Table 9H.2 3 - CABLE 'IYPE AND CONFIGURATION FOR UL TESTS * - Cable Type Cables Per Combustible Tray 1 ft Leading - Wide Blus Per Sq Ft of Tray 7/C #14AWG XLPE FR 94 141523 7/C #14AWG Tefzel 94 22508 7/C #14AWG Tefzel 202 48368 19/C #14AWG XLPE FR 37 135737
- 19/C #14AWG Teftel 37 17579 19/C #14AWG Teft.el 58 27556 l ?O i
l^ l {' O. (Referenec 1) Amendment 17 9D.210 l
ABWR - maan Standard Plant _ nev n Table 9B.2-4 g.
SUMMARY
OF BURNING RATE CALCULATIONS hlaterial Source of Data BurnIne Rate B u r n I n c. R a t e (Btus/ min per sq ft (Btus/ min per sq ft of surface aiva) of eable tray or floor) Cross-linked polyethylene F1VE bench 918 2884 scale burning data (Ref. 2) Cross-linked polyethylene Estimated from S8S to 1062 1847 to 3336 tests at UL (Ref. 5) Tefzel Estimated from 1% to 385 616 to 1209 tests at UL (Ref. 5) Ventilation limited F1VE P1 ant 72.5 (Three air changes per hour) Screening (Btus per min per Guide, sq ft of floor total Equation 47 of area) a t t a c b . 10.7 (Ref. 2) Design normal maximum limit Typieai for 356 power houses (Btus per min per (Ref.1) sq ft of floor total area) ASTM E 119 1333 Fire barrier capability curve for three (Btus per min per hours sq f: ci floor total area) O Amendment 17 90.2 11
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ABWR uuuan Standard Plant an n 9B.3 REFERENCES O 1. Stello, Victor, Jr., Design Requirements Related To The Evolutionary Advanced Light Water Reactors (ALWRS), Policy issue, SECY-89-013, The Commissioners, United States Nuclear Regulatory Commission, January 19,1989
- 2. Cote, Arthur E., NFPA Fire Protection Handbook, National Fire Protection Association, Sixteenth Edition
- 3. C. F. Draun & Co., TVA STRIDE Fire Hazard Analysis, Project 4S&P, Rev.1, General Electric Company, May 1977
- 4. Professional Loss Control, Fire Vulnerability Evaluation Methodology (FIVE) Plant Screening Guide, Draft, EPR17.REV, Contract No. RP 3000-41, Electric Power Research Institute, Palo Alto, CA,1990
- 5. Flame Tests , A report on tests conducted by Underwriters Laboratories, lue., at Northbrook, Illinois, September 27,28, and 29,1976, E.1. Du Pont De Nemours & Co. (lue.) E-12952 O
9113-1 Amendment 17
ABWR nuimu Sipndard Plant m O cuirTERio TABLE OF CONTENTS Section llil.c . Eagt 10 STEAM AND POWER CONVERSION SLTEM 10.1
SUMMARY
DESCRilTION 10.1 1 10.1.1 Protective Features 10.1 2 10.2 TURBINE GENERATOR 10.2-1 10.2.1 Design Bases 10.2-1 10.2.2 Description 10.2 2 10.23 Turbine Disk integrity 10.2-6 10.2,4 Evaluation 10.2-8 10.2.5 References 10.2-9 103 M AIN STEAM SUPPLY O 103.1 Design Bares 103 1 10 3.2 Descrircon 10 3-2 1033 Evaluation 10 3-4 103.4 Inspection and Testing . Requirements 103 5 10 3.5 Water Chemistry (PWR) 10 3-5 103.6 Steam and Feedwater System Materials 10 3-5 10.4 OTHER FEATURES OF T!!E M AIN STEAM AND POWER CONVERSION SYSTEM 10.4.1 Main Condensers 10.4 1 10.4.2 Main Condenser Evacuation System 10.4-4 10.4 3 Turbine Gland Scaling System 10.4-5.1 l O 10-il j i Amendmcat 17
ABWR nasima Standard Plant nev, A CHAPTER 10 $ ] TABLE OF CONTENTS (continued) Sectlon 1 1112 Eagt 10.4.4 Turbine Dypass System 10.4-7 10.4.5 Circulating Water System 10.4-9 10.4.6 Condeusate Cleanup System 10.4 10 10.4.7 Condensate and Feedwater System 10.4-13 10.4.8 Steam Generator Blowdown Systcm (PWR) 10.4-17 10.4.9 Auxiliary Feedwater System (PWR) 10.4-17 10.4.10 Interfaces 10.4-17 O e Amendment 17
l ABWR moom Standard Plant an A (1) Forgings are rough machined with minimum G material samples using correlation methods which stock allowance prior to heat treatment. l
-\j are as conservative, or more so, than those presented in Reference 1.
(2) Each finished machined rotor is subjected to l Turbine operating procedures are, employed to 100-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 FATT, and (b) as denned above, sufficient to main- restrictive than those specificd for Class 1 tain the fracture toughnen to tangential sitess ratio components in the AShtE Boiler and Pressure at or above 2 /mch. Vessel Code, Sections til and V, and include the requirement that subsurface sonic indications 10.233 liigh 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 l 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 \ consistent with ASThi recommendations. spin tested at 20-percent overspeed. 10.23.4 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 ( N 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. Ilydro-l (1) Turbine shaft bearings are designed to retain static tests are performed on all coolers, All piping is their structural integrity under normal pressure tested for leaks. hiotor-operated valves are operating loads and anticipated transients, factory leak tested and inplace tested once installed. including those leading to turbine trips. 10.23.6 Insenice Inspection l(2) The multitude of natural critical frequencies of the turbine shaft assemblies existing between The inservice 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, cause no distress to the unit during operation. such as couplings, coupling bolts, turbine shafts, low-pressure turbine buckets, low-pressure and 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 gradients does not exceed 0.75 of the yield ash 1E Section 111 components, as required by the strength of the materials at 115 percent of rated AShiE Boiler and Pressure Vessel Code, Section XI, speed. turbine inspection is performed in sections during the refueling outages so that in 10 years total inspection 10.23.5 Presenice inspection has been completed at least once. The preservice procedures and acceptance cri- This inspection consists of visual and surface teria are as follows: examinations as indicated below: Arnendment 8 10 2-7 )
1 l ABWR uuiom Standard Plant wA (1) Visual examination of all accessible surfaces (b) Erosion of valve seats and stems of totors (c) Deposits on stems and other valve parts (2) Visual and surface examination of all low- which could interfere with valve operation pressure buckets (d) Distortions, misalignment (3) 100-percent visual examination of couplings and coupling bolts inspection of all valves of one type will be conducted if any unusual condition is discovered The inservice inspection of valves important to overspeed protection includes the following: 10.2A Evaluation (1) All main stop valves, control valves, extrac- The turbine-generator is not nuclear safety tion nonreturn valves, and CBIVs will be related and is not needed to effect or support a safe tested under load. Test controls installed on shutdown of the reactor, the main control room turbine panel and permit full stroking of the stop valve, control The turbine is designed, constructed, and in-valves, and CBIVs. Valve position indication spected to minimize the possibility of any major is pr"vided on the panel. No load reduction component failure. is necessary before testing main stop and control valves, and CBIVs. Extraction nonte- The turbine has a redundant, testable overspeed turn valves are tested by equalizing air trip system to minimize the possibility of a turbine pressure across the air cylinder. Movement overspeed event. of the valve arm is observed upon action of the spring closure mechanism. Unrestrained stored energy in the extraction steam system has been reduced to an acceptable (2) Main stop valves, control valves, extraction minimum by the addition of nonreturn valves in nonreturn valves, and CBlVs will be tested at selected extraction lines. least once a week by closing each valve and observing by the valve position indicator that The turbine-generator equipment shielding re-l it moves smoothly to a fully closed position. At least once per month, closure of each quirements and the methods of access control for all areas of the turbine building ensure that the dose main stop valve, control valve and CBIV criteria specified in 10CFR20 for operating personnel during test will be verified by direct observa- are not exceeded, tion of the vah*e motion. All areas in proximity to turbine generator Tightness tests of the main stop and control equipment are zoned according to expected valves are performed at least once per main- occupancy times and radiation levels anticipated tenance cycle by checking the coastdown under normal operating conditions. characteristics of the turbine from no load 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 scheduled so that each valve is inspected at 3 will be reduced by administrative controls enacted by 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 will be 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 turbine exhaust hood and the condenser is made by means of a stainless steel expansion joint. Arnendment 17 1028
MM . 23A6100AJ ' Standard Plant _ nm A _W
- welds in locations of restricted direct physical and visual accessibility.
(a) The performance qualification should require testing of the welds when condi-tions of accessibility to production welds are less_than 30 to 35 cm (12-14 inches) in any direction from the joint. (b) RequaMeation is required for different accessibility conditions or when other
. essential variables listed in the Code, Section IX, are changed.
. (c) The qualification and requalification tests required by (a) and (b) above may be waived provided that the joint is to be -
100% radiographed or ultrasonically exam-Ined after completion of the weldment. Examination procedures and acceptance standards should meet the requirements of the ASME Code Section 111. Records of
. the examination reports and radiographs should.be retained and made part of the Quality Assurance documentation of the completed weld.
(2) ' Regulatory Guide 1.37, Quality Assurance Requirementsfor Cleaning of Fluid Systems and Associated Components of Water Cooled -
. Nuclear Power Plants, describes acceptable -
procedures for cleaning and handling Class 2 components of the steam and feedwater systems. Vented tanks with deionized or demineralized water are an acceptable source of water for final cleaning or flushing of
- finished surfaces. The oxygen content of the water in these vented tanks need not be controlled.
(3) ' Acceptance criteria for nondestructive exami-nation of tubular products are given in the ASME Code, Section 111, Paragraphs NC 2550 through 2570. i Amendment 17 10.3-3 _ - - . _ . . __ --- - _ _ _ - - ._ _ . . . ~ .- . . . -
ABWR nuiow Rev.A Standard Plant I !O SECTION 10.4 CONTENTS
&ction Etic Ibn 10.4.1 Main Condensers 10.4-1 10.4.1.1 Design Bases 10.4-1 10.4.1.1.1 Safety Design Bases 10.4-1 10.4.1.1.2 Power Generation Design Bases 10.4-1 10.4.1.2 Description 10.4 1 10.4.1.2.1 General Description 10.4-1 10.4.1.2.2 Component Description 10.4-2 10.4.1.23 System Operation 10.4 2 10.4.1 3 Evaluation 10.4-3 10.4.1.4 Tests and Inspections 10.4-3 10.4.1.5 Instrumentation Applications 10.4-3 10.4.1.5.1 Ilotwell Water Level 10.4-3 10.4.1.5.2 Pressure 10.4-3 10.4.1.5 3 Temperature 10 4-4 10.4.1.5.4 Leakage 10.4-4 10.4.2 Maln Condenser Evacuation System 10.4 4 10.4.2.1 Design Bases 10.4 4 10.4.2.1.1 Safety Design Bases 10.4-4 10.4.2.1.2 Power Generation Design Bases 10.4 4 10.4.2.2 Description 10.4-4 10.4.23 Evaluation 10.4-5 10.4.2.4 Tests and Inspections 10.4-5 O
10A-ii l Amendment 3
-ABWR na6ima Standard Plant nev A SECUON 1.0.4 h CONTENTS (Continued)
Section Title Page 10.4.2.5 Instrumentation Applications 10.4-5 10.4.2.5.1 Steam Jet Air Ejectors 10.4-5 10.4.2.5.2 Mechanical Vacuum Pump 10.4-5 10.4 3 Turbine Gland Seal Sntem 10.4-5.1 l 10.43.1 Design Bases 10.4-6 10.4 3.1.1 Safety Design Bases 10.4-6 10.4 3.1.2 Power Generation Design Bases 10.4-6 10.4 3.2 Description 10.4-6 10.4 3.2.1 Genetal Description 10.4-6 10.4 3.2.2 System Operation 10.4-6 10.433 Evaluation 10.4-6 O 10.4 3.4 Tests and Inspections 10.4-6 10.4 3.5 Instrumentation Application 10.4-7 10.43.5.1 Gland Steam Condenser Exhausters 10.4-7 10.43.5.1.1 Pressure 10.4-7 10.43.5.1.2 Level 10.4-7 10.43.5.13 Effluent Monitoring 10.4-7 l 10.4 3.5.2 Sealing Steam Header 10.4-7 10.4.4 Turbine Bvnass System 10.4-7 10.4.4.1 Design Bases 10.4 7 10.4.4.1.1 Safety Design Bases 10.4-7 10.4.4.1.2 Power Generation Design Bases 10.4-7 10.4.4.2 Description 10.4-7 10.4-iii Amendment 17
ABWR umam Standard Plant neo SECTION 10A CONTENTS (Continued) 8tCdQIl lillC EllEC 10.4.4.2.1 General Description 10.4-7 10.4.4.2.2 Ccmponent Description 10.4-7.1 10.4.4.2 3 System Operation 10.4-8 10.4.43 Evaluation 10.4 8 10.4.4.4 Inspection and Testing Requirements 10.4-8 10.4,4.5 Instrumentation Applications 10.4 8 10.4.5 Circulatine Water System 10.4-9 10.4.5.1 Design Bases 10.4 9 10.4 5.1.1 Safety Design Bases 10.4 9 10.4 3.1.2 Power Generation Design Bases 10.4-9 10.4.5.2 Description 10.4-9 10.4.5.2.1 General Description 10.4-9 10.4.5.2.2 Component Description 10.4-9 10.4.5.23 System Operation 10.4 9 10.4.5 3 Evaluation 10.4-10 10.4.5.4 Tests and Inspections 10.4 10 10.4.5.5 Instrumentation Applications 10.4-10 10.4.5.6 Flood Protection 10.4-10 l 10.4.6 Condensate Cleanun Snism 10.4 10 10.4.6.1 Design Bases 10.4 11 10.4.6.1,1 Safety Design Bases 10.4-11 10.4.6.1.2 Power Generation Design Bases 10.4-11 10.4 iv Amendment 17
ABWR mam Standard Plant awa SECTION 10A CONTENTS (Continued) Section Title Page 10.4.6.2 System Description 10.4-11 10.4.6.2.1 General Description 10.4-11 10.4.6.2.2 Component Description 10.4-11 10.4.6.23 System Operation 10.4-11 10.4.6 3 Evaluation 10.4-12 10.4.6.4 %sts and Inspections 10.4-12 10.4.6.5 h.strumentation Applications 10.4 12 10.4.7 Condentate And Feedwater System 10.4-13 10.4.7.1 Design Bases 10.4-13 10.4.7.1.1 Safety Design Bases 10.4-13 0 10.4.7,1.2 Power Generation Design Bases 10.4-13 10.4.7.2 Description 10.4-13
~
10.4.7.2.1 General Description 10.4-13 10.4.7.2.2 Component Description 10.4-14 10.4.7.2 3 System Operation 10.4 16 10.4.7 3 Evaluation 10.4-16 10.4.7.4 Tests and Inspections 10.4-16 10.4.7.4.1 Presenice Testing 10.4-16 10.4.7.4.2 Inservice Inspections 10.4-16 10.4.7.5 Instrumentation Applications 10.4-16 10.4.8 Steam Generator Itlowdown System (PWR) 10.4-17 10.4.9 Auxiliary Feedwater System (P%TO 10.4-17 10.4-v Amendment 17
ABWR ma. Standard Plant , . _ _ an. A s_ ,) SECTION 10.4 ! l CONTENTS (Continued) Stfilm litic Eage 10.4.10 Interfneen 10.4 17 10.4.10.1 Radiological Analyds of the TGSS Efauents 10.4 17 t 1 O I 10.4-v.1 Amendment 17 i _ _ _ _ _ _ . _ _ ____.__m. . __ m -______. _ ___
ABWR new Standard Plant mA system, the re.ctor coolant system can still be safely at the resulting static head, inspecting all tube joints, (N d cooled dow using only nuclear island systems. accessible welds, and surfaces for visible leakage and/or excessive deflection. Each ccmdenser water 10.4.13 Evaluation box is to receive a field hydrostatic test with all joints and external surfaces inspected for leakage. During operation, radioactive steam, gases, and condensate are present in the shells of the main 10.4.1.5 Instrumentatiou Applications condenser. The anticipated inventory of radioactive contaminants during operation and shutdown is 10.4.1.5.1 Ilotwell Water Iml discussed in Sections 11.1 and 113. The condenser hotwell water levelis measured Necessary shleiding and (ontrolled access for by twe level transmitters. These transmitters provide the main condenser are provided (see Sections 12.1 signals to ain Indicator, annunciator trip units, the and 123). plant computer, andthe hotwell level control system. Level is controlled by two sets of modulating control liydrogen 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 valve. evacuation of the main condenser. During shut-down, significant hydrogen buildup in the main One set of valves allows water to fimy 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. f t.c level drops below setpoint. If the level increases above another set point, the second set of valves h1ain condenser tubesiae circulating water is located on the discharge of the condensate pumps, treated to limit algae growth and present los pterm open to allow condensate to be pumped back to the corrosion of the tubes and other con nornts. Cor- storage tank. rosion of the outside of the condenser tubing is O
\_/
prevented by maintaining strict water quality u3ing the cond: sate cleanup systern described in Sulnec-10.4.1.8.2 Pressure tion 10.t.o. The construction materials used for IN Condenser pressure is measured by gauges, pres-mair condenser ere selected such that the potential sure switches, and electronic pressure transducers. for corrosion b) galvanic and other effects is The pressure switches provide input signals to the minimized. turbine control system and the annunciator. Two pressure transducers provide input signals 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 wili not tion system and steam bypass and pressure regula. adversely affect .,ny equipment required for sale tion system. In addition, four independent and re-shutdown of the reactor, dundant safety-related pressure transmitters provide inpt.t 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 levels, an annunciator is activated. A further valves. The consequences of a turbine trip are dis-cussed in Subsection 15.23. Should the turbine stop, increase in pressure results in a turb:ne trip. As
- control or bypass valves fail to close on loss of con- pressure increases toward a cornplete loss of denser vacuum, two rupture diaphrams on each tur- .acuum, the main steam isolation salves and the bine exhaust hood protect the condenser and turbine imbi- bypass valves are closed to prevent overpres-exhaust boods against overpressure. surization of the condenser shell.
l l t 10.4.1.4 Tests and Inspections 'Ihe approximate setpoints for these functions a.e as follows: l Each condenser shellis to receive a field hydro-static test before initial operation. This test will (1)liigh condenser pressure turbine alarm at 24 p Q consist of filling the condenser shell with water and. inches flg. Vacuum Amen 6cnt 3 1003 i l
ABWR n-43 , Rn A Standard Plant (2)lligh condenser pressure turbine trip at 22 plant power operation, and to the turbir.a building inches lig vaccum compartment exhaust system at the begirining of l cach start up. {v (3) Bypass valve closure at 12 inches lig vacuum i 10.4.2.1 Design Itases (4) Main steam isolation valve closes at 7 to 10 inches lig vacuum 10.4.2.1.1 Safety Design Itases Condenser pressure is an input to the reacter 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 Itases specific preset valve. Runback is automatically initiated when required to avoid a turbine trip on Power Generation Desien Basis One The MCES high condenser pressure, is designed to remove air and other power cycle non-condensable gases from the condenser during plant 10.4.1.F.3 Temperature startup, cooldown, and power operation and exhaust them to the offgas system or turbine building *e Temperature is measured in each LP turbine compartment exhaust system. , exhaust hood by pneumatic temperature controllers. The controllers modulate a control valve in the water Power Generation Design Unis Two The MCES spray line protecting the exhaust hoods from over- establishes and maintains a vacuum in the condenser heating. during power operation by the use of steam jet air ejectors, and by the mechanical vacuum pump during Circulating water temperatures are moaitored early startup. l upstream and downstream of each con fenser tube bundle and are fed to the plant computer and a main 10.4.2.2 Descriptkm l control room recorder for use during periodic condenser performance evaluations. The condenser evacuation system is illustrated in Figure 10.41, The system consists of two 1004ca- @g 10.4.1.5.4 12aknge pacity, double stage, steam jet air ejector (SJAE) &$ units (complete with intercondenser) for power plant Leakage of circulating water into the condenser operation, and a mechanical vacuum pump for use shellis uonitored 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 normally in l MR 9.3.2. ope ation and the other is on standby. b6 Conductivity of the condensate is continuously During the initial phase of startup, when the monitored at selected locations in the condenser, desired rate of air and gas removal exceeds the Conductivity and sodium are continuously monitored capacity of the steam jet air ejectors, and nuclear at the discharge of the condensate pumps. liigh steam pressure is not adequate 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 10A.2 Main Condenser Evacuation vacuum compartmentpump exhaust is then system routed since there isto thenthe littleturbine 88 buildi System or no effluent radioactivity present. Radiation 66 detectors in the turbine building compartment Noncondensable gases are removed from the exhaust svstem and plant vent alarm in the main power cycle by the main condenser evacuation control room if abnormal radioactivity is detected system (MCES). The MCES removes the hydrogen (see Section 7.6) Radiation monitors are provided and oxygen produced by radiolysis of water in the on the main steam lines which trip the vacuum pump l reactor, and other power cycle noncondensable if abnormal radioactivity is detected in the steam i gases, and exhausts them to the offgas system during being supplied to the condenser. l ' 1044 Amendment 17
ABWR mmm Standitrd Pirmt ._ mA The steam Jct air ejectors are placed in senice 10.4.2.4 Tests and inspections b] / l to reve the gases from the main condenser after a press of about 10 to 15 in lig absolute is estab- Testing and inspection of the system is per-lishe t the main condenser by the mechanical formed prior to plant operation in accordance with vacuum pump and when sufficient nuclear steam applicable codes and staadards. pre 6sure is available. Components of the system are continuously mon. During normal power operation the steam jet !tored during operation to ensure satisfactory perfor-air injectors are norrnally driven by crossaround mance. Periodic inservice tests and ir snections of steam, with the main steam supply on automatic the evacuation system are performed in sonjunction lh standby. The main steam supply, however,is normally used during startup and low load operation, with the scheduled maintenance outages. and auxiliary steam is available for normal use of the 10.4.2.5 Instrumentation Appilcations i steam jet air ejectors during early Startup, should the l mechanical vacuum pump prove to be unasailable. Local and remote indicating devices for such j parameters as pressure, temperature, and flow , 10.4.23 Evaluation indicators are provided as required for monitoring the system operation. The offgas from the main condenser is one source of radioactive gas in the station. Normally it 10.4.2.5.1 StearnjetAirQectors includes the activation gases nitrogen 16, oxygen-19, and nitrogen 13, plus the radioactive noble-gas Steam pressure and flow is continuously monb parents of strontium 89, strontium 90, and cesium 07. An inventory of radioactive ccatamb tored and controlled lines. Redundant in the ejector pressure controllers steam supply l sense steam nants in the effluent from the steam jet air ejectors is pressure at the second stage inlet and modulate the evaluated in Section 113. steam supply control valves upstream of the air ejectors. The t. team flow transmitters provide inputs ag
\ Steam supply to the second stage ejector is to logic devices. These logic devices provide for $$
l maintained at a minimum specified flow rate to isolating the offgas flow from the air ejector unit on a l ensure adequate dilution of hydrogen and prevent two out of three logic, should the steam flow drop the offgas from reaching the flammable limit of below acceptable limits for offgas stream dilution, hydrogen. pgl 10.4.2.5.2 Mechanical Vacuum Pump rd W Pressure is measured on the suction line of the The MCES has no safety related function as mechanical vacuum pump by a pressure switch, discussed in Section 3.2. Failure of the system will Upon reaching a preset vacuum, the pressure switch not comprom be any safety related system or compo- energi7cs a solenoid valve which allows additional nent and will not prevent safe reactor shutdown, seal water to be pumped to the vacuum purnp. Seal l pump discharge pressure is locally monitored. Seal Should the system fail completely, a gradual water cooler discharge temperature is measured by a reduction in condenser vacuum would result from temperature indicating switch. On high temperature, the buildup of noncondensable gases. This reduction the switch activates an annuciator in the main control in vacuum would first cause a lowering of turbine room. The vacuum pump exhaust stream is l cycle efficiency due to the increase in turbine exhaust discharged to ti.e turbine building compartment MR gg pressure. If the MCES remained inoperable, exhaust system v>hich provides for radiation NN monitoring of the system effluents prior to their
$$ condenser pressure would then reach the turbine trip set point and a turbine trip would result. The loss of release to the monitored vent stack and the condenser vacuum incident is discussed in Subsection atmosphere.
15.2.5. The vacuum pump is ' ripped and its discharge l O U valve is closed upon receivwg a main steam high-high radiation signal. Amendment 17 10.4 5
ABWR nuww Standard Plant nev ^ 10.4.3 Turbine Gland Seal Splem The turbine F and l seal system (TGSS) prevents the escape of radioactive steam from the turbine shaft / casing penetrations and valve stems and prevents air inleakage through subatmospherie turbine glands. O { 1 l l 1 I l O Amendment 17 10 & 5.1
ABWR mem S111RLllinLP.latit - w^ 10.4 3.1 Design Itases supplied from the main steam line or ausiliary steam (V] header. Above approximately 50% load, however, 10.43.1.1 Safety Design liases sealing steam is normally provided from the heater 6E drain tankvent header. At allloads, gland sealing ?6 The TGSS 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 appreciable radioatlivity releases even if highly 10.43.1.2 Power Generatinn Design Itases abnormallevels of radioactive contaminants are present in the process sicam, due to unanticipated Power Generation Design flash Orts The TGSS is fuel failure in the reactor, designed to prevent atmospherie air leakage into the turbine casings and to prevent radioactive steam The outer portion of all glands of the turbine and leakage out of the casings of the turbine. generator. main steam valves are connected to the gland steam condenser which is maintained at a slight vacuum by Power Generation Design I! asis Tra The TGSS the exhauster blower. During plant creration, the reti.irns the condensed sicam to the condenser and gland steam condeaser and one of the two installed exhausts the noncon densable gases,sia 'he turbine 100% capacity motor driven blowers are in l l building compartment exhaust system, to the plant vent. operation. The exhauster blower to the turbine building compartrnent exhaust system effluent ,, stream is continuously monitored prior to being Power Generation Design Ihcis Three The TGSS discharged. The gland steam condenser is cooled by has enough capacity to handle steam and air flows main condensate now, resulting from twice the normal packing clearances. 10.433 1: valuation 10.43.2 Description p The turbine gland seal system is designed to l 10.43.2.1 General Description prevent leakage of radioactive steam from the snain I (] turbine shaft glands and the valve sts as. The l The turbine gland sealing system is illustrated in high pressure turbine shaft seals must accommodate l Figure 10.4-2. The turbine gland seal system consists a range of turbine shell pressure from full vacuum to of a sealing steam pic:.sure regulator, sealing steam approximately 220 psia. The low.picssure turbine header, a gland steam condenser, with two full- shaft seats operate against a vacuum at all times. capacity exhausser 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 10.4 3.2.2 System Operatinn a slight vacuum. The radioactive content of the sealing steam which eventually exhausts to the plant The annular space through which the turbine vent and the atmosphere is evaluated in Section 113 7g shaft penetrates the casing is scaled by steam and makes a negligible contribution to overall plant gg supplied to the shaft seals. Where the gland seals radiation release, in addition, the auxiliary steam operate against positive pressure, the sealing steam system is designed to provide a 10001 backup to the acts as a buffer and flows either inwards for collec- normal gland seal process steam supply. A fu'.', ca-tion ut an intermediate leakoff 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. opesate against vacuum, the seahng steam either is drawn into the casing or leaks outward to a vent Relief valves on the seal steam header prevent 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-d The seal steam header pressure is regulated formed prioi to plant operation. Components of the system are continuously monitored during operation g g l au 3matically by a pressure controller. During g g l Startup and low load operation, the seal steam is Arne ndment 17 1044
ABWR mum i Standard Plant / rv. A (m) to ensure that they are functioning satisfactorily. 10.4.4.1 Design liases V' Periodic tests and inspections may be performed in conjunction with maintenance outages. 10.4.4.1 ', Safety Design Itases 10.433 Instrumentation Appliestion The TDS does not serve or support any safety function and has no safety design basis. 10.43.5.1 Gland Steam Condenser lihausters 10.4.4.1.2 Power Generation Design liases 10.4.3.5.1.1 Pressurt Power Generation Design Basis on.g The TDS has Gland steam condenser exhauster suction pres- the capacity to bypass 33 percent of the rated main sure is continuously monitored and reported to the steam flow to t*2e main condenser, main control room and plant computer. A low vacuum signal actuales a main control room annun. Power Generation Desien Basis Two . The TBS is ciator, designed to bypass steam to the main condenser during plant startup and to permit a normal manual 10.43.5.1.2 Lori cooldown of the reactor coolant system from a hot shutdown condition to a point consistent with initia-Water leveh in the gland steam condenser drain tion of residual heat removal .ystem operation, leg are monitored and makeup is added as required l to maintain loop sealintegrity. Abnormalleveh are Power Generation Desien Basis _Three - The TBS is annunciated in the main control room, designed,in conjun tion with the reactor systems, to provide for a 40 percent electrical step- oud reduc. 10.43.5.1.3 ffiluent Monitoring tion without reactor trip. The systems will also allow a turbine trip but without lifting the main steam A The TGSS effluents are first monitored by a reljef and safety valves. (j system dedicated continuous radiation monitor 10.4.4.2 De*cription installed on the gland steam condenser exhauster blower discharge. liigh monitor readings are alarmed in the main control room. The system 10.4.4.2.1 General Description effluents are then discharged to the turbine building campartment cAhaust system and the plant vent stack The TBS is shown in Figure 10.31, Main Steam where further effluent radiation monitoring is System. The TBS consists of a three valve chest that performed. See Subsection 10.4.10.1 for interface is connected to the main steam lines upstream of the requirements pertaining to the radiological analysis turbine stop valves, and of three dump lincu that of the TGSS cffluents. connect separately each regulating valve outlet to one condenser shell. The system is designed to 10.43.5.2 Scaling Steem ilcader bypass 33 percent of the rated main steam flow directly to the condenser. The system and its Scaling steam header pressure is monitored and components are shown in Figures 10.4-10 and reported to the main control room and plant com- 10.4-11. puter, licader steam temperature is also measured and recorded. The turbine bypass system, in combination with the reactor systems, provides the capability to shed 40 10.4.4 Turbine Ilypass Systern percent of the turbine generator rated load without l reactor trip and without the operation of relief and The turbine bypass system (TBS) provides the safety vahes. A load rejection in excess of 40 capability to dischar,;c main steam from the reactor percent is expected to result in scactor trip but directly to the conder4ser to minimize step load without operation of any steam relief and safety reduction transient effects on the reactor coolant valve. system. The system is also used to discharge main (,_) steam during reactor hot standby and cooldown op-V erations. Arnendment 17 104 7
ABWR nauwai Standard Plant Rev A 10.4.4.2.2 Co.nponeut Description One valve chest is provided and houses three individual bypass valves. Each bypass valve is an angle body type valve operated by hydraulic Guid pressure with spring action to close. The valve chest assembly includes hydraulie supply and drain piping; three hydraulie accumulator:, one for each bypass valve; servo Valves; fast acting servo Valves; and, valve position transmitter $. The turbine bypass valves are provided with a separate hydraulic fluid power unit. The unit includes high pressure Guid pumps, filters, and heat exchangers, liigh pressure hydraulic fluid in provided at the bottom valve actuator and drained O l l l l O Amendment 17 10 4-7.1
AllWR mama Slamlard PItint nm a p back to the fluid reservoir. Sparger piping Tbc tut'eine bypass system vahes and piping con-distribut:s the steam within the condenser. form to the applicable codes as referenced in Chap-ter 3. 10.4.4.2.3 System Operation 10.4.4.3 Esaluation The turbine bypass vah'es are opened by a signal received from the steam bypass and pressure regula. The TBS does not serve or support any safety tion system whenever the actual steam pressure function and has no safety design basis. There is no exceeds the preset steam pressure by a small margin. safety-related equipment in the vicinity of the TBS. This occurs when the amount of steam generated by All high energy lines of the TBS are located in the the reactor cannot be entirely used by the turbine. turbine building. This bypass demand signal causes fluid pressure to be applied to the operating cylinder which opens the The effects of a malfunction of the turbine bypass first of the individual valves. As the bypass demand system valves and the effects of such a failure on increases, additional bypass valves are opened other systems and components are evaluated in dumping the steam to the condenser. The bypass Chapter 15. valves are equipped with fast acting servo valves to allow rapid opening, of bypass valves upon turbine 10.4.4.4 Inspection and Testing Requirements trip or generator load rejection. Before the system is placed in service, all turbine The bypass valves automatically trip closed bypass valves are tested for operability. The steam whenever the vacuum in the main condenser falls lines are hydrostatically tested to confirm lenktight-below a preset value. The b, pass valves are also ness. Pipe weld joints are inspected by radiography closed on loss of electrical power or hydraulic system per ASME !!!, Class 2 requirements upstream and pressure. The bypass valve hydraulic accumulators ANSI B31.1 downstream of the valve chest. The have the capability to stroke the valves at least 3 bypass valves may be tested while the unit is in O V times should the hydraulic power unit fail. operation. Periodic inspections are performed on a rotating basis within a pieventive maintenance When the reactor is operating in the automatic program in accordance wnh manufacturer's recom. load following mode, a 10% load reduction can be mendations. accommodated without opening the bypass valves, l and a 25% load reduction can be accommodated 10.4.4.5 Instrumentation Applications with momentary openirg of the bypass valves. These load changes are acccmplished by change in reactor Main steam pressure is measured in the equalin recirculating flow without any control rod motion. ing header by electronic pressure transmitters. Each transmitter supplies a signal to a corresponding When the plant is at uro power, hot standby or pressure regulator. Under normal conditions the initial cooldown, the system is operated manually by regulator output signal that has been selected by the the control room operator. The measured main operator will be transmitted to the steam bypass and steam system pressure is then compared against, and presst.re regulator system. If, however, the error l regulated to, the pressure set by the operator. detection circuitry detects failure in one of the signah, the valid signal is used and an annunciator is The turbine bypass control system can malfune- activated to warn the operator of the failed regulator tion in either the open or closed mode. The effects output. of both these potential failure modes on the NSSS and turbine system are addressed in Chapter 15.0. If Input to the system also includes load demand I the bypast valves fail open, additional heat load is and load reference signals from the turbine speed placed on the condenser. If this load is great load control system. The steam bypass and pressure enough, the turbine is tripped on high high con- regulation system uses these three signals to position denser pressure. Ultimate overpressure protection the turbine control valves, the bypass valves, and the for the condenstr is provided by rupture discs. If the reactor recirculation Gow control valves. A complete p bypass valves fait closed, the relief valves permit description of the control system is included in Q controlled cooldown of the reactor. Chapter 7. Amendment 3 10M
ABWR = > wu b Ild M i b illi Htv. A fm 10A.5 Circulating Water System each pump is fitted with a butterfly valve. This ( ) arrangement permits isolation and maintenance of V The circulating water system (CWS) provides any one pump while the others rernain in operation. cooling water for removal of the power cycle waste heat from the main condensers and transfers this The circulating water system and condenser is heat to the uhlmate heat sink. designed to permit isolation of each set of the ihrce series connected tube bundles to permit repair of 10.43.1 Design Itases leah and cleaning of water boxes while operating at reduced power. 10.43.1.1 Safety Design Itases The circulating water system includes water box The CWS does not serve or support any safety vents to help fill the condenser water boxes during function and has no safety design basis. startup and removes accumulated air and other gases from the water boxes during normal operation. 10.43.1.2 Power Generation Design Hases A chemical additive subsystem is also provided to Power Generation Desien Basis Q.Dr The CWS prevent the accumulation of biological growth and supplies cooling water at a sufficient flow rate to chemical deposits within the wetted surfaces of the condense the steam in the condenser, as required for system. optimum heat cycle efficiency. 10.43.2.2 Component Description Power Generation Desien Hsis Two The CWS is automatically isolated in the event of gross leakage Codes and standards applicable to the CWS are into the condenser pit to prevent flooding of the listed in Table 3.2-1. The system is designed and turbine building. corstructed in accordance with quality group D spec. ifications Table 10.4-3 provides design parameters Q 10.4.5.2 Description for the major components of the circulating water C/ system. 10.43.2.1 General Description 10.43.2.3 Spton Operation The circulating water system is illustrated in Figure 10.4 3. The circulating water system consists The CWS operates continuously during power of the following components: screen house and generation including startup and shutdown. Pumps intake screens; pumps; condenser water boxes and and condenser isolation valve actuation is controlled piping and valves; tube side of the main condenser; by locally mounted hand switches or by remote water box fill and drain subsystem; and related manual switches located in the main control room. support facilities such ns for system water treatment aad general maintenance. The circulating water pumps are tripped and the pump and condenser valves are closed in the event of The ultimate heat sink is designed to maintain a system isolation signal from the condenser pit the temperature of the water entering the circulation high-high level switches. A condenser pit high level water system within the range of 32 F to 100 F. The alarm is prmided in the control room. The pit water circulating water system is designed to deliver water level trip is set high enough to prevent inadvertent to the main condenser within a temperature range of plant trips from unrelated failures, such as a sump 40 F to 100"F. The 40"F minimum ternperature is overflow. maintained, when needed, by warm water recircula-tion. Draining of any set of series connected con-denser water boxes is initiated by closing the The cooling water is circulated by three fixed associated condenser isolation valves and opening speed motor driven pumps. the drain connection and wates box vent valve. When the suction standpipe of the condenser drain The pumps are arranged in parallel and dis- pump is filled, the pump is manually started. A low ( charge into a common header. The discharge of level switch is provided in the standpipe, on the Amendment 3 1049
ABWR moom Standard Plant Rn A suction side of the drain pump. This switch will tion valves are interlocked with the circulating water automatically stop the pump in the esent of low pumps so that when a pump is started, its discbarge water level in the standpipe to protect the pump valve will be opening while the pump is coming up to from excessive caviistion. speed, thus assuring there is water flow through the pump. When the pump is stopped, the discharge 10.4.5.3 Evaluation valve closes automatically to prevent or minimize backward rotation of the pump and motor. The CWS is not a safey-related system; however, a flooding analysis of the turbine building is Level switches rnonitor water levelin the con-performed on the CWS postulating a complete denser discharge water boxes and provide a permis-rupture of a single expansion joint. The analysis sive for trarting the circulating water pumps.These assumes that the flow into the condenser pit comes level switches ensure that the supply piping and the from both the upstream and downstream side of the conden:ct are full of water prior to circulating water break and, for conservatism, it assumes that one pump startup thus preventing water pressure surges system isolation valve does not fully close, from damaging the supply piping or the condenser. Based on the above conservative assumptions, To satisfy the bearing lubricating water and shaft the CWS and related facilities are designed such that sealing water interlocks during startup, the circulat-the selected combination of plant physical arrange. ing water pump bearing lubricating and shaft seal ment and system protective features ensures that all flow switches, located in the lubricating seal water credible potential circulating water spills inside the supply lines, must se'nse a minimum flow to provide turbine building remain confined inside the con- pump start permissive. denser pit. Further, plant safety is ensured in case of multiple CWS failures or other negligible probability Monitoring the performance of the circulating CWS related events by the plant safety related gen. water system is accomplished by differential pressure cral flooding protection provisions that are discussed transducers across each half of the cor. denser with in Section 3.4. remote differential pressure indicators located in the main control room. Thermal element signals from 10.4.5.4 Tests and Inspections the supply and discharge sides of the condenser are transmitted to the plant computer for recording, The CWS and related systems and facilities are display and condenser performance calculations. lested and checked for leakage integrity prior to initial plant startup and, as may be appropriate, To prevcnt icing and freeze up when the ambient following major maintenance and inspection. temperature of the ultimate heat sink falls below 32"F, warm water from the discharge side of the i All active and selected passive components of condenser is recirculated back to the screen house l the circulating water system are accessible for intake. Thermat elements, located in each condenser inspection and maintenance / testing during normal supply line and monitored in the main control room, power station operation. are utilized in throttling the warm water recirculation valve, which mi ains the minimum inlet tempera. 10.4.5.5 Instrumentation Applications ture of appreamately 40 F. Temperature monitors are provided upstream 10.4.5.6 Flood Protection and downstream of each condenser shell section. See response to Question 430.73(b), protection Indication is provided in the control room to against a CWS pipe, water box or expansion joint identify open and closed positions of motor-operated failure, butterfly valves in the CWS piping. 10.4.6 Condensate Cleanup System All major circulating water system valves which
- control the flow path can be operated by local The condensate cleanup system (CCS) purifies controls or by remote manual switches located on and treats the condensate as required to maintain the main control board. The pump discharge isola- reactor feedwater purity, using filtration to remove Amendment 17 10 4-10
ABWR m eio m Standard Plant n e v. /,
~
corrosion products, ion exchange to remove con-(u )'; denser leakage and other impurities, and water treatment additions to minimize corrosion / erosion product releases in the power cycle. i l I l (~~\/
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%J Amendment 17 10410.1
ABWR neww Sumdel Plant nu g ( l
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through the auxiliary condensers (off-gas recombiner condenser / coolers, gland steam condenser, and The applicant referencing the ABWR design shall perform a radiological analysis of the TGSS k steam jet air eje. tor condensers) and maintains effluents based on conservative site specific condensate pump rninimum flow. Measurements of parameters. From this analysis, the applicant shall pump suction and discharge pressures are provided determine the various actions to be taken if and for all purnps in the system, when the TGSS cffluent radiation monitor detects preset levels of effi.icnt contamination 5, including the The high pressure feedwater heater isolation level at which the TGSS steam supply will be valves are interlocked such that if a string of heaters switched over to auxiliary steam. (See Subsection were to be remosed from senice the extraction non. 10.3.5.1). return valves and/a isolation valves for those heaters 8 would automatically close and the heater string
$ bypass valve open. The low pressure feedwater heater isolation valves are interlocked such that, if a string of heaters were removed from service, the extractions to the affected heaters which are equipped with nonreturn valves would automatically close.
Sampling means are provided for monitoring the quality of the condensate and final feedwater, as described in Subsection 9.3.2. Temperatute mea-surements are provided for each stage of feedwater heating. Steam pressure measurements are provided at each feedwater heater, levelinstrumentation and controls are provided for automatically regulating the heater drain flow rate to maintain the proper
\ . levelin each feedwater heater shell or heater drain tank. liigh level control valves provide autornatic 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 makeup and rejection are controlled by the condenser hotwelllevel controllers. 10.4.8 Steam GeneratorIllowdown System (PWR) i l Not applicable to AllWR. l ( 10.4.9 Auxiliary Feedrater System (PWR) Not applicable to ABWR. 10.4.10 Interfaces 10..t.10.1 Radiological Analysis of the TGSS Emuents O G Amendment 17 10 4-17 1 l
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ABWR mamn 1;hmdard Plant an ti os 1 SECTION 11.4 CONTENTS Section Illie Page 11.4.1 Des!cn liases 11.4 1 11.4.1.1 Design Olijectives 11.4 1 11.4.1.2 Design Criteria 11.4.1 11.4.2 SnLtm DescrlRlhtn 11.4 1 11.4.2.1 General Description 11.4-1 11.4.2.2 Component Description 11.4-2 11.4.* .2.1 Thinfilm Layer 11.4 2 11.4.2.2.2 Pelletizer 11.4 2 11.4.2.2 3 Pellet Filling Machine 11.4 2
, 11.4.2.2.4 Mixing Tank 11.4 2 (v )
11.4.2.2.5 Drum Conveyor 11.4-3 11.4.2.2.6 incinerator 11.4 3 11.4.23 System Operation 11.4-4 11.4.23.1 Slurry Waste 11.4 4 11.4.2 3.2 Dry Active Waste (DAW) 11.4-4 11.4.233 Emironmental and Exposure Control 11.4-4 11.4.23.4 Malfunction Analysis 11.4 4 11.4.23.5 Shipment 11.4-4 11.4 3 Initth! cts 11.4 5 11.4 3.1 Cement-Gases Solidification System 11.4 5 rN l 11.4 il l l l Ar..cndment 17 l
ABWR ammax Standard Plant Jird! 4 I SECTION 11.4 TAllLES g 1 Table Illk Eage 11.4 1 Solid Waste Management System Components 11A 5 11.4 2 Calculated Solid Waste Volumes and Specific Activities 11.4 7 11.4-3 Calculated Solid Waste isotopic Distribution 11.4 8 1 l i O 1 l l l l l l 1 l 11.4 iii Amendment 6 l
I ABWR mmox Standard Plant nev. n 11.4 SOLID WASTE MANAGEMENT SYSTEM t i i t GE PROPRIETARY provided under segmrate awer . l'agg Mpendment : 11.4 1 6 , i 11,4 2 16 4 11.4 3 17 11.4-4 17 ; 11.4 5 17 11,4 6 6 11.4 7 13 >- 11,4 8 6 11,4 9 6 t f I
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ABWR mamu l Standnrd finnt nuv n o SECTION 11.5
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CONTENTS Sectlon 11Lis Eage 11.5.1 Desien linita 11.5 1 11.5.1.1 Design Objectives 11.5 1 11.5.1.1.1 Radiation hionitors Required fct Safety & Protection 11.5 1 11.5.1.1.2 Radiation hionitors Required for Plant Operation 11.5 1 11.5.1.2 Design Critcria 11.5 2 11.5.1.2.1 Radiation hionitors Required for Safety 11.5 2 11.5.1.2.2 Radiatien hionitors Required for Plant Operation 11.52 11.5.2 System Description 11.5-3 11.5.2.1 Radiation hionitors Required for Safety 11.5-3 m 11.5.2.1.1 hiain Steamline (htSL) Radiation hionitoring 11.5-3 [V) 11.5.2.1.2 Reactor Building IIVAC Radiatian hionitoring 11.5-3 11.5.2.1.3 Fuelllandling Area Ventilation Exhaust Radiation hionitoring 11.5-4 11.5.2.1.4 Standby Gas Treatment Radiation hionitoring 11.5-4 11.5.2.1.5 Control Building IIVAC Radiation hionitoring 11.5 . 11.5.2.2 Radiaticn hionitors Required for Plant Operation 11.5 6 11.5.2.2.1 Off gas Pretreatment Radiation Monitoring 11.5-6 11.5.2.2.2 Off gas Post. Treatment Radiation hionitor 11.5-6 11.5.2.2.3 Carbon Bed Vault Radiation hionitoring System 11.57 11.5.2.2.4 Plant Vent Discharge Radiation hionitoring 11.5 8 11.5-il rx Amendment 17 l
ABWR maux Standard Plant RIN. H SECTION 11.5 CONTENTS Nontinued) h Section Illig Bige 11.5.2.2.5 Radwaste Effluent Radiation Monitoring 113-8 11.5.2.2.6 Reactor Building Cooling Water Radiation Monitoring 11.5 9 11.5.2.2.7 Radwaste Building IIVAC Exhau<*. Monitoring 11.5-9 11.5.2.2.8 Turbine Building Compartmera Exhaust Monitoring 11.5 9 11.5.2.2.9 Turbine Gland Condenser Steam Exhaust Discharge Monitoring 11.5-9.1 11.3.2.2.10 Drywell Sumps Drain Line Rad;2 ion Monitoring 11.5-9.1 11.53 Emuent Monitorine and Sampline 11.5 9.1 11.53.1 Basis for Monitor Location Selection 11.5-10 11.53.2 Expected Radiation levels 11.5 10 11.533 Instrumentation 11.5-10 11.5 3.4 Setpoints 11.5 10 O1l 11.5.4 Process Monitorine and Sumnline 11.5 10 11.5.4.1 Implementation of General Design Criterion 60 11.5 10 l 11.5.4.2 Implementation of General Design Criterion M 11.5 10 11.5.43 Basis for Monitor Location Sclection 11.5-10 11.5.4.4 Expected Radiation Levels 11.5-10 11.5.4.5 Instrumentation 11.5-10 11.5.4.6 Setpoints 11.5-11 11.5.5 Calibration and Maintenance 11.5 11 11.5.5.1 Inspection and Tests 11.5 11 11.5.5.2 Calibration 11.5-11 11.5.5 3 Maintenanca. 11.5 12 11.5 lii .i.mendment 17
_~ - _ , - - - - - - _ . - - - - . . - ~ - _ - . - _ - . - - _ ABWR mmx Standard Plant niv n 1L5 PROCESS AND EFFLUENT (4)- Control building h /AC air suply radiation RADIOLOGICAL MONITORING AND monitoring; SAMPLING SYSTEMS (5) Standby gas treatment system off gas l The process and effluent radiological radiation monitoring monitoring and sampling systems are provided to , allow determination of the content of radioactive 11.5.1.1J Radiation Monitors Required for material in various gaseous and liquid process Plant Operation and effluent streams. The der.*gn objective and criteria are based on the following requirements: The main objective of this radiation monit. '
.aring is to provide operating personnel with (1) Radiation instrumentation required for measurements of the content of radioactive safety and protection, material in all effluent and important process streams. This demonstrates compliance with l
-(2) Radiation instrumentation required for plant normal operational technical monitor and plant operation. specifications by providing gross radiation level monitoring and by collection of halogen 4 All critical radioactive release points / paths and particulates on filters (gaseous effluents) within the plant are indentified and monitored by as required by Regulatory Guide 1.21.
this system. All other release points / paths of Additional objectives are to initiate discharge the plant are located in clean areas where valve isolation on the offgas or liquid radwaste radiological monitoring is not required. systems if predetermined release rates are exceeded, and to provide for sampling at certain 11.5.1 Design llases radiation monitor locations to allow determination of specific radionuclide content. 11.5.1.1 Design Objecthes The process radiation monitoring systems also 11.5.1.1.1 Radiation Monitors Requireil for provides the following design objectives: O Safety and Protection (1) Monitors gaseous effluent streams The main objective of this radiation monitor. ing is to initiate appropriate protective action (a) Plant vent discharge through stack
-to. limit the potential release of radioactive
- materials from the reactor vessel and primary and (b) Turbine building compartment exhaust secondary containment if predetermined radiation levels are exceeded in major process / effluent (c) Radwaste building ventilation exhaust streams. Another objective is to provide control room personnel with an indication of the (d) Turbine gland steam condenser exhaust.
radiation levels in the major process / effluent streams plus alarm at.nunciation if high radiation (2) Monitors liquid effluent streams levels are detected. (a) Radwaste effluent radioactivity The process radiation monitoring system provides-the iollowing design objeetives: (b) Drywell sumps drain radioactivity
- (1)- Main steamline tunnel area radiation monitoring; (3) Monitors gaseous process streams (2) Reacter building heating, ventilating, and' (a) Off gas pre-treatment sampling air conditioning (i!VAC) exhaust air rad-lation monitoring; (b) Off-gas post treatment sampling -
(3) Fuel handling area llVAC exhaust air rad- (c) Carbon bed vault gross gamma radiation intion monitoring; levels Amendment 17 115-1 - __.__________.______._.=_~,__.._,m. . - _ _ , _ . . _ _ _ _ , _ -
ABWR 2mim RIV 11 Sinndard Plarit (4) Monitors 1iquid process streams (6) Assure an extremely high probability of accomplishing safety functions in the event (a) Reactor building closed cooling water of anticipated operational occurrences intersystem radiation leakage (7) Initiate prompt protective action prior to 113.1.2 Design Criteria exceeding plant technical specification limits Design criteria of this system are based on meeting the relevant requirements of General (8) Provide warning of increasing radiation Design Criteria (GDC) 60,63, and 64 of 10CFR50, levels indicative of abnormal conditions by Appendix A,in accordance with SRP 11.5 of NUREG. alarm annur. cia *. ion 0800. These GDCs are in addition to those GDCs that are specified in Subsection 7.6.2.2 for (9) Insof ar as practical, provide self-system instrumentation. monitoring of components to the extent that power failure or component malfunction Also, the system is designed to meet the causes annunciation and channel trip applicable provisions of 10CFR20.106, RG 1.21 and RG 1.97. (10) Register full-scale output if radiation detection exceeds full scale The safety related process radiation monitoring subsystems are classified Safety Class (11) Use instrumentation with sensitivities and 2, Seismic Category 1. These subsystems conform ranges compatible with anticipated to the quality assurance requirements of 10CFR50, radiation levels. Appendix U. 11.5.1.2.1 Radiation Monitors Required for Safety
'1he design criteria for the main steamline and containment ventilation exhaust plenum radiation monitoring includes the following functional 11 5.1.2.2 Radiation Monitors Required for requirements: Plant Operation (1) Withstand the effect of natural phenomena The design criteria for operational radiation (e.g., earthquakes) without loss of monitoring shall include the following capability te perform their functions; functional requirements:
(2) Perform the intended safety functions in the (1) Prnvide continuous indication of radiation environment resulting from normal and levels in the main control room abnormal conditions (e.g., loss of HVAC and isolation events) (2) Provide warning of increasing radiation levels indicative of abnormal conditions by (3) Meet the reliability, testability, indepen- alarm annunciation dence, and failure mode requirements of engineered safety features (3) Insofer as practical, provide self-monitoring of components to the extent that (4) Provide continuous output in the main power failure or component malfunction control room causes annunciatior, and discharge valve isoit ion channel trip (5) Permit checking of the operational availability of each channel during reactor (4) Monitor a sample representative of the bulk operation with provisions for calibration stream or volume function and instrument checks 11.5 2 Amendment 17
ABWR uuim Sjtudard Plant FFV H 7ss (5) Incorporate provisions for calibration, and ( ) functional checks v (6) Use instruments with sensitivities and ranges compatible with anticipated radiation levels (7) Register full scale output if radiation detection exceeds full scale, The radiation system that monitors discharSes from the gaseous and liquid radwaste treatment system shali have provisions to alarm and to initiate automatic closure of the waste tD U i ~s N-l Amendment 17 115-21 l
ABWR mm Standard Plant RFV H discharge valve on the affected treatment system !s visually displayed on the affected radiation (N prior to exceeding the normal operation limits monitor. A high high or inoperative trip in the h specified in technical specifications as required by Regulatory Guide 1.21. radiation monitor results it. a channel trip which is provided to the reactor protection system (RPS) anc to the Icak detection and isolation system (LDS). Any two out of four channel trip results in initiation of main steamline isolation valve closure, reactor scram, main condenser mechanical vacuum pump 11.5.2 System Description (MVP) shutdown, and MVP line discharge valve closure. A high trip actuates a MSL high 11.5.2.1 Radiation Monitors Requind for Safety control room annunciator common to all l channels liigh and low trips do not result in a Information on these monitors is presented in channel trip. Each radiation monitor displays Table 11.51 and the arrangements are shown in the measured radiation level in mR/hr. l Subsection 7.6.1.2. 11.5.2.1.2 Reactor Building IIVAC Radiation 11.5.2.1.1 Main Steamline (MSL) Radiation Monitoring Monitoring Th' subsystem monitors the radiation level This subsection monitors the gamma radiation in il secondary containment of the reactor l level exterior to the main steamlines in the MSL building ventilation system exhaust duct. A tunnel. The normal radiation level is produced high activity level in the ductwork could be due primarily by coolant activation gases plus to fission gases from a leak or an accident. smaller quantities of fission gases being transported with the steam. In the event of a The system consists of four redundant instru. gross release of fission products from th: core, ment channels. Each channel consists of a [d the monitoring channels provide trip signals to digital gamma sensitive GM detector and a the leak detection and isolation system, coatrol room radiation monitor. Power is supplied to each channel, A, B, C, and D The MSL !adiation monitors consists of fo;r monitors from vital 120 Vac Divisions 1, 2, 3 redundant instrument channels. Each channel and 4 respectively. A two pen recorder powered l consists of a local detector (ion chamber) and a from the 120 Vec instrument bus allows the control room radiation monitor with a trip output of any two channels to be recorded by the auxiliary unit. Power for channels A, B, C, and use of selection switches. D monitors is supplied from vital 120 Vac divisions 1, 2, 3 and 4 respectively. All four The detectors are located adjacent to the l channels are physically and electrically exhaust ducting epstream of the ventilating independent of each other. system isolation valves and monitor the llVAC vent exhausts from the primary containment The detectors are physically located near the during purging and from the secondary main steamlines (MSL) just downstream of the out- containment. These detectors have sufficient board main steamline isolation valves in the sensitivity to detect high radiation levels stream tunnel. The detectors are geometrically during primary containment purge to alert the arranged and are capable of detecting significant operator for corrective action and to initiate increases in radiat. ion level with any number of the appropriate measurei. main steamlines in operation. Table 11.5-1 lists the location and range of the detectors. Each radiation monitor has four trip circuits: two upscal:, one downscale and one inoperative Each radiation monitor has four trip similar to MSL radiation monitors. circuits: two upscale (high high and high), one downscale (low), and one inoperative. Each trip V Amendment 17 11.5-3
ABWR mamax rov n Standard Plant A high high or inoperative /downscale trip in the radiation monitor results in a channel trip which is provided to LDS Any two.out of four channel trips will result in the initiation by LDS of the standby gas treatment system (SGTS) and in the isolation of the secondary containment (including closure of the containment purge and 11.5.2.1.3 Furillandling Area Ventilation vent valves and closure of the reactor building Exhaust Radiation l ventilating exhaust isolation valves). l This subsystem monitors the off gas radiation l The high high trip will initiate an alarm in level in the fuel handling area ventilation I the control room common to all channels. exhaust duct. The system consists of four ( channels which are physically and electrically j A downscale inoperative trip is displayed on independent of each other. Each channel l the radiation monitor and actuates a control room consists of a digital gamma sensitive GM l ) annunciator common to all four channels, detector and a control room radiation monitor. Power for channels (A, B, C, and D) is supplied The high radiation trip is provided and from the vital 120 Vac divisions 1,2,3 and 4 actuates a control room annunciator common to all respectiv-ly, channels. Each radiation monitor has four circuits: two upscale, one downscale and one inoperative similar to the MSL radiation monitors. This Each radiation monitor will display the subsystem performs the same trip functions as measured radiation level. those described in Subsection 11.5.2.1.2 for the reactor building HVAC exhaust radiation monitoring. O 11.5.2.1.4 Standby Gas Treatment Radiation Morutoring This subsystem monitors the off-gas radiation level in the SGTS exhaust duct to the stack using four channels. Two ionization chamber detectors are physically located downstream of the exhaust and heat removal fans and dampers on the exhaust duct to the stack. Two other scintillation detectors are used during off gas sampling of the gas exhaust to the stack. O 11.5 1 Amendment 17
ABWR mum Slandard Pf arit Rivl) The subrystem consists of four instrumented any source, and will provide isolation of intake l C channels. Each channel consir,ts of a detector of leakage from accident sources escaping from and a main control room radiation monitor, other plant buildings. Power for the channels is supplied from the non.IE vital 120Vac source. Each radiation monitor has four trip circuitt: two upscale, one/ inoperative and one downscale. All trips are displayed on the appropriate radiation monitor and each actuates a common main control room annunciator for high.high, high and Each radiation channel consists of a digital low / inoperative indications. gamma. sensitive GM detector and a radiation monitor which is located in the control room. 11.5.2.1.5 Control llullding IIVAC lladiation Monitoring Each radiation monitor has four trip circuits: two upscale, one/ inoperative and one The control building IIVAC radiation downscale- All trips are displayed for the monitoring subsystem is provided to detect high appropriate radiation monitor and each actuates radiation level in the cormal outdoor air supply, a control room annunciator, automatically close the outdoor air intake and j the exhaust dampers, and initiate automatically the outdoor air cleanup system in the emergency recirculation air supply loop. The emergency O recirculation fans shall be started and area exhaust fans stopped on high radiation. The radiation monitors for each of the control building, IIVAC systems consist of four l redundant channels to monitor the air supply to the building. Each radiation monitor is physically separated and powered from separate vital 120 Vac divisional power busses. Failure of one channel will not cause isolation of the llVAC system. l The monitors meet the requirements for Class 1E components to provide appropriate reliability. The system will wirn of the presence of significant air contamination in inlet air, from Amendment 17 11 M
1 ABWR maa Standard Plant Rtx n 11J.2.2 Radiation Monitors Required for Plant The radiation level output by the monitor can Operation be directly correlated to the concentration of the noble gases by using a seminutomatie vial I Information on these monitors is presented in sampler panel to obtain a grab sample. To draw Table 1131. a sample, a scrum bottle is inserted into a sampler holder, the sample lines are evacuated, l 113.2.2.1 Off-gas Pretrtatment Radiation and a solenold operated sample valve is opened Monitoring to allow offgas to enter the bottle. The bottle is then removed and the sample is analyzed in This subsystem monitors radioactivity in the the counting room with a multichannel gamma condenser offgas at the discharge of the delay pulse height analyzer to determine the concen-pipe after it has passed through the offgas tration of the various noble gas radionuclides. condenser and moisture separator. The monitor A correlation between the observed activity and detects the radiation level which is attributable the monitor reading permits calibration of the to the fission gases produced in the reactor and monitor, transported with steam through the turbine to the condenser, 113.2.2.2 Off-gas Post Treatment Radiation Monitoring A continuous sample is extracted from the offgas pipe via a stainless steel sample line. This system monitors radioactivity in the it is then passed through a sample cleober and a offgas piping downstream of the offgas system sample panel before being returnd to the suction charcoal adsorbers and upstream of the offgas side of the steam jet air ejector (SJAE). The system discharge valve. A continuous sample is sample chamber is a stainless steel pipe which is extracted from the offgas system piping, passed internally polished to minimize plateout. It can through the offgas post treatment sample panel be purged with room air to check detector for monitoring and sampling and returned to the response to background radiation by using a offgas system piping. The sample panel has a three way solenoid operated valve. The valve is pair of filters (one for particulate collection controlled by a switch located in the main and one for halogen collection) in parallel control room. The sample panel measures and (with respect to flow) with two identical GM l indicates sample line flow. Two digital detectors. Two radiation monitors in the main gamma sensidve GM detectors are positioned control room analyze and visually display the adjacent to the vertical sample chamber and are measured gross radiation level. connected to radiation monitors in the main control room. Power is supplied from 120 Vac instrument bus for radiation monitor and detector and for the The sample panel shielded chambers can be sample and vital sampler panels. purged with room air to check detector response to background radiation by using solenoid valves The radiation monitor has four trip circuits: opera:cd from the control room. The sample l two upscale (high high and high), one downscale panel measures and indicates sample line flow. and one inoperative. A solenoid operated check source for each detector assembly operated from the control room The trip outputs are used for alarm function can be used to check operability of the gross only. Each trip is visually displayed on the radiation channel, radiation monitor and actuates a control room annunciator: offgas high high, offgas high, and Power is supplied from a 120-Vac instrument offgas downscale/ inoperative. High or low sample bus to the radiation monitors and to the two pen line flow measured at the sample panel actuates a main control room offgas sarnple high. low flow annunciator. i Amendment 17 11.5 4 !
ABWR mmu Standard Plant niv n recorder. A 120 Vac local bus supplies the
; j sample panel. %)
Each radiation monitor has four trip circuits: two upscale (high high and high), one downcale (low) and one inoperative. Each trip is visually displayed on the radiation monitor. The trips actuate corresponding main control room annunciators: offgas post treatment high high radiation, offgas post treatment high radiation, and offgas post treatment downscale/ inoperative. liigh or low flow measured at the sample panel actuates the abnormal flow annunciator in the control room. A trip auxiliary unit in the control room takes the high high and downscale trip /inoper-ative outputs to initiate closure of the oligas system discharge and bypass valves. The high high trip setpoints are determined so that valve closure is luitiated prios en exceeding techniral specification limits. Any one high upscala trip initiates closure of offgas system bypass ;ine valve and permits operiing of the treatment line valve. O) (d' A vial sampler panel .imilar to the pre-treatment sampler panel is provided for grab sample collection to allow isotopic analysis and gross monitor calibration. 11.5.2.23 Carbon lied Vault Radiation Monitoring The carbon vault is monitored for gross gamma radiation !evel with a single instrument chantiel. The channel includes a digital sensor and converter, and a radiation monitor. The sensor is located outside the vault on the HVAC exhaust line from the vault. The radiation monitor is located in the main control rcom. The channel provides for sensing and readout of gross gamma radiation over a range of six logarithmic 6 decades (1 to 10 mR/hr). The monitor has one adjustable upscale trip circuit for alarm and one downscale trip for iistrument trouble. Power is supplied from 120 Vac instrument bus. I l V Amendmeret 17 11 3-7 i
ABWR msmx ! Standard Plant nrv.n ! 11.5.2.2.4 Plant Vent 1)ischarge Radiation Monitoring This system monitors the plant vent discharge for gross radiation level during normal plant operation and collects halogen and particulate samples for laboratory analysis. The discharge { through this coramon plant vent includes HVAC j exhausts from the reactor, turbine, radwaste and j service buildings. Also, this system utilizes a l high range radiation monitor that measures fission products in plant gaseous effluents during and following an accident. l A representative sample is continuously extracted from the ventilation ducting through ., two isokinetir nrobes in accordance with ANSI W N13.1 and passed through the containment ventilation sample panels for monitoring and sampling, and returned to the ventilation ducting. Each sample panel has a pair of filters (one for particulate collection and one for halogen collection ) in parallel (with respect to flow) for continuous gaseous radiation sampling. 11.5.2.2.5 Radwaste EIT1uent Radiation l The gross radiation detection assembly coo ists Monitoring of a shielded chamber, beta-gamma-sensitive GM tubes, and a check source. The extended range This subsystem continuously monitors the detector assembly consists of an ionization radioactivity in the redwaste effluent prior to chan.ber which measure radiation levels up to its d:scharge and drainage. l 10 pCi/cc. A radiation monitor in the main 5 control room analyzes and visually displays the Liquid waste can be discharged from the measured radiation IcVel, sample tanks containing liquids that have been processed through one or more treatment systems The sample panel shielded chambers can be such as evaporation, filtration, and ion purged with room air by using two solenoid valves exchange. Prict to discharge, the liquid is operated from tL control room to check detector extracted from the liquid drain treatment response to background radiation, thus checking process pipe, passed through a liquid sample operal'ity of .he gross radiation channel, panel which conta ns a detection assembly for gross radiation monitoring, and returned to the Power is supplied from 120 Vac local bus for process pipe. The detection assembly consists the radiation monitor and for the sample panel, of a scintillation detector n iunted in a shielded sample chamber equipped with a checx The radiation monitor initiates trips for source. A radiation monitor in the control room alarm indications on high high, high, and low analyzes and visually displays the measured radiation from each detector assembly. Also, the gross radiation level. sampled line is monitored for high or low flow indications and alarming. The sample panel charrSer and lines can be drained to allow assessment of background Table 11.5-2 pr:sents the gaseous and airborne buildup. The panel measures and indicates monitors for the effluent radiation monitoring sauple liv flow. A solenoid-operated check system. some operated from the control room can ba: used to chech operability of the channel. O Amendment 17 1158
ABM 2w=xx Standard Plant uv. n Based on acceptable radiation levels, The trip ignals t.re anuunciated in the _(q)~ discharge is permitted at a specified release radwaste bui! Jing control room and in the du rate and' dilution rete. contro; room. The radiation monitor has three trip Each radiation monitor visually displays the circuits. Two upschte trips (high high 'and radiation level and supplies an output signal to high), and oae downscale/ inoperative trip. The the computer, high-high u pscale t rip a n d -th e downscale/ inoperative trip are used to stop the ' A gamma check source is provided for channel HCW cffluent pump. .Also, the two upscale trips calit, ration.
-and the low downscale/ inoperative trip actuate
- annunciators in the main control room and in the
_ radwaste building control room. Table 11.5 3
~
describes _the liquid monitors used for process radiation monitoring. 11.5.2.2.6 - Reactor Building Cooling Water Radiation Monitoring - This subsystem consists of three channels: one for each RCW A,' B and C loop for monitoring
=lntersystem radiation leakage into the reactor building cooling water system.
Each channel consists of a scintillation detector w'nch is located in a well near the RCW heat exchanger exit pipe. Radiation detected-O. from.the three channels are multiplexed and fed b- l into a common radiation monitor. This monitor
. provides individual channel trips on high
- radiation; level and; nwnscale/ inoperative indication for annunciation in the control room.
Power to the monitorm is provided from the non.1E vital 120 Vac source. l 11.5.2.2.7 Radwaste Building IIVAC Exhaust
' Monitoring . 11.5.2.2.8 Turbine Building Compartment Exhaust Monitoring This subsystem monitors the radwaste building ventilation discharge to the stack, including This subsystem monitors the vent discharge in
- radwaste storag: tank vents, for gross radiation the turbine building con.partment for gross level. _The ~ system consists of two redundant radiation levels _ The monitoring is provided by instrument channels, each channel having a local- four channels (two redundant sets) .Two-detector, a converter, and a main conhol room redundant channels monitor radiation in the radiation monitor. Power is supplied to each equipment area air exhaust duct and the other
! channel by.the 120 Vac instrument bus, two redundant channels monitor the radiation in the SJAE arca air exhaust duct. Each channel uses a digital detector located adjacent to the Each radiation' monitor provides two trip monitored exhaust duct. The outputs from cach circuits: one for upscale (high) radiation and set of detectors-are multiplexed and then fed one for downscale/ inoperative trip. into two separate process radiation monitor for LU l' Amendment 17 39 I-i
ABWR mamax
-prv n Standard Plant display, recording and annunciation. Each are monitored for radioactivity releases in monitor provides alarm trips on radiation high accordance with Criterion 64 of General Design and on radiation low (downscale/ inoperative). Criteria,10CFR50, Appendix A, as follows:
11.5.2,2.9 Turbine Gland Condenser Steam Exhaust Discharge Monitoring
' This subsystem monitors the off gas releases to the stack from the turbine gland seal system.
The off-gas releases are continuously sampled and monitored for noble gases by a scintillation detector. The output signal is multiplexed and then fed to a shared radiation monitor in the main control for display, recording and annunciation. This monitor provides t 'o trip alarms, one on radi8 tion high and.one on radiation low (downscale/ inoperative).. A grab sample of the off gas is provided for laboratory analysis. Also, samples of halogens and particulates are collected on filters for periodic analysis. A gamma source check is provided for channel calibration purposes. l 11.5.2.2.10 Drywell Sumps Drnin Line Radiation Monitoring This subsystern monitors the radiation levelin the liquid waste that is transferred in the drain line from the drywell LCW and HCW sumps to the radwaste system. One monitoring channel is l provided in each sump drain line. Each channel uses an ionization chamber which is located on the drain line from the sump juit downstream fiom the outboard isolation valve. The output from each sen>or is multiplexed and then fed into a l shared radiation monitor for display, recordin; and annunciation. I The radiation monitor provides three trip circuits: two upscale (radiation high-high and l high), one downscale/ inoperative. The high-high signal is used to close the outboard isolation valve in its respective drain line. All trips are annunciated in the maia control room. 11.5.3 Effhient Moultoring and Sampling All potentially radioactive effluent materials O Amendment 17 11.5-9 1
ABWR mama Standard Plant unv. n
,, l (1) Liquid releases are monitored for gross (1) Off-gas post. treatment
( ) gamma radioactivity; v (2) Reactor building HVAC air exhaust. (2) Solid wastes are monitored for gross gamma radioactivity; and (3) Fuel handling area air exhaust. l(3) Gaseous releases are monitored for gross (4) Drywell sump liquid waste drain gamma radioactivity. (5) Radwaste ef0uent 11.53.1 Basis for Monitor Location Selection 11.5.4.2 Impleraentation of General Design Monitor locations are selected to assure that Criterla 64 all effluent materials comply with regulatory requirements as covered in Regulatory Guide 1.21. Radiation levels in radioactive and poten, tially radioactive process streams are monitored 11.5J.2 Expected Radiation levels for radioactivity releases. The.e include: Expected radiation levels are within the (1) Main steamlir.e ranges specified in Tables 11.5 2 and 11.5-3. (2) Off-gas pre-treatment and post treatment 11.533 Instrumentation (3) Carbon bed vault The process radiation monitors used for measuring radioactivity are listed in Table (4) Intersystem leakage into reactor building 11.5-1. cooling water Graf samples are analyzed to identify and 11.5.43 Basis for Monitor Location Selection (o) quar.tify the specific radic suclides in effluents and wastes. The results from the sample analysis Monitor locations are selected to usure are used to establish relationships between the compliance with Regulatory Guide 1.21 in that gross gartma monitor readings and concentrations sample points are located where there is a or release rates of radionuclides in continuous minimum of disturbance due to fittings and other effluent releases, physical characteristics of the equipment and components. Sample nozzles are inserted into 11.53.4 Setpoints the flow or liquid volume to ensure sampling the bulk volume of pipes and tanks. In the case of Th. radiation level trip setpoints for both liquid and gas flow, care is taken to actuation of automatic control features that assure that individual samples are actually initiate actuation of isolation valves, dampers representative of the effluent mixture. A more or diversior. valves are specified in the plant jetailed discussion is given in ANSI N13.1. technical specifications as indicated in Table 11.5 1. 11.5.4.4 Expected Radiation 14vels Expected radiation levels are listed in Tables 11.5 2 and 11.5-3. 11.5.4. Process Monitoring hnd Sampling 11.5.4.5 Instrumentation 11.5.4.1 Implementation of General Design Criterlon 60 The process radiation monitors used for measuring radioactivity are listed in Table All potentially significant radioactive dis- 11.5-1. l
, charge paths are equipped with a control system ,
() tn automatically isolate the discharge on indi- I v cation of a high radiation level. These include: l Amendment 17 115-10 l l l
MM - 2w100AK - nW n
)'
Sandard Plant _ _ _ Grab samples are analyzed to identify and (4) Control building HVAC
; . quantify the _ specific radionuclides in pro _ cess =* 1 streams. The results from the sample analysis - (5) Reactor building cooling water system
- are used to establish relationsl.ips between the ,
c gross gamma monitor readings and concentration (6) SGTS and radionuclides in the process streams. (7) Turbine buildiag compartment exhaust 11.5.4.6 Setpoints , (8) Offgas preueatment
- Tl.e radiation trip set points for the various monitors are listed in Table 11.51. (9) Carbon bed sault 11.5.5 Calibretion and Maintenance - The following monitors include built.in check sources and purge systerns which can be operated l11.5.5.1 Inspection and Tests from the main control room:-
'During re. actor operation, daily checks of (1) Offgas post. treatment system operability are made by observing channel behavior. At periodic intervals during teactor (2). Plant vent discharge _
operation, the detector response of each monitor provided with a remc,tely pesitioned check source (3) Liquid waste discharge will be recorded together whh the ir,strument background count rate to ensure proper function. (4) - SGTS discharge ing of _the monitors. Any _ detector whose re. sponse cannot be verified by observation during (5) Radwaste building exhaust
-normal operation or by using the remotely posi-tioned check source will have its response (6) Gland steam condenstr exhaust checked with portable check source. A record O--
will be maintained showing the background - radiation level and-the detector response. 11.5.5.2 Cilibration The system has ' electronic testing and cali _ The continuous radiation monitor calibration - ! brating equipment which permits channel testing is according to certified National Bureau of
-without relocating or dismounting channel compo- Standards of commercial radionuclide standards, nents. An irternal trip test circuit' adjustable and is accurate to at least + or. 15E The over the full range of the readout meter is used- snurce-detector geometry during primary cali-for testing. Each channel is tested at least brction is identical to the sample deacetor gen.
semiannually prior to_ performing a calibration metry in actual use. Secondary standards which check. Verification'of channel operation and were counted in reproducibic geometry during the trip function will be done at this time if it can primary calibration are supplied with each con-
-be done without jeopardizing plant safety. The - tinuous monitor for calibration after installa-test will be documented, tion. Each continuous' monitor is calibrated-during plant operation or during the refueling -
The following monitors have alarm trip outage if the detector is not readily ac-
~
l-_ L circuits'which can be tested by using test cessible. A calibration can also be performed signals or portable gamma sources: by using liquid or gaseous radionuclide stan-dards or by analyzing particulate iodine or gas-(1)' Main steamline cous grab samples with laboratory instruments. (2) Reactor building HVAC l (3) ~ Fuel handling area IIVAC
.g p
u Amendment 17 11.5-11
. , .. _ _ _ _ _ . . _ _ _ . _ - - _ _ _ _ _ _ _ _ . - - _ _ _ = _ - - . - - -
ABWR mama RIN,11 Standard Plant The following monitors display the gross gamma addition to the replacement or adjustment of any signal in counts / min components required after performing a test or calibration check. If any work is perforined (1) Off gas post-treatment which would affect the calibration, a recali-bration is performed at the completion of the (2) Plant vent discharge work. 11.5.5.4 Audit 4 and Verlucations Audits ?.nd verification during normal plant operation are out-of scope for the Standard ABWR Plant. (3) Radwaste effluent discharge (4) Gland steam condenser exhaust (5) Reactor building cooling water system The following monitors are calibrated to provide measurements of the gross gamma dose rate in mR/hr: (1) Main steamline (2) Reactor building IIVAC (3) Fuel handling arca HVAC (4) Carbon bcd vault (5) Control building HVAC (6) Turbine building compartment exhaust (7) Radwaste building HVAC exhaust (8) Off-gas pre-treatment (9) Drywell sump liquid drain line 11.5.5.3 M;ir.tenance All t hsnnel detectors, electronics, and receder tre serviced and maintained on an annual basis or in accordance with manufacturers recommendations to ensure reliable operations. Such maintenance includes cleaning, lubrication, and nsurance of free movement of the recorder in Amendment 17 115 12 9
,l l
ABWR- naam. Standard Plant imv. n TAllLE 11.5-1 -
. PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS St1D91111 Monitored ' No. - Detedor - Sample IJne - Channel Warning ACI' -frocess of 1 or Detector Range Alarm ItlD Sntle '
Chan- lascation (Note 1) Ilthi A. Saf: tv-Related Monitors - Main 4 team. 4 IC - Immediately. 110hiR/hr above full technical 6 dec, log line tunnel - downstream (106 to power . speci cation area of plant main 10 background, steamline Amps) below trip isolation valve Reactor 4 S/C Exhaust duct . 0.01 to above back- technical 4 dec. log building upstream of 100 'mR/hr ground, specification liVAC - exhaust ven- below trip exhaust - tilation
-isolation valve Control 8' S/C Intake duct . 0.01 to above back- technical 4 dec. log building _ upstream of 100mR/hr ground, specification iilVAC air . intake venti- below trip supply lation isola-tion valve.
Standby gas 2 S/D SGTS exhaust 0,gto above back- technical 6 dec. log treatment - air duct tocpm. ground,- specification system downstream - below trip off gas 2 IC of exhaust 10 .,13 to abose back- Nonc 6 dec, log ad 1. cat re- 10 grou.;d moval fans Amps and dampers
- Fuel - '4 S/C locally above. 0.1 to 10 above back- 4 dec. log ! handling operating mR/hr ground, - technical area air ~ floor below trip specificat' ion exhaust -
4 Channels for cach air intake l Amendment 17.. till3 l l l l
. _ . - - . . - . . . . _ . . - . . . . . _ . . . _ . . . . . , . _ _ . . . . . . . . . . , _ . . . _ . _ _ . _ . . - , . . . . . _ - - _ , _ _ , . . . . . _ - . ~ . _ . , ~ . .
7 .. . ABWR uumu Rf?V D Standard Plant TAllLE 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS (Continued) h Sitnoint No. Detector Sample Line Channel Warning ACF Monitored Process of Ins or Detector Range Alacu IIiD Stn}I Chan- Leta.tiqn (Note 1) Etli B. Monitors Reautred for Plant Operation Radwaste liquid dis-1 S/D Sample line 10
)
10' to above back-ground, Technical specification 5 dec. log charge counts /mic below trip RCW 11x above back- None 5 dec, log Reactor 3 S/D 10'fto ground building line exit 10 cooling water counts / min system 6 5 dec. log Offgas 2 GMB Sample line 10 to 10 above back- Technical counts / min ground, specification post treatment below trip 6 6 dec. log Sample line 1 to 10 at tech spec None Offgas 1 S/C pre- mR/hr report level treatment 6 6 dec. log On charcoal 1 to 10 above None Charcoal 1 S/C vault vault HVAC mR/hr background exhaust line 6 5 dec. log GM B Sample line 10 to 10 at quarterly None Plant vent 1 discharge counts / min tech spec level 1 IC Sample line 10 to above back- None 6 dec log 10 ground, Amps below trip 5 2 GM-B Exhaust ducts 0.01 to 100 above back- None 4 dec. log k Radwaste building mR/hr ground, [ HVAC vent below trip 113-14 9 Amendment 17 l l
ABWRt -max Standard Plant an in TABI.E 11.5 1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS (Continued) Ertnoint Monitored : No. Detector Sample Line Channel Warning ACF fr.9Cf5.8 of Dpg or Detector Range AlatB1 hlD ' ralf. d Chan. Location fNote 1) El. B. hiQDitors Required for Plant Ooeraden-T/B compart 4 S/C Exhaust duct 0.01 to 100 above back- None 4 dec. log , ment exhaust mR/hr ground 0.ywell . 2 IC Drain line I to'10 above back- Teci.nicad 6 dec. log sump - from LCW & ' mR/hr ground specification liquid itCW sumps drain 5 Gland steam . :1 S/D ' Sample line 0.1 to 10 above back- None 6 dec. log
. condense: cpm ground exhaust
- discharge Legend ACF Automatic Control Function GM.B Beta. Sensitive GM Detector IC lon Chamber
-S/C Digital Gamma Sensitive GM Detector S/D- Scintillation Detecter Note 1 The channel range specified in this tnble is the equipment measuring or display range of the indicated parameter. Refer to Tables 11.5 2 &
- 11.5.* %r the dynamic detection range of the monitoring channel
- expressed as concentration in units of microcuries per cubic centimeter, referenced to a specific nuclide.
l O Amendment 17 11 3-15 __ ~ . . _ _ , - - - . . _ _ _ . _ . _ _ __ . - -. . . . . , , . . _ . . . , , , _ ,
lI ABWR moux Standard Plant nrv.n TABLE 11,5-2 PROCESS RADIATION MONITORING SYSTEM (GASEOUS AND AIRBORNE MONITORS) Dynamic Principal l Radiation Configu- Detection Radionuclides Expected Alarms i hl.pgligt ration Iygg Sensitivity Range Measured Activitv* * &Idp3 l i Offgas Offline B-GM 0.25 cpm /pCi/ 10-5 to Xe-133' 5x10-5 plown/L post treat- cm 3 101 pCi/cc INOP/ Low I ment Filter P pCi/cc Cs-137 High High-High Filter-1 1-131 Offgas Adjacent S/C 700 mR/hr 10-3 to 4 Noble gas ~03pCi/cc High High pre-treat- to per pCi/cc pCi/cc fission pro. High ment sample ducts Low /INOP chamber FlowII/L Main steam- Adjacent IC 3.721010 100 106 Coolant ~ 100 mR/hr INOP line radia- to steam Amp /R/hr mR/hr activation Low tion lines (Co-60)* gases liigh High High Charcoal Inline S/C 0.5 mR hr 10'3 to Noble gases Negligible High O vault per 10' 103 Low /INOP pCi/cc pCi/cc T/B compart- Inline S/C 0.5 mR hr 10-5 to Xe-133' ~4 x 10-5 High ment exhaust per 10- 10 1 Xe-135 pCi/cc - Low /INOP pCi/cc pCi/cc Reactor Inline S/C 0.5 mR ehr 10-5 to Noble gases ~ 4 x 10'5 INOP building HVAC per 10'8 10-I Xe-133* pCi/cc Low air exhaust Ci/cc pCi/cc Xe-135 High High High Isolate Phn' vent Offline B-GM 250 epm /pCi 10-7 to Xe-133* ~5 x 10-5 Flow H/L discharge Filter-P per cc 10-1 Cs-137 pCi/cc INOP/ Low (normal pCi/cc High range) Filter-I I-131 High High Sensitivity based upon this radionuclide.
** Erpected activities are estimated based on cristingplants.
Amendment 17 11.5-16
3 ABWR- 2wmx l Standard Plant nwn TABLE 11.5 2 PROCESS RADIATION MONITORING SYSTEM (GASEOUS AND AIRBORNE MONITORS)(Continued) Dynamic Principal Radiation Configu . __ _ Detection Radionuclides Expected Alarms Monitor En1[egt. Iypt Stas.111111r Eanat Measurest Activit v** & Trips Main stack Offline IC ' 1.6x104 1g2go xe.133, 5xigs pio, ii/L (high range) A/pCi/cc 105 Ci/cc INOP/ Low pCi/cc liigh liigh lligh Radwaste Offline Il-GM 1&5 to Xc133' ~1g5 Ilighdligh building 0.510mR{hr Filter P per - 10 3 Ct 137 pCi/cc liigh ventilation Filter 1 pCi/cc pCi/cc 1131 Low /INOP discharge Flow II/L Glandsteam Offline S/D 1.3.) cpm x 10~5to Xe 133 6 Iligh condenser Fliter P 103 101 Cs-137' pCi/cc tow /Inop exhaust Filtcr l' perpCi/cc pCi/cc discharge 1131 (-
\
Control bldg. . Inline IIVAC air S/C 0.5 mR hr per 10' 10-5 to Xe-133' Negligible Iligh Iligh/ 10-1 INOP intake - pCi/cc pCi/cc 1ligh Low Standby gas - Offline S/D- 1.33 cpm x 10 7tol0'l Noble gases -5x10 7 liigh liigh
= treatment - Filter.P 105 pCi/cc Cs 137' pCi/cc fligh Exhaust Filter 1 pCi/cc l131 Iow/INOP Inline IC 1.6x10-10 10-2 to Noble gases ~ 5 x 10 7 pCi/cc . ItE pCi/cc pCi/cc Fuel handling ' Inline S/C 34 mR/hr 10-3 to Noble gases ~6 x 10'3 liigh liigh area exhaust per pCi/cc 102 pCi/cc pCi/cc liigh (Cs-137)' low INOP isolate-Sensitivity based upon this radionuclide.
" 'Etpected activities are estimated and are based on exirting plants.
P = Particulate Filter 1 - Iodine or Cnarcoat Filter
-O Amendment 17 115-17
ABWR 33AC00AK Standard Plant nuv.n TABL'd 11.5-3 PROCESS RADIATION MONITORING SYSTEM (LIQUID MONITORS) Dynamic Principal Radiation Configu- Sensit- Detection Radlonuclides Expected Alarms Monitor ration 11111 Eange Measured Activity" & Trips Radwaste effluent Offline 1.33 x105 to-7 to Cs 137' ~ 104 High/High radiation monitor epm /pCi 10-2 Co-60 Ci/cc High percc pCi/cc Low /INOP Isolate Reactor building Inline 1.2x104 10-5 to Cs-137* ~6 x 10-5 High cooling water epm /pCi 100 Co-60 pCi/cc low /INOP system radiation per cc pCi/cc monitor Drywell Sump Inline 30 mR/hr 10-2 to Gross ~5 x 10-2 High High l Drain 104 Gamma pCi/cc High I pCi/cc Cs-137 ' Low /INOP Isolate Sensitivity based upon this radionuclide.
" Expected activities are estimated and are based on existingplants.
O 11.5-18 Amendment 17
_ , _ _ . _ . . . . . . _ _ _ , . _ . , _ . _ . _ _ . . _ _ _ _ _ . ~ _ . _ _ _ . - d
-ABWR mmu Standard Plant Rtw. n TABLE I1.5-4 RADIOLOGICAL: ANAIXSIS
SUMMARY
OF LIQUID PROCESS SAMPLES Grrb Sample SenstthIty Samnle DescrIDll2B freau *ncy Analysis pCl/ml harysht
- 1. Reactor Coolant Filtrate Daily (a) Gross gamma 10-6 Evaluate reactor water ac-tisity Crud Daily (a) Gross gamma 10-6 Eval uate crud activity i Filtrate Weekly (b) 1-131,1-133 10-7 Evaluate fuel cladding -
integrity Crud and filtrate Weekly Gamma spectrum . 5 x 10-7 Deterndne radionuclides present in system
- 2. Reactor water cleanup Biweekly Gross gamma 10*6 Evaluate cleanup systern efficiency
- 3. Condenser demineralizer
- l. Influent . Monthly Gross gamma 10-6 Evaluate leakage
.. Effluent Monthly Gross gamma - 10 6 Evaluate demineralizer performance h
t/
- 4. Condensate storage tank Weekly Gross p y 10-6 Evaluate water radioa-ctivity
- 5. Fuel pool filter -
demineralizer Inlet and outlet Periodically Gross B-y 10-6 . Evaluate system perform-ance
- 6. LCW collector sampling Periodically Gross p-y 10 6 Evaluate system perform-tanks (4) ance
- 7. IICW collector tanks (2) Periodically Gross 7 10-6 Evaluate system perform-anCC
- 8. HSD sample tanks (2) Periodically Gross p y 10-6 Evaluate system perform-ance 3
0
. Amendment 17 115-19 o .
ABWR swer. Standard Plant RIN. H TABLE 11.5-4 RADIOLOGICAL ANALYSIS
SUMMARY
OF LIQUID PROCESS SAMPLES (Continued) Grab Sample Sensitivity Samole Descrictl2B FERECDgy Annivsls DCl/ml INrpose
- 9. Solid waste supply tank Periodically Gross p 7 10-6 Compare activity with (evaporator bottoms) that determined by drum readings
- 10. HCW distillate tank Periodically Gross S.y 10-6 Evaluate evaporator per.
(evaporator) formance
- 11. Reactor building cooling Weekly Gross S y 10-6 Evaluate intersystem water system leakage l
l (a) Daily meansfiw timesper week. (b) Perfon..cd morefrequently ifincrea:e noted on daily gamma count. O\: TABLE 11.5-5 RADIOLOGICAL ANALYSIS
SUMMARY
OF GASEOUS PROCESS SAMPLES Sample Sensitivity Sqmple Description Frecuency bnalysis pCi/ml Purnose
- 1. Containment atmosphere Periodically Grost a & 10-11 Determine riced for respi-(drywell) and prior to Tritium 10-6 ratory equipment l entry
- 2. Offgas monitor sample Weekly Gamma spectrum 10-10 Determir.e ofigas activity
- 3. Offgas vent sample Weekly Gross g(a) 10-11 Determine offgas system I131(b) 10-10 cleanup Gamma spectrum 10-10 t0) On particulatefilter.
(b) On charcoal cartridge. Amendment 17 11.5-20
.ABWR umma Standard Plarti ni!v. n i TAHLE 11.5 7 -
RADIOLOGICAL ANALYSIS
SUMMARY
OF GASEOUS EFFLUENT SAMPLES Sample . Sensitivity Salmsle Destdatll:A EttsutDc!' ADalnla DCl/mi Purpose l1. Detergent drain tanks llatch(a) Gro s Gamma 10-7 IIffluent discharge record
- 2. Liquid radwaste Weekly (b) lla/La,140 and 5 x 10'7 IIIfluent discharge accord
- effluent 1131 composite of all Monthly Gamma spectrum 5 x 10-7 Tritium 10 5 Gross alpha 10 7 Dissolved gas (C) 10 5 Ouarterly Sr/89/90 5 x 10 8 l
l3. Circulating water Weekly grat> Gross Gamma 10 ' IIffluent discharge record dicent line of Tritium 10-5 (backup sample) continuously collected
.. proportional sample 4 Reactor Service Weekly Gross Gamma 10 7 Ilffluent dischar;c record Water Tritium 10-5 (0) If tank is to be discharged, analysis will be perforrned on each batch, if tank is n%r 'c I;-
discharged, analysis will be performed periodically to evaluate equipment perforn ~cs, lb) Typicalbatch of average release. Allothersarnples areproportionalcomposites.
.(c) - llno discharge event occurs during the week, frequency shall be so adjusted.
l i l l r V. I Amendment 17 315 21
-, . _ , ~ _ _ ..___u... , , _ - - . . . . _ . . _ , _ _
ABWR 2346iooix Standard Plant , REV.B TAULE 11.5-7 g RADIOLOGICAL ANALYSIS
SUMMARY
OF GASEOUS EFFLUENT SAMPLES l l Sample Sensitivity l Samole Descrintion fngynn Analysis pCl/ml furpose
- 1. Plant vent exhaust thru Weekly Gross B (a) 10-11 Effluent record i stack
- l131(b) 10-10 l Ba/La 140 (a) 10-9 ;
i I Monthly Gamma spectrum (a) 10-10 l 133 and 135(b) 10-10 Tritium 10-6 Quarterly Sr 89 and 90(a) 10-11 Gross alpha (a) 10-11
- 2. Gland steam condenser As above As abcve Effluent record exhaust discharge O
(0) On paniculatefilter. (b) On charcoalcartridge. This includes off-gas exhausts from the reactor building, turbine building, radwaste building and service building. O Amendment 17 11522
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Standard Plant Riv c-
- Table 15.0 2 4
O-
. RESULTS
SUMMARY
OF SYSTEM RESPONSE ANALYSIF TRANSIENT EVENTS (Cont.) - Mas Core - No. Max. Max. Average of Duration Mac M u. Vesel Steam - Surfue Valves ' of Sub Neutron - Dome . Hottom ' une fleat lius A Freq. First Bicm4 m Section Figure llux Pressure Preuure Presure (% of in Cate- Dicw- (uconds) - d 'd- Desenntion ftlGB (Kr!Crog) 2 fKr/Cm2g] pat /Cm2 g) - ggi 39 gg,, g73 62 15 A.1.2 RWE-Startup SEE HiAT 15A.2 RWE et Pcmcr SEE n!XT 15A3 Control Rod SEE RAT Misoperation
' 15A 4 - Abnormal Startup SEE MIXT of One Reactor Internal Pump -
15A3 15 4-2 Fast Runout 89.8 71.1 723 70.6 116.1 a 0 0 of One Reactor - Internal Pump 15A3 15A-3 Fast Runout 115,0 72.5 74.7 713 1683 "" a+ 0 0 of All Reactor
~
Internal Pumps l 15 A.7 Misplaced Bundic SEE RIXT Accident - 15.4.8 - Rod Ejection Accident SEE RIAT - 15 A.9 Control Rod Drop Accident SEE TEXT 15 3 - Increase in Reactor Coolant Inventory 1$11 15 3-1 Inadvertent 102.0 711 75 6 71,6 100.0 a 0 0 IIPCF Startup Frequency definition is discussed in Subsection 15.04.1 _ Not limiting ( e Subsection 15.0.4.5.) Transients initiatedfrom low power. a Wderate Frequency b Infrequent c Limiting Fault
-+ This event should be classified as a limitingfault. However, criteria for inoderatefrequent incidents are conservatively ap;> lied.
Amendment 17 150-10
. - _ _-__-_----_______.__-_-____-_:_______.. ....._-._a
ABWR nmma Standard Plant inw. g Table 15.0 3
SUMMARY
OF ACCIDENTS FAILED FUEL RODS GE NRC SUBSECTION CALCULATED WORST CASE I.D. TITLE YALUE ASSUMITION 15 3.1 Trip of All Re.setor Internal Pumps None ~ 60% 1533 Seizure of one Reactor Internal Pamp None None 15 3.4 Reactor Internal Pump Shaft Break None None 13.6.2 Instrument Line Break None None 15.6.4 Steam System Pipe Break Outside None None Containment 15.6.5 LOCA Within RCPB None 1007c 15.6.6 Feedwater Line Break None None 15.7.1.1 Main Corsdenser Gas Treatment N/A N/A System Failure 15.7 3 Liquid Radwaste Tank Failme N/A N/A 15.7.4 Fuel-liandling Accident < 125 125 15.7.5 Cask Drop Accident None All Rods in Cask i Table 15.0-4 CORE W1DE TRANSIENT ANALYSIS RESULTS TO BE PROVIDED FOR DIFFERENT CORE DESIGN MAX, CORE MAX. AVERAGE NEUTRON SURFACE FLUX llEAT FLUX DELTA TRANSIENT (%NHR) ('fr$ BR) QB FIGUR.E Closure of One Turbine Control Valve X X X X Load Rejection with all Bypass Valves X X X X Failure Runout of 2 Feedwater Pumps X X X X 15 &11 Amendment 15
ABWR m-n Standard I'lant RIV C (igure 15.3 2 graphically shows this event all ten reactor internal pumps (RIPS) at the hV with the minimum specified rotating inertia for same speed. As presented in Subsection the RIPS. The vessel water level swell due to 15.1.2.1.1, no credible single failure in the rapid flow coastdown is expected to reach the control system will result in a minirnum demand high level trip, thereby tripping the main to all RIPS. A voter or actuator failure may turbine and feed pumps. Subsequent events, such result . inadvertent runback of one RIP at as initiation of the RCIC system occurring late its maximum drive speed (-40%/sec.). In this in this event, have no significant effect on the case, the RFCS will sense the core flow change results. The peak clad temperature during this and command the remaining RIPS to increase event is calculated to be less than 6000C, speeds and thereby automatically mitigate the which is Niow the applicable limit of 12000C. transient and maintain the core flow. . i 153.1.4 Barrier Performance As tresented in Subsection 15.1.2.1.1, multiple failures in the control system might 15 3.1.4.1 Trip of Three Reactor Internal Pumps cause the RFCS to erroneously issue a minimum demand to all RIPS. Should this occur, all RIPS The results shown in Figure 1531 indic4te could reduce speed simultaneously. Each RIP that peak pressures stay well below the 96.7 drive has a speed limiter which limits the Kg/cm2g limit allowed by the applicable code. maxin um speed change rate to 5%/sec. Iloweser, Therefore, the barrier pressure boundary is not the probability of this event occurring is low threatened. (less than 7 x 10 6 failures per reactor Scar); and hence, the event should be considered 15 3.1.4.2 Trip of All Reactor internal Pumps as a limiting fault. Ilowes er, criteria for moderate frequent incidents are conservatively The results shown in Figure 15.3.2 indicate applied. that peak pressures stay well below the limit allowed by the applicable code. Therefore, the 153.2.1.2 Frequency Classification O barrier pressure boundary is not threatened. 1.AJ.2.1.2.1 Fast Runback of One Reactor 153.1.5 Radioh>gical Consequences Internal Pump Trip of all 10 internal pumps due to a loss of The failure rate of a voter or an actuator is power supply is considered extremely unlikely to about 0.0088 f ailures per reactor year. result in perforation of fuel under conditions of However, L is analyzed as an incident of boi ing transition. The release of fission moderate frequency. products would be, however, rauch less than that assumed in the Loss of Coolant Accident for an 153.2.1.2.2 Fast Runhack of All Reactor event of equal probability. Therefore. the Internal Pumps radiological exposures noted in Subsection 15.6 5 cover the consequences of this event. This event should be classified as a limiting fault esent. Howeser, criteria for moderate 15.3.2 Recirculation Flow Control frequent incidents are conservatively applied. Failure--Decreasing Flow 153.2.2 Sequence of Esents and Systems 153.2.1 Identification of Causes and Operation l
- l Frequency Classification 153.2.2.1 Sequence of Esents 15 3.2.1.1 Identification of Causes 15 3.2.2.1.1 Fast Runback of One Reactor The recirculation flow control system (RFCS) Internal Pump uses a triplicated, f ault-tolerant digit al control system, instead of an analog system as Table 153 3 lists the sequerice of events for used in BWR 2 through BWR 6. The RFCS controls Figure 153 3.
Amendment 17 15 3-3 I 1_________________-_________-__-_-._--
ABWR mamn Standard Plant RPV.A 15 3.2.2.1.2 Fast Runhack of All Reactor Failure can result it. the maximum speed of Internal Puraps the RIP decreasing at a rate of 40%/sec as limited by the pump drive. Table 15.3-4 lists the sequence of esents for Figure 15.3 4. 153.23.1.2 Fast Runback of All Reactor Internal Pumos 15.3.2.2.13 Identification of Operator Actions A downscale failure of the master controller 153.2.2.13.1 Fast Runback of One Reactor will generate a zero flow demand signal to all Internal Pump RIPS Each individual RIP drive has a speud limiter which limits the maximum speed decrease As soon as possible, the operator verifies to a rate of 5%/sec. Core flaw decreases to that no operating limits are being exceeded. The approximately 40% of rated. This is the flow operator determines the cause of failure prior to expected when the RIPS are traintained at their returning the system to normal. minimum speeds. 153.2.2.1.3.2 Fast Rucback of All Reactor 153.23.2 Results Internni Pumps 153.23.2.1 Fast Runback on One Reactor As soon as postible, the operator verifies Internal Pump that no operating limits are being exceeded. If they are, corrective actions must be initiated. Figure 15.3 3 illustrates the fast runback of Also, the operator determines the cause of the one RIP event with the maximum rate which is failures r** to returning the system to normal, limited by hydraulic means. The MCPR remains above the safety limit. Therefore, this event F.3.2.2.2 Systems Operation does not have to be reanalyzed for specific core configurations. 15.3.2.2.2.1 Fast Runback of One Reactor l Internal Pump 15.3.2.3.2.2 Fast Runback of All Reactor l Internal Pumps Normal plant instrumentation and control is assumed to fanction. Figure 15.3 4 illustrates the expected event. Design of limiter operation is intended 15.3.2.2.2.2 Fast Ru'iback of All Reactor to render this event to be less severe than the Internal Pumps trip of all RIPS. No fuel damage is expected to occur. Therefore, this event does not have to Normal plant instrumentation and control is be reanalyzed for specific core configurations, assumed to function. 153.2.4 Barrier Performance 153.23 Core and System Performance 153.2.4.1 Fast Runback of Gne Reactor Internal 153.2.3.1 Input Parameters and Initial Pump l
- Conditions Peak pressures are less than those for the 153.23.1.1 Fast Runback of one Reactor Fast Runhack of All RIPS presented in Internal Pump Subsection 15.3.2.4.2.
O 15 3-4
~ . . ~ - .- .- - - - ._. .- - - . ~. - - ~ . - ABWR mmn .
EinmAard Plaa' su u the ovetcurrent protection logle of the controls. Nu protection systems ticilon is O electrical bus which supplies the power to the idie RIP. This electrie but is tripped by the the event. anticipated. No ESP action occers as a result of protection logic. Consequently, the other RIPS powered by this electrical bus are also tripped. 15.4.4.3 Core and System Performance Therefore, an abnottahl restart of the idle RIP i becomes a trip of one or two RIPS, which is An abnormal restart of a idle RIP becomes a presented in Subsection 15.3.1. trip of one or two RIPS event which is presented in Subsection 15.3.1. 15 4.4.1.1.1 Normal Restart of Reactor late. nal Pump 15.4.4.4 Barrier l'erformance This transient is categorized as an incident No evaluation of nreier performance is of mode. ate frequency. required for this event because no significant pressure increases are incurred duririg this 15.4.4.1.1.2 Abnormal Startup of l<lte Reactor transient (see Subsection 15.3.1), i Internal Pump at High Power l 15.4.4.5 Radiological Consequences i This transient sho ald be considered .s a limiting fault. However, crite/ al for moderate An evaluation of the radiological !
; frequent incidents are conservatively applied, consequet..es is not required for this event, becacte no radioactive materialis released from 15.4.43 Sequence of Dents and Systems the fuel, i
- Operstlen -
15A.5 Recirculation Flow Control Fallure 15.4.4.2.1 Sequence of Dents With Increasing Flow , Table 15.4 3 lists the_ sequence of events for 15.4.5.1 Identification of Causes and Frequene) ; an abnormal startup of an idle RIP. Classification . r 15.4.4.2.1.1 Operator Actions 15.4.5.1.1 Identification of Causes The norlnal sequence of operator *ctions The AllWR recirculation flow control system expected in starting the idle loop is tas (RFCS) uses a triplicated, fault. tolerant digital follows. The operator should control system. The RFCS controls all ten - reactor internal pumps (RIPS) at the same speed. (1) adjust rod pattern, as necessary, for new . As presented in Subsection 15.1.2.1.1, n o !
. power level following idle RIP start; credible single failure in the control system results in a maximum demand to all RIPS. A voter 1
(?) reduce the speed of the running RIPS to or actuator failure may result in an inadvertent their minimum speeds; runout of one RIP at its maximum drive speed - (~40%/sec). In this case, the RFCS senses the ; (3)' start the U : loop pump and adjust speed to core flow change and commands the remaining RIPS match th woning RIPS (monitor reactor to decrease speed and thereby automatically power); arid mitigate the transient and maintains the core flow, i (4) readjust power, as necessary, to satisfy plant ret,uirements per standard procedure. - As presented in Subsection 15.1.2.1.1, multiple fai6tes in the control system might 15.4.4.2.2 Systems Operation cause the RFCS to erroneously issue a maximum demand to all RIPS. Should this occur, all RIPS This event assumes and takea credit for normal could increase speed simultaneously. Each RFCS functioning of plant instrumentation and Amendment 17 1345 _ _ - _ . _ -n __.m._ _ ,_.u._,. . _ _ _ _ . - . _ _ _ . _ , . . _ . _ . . . . . - . . _ , _ _ _ _ . . _ . . , - . . . . . _
ABWR a m maAn RN c Standard Plant processing channel has a speed demand limiter action is to hold reactor pressure and condenser which limits the maximum speed change rate to vacuum for testart after the malfunction has bon 5%/see flowever, the probability of this esent repaired. The following is the sequence of occurring is low (less than 7 x 10 5 ailures f operator actions expected during the course of per reactor yeat); and hence the event should be the eveet, assumlag testart. The operator considered as a limiting fault. should: 15.4.5.1.2 Frequency Classification (1) observe that all rods are in; 15.4.5.1.2.1 Tast Runout of One Reactor Internal (2) check the reactor water level and maintain Pump above low level (L2) trip to pi > nt RCIC initiation; Tbc failure rate of a vt,ter or an actuator is about 0.0088 failure per reactor year. Ilowever, (3) switch the reactor mode r '"h to the STARTUP l It is analyzed as an incident of moderate position; l frequency. (4) maintain vacuum and turbine seals; 15.4.5.1.2.2 Fast Runout of All Reartr. Internal l j Pumps (5) transfer the recirculation flow controller to the manual position and reduce setpoint to This event should be considered as a limiting zero; fault, flowever, criteria for moderate frequent l lucidents are conservatively applied. (6) survey maintenance requirements and complete l the scram report; j 15 4.5J Sequence of Events and Systems Operation (7) monitor the turbine coastdown and auxiliary systems; and 15.4.5.2.1 Sequence of Events (8) establish a restart of the reactor per the 15.4.5.2.1.1 rest Runout of One Reactor internal normal procedure. P9mn
!!.4.512 Systems Operation Table 15.4 4 lists the sequence of events for Figure 15.4 2. The analysis of this transient assumes and takes credit for normal functioning of plant 15.4.5.2.1.2 Fast Runout of All Reactor internal instrumentation and controls and the reactor Pumps protection system. Operation of engineered s safeguards is not expected.
Table 15.4 5 lists the sequence of eveats for Figure 15.4 3. 15.4.53 Core and System Performance 15.4.5.2.1 3 Identitication of Operrtor Actions 15.4.5 3.1 loput Parameters and inillal Conditions The operator should: In each of these events, the most severe (1) transfer flow control to manual and reduce consequences result when initial conditions are now to minimum, and established for operation at the low end of the rated flow control rod line. Specifically, this (2) identify cause of failure, is 59% NBR power and 42% core flow. The maximum speed increuf,g rate of 40%/sec is assumed for Reactor pressure is controlled as required, one RIP runcut. depending on whether scram occurs, and,if scram occurs whether a restart or cooldown is planned. For all Rips runout,5%/r,ec is assumed for the in general, following a scram the corrective speed limit. The maximam core flow achiesed by Amendtnent 17 1546
ABWR m-u Standard Plant litv c 15A INCREASE IN REACTOR COOLANT INVENTORY 15.5.1J.1 luput Parameter and initial Conditions 15.5.1 InadycHent IIPCF Stadup The water temperature of the .lPCP system is 15.5.1.1 Identification of Causes and rtrquene) assumed to be 400F with an enthalpy of 11 Classincation litu/lb.
-15.5.1.1.1 Identification of Causes inadvertent startup of the llPCP system is chosen to be analyred, because _it provides the hianual startup of the llPCP system is greatc61 ausiliary source of co!d water into the postulated for this analysis (i.e., operator vessel.
error). 15.5.13J Results 15.5.1.1J Frequency Classincation Figure 15.$ 1 shows the simulated Itansient This transient disturbance is categorired as event for the manual flow control mode.11 an incident of moderate frequency. begins with the introduction of cold water into the upper core plenum. Within i see, the full 15.5.1J Sequence of f. vents and System llPCF flow is established at approsimately 32% Operation (.f rated feedwater flow rate. This flow is nearly 13Fe of the llPCF flow at rated pressure. 15.5.1J.1 Sequence of thents No delays are considered because they are not relevant to the analysis. Table 15.51 lists the sequence of events for Figure 15.51. Addition of cooler water to the upper plenum
.O 1 53.1J.1.1 Identincation of 0perator Actions causes a reduction in steam flow, which results in some depressuriration as the pressure regulator responds to the event. The flux level Relatively small changes are be experienced settles out slightly below operating level.
In plant enndirlant. The operator should, after Pressure and thermal variations are relatively hearing the alarm that the llPCF nas cc.mmenced small and no significant consequences are operation, check reactor water level and drywell aparienced htCPR remains above the safety _ pressure. If conditions are normal, the operatot limit and, thereferr. fuel thermal margins are , shuts down the system. maintained. Therefore, this event dm not have to be reanalyred for specific core configura. 15.5.1J.2 System Operation tions. To properly simulate the expected sequence of 15.5.133 Consideration of tintertaintles events, the analysis of this ment assumes normal functioning of plant instrucentation and important analytical f actors, including controls specifically, the pressure regulation reactivity coefficient and feedwater temperature and the vessel level control which respond change, are assumed to be at the worst directly to this r gnt, conditions so that any sieviations in the actual plant parameters will produce a less severe Required operation of engine red safeguards transient. other than what is described is not expected for this transient event. 15.5.1.4 Itarrier Performance The system is assumed to be in the manual flow Figure 1$.51 shows a slight prcuure control mode of operation. reduction from initial conditions; therefore, no further evaluation is required as RCPil piruute O- - 15.5.13 Core and System Performance margins ate maintained.
-- Arnendment 17 lbl
ABWR 2WIMAH Standard Plant RIY A 15.5.1J Radiological Consequences Because no activity is released during this event, a detalled evaluation is not required. 15.5.2 Chemical Volume Contml System Malfunction (or Operator Ermr) This section is not applicable to BWR. 15.5.3 IlWR Transients Which Inetesse Reactor Coolant Inventory These events are esented anc considered in Section 15.1 and 15.2. O O 133 2 l
ABWR mi-nir c Elandard Plant o SECTION 15.6 ft ') TAllLES Inhle Illte Pilge 15.6 1 Instrument Line lireak Accident Parameters 15.6 11 15.6 2 Instrument Line Break Accident Isotopic Inventory 15.6 12 15.6 3 lastrument Line lireak Accident Results 15.6-13 15.6-4 Sequence of IIvents for Steamline lireak Outside Containment 15.6 14 15.6-5 5teamline lireak Accident Parameters 15.6-15 15.6-6 hiain Steamline lireak Accident Activity Released to l'nvironment in Curies 15.6 16 15.6-7 hiain Steamline Ilicak hieteorology Parameters and Radiological Effects 15617 15.6-8 less of Coolant Accident Parameter 15.6 18 m. 15.6-9 Primary Containment Accident in Curies 15.6 19 15.6-10 Integrated Activity Released to Emironment in Curies 15.6-20 15.6-11 Integrated Control Room Activity in Curies second 15.6 21 15.6-12 Control Room hieterology and Doses 15.6 22 15,6 13 less of Coolant Accident Site !!oundary 2 llour Dose in REh! 15.6 22 15.6 14 Loss of Coolant Accident lew Population 7ene floundary Doses 15.6 23 15.6-15 Sequence of Esents fy Feedwater Line lircak Outside Containment 15.6 24 15.6-16 I cedwater Line lireak Accident Parameters 15.6 25 15.6 17 l'ecdwater Line Ilreak Isotopic Release in Curies 15 6-26 15.6-18 Feedwater Line Break hieteorology and l nose Resuiis 25.6-27 15.64i Amendment 17
+Mdi%CT)MDi. -
Y i' u x,-((grR 2m"c^ii RIT C k f/ _ dardHant v i SECTION 15,6 1LLUSTRATIONS Uguts 11 tic b ec 15.6-1 Steam flow Schematic for Steamline Break Outside Containment 15.6 23 154 2 LOCA Radiological Analysis 15.6-29
' 15.6 3 Airborne lodine in Primary Containment During Blowdown l'hase 15.6 29.1 15.6 4 ABWR Plant Layout 15.6-29.1 15.6-5 Leakage l'ath for reedwater Line Break Outside Containment 154 30 0
Amendment 8
i J ABWR zwmu Standard Plant nty c 15.6 DECREASE IN REACTOR COOLAhT 15.6.2.1.2 Frequency Classtrication l Oi INVENTORY This event is categorized as a limiting l 15.6.1 Inadvertent Safety / Relief Valve fault. Opening 15.6.2.2 Sequence of Eients and Systems l This event is presetated and analyzed in Operations i Subsection 15.1.4. 15.6.2.2.1 Sequence of Events 15.6.2 Failure of Small Line Carrying Primary Coolant Outside Containment The leak may result in noticeable increases in radiation, temperature, humidity, or noise
- This event postulates a sirsll steam or liquid levels in the secondary containment or abnormal ,
line pipe break inside or outside the primary indications of actuations caused by the affected ) containment, but within a controlled release instrument. structure. To bound the event, it is assumed 6 bat a small instrument line, instantaneously and Termination of the analyzed event is circumferentially, breaks at a location where it dependent on operator action. The action is may not be able to be isolated and where initiated with the discovery of the un' nlatable detection is not automatic or apparent. This leak. The action consist. of the 3rderly l event is less limiting than the postulated events shutdown and depressurization of the reactor presented in Sections 15.6.4,15.6.5, and 15.6.6. vessel. This postulated event represents the envelope evaluation for small line failure inside and 15.6.2.2.2 Systems operation outside the primary containment relative to sensitivity for detection. 15.6.2.1 Identification of Causes and A presentation of plant and reactor frequency Classification protection system action and ESF action is given in Sections 63,73, and 7.6. 15.6.2.1.1 Identification of Causes 15.6.2.23 The Effect of Single railures There is no identified specific event or and 0perator Errors circumstance which results in the failure of an instrument line. These lines are designed to There is no single failure or operator error high quality, engineering standards, seismic and that will significantly affect the system environmental requirements. liowever, for the response to this event. purpose of evaluating the consequences of a small line rupture, the failure of an instrument line 15.6.2.3 Core and System Performance ! is assumed to occur. Instrument line breaks, because of their A circumferential rupture of an instrument small size, are substantially less limiting from line which is connected to the primary coolant a core and systen4 performance standpoint that system is postulated to occur outside the the events examined in Sections 15.6.4, 15.6.5, drywell, but inside the reactor building. This and 15.6.6. Consequently, instrument line event could conceivably occur also in the - breaks are considered to be bounded specifically drywell, liowever, the associated effects would by the steam line break (Section 15.6.4). not be as significant as those from the failure Details of this calculation, including those in the reactor building. pertinent to core and system performance, are presented in Section 15.6.43. l O Amenament 17 l$ M
ABWR m-n Standard Plant nrv c 15.6.2.3.1 Input Parameters and initial iodine available in the flashed water is Conditions transported via the ilVAC system or blowout panels to the environment without prior All information concerning ECCS models treatment by the standby gas treatment system. employed, input parameters, and detailed results Other isotopes in the water contribute only for a more limiting (steam line break) event are negligibly to the iodine dose. presented in Section 6.3. 15.6.2.5.2 Fission Product Release 15.6.2.3.2 Resutts The iodine activity in the coolant is No fuel damage or core uncovering occurs as a assumed to be at the maximum equilibrium I result on this accident. Similarly, inst r u m e nt technical specification limit (see Section line breaks are within the spectrum considered in 15.6.4.5.1.1, case 1) for continious operation. ECCS performance calculations presented in The iodine released to the reactor building Section 6.3.3. atmosphere and to the environment are presented in Table 15.6-2. 15.6.2.4 Itarrier Performance 15.6.2.5.3 Rosulis The following assumptions and conditions are the basis for the mass loss during the release Results of the analysis are found in Table period of this event: 15.6 3 are are within the 10% of 10CFR100 specified in the standard review plan.
- 1. The instrument line releases coolant into the reactor building for a period of ten 15.6.3 Steam Generator Tube Failure minutes at normal operating temperature and pressure. Following this ten minute period. This section is not applicable to the direct the operator is assumed to have isolated the cycle 11WR.
event and taken steps to SCRAM the reactor to reduce reactor pressure over a period of 15,6.4 Steam System Piping Break Outside 4.5 hours. Containment
- 2. The flow from the instrument line is limited This event involves postulating a large by reactor pressure and a 1/4 inch diameter steam line pipe break outside containment. It flow restricting orifice inside the is assumed that the largest steam line, drywell. The Moody critical blowdown model instantaneously and circumferentially breaks at is applicable, and the flow is critical at a location down-tream of the outermost isolation the orifice (Reference 1). valve. The plani is designed to immediately detect such an occurrence, initiate isolation of The total integrated mass of fluid released the broken line and actuate the necessary into the reactor building is 12,000 lbs with protective features. This postulated event approximately 5,000 lb being flashed to steam. represents the envelope evaluation of steam line failures outside containment.
15.6A.1 Identincatkm of Causes and 15.6.2.5 Radiological Analpis Frequency Classification 15.6.2.5.1 General 15.6A.I.1 Identification of Causes 1 The radiological analysis is based upon A main steam line break is postulated l conservatise assumptions considered acceptable to without the cause being identified. These lines I the NRC. Though the standard review plan does are designed to high quality engineering codes not provide detailed guidance, the assumptions and standards, and to seismic and environmental found in Table 15,61 assume that all of the requirements. Ilowever, for the purpose of Amendment 17 15 42
ABWR momn Silindllid.13aul klYf 15.6.63.2 Qualitathe Results The analysis is based on a conservatise [ assessment of this accident. The specific The feedwater line break outside the models, assumptions and the program used for containment is less limiting *,han either of the computer evaluation are presented in Reference steam line breaks outsib the containment 2. Specific values of parameters used in the (analysis presented in Sections 6.3 and/or evaluation are presented in Table 15.616. A 15.6.4), the feedwater line break inside the schematic diagram of the leakage path for this containment (analysis presented in Subsections accident is shown in l'igure 15.6 3. 6.3.3 and 15.6.5). 1!.6.6.5.2.1 Iission Product Helease The reactor vessel is isolated on water level 1.5, and the RCIC and the lipCF systems together There is no fuel damage as a consequence of restore the reactor water level to the normal this accident, in addition, an insignificant elevation. The fuel is covered throughout the quantity of activity (compared to that existing transient and there are no pressure or in the main condenser hotwell prior to occur-temperature transients sufficient to cause fuel rence of the breal) is released from the con. damage. tained piping system prior to isolation closure. 15.6.633 Consideration of Uncertainties The iodine concer.' ion assumed s that of the maximum equilibrium reactor water This event was conservatively analyzed and concentration given in Submtion 15.6.4.5.1.1, uncertainties were adequately considered (see case 1, subject to a 29 carryover of iodine in Section 6.3 for details), the water to steam condensate. Noble gas activity in the condensate is negligible since 15.6.6.4 Harrier Performance the air ejectors remose all noble gases from the condenser. Accidents that results in the release of O radioactive materials outside the containment are the result of postulated breaches in the reactor 15.6.6.5.2.2 Fission Product Transport to the I:nvironment coolant pressure boundary or the steam power. conversion system boundary. A break The transport pathway consists of liquid spectrum analysis for the complete range of release from the break, carryoser to tht turbine reactor conditions indicates that the limiting building atmosphere due to flashing and fault event for breaks outside the containment is partitioning und unfiltered release to the a complete severance of one of the main steam environment through the turbine building lines as presented in Subsection 15.6.4. The ventilation system. feedwater system piping break is less sesere that the main steam line break. Results of analysis Of the 950,500 lb of condensate release from of this event can be found in Subsections 6.2.3 the break, 231,000 lb flashes to steam. or 6.2.4. Takig no credit for holdup, decay or 15.6.6.5 Radiological Consequences plate out during transport through the turbine building, the release of activity to the 15.6.6.5.1 Design llasis Analpis environment is presented in Table 15 617. 'lhe relcase is assumed to take place within 2 hours The NRC provides no specific regulatory of the occurrence of the break, guidelines for the evaluation of this accident; therefore, the analysis presented is based upon 15.6.6.5.2 3 Hesults consenative assumptions considered acceptable to the NRC. The calculated exposures for the analpis are presented in Table 15.618 and are a small 15.6.6.5.2 Analpis fraction of 10CFR100 guidelines. Amendment 17 6 fM
ABWR a miman Silmdard Plant new c 15.6.7 References
- 1. F.J. Moody, Ataximum Two. Phase l'essel Blowdownfrom Pipes, ASME Paper Number 65-WA/HT 1 March 15,1965.
- 2. II.A. Careway, V.D. Nguyen, and P.P.
Stanesvage, Radiological Accident Evaluation . The CONAC03 Code, December 1981 (NEDO-211431).
- 3. II.A. Careway, V.D. Nguyen, and D.G. Weiss, Control Room Accident hposure Evaluation, CRDOS frogram, Febr u a ry 1981 (NEDO 2fXY)A).
- 4. K.G. M urphy, and K.M. Campe, Nuc/ car rower Plant ControlRoom l'entilation System Design for hiceting General Design Criteria 19.13th ASC Air Cleaning Conference, June 1974.
- 5. LS.1xe, Increasing Main Steam Isolation l'alve Leakage Rate Limits and Elimination of Leakage Control Systems, November 1988 (NEDC 31643 P).
- 6. J.V. Ramsdell, Atmospheric Diffusion for Control Room Habitability Assessments, M ay 19S8 (NUREG/CR 5055).
O I Amc.ndment g 15610
ABWR me Standard.flant ni v c m
) Table 15.61 INSTRUA!ENT LINE ilREAK ACCIDENT PAllAh1ETERS I Data and assumptions used to estimate Source terms A. Power level 4(K)5 htWt
- 11. hiats of fluid released 12,(AK) Ib C. hiau ofIluid flashed to steam 5(AK)Ib D. Duration of accident 8 hours E. Number of bundles in core 872 11 Data and assumptions used to estimate activity released A. Jodine water concentration 15.6.4.5.1.1, case 1
- 11. lodine Spiling 1131 2.1 Curie / bundle I132 3.2 1133 5.0 11M 5.4 1 135 4.8 C. lodine platcout fraction 5(VE D. Reactor 11uilding Flow rate AA)'li / hour
/\ G E. SGTS Filter Efficiency None assumed w]
!!! Dispersion and Dose Data A. hieteorology Table 15.6 3
- 11. Iloundary and LPZ distances Table 15.6-3 C. hiethod of Dose Calculation Reference 2 D. Dose conversion Assuinptions Reference 2, RG 1.109, and ICRP 30 E. ActivityInventory/ releases Table 15.6-2 >
F. Dose Evaluations Table 15.6 3 /O V Amendment 17 15 f Il
1 ABWR mmen SIAltdard Plarit nw c Table 15.6 2 l INSTRUMENT LINE IIREAK ACCIDENT ISOTOPIC INVENTORY REACTOR IlUILDING INVENTORY IN CURIES l ISOTOPE 1.h11N 10.htlN 1.llOUR 2 IlOUR 4 ilOUR 8 llOUR I.31 1.02E-03 8.85E-03 7.02E-01 4.68E-01 3.74E41 1.2412-04 1132 9.94E 03 8.4112 02 6.23E + 00 3.90E + 00 3.15E + 00 3.17E-04 1133 7.01E 03 6.05E-02 4.73E + 00 3.13E + 00 2.51E 4 00 7.36E-04 1134 1.95E 02 1.60E-01 1.0$E + 01 6.12E + 00 5.03E 4 00 7.15E-05 1135 1.02E-02 8.78E-02 6.81E + 00 4.44E 4 00 3.56E + 00 7.85E-04 TT)TAL 4.77E 02 4.0!!M1 2.90E + 0! 1.81E + 0! 1.460 + 01 2.03 0-03 ISOTPIC RELEASE TO ENVIRONMENT IN CURIES ISOTOPE 1. AllN 10h11N 1 ilOUR 2 IlOUR 4 ilOUR 8 IlOUR l131 1.72E-05 1.56E 03 7.4SE-01 1.ME + 00 3.43E + 00 3.80E + 00 1132 1.67E-04 1.49E42 6.82E 4 00 1.61E + 01 2.94E + 01 3.21 E + 01 1133 1.18E-04 1.07E-02 5.05E 4 00 1.24E + 01 2.30E + 01 2.55E + 0! 1 134 3.2SE-M 2.87E-02 1.20E + 01 2.6SE 4 01 4.77E + 01 5.13E + 0! 1 135 1.72E-N 1.55E-02 7.32E + 00 1.76E + 01 3.28E + 01 3.62E + 01 TOTAL 8.01E 04 7.130 02 3.20E + 0! 7.50E+ 0! 1.36E+ 02 1.491:+ 02 O Arnendment 17 15412
AllWR mimn Standa[(U'lan! Riv. c Table 15,6 3 !n) INSTRUMENT 1.1NE IlitEAK ACCIDENT RESUI.TS hil:Tt:oRo!.Om'* AND DOSE kt:SI'!*lS hieteorology Distanct 1hyroid Dose Whole Body Dose 3 (sec/m ) (en) (Hem) (Rem) 8.59E-03 mas 30 o.0!!-01 2.19E-04 8(O 7,61!-01 1.5E 02 1.11E-04 140 3.9E-01 7.9E-03 5.61E 05 32(O 2.0l! 01 4.Oli-03 ! 3.73E-05 48m 1.311 01 2.611 03 l 1 1
' hieteorology calculated using flegulatory Guide 1.145 for a ground level 1.0 m/s.17 stability release.
- mas" - masimum meteorology to meit 10% of 10C111100 limits.
(D ( \v) / g Amendment 17 13f-13
i ABWR mwan ; Standard Plant ni:v. n Table 15.6 4 SEQUENCE 01 EVENTS FOR STEAM LINE lillEAK OUTSIDE CONTAINMENT I Time hill}}