Text

Unit 2, Updated Final Safety Analysis Report, Revision 33
	ML22298A056
Person / Time
Site:	Salem, Hope Creek
Issue date:	10/24/2022
From:	; Public Service Enterprise Group
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML22298A053	List: LR-N22-0086, Unit 2, Updated Final Safety Analysis Report, Revision 33; ML22298A054;
References
	LR-N22-0086, NEI 99-04
	Download: ML22298A056 (1)
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{{#Wiki_filter:MASTER TABLE OF CONTENTS Section Title Volume 1 INTRODUCTION AND GENERAL DESCRIPTION l OF PLANT 1.1 PROJECT IDENTIFICATION l 1.2 PLANT SITE

SUMMARY

1 1.3

SUMMARY

PLANT DESCRIPTION 1 1.4 IDENTIFICATION OF CONTRACTORS 1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL 1 INFORMATION 1 1.6 LIST OF ACRONYMS 1 2 SITE CHARACTERISTICS 1 2.1 GEOGRAPHY AND DEMOGRAPHY l 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND 1 MILITARY FACILITIES 2.3 METEOROLOGY 1 APP. 2.3A JOINT FREQUENCY DISTRIBUTIONS OF WIND 1 SPEED AND DIRECTION BY LAPSE RATE DELTA STABILITY CLASSES: JUNE 1969 TO NOVEMBER 1971 2.4 HYDROLOGIC ENGINEERING l 2.5 GEOLOGY ANO SEISMOLOGY l 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT 1 AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA l 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, l AND SYSTEMS 3.3 WIND AND TORNADO LOADINGS 1 i SGS-UFSAR Revision 14 December 29, 1995

MASTER TABLE OF CONTENTS (Cont) Section Title Volume 3.4 WATER LEVEL (FLOOD) DESIGN 1 3.5 MISSILE PROTECTION 1 3.6 PROTECTION AGAINST DYNAMIC EFFECTS 1 ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING APP. 3.6A DESCRIPTION OF BACKDRAFT DAMPER 1 3.7 SEISMIC DESIGN 2 3.8 DESIGN OF CATEGORY I STRUCTURES 2 3.9 MECHANICAL SYSTEMS AND COMPONENTS 2 APP. 3.9A BOLTED CONNECTIONS FOR LINEAR COMPONENT 2 SUPPORT 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 2 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT 2 3.12 CONFORMANCE TO RULES ISSUED AFTER PLANT 2 LICENSING APP. 3A PUBLIC SERVICE ELECTRIC & GAS POSITIONS ON USNRC REGULATORY GUIDES 2 4 REACTOR 2 4.1

SUMMARY

DESCRIPTION 2 4.2 MECHANICAL DESIGN 2 4.3 NUCLEAR DESIGN 2 4.4 THERMAL ANO HYDRAULIC DESIGN 2 4.5 RELOAD ANALYSES 2 s REACTOR COOLANT SYSTEM AND CONNECTED 3 SYSTEMS 5.1

SUMMARY

DESCRIPTION 3 ii SGS-UFSAR Revision 15 June 12, 1996

MASTER TABLE OF CONTENTS {Cont) Section Title Volume 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 3 5.3 THERMAL HYDRAULIC SYSTEM DESIGN 3 5.4 REACTOR VESSEL AND APPURTENANCES 3 5.5 COMPONENT AND SUBSYSTEM DESIGN 3 5,6 INSTRUMENTATION APPLICATION 3 6 ENGINEERED SAFETY FEATURES 3 6,1 CRITERIA 3 6,2 CONTAINMENT SYSTEMS 3 6.3 EMERGENCY CORE COOLING SYSTEM 3 6.4 HABITABILITY SYSTEMS 4 7 INSTRUMENTATION AND CONTROLS 4

7.1 INTRODUCTION

4 7.2 REACTOR TRIP SYSTEM 4 7.3 ENGINEERED SAFETY FEATURES INSTRUMENTATION 4 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 4 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 4 7.6 ALL OTHER INSTRUMENTATION REQUIRED FOR SAFETY 4 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 4 7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) 4 7.9 PHYSICAL SEPARATION AND ELECTRICAL ISOLATION 4 7.10 SAFETY PARAMETER DISPLAY SYSTEM (SPDS) 4 APP. 7A PROGRAM ELEMENT DESCRIPTIONS RESULTING FROM 4 THE IMPLEMENTATION OF THE MAY 6, 1983 ORDER (REACTOR TRIP BREAKER ATWS EVENT} 8 ELECTRICAL SYSTEMS 4

8.1 INTRODUCTION

4 8.2 OFFSITE POWER SYSTEM 4 8.3 ONSITE POWER SYSTEM 4 iii SGS-UFSAR Revision 23 October 17, 2007

MASTER TABLE OF CONTENTS (Cont) Section Volume 9 AUXILIARY SYSTEMS 4 9.1 FUEL STORAGE AND HANDLING 4 9.2 WATER SYSTEMS 4 9.3 PROCESS AUXILIARIES 5 9.4 HEATING, VENTILATION AND AIR CONDITIONING SYSTEMS s 9.5 OTHER AUXILIARY SYSTEMS 5 10 STEAM AND POWER CONVERSION SYSTEM 5 10.l

SUMMARY

DESCRIPTION 5 10.2 TURBINE GENERATOR 5 10.3 MAIN STEAM SYSTEM 5 10.4 OTHER FEATURES OF THE STEAM ANO POWER CONVERSION SYSTEM 6 11 RADIOACTIVE WASTE MANAGEMENT 6 11.1 SOURCE TERMS 6 11.2 LIQUID WASTE SYSTEM 6 11.3 GASEOUS WASTE SYSTEMS 6 11.4 RADIOLOGICAL MONITORING 6 11.5 SOLID RADWASTE SYSTEM 6 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 6 12 RADIATION PROTECTION 6 12.l SHIELDING 6 12.2 VENTILATION 6 12.3 RADIATION PROTECTION PROGRAM 6 12.4 ALARA PROGRAM 6 iv SGS-UFSAR Revision 9 July 22, 1989

MASTER_ TABLE OF CONTENTS (Cont) I sebtion Title Volume 13 CONDUCT OF OPERATIONS 6 13.1 ORGANIZATION STRUCTURE 6 13.2 TRAIN I NG,'f;;PROGRAM 6 13.3 EMERGENCY PLANNING 6 13.4 REVIEW AND AUDIT 6 13.5 PLANT PROCEDURES 6 13.6 PLANT RECORDS 6 13.7 SECURITY 6 14 INITIAL TESTS AND OPERATION 6

14.1 DESCRIPTION

OF TEST PROGRAM 6 14.2 INITIAL TESTS AND OPERATION 6 14.3 FINAL STATION PREPARATION 6 14.4 INITIAL TESTING OF THE OPERATING REACTOR 6 14.5 TEST PROGRAM ORGANIZATION - UNIT 1 6 14.6 TEST PROGRAM ORGANIZATION - UNIT 2. 6 15 ACCIDENT ANALYSIS _6 15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL TRANSIENT~ 6 15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY 7, 15.3 CONDITION I II INFREQUENT FAULTS 7 APP. 15.3A DELETED 7 15.4 CONDITION IV -*LIMITING FAULTS 7 15.5 ANTICIPATED TRANSIENTS WITHOUT SCRAM 7 16 TECHNICAL. SPECIFICATIONS 7 V SGS-UFSAR Revision 15 June 12, 1996

MASTER TABLE OF CONTENTS (Cont) Section Title Volume 17 QUALITY AS$URANCE 7 17 .1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION PHASES 7 17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE 7 APPENDIX A TMI LESSONS LEARNED 7 I APPENDIX B LICENSE RENEWAL FINAL SAFETY ANALYSIS REPORT 7 SUPPLEMENT r vi SGS-OFSAR Revision 26 May 21, 2012

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV i 14 2-vii 16 T2.1-11 6 ii 15 2-viii 16 T2.1-12 6 iii 23 2-ix 6 F2.1-1 6 iv 9 2-x 6 F2.1-2 6 v 15 2-xi 6 F2.1-3 6 vi 26 2-xii 27 F2.1-4 6 1-i 6 2-xiii 6 F2.1-5 6 1-ii 6 2.1-1 6 F2.1-6 6 1-iii 7 2.1-2 6 F2.1-7 6 1.1-1 19 2.1-3 6 F2.1-8 6 1.1-2 25 2.1-4 6 F2.1-9 6 1.2-1 6 2.1-5 6 F2.1-10 6 1.2-2 6 2.1-6 6 F2.1-11 6 1.2-3 6 2.1-7 6 F2.1-12 6 F1.2-1 17 2.1-8 6 F2.1-13 6 F1.2-2 27 2.1-9 6 F2.1-14 6 1.3-1 6 2.1-10 6 F2.1-15 6 1.3-2 27 2.1-11 6 F2.1-16 6 1.3-3 19 2.1-12 6 F2.1-17 6 1.3-4 24 2.1-13 6 F2.1-18 6 1.3-5 25 2.1-14 6 2.2-1 6 1.3-6 14 2.1-15 6 2.2-2 6 1.3-7 25 2.1-16 6 2.2-3 6 1.3-8 14 2.1-17 6 2.2-4 6 1.3-9 15 2.1-18 6 2.2-5 6 1.4-1 6 2.1-19 6 2.2-6 6 1.4-2 18 2.1-20 6 2.2-7 6 1.5-1 6 2.1-21 6 2.2-8 6 1.5-2 6 2.1-22 6 2.2-9 23 1.5-3 6 2.1-23 25 2.2-10 33 1.5-4 6 2.1-24 6 2.2-11 33 1.6-1 10 2.1-25 6 2.2-12 23 1.6-1a 10 2.1-26 6 2.2-12a 26 1.6-1b 10 2.1-27 6 2.2-12b 13 1.6-2 16 2.1-28 6 2.2-13 12 1.6-3 16 T2.1-1 sh1 of 2 6 2.2-14 12 1.6-4 10 T2.1-1 sh2 of 2 6 2.2-15 16 1.6-4a 10 T2.1-2 6 T2.2-1 sh1 of 2 6 1.6-4b 10 T2.1-3 6 T2.2-1 sh2 of 2 6 1.6-5 16 T2.1-4 sh1 of 2 6 T2.2-2 sh1 of 1 33 1.6-6 22 T2.1-4 sh2 of 2 6 T2.2-2 sh2 of 2 R 1.6-7 6 T2.1-5 6 T2.2-3 16 1.6-8 15 T2.1-6 6 T2.2-4 sh1 of 2 26 1.6-9 6 T2.1-7 6 T2.2-4 sh2 of 2 33 1.6-10 10 T2.1-8 6 F2.2-1 6 2-i 6 T2.1-9 sh1 of 3 25 2.3-1 6 2-ii 11 T2.1-9 sh2 of 3 6 2.3-2 6 2-iii 16 T2.1-9 sh3 of 3 6 2.3-3 6 2-iv 6 T2.1-10 sh1 of 3 6 2.3-4 25 2-v 6 T2.1-10 sh2 of 3 6 2.3-5 6 2-vi 6 T2.1-10 sh3 of 3 6 2.3-6 6 Page 1 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 2.3-7 23 2.3A-5 R 2.3A-56 R 2.3-8 33 2.3A-6 R 2.3A-57 R 2.3-9 33 2.3A-7 R 2.3A-58 R 2.3-10 33 2.3A-8 R 2.3A-59 R 2.3-10a R 2.3A-9 R 2.3A-60 R 2.3-10b R 2.3A-10 R 2.3A-61 R 2.3-11 18 2.3A-11 R 2.3A-62 R 2.3-12 16 2.3A-12 R 2.3A-63 R 2.3-13 16 2.3A-13 R 2.3A-64 R 2.3-14 21 2.3A-14 R 2.3A-65 R 2.3-14a 16 2.3A-15 R 2.3A-66 R 2.3-14b 16 2.3A-16 R 2.4-1 27 2.3-15 16 2.3A-17 R 2.4-2 27 T2.3-1 6 2.3A-18 R 2.4-3 27 T2.3-2 6 2.3A-19 R 2.4-4 6 T2.3-3 6 2.3A-20 R 2.4-5 6 T2.3-4 6 2.3A-21 R 2.4-6 16 T2.3-5 6 2.3A-22 R 2.4-7 6 T2.3-6 6 2.3A-23 R 2.4-8 6 T2.3-7 6 2.3A-24 R 2.4-9 6 T2.3-8 6 2.3A-25 R 2.4-10 6 T2.3-9 18 2.3A-26 R 2.4-11 6 T2.3-10 18 2.3A-27 R 2.4-12 6 T2.3-11 6 2.3A-28 R 2.4-13 23 T2.3-12 6 2.3A-29 R 2.4-14 23 T2.3-13 18 2.3A-30 R 2.4-15 23 T2.3-14 18 2.3A-31 R 2.4-16 6 T2.3-15 18 2.3A-32 R 2.4-17 6 T2.3-16 sh1 of 1 33 2.3A-33 R 2.4-18 6 T2.3-16 sh2 of 3 R 2.3A-34 R 2.4-19 6 T2.3-16 sh3 of 3 R 2.3A-35 R 2.4-20 6 T2.3-16a 21 2.3A-36 R 2.4-21 6 T2.3-17 6 2.3A-37 R 2.4-22 6 T2.3-18 6 2.3A-38 R 2.4-23 6 T2.3-19 6 2.3A-39 R 2.4-24 6 T2.3-20 15 2.3A-40 R 2.4-25 6 T2.3-21 16 2.3A-41 R 2.4-26 6 T2.3-22 16 2.3A-42 R 2.4-27 25 F2.3-1 6 2.3A-43 R 2.4-28 6 F2.3-2 6 2.3A-44 R 2.4-29 6 F2.3-3 6 2.3A-45 R 2.4-30 6 F2.3-4 6 2.3A-46 R T2.4-1 6 F2.3-5 6 2.3A-47 R T2.4-2 6 F2.3-6 6 2.3A-48 R T2.4-3 sh1 of 4 6 F2.3-7 33 2.3A-49 R T2.4-3 sh2 of 4 6 F2.3-8 33 2.3A-50 R T2.4-3 sh3 of 4 25 APP 2.3A 18 2.3A-51 R T2.4-3 sh4 of 4 25 2.3A-1 R 2.3A-52 R T2.4-4 sh1 of 2 6 2.3A-2 R 2.3A-53 R T2.4-4 sh2 of 2 6 2.3A-3 R 2.3A-54 R T2.4-5 sh1 of 3 6 2.3A-4 R 2.3A-55 R T2.4-5 sh2 of 3 6 Page 2 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV T2.4-5 sh3 of 3 6 F2.5-11 6 3.1-28 12 F2.4-1 6 F2.5-12 6 3.1-29 12 F2.4-2 27 F2.5-13 6 3.1-30 12 F2.4-3 27 3-i 12 3.1-31 12 F2.4-4 6 3-ii 16 3.1-32 20 F2.4-5 6 3-iii 24 3.1-33 12 F2.4-6 6 3-iv 25 3.1-34 12 F2.4-7 23 3-v 6 3.1-35 12 F2.4-8 6 3-vi 6 3.1-36 12 F2.4-9 6 3-vii 16 3.1-37 12 F2.4-10 6 3-viii 23 3.1-38 12 F2.4-11 6 3-ix 6 3.1-39 25 2.5-1 6 3-x 15 3.1-39a 25 2.5-2 6 3-xi 28 3.1-39b 12 2.5-3 6 3-xii 27 3.1-40 12 2.5-4 6 3-xiii 6 3.1-40a 7 2.5-5 6 3-xiv 6 3.1-40b 7 2.5-6 6 3-xv 27 3.1-41 6 2.5-7 6 3-xvi 27 3.1-42 6 2.5-8 6 3-xvii 27 3.1-43 6 2.5-9 6 3-xviii 6 3.1-44 25 2.5-10 6 3-xix 27 3.1-45 6 2.5-11 6 3-xx 13 3.1-46 6 2.5-12 6 3-xxi 27 3.1-47 6 2.5-13 6 3.1-1 12 3.1-48 6 2.5-14 6 3.1-2 12 3.1-49 25 2.5-15 6 3.1-3 12 3.1-49a R 2.5-16 6 3.1-4 32 3.1-49b R 2.5-17 6 3.1-5 12 3.1-50 6 2.5-18 6 3.1-6 12 3.1-51 25 2.5-19 6 3.1-7 12 3.1-52 25 2.5-20 6 3.1-8 12 3.1-53 25 2.5-21 6 3.1-9 12 3.1-54 25 2.5-22 6 3.1-10 16 3.1-55 R 2.5-23 6 3.1-11 12 3.2-1 6 2.5-24 6 3.1-12 12 3.2-2 17 T2.5-1 6 3.1-13 12 3.2-3 6 T2.5-2 sh1 of 2 6 3.1-14 12 3.2-4 6 T2.5-2 sh2 of 2 6 3.1-15 12 3.2-5 18 T2.5-3 sh1 of 2 6 3.1-16 12 3.2-6 18 T2.5-3 sh2 of 2 6 3.1-17 12 3.2-7 6 F2.5-1 6 3.1-18 12 3.2-8 18 F2.5-2 6 3.1-19 12 3.3-1 6 F2.5-3 6 3.1-20 12 3.3-2 6 F2.5-4 6 3.1-21 12 F3.3-1 6 F2.5-5 6 3.1-22 25 F3.3-2 6 F2.5-6 6 3.1-23 12 F3.3-3 6 F2.5-7 6 3.1-24 12 F3.3-4 6 F2.5-8 6 3.1-25 12 F3.3-5 6 F2.5-9 6 3.1-26 12 3.4-1 15 F2.5-10 6 3.1-27 12 3.4-2 27 Page 3 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 3.4-3 23 3.6-2 24 3.6-49 25 3.4-4 23 3.6-3 24 3.6-50 31 3.4-5 16 3.6-4 20 3.6-51 25 3.4-6 23 3.6-5 33 3.6-52 25 F3.4-1 6 3.6-6 33 3.6-53 25 F3.4-2 23 3.6-7 23 3.6-54 25 F3.4-3 23 3.6-8 23 3.6-55 31 F3.4-4 6 3.6-9 25 3.6-56 25 3.5-1 14 3.6-9a 25 3.6-57 33 3.5-2 22 3.6-9b 7 T3.6-1 33 3.5-3 32 3.6-10 7 T3.6-2 16 3.5-4 32 3.6-11 7 T3.6-3 6 3.5-5 32 3.6-12 16 F3.6-1 33 3.5-6 17 3.6-13 7 F3.6-2 20 3.5-7 28 3.6-14 7 F3.6-3 23 3.5-8 28 3.6-15 24 F3.6-4 20 3.5-9 24 3.6-16 18 F3.6-5 20 3.5-10 28 3.6-17 18 F3.6-6 6 3.5-11 7 3.6-18 16 F3.6-7 24 3.5-12 27 3.6-19 16 F3.6-8 18 3.5-13 21 3.6-20 6 F3.6-8a 18 3.5-14 21 3.6-21 21 F3.6-9 6 3.5-15 21 3.6-21a 16 F3.6-10 22 3.5-16 21 3.6-21b 16 F3.6-10a 22 3.5-17 16 3.6-22 24 F3.6-11 6 3.5-18 21 3.6-23 16 F3.6-12 6 3.5-19 32 3.6-24 16 F3.6-13 6 3.5-20 21 3.6-25 6 F3.6-14 6 3.5-21 23 3.6-26 25 F3.6-15 6 3.5-21a 23 3.6-27 16 F3.6-16 6 3.5-21b 21 3.6-28 16 F3.6-17 6 3.5-21c R 3.6-29 6 F3.6-18 6 3.5-21d R 3.6-30 27 F3.6-19 6 3.5-22 32 3.6-31 24 F3.6-20 6 3.5-23 32 3.6-32 24 F3.6-21 6 T3.5-1 21 3.6-33 6 F3.6-22 6 T3.5-2 21 3.6-34 27 F3.6-23 6 T3.5-3 21 3.6-35 27 F3.6-24 15 F3.5-1 27 3.6-36 27 F3.6-25 6 F3.5-2 27 3.6-37 7 F3.6-26 27 F3.5-3 27 3.6-38 29 F3.6-27 27 F3.5-4 21 3.6-39 25 F3.6-28 17 F3.5-5 6 3.6-40 27 F3.6-29 6 F3.5-6 21 3.6-41 16 F3.6-30 6 F3.5-6a 16 3.6-42 16 APP 3.6A 6 F3.5-7 21 3.6-43 25 3.6A-1 15 3.6-1 33 3.6-44 25 3.7-1 23 3.6-1a 24 3.6-45 24 3.7-2 6 3.6-1b 18 3.6-46 25 3.7-3 23 3.6-1c 24 3.6-47 25 3.7-4 6 3.6-1d 18 3.6-48 25 3.7-5 6 Page 4 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 3.7-6 16 3.8-14 6 3.8-65 16 3.7-7 6 3.8-15 6 3.8-66 27 3.7-8 6 3.8-16 17 3.8-66a 13 3.7-9 6 3.8-17 6 3.8-66b 13 3.7-10 6 3.8-18 20 3.8-67 6 3.7-11 6 3.8-19 17 3.8-68 29 3.7-12 6 3.8-20 17 3.8-69 27 3.7-13 6 3.8-21 17 3.8-70 13 3.7-14 6 3.8-22 17 3.8-71 6 3.7-15 6 3.8-23 6 3.8-72 23 3.7-16 6 3.8-24 17 3.8-73 23 3.7-17 24 3.8-25 6 3.8-74 27 3.7-18 6 3.8-26 6 3.8-75 32 3.7-19 6 3.8-27 6 3.8-76 16 3.7-20 6 3.8-28 17 3.8-77 25 3.7-21 6 3.8-29 27 3.8-78 17 3.7-22 30 3.8-30 23 3.8-78a 25 3.7-23 6 3.8-31 27 3.8-78b 17 3.7-24 6 3.8-32 6 3.8-79 16 3.7-25 6 3.8-33 6 3.8-79a 16 3.7-26 6 3.8-34 6 3.8-79b 16 3.7-27 6 3.8-35 6 3.8-80 17 3.7-28 6 3.8-36 6 3.8-81 27 3.7-29 6 3.8-37 6 3.8-82 17 3.7-30 16 3.8-38 6 T3.8-1 6 F3.7-1 6 3.8-39 6 T3.8-2 6 F3.7-2 6 3.8-40 6 T3.8-3 6 F3.7-3 6 3.8-41 6 T3.8-4 6 F3.7-4 6 3.8-42 6 T3.8-5 6 F3.7-5 6 3.8-43 6 T3.8-6 6 F3.7-6 6 3.8-44 6 T3.8-7 6 F3.7-7 6 3.8-45 6 T3.8-8 6 F3.7-8 6 3.8-46 6 T3.8-9 6 F3.7-9 6 3.8-47 6 T3.8-10 6 F3.7-10 6 3.8-48 6 T3.8-11 sh1 of 2 6 F3.7-11 6 3.8-49 6 T3.8-11 sh2 of 2 6 F3.7-12 6 3.8-50 25 F3.8-1 6 F3.7-13 6 3.8-51 6 F3.8-2 27 3.8-1 27 3.8-52 6 F3.8-3 6 3.8-2 27 3.8-53 6 F3.8-4 27 3.8-3 29 3.8-54 6 F3.8-5 27 3.8-4 17 3.8-55 13 F3.8-6 27 3.8-5 17 3.8-56 6 F3.8-7 27 3.8-6 6 3.8-57 6 F3.8-8 6 3.8-7 6 3.8-58 18 F3.8-9 27 3.8-8 6 3.8-59 6 F3.8-10 27 3.8-9 6 3.8-60 6 F3.8-11 6 3.8-10 6 3.8-61 6 F3.8-12 6 3.8-11 6 3.8-62 6 F3.8-13 6 3.8-12 6 3.8-63 6 F3.8-14 6 3.8-13 6 3.8-64 27 F3.8-15 6 Page 5 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F3.8-16 6 3.9-10 6 T3.10-1 sh7 of 9 14 F3.8-17 6 3.9-11 6 T3.10-1 sh8 of 9 13 F3.8-18 6 3.9-12 20 T3.10-1 sh9 of 9 25 F3.8-19 6 3.9-13 20 3.11-1 21 F3.8-20 6 3.9-14 33 3.11-2 21 F3.8-21 6 3.9-15 22 3.11-3 6 F3.8-22 6 3.9-16 24 3.11-4 21 F3.8-23 6 3.9-16a 20 3.12-1 15 F3.8-24 6 3.9-16b 25 3.12-2 24 F3.8-25 6 3.9-16c 20 APP 3A 6 F3.8-26 6 3.9-16d 23 3A-1 24 F3.8-27 6 3.9-17 20 3A-2 24 F3.8-28 6 3.9-18 20 3A-3 24 F3.8-29 6 3.9-19 6 3A-4 23 F3.8-30 6 3.9-20 6 3A-5 29 F3.8-31 6 3.9-21 6 3A-6 17 F3.8-32 6 3.9-22 33 3A-7 13 F3.8-33 6 F3.9-1 18 3A-7a R F3.8-34 6 F3.9-2 sh1 of 2 6 3A-7b R F3.8-35 27 F3.9-2 sh2 of 2 6 3A-8 32 F3.8-36 27 APP 3.9A 6 3A-9 16 F3.8-37 17 3.9A-1 6 3A-10 32 F3.8-38 27 3.9A-2 6 3A-11 23 F3.8-39 6 3.9A-3 6 3A-12 31 F3.8-40 6 3.9A-4 6 3A-13 6 F3.8-41 6 3.9A-5 6 3A-14 31 F3.8-42 6 3.9A-6 6 3A-15 6 F3.8-43 6 3.9A-7 6 3A-16 31 F3.8-44 6 3.9A-8 6 3A-17 24 F3.8-45 6 3.9A-9 6 3A-18 6 F3.8-46 6 3.9A-10 6 3A-19 22 F3.8-47 6 3.9A-11 6 3A-20 24 F3.8-48 6 3.9A-12 6 3A-21 24 F3.8-49 6 3.9A-13 6 3A-22 13 F3.8-50 6 3.9A-14 6 3A-23 16 F3.8-50a 13 3.9A-15 6 3A-23a 30 F3.8-51 6 3.9A-16 6 3A-23b 13 F3.8-52 6 3.9A-17 6 3A-24 32 F3.8-53 6 3.9A-18 6 3A-25 6 F3.8-54 6 3.9A-19 6 3A-26 30 F3.8-55 6 3.10-1 6 3A-27 30 F3.8-56 27 3.10-2 6 3A-28 13 3.9-1 23 3.10-3 6 3A-29 23 3.9-2 23 3.10-4 6 3A-30 23 3.9-3 6 3.10-5 6 3A-31 25 3.9-4 6 T3.10-1 sh1 of 9 32 3A-32 13 3.9-5 6 T3.10-1 sh2 of 9 13 3A-32a 13 3.9-6 20 T3.10-1 sh3 of 9 13 3A-32b 26 3.9-7 6 T3.10-1 sh4 of 9 13 3A-33 10 3.9-8 6 T3.10-1 sh5 of 9 13 3A-34 10 3.9-9 6 T3.10-1 sh6 of 9 25 3A-35 25 Page 6 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 3A-36 10 T3B.7-2 sh1 R 4.2-2 18 3A-37 16 T3B.7-2 sh2 R 4.2-3 6 3A-38 6 T3B.7-2 sh3 R 4.2-4 17 3A-39 16 T3B.7-3 sh1 R 4.2-5 25 3A-40 28 T3B.7-3 sh2 R 4.2-6 11 3A-41 23 T3B.7-4 R 4.2-7 21 3A-42 26 3B-21 R 4.2-8 21 3A-42a 24 3B-22 R 4.2-9 28 3A-42b 23 3B-23 R 4.2-10 21 3A-43 6 3B-24 R 4.2-11 28 3A-44 6 3B-25 R 4.2-12 28 3A-45 16 F3B-1 R 4.2-13 31 3A-46 16 F3B-2 R 4.2-13a 18 3A-47 26 F3B-3 R 4.2-13b 18 3A-48 26 F3B-4 R 4.2-13c 28 3A-49 6 F3B-5 R 4.2-13d 28 3A-50 23 F3B-6 R 4.2-14 17 3A-51 16 F3B-7 R 4.2-15 30 3A-51a 16 F3B-8 R 4.2-16 6 3A-51b 23 F3B-9 R 4.2-17 22 3A-52 16 F3B-10 R 4.2-18 22 3A-53 21 F3B-11 R 4.2-19 11 3A-54 24 F3B-12 R 4.2-20 6 3A-55 28 4-i 11 4.2-21 6 T3A-1 28 4-ii 6 4.2-22 6 F3A-1 6 4-iii 17 4.2-23 11 F3A-2 6 4-iv 31 4.2-24 17 F3A-3 16 4-v 18 4.2-25 18 F3A-4 16 4-vi 6 4.2-26 25 3B-1 R 4-vii 17 4.2-27 17 3B-2 R 4-viii 18 4.2-28 22 3B-3 R 4-ix 23 4.2-28a 11 3B-4 R 4-x 19 4.2-28b 11 3B-5 R 4-xi 18 4.2-29 6 3B-6 R 4-xii 17 4.2-30 8 3B-7 R 4-xiii 17 4.2-31 6 3B-8 R 4-xiv 18 4.2-32 19 3B-9 R 4-xv 23 4.2-33 6 3B-10 R 4.1-1 28 4.2-34 6 3B-11 R 4.1-2 18 4.2-35 6 3B-12 R 4.1-3 17 4.2-36 6 3B-13 R 4.1-4 28 4.2-37 6 3B-14 R T4.1-1 sh1 of 5 33 4.2-38 6 3B-15 R T4.1-1 sh2 of 5 33 4.2-39 11 3B-16 R T4.1-1 sh3 of 5 18 4.2-40 11 3B-17 R T4.1-1 sh4 of 5 20 4.2-40a 11 3B-18 R T4.1-1 sh5 of 5 20 4.2-40b 11 3B-19 R T4.1-2 sh1 of 2 31 4.2-41 6 3B-20 R T4.1-2 sh2 of 2 6 4.2-42 6 T3B.7-1 sh1 R T4.1-3 6 4.2-43 6 T3B.7-1 sh2 R 4.2-1 6 4.2-44 18 Page 7 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 4.2-45 11 F4.2-3c 19 4.3-28 19 4.2-46 23 F4.2-4 6 4.3-29 33 4.2-47 18 F4.2-5 17 4.3-30 6 4.2-48 6 F4.2-6 6 4.3-31 17 4.2-49 11 F4.2-7 17 4.3-32 17 4.2-50 25 F4.2-8 6 4.3-33 17 4.2-51 17 F4.2-9 17 4.3-34 17 4.2-52 17 F4.2-10 6 4.3-35 17 4.2-53 19 F4.2-11 17 4.3-36 17 4.2-54 25 F4.2-12 6 4.3-37 6 4.2-55 25 F4.2-13 6 4.3-38 17 4.2-56 25 F4.2-14 17 4.3-39 25 4.2-57 22 F4.2-15 17 4.3-40 17 4.2-58 22 F4.2-16 18 4.3-41 25 4.2-58a 13 F4.2-17 18 4.3-42 17 4.2-58b 13 F4.2-17a 18 4.3-43 17 4.2-59 6 F4.2-18 6 4.3-44 18 4.2-60 6 F4.2-19 18 4.3-45 26 4.2-61 6 F4.2-20 11 4.3-46 17 4.2-62 6 F4.2-21 6 4.3-47 17 4.2-63 6 F4.2-22 6 4.3-48 17 4.2-64 6 F4.2-23 6 4.3-49 17 4.2-65 11 F4.2-24 6 4.3-50 17 4.2-66 25 F4.2-25 6 4.3-51 17 4.2-67 17 4.3-1 6 4.3-52 6 4.2-68 18 4.3-2 17 4.3-53 19 4.2-69 17 4.3-3 33 4.3-54 17 4.2-70 17 4.3-4 25 4.3-55 31 4.2-71 17 4.3-5 20 4.3-56 31 4.2-72 11 4.3-6 6 4.3-57 31 4.2-73 22 4.3-7 25 4.3-58 31 4.2-74 6 4.3-8 6 4.3-59 31 4.2-75 20 4.3-9 27 4.3-60 31 4.2-76 6 4.3-10 23 4.3-61 31 4.2-77 6 4.3-11 25 4.3-62 31 4.2-78 17 4.3-12 23 4.3-63 31 4.2-79 16 4.3-13 17 4.3-64 33 4.2-80 16 4.3-14 6 T4.3-1 sh1 of 3 20 4.2-81 22 4.3-15 19 T4.3-1 sh2 of 3 23 4.2-82 25 4.3-16 17 T4.3-1 sh3 of 3 24 4.2-83 30 4.3-17 17 T4.3-2 sh1 of 2 24 T4.2-1 6 4.3-18 17 T4.3-2 sh2 of 2 24 T4.2-2 18 4.3-19 17 T4.3-3 sh1 24 F4.2-1 18 4.3-20 17 T4.3-4 6 F4.2-2 11 4.3-21 17 T4.3-5 6 F4.2-2a 17 4.3-22 17 T4.3-6 6 F4.2-2b 18 4.3-23 23 T4.3-7 25 F4.2-2c 28 4.3-24 20 T4.3-8 6 F4.2-3 17 4.3-25 19 T4.3-9 6 F4.2-3a 17 4.3-26 19 T4.3-10 6 F4.2-3b 18 4.3-27 19 T4.3-11 6 Page 8 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F4.3-1 17 4.4-14 25 4.4-59 6 F4.3-2 17 4.4-15 29 4.4-60 6 F4.3-3 17 4.4-15a 18 4.4-61 6 F4.3-4 17 4.4-15b 11 4.4-62 6 F4.3-5 17 4.4-16 11 4.4-63 6 F4.3-6 17 4.4-17 18 4.4-64 6 F4.3-7 17 4.4-17a 11 4.4-65 6 F4.3-8 17 4.4-17b 11 4.4-66 11 F4.3-9 17 4.4-18 6 4.4-66a 25 F4.3-10 17 4.4-19 25 4.4-66b 18 F4.3-11 17 4.4-20 18 4.4-66c 29 F4.3-12 17 4.4-21 18 4.4-66d 18 F4.3-13 17 4.4-21a 18 T4.4-1 sh1 of 3 33 F4.3-14 17 4.4-21b 18 T4.4-1 sh2 of 3 33 F4.3-15 17 4.4-22 19 T4.4-1 sh3 of 3 28 F4.3-16 17 4.4-23 19 T4.4-1 sh4 of 4 R F4.3-17 17 4.4-24 18 T4.4-2 sh1 of 2 18 F4.3-18 17 4.4-25 18 T4.4-2 sh2 of 2 R F4.3-19 17 4.4-26 11 T4.4-3 19 F4.3-20 17 4.4-27 11 T4.4-4 6 F4.3-21 17 4.4-28 11 F4.4-1 18 F4.3-22 17 4.4-29 18 F4.4-1a 18 F4.3-23 17 4.4-30 11 F4.4-2 18 F4.3-24 17 4.4-31 33 F4.4-2a 18 F4.3-25 17 4.4-32 6 F4.4-3 6 F4.3-26a 17 4.4-33 6 F4.4-4 6 F4.3-26b 22 4.4-34 33 F4.4-5a 11 F4.3-27 17 4.4-35 18 F4.4-5b 11 F4.3-28 17 4.4-36 6 F4.4-5c 18 F4.3-29 17 4.4-37 18 F4.4-6 6 F4.3-30 17 4.4-38 19 F4.4-7 6 F4.3-31 17 4.4-39 6 F4.4-8 6 F4.3-32 17 4.4-40 6 F4.4-9 6 F4.3-33 17 4.4-41 6 F4.4-10 6 F4.3-34 17 4.4-42 6 F4.4-11 6 F4.3-35 17 4.4-43 19 F4.4-12 6 4.4-1 6 4.4-44 6 F4.4-13 6 4.4-2 19 4.4-45 6 F4.4-14 6 4.4-2a 19 4.4-46 6 F4.4-15 6 4.4-2b 18 4.4-47 6 F4.4-16 21 4.4-3 33 4.4-48 6 F4.4-17 21 4.4-4 33 4.4-49 6 4.5-1 23 4.4-5 19 4.4-50 6 4.5-2 17 4.4-6 18 4.4-51 6 T4.5-1 sh1 23 4.4-7 18 4.4-52 11 T4.5-2 sh1 23 4.4-8 20 4.4-53 18 F4.5-1 23 4.4-9 18 4.4-54 19 F4.5-2 25 4.4-10 18 4.4-55 6 F4.5-3 23 4.4-11 6 4.4-56 6 F4.5-4 25 4.4-12 6 4.4-57 6 5-i 6 4.4-13 19 4.4-58 6 5-ii 6 Page 9 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 5-iii 19 5.2-5 19 5.2-54 19 5-iv 30 5.2-6 19 5.2-55 19 5-v 30 5.2-7 21 5.2-56 19 5-vi 6 5.2-8 33 5.2-56a 19 5-vii 24 5.2-9 6 5.2-56b 19 5-viii 19 5.2-10 6 5.2-57 19 5-ix 24 5.2-11 24 5.2-57a 10 5-x 24 5.2-12 6 5.2-57b 10 5-xa 19 5.2-13 6 5.2-58 22 5-xb 7 5.2-14 24 5.2-59 24 5-xi 27 5.2-15 16 5.2-60 6 5-xii 6 5.2-16 33 5.2-61 6 5-xiii 27 5.2-17 6 5.2-62 24 5-xiv 27 5.2-18 6 5.2-63 16 5.1-1 6 5.2-19 24 5.2-64 33 5.1-2 27 5.2-20 6 5.2-65 33 5.1-2a R 5.2-21 6 5.2-66 33 5.1-2b R 5.2-22 6 5.2-67 6 5.1-3 6 5.2-23 6 5.2-68 32 5.1-4 24 5.2-24 6 5.2-69 32 5.1-5 22 5.2-25 6 5.2-70 31 5.1-6 27 5.2-26 19 5.2-71 16 5.1-7 27 5.2-27 24 5.2-72 16 5.1-8 27 5.2-28 24 5.2-73 24 5.1-9 6 5.2-29 24 5.2-74 24 T5.1-1 sh1 of 1 24 5.2-30 24 5.2-75 33 T5.1-1 sh2 of 2 R 5.2-31 24 T5.2-1 15 F5.1-1 22 5.2-32 24 T5.2-2 28 F5.1-2 6 5.2-33 25 T5.2-3 sh1 of 2 7 F5.1-3 24 5.2-34 25 T5.2-3 sh2 of 2 24 F5.1-3a 18 5.2-35 24 T5.2-4 16 F5.1-4 30 5.2-36 6 T5.2-5 sh1 of 2 24 F5.1-5 6 5.2-37 6 T5.2-5 sh2 of 2 24 F5.1-6a sh1 of 3 27 5.2-38 6 T5.2-5a 19 F5.1-6a sh2 of 3 27 5.2-39 24 T5.2-6 6 F5.1-6a sh3 of 3 27 5.2-40 6 T5.2-7 sh1 of 2 7 F5.1-6b sh1 of 3 27 5.2-41 25 T5.2-7 sh2 of 2 22 F5.1-6b sh2 of 3 27 5.2-42 24 T5.2-8 sh1 of 2 20 F5.1-6b sh3 of 3 27 5.2-43 20 T5.2-8 sh2 of 2 20 F5.1-6c 6 5.2-43a 15 T5.2-9a sh1 of 2 22 F5.1-7 27 5.2-43b 15 T5.2-9a sh2 of 2 18 F5.1-8 27 5.2-44 6 T5.2-9b 24 F5.1-9 27 5.2-45 6 T5.2-10 sh1 of 2 24 F5.1-10 27 5.2-46 6 T5.2-10 sh2 of 2 24 F5.1-11 27 5.2-47 19 T5.2-10a sh1 of 2 22 F5.1-12 6 5.2-48 19 T5.2-10a sh2 of 2 18 F5.1-13 27 5.2-49 7 T5.2-11 sh1 of 2 6 5.2-1 6 5.2-50 6 T5.2-11 sh2 of 2 6 5.2-2 6 5.2-51 20 T5.2-12 25 5.2-3 20 5.2-52 15 T5.2-13 sh1 of 4 24 5.2-4 20 5.2-53 6 T5.2-13 sh2 of 4 22 Page 10 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV T5.2-13 sh3 of 4 6 F5.2-14 6 5.5-6 30 T5.2-13 sh4 of 4 24 F5.2-15 6 5.5-7 31 T5.2-14 19 F5.2-16 6 5.5-8 30 T5.2-15 22 F5.2-17 6 5.5-9 30 T5.2-16 24 F5.2-18 22 5.5-10 30 T5.2-17 24 F5.2-19 7 5.5-11 24 T5.2-18 sh1 of 1 24 F5.2-20 7 5.5-12 24 T5.2-18 sh2 of 2 R F5.2-21 7 5.5-13 24 T5.2-19 24 F5.2-22 18 5.5-14 24 T5.2-20 24 F5.2-23 18 5.5-15 24 T5.2-21 24 F5.2-24 18 5.5-16 24 T5.2-22 sh1 of 1 24 5.3-1 24 5.5-17 24 T5.2-22 sh2 of 2 R 5.3-2 24 5.5-18 24 T5.2-23 sh1 of 1 24 5.4-1 6 5.5-19 24 T5.2-23 sh2 of 2 R 5.4-2 22 5.5-20 24 T5.2-24 24 5.4-3 6 5.5-21 24 T5.2-25 24 5.4-4 6 5.5-22 14 T5.2-26 sh1 of 4 24 5.4-5 19 5.5-23 6 T5.2-26 sh2 of 4 24 5.4-6 19 5.5-24 24 T5.2-26 sh3 of 4 24 5.4-7 19 5.5-25 19 T5.2-26 sh4 of 4 24 5.4-8 19 5.5-26 27 T5.2-27 25 5.4-9 19 5.5-27 6 T5.2-28 26 5.4-10 19 5.5-28 25 T5.2-29 12 5.4-10a R 5.5-29 29 T5.2-30 19 5.4-10b R 5.5-30 11 T5.2-31 19 5.4-10c R 5.5-30a 10 T5.2-32 19 5.4-10d R 5.5-30b 10 T5.2-33 sh1 of 2 18 5.4-11 19 5.5-31 11 T5.2-33 sh2 of 2 18 5.4-12 6 5.5-31a 10 T5.2-34 18 5.4-13 16 5.5-31b 10 T5.2-35 18 5.4-14 19 5.5-32 6 T5.2-36 18 5.4-15 R 5.5-33 17 T5.2-37 19 5.4-16 R 5.5-34 7 T5.2-38 19 T5.4-1 19 5.5-35 6 T5.2-39 19 T5.4-2 6 5.5-36 6 T5.2-40 19 T5.4-3 19 5.5-37 20 T5.2-41 19 T5.4-4 19 5.5-38 14 T5.2-42 19 T5.4-5 19 5.5-39 27 F5.2-1 6 T5.4-6 7 5.5-40 28 F5.2-2 6 T5.4-7 19 5.5-41 27 F5.2-3 6 T5.4-8 19 5.5-42 27 F5.2-4 6 T5.4-9 19 5.5-42a 14 F5.2-5 6 T5.4-10 19 5.5-42b 12 F5.2-6 6 F5.4-1 7 5.5-43 20 F5.2-7 24 F5.4-2 7 5.5-44 20 F5.2-8 24 F5.4-3 7 5.5-45 6 F5.2-9 24 5.5-1 6 5.5-46 6 F5.2-10 24 5.5-2 30 5.5-47 6 F5.2-11 6 5.5-3 28 5.5-48 17 F5.2-12 6 5.5-4 6 5.5-49 6 F5.2-13 19 5.5-5 30 5.5-50 6 Page 11 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 5.5-51 6 6-vi 6 6.2-34 25 5.5-52 6 6-vii 16 6.2-35 24 5.5-53 6 6-viii 27 6.2-36 19 5.5-54 22 6-ix 6 6.2-37 26 5.5-55 17 6-x 6 6.2-38 6 5.5-56 24 6-xi 12 6.2-39 20 5.5-56a R 6-xii 27 6.2-40 18 5.5-56b R 6.1-1 6 6.2-41 6 5.5-57 27 6.1-2 6 6.2-42 6 5.5-58 27 6.1-3 13 6.2-43 6 5.5-59 27 6.1-4 6 6.2-44 16 5.5-59a R 6.1-5 6 6.2-45 6 5.5-60 24 6.1-6 6 6.2-46 25 5.5-61 24 6.1-7 23 6.2-47 6 5.5-62 24 6.1-8 12 6.2-48 6 5.5-63 24 6.1-9 6 6.2-49 6 T5.5-1 sh1 of 4 19 6.1-10 6 6.2-50 15 T5.5-1 sh2 of 4 19 6.1-11 6 6.2-51 6 T5.5-1 sh3 of 4 6 6.2-1 6 6.2-52 23 T5.5-1 sh4 of 4 17 6.2-2 24 6.2-53 23 T5.5-2 sh1 of 2 17 6.2-3 6 6.2-54 6 T5.5-2 sh2 of 2 16 6.2-4 15 6.2-55 16 T5.5-3 20 6.2-5 6 6.2-56 6 F5.5-1 6 6.2-6 6 6.2-57 21 F5.5-2a sh 1 of 2 27 6.2-7 6 6.2-58 16 F5.5-2a sh 2 of 2 27 6.2-8 6 6.2-59 18 F5.5-2b sh 1 of 2 27 6.2-9 6 6.2-59a 18 F5.5-2b sh 2 of 2 27 6.2-10 16 6.2-59b 20 F5.5-3 27 6.2-11 20 6.2-60 19 F5.5-4 27 6.2-12 20 6.2-61 16 F5.5-5 6 6.2-13 23 6.2-62 20 F5.5-6 27 6.2-14 27 6.2-63 6 F5.5-7 27 6.2-15 6 6.2-64 25 5.6-1 27 6.2-16 16 6.2-65 23 5.6-1a 16 6.2-17 27 6.2-66 16 5.6-1b 10 6.2-18 27 6.2-67 26 5.6-2 10 6.2-19 27 6.2-68 6 5.6-3 16 6.2-20 9 6.2-69 6 5.6-4 17 6.2-21 23 6.2-70 6 5.6-5 22 6.2-22 6 6.2-71 33 5.6-6 22 6.2-23 23 6.2-72 6 5.6-6a 13 6.2-24 23 6.2-73 6 5.6-6b 13 6.2-25 6 6.2-74 23 5.6-7 18 6.2-26 6 6.2-75 6 5.6-8 6 6.2-27 18 6.2-76 13 5.6-9 6 6.2-28 24 6.2-76a 13 6-i 6 6.2-29 26 6.2-76b 13 6-ii 6 6.2-30 23 6.2-77 23 6-iii 23 6.2-31 27 6.2-78 25 6-iv 20 6.2-32 18 6.2-79 31 6-v 26 6.2-33 18 6.2-80 31 Page 12 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 6.2-81 23 F6.2-12 27 F6.2-58 23 6.2-82 23 F6.2-13 27 6.3-1 6 6.2-83 R F6.2-14 6 6.3-2 6 6.2-84 R F6.2-15 6 6.3-3 6 6.2-85 R F6.2-16 16 6.3-4 27 T6.2-1 6 F6.2-17 6 6.3-5 6 T6.2-2 6 F6.2-18 16 6.3-6 16 T6.2-3 6 F6.2-19 17 6.3-7 20 T6.2-4 sh1 of 2 6 F6.2-20 6 6.3-8 24 T6.2-4 sh2 of 2 6 F6.2-21 16 6.3-9 24 T6.2-5 6 F6.2-22 20 6.3-10 24 T6.2-6 sh1 24 F6.2-23 16 6.3-11 6 T6.2-7 18 F6.2-24 16 6.3-12 20 T6.2-8 6 F6.2-25 20 6.3-13 21 T6.2-9 sh1 23 F6.2-26 20 6.3-14 21 T6.2-10 sh1 of 12 25 F6.2-27 20 6.3-15 19 T6.2-10 sh2 of 12 R F6.2-28 6 6.3-16 30 T6.2-10 sh3 of 12 R F6.2-29 17 6.3-17 18 T6.2-10 sh4 of 12 R F6.2-30 17 6.3-18 6 T6.2-10 sh5 of 12 R F6.2-31 16 6.3-19 18 T6.2-10 sh6 of 12 R F6.2-32 16 6.3-20 26 T6.2-10 sh7 of 12 R F6.2-33 20 6.3-21 17 T6.2-10 sh8 of 12 R F6.2-34 25 6.3-22 6 T6.2-10 sh9 of 12 R F6.2-35 18 6.3-23 6 T6.2-10 sh10 of 12 R F6.2-36 19 6.3-24 6 T6.2-10 sh11 of 12 R F6.2-37 24 6.3-25 6 T6.2-10 sh12 of 12 R F6.2-38 23 6.3-26 19 T6.2-11 16 F6.2-39 24 6.3-27 6 T6.2-12 sh1 of 3 26 F6.2-40 16 6.3-28 6 T6.2-12 sh2 of 3 R F6.2-41 16 6.3-29 18 T6.2-12 sh3 of 3 R F6.2-42 16 6.3-30 27 T6.2-13 sh1 of 4 26 F6.2-43 16 6.3-31 24 T6.2-13 sh2 of 4 R F6.2-44 20 6.3-32 24 T6.2-13 sh3 of 4 R F6.2-45 12 6.3-33 13 T6.2-13 sh4 of 4 R F6.2-45a 23 6.3-33a 13 T6.2-14 6 F6.2-45b 23 6.3-33b 13 T6.2-15 sh1 23 F6.2-45c 16 6.3-34 24 T6.2-16 6 F6.2-45d 16 6.3-35 17 T6.2-17 6 F6.2-45e 18 6.3-36 17 F6.2-1 6 F6.2-46 6 6.3-36a 24 F6.2-2 6 F6.2-47 6 6.3-36b 24 F6.2-3 6 F6.2-48 6 6.3-36c 24 F6.2-4a 27 F6.2-49 6 6.3-36d 16 F6.2-4b 27 F6.2-50 6 6.3-37 27 F6.2-5 27 F6.2-51 6 6.3-38 19 F6.2-6 27 F6.2-52 6 6.3-39 25 F6.2-7 27 F6.2-53 6 6.3-39a 7 F6.2-8 27 F6.2-54 6 6.3-39b 7 F6.2-9 27 F6.2-55 6 6.3-40 19 F6.2-10 27 F6.2-56 23 6.3-41 26 F6.2-11 27 F6.2-57 23 6.3-42 16 Page 13 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 6.3-43 6 F6.3-1b sh 1 of 4 27 7.2-6 6 6.3-44 16 F6.3-1b sh 2 of 4 27 7.2-7 25 6.3-45 6 F6.3-1b sh 3 of 4 27 7.2-8 6 6.3-46 6 F6.3-1b sh 4 of 4 27 7.2-9 27 6.3-47 7 F6.3-2 27 7.2-10 6 6.3-48 29 F6.3-3 23 7.2-11 6 6.3-49 16 F6.3-4 6 7.2-12 16 6.3-50 21 6.4-1 27 7.2-13 11 6.3-51 7 6.4-1a 27 7.2-13a 11 6.3-52 6 6.4-1b 26 7.2-13b 30 6.3-53 16 6.4-1c 27 7.2-14 30 6.3-54 6 6.4-1d 23 7.2-15 11 6.3-55 6 6.4-2 26 7.2-16 6 6.3-56 9 6.4-3 33 7.2-17 6 6.3-56a 9 6.4-4 33 7.2-18 6 6.3-56b 9 6.4-4a R 7.2-19 6 6.3-57 6 6.4-4b R 7.2-20 6 6.3-58 19 6.4-5 R 7.2-21 26 6.3-59 6 T6.4-1 15 7.2-22 25 6.3-60 9 T6.4-2 16 7.2-23 14 6.3-61 9 T6.4-3 sh1 26 7.2-24 16 6.3-62 16 7-i 24 7.2-25 18 6.3-63 24 7-ii 15 7.2-26 6 6.3-64 24 7-iii 13 7.2-27 31 T6.3-1 21 7-iv 30 7.2-28 31 T6.3-2 19 7-v 23 7.2-29 11 T6.3-3 27 7-va 21 7.2-30 31 T6.3-4 16 7-vb 15 7.2-31 6 T6.3-5 27 7-vi 23 7.2-32 25 T6.3-6 sh1 of 4 25 7-vii 31 7.2-33 6 T6.3-6 sh2 of 4 25 7-viii 27 7.2-34 6 T6.3-6 sh3 of 4 25 7-ix 17 7.2-35 25 T6.3-6 sh4 of 4 25 7.1-1 6 7.2-36 6 T6.3-7 sh1 25 7.1-2 6 7.2-37 23 T6.3-8 16 7.1-3 6 7.2-38 30 T6.3-9 sh1 of 3 6 7.1-4 6 7.2-39 15 T6.3-9 sh2 of 3 7 7.1-5 6 T7.2-1 sh1 of 7 6 T6.3-9 sh3 of 3 7 7.1-6 6 T7.2-1 sh2 of 7 16 T6.3-10 6 7.1-7 6 T7.2-1 sh3 of 7 25 T6.3-11 sh1 of 2 15 7.1-8 6 T7.2-1 sh4 of 7 17 T6.3-11 sh2 of 2 15 7.1-9 6 T7.2-1 sh5 of 7 17 T6.3-12 26 7.1-10 6 T7.2-1 sh6 of 7 17 T6.3-13 sh1 of 2 19 7.1-11 24 T7.2-1 sh7 of 7 26 T6.3-13 sh2 of 2 26 7.1-12 24 T7.2-2 sh1 of 4 21 T6.3-14 sh1 of 3 7 7.1-13 24 T7.2-2 sh1a of 4 9 T6.3-14 sh2 of 3 6 7.1-14 16 T7.2-2 sh2 of 4 25 T6.3-14 sh3 of 3 6 7.2-1 6 T7.2-2 sh3 of 4 31 F6.3-1a sh 1 of 4 27 7.2-2 6 T7.2-2 sh4 of 4 21 F6.3-1a sh 2 of 4 27 7.2-3 27 T7.2-3 sh1 of 2 6 F6.3-1a sh 3 of 4 27 7.2-4 6 T7.2-3 sh2 of 2 6 F6.3-1a sh 4 of 4 27 7.2-5 6 T7.2-4 sh1 25 Page 14 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV T7.2-4 sh2 R 7.3-36 13 T7.3-4 sh45 of 47 16 T7.2-5 sh1 25 T7.3-1 sh1 of 2 17 T7.3-4 sh46 of 47 16 T7.2-5 sh2 R T7.3-1sh2 of 2 21 T7.3-4 sh47 of 47 16 F7.2-1 6 T7.3-2 sh1 24 T7.3-5 sh1 of 15 16 F7.2-2 27 T7.3-3 sh1 of 2 16 T7.3-5 sh2 of 15 17 F7.2-3 6 T7.3-3 sh2 of 2 16 T7.3-5 sh3 of 15 16 F7.2-4 17 T7.3-4 sh1 of 47 16 T7.3-5 sh4 of 15 17 F7.2-5 16 T7.3-4 sh2 of 47 16 T7.3-5 sh5 of 15 16 F7.2-6 6 T7.3-4 sh3 of 47 16 T7.3-5 sh6 of 15 17 F7.2-7 17 T7.3-4 sh4 of 47 16 T7.3-5 sh7 of 15 16 7.3-1 6 T7.3-4 sh5 of 47 16 T7.3-5 sh8 of 15 25 7.3-2 6 T7.3-4 sh6 of 47 16 T7.3-5 sh9 of 15 16 7.3-3 6 T7.3-4 sh7 of 47 16 T7.3-5 sh10 of 15 17 7.3-4 16 T7.3-4 sh8 of 47 16 T7.3-5 sh11 of 15 6 7.3-5 16 T7.3-4 sh9 of 47 16 T7.3-5 sh12 of 15 17 7.3-6 6 T7.3-4 sh10 of 47 16 T7.3-5 sh13 of 15 16 7.3-7 16 T7.3-4 sh11 of 47 14 T7.3-5 sh14 of 15 17 7.3-8 23 T7.3-4 sh12 of 47 16 T7.3-5 sh15 of 15 16 7.3-9 6 T7.3-4 sh13 of 47 16 T7.3-6 sh1 of 14 16 7.3-10 13 T7.3-4 sh14 of 47 23 T7.3-6 sh2 of 14 16 7.3-10a 13 T7.3-4 sh15 of 47 16 T7.3-6 sh3 of 14 16 7.3-10b 13 T7.3-4 sh16 of 47 16 T7.3-6 sh4 of 14 16 7.3-11 6 T7.3-4 sh17 of 47 16 T7.3-6 sh5 of 14 16 7.3-12 23 T7.3-4 sh18 of 47 14 T7.3-6 sh6 of 14 16 7.3-13 11 T7.3-4 sh19 of 47 14 T7.3-6 sh7 of 14 16 7.3-14 6 T7.3-4 sh20 of 47 14 T7.3-6 sh8 of 14 16 7.3-15 8 T7.3-4 sh21 of 47 23 T7.3-6 sh9 of 14 16 7.3-16 15 T7.3-4 sh22 of 47 23 T7.3-6 sh10 of 14 16 7.3-16a 8 T7.3-4 sh22a of 47 23 T7.3-6 sh11 of 14 6 7.3-16b 8 T7.3-4 sh23 of 47 16 T7.3-6 sh12 of 14 6 7.3-17 6 T7.3-4 sh24 of 47 16 T7.3-6 sh13 of 14 16 7.3-18 6 T7.3-4 sh25 of 47 16 T7.3-6 sh14 of 14 16 7.3-19 6 T7.3-4 sh26 of 47 16 T7.3-7 sh1 of 2 16 7.3-20 16 T7.3-4 sh27 of 47 16 T7.3-7 sh2 of 2 25 7.3-21 30 T7.3-4 sh28 of 47 16 T7.3-8 sh1 25 7.3-22 6 T7.3-4 sh29 of 47 16 T7.3-8 sh2 R 7.3-23 20 T7.3-4 sh30 of 47 16 T7.3-8 sh3 R 7.3-24 6 T7.3-4 sh31 of 47 16 T7.3-8 sh4 R 7.3-25 24 T7.3-4 sh32 of 47 16 T7.3-8 sh5 R 7.3-26 30 T7.3-4 sh33 of 47 16 T7.3-8 sh6 R 7.3-27 16 T7.3-4 sh34 of 47 16 T7.3-9 sh1 25 7.3-28 25 T7.3-4 sh35 of 47 16 T7.3-9 sh2 R 7.3-29 17 T7.3-4 sh36 of 47 16 T7.3-9 sh3 R 7.3-30 6 T7.3-4 sh37 of 47 16 T7.3-9 sh4 R 7.3-31 13 T7.3-4 sh38 of 47 16 T7.3-9 sh5 R 7.3-31a 13 T7.3-4 sh39 of 47 16 T7.3-9 sh6 R 7.3-31b 13 T7.3-4 sh40 of 47 16 F7.3-1 6 7.3-32 18 T7.3-4 sh41 of 47 16 F7.3-2 6 7.3-33 18 T7.3-4 sh42 of 47 16 F7.3-3 13 7.3-34 6 T7.3-4 sh43 of 47 16 F7.3-4 20 7.3-35 12 T7.3-4 sh44 of 47 16 7.4-1 6 Page 15 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 7.4-2 6 7.6-5 19 F7.7-2 16 7.4-3 6 7.6-6 24 F7.7-3 6 7.4-4 6 7.6-7 29 F7.7-4 21 7.4-5 6 7.6-8 24 F7.7-5 6 7.5-1 16 F7.6-1 6 F7.7-6 6 7.5-2 6 7.7-1 31 F7.7-7 29 7.5-3 6 7.7-2 31 F7.7-8 29 7.5-4 17 7.7-3 31 F7.7-9 29 7.5-5 11 7.7-4 6 7.8-1 21 7.5-6 6 7.7-5 31 7.8-2 21 7.5-7 6 7.7-6 18 7.8-3 10 7.5-8 15 7.7-7 32 7.8-4 24 7.5-9 6 7.7-8 21 7.8-5 21 7.5-10 6 7.7-9 21 7.8-6 21 7.5-11 14 7.7-9a 29 7.8-7 21 7.5-12 6 7.7-9b 29 7.8-8 10 7.5-13 20 7.7-10 29 7.8-9 25 7.5-14 11 7.7-11 21 7.8-10 10 7.5-15 30 7.7-12 31 7.8-11 25 T7.5-1 sh1 of 4 18 7.7-13 31 7.8-12 25 T7.5-1 sh2 of 4 17 7.7-14 23 7.8-13 10 T7.5-1 sh3 of 4 18 7.7-15 23 7.9-1 11 T7.5-1 sh4 of 4 23 7.7-16 23 7.9-2 11 T7.5-2 sh1 of 11 6 7.7-17 25 7.9-3 11 T7.5-2 sh2 of 11 6 7.7-17a 18 7.9-4 11 T7.5-2 sh3 of 11 25 7.7-17b 18 7.9-5 11 T7.5-2 sh4 of 11 15 7.7-18 18 7.10-1 33 T7.5-2 sh5 of 11 18 7.7-19 18 7.10-2 33 T7.5-2 sh6 of 11 6 7.7-20 18 7.10-3 19 T7.5-2 sh7 of 11 6 7.7-21 27 7.10-4 19 T7.5-2 sh8 of 11 16 7.7-21a 31 7.10-5 14 T7.5-2 sh9 of 11 17 7.7-21b 14 T7.10-1 sh1 of 2 14 T7.5-2 sh10 of 11 25 7.7-22 6 T7.10-1 sh2 of 2 14 T7.5-2 sh11 of 11 25 7.7-23 11 T7.10-2 sh1 of 2 14 T7.5-3 30 7.7-24 6 T7.10-2 sh2 of 2 14 T7.5-4 sh1 of 5 30 7.7-25 21 T7.10-3 14 T7.5-4 sh2 of 5 31 7.7-26 15 7A-1 23 T7.5-4 sh3 of 5 30 7.7-26a 11 7A-2 15 T7.5-4 sh4 of 5 16 7.7-26b 11 7A-3 22 T7.5-4 sh5 of 5 30 7.7-27 6 7A-4 29 T7.5-5 sh1 of 3 20 7.7-28 16 7A-5 30 T7.5-5 sh2 of 3 25 7.7-29 18 7A-6 27 T7.5-5 sh3 of 3 20 7.7-30 18 7A-7 27 7.6-1 11 T7.7-1 21 8-i 26 7.6-1a 11 T7.7-2 sh1 of 2 17 8-ii 14 7.6-1b 10 T7.7-2 sh1 of 2 17 8-iii 32 7.6-2 11 T7.7-3 31 8-iv 27 7.6-2a 10 T7.7-4 31 8.1-1 25 7.6-2b 10 T7.7-5 31 8.1-2 27 7.6-3 6 T7.7-6 31 8.1-3 16 7.6-4 6 F7.7-1 31 8.1-4 24 Page 16 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 8.1-5 29 F8.3-1 22 9.1-11 24 8.1-5a 10 F8.3-2 16 9.1-12 18 8.1-5b 10 F8.3-2a 15 9.1-13 25 8.1-6 6 F8.3-3a 27 9.1-14 25 8.1-7 18 F8.3-3b 27 9.1-15 18 8.1-8 16 F8.3-4 27 9.1-16 24 8.1-9 16 F8.3-4a 27 9.1-17 25 8.1-10 18 F8.3-4b 27 9.1-18 22 8.1-11 26 F8.3-4c 27 9.1-19 22 8.1-11a 26 F8.3-5 sh 1 of 2 27 9.1-20 23 8.1-11b 16 F8.3-5 sh 2 of 2 27 9.1-20a 23 8.1-12 33 F8.3-6 sh 1 of 2 27 9.1-20b 12 8.2-1 27 F8.3-6 sh 2 of 2 27 9.1-21 22 8.2-2 6 F8.3-7 sh 1 of 2 27 9.1-22 22 8.2-3 25 F8.3-7 sh 2 of 2 27 9.1-23 25 8.2-4 15 F8.3-8 sh 1 of 2 27 9.1-24 25 F8.2-1 27 F8.3-8 sh 2 of 2 27 9.1-25 22 F8.2-2 25 9-i 32 9.1-26 22 8.3-1 26 9-ii 6 9.1-27 21 8.3-2 27 9-iii 33 9.1-28 6 8.3-3 15 9-iv 32 9.1-29 24 8.3-4 32 9-v 16 9.1-30 25 8.3-5 22 9-vi 24 9.1-31 25 8.3-5a 33 9-vii 21 9.1-32 25 8.3-5b 27 9-viii 13 9.1-33 6 8.3-6 27 9-ix 13 9.1-34 16 8.3-7 30 9-x 6 9.1-35 25 8.3-8 14 9-xi 27 9.1-36 25 8.3-8a 25 9-xia 27 9.1-37 25 8.3-8b 12 9-xib 15 9.1-38 25 8.3-9 25 9-xii 27 9.1-39 25 8.3-10 6 9-xiii 27 9.1-40 25 8.3-11 6 9-xiv 27 9.1-41 25 8.3-12 14 9.1-1 27 T9.1-1 6 8.3-12a 13 9.1-2 32 T9.1-2 sh1 of 3 18 8.3-12b 13 9.1-2a 32 T9.1-2 sh2 of 3 19 8.3-13 16 9.1-2b 32 T9.1-2 sh3 of 3 12 8.3-14 18 9.1-3 25 T9.1-3 6 8.3-15 6 9.1-3a 15 T9.1-4 sh1 of 6 23 8.3-16 27 9.1-3b 22 T9.1-4 sh2 of 6 31 8.3-17 32 9.1-4 25 T9.1-4 sh3 of 6 23 8.3-18 14 9.1-4a 15 T9.1-4 sh4 of 6 31 8.3-19 16 9.1-4b 25 T9.1-4 sh5 of 6 31 8.3-20 33 9.1-5 27 T9.1-4 sh6 of 6 31 T8.3-1 11 9.1-6 32 F9.1-1 27 T8.3-2 sh1 23 9.1-7 27 F9.1-2a 15 T8.3-3 22 9.1-8 15 F9.1-2b 15 T8.3-4a 14 9.1-9 15 F9.1-3 15 T8.3-4b 14 9.1-10 18 F9.1-3a 15 T8.3-5 19 9.1-10a 33 F9.1-3b 15 T8.3-6 32 9.1-10b 33 F9.1-3c 15 Page 17 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F9.1-3d 15 9.2-34 33 9.3-10 6 F9.1-3e 15 T9.2-1 16 9.3-11 6 F9.1-3f 15 T9.2-2 6 9.3-12 6 F9.1-4a 27 T9.2-3 27 9.3-13 6 F9.1-4b 27 T9.2-4 sh1 of 2 26 9.3-14 27 F9.1-5 6 T9.2-4 sh2 of 2 26 9.3-15 18 F9.1-6 22 T9.2-5 6 9.3-15a 13 F9.1-6a 22 F9.2-1a sh1 of 7 27 9.3-15b 13 F9.1-7 22 F9.2-1a sh2 of 7 27 9.3-16 20 F9.1-7a 22 F9.2-1a sh3 of 7 27 9.3-17 6 F9.1-8 22 F9.2-1a sh4 of 7 27 9.3-18 6 F9.1-9 6 F9.2-1a sh5 of 7 27 9.3-19 27 9.2-1 27 F9.2-1a sh6 of 7 27 9.3-20 27 9.2-2 18 F9.2-1a sh7 of 7 27 9.3-21 19 9.2-3 16 F9.2-1b sh1 of 6 27 9.3-22 21 9.2-4 24 F9.2-1b sh2 of 6 27 9.3-23 21 9.2-5 27 F9.2-1b sh3 of 6 27 9.3-24 21 9.2-6 30 F9.2-1b sh4 of 6 27 9.3-25 20 9.2-6a 33 F9.2-1b sh5 of 6 27 9.3-26 6 9.2-6b 28 F9.2-1b sh6 of 6 27 9.3-27 6 9.2-7 18 F9.2-2a 19 9.3-28 13 9.2-8 31 F9.2-2b 27 9.3-29 14 9.2-8a 24 F9.2-3 27 9.3-30 21 9.2-8b 18 F9.2-4a sh1 of 3 27 9.3-31 21 9.2-9 31 F9.2-4a sh2 of 3 27 9.3-32 20 9.2-9a 18 F9.2-4a sh3 of 3 27 9.3-33 21 9.2-9b 27 F9.2-4b sh1 of 3 27 9.3-34 25 9.2-10 18 F9.2-4b sh2 of 3 27 9.3-35 21 9.2-11 27 F9.2-4b sh3 of 3 27 9.3-36 18 9.2-12 6 F9.2-5 6 9.3-37 14 9.2-13 6 F9.2-6 6 9.3-38 29 9.2-14 16 F9.2-7 6 9.3-39 25 9.2-15 6 F9.2-8 6 9.3-39a 14 9.2-16 6 F9.2-9 6 9.3-39b 12 9.2-17 14 F9.2-10 6 9.3-40 14 9.2-18 6 F9.2-11 sh1 of 7 15 9.3-41 20 9.2-19 13 F9.2-11 sh2 of 7 15 9.3-42 6 9.2-20 6 F9.2-11 sh3 of 7 15 9.3-43 6 9.2-21 6 F9.2-11 sh4 of 7 15 9.3-44 27 9.2-22 16 F9.2-11 sh5 of 7 27 9.3-45 26 9.2-23 6 F9.2-11 sh6 of 7 27 9.3-46 25 9.2-24 6 F9.2-11 sh7 of 7 27 9.3-47 14 9.2-25 6 9.3-1 27 9.3-48 16 9.2-26 6 9.3-2 27 9.3-49 6 9.2-27 25 9.3-3 27 9.3-50 6 9.2-28 28 9.3-4 20 9.3-51 19 9.2-29 23 9.3-5 20 9.3-52 6 9.2-30 32 9.3-6 27 9.3-53 6 9.2-31 32 9.3-7 6 9.3-54 21 9.2-32 32 9.3-8 6 9.3-55 20 9.2-33 32 9.3-9 17 9.3-56 6 Page 18 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 9.3-57 27 F9.3-4b sh3 of 3 27 9.4-19 22 9.3-58 23 F9.3-5a sh1 of 2 27 9.4-20 23 9.3-59 23 F9.3-5a sh2 of 2 27 9.4-21 27 9.3-60 20 F9.3-5b sh1 of 2 27 9.4-22 22 9.3-61 16 F9.3-5b sh2 of 2 27 9.4-23 6 9.3-62 6 F9.3-6a sh1 of 3 27 9.4-24 23 9.3-63 16 F9.3-6a sh2 of 3 27 9.4-25 24 T9.3-1 8 F9.3-6a sh3 of 3 27 9.4-26 24 T9.3-2 6 F9.3-6b sh1 of 3 27 9.4-27 22 T9.3-3 6 F9.3-6b sh2 of 3 27 9.4-28 22 T9.3-4 16 F9.3-6b sh3 of 3 27 9.4-29 24 T9.3-5 22 F9.3-7a sh1 of 3 27 9.4-30 24 T9.3-6 sh1 of 12 29 F9.3-7a sh2 of 3 27 9.4-31 6 T9.3-6 sh2 of 12 23 F9.3-7a sh3 of 3 27 9.4-32 12 T9.3-6 sh3 of 12 23 F9.3-7b sh1 of 3 27 9.4-33 27 T9.3-6 sh4 of 12 14 F9.3-7b sh2 of 3 27 9.4-34 22 T9.3-6 sh5 of 12 6 F9.3-7b sh3 of 3 27 9.4-35 16 T9.3-6 sh6 of 12 14 F9.3-8a 27 9.4-36 25 T9.3-6 sh7 of 12 21 F9.3-8b 27 9.4-37 25 T9.3-6 sh8 of 12 14 F9.3-9 6 9.4-38 16 T9.3-6 sh9 of 12 26 F9.3-10 6 9.4-39 27 T9.3-6 sh10 of 12 6 9.4-1 28 9.4-39a 14 T9.3-6 sh11 of 12 6 9.4-1a 27 9.4-39b 14 T9.3-6 sh12 of 12 23 9.4-1b 31 9.4-40 14 T9.3-7 6 9.4-2 16 9.4-41 25 F9.3-1a sh1 of 4 27 9.4-3 24 9.4-42 27 F9.3-1a sh2 of 4 27 9.4-4 19 9.4-43 16 F9.3-1a sh3 of 4 27 9.4-4a 19 9.4-44 24 F9.3-1a sh4 of 4 27 9.4-4b 17 9.4-45 24 F9.3-1b sh1 of 2 27 9.4-4c 16 9.4-46 17 F9.3-1b sh2 of 2 27 9.4-4d 19 9.4-47 33 F9.3-2a sh1 of 3 27 9.4-4e 16 F9.4-1a 27 F9.3-2a sh2 of 3 27 9.4-4f 17 F9.4-1b 27 F9.3-2a sh3 of 3 27 9.4-5 23 F9.4-2a sh1 of 3 27 F9.3-2b sh1 of 3 27 9.4-6 23 F9.4-2a sh2 of 3 27 F9.3-2b sh2 of 3 27 9.4-7 22 F9.4-2a sh3 of 3 27 F9.3-2b sh3 of 3 27 9.4-8 23 F9.4-2b sh1 of 3 27 F9.3-2c 27 9.4-8a 23 F9.4-2b sh2 of 3 27 F9.3-3a sh1 of 4 27 9.4-8b 16 F9.4-2b sh3 of 3 27 F9.3-3a sh2 of 4 27 9.4-9 27 F9.4-3a sh1 27 F9.3-3a sh3 of 4 27 9.4-10 16 F9.4-3b 27 F9.3-3a sh4 of 4 27 9.4-11 23 F9.4-4a sh1 of 3 27 F9.3-3b sh1 of 4 27 9.4-12 33 F9.4-4a sh2 of 3 27 F9.3-3b sh2 of 4 27 9.4-13 23 F9.4-4a sh3 of 3 27 F9.3-3b sh3 of 4 27 9.4-14 33 F9.4-4b sh1 of 3 27 F9.3-3b sh4 of 4 27 9.4-15 31 F9.4-4b sh2 of 3 27 F9.3-4a sh1 of 3 27 9.4-15a R F9.4-4b sh3 of 3 27 F9.3-4a sh2 of 3 27 9.4-15b R F9.4-5a 27 F9.3-4a sh3 of 3 27 9.4-16 20 F9.4-5b 27 F9.3-4b sh1 of 3 27 9.4-17 27 F9.4-6a 27 F9.3-4b sh2 of 3 27 9.4-18 22 F9.4-6b 27 Page 19 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 9.5-1 21 10-ii 16 10.3-14 28 9.5-2 29 10-iii 30 10.3-15 19 9.5-3 29 10-iv 18 10.3-16 19 9.5-4 29 10-v 27 10.3-17 18 9.5-5 29 10-vi 27 10.3-18 29 9.5-6 13 10-vii 27 10.3-19 29 9.5-7 22 10.1-1 6 T10.3-1 12 9.5-8 33 10.1-2 6 T10.3-2 12 9.5-9 13 10.2-1 26 F10.3-1a sh1 of 6 27 9.5-10 21 10.2-2 27 F10.3-1a sh2 of 6 27 9.5-11 21 10.2-3 27 F10.3-1a sh3 of 6 27 9.5-12 21 10.2-4 23 F10.3-1a sh4 of 6 27 9.5-13 21 10.2-5 32 F10.3-1a sh5 of 6 27 9.5-14 21 10.2-6 26 F10.3-1a sh6 of 6 27 9.5-15 32 10.2-6a 31 F10.3-1b sh1 of 6 27 9.5-16 21 10.2-6b 31 F10.3-1b sh2 of 6 27 9.5-17 21 10.2-7 31 F10.3-1b sh3 of 6 27 9.5-18 21 10.2-8 31 F10.3-1b sh4 of 6 27 9.5-19 33 10.2-9 6 F10.3-1b sh5 of 6 27 9.5-20 21 10.2-10 6 F10.3-1b sh6 of 6 27 9.5-21 21 10.2-11 27 F10.3-2 15 9.5-22 33 10.2-12 27 F10.3-3 sh1 of 2 27 9.5-23 33 10.2-13 6 F10.3-3 sh2 of 2 27 9.5-24 21 F10.2-1 sh1 of 3 27 10.4-1 15 9.5-25 33 F10.2-1 sh2 of 3 27 10.4-2 27 9.5-26 21 F10.2-1 sh3 of 3 27 10.4-3 20 9.5-27 24 F10.2-2a sh1 of 3 27 10.4-4 27 9.5-28 29 F10.2-2a sh2 of 3 27 10.4-5 16 9.5-29 13 F10.2-2a sh3 of 3 27 10.4-6 23 9.5-30 13 F10.2-2b sh1 of 3 27 10.4-7 27 9.5-31 29 F10.2-2b sh2 of 3 27 10.4-8 16 9.5-32 21 F10.2-2b sh3 of 3 27 10.4-9 27 9.5-33 13 F10.2-3 sh1 of 2 27 10.4-10 28 9.5-34 24 F10.2-3 sh2 of 2 27 10.4-11 28 9.5-35 13 F10.2-4 sh1 of 4 27 10.4-12 24 9.5-36 14 F10.2-4 sh2 of 4 27 10.4-13 27 9.5-37 16 F10.2-4 sh3 of 4 27 10.4-14 24 9.5-38 17 F10.2-4 sh4 of 4 27 10.4-15 18 9.5-39 28 10.3-1 16 10.4-15a 24 9.5-40 32 10.3-2 27 10.4-15b 24 9.5-41 16 10.3-3 16 10.4-16 27 9.5-42 23 10.3-4 24 10.4-17 26 9.5-43 22 10.3-5 16 10.4-18 17 9.5-44 23 10.3-6 15 10.4-19 28 9.5-45 16 10.3-7 22 10.4-20 21 9.5-46 16 10.3-8 16 10.4-21 21 F9.5-1 13 10.3-9 18 10.4-22 24 F9.5-2 sh1 of 3 27 10.3-10 27 10.4-23 24 F9.5-2 sh2 of 3 27 10.3-11 27 10.4-24 24 F9.5-2 sh3 of 3 27 10.3-12 29 10.4-25 30 10-i 16 10.3-13 19 10.4-26 30 Page 20 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 10.4-27 30 F10.4-15 6 11.2-6 14 T10.4-1 sh1 of 4 26 F10.4-16 6 11.2-7 26 T10.4-1 sh2 of 4 30 F10.4-17a 27 11.2-8 26 T10.4-1 sh3 of 4 30 F10.4-17b 27 11.2-9 26 T10.4-1 sh4 of 4 28 F10.4-18a sh1 of 3 27 11.2-10 26 T10.4-2 33 F10.4-18a sh2 of 3 27 11.2-10a 18 T10.4-3 18 F10.4-18a sh3 of 3 27 11.2-10b 13 T10.4-4 18 F10.4-18b sh1 of 3 27 11.2-11 26 T10.4-5 sh1 of 2 18 F10.4-18b sh2 of 3 27 11.2-12 27 T10.4-5 sh2 of 2 R F10.4-18b sh3 of 3 27 11.2-13 14 T10.4-6 18 F10.4-19 18 11.2-14 6 F10.4-1 sh1 of 2 27 11-i 16 11.2-15 6 F10.4-1 sh2 of 2 27 11-ii 23 11.2-16 6 F10.4-2 27 11-iii 6 11.2-17 18 F10.4-3a sh1 of 5 27 11-iv 16 T11.2-1 19 F10.4-3a sh2 of 5 27 11-v 23 T11.2-2 14 F10.4-3a sh3 of 5 27 11-vi 15 T11.2-3 sh1 of 2 19 F10.4-3a sh4 of 5 27 11-vii 27 T11.2-3 sh2 of 2 19 F10.4-3a sh5 of 5 27 11.1-1 28 T11.2-4 sh1 of 2 14 F10.4-3b sh1 of 4 27 11.1-2 23 T11.2-4 sh2 of 2 14 F10.4-3b sh2 of 4 27 11.1-3 20 T11.2-5 14 F10.4-3b sh3 of 4 27 11.1-4 16 T11.2-6 sh1 of 2 6 F10.4-3b sh4 of 4 27 11.1-5 16 T11.2-6 sh2 of 2 6 F10.4-4 27 11.1-6 16 T11.2-7 sh1 of 2 6 F10.4-5a sh1 of 3 27 11.1-7 6 T11.2-7 sh2 of 2 6 F10.4-5a sh2 of 3 27 11.1-8 24 T11.2-8 sh1 of 3 6 F10.4-5a sh3 of 3 27 11.1-9 6 T11.2-8 sh2 of 3 6 F10.4-5b sh1 of 3 27 11.1-10 16 T11.2-8 sh3 of 3 6 F10.4-5b sh2 of 3 27 11.1-11 28 T11.2-9 6 F10.4-5b sh3 of 3 27 T11.1-1 sh1 23 F11.2-1a sh1 of 5 27 F10.4-6a sh1 of 6 27 T11.1-2 16 F11.2-1a sh2 of 5 27 F10.4-6a sh2 of 6 27 T11.1-3 16 F11.2-1a sh3 of 5 27 F10.4-6a sh3 of 6 27 T11.1-4 16 F11.2-1a sh4 of 5 27 F10.4-6a sh4 of 6 27 T11.1-5 16 F11.2-1a sh5 of 5 27 F10.4-6a sh5 of 6 27 T11.1-6 16 F11.2-1b sh1 of 3 27 F10.4-6a sh6 of 6 27 T11.1-7 sh1 of 2 24 F11.2-1b sh2 of 3 27 F10.4-6b sh1 of 6 27 T11.1-7 sh2 of 2 16 F11.2-1b sh3 of 3 27 F10.4-6b sh2 of 6 27 T11.1-8 sh1 24 11.3-1 6 F10.4-6b sh3 of 6 27 T11.1-9 6 11.3-2 23 F10.4-6b sh4 of 6 27 T11.1-10 sh1 of 2 24 11.3-3 6 F10.4-6b sh5 of 6 27 T11.1-10 sh2 of 2 24 11.3-4 15 F10.4-6b sh6 of 6 27 T11.1-11 24 11.3-5 27 F10.4-7 24 T11.1-12 sh1 23 11.3-6 15 F10.4-8 24 T11.1-13 sh1 23 11.3-7 13 F10.4-8a 18 T11.1-14 sh1 23 11.3-8 15 F10.4-9 18 T11.1-15 sh1 23 11.3-9 13 F10.4-10 18 11.2-1 6 11.3-10 6 F10.4-11 18 11.2-2 23 11.3-11 20 F10.4-12 18 11.2-3 25 11.3-12 31 F10.4-13 6 11.2-4 17 11.3-13 31 F10.4-14 6 11.2-5 27 11.3-14 23 Page 21 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 11.3-15 29 T11.4-3 sh1 of 2 25 T12.1-1 6 11.3-16 27 T11.4-3 sh2 of 2 31 T12.1-2 6 11.3-17 R T11.4-4 sh1 of 2 16 T12.1-3 23 11.3-18 R T11.4-4 sh2 of 2 31 T12.1-4 6 11.3-19 R F11.4-1 27 T12.1-5 22 11.3-20 R F11.4-2 19 12.2-1 6 11.3-21 R F11.4-3 16 12.2-2 6 11.3-22 R F11.4-4 sh 1 of 4 25 12.3-1 22 T11.3-1 19 F11.4-4 sh 2 of 4 21 12.3-2 24 T11.3-2 19 F11.4-4 sh 3 of 4 21 12.3-3 24 T11.3-3 19 F11.4-4 sh 4 of 4 27 12.3-3a R T11.3-4 19 F11.4-5 sh 1 of 2 26 12.3-3b R F11.3-1a sh1 of 3 27 F11.4-5 sh 2 of 2 27 12.3-3c R F11.3-1a sh2 of 3 27 F11.4-6 21 12.3-3d R F11.3-1a sh3 of 3 27 F11.4-7 sh 1 of 2 19 12.3-4 24 F11.3-1b sh1 of 3 27 F11.4-7 sh 2 of 2 19 12.3-5 24 F11.3-1b sh2 of 3 27 F11.4-8 27 12.3-6 26 F11.3-1b sh3 of 3 27 F11.4-9 21 12.3-7 26 11.4-1 27 11.5-1 26 12.3-8 15 11.4-2 19 11.5-2 26 12.3-9 15 11.4-3 25 11.5-3 26 12.4-1 24 11.4-4 19 11.5-4 32 12.4-2 21 11.4-5 27 11.6-1 16 13-i 31 11.4-5a R 11.6-2 16 13-ii 31 11.4-5b R T11.6-1 15 13-iii 31 11.4-6 27 T11.6-2 sh1 of 9 6 13-iv R 11.4-7 27 T11.6-2 sh2 of 9 6 13.1-1 31 11.4-8 25 T11.6-2 sh3 of 9 6 13.1-1a R 11.4-9 25 T11.6-2 sh4 of 9 6 13.1-1b R 11.4-10 24 T11.6-2 sh5 of 9 6 13.1-2 31 11.4-11 30 T11.6-2 sh6 of 9 6 13.1-2a R 11.4-12 27 T11.6-2 sh7 of 9 6 13.1-2b R 11.4-13 21 T11.6-2 sh8 of 9 6 13.1-3 R 11.4-14 15 T11.6-2 sh9 of 9 6 13.1-3a R 11.4-15 27 12-i 16 13.1-3b R 11.4-16 28 12-ii 26 13.1-4 R 11.4-17 30 12-iii 7 13.1-5 R 11.4-18 26 12.1-1 6 13.1-6 R 11.4-19 22 12.1-2 6 13.1-6a R 11.4-20 31 12.1-3 6 13.1-6b R 11.4-21 30 12.1-4 6 13.1-7 R 11.4-22 18 12.1-5 6 13.1-8 R 11.4-23 28 12.1-6 29 13.1-8a R T11.4-1 sh1 of 4 25 12.1-7 6 13.1-8b R T11.4-1 sh2 of 4 23 12.1-8 29 13.1-9 R T11.4-1 sh3 of 4 28 12.1-9 25 13.1-10 R T11.4-1 sh4 of 4 30 12.1-10 16 13.1-11 R T11.4-2 sh1 of 4 25 12.1-11 16 13.1-12 R T11.4-2 sh2 of 4 30 12.1-12 16 13.1-13 R T11.4-2 sh3 of 4 28 12.1-13 21 13.1-14 R T11.4-2 sh4 of 4 19 12.1-14 15 13.1-15 R Page 22 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 13.1-16 R 14-i 10 14.5-10 6 13.1-17 R 14-ii 6 F14.5-1 6 13.1-18 R 14-iii 6 F14.5-2 6 13.1-18a R 14-iv 6 14.6-1 6 13.1-18b R 14.1-1 6 14.6-2 6 13.1-19 R 14.1-2 6 14.6-3 6 13.1-20 R 14.1-3 6 14.6-4 6 T13.1-1 31 14.2-1 6 14.6-5 6 F13.1-1 31 T14.2-1 sh1 of 11 6 14.6-6 6 F13.1-2 31 T14.2-1 sh2 of 11 6 14.6-7 6 F13.1-3 31 T14.2-1 sh3 of 11 6 F14.6-1 6 F13.1-4 31 T14.2-1 sh4 of 11 6 F14.6-2 6 F13.1-5 R T14.2-1 sh5 of 11 6 15-i 18 F13.1-5a R T14.2-1 sh6 of 11 25 15-ii 31 F13.1-6 R T14.2-1 sh7 of 11 6 15-iii 24 F13.1-7 R T14.2-1 sh8 of 11 6 15-iv 24 F13.1-8a R T14.2-1 sh9 of 11 6 15-v 25 F13.1-8b R T14.2-1 sh10 of 11 6 15-vi 16 F13.1-8c R T14.2-1 sh11 of 11 6 15-vii 16 F13.1-8d R 14.3-1 6 15-viia 16 F13.1-8e R 14.3-2 6 15-viib 10 F13.1-9 R 14.3-3 6 15-viii 24 F13.1-10 R 14.3-4 6 15-ix 24 F13.1-10a R 14.3-5 6 15-x 23 F13.1-10b R T14.3-1 sh1 of 2 6 15-xi 33 F13.1-10c R T14.3-1 sh2 of 2 6 15-xii 33 F13.1-10d R 14.4-1 12 15-xiii 33 F13.1-10e R 14.4-1a 12 15-xiiia R 13.2-1 29 14.4-1b 26 15-xiiib R 13.2-2 30 14.4-1c 12 15-xiv 33 13.3-1 19 14.4-1d 12 15-xv 33 13.4-1 24 14.4-2 6 15-xvi 33 13.5-1 30 14.4-3 6 15-xvii 33 13.5-2 25 14.4-4 6 15-xviii 33 13.5-2a R 14.4-5 18 15-xix 33 13.5-2b R 14.4-6 18 15-xixa R 13.5-2c R T14.4-1 sh1 of 5 12 15-xixb R 13.5-2d R T14.4-1 sh2 of 5 6 15-xixc R 13.5-3 25 T14.4-1 sh3 of 5 21 15-xixd R 13.5-4 20 T14.4-1 sh4 of 5 25 15-xx 33 13.5-5 20 T14.4-1 sh5 of 5 26 15-xxi 33 13.5-6 20 14.5-1 6 15-xxii 33 13.5-7 25 14.5-2 6 15-xxiii 33 T13.5-1 sh1 25 14.5-3 6 15-xxiiia R T13.5-1 sh2 25 14.5-4 6 15-xxiiib R T13.5-1 sh3 25 14.5-5 6 15-xxiv 33 T13.5-2 sh1 22 14.5-6 6 15-xxiva R T13.5-2 sh2 22 14.5-7 6 15-xxivb R 13.6-1 19 14.5-8 6 15-xxv 33 13.6-2 19 14.5-9 6 15-xxvi 33 13.7-1 19 F14.5-1 6 15-xxvii 33 Page 23 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV 15-xxviii 33 F15.1-5 18 15.2-43 29 15-xxix 33 F15.1-6 6 15.2-43a 24 15-xxx 33 F15.1-7 6 15.2-43b 29 15-xxxi 33 F15.1-8 18 15.2-44 29 15-xxxii 33 F15.1-9 18 15.2-45 29 15-xxxiii 33 15.2-1 15 15.2-46 29 15-xxxiv 33 15.2-2 16 15.2-47 18 15-xxxv 33 15.2-3 28 15.2-48 18 15-xxxvi 33 15.2-4 6 15.2-49 31 15-xxxvii 33 15.2-5 11 15.2-50 18 15.1-1 6 15.2-5a 11 15.2-51 18 15.1-2 16 15.2-5b 11 15.2-51a 9 15.1-3 6 15.2-6 18 15.2-51b 9 15.1-4 23 15.2-7 18 15.2-52 22 15.1-5 19 15.2-8 18 15.2-53 22 15.1-6 33 15.2-8a 7 15.2-54 22 15.1-7 33 15.2-8b 7 15.2-55 22 15.1-8 18 15.2-9 18 15.2-56 25 15.1-9 18 15.2-10 18 15.2-57 31 15.1-10 28 15.2-11 18 15.2-58 18 15.1-10a 7 15.2-12 25 15.2-59 18 15.1-10b 7 15.2-13 18 15.2-60 27 15.1-11 18 15.2-14 19 15.2-60a 21 15.1-12 18 15.2-15 18 15.2-60b 16 15.1-13 16 15.2-16 31 15.2-61 24 15.1-14 18 15.2-17 15 15.2-62 31 15.1-15 6 15.2-18 18 T15.2-1 sh1 of 8 18 15.1-16 6 15.2-19 30 T15.2-1 sh2 of 8 18 15.1-17 6 15.2-20 30 T15.2-1 sh3 of 8 32 15.1-18 6 15.2-21 30 T15.2-1 sh4 of 8 24 15.1-19 18 15.2-22 30 T15.2-1 sh5 of 8 24 15.1-20 18 15.2-23 32 T15.2-1 sh6 of 8 24 15.1-21 18 15.2-24 30 T15.2-1 sh6a of 8 29 15.1-22 18 15.2-25 6 T15.2-1 sh7 of 8 18 15.1-23 11 15.2-26 14 T15.2-1 sh7a of 8 18 15.1-24 31 15.2-27 18 T15.2-1 sh8 of 8 32 15.1-25 31 15.2-28 18 T15.2-2 15 15.1-26 31 15.2-29 18 T15.2-3 6 T15.1-1 19 15.2-30 18 F15.2-1 16 T15.1-2 sh1 of 4 21 15.2-31 18 F15.2-2 16 T15.1-2 sh2 of 4 21 15.2-32 24 F15.2-3 16 T15.1-2 sh3 of 4 31 15.2-33 24 F15.2-4 18 T15.1-2 sh4 of 4 31 15.2-34 24 F15.2-5 18 T15.1-3 18 15.2-35 24 F15.2-6 18 T15.1-4 6 15.2-36 18 F15.2-7 18 T15.1-5 16 15.2-37 24 F15.2-8 18 T15.1-6 16 15.2-38 19 F15.2-9 18 F15.1-1 18 15.2-39 26 F15.2-10 18 F15.1-2 18 15.2-40 19 F15.2-11 sh1 of 2 31 F15.1-3 18 15.2-41 19 F15.2-11 sh2 of 2 R F15.1-4 18 15.2-42 26 F15.2-12 18 Page 24 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F15.2-13 18 15.3-2 16 F15.3-16 18 F15.2-14 18 15.3-3 24 F15.3-16a 18 F15.2-15 18 15.3-4 24 F15.3-16b 18 F15.2-16 11 15.3-5 24 F15.3-17 24 F15.2-17 11 15.3-6 24 F15.3-17a 24 F15.2-18 11 15.3-7 15 F15.3-18 24 F15.2-19 18 15.3-8 25 F15.3-18a 24 F15.2-20 24 15.3-9 25 F15.3-19 24 F15.2-21 24 15.3-10 25 F15.3-19a 24 F15.2-22 24 15.3-11 25 F15.3-20 24 F15.2-23 32 15.3-11a 25 F15.3-21 24 F15.2-24 24 15.3-11b 25 F15.3-22 24 F15.2-25 24 15.3-12 14 F15.3-23 24 F15.2-25a 24 15.3-13 28 F15.3-24 24 F15.2-25b 24 15.3-14 16 F15.3-25 24 F15.2-25c 24 15.3-15 31 F15.3-26 24 F15.2-26 18 15.3-16 25 F15.3-27 24 F15.2-27 18 15.3-17 23 F15.3-28 24 F15.2-28 18 15.3-18 16 F15.3-29 24 F15.2-28a 24 15.3-19 15 F15.3-30 24 F15.2-28b 24 15.3-20 18 F15.3-31 24 F15.2-28c 24 15.3-21 31 F15.3-32 24 F15.2-28d 24 T15.3-1 24 F15.3-33 24 F15.2-28e 24 T15.3-1a 24 F15.3-34 24 F15.2-28f 24 T15.3-2 24 F15.3-35 24 F15.2-29a 29 T15.3-2a 24 F15.3-36 24 F15.2-29b 29 T15.3-3 24 F15.3-37 24 F15.2-29c 29 T15.3-3a 24 F15.3-38 24 F15.2-29d 29 T15.3-4 18 F15.3-39 24 F15.2-29e 29 F15.3-1 15 F15.3-40 24 F15.2-29f 29 F15.3-2 24 F15.3-41 24 F15.2-30 18 F15.3-2a 24 F15.3-42 24 F15.2-31 18 F15.3-3 24 F15.3-43 24 F15.2-32 18 F15.3-3a 24 F15.3-44 24 F15.2-33 18 F15.3-4 24 F15.3-45 24 F15.2-34 18 F15.3-4a 24 F15.3-46 24 F15.2-35 18 F15.3-5 24 F15.3-47 24 F15.2-36 18 F15.3-5a 24 F15.3-48 24 F15.2-37 18 F15.3-6 24 F15.3-49 24 F15.2-38 18 F15.3-6a 24 F15.3-50 24 F15.2-39 18 F15.3-7 24 F15.3-51 24 F15.2-40 18 F15.3-7a 24 F15.3-52 15 F15.2-41 24 F15.3-8 24 F15.3-53 24 F15.2-42 24 F15.3-8a 24 F15.3-54 24 F15.2-43a 24 F15.3-9 25 F15.3-55 24 F15.2-43b 24 F15.3-10 25 F15.3-56 24 F15.2-43c 24 F15.3-11 25 F15.3-57 24 F15.2-44 24 F15.3-12 25 F15.3-58 24 F15.2-45 24 F15.3-13 25 F15.3-59 24 F15.2-46 18 F15.3-14 18 F15.3-60 24 15.3-1 24 F15.3-15 18 F15.3-61 24 Page 25 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F15.3-62 24 15.4-36 24 15.4-81 24 F15.3-63 24 15.4-37 22 15.4-82 33 F15.3-64 24 15.4-38 24 15.4-83 24 F15.3-65 24 15.4-39 24 15.4-84 24 F15.3-66 24 15.4-40 23 15.4-85 24 F15.3-67 24 15.4-41 23 15.4-86 24 F15.3-68 24 15.4-42 16 15.4-87 33 15.3A-1 15 15.4-43 18 15.4-88 33 15.4-1 23 15.4-44 15 15.4-89 33 15.4-2 15 15.4-45 19 15.4-90 33 15.4-2a 20 15.4-46 24 15.4-91 28 15.4-2b 24 15.4-47 24 15.4-92 24 15.4-3 15 15.4-48 24 15.4-93 33 15.4-3a 15 15.4-49 26 15.4-94 33 15.4-3b 15 15.4-50 18 15.4-95 24 15.4-4 19 15.4-51 18 15.4-96 24 15.4-5 15 15.4-52 23 15.4-97 24 15.4-6 15 15.4-53 23 15.4-98 24 15.4-7 15 15.4-54 16 15.4-99 24 15.4-8 15 15.4-55 23 15.4-100 24 15.4-9 10 15.4-56 23 15.4-101 24 15.4-10 10 15.4-57 10 15.4-102 24 15.4-11 10 15.4-58 18 15.4-103 33 15.4-12 10 15.4-59 33 15.4-104 24 15.4-13 10 15.4-60 10 15.4-105 24 15.4-14 10 15.4-60a 33 15.4-106 24 15.4-15 10 15.4-60b 33 15.4-107 24 15.4-16 23 15.4-61 24 15.4-108 24 15.4-17 23 15.4-62 20 15.4-109 24 15.4-18 23 15.4-63 28 15.4-110 24 15.4-18a 28 15.4-64 20 15.4-111 24 15.4-18b 28 15.4-65 22 15.4-112 28 15.4-19 23 15.4-66 18 15.4-113 24 15.4-20 23 15.4-67 25 15.4-114 24 15.4-21 23 15.4-68 23 15.4-115 28 15.4-22 28 15.4-69 18 15.4-116 24 15.4-22a 23 15.4-70 6 15.4-117 24 15.4-22b 16 15.4-71 6 15.4-118 24 15.4-23 28 15.4-72 6 15.4-119 24 15.4-24 23 15.4-73 6 15.4-120 24 15.4-25 16 15.4-74 33 15.4-121 33 15.4-26 16 15.4-75 33 15.4-122 R 15.4-27 28 15.4-75a 18 15.4-123 R 15.4-28 28 15.4-75b 7 15.4-124 R 15.4-29 28 15.4-76 23 15.4-125 R 15.4-30 23 15.4-77 23 15.4-126 R 15.4-31 18 15.4-77a 23 15.4-127 R 15.4-32 23 15.4-77b 23 15.4-128 R 15.4-33 24 15.4-78 33 15.4-129 R 15.4-34 23 15.4-79 33 15.4-130 R 15.4-35 31 15.4-80 24 15.4-130a R Page 26 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV T15.4-1 sh1 of 3 15 T15.4-15C sh1 of 4 33 T15.4-21 sh7 of 18 23 T15.4-1 sh2 of 3 24 T15.4-15C sh2 of 4 33 T15.4-21 sh7a of 18 24 T15.4-1 sh2a of 3 24 T15.4-15C sh3 of 4 33 T15.4-21 sh8 of 18 24 T15.4-1 sh3 of 3 33 T15.4-15C sh4 of 4 33 T15.4-21 sh9 of 18 23 T15.4-1 sh3a of 3 33 T15.4-15D sh1 of 4 33 T15.4-21 sh10 of 18 23 T15.4-1 sh3b of 3 18 T15.4-15D sh2 of 4 33 T15.4-21 sh11 of 18 24 T15.4-2 sh1 of 2 15 T15.4-15D sh3 of 4 33 T15.4-21 sh12 of 18 23 T15.4-2 sh2 of 2 15 T15.4-15D sh4 of 4 33 T15.4-21 sh13 of 18 23 T15.4-3 sh1 of 3 15 T15.4-16 sh1 of 5 24 T15.4-21 sh14 of 18 24 T15.4-3 sh2 of 3 23 T15.4-16 sh2 of 5 24 T15.4-21 sh15 of 18 24 T15.4-3 sh3 of 3 23 T15.4-16 sh3 of 5 24 T15.4-21 sh16 of 18 23 T15.4-4 15 T15.4-16 sh4 of 5 24 T15.4-21 sh17 of 18 23 T15.4-5 15 T15.4-16 sh5 of 5 24 T15.4-21 sh18 of 18 24 T15.4-5a sh1 23 T15.4-16A sh1 of 4 33 T15.4-22 24 T15.4-5b sh1 of 2 23 T15.4-16A sh2 of 4 33 T15.5-22A 33 T15.4-5b sh2 of 2 28 T15.4-16A sh3 of 4 33 T15.4-23 24 T15.4-5c sh1 28 T15.4-16A sh4 of 4 33 T15.4-23A 33 T15.4-5d sh1 23 T15.4-17 sh1 of 5 24 T15.4-24 sh1 of 2 33 T15.4-5d sh2 23 T15.4-17 sh2 of 5 24 T15.4-24 sh2 of 2 24 T15.4-5e sh1 23 T15.4-17 sh3 of 5 24 T15.4-25 24 T15.4-6 33 T15.4-17 sh4 of 5 24 T15.4-25A 33 T15.4-6a sh1 23 T15.4-17 sh5 of 5 24 T15.4-26 15 T15.4-6b sh1 26 T15.4-17A sh1 of 4 33 T15.4-27 9 T15.4-7 sh1 23 T15.4-17A sh2 of 4 33 T15.4-28 24 T15.4-7 sh2 R T15.4-17A sh3 of 4 33 T15.4-28A 33 T15.4-7a sh1 26 T15.4-17A sh4 of 4 33 T15.4-29 24 T15.4-7b sh1 23 T15.4-18 sh1 of 3 24 T15.4-29A 33 T15.4-7b sh2 23 T15.4-18 sh2 of 3 24 T15.4-30 24 T15.4-7c sh1 26 T15.4-18 sh3 of 3 24 T15.4-30A 33 T15.4-8 23 T15.4-18A sh1 of 3 33 T15.4-31 24 T15.4-9 23 T15.4-18A sh2 of 3 33 T15.4-31A 33 T15.4-10 20 T15.4-18A sh3 of 3 33 T15.4-32 24 T15.4-11 16 T15.4-19 sh1 of 5 24 T15.4-32A 33 T15.4-12 33 T15.4-19 sh2 of 5 24 T15.4-33 24 T15.4-12a sh1 23 T15.4-19 sh3 of 5 24 T15.4-33A 33 T15.4-12b sh1 26 T15.4-19 sh4 of 5 24 T15.4-34 24 T15.4-13 33 T15.4-19 sh5 of 5 24 T15.4-35 33 T15.4-14 sh1 of 2 33 T15.4-19A sh1 of 5 33 F15.4-1 15 T15.4-14 sh2 of 2 33 T15.4-19A sh2 of 5 33 F15.4-2 15 T15.4-15 sh1 of 1 24 T15.4-19A sh3 of 5 33 F15.4-3 15 T15.4-15 sh2 of 2 R T15.4-19A sh4 of 5 33 F15.4-4a 15 T15.4-15A sh1 of 5 24 T15.4-19A sh5 of 5 33 F15.4-4b 15 T15.4-15A sh2 of 5 24 T15.4-20 sh1 of 3 23 F15.4-4c 15 T15.4-15A sh3 of 5 24 T15.4-20 sh2 of 3 23 F15.4-5 15 T15.4-15A sh4 of 5 24 T15.4-20 sh3 of 3 23 F15.4-6 15 T15.4-15A sh5 of 5 24 T15.4-21 sh1 of 18 23 F15.4-7a 15 T15.4-15B sh1 of 5 24 T15.4-21 sh2 of 18 24 F15.4-7b 15 T15.4-15B sh2 of 5 24 T15.4-21 sh3 of 18 23 F15.4-7c 15 T15.4-15B sh3 of 5 24 T15.4-21 sh4 of 18 23 F15.4-8 15 T15.4-15B sh4 of 5 24 T15.4-21 sh5 of 18 23 F15.4-9 15 T15.4-15B sh5 of 5 24 T15.4-21 sh6 of 18 23 F15.4-10a 15 Page 27 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F15.4-10b 15 F15.4-43c 15 F15.4-62 16 F15.4-10c 15 F15.4-44 15 F15.4-63 16 F15.4-11 15 F15.4-45 15 F15.4-64 16 F15.4-12 15 F15.4-46a 15 F15.4-65 16 F15.4-13a 15 F15.4-46b 15 F15.4-66 16 F15.4-13b 15 F15.4-47 15 F15.4-67 16 F15.4-14 15 F15.4-47a 16 F15.4-68 33 F15.4-15 15 F15.4-47b 16 F15.4-69 33 F15.4-16a 15 F15.4-47c 16 F15.4-70 33 F15.4-16b 15 F15.4-48 31 F15.4-71 18 F15.4-17 15 F15.4-49 31 F15.4-72 18 F15.4-18 15 F15.4-50 19 F15.4-73 10 F15.4-19a 15 F15.4-50a 19 F15.4-74 10 F15.4-19b 15 F15.4-50b 19 F15.4-75 18 F15.4-20 15 F15.4-50c 19 F15.4-76 33 F15.4-21 15 F15.4-50d 18 F15.4-77 33 F15.4-22a 15 F15.4-50e 18 F15.4-78 18 F15.4-22b 15 F15.4-50f 18 F15.4-78a 33 F15.4-22c 15 F15.4-51a 24 F15.4-78b 33 F15.4-23 15 F15.4-51b 24 F15.4-79 24 F15.4-24 15 F15.4-51c 24 F15.4-80 24 F15.4-25a 15 F15.4-51d 18 F15.4-81 24 F15.4-25b 15 F15.4-51e 18 F15.4-82 24 F15.4-25c 15 F15.4-51f 18 F15.4-83 24 F15.4-26 15 F15.4-52a 19 F15.4-84 24 F15.4-27 15 F15.4-52b 19 F15.4-85 18 F15.4-28a 15 F15.4-52c 19 F15.4-86 24 F15.4-28b 15 F15.4-52d 18 F15.4-87 24 F15.4-29 15 F15.4-52e 18 F15.4-88 24 F15.4-30 15 F15.4-52f 18 F15.4-89 24 F15.4-31a 15 F15.4-53a 19 F15.4-90 24 F15.4-31b 15 F15.4-53b 19 F15.4-91 24 F15.4-31c 15 F15.4-53c 19 F15.4-92 9 F15.4-32 15 F15.4-53d 18 F15.4-93 6 F15.4-33 15 F15.4-53e 18 F15.4-94 6 F15.4-34a 15 F15.4-53f 18 F15.4-95 6 F15.4-34b 15 F15.4-54 18 F15.4-96 24 F15.4-34c 15 F15.4-55 15 F15.4-97 24 F15.4-35 15 F15.4-56 15 F15.4-98 24 F15.4-36 15 F15.4-57 15 F15.4-99 24 F15.4-37a 15 F15.4-58 15 F15.4-100 24 F15.4-37b 15 F15.4-59 9 F15.4-101 15 F15.4-38 15 F15.4-60a 24 F15.4-102 15 F15.4-39 15 F15.4-60b 24 F15.4-103 9 F15.4-40a 15 F15.4-60c 24 F15.4-104 9 F15.4-40b 15 F15.4-60d 24 F15.4-105 9 F15.4-40c 15 F15.4-60e 24 F15.4-106 9 F15.4-41 15 F15.4-60f 24 F15.4-107 9 F15.4-42 15 F15.4-60g 24 F15.4-108 9 F15.4-43a 15 F15.4-60h 24 F15.4-109 9 F15.4-43b 15 F15.4-61 23 F15.4-110 9 Page 28 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV F15.4-111 9 17.2-20 R A-14 6 F15.4-112 33 17.2-21 R A-15 16 F15.4-113 33 17.2-22 R A-16 23 F15.4-114 33 17.2-23 R A-17 23 F15.4-115 33 17.2-24 R A-18 23 F15.4-116 33 17.2-25 R A-19 23 F15.4-117 33 17.2-26 R A-20 6 15.5-1 24 17.2-27 R A-21 13 16-1 6 17.2-28 R A-22 6 17-i 23 17.2-29 R A-23 6 17-ii 23 17.2-30 R A-24 14 17-iii 23 17.2-31 R A-25 23 17-iv R 17.2-32 R A-26 23 17.1-1 6 17.2-33 R A-27 6 17.2-1 23 17.2-34 R A-28 6 17.2-2 R 17.2-35 R A-29 6 17.2-3 R 17.2-36 R A-30 6 17.2-4 R 17.2-37 R A-31 6 17.2-4a R 17.2-38 R A-32 6 17.2-4b R 17.2-39 R A-33 6 17.2-5 R 17.2-40 R A-34 23 17.2-6 R 17.2-41 R A-35 23 17.2-6a R T17.2-1 sh1 of 1 23 A-36 16 17.2-6b R T17.2-1 sh2 of 5 R A-36a 18 17.2-7 R T17.2-1 sh3 of 5 R A-36b 16 17.2-7a R T17.2-1 sh4 of 5 R A-37 18 17.2-7b R T17.2-1 sh5 of 5 R A-38 20 17.2-7c R F17.2-1 23 A-39 18 17.2-7d R F17.2-2 9 A-40 23 17.2-7e R F17.2-3 9 A-41 18 17.2-7f R F17.2-4 9 A-41a 20 17.2-7g R APP A 6 A-41b 16 17.2-7h R A-i 6 A-42 23 17.2-7i R A-ii 16 A-43 16 17.2-7j R A-iii 14 A-44 6 17.2-8 R A-iv 6 A-45 6 17.2-8a R A-1 6 A-46 6 17.2-8b R A-2 16 A-47 6 17.2-9 R A-3 16 A-48 6 17.2-10 R A-4 6 A-49 6 17.2-11 R A-5 17 A-50 23 17.2-12 R A-06 6 A-51 17 17.2-13 R A-7 6 A-52 28 17.2-14 R A-8 17 A-53 15 17.2-15 R A-9 6 A-54 14 17.2-15a R A-10 14 A-55 14 17.2-15b R A-10a 15 A-56 16 17.2-16 R A-10b 13 A-57 16 17.2-17 R A-11 7 A-58 6 17.2-18 R A-12 6 A-59 6 17.2-19 R A-13 6 A-60 6 Page 29 of 30

LIST OF CURRENT PAGES SALEM UFSAR REV 33 PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV PAGE/TABLE/FIGURE REV A-61 6 B-15 26 B-66 26 A-62 6 B-16 26 B-67 26 A-63 6 B-17 26 B-68 26 A-64 6 B-18 26 B-69 26 A-65 6 B-19 26 B-70 26 A-66 6 B-20 26 B-71 26 A-67 6 B-21 26 B-72 26 A-68 6 B-22 26 B-73 26 A-69 22 B-23 26 B-74 26 A-70 6 B-24 26 B-75 26 A-71 7 B-25 26 B-76 26 A-72 16 B-26 26 B-77 26 A-73 6 B-27 26 B-78 26 A-74 14 B-28 26 B-79 26 A-75 14 B-29 26 B-80 26 A-76 14 B-30 26 B-81 26 A-77 14 B-31 26 B-82 26 A-78 14 B-32 31 B-83 26 A-79 14 B-33 26 B-84 26 A-80 27 B-34 26 B-85 26 TA-1 14 B-35 26 B-86 26 TA-2 14 B-36 26 B-87 26 TA-3 14 B-37 26 B-88 26 TA-4 sh1 of 3 21 B-38 26 B-89 26 TA-4 sh2 of 3 21 B-39 26 B-90 26 TA-4 sh3 of 3 23 B-40 26 B-91 26 TA-4 sh4 of 4 R B-41 26 B-92 26 TA-5 sh1 of 3 21 B-42 26 B-93 26 TA-5 sh2 of 3 21 B-43 26 TA-5 sh3 of 3 23 B-44 26 TA-5 sh4 of 4 R B-45 26 FA-1 6 B-46 26 FA-2 6 B-47 26 FA-3 6 B-48 26 FA-4 6 B-49 26 FA-5 14 B-50 26 FA-6 14 B-51 26 B-1 26 B-52 26 B-2 26 B-53 26 B-3 33 B-54 33 B-4 26 B-55 33 B-5 26 B-56 33 B-6 26 B-57 26 B-7 26 B-58 26 B-8 26 B-59 26 B-9 26 B-60 26 B-10 26 B-61 26 B-11 26 B-62 26 B-12 26 B-63 26 B-13 26 B-64 26 B-14 26 B-65 26 Page 30 of 30

SECTION 1

Section TABLE OF CONTENTS INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Title

1. 1 PROJECT IDENTIFICATION 1. 1-1 1.2 PLANT SITE

SUMMARY

1.2-1 l.2.1 Site Description 1.2-1 1.2.2 Meteorology 1.2-1 1.2.3 Geology and Hydrology 1.2-1 1.2.4 Seismology 1.2-2 l

2. 5 Marine Ecology 1.2-2 1.2.6 Environmental Radiation Monitoring 1.2-2

1. 2. 7 Facility Safety Conclusions 1.2-2 1.3

SUMMARY

PLANT DESCRIPTION 1.3-1

1. 3. 1 Structures 1.3-2 1.3.2 Nuclear Steam Supply System 1.3-3 1.3.3 Reactor and Plant Control 1.3-4 1.3.4 Waste Disposal System 1.3-4 1.3.5 Fuel Handling System 1.3-5 1.3.6 Turbine and Auxiliaries 1.3-5

1. 3. 7 Electrical System 1.3-6 1.3.8 Engineered Safety Features 1.3-7 1.4 IDENTIFICATION OF CONTRACTORS 1.4-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 17 x 17 Fuel Assembly 1.5-1 1.5.1.1 Rod Cluster Control Spider Tests 1.5-1 1.5.1.2 Grid Tests 1.5-1 1.5.1.3 Fuel Assembly Structural Tests 1.5-1 1-i SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS (Cont) Section Title Page 1.5.1.4 Guide Tube Tests 1.5-2 1.5.1.5 Prototype Assembly Tests 1.5-2 1.5.1.6 Departure from Nucleate Boiling Tests 1.5-2 1.5.I.7 Incore Flow Mixing 1.5-2

1. 5. 2 Other Programs 1.5-2 1.5.2.1 Generic Programs of Westinghouse 1.5-2 1.5.2.2 LOCA Heat Transfer Tests 1.5-3 1.5 .3 References for Section 1.5 1.5-3 J.6 LIST OF ACRONYMS 1.6-1 SGS-UFSAR 1-ii Revision 6 February 15, 1987

LIST OF FIGURES

Figure 1.2-1 General Site Plan Title
SGS-UFSAR 1-iii Revision 7 July 22, 1987

SECTION 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 PROJECT IDENTIFICATION This Updated Final Safety Analysis Report is submitted pursuant to the requirements of 10CFRS0.71 by Public Service Electric and Gas Company (PSE&G) for the two nuclear power units at its Salem Generating Station. PSE&G and Westinghouse Electric Corporation have jointly participated in the design and construction of each unit. On August 21, 2000, the operating licenses for Salem Units 1 & 2 were transferred from PSE&G to PSEG Nuclear LLC. Each unit employs a pressurized water reactor nuclear steam supply system furnished by Westinghouse which is similar in design concept to several other projects licensed by the Nuclear Regulatory Commission. The only systems shared by the two units are Compressed Air, the Control Room Area intake air radiation monitors, parts of the Control Room Area Ventilation System, bulk Nitrogen Supply, Demineralized Water, and the Solid Radwaste Handling system. There are a minimum of shared components; chemical drain and laundry hot shower tanks and pumps and the 20,000 barrel Bulk Fuel Oil Storage Tank are the only components in common. I The licensed core power for both units is 3459 MWt. The approximate values for gross and net electrical outputs are 1178 MWe and 1135 MWe respectively for Unit 1 and 1182 MWe and 1139 MWe respectively for Unit 2. The reactors are expected to be capable of outputs of approximately 3570 MWt, which corresponds I to the valves-wide-open rating of the turbine generators of 1209 MWe gross and 1163 MWe net for Unit 1 and 1220 MWe gross and 1174 MWe net for Unit 2. The containment and engineered safety features for both uni ts have been designed and evaluated at the maximum power rating of 3570 MWt. Loss-of-coolant accidents and those postulated accidents havin~ offsite dose consequences have been analyzed at the power rating of 3570 MWt. 1.1-1 SGS-UFSAR Revision 19 November 19, 2001

The remainder of Section 1 of this report summarizes the principal design features and safety criteria of the nuclear units, pointing out the similarities and differences with respect to other pressurized water nuclear power plant.s employing the same technology and basic engineering features as the Salem Generating Station. Sec::icn 2 contains a description and evaluation of the site and environs, supporting the suitability of the site for a nuclear plant of the size and type described. Section 3 discusses the identification, description, and discussion of principal archjt.ec::ural and engineering design of structures, components, equipment, and systems important to safety. The reactor is described in Section 4. Section 5 discusses the Reactor Coolant Systerr, and related systems, and Sections 6 through 1 the emergency and auxiliary systems. Section 13 describes the Cornpa;-iy' s program for organization and training of plan:: personnel. Section 14 contains an outline and description of the initial tests and operations associated with plant startup. Section 15 is a safety evaluation summarizing the analyses which demonstrate the adequacy of the Reactor Protection System, and the engineered safety features. The consequences of various postulated accidents are within the guidelines set forth in the Nt;.clear Regulatory Commission's Rules and Regulations and 10CFR50.67, Accident Source Term. 1.1-2 SGS-UFSAR Revision 25 October 26, 2010

1.2 PLANT SITE

SUMMARY

1.2.1 Site Description The approximately 700 acre Salem site is located along the eastern shore of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey about 8 miles southwest of Salem, New Jersey. The population density of the area surrounding the site is low. Distance to the site boundary is about 4200 feet. The nearest residence is approximately 3.4 miles west of the site in Bay View Beach, Delaware. Other nearby residences are located 3.5 miles east-northeast and 3.5 miles northwest of the site. The population center distance is 15.5 miles. The area is primarily utilized for agricultural pursuits, with heavy industry located generally 15 miles and beyond to the north of the site.

1.2.2 Meteorology The meteorological data pertinent to the Salem site has been reviewed, and there is no reason to anticipate unusual meteorological problems. The terrain is open and extremely flat, and the land-sea interaction favors a vigorous wind flow. A meteorological tower facility was established northwest of the reactor area on the site to provide actual site meteorological data. This data collection program has been terminated as sufficient data has been collected and analyzed to describe the dispersion parameters. The tower has been relocated east of the site. 1.2.3 Geology and Hydrology An investigation of Salem site geology and hydrology was completed in 1967. The nearest known faulting is approximately 25 miles from the site. Test borings at the site indicate that subsurface conditions are adequate to support the structures. The regional direction of ground water movement is toward the Delaware River, 1.2-1 SGS-UFSAR Revision 6 February 15, 1987

and all surface drainage at the Salem site flows directly into the river. 1.2.4 Seismology The site is located in a region which has experienced only infrequent minor earthquake activity. No known faults exist in the basement rock or sedimentary deposits in the immediate vicinity of the site. Significant earthquake motion is not expected at the site during the life of the facility. The plant was conservatively designed to respond elastically, with no loss of function, to horizontal ground accelerations as high as 10 percent of gravity, and the design was checked for a hypothetical acceleration of 20 percent of gravity. 1.2.5 Marine Ecology A thorough study of the biological makeup of the Delaware Estuary is being conducted. The study has continued since the plant went into operation to determine the effects (if any) of plant operation on the ecology of the Estuary. 1.2.6 Environmental Radiation Monitoring An environmental radiation monitoring program for the site and surrounding area is being conducted. This program has continued since the plant went into operation to determine the effects (if any) of plant operation on radiation levels in the environment. 1.2.7 Facility Safety Conclusions The safety of the public and plant operating personnel and reliability of plant equipment and systems have been the primary considerations in the plant design. The approach taken in fulfilling the safety consideration is three-fold. First, careful attention has been given to design so as to prevent the release of 1.2-2 SGS-UFSAR Revision 6 February 15, 1987

radioactivity to the environment under conditions which could be hazardous to the health and safety of the public. Second, the plant has been designed so as to provide adequate protection for plant personnel wherever a potential radiation hazard exists. Third, Engineered Safety Features have been designed with redundancy and diversity, and to stringent quality standards. Based on the over-all design of the plant including its safety features and the analyses of possible incidents including the design basis accident, it is concluded that the Salem Generating Station can be operated without undue risk to the health and safety of the public .

SGS-UFSAR 1.2-3 Revision 6 February 15, 1987

e LEGEND

1. Unit 1 Containment 8. Admlnilnnrlian luldlng

2. Unit 2 Contli11ment 9. Circulllllnl waw ln1ak4I

3. Auxiliary laildlng 10.SemceW-lntake

4. Unit 1 FUii Hllncling 11. 500 KV Swilctryard

5. Unit 2 PUii Handing , 12. Fuel Oil Storagt Tri

6. Serva ....... 13. Clean faciltiN Buiklng

7. Turbine Genlnltor Are

14. Cwnttolld Facl'lties luilding

15. Unit 3 Gas Tt.WbiM Il

16. SERVICE WATER ACCUMULATOR ENCLOSURE I

I D .. C: l* J

    ....JI       ,s ~

O[JCD 12 13 Hope Creek Generating Stauon

                                                                                 ~f..
                                                                       ~~

fJiV REVISION 17 OCTOBER 16 1998 General Site Plan PUBLIC SERVICE ELECJRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 1.2-1

Fl.2-2 entionally Re r to plant 204803 in DCRMS SGS-UFSAR Revision 27 November 25, 20

1.3

SUMMARY

PLANT DESCRIPTION

The inherent design of the pressurized watert closed-cycle reactor minimizes atmosphere.

the Four quantities barriers of accumulation and the environment. fission exist products between the released to the fission product These are the uraniwn dioxide fuel matrix, the fuel cladding, the reactor vessel and coolant loops, and the reactor containment. The consequences of a breach of the fuel cladding are greatly reduced by the ability of the uranium dioxide lattice to retain fission products. Escape of fission products through a fuel cladding defect would be contained within the pressure vessel, loops, and auxiliary systems. Breach of these systems or equipment would release the fission products to the reactor containment where they would be retained. The reactor containment is designed to adequately retain these fission products under the most severe accident conditions, as analyzed in Section 15. Several engineered safety features have been incorporated into the plant design to reduce the consequences of a loss-of-coolant accident (LOCA). These safety features include an Emergency Core Cooling System (ECCS). This system automatically delivers borated water to the reactor vessel for cooling the core under high and low reactor coolant pressure conditions. The ECCS also serves to insert negative reactivity into the core in the form of borated water during plant cooldown following a steam line break or an accidental steam release. Other safety features which have been included in the reactor containment design are a Containment Fan Cooler System which acts to effect depressurization of the containment following a LOCA and to remove particulate matter from the containment atmosphere, and a Containment Spray System which acts to depressurize the containment and remove elemental iodine from the atmosphere by washing action. The Containment Spray System provides redundant backup by an alternate principle for the Containment Fan Cooler System for heat removal .

SGS-UFSAR 1.3-1 Revision 6 February 15, 1987

1.3.1 Structures The major structures include a separate and independent Containment and Fuel Handling Building for each reactor, a common Auxiliary Building with holdup tank vault, a common Turbine Building and a common Administration and Service Building. General layouts of the Reactor Containment, Auxiliary Building, and interior component arrangements are shown on Figures 1. 2-1, 5 .1-12 and Plant Drawings 204803, 204804, 204805, 204806, 204807 and 204808. Seismic Criteria For Category I ( seismic) equipment, dynamic methods or conservative static equivalents were used to determine that components and structures will operate or maintain their integrity, as required. For Category II (seismic) equipment, static methods were used and non-seismic equipment meets applicable codes. Definition of Seismic Categories Particular structures and equipment are classified according to seismic design. The seismic definitions are:

1. Category I (seismic)

Those structures, mechanical components, the Reactor Protection System, and Engineered Safety Features Actuation System whose failure might cause or increase the severity of a LOCA. Also, those structures and components vital to safe shutdown and isolation.

1. 3-2 SGS-UFSAR Revision 27 November 25, 2013

2. Category II (seismic)

Those structures and mechanical components that are not Category I (seismic), but which function in direct support of reactor operation. 1.3.2 Nuclear Stearn Supply System The Nuclear Stearn Supply System for each unit consists of a pressurized water reactor, Reactor Coolant System (RCS), and associated auxiliary fluid systems. The RCS is arranged as four closed reactor coolant loops connected in parallel to the reactor vessel, each containing a reactor coolant pump and a steam generator. An electrically heated pressurizer is connected to one of the loops. The reactor core is composed of uranium dioxide pellets enclosed in Zircaloy-based tubing with welded end plugs. The tubes are supported in assemblies by spring clip grid structures. The control rods consist of clusters of stainless steel clad silver-indium-cadmium absorber rods located within the fuel assemblies. The nuclear fuel is typically loaded in three regions, with the new fuel being introduced into the core interior and by its third cycle of operation being discharged from the core's outermost region to spent fuel storage. The reactor vessel and reactor internals contain and support the fuel and control rods. The reactor vessel is cylindrical with hemispherical heads and is clad with stainless steel. The pressurizer is a cylindrical pressure vessel with hemispherical heads and is equipped with electrical heaters and spray nozzles for system pressure control. 1.3-3 SGS-UFSAR Revision 19 November 19, 2001

The st~am generators are vertical U-tube units utilizing Inconel tubes. Integral moisture separating equipment reduces the moisture content of the steam at the turbine throttle to S 0.25 percent for Unit 1 and S 0.1 percent for Unit 2. The reactor coolant pumps are vertical single stage centrifugal pumps equipped with controlled leakage shaft seals. Auxiliary systems are provided to charge the RCS, add makeup water, purify reactor .coolant water, provide chemicals for corrosion inhibition and reactor control, cool system components, remove residual heat when the reactor is shutdown, cool the spent fuel storage pool, sample reactor coolant water, provide for emergency safety injection, and vent and drain the RCS.

1. 3. 3 Reactor and Plant Control The reactor is controlled by a coordinated combination of soluble neutron absorbers and mechanical control rods. The control system allows the plant to accept step load changes of 10 percent and ramp load changes of 5 percent per minute over the load range of 15 to 95 percent power under normal operating conditions.

Complete supervision of each reactor and turbine generator is accomplished from each unit's control room.

1. 3. 4 Waste Disposal System The Waste Disposal Systems provide all the equipment necessary to collect, process, and prepare for disposal, all radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation.

After collection, liquid wastes are evaporated and/or demineralized if necessary to reduce activity levels. The treated water from the demineralizers or the evaporator distillate may be recycled for use in the plant or may be discharged via the

1. 3-4 SGS-UFSAR Revision 24 May 11, 2009

condenser discharge aL concentrations well within the limits set forth in 10CFR20. The evaporator concentrates and spent demineralizer resins are solidified, drummed, and shipped from the site for ultimate disposal in an authorized location. Gaseous wastes are collected and held up for radioactive decay, after which they rnay be reused for blanketing tanks. Decayed gases are discharged to the environment in a controlled manner which maintains the offsite dose well below the limits set forth in 10CFR20. 1.3.5 Fuel Handling System E:ach reactor is refueled with equipment designed to handle spent fuel Gnder water from the time it leaves the reactor vessel until it is placed in a cask for dry storage at the Independent Spent Fuel Storage Installation (ISFSI) or for shipment from the site. Underwater transfer of spent fuel provides an optically t~ansparent radiation shield as well as a reliable source of coolant for removal of decay heat. T~is system also provides capability for receiving, handling, and storing new fuel. The Spent Fuel Pool Cooling System has been redesigned to include a second permanent spent fuel pool cooling pump. 1.3.6 Turbine and Auxiliaries The turbine is a four casing, tandem-compound, six flow exhaust, 1800 rpm unit with 44-inch last stage blades. There are six combination moisture separator-steam reheat er assemblies. The turbine generators are rated as described in Section 1, with saturated inlet steam conditions of 765 psia, exhausting at 1.5 inches of mercury absolute, at zero percent makeup. There are six stages of feedwater heating. The turbine is equipped with an Electro-Hydraulic Control System, which uses an electronic controller and a high-pressure fire resistant fluid system to control valve movement.

1. 3-5 Revision 25 October 26, 2010

The condenser is of the single pass deaerating type. There are three strings of feedwater heaters, three one-third size condensate and heater drain pumps and two one-half size feedwater pumps. Drains from the two highest feedwater heaters are pumped into the Condensate System and drains from the four lowest feedwater heaters are cascaded to the condenser. 1.3.7 Electrical system Each main generator is a 1300 MVA, 25 kV, 3 phase, 60 cycle, 0.9 pf, 1800 rpm, 75 psig hydrogen inner-cooled unit with water cooled stator windings. Field excitation is provided by a direct shaft driven brushless excitation system. Each generator is connected to the primary side of three single phase main stepup transformers through isolated phase buses. The secondary side of each main transformer delivers power to the 500 kV switchyard. The station service systems consist of a 13.8 kV north ring bus, and 13.8 kV south hue sections, auxiliary and station power transformers, 4160 V, 460 V, 230 V, and 115 V ac and 250 V, 125 V, and 28 V de buses and equipment. A third 500 kV system tie, the 13.8 kV north ring bus and 13.8 kV south bus sections, arrangement replaces the 69 kV single source described in the Preliminary Safety Analysis Report. This provides a superior power supply system to the station. Three diesel-generators per unit are provided as onsite sources of power in the event of complete loss of normal ac power. These generators power the post-accident containment cooling equipment as well as the safety injection, centrifugal charging, and residual heat removal pumps to assure an acceptable post loss-of-coolant containment pressure transient and adequate core cooling. Two-out-of-the three diesel-generators can handle the electrical load required for a unit in the event of a LOCA. SGS-UFSAR Revision 14 December 29 1 1995

1.3.8 Engineered Safety Features The engineered safety features provided for each unit have sufficient redundancy of components and power sources such that under the conditions of a LOCA they can maintain the integrity of the containment and maintain the exposure of the public below the limits set forth in 10CFR50. 67, even when I operating with partial effectiveness. The engineered safety features incorporated in the design of each unit and the functions they serve are summarized below.

1. The ECCS injects borated water into the RCS. This system limits damage to the core and limits the energy and fission products released into the containment following a LOCA.

The system has been extensively redesigned by Westinghouse. The basic changes in the redesigned system are the use of two charging pumps from the Chemical and Volume Control System for high head injection in addition to their normal charging function and the relocation of the boron injection tank to the discharge side of these pumps. The design of these pumps was changed from reciprocating to centrifugal. Piping, valving, and instrumentation were also revised as a result of the system redesign.

2. A steel-lined concrete containment vessel consisting of reinforced concrete cylindrical wall, a hemispherical dome, and a reinforced concrete base with testable high integrity penetrations.

3. Reactor containment fan coolers and filters to reduce containment pressure and filter particulate matter following a LOCA.

1. 3-7 SGS-UFSAR Revision 25 October 26, 2010

4. A Containment Spray System to reduce containment pressure and remove iodine from the containment atmosphere.

5. The containment Isolation System incorporates valves and controls on piping systems penetrating the containment structure. These valves are arranged to provide two barriers between the RCS or containment atmosphere and the environment. System design is such that failure of one valve to close will not prevent isolation, and no manual operation is required for immediate isolation.

Automatic isolation is initiated by a containment isolation signal, derived for Phase A isolation by the safety injection signal and high-high containment pressure signal for Phase B isolation.

6. Power sources for the engineered safety features for each unit are provided by two 4 kV power circuits fed from the 500 kV system through the south 13 kV substation in the 500 kV switchyard.

The 500 kV switchyard arrangement consists of three 500 kV transmission lines connected to a breaker-and-a-half design with four 500-13 kV transformers. Two of them are connected to the 500 kV main bus section 1, the other two are connected to Section 2. SGS-UFSAR 1.3-8 Revision 14 December 29, 1995

Two 500-13 kV transformers provide power to the south 13kV bus sections (one transformer per section) while the other two transformers feed the north 13kV ring bus. Each south 13kV bus section feeds two 13-4 kV transformers, one for each unit, to provide off-site power for the engineered safety features and new Circulating Water Switchgear. The north 13 kV ring bus is normally operated split to allow one 500-13 kV transformer to feed two (one for each unit) 13-4 kV transformers for Group buses. Should one out of two 500-13kV transformers feeding the north 13kV ring bus be out of service, the ring bus will be realigned to provide power to all four 13-4kV transformers for both unit group buses from the remaining transformer. If one out of two 500-13kV transformers feeding the south 13kV bus is out of service, transformers connected to the ring bus will be realigned in such a way that one transformer replaces the lost one while the other provides power to all four 13-4kV transformers for the group buses. During this 500-13kV transformer swap over period, the double ended 4kV vital buses receive power from the second off-site power source. Reliable diesel-generator power is provided for the engineered safeguards loads in the event of failure of station auxiliary power. In addition, if external auxiliary power to the station is lost concurrent with an accident, power is available for the engineered safeguards from the diesel-generatore, which are capable of supplying the engineered safeguards load to assure protection of the health and safety of the public in the event of a LOCA.

7. All components necessary for the proper operation of the engineered safety features are operable from the control room *

SGS-UFSAR 1.3-9 Revision 15 June 12, 1996

1.4 IDENTIFICATION OF CONTRACTORS The Salem Generating Station was designed and constructed by Public Service Electric & Gas (PSE&G). Westinghouse Electric Corporation designed and furnished the nuclear steam supply equipment and systems including the fuel assemblies. PSE&G contracted United Engineers and Constructors Inc. of Philadelphia, Pennsylvania, to supervise field erection. PSE&G also engaged several consultants to provide technical assistance in various areas. These consultants are listed below. Consultant Program Southwest Research Institute, Quality Control San Antonio, Texas

s. M. Stoller Corporation, New York, Reactor Core and Nuclear New York Fuel Cycle Smith - Singer Meteorologists, Inc., Meteorology (Now Meteorological Evaluation systems, Inc. )

Amityville, Long Island, New York Dames and Moore, Cranford, New Jersey Geology, Hydrology, Seismology Pritchard - Carpenter, Consultants, Hydrology Coral Gables, Florida Radiation Management Corporation, Radiation Monitoring, Philadelphia, Pennsylvania Emergency Planning Ichthyological Associates, Marine Ecology Middletown, Delaware 1.4-1 SGS-UFSAR Revision 6 February 15, 1987

Porter-Gertz, Consultants, Inc., Radiation Monitoring, (Now Porter Consultants) Emergency Planning Ardmore, Pennsylvania I Framatome Technologies, Inc., (FTI) Unit 1 Steam Generator of Lynchburg, Va. and Raytheon Corp. changeout During the operational phase, various consultants and contractors have been employed to support station operation. These organizations are selected and perform the applicable service in accordance with the Salem QA Manual. 1.4-2 SGS-UFSAR Revision 18 April 26, 2000

1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION

One of the design bases for the Salem Generating Station has been to utilize well-developed and proven design concepts, systems, and equipment, in order to minimize the potential for schedule overruns and to enhance the reliability of operation.

cost and As a consequence, there have been few requirements for research and development programs to confirm the adequacy of the design. Those programs identified for Salem have been satisfactorily completed, as described in Section 1.5.1. Other programs were identified as valuable to define margins of conservatism or possible design improvements. Relevant programs in this latter category are described in Section 1.5.2 1.5.1 17 x 17 Fuel Assembly A comprehensive test program for the 17 x 17 assembly has been successfully completed by Westinghouse. Reference 1 contains a summary discussion of the program. The following sections present specific references documenting individual portions of the program. 1.5.1.1 Rod Cluster Control Spider Tests Rod cluster control spider tests have been completed. For a further discussion of these tests, refer to Section 4.2.3.4. 1.5.1.2 Grid Tests Verification tests of the structural adequacy of the grid design have been completed. Refer to Section 4.2.3.4 and Reference 2 for a discussion of these tests. 1.5.1.3 Fuel Assembly Structural Tests Fuel assembly structural tests have been completed. Refer to References 2 and 3 for a discussion of these tests. 1.5-1 SGS-UFSAR Revision 6 February 15, 1987

1.5.1.4 Guide Tube Tests Verification tests of the structural adequacy of the guide tubes have been completed. of these tests. Refer to References 3 and 4 for a discussion 1.5.1.5 PrototyPe Assembly Tests Verification tests of the integrated fuel assembly and rod cluster control per£ormance have been completed. Refer to References 3 and 4 for a discussion of these tests. 1.5.1.6 Departure from Nucleate Boiling Tests The test program for experimentally determining the effect of the fuel assembly geometry on the departure from nucleate boiling (DNB) heat flux has been completed. Refer to Reference 5 for a discussion of these tests. 1.5.1.7 Incore Flow Mixing The experimental test program to determine the effects of the fuel assembly geometry on mixing has been completed. Refer to Reference 6 for a discussion of these tests. 1.5.2 Other Programs 1.5.2.1 Generic Programs of Westinghouse Reference 7 summarizes ongoing safety-related research and development programs that are being carried out for, or by, or in conjunction with the Westinghouse Nuclear Energy System Division and that are applicable to Westinghouse pressurized water reactors. SGS-UFSAR 1.5-2 Revision 6 February 15, 1987

1.5.2.2 LOCA Heat Transfer Tests Experimental test programs to determine the thermal-hydraulic characteristics of 17 x 17 fuel assemblies and to obtain experimental re flooding transfer data under simulated loss-of-coolant accident (LOCA) conditions have been completed. Refer to Refer*ence 8 for a discussion of these tests. A single rod burst test program to quantify the maximum assembly flow blockage which is assumed in the LOCA analyses has been completed. Refer to Reference 9 for a discussion of these tests. The results of these two test programs have been used in the Emergency Core Cooling System analyses in Chapter 15. 1.5.3 References for Section 1.5 I. Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries - Spring 1976,." June 1976 ..

2. Gesinski, L. and Chiang, D. , "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary) and WCAP-8288 (Non-Proprietary), December 1973.

3. DeMario, E. E. , "Hydraulic Flow Test of the 17 x 17 Fuel Assembly, 11 WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February 1974.

4. Cooper, F. W., Jr., 0 17 x 17 Driveline Component Tests -

Phase IB, II, III, D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974.

5. Hill, K. W., et aL, "Effects of 17 x 17 Fuel Assembly Geometry on DNB, 11 WCAP-8296-P-A (Proprietary) and WCAP-8297-A (Non-Proprietary), February 1975.

1.5-3 SGS-UFSAR Revision 6 February 15, 1987

6. Cadek, F. F. ; Motley, F. E. ; and Dominicis, D. P. , "Ef feet of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid, 11 WCAP-7941-P-A (Proprietary) and WCAP-7959-A (Non-Proprietary), January 1975.

7. Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries .. Winter 1977 - Summer 1978," WCAP-8768, Revision 2, October 1978.

8. "Westinghouse ECCS Evaluation Model - October 1975 Version, 11 WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),

November 1975.

9. Kuchirka, P. J., "17 x 17 Design Fuel Rod Behavior During Simulated Loss-of-Coolant Accident Cqnditions, 1 WCAP-8289 (Proprietary) and WCAP-8290 (Non-Proprietary), November 1974.

SGS-UFSAR 1.5-4 Revision 6 February 15, 1987

1.6 LIST OF ACRONYMS The following is an alphabetical listing of the most frequently used acronyms in this report. AEC Atomic Energy Commission AFST Auxiliary Feedwater Storage Tank AFW Auxiliary Feedwater AIF Atomic Industrial Forum ALA.RA - As Low as is Reasonably Achievable ALP Actuation Logic Processor (AMSAC) ALS Actuation Logic System (AMSAC) AMSAC - Actuation Mitigation System Actuation Circuitry ANS American Nuclear Society ANSI American National Standards Institute AO Axial Offset ASTM American Society for Testing and Materials ATWS Anticipated Transient Without SCRAM BIT Boron Injection Tank BNWL Battelle Northwest Laboratory BOL Beginning-of-Life 1.6-1 SGS-UFSAR Revision 10 July 22, 1990

BOP Balance-of-Plant BTP Branch Technical Position

1. 6- la SGS-UFSAR Revision 10 July 22, 1990

THIS PAGE INTENTIONALLY BLANK 1,6-lb SGS-UFSAR Revision 10 July 22, 1990

BWR Boiling Water Reactor CAACS control Area Air Conditioning system CAP Chemical Analysis Panel CASP Containment Air Sampling Panel CCP Centrifugal Charging Pump CERC coastal Engineering Research Center CFR Code of Federal Regulations CIS Containment Isolation system CPS Condensate Polishing System CRDM Control Rod Drive Mechanism CRS control Room supervisor CSAS Containment spray Actuation system css containment Spray System eves Chemical and Volume Control System CVTR Carolina-Virginia Tube Reactor cws Circulating Water System DAS Data Acquisition system DBA Design Basis Accident DBE Design Basis Earthquake DCRDR Detailed Control Room Design Review 1.6-2 SCS-UFSAR Revision 16 January 31, 1998

DEPS Double-Ended Pump Suction DF Decontamination Factor DNB - Departure from Nucleate Boiling DNBR - Departure from Nucleate Boiling Ratio DOT Department of Transportation d/p differential pressure ORF Dose Reduction Factor DTT Ductility Transition Temperature DVRPC Delaware Valley Regional Planning commission E&CD Engineering and Construction Department EACS - Emergency Air conditioning system ECCS Emergency Core Cooling System EOL - End-of-Life EPD - Electric Production Department EPRI Electrical Power Research Institute EPZ Emergency Planning Area ESF Engineered Safety Features FPS Fire Protection system 1.6-3 SGS-UFSAR Revision 16 January 31, 1998

FSAR GDC GM Final Safety Analysis Report General Design Criteria Geiger-Mueller GPM Gallons Per Minute GWD Gigawatt Day GWS Gaseous Waste System HED - Human Engineering Deficiency HEPA - High-Efficiency Particulate Air HFP - Hot Full Power HHW - High-High Water HP High Pressure I&C Instrumentation and Control ICE Instrumentation Controls and Electrical IEEE Institute of Electrical and Electronics Engineers I/0 Input/Output LCR License Change Request LDP Lighting Distribution Panel LP Low Pressure LPG Liquified Petroleum Gas 1.6-4 SGS-UFSAR Revision 10 July 22, 1990

LPM Loose Parts Monitoring 1.6-4.a SGS-UFSAR Revision 10 July 22, 1990

THIS PAGE INTENTIONALLY BLANK SGS*UFSAR

1. 6-4b Revision 10 July 22, 1990

LNG Liquified Natural Gas LOCA Loss-of-Coolant Accident LOFT Loss of Fluid Test LSP Liquid Sampling Panel LWS Liquid waste System MCD Minor Civil Division MEL Master Equipment List (Section 17.2) MEL Moderate Energy Lines MI Mechanical and Integrated MIG Manual Inert Gas MMA Manual Metal Arc MMI Modified Mercalli Intensity MOL Middle-of-Life MSIV Main Steam Isolation Valve MSL Mean Sea Level MSR Moisture Separator-Reheater KWD/ MTU Megawatt Days per Metric Ton of Uranium NBS National Bureau of Standards NBU Nuclear Business Unit NCO Nuclear control Operators 1.6-5 SGS-UFSAR Revision 16 January 31, 1998

NOT Nil Ductility Transition NEMA National Electric Manufacturers' Association NFPA National Fire Protection Association NIS Nuclear Instrumentation System NML Nuclear Mutual Limited NPSH Net Positive Suction Head I NOS Nuclear Oversight Department NRB Nuclear Review Board NRC Nuclear Regulatory Commission NSR - Nuclear Safety Review Department NSSS Nuclear Steam Supplier System NWS National Weather Service OBE Operating Basis Earthquake OD Outside Diameter O&M Operations and Maintenance ORNL Oak Ridge National Laboratories OB - Operations Superintendent OSHA Occupational and Safety Health Act OTG Operational Test Group PASS Post Accident Sampling System 1.6-6 SGS-UFSAR Revision 22 May 5, 2006

PLUS Parcel Land Use System PMH Probab1e Maximum Hurricane POPS Pressurizer overpressure Protection System. PORC Preoperational Testing Review committee PORV Power-Operated Relief Valve PRT Pressurizer Relief Tank PSAR Preliminary Safety Analysis Report PSD Public Service Datum PSE&G Public Service Electric & Gas PSSUG Public Service Electric & Gas Startup Group* PVRC Pressure Vessel Research Committee PWR Pressurized Water Reactor QA Quality Assurance QC Quality control RA Reduction Area RAMPS Repair and Maintenance Procedure System RCC Rod Cluster control RCS Reactor coolant syste~ RCCA Rod Cluster Control Assembly

SGS-UFSAR 1.6-7 Revision 6 February 15, 1987

RCFC Reactor containment Fan cooler RCS Reactor coolant System RCL Reactor Coolant Loop RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RCS Reactor coolant system REMP Radiological Environmental Monitoring Program REP Radiation Exposure Permit RG Regulatory Guide RHR - Residual Heat Removal

       - Radiation Monitoring System RMS RMS       Root-Mean-square RPS    -  Reactor Protection System RSE       Reload Safety Evaluation RTD       Resistance Temperature Detector RVED      Reactor Vessel Examination Device RWST      Refueling Water Storage Tank SBO       Station Blackout SEC       Safeguards Equipment Control SER       Safety Evaluation Report SGS-UFSAR 1.6-8 Revision 15 June 12, 1996

SGB -: Steam Generator Blowdown SGS - Salem Generating Station SIS - Safety Injection System SMSA - Standard Metropolitan Statistical Area SORC - Station Operations Review Committee SPDS - Safety Parameter Display System SRP - Standard Review Plan SRSS - Square-Root-of-the-Sum-of-the-Square SSE - Safe Shutdown Earthquake SSG - Salem Startup Group SSPS - Solid State Protection System STA - Shift Technical Advisor STGD - Steam Turbine - Generator Division SWRI - Southwest Research Institute SWS - Service Water System TDC - Thermal Diffusion Coefficient TDH - Total Dynamic Head TDS - Total Dissolved Solid TLD - Thermoluminescent Dosimeter 1.6-9 SGS-UFSAR Revision 6 February 15, 1987

T/MS TSC Three Mile Island Test/Maintenance System (AMSAC) Technical Support Center UE&C United Engineers & Constructors UFSAR Updated Fi~al Safety Analysis Report UPS Uninterruptible Power System USE Upper Shelf Energy UTG United Engineers and Constructors Test Group UTS Ultimate Tensile Stress VCT Volume Control Tank WDS Waste Disposal System w.g. - water gage W'ILMAPCO

Wilmington Metropolitan Area Planning Council WMID Wisconsin-Michigan Inspection Device W'OL Wedge Opening Loading 1.6-10 SGS-UFSAR Revision 10 July 22, 1990

SECTION 2 SITE CHARACTERISTICS

TABLE OF CONTENTS Section Page 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1-1 2.1.1 Site Location 2.1-1 2.1.2 Site Description 2.1-2 2.1.2.1 Exclusion Area Control 2.1-2 2.1.2.2 Boundaries for Establishing Effluent Release Limits 2.1-3 2.1.3 Population Distribution 2.1-3 2.1.3.1 Population Within 10 Miles 2.1-4 2.1.3.1.1 Population Projections for Oto 10 Mile Area 2.1-5 2.1.3.1.2 Population Update Within 10 Miles 2.1-5 2.1.3.2 Population Between 10 and 50 Miles 2.1-10 2.1.3.2.1 Population Projections for 10 to 50 Mile Area 2.1-10 2.1.3.2.2 Population Update 10 to 50 Miles 2.1-10 2.1.3.3 Low Population Zone 2.1-14 2.1.3.4 Transient Population 2.1-15 2.1.3,5 Population Center 2.1-15 2.1.3.6 Public Facilities and Institutions 2.1-16 2.1.3.6.1 Schools 2.1-16 2.1.3.6.2 Hospitals and Nursing_Homes 2.1-17 2.1.3.6.3 Correctional Institutions 2.1-17 2.1.3.6.4 Recreational Facilities 2.1-18 2.1.3.7 Population Projection Methodology 2.1-18 2.1.4 Use of Adjacent Land 2.1-22 2.1.4.1 Recreational Land Use 2.1-25 2.1.5 References for Section 2.1 2.1-25 SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS (Cont) Section Title bu. 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACIUTIES 2.2-1 2.2.1 Location and Routes 2.2-1 2.2.2 Descriptions 2.2-1 2.2.2.1 Missile Bases or Missile Sites 2.2-1 2.2.2.2 Manufacturing Plants 2.2-2 2.2.2.3 Chemical Plants and Storage Facilities 2.2-2 2.2.2.4 Oil and Gas Pipelines and Tank Farms 2.2-2 2.2.2,5 Transportation Complexes (Harbors, R.ailway Yards, Airports) 2.2-2 2.2.2.6 Transportation Routes (Highways, Railway, and Waterways) 2.2-3 2.2.2.7 Petroleum Wells, Mines, or Quarries 2.2-5 2.2.3 Evaluations 2.2-5 2.2.3.1 Barge Transportation 2.2-5 2.2.3.2 Hazardous Chemicals* Onsite 2.2-9 2.2.3.3 Hazardous Chemicals* Offsite 2.2-12 2.2.4 References for Section 2.2 2.2-13 2.3 METEOROLOGY 2.3-1 2.3.1 Regional Climatology 2.3-1 2.3.1.1 Data Sources 2.3-1 2.3.1.2 General Climate 2.3-1 2.3.1.2.1 Precipitation 2.3-1 2.3.1.2.2 Humidity. Vinds 2.3-2 2.3.1.3 Severe \leather 2.3-2 2.3.2 Local Meteorology 2.3-2 2.3.3 Onsite Meteorological Measurements 2.3-3 Program 2.3.3.1 Preoperational Data Collection Program 2.3.3.1.1 Data Summaries and Turbulence Classifications 2.3-4 2.3.3.2 Operational Data Collection Program 2.3-7 2-ii SGS-UFSAR Revision ll July 22, 1991

TABLE OP CONTENTS (COnt) Section Title 2.3.4 Short-Term Diffusion Estimate 2.3-11 2.3.4.l Objective 2.3-ll 2.3.4.2 Accident Assessment 2.3-12 2.3.4.2.l Methodology 2.3-12 2.3.4.2.2 Meteorological Data 2.3-13 2.3.4.2.2.l Representativeness 2.3-13 2.3.4.2.2.2 Joint Frequency Distributions 2.3-13 2.3.4.3 Atmospheric Diffusion Model 2.3-14 2.3.4.4 Diffusion Eati.mates 2.3-14& 2.3.4.4.l Exclusion Area Boundary 2.3-l4a 2.3.4.4.2 Low Population Zone 2.3-14a 2.3.5 Long-Term Diffusion Estimate 2.3-14b 2.3.5.l Objective 2.3-14b 2.3.5.2 Calculations 2.3-l4b 2.3.6 References for Section 2.3 2.3-15 Appendix 2.3A Joint Frequency Distributions of Wind 2.3A-l Spead and Direction by Lapse Rate Delta Temperature Stability Classes: June 1969 to November 1971 2.4 HYDROLOGIC ENGINEERING 2.4-1 2.4.l Hydrologic Description 2.4-1 2.4.1.1 Site and Facilities 2.4-1 2.4.1.2 Hydrosphere 2.4-4 2.4.2 Floods 2.4-5 2.4.3 Probable Maximum Flood 2.4-6 2.4.J.l Probable Maximum Precipitation 2.4-6 2.4.4 Potential Dam Failures 2.4-7 2.4.5 Probable Maximum Surge and Seiche 2.4-7 Flooding 2.4.5.l Probable Maximum Winds and Associated Meteorological Parameters 2.4-7 2.4.5.2 surge and Seiche History 2.4-12 2.4.5.3 Surge and Seiche Sources 2.4-12 2.4.5.4 Wave Action 2.4-12 2.4.5.5 Resonance 2.4-13 2.4.5.6 Runup 2.4-13 2.4.5.7 Protective Structures 2.4-14 2.4.6 Probable Maximum Tsunami Flooding 2.4-14 2-iii SGS-UFSAR Revision 16 January 31, 1998

TABLE OF CONTENTS (Cont) Section 2.4.7 2.4.8 Ice Flooding Title Cooling Water Canals and Reservoirs 2.4-15 2.4-16 2.4.9 Channel Diversions 2.4-16 2.4.10 Flood Protection Requirements 2.4-16 2.4.11 Low Water Considerations 2.4-16 2.4.11.1 Low Flow in Rivers and Streams 2.4-16 2.4.11.2 Low Water Resulting from Surges, Seiches, and Tsunamis 2.4-16 2.4.11.3 Historical Low Water 2.4-19 2.4.11.4 Future Control 2.4-19 2.4.11.5 Plant Requirements 2.4-19 2.4.11.6 Heat Sink Dependability Requirements 2.4-19 2.4.12 Environmental Acceptance of Effluents 2.4-19 2.4.13 Groundwater 2.4-21 2.4.13.1 Description and Onsite Use 2.4-21 2.4.13.2 Sources 2.4-22 2.4.13.3 Accident Effects 2.4-26 2.4.13.4 Monitoring or Safeguard Requirements 2.4-27 2.4.13.5 Technical Specifications and Emergency Operation Requirements 2.4-27 2.4.14 References for Section 2.4 2.4-27 2.4.15 Bibliography for Section 2.4 2.4-28 2.5 GEOLOGY AND SEISMOLOGY 2.5-1 2.5.1 Basic Geologic and Seismic Information 2.5-1 2.5.1.1 Regional Geology 2.5-2 2.5.1.1.1 Physiography 2.5-2 2.5.1.1.2 History and Tectonics 2.5-3 2.5.1.1.3 Stratigraphy 2.5-5 2.5.1.1.4 Structure 2.5-6 2.5.1.1.5 Groundwater 2.5-8 2.5.1.2 Site Geology 2.5-9 2-iv SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS (Cont)

Section 2.5.2 2.5.2.1 Title Vibratory Ground Motion Geologic Conditions at Site 2.5-11 2.5-11 2.5.2.2 Tectonic Conditions 2.5-11 2.5.2.3 Behavior During Prior Earthquakes 2.5-12 2.5.2.4 Geotechnical Properties 2.5-13 2.5.2.5 Seismicity 2.5-13 2.5.2.6 Correlation of Epicenters with Geologic Structures 2.5-16 2.5.2.7 Identification of Active Faults 2.5-17 2.5.2.8 Description of Active Faults 2.5-17 2.5.2.9 Maximum Earthquake 2.5-17 2.5.2.10 Safe Shutdown Earthquake 2.5*18 2.5.2.11 Operating Basis Earthquake 2.5-19 2.5.2.12 Response Spectra 2.5-19 2.5.3 Surface Faulting 2.5*20 2.5.4 Stability of Subsurface Materials 2.5-20 2.5.5 Slope Stability 2.5-20 2.5.6 References for Section 2.5 2.5-20 2.5.7 Bibliography for Section 2.5 2.5-21
SGS-UFSAR 2-v Revision 6 February 15, 1987

LIST OF TABLES Table 2.1-1 Title New Jersey Population Projections to 2020 2.1-2 Population Projections Available for MCD's of Counties Within 50 Miles of Salem 2.1-3 Persons per Household Factors 2.1-4 Population Estimates of Cities and Towns Within 10 Miles of the Site 2.1-5 Resident Population Distribution by Zone and Sector, 0 to 10 Miles from SGS 2.1-6 Resident Population Distribution by Zone and Sector 10 to 50 Miles from SGS 2.1-7 Land Use in Five Surrounding Counties 2.1-8 Agricultural Statistics 2.1-9 Schools Located in EPZ by Emergency Planning Area 2.1-10 Recreational Facilities in SNGS Local Area 2.1-11 Health Care Facilities 2.1-12 Correctional Facilities/Jails 2.2-1 Industries within Ten Miles of the Site 2.2-2 Hazardous Chemicals Stored Onsite 2-vi SGS-UFSAR Revision 6 February 15, 1987

LIST OF TABLES (Cont) 2.2-3 DELETED 2.2-4 Estimates of Hazardous Chemical Traffic 2.3-1 Percentage of Daye with Various Hydrometers - Dover Delaware Air Force Base (1942-1965) 2.3-2 snowfall - Philadelphia International Airport 2.3-3 Snowfall - Trenton Airport 2.3-4 Distribution of Peak Winds - Philadelphia International Airport 2.3-5 Distribution of Hourly Temperatures - Temperature classes 2.3-6 Precipitation 2.3-7 Percentage of Hours with Fog 2.3-8 Percentage Frequency of Turbulence Classes 2.3-9 Percentage Frequency of Lapse Rates 2.3-10 Relation Between Lapse Rates and Turbulence Classes 2 .3-11 Average Horizontal Range 2.3-12 Average Horizontal Range for Wind Directions Between 130 and 160 Degrees 2.3-13 Percentage Frequency of Wind Speed Classes 2-vii SGS-UFSAR Revision 16 January 31, 1998

LIST OF TABLES (Cont) Table Title 2.3-14 Hean Annual Wind Speeds at Various Levels 2.3-15 Wind Data Recovery (June 1969 - Hay 1970) 2.3-16 Meteorological Instrumentation 2.3-16A Data Acquisition system Hardware 2.3-17 Vent Release - Exit Velocity of 7.2 H/Seconda - Undepleted X/Q at Ground Level Applicable to Long Term (Routine) Gaseous Releases - sector Annual X/Q at Ground Level 2.3-18 vent Release - Exit Velocity of 7.2 M/Seconds - Undepleted X/Q at Ground Level Applicable to Long Term (Routine) Gaseous Releases - sector Annual X/Q at Ground Level 2.3-19 Vent Release - Exit Velocity of 7.2 M/Seconds - Undepleted X/Q at Ground Level Applicable to Long Term (Routine) Gaseous Releases - Sector Annual X/Q at Ground Level 2.3-20 Vent Release - Exit Velocity of 7.2 M/Seconds - Undepleted X/Q at Ground Level Applicable to Long Term (Routine) Gaseous Releases - sector Annual X/Q at Ground Level 2.3-21 Accident X/Q Estimates 2.3-22 Accident X/Q values at LPZ by Sector 2.4-1 summary of Maximum Stillwater Elevation Determinations 2.4-2 Agencies and Individuals contacted 2-viii SGS-UFSAR Revision 16 January 31, 1998

LIST OF TABLES (Cont)

Table 2.4-3 Title Hydrologic Characteristics of Geologic Formations 2.4-4 Public Water Supplies 2.4-5 Private Water Wells in Vicinity of the Site 2.5-1 List of References - Agencies and Individuals Interviewed 2.5-2 Modified Mercalli Intensity (Damage) Scale of 1931 2.5-3 Significant Earthquakes within 100 Hiles of Salem, New Jersey
SGS-UFSAR 2-ix Revision 6 February 15, 1987

LIST OF FIGURES Figure 2.1-1 Title General Site Location 2.1-2 General Site Location Oto 60 Miles 2.1-3 Site Environs Detail 2.1-4 Aerial Photograph of Site 2.1-5 Area Plot Plan of Site 2.1-6 Resident Population Distribution (0 to 10 Miles) - 1970 and 1980 2.1-7 Resident Population Distribution (Oto 10 Miles) - 1970 and 1990 2.1-8 Resident Population Distribution (Oto 10 Hiles) - 1970 and 2000 2.1-9 Resident Population Distribution (0 to 10 Miles) - 1970 and 2010 2.1-10 Resident Population Distribution (Oto 10 Miles) - 1970 and 2020 2.1-11 Regional Resident Population Distribution (10 to 50 Miles) 1970 and 1980 2.1-12 Regional Resident Population Distribution (10 to 50 Miles) 1970 and 1990 SGS-UFSAR 2-x Revision 6 February 15, 1987

LIST OF FIGURES (Cont)

Figure 2.1-13 Title Regional Resident Population Distribution (10 to 50 Miles) 1970 and 2000 2.1-14 Regional Resident Population Distribution (10 to 50 Miles) 1970 and 2010 2.1-15 Regional Resident Population Distribution (10 to 50 Miles) 1970 and 2020 2.1-16 10 Mile EPZ Boundary 2.1-17 Evacuation Time Estimates

2. 1-18 Evacuation Time Estimates

2.2-1 2.3-1 Site Vicinity Map Showing Major Facilities Sources of Meteorological Records 2.3-2 Two Year Wind Rose - All Hours 2.3-3 Two Year Wind Rose - Only Hours with a Stabile Stability 2.3-4 Definition of Turbulence Classes 2.3-5 Diurnal Variation of Lapse Rate - June 1970 2.3-6 Diurnal Variation of Lapse Rate - December 1970 2.3-7 SNGS Meteorological Tower Schematic 2-xi SGS-UFSAR Revision 6 February 15, 1987

LIST OF FIGURES (Cont) Figure Title 2.3-8 Meteorological Data Acquisition Display System 2.4-1 Map of Area 2.4-2 Deleted: Refer to Plant Drawing 232091 2.4-3 Deleted: Refer to Plant Drawing 211612 2.4-4 Site Location in Relation to the Surrounding Area 2.4-5 Typical Rainfall Intensity-Duration-Frequency Curves for New Jersey Area 2.4-6 Map of Delaware Bay Showing Storm Track and Point of Calculations for Maximum Hurricane Surge 2.4-7 Deleted 2.4-8 Map of Delaware Bay Showing Location of PMH for Maximum Low Water Conditions 2.4-9 Yard-Fresh Water Well Locations 2.4-10 Public Water Supplies in Vicinity of Site

2. 4-11 Map of Area-Known Water Wells in New Jersey in Vicinity of Site 2.5-1 Regional Physiographic Map 2.5-2 Geological Section - Coastal Plain 2.5-3 Regional Geologic Map 2-xii SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure Title 2.5-4 Regional Tectonic Map 2.5-5 Geologic Columnar Section - Site Area 2.5-6 Plot Plan 2.5-7 Plot Plan - Detail A 2.5-8 Sub-Surface Sections 2.5-9 Columnar Section Showing Geophysical Data 2.5-10 Soil Profile and Material Properties at Salem Nuclear Generating Station Site

2. 5-11 Epicentral Location Map 2.5-12 Ground Response Spectra - Safe Shutdown Earthquake 2.5-13 Ground Response Spectra - Operating Basis Earthquake

SGS-UFSAR 2-xiii Revision 6 February 15, 1987

SECTION 2

SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGRAPHY 2.1.1 Site Location The Salem site is located on the southern part of Artificial Island on the east bank of the Delaware River in Lower Alloways Creek Township, Salem County, New Jersey. The point of intersection of the centerlines of the two Containment Buildings and the Auxiliary Building is located at Latitude 39° 27 min 46 sec north and Longitude 75° 32 min 08 sec west. The Universal Transverse Mercator coordinates of the reactor site are 4,368,100 m north and 454,070 m east, Zone 18. While called Artificial Island, the site is actually connected to the mainland of New Jersey by a strip of tideland formed by hydraulic fill from dredging operations on the Delaware River by the U. S. Army Corps of Engineers. The site is 15 miles south of the Delaware Memorial Bridge, 18 mil~s south of Wilmington, Delaware, 30 miles southwest of Philadelphia, Pennsylvania, and 7-1/2 miles southwest of Salem, New Jersey. The location of the site with respect to major cities in the northeast is shown on Figures 2.1-1 and 2.1-2.

Salem Generating Station (SGS) is located on a 700-acre site which is owned by Public Service Electric & Gas (PSE&G). Access to the site is achieved by a road (constructed by PSE&G) that connects with Alloways Creek Neck Road, about 2-1/2 miles east of the site. The location of the site with respect to the surrounding area is shown on Figure 2.1-3, a U. S. Geological Survey map. An aerial photograph is presented on Figure 2.1-4 .

SGS-UFSAR 2.1-1 Revision 6 February 15, 1987

2.1.2 Site Description The location of the site boundary and significant plant features is shown on Figure 2.1-5. The site exclusion area is defined as follows: Land The land exclusion area is defined as that area bounded by the property line as shown on Figure 2.1-5. This land is owned by and under the control of PSE&G. The minimum distance between the reactors and the exclusion area boundary (property line) is 1270 meters. Water The water portion of the exclusion area is defined as that area bounded by the locus of points 1270 meters from the Containment Buildings of either Units 1 or 2 and also falling within the Delaware River. The 1270 meters is consistent with the minimum land exclusion area distance. Discussion of the exclusion of people, property, and river traffic from that portion of the exclusion area which extends over the river is included as part of the detailed Emergency Plan, Section 13.3. 2.1.2.1 Exclusion Area Control PSE&G owns and has control of the 700-acre land area that comprises the exclusion area. Control of the water portion of the exclusion area is described in Section 13.3, Emergency Plan. SGS-UFSAR 2.1-2 Revision 6 February 15, 1987

2.1.2.2 Boundaries for Establishing Effluent Release Limits

The line land boundary on which technical drawing defining land owned by PSE&G.

specification limits on release of gaseous radioactive effluents is based on the property Figure 2. 1-5 is a scale showing the property line in relation to the reactor units. Distances from both the Units 1 and 2 vents to the property line in any direction may be scaled from this drawing. The minimum distance from the vents to the property line is nominally 1270 meters. 2.1.3 Population Distribution The sources used for the 1970 distribution of population were the U. S. Bureau of the Census counts for 1970 (1), census and topographic maps (2,3), aerial photos (4), and field check surveys. The Bureau of the Census published various levels of population data: county, minor civil division (MCD), census tract, and block. In 1970, the study area (within 50 miles of the Salem site), included portions of 4 states, 24 counties, 338 MCDs, hundreds of tracts, and thousands of census blocks. Population distribution about SGS is provided in the sector format required. Concentric circles with the required radii for distances to 10 miles and IO to 50 miles are provided. The circles are then divided into 22.5 degree segments, each centered on one of the 16 cardinal compass points (e.g., north, north-northeast, etc.). Projections on IO-year intervals to the year 2020 are provided on the above described sector format. Population projections are based on 1970 census data. Projections based on current census data will be provided after the validity of the current projection assumption and calculational techniques have been analyzed .

SGS-UFSAR 2.1-3 Revision 6 February 15, 1987

2.1.3.1 Population Within 10 Miles For the 96 sectors within 10 miles of the site, the following method of distribution was used. It was felt, and subsequently proven, that the "area" method for distribution, successfully used beyond 10 miles depends on large sector area. Within 10 miles, the sector size is significantly smaller. For example, an average size MCD beyond 10 miles falls into two sectors. Within 10 miles, a similar-sized MCD would include over 5 sectors. The assumption of evenly distributed population within a MCD is only valid beyond 10 miles. This is clearly seen in the area within Smiles of the site which is mostly marsh. An even distribution of people throughout a sector would place residents in uninhabitable areas. Beyond 10 miles, this factor is not of great concern since habitable land is also included in the larger sectors. To arrive at an accurate portrayal of the 1970 population distribution within 10 miles, a house count was made from topographic maps. The count was field checked within 5 miles. The total house count for a MCD or census tract was divided into the 1970 census population resulting in persons per household factor of 2.99 to 3.36 as shown in Table 2.1-1. The factors were then applied to the houses in their respective MCDs. By totaling the population in each sector, the 1970 distribution was derived. The results are shown on Figures 2.1~6 through 2.1-10. One problem encountered was the lack of updated maps in some sectors. A review of township population growth showed that in the areas not covered by the 1970 photo-revised topographic maps (concurrent with the census data), the township growth was minimal. Any growth was in areas already populated and beyond 5 miles of the site. The field check proved this to be true. The house count method assumed that growth to 1970 is proportional to development already mapped. To be consistent, house counts were made for all MCDs which were partially within 10 miles but which extended beyond 10 miles. This dot-map distribution method is more precise than the area 2.1-4 SGS-UFSAR Revision 6 February 15, 1987

distribution method which assumes equal population density throughout the MCD. However, as noted above, house counts were only necessary within 10 miles where the sector sizes required greater precision. The great amount of swampy and marshy areas found around the site is not populated. The proximity of the Delaware River is also responsible for the low density within 5 and 10 miles. The population densities for land area only were 29 people per square mile and 109 people per square mile within 5 and 10 miles respectively, in 1970. The nearest residence is approximately 3.4 miles west-northwest of the site in Bay View Beach, Delaware. Other nearby residences are located 3.5 miles east-northeast, and 3.7 miles northwest of the site. 2.1.3.1.1 Population Projections for Oto 10 Mile Area The methodology for population projection is described in Section 2.1.3.7. The allocation of 1970 population to the rose within 10 miles was based on house distribution. Projected population distribution is assumed to be similar. Within 5 miles, a land survey field check was made. The area is marsh and meadowland, not suitable in its present form for residential development. Thus, the areas presently developed are assumed to be the focal points of further development. Figures 2.1-6 through 2.1-10 show the estimated future population distributions. 2.1.3.1.2 Population Update Within 10 Miles Updated population is provided for the most current estimate of population within 10 miles. The updated population is provided in Tables 2.1-2 and 2.1-3. The population estimates are essentially current (March 1980), although they are based on U.S. Bureau of Census figures of July 1, 1977 and Dresdner Associates' surveys (5) conducted in September 1979 and March 1980. There has been little population 2.1-5 SGS-UFSAR Revision 6 February 15, 1987

change by sector since July 1, 1977. Changes that have occurred tend to emphasize the conservative (high) nature of the population estimate, including decreased family size, out-migration in older communities, and decline in new housing starts. The methodology of allocating and reporting population is consistent with that already described. Distribution of Population, 0 to 5 Miles, New Jersey The distribution of population in New Jersey within Oto 5 miles of the SGS is based on a comprehensive land use survey of dwelling units factored by an estimated average household size.

1. The O to 5 mile area from the SGS was divided into 35 sector/zones based on the standard Nuclear Regulatory Commission (NRC) designators for population distribution maps. The sectors located in New Jersey are north, north-northeast, northeast, east-northeast, east, east-southeast, and southeast in the Oto 5 mile area.

2. A survey of land uses within the O to 5 mile area identified all residential units by zone and sector in New Jersey. The reasonableness of this survey was confirmed by sample counts from aerial photos, the USGS maps, and municipal master plans.

3. Based on the 1970 average household size by community, the population of each sector was determined by multiplying the number of dwelling units in the sector by the average household size of the community in the sector. Where more than one community was within a sector, the average household size of the community with the largest population was assumed to be reasonable.

The resultant figure is considered conservative (a high estimate) because all literature indicates that average household size has decreased since 1970. 2.1-6 SGS-UFSAR Revision 6 February 15, 1987

Distribution of Population, 0 to 5 Miles, Delaware The distribution of population within Oto 5 miles in Delaware of the SGS is based on small area (sub-municipal) population estimates made by the Wilmington Metropolitan Area Planning Council (WILMAPCO).

1. The O to 5 mile area from the SGS was divided into 45 sector/zones based on standard NRC designators for population distribution maps. The sectors located in Delaware are south-southeast, south, south-southwest, southwest, west-southwest, west, west-northwest, northwest, and north-northwest.

2. The entire portion of the Oto 5 mile area in Delaware is included in the WILMAPCO Parcel Land Use System (PLUS). This program presented small area, sub-municipal (cells) population data for the year 1976, and estimated current for 1980 .

3. Each small area, or cell, was assigned to a zone/sector.

Where a cell was located in more than one zone/sector, its proportional area was allocated to each cell.

4. The population of each cell was then proportionately distributed to each zone/sector in the Oto 5 mile ring in *Delaware. This proportional distribution was based on the assumption that population is generally evenly distributed throughout the cell. This distribution was validated by a "windshield" survey, examination of aerial photos, and review of USGS maps. On the basis of this validation, transfers of population from one sector/zone to another (but within the same cell) were undertaken to account for grossly unequal distributions of population within a cell. A typical example would be where the wetlands portion of a cell was located in one zone/sector, and the built up portion located in an 2.1-7 SGS-UFSAR Revision 6 February 15, 1987

adjacent zone/sector. In such a case, it was assumed that population was concentrated in the developable section of the cell. Distribution of Population, 5 to 10 Miles, New Jersey The allocation of population within 5 to 10 miles of the SGS was based on a count of dwelling units except for the City of Salem, New Jersey, where the population was based on a Census update. I. The 5 to 10 mile area from the SGS was divided into 35 sector/zones based on the standard NRC designators for population distribution maps. The sectors located in New Jersey are north, north-northeast, northeast, east-northeast, east, east-southeast, and southeast in the 5 to 10 mile area, with zones at one mile intervals.

2. A survey of land uses within the 5 to 10 mile area identified all residential units (outside of boroughs and cities) by zone and sector, except for the City of Salem.

3. The population of the City of Salem was taken from the Census update (see Sources) and allocated to the appropriate sector based on its aerial distribution.

4. Based on the 1970 average household size by community, the population of each sector was determined by multiplying the number of dwelling units in the sector by the average household size of the community in the sector. Where more than one community was within the sector, then the average household size of the community with the largest population was assumed to be reasonable. The resultant figure was considered conservative (a high estimate) because all literature indicates that average household size has decreased since 1970.

2.1-8 SGS-UFSAR Revision 6 February 15, 1987

Distribution of Population, 5 to 10 Miles, Delaware The distribution of population in Delaware within 5 to 10 miles of SGS is based on small area, sub-municipal population estimates made by WILMAPC0.

1. The 5 to 10 mile area from the SGS was divided into 45 sector/zones based on standard NRC designators for population distribution maps. The sectors located in Delaware are south-southeast, south, south-southwest, southwest, west-southwest, west, west-northwest, northwest, and north-northwest.

2. The entire portion of the 5 to 10 mile area in Delaware is included in WILMAPCO' s PLUS program, except for a small section of Kent County. This program presented small area, sub-municipal (cells) population data for the year 1976 and estimated current for 1980 .

3. Each small area, or cell, was assigned to a zone/sector.

Where a cell was located in more than one zone/sector, its proportional area was allocated to each cell.

4. The population of each cell was then proportionately distributed to each zone/sector in the 5 to 10 mile ring in Delaware. This proportional distribution was based on the assumption that population is generally evenly distributed throughout the cell. This distribution was validated by a "windshield" survey, examination of aerial photos, and review of USGS maps. On the basis of this validation, transfer of population from one sector/zone to another (but within the same cell) were undertaken to account for grossly unequal distributions of population within a cell. A typical example would be where the wetlands portion of a cell was located in one zone/sector, and the built up portion located in an adj a cent zone/ sector. In such a case, it was assumed 2.1-9 SGS-UFSAR Revision 6 February 15, 1987

that population was concentrated in the developable section of the cell. 2.1.3.2 Population Between 10 and 50 Miles This area from 10 to 50 miles is divided into 64 sectors ranging in size from 59 square miles to 177 miles. The great majority of MCDs are divided between two sectors. For this reason, the MCD was chosen as the unit to be studied from 10 to 50 miles. Only in Philadelphia County, which is one MCD, were census tracts used. This was due to the size of Philadelphia which falls within four sectors and partially beyond the 50 mile radius circle. Census tracts more accurately portray the 1970 distribution in urban areas as they are smaller in size than MCDs. However, for most areas beyond 10 miles, they were not available. In many of the rural areas. census tracts are contiguous with MCDs. The 1970 population data on the MCD level was allocated to the sectors assuming equal distribution throughout the sector. This percentage of each MCD within a sector was calculated. This percentage was multiplied by the MCD population to obtain the population in the sector. The procedure was repeated for all land areas within a sector. The sum of these computations for each sector yielded its 1970 population. The results are shown on Figures 2.1-11 through 2.1-15. 2.1.3.2.1 Population Projections for 10 to 50 Mile Area The population derived from the MCDs, as discussed above, were allocated to the rose in the same manner as the 1970 populations. The results are shown on Figures 2.1-11 through 2 .1-15. The methodology for these projections is described in Section 2.1.3.7. 2.1.3.2.2 Population Update 10 to 50 Miles Updated population is provided for the most current estimate of population for the area 10 to 50 miles from SGS. This population 2.1-10 SGS-UFSAR Revision 6 February 15, 1987

distribution is tabulated in Table 2 .1-4. The population estimates are essentially current (March 1980), although they are based on U.S. Bureau of Census figures of July 1, 19 77 and Dresdner Associates' surveys (5) conducted in September 1979 and March 1980. There has been little population change by sector since July 1, 1977. Changes that have occurred tend to emphasize the conservative (high) nature of the population estimate, including decreased family size, out-migration in older communities and decline in new housing starts. The methodology of allocating and reporting population is consistent with that already described. Distribution of Population, 10 to 50 Miles, New Jersey The distribution of population within 10 to 50 miles in New Jersey of SNGS is based on updated Bureau of Census reports.

1. The 10 to 50 mile area from SGS in New Jersey was divided into 46 sector/zones. The sectors located in New Jersey are north, north-northeast, northeast, east-northeast, east, east-southeast, and southeast.

2. The population for the entire portion of the 10 to 50 mile area in New Jersey is included in the Bureau of Census, P-25 series, Report No. 843. This report, entitled "Population Estimates and Projections,"

contains current estimates of the July 1977 population for all counties, incorporated places, and active MCDs.

3. Each municipality was assigned a zone/sector. Where a municipality was located in more than one sector> a proportional area was allocated to each one.

4. The population of each sector/zone was based on the percentage of aerial distribution, assuming equal distribution of population through the municipality.

2.1-11 SGS-UFSAR Revision 6 February 15, 1987

5. Equal distribution of population throughout the municipality was assumed excluding wildlife refuges, state parks, coastal wetlands, and marshlands.

6. Total population distribution by sector.

Distribution of Population, 10 to 50 Miles, Pennsylvania The distribution of population from 10 to 50 miles from SGS in Pennsylvania was determined by Census Bureau update reports.

1. The 10 to 50 mile area from SGS in Pennsylvania was divided into 30 sector/zones. The sectors in Pennsylvania are north, north-northeast, northeast, west-northwest, northwest, and north-northwest, and fall into the 20 to 50 mile zones.

2. The population for the entire portion of the 20 to 50 mile area in Pennsylvania is included in the Bureau of Census, P-25 series, Report No. 851. This report, entitled "Population Estimates and Projections,"

contains current estimates of July 1977 populations for all counties, incorporated places, and active minor civil divisions.

3. Each municipality was assigned a zone/sector. Where a municipality was located in more than one sector or zone, a proportional area was allocated to each.

4. The population of each zone/sector was based on the percentage of aerial distribution, assuming equal population throughout the municipality.

5. Equal distribution of the population throughout the municipality was assumed excluding wildlife refuges, state parks, coastal wetlands, and marshlands.

2.1-12 SGS-UFSAR Revision 6 February 15, 1987

6. Total population distribution by sector .

Distribution of Population, 10 to 50 Miles, Delaware The distribution of population with 10 to 50 miles in Delaware of SGS is based on updated Bureau of Census reports.

1. The 10 to 50 mile area from SGS in Delaware was divided into 43 sector/zones. The sectors are north, north-northwest, northwest, west-northwest, west, west-southwest, southwest, and south-southwest.

2. The population estimates for this area are available from the Bureau of the Census, P-25 series, Report No.

821. This report, entitled "Population Estimates and Projections, 11 contains current estimates of the July 1977 population for all counties, incorporated areas, and active MCDs. Much of the Delaware and Maryland populations remain unincorporated, meaning that this portion of the populace is represented in the county total only. To determine the number of unincorporated people per county, the total incorporated population was subtracted from the county total. This portion of the population was then equally allocated, based on a percentage of developed land area for each sector/zone.

1. Each governmental unit was assigned a sector/zone.

Where a governmental unit was located in more than one, a proportional area was allocated to each sector or zone.

2. The population of each sector/zone was based on the percentage of aerial distribution for incorporated and unincorporated areas. Equal population distribution was assumed for each .

2.1-13 SGS-UFSAR Revision 6 February 15, 1987

Distribution of Population, 10 to SO Miles, Maryland The distribution of population within 10 to 50 miles in Maryland of SGS is based on updated Bureau of Census reports.

1. The 10 to 50 mile area from SGS in Maryland was divided into 41 sector/zones. The sectors are northwest, west-northwest, west, west-southwest, southwest, and south-southwest.

2. The population estimates for this area are available from the Bureau of Census, P-25 Series, Report No. 833.

This report contains current population estimates for July 1977 for all counties, incorporated areas, and active MCDs. Much of the Maryland population remains unincorporated and, therefore, is represented only in the county totals. To determine the number of unincorporated people per county, the total incorporated population was subtracted from the county total. This portion of the population was then equally allocated, based on a percentage of developed land area for each sector/zone.

1. Each governmental unit was assigned a sector/zone.

Where a governmental unit was located in more than one, a proportional area was allocated to each sector or zone.

2. The population of each sector/zone was based on the percentage of aerial distribution for incorporated and unincorporated areas. Equal population distribution was assumed for each.

2.1.3.3 Low Population Zone The radius of the low population zone (defined in 10CFRlOO) is 5 miles. This distance is based on plant design and protective 2.1-14 SGS-UFSAR Revision 6 February 15, 1987

action considerations. The update population (1980) for the area within the Smile low population zone is 1298 persons

2.1.3.4 Transient Population Within 5 miles of the sitet there are no major seasonal or daily additions to the population with the exception of the Salem Station and Hope Creek Station construction and outage support crews and onsite visitor's center. The center has a seating capacity of 140 persons. The area is marsh and meadowland which attracts only limited numbers of hunters and trappers.

A list of the transient population attracted by the recreational facilities around the site is provided in Table 2.1-5. Pleasure craft are used on the Delaware River and Alloways Creek. Prime usage occurs on weekends and holidays. The boats range from 14 feet to 35 feet in length and might accommodate an average maximum of 120 passengers .

The only other major source of transients in the vicinity is the Delaware River traffic. Annual passenger traffic according to the U.S. Corp of Engineers is over four million people (6). This number seems high and might include double counting at the various ports north of the site. It should be stressed that river traffic does not remain within 5 miles of the site vicinity longer than the time required to traverse the river, normally less than 1 hour.

2.1.3.5 Population Center The nearest population center of about 25,000 (as defined in 10CFRlOO) is Wilmington, Delaware, 18 miles north of the site. The 1970 population of Wilmington is listed in the U.S. Census report as 80,386, a decrease of 16 percent from the 1960 U.S. Census report population of 95,287. Bridgeton, New Jersey, 15.5 miles east of the site, is listed in the U.S. Census report as 2.1-15 SGS-UFSAR Revision 6 February 15, 1987

having a 1970 population of 20,453, a decrease of 2.5 percent from the 1960 U.S. Census report. Wilmington is the center of a Standard Metropolitan Statistical Area (SMSA). The Wilmington SMSA has a population in excess of 300,000. Philadelphia, Pennsylvania, and Camden, New Jersey, are part of the SMSA with a population in excess of 3. 5 million, beginning about 30 miles north-northeast of the site. Baltimore, Maryland, with a population of less than 1 million is located 50 miles west of the site. The City of Salem, located 8 miles north-northeast of the site, had a 1970 population of 7648. 2.1.3.6 Public Facilities and Institutions An area of approximately 10 miles radius (slightly larger and irregular) has been defined as the Emergency Planning Area (EPZ) for the Salem site. The EPZ area, as defined in NUREG-0654, Rev. 1, dated November 1980, obtains the irregular shape by virtue of being defined by political and physical boundaries. This area is slightly larger than a IO-mile radius, a description of which is provided on Figure 2.1-16. All information related to special facilities, including public facilities and institutions, are those facilities which reside in this area. Additional information with respect to the facilities and related transient population is provided in the Salem Generating Station Emergency Plan and in references contained in Reference 28 of this plan. Total transient population and special facilities population is provided on Figures 2.1-17 and 2.1-18. 2.1.3.6.1 Schools There are a total of 24 schools in this area. The schools located closest to the site are Lower Alloways Creek Township School, located 6.5 miles east with a total population (students and instructors) of 285, and the Corbit School (185 persons) located 2.1-16 SGS-UFSAR Revision 6 February 15, 1987

6.5 miles west in Odessa, Delaware. A listing of the schools is provided in Table 2.1-6 and identified on Figure 2.1-16 . 2.1.3.6.2 Hospitals and Nursing Homes There are two hospitals and one nursing home located within the 10 mile EPZ. The Salem County Memorial Hospital is a public facility located in Salem, New Jersey, 10 miles north-northeast of the site. It has a bed capacity of 168. There is also a daytime facility (Association of Retarded Citizens in New Jersey) with an attendance of 80 persons. The Governor Bacon Health Center is located near Delaware City, Delaware and is 8.5 miles north-northwest of the Salem site. It is operated by the State Division of Mental Health and Retardation primarily for emotionally and mentally ill children. Present capacity is 222 patients with a daytime staff of 66. Salem Nursing Center, 8 miles north-northeast of the site, has a capacity of 110 patients. Table 2.1-7 lists the hospital and nursing homes with current patient and staff population. 2.1.3.6.3 Correctional Institutions There are two correctional institutions within or very near 10 miles of the site. The nearest institution is the Salem County Jail located 8 miles north-northeast of the site with a capacity of 115 inmates and an average of 75 inmates. The Delaware State Correctional Institution has a total capacity of 775 inmates. Inmate average population as of the beginning of 1981 was 900 inmates. The institution is located in Smyrna, Delaware, 12 miles south-southwest of the Salem Site. Table 2.1-11 lists the correctional institutions

2.1-17 SGS-UFSAR Revision 6 Febraary 15, 1987

2.1.3.6.4 Recreational Facilities Recreational facilities which include State Parks, wildlife refuge areas and boating access areas are tabulated in Table 2.1-5. major recreational facilities populations are Fort Delaware, with the largest The transient 9 miles north-northwest, with a peak summer day attendance of 1200 persons, and Fort Mott State Park, 9. 5 miles north, with an annual attendance for the same period of 500 persons. The boating access areas are Augustine Beach and Woodland Beach, located 5 miles west*northwest and 9.8 miles south-southeast, respectively. Both of these access areas are heavily used between April 1 and September 30; however no attendance statistics are available. Five wildlife refuges are within 10 miles of the site. Artificial Island Wildlife Area is the closest, as it adjoins the site. The northern part of the island and the marshes connecting the island to the mainland are owned by the U.S. Government. Some of this area is leased to hunting and fishing clubs. Adjacent property extending for 3 miles south on the New Jersey Coast is owned by the State of New Jersey and operated as a fish and game preserve for limited use by sportsmen. The closest attraction, although not strictly a recreational facility, is the site Visitor's Center with a seating capacity of 140 persons. 2.1.3.7 Population Projection Methodology This section describes the procedures used to project the population of the year 2020 and to allocate it to the rose format. It also describes exceptions and their impact. The basis for population projection shown on Figures 2.1-6 through 2.1-15 is a form of cohort-survival analysis. Utilizing data projected on a national level, projections are based on proportions or shares at the MCD level. This step-down method is a systematic approach (7) relying on three assumptions. These are that historic trends of birth, death, and migration will continue. 2.1-18 SGS-UFSAR Revision 6 February 15, 1987

The Bureau of Census (8,9) formulates projections for the nation to 2020 and for the states to 1990. They project for a range of fertility rates: 3.35 (A) through 2.11 (E); and a choice of migration patterns: the same as presently observed (I), no migration (III) and a mixture (II). The A and B fertility rates have been declared unrealistically high and as of ¥972, only C, D, and E rates are used in Federal projections. The migration patterns projected are I and III. For consistent conservatism, the projections for Salem reflect a C fertility rate, or 2. 78 children per woman, and both I and II I migration rates. The numbers shown on Ffgures 2 .1-6 through 2 .1-10 reflect the IC and IIIC projections. In the step down method, the change in proportion is all important. Thus, the state projections were extended from 1990 to 2020 by calculating the change in share of state to nation from 1940 to 1990. The change from 1980 to 1990 was considered characteristic and was reapplied to determine the state population in 2000, 2010, and 2020. An example is shown in Table 2. 1-8. The proportion method was carried down to the county level using projections from state or regional planning commissions (10-14). Although the numbers were discarded, the ratio of county to state population was retained. The change in proportion was applied to the federally projected percentage for the state to yield a projected county population. It was felt that the state or regional planning staffs were cognizant of the areas of growth within their region, but that the absolute number might not be reliable. Thus the total county population of 24 counties was derived from the IC and IIIC Federal/state populations. In the same manner, the MCD populations were calculated. This time, however, the data was only partially complete. Many rural planning boards have not made projections for their counties . 2.1-19 SGS-UFSAR Revision 6 February 15, 1987

Other commissions have made only limited or short range projections. Although this last is the most realistic and sensible approach, the projections had to be e:xtended to 2020. Accordingly, the county data was reviewed and categorized based on type of projection available. Table 2.1-9 lists the counties and the categories. The basis for classification is discussed below:

1. Near-complete projections - The Delaware Valley Regional Planning Commission (DVRPC) has jurisdiction over seven of the counties in this study. Its projections at the MCD level are by decade to 2000 (10). York and Salem Counties have planning commissions that project MCD populations by decade to 2010 (15~16). These near-complete projections were used to determine projections to 2020.

As with the step down from state to countyt this study utilized the proportions but not the absolute numbers projected. Again it was felt that local agencies have a grasp of where growth will occur within their regions, but the local policies of boosterism or isolationism will bias the numbers. Thust using the proportions of growth for an MCD, and the projected county population, the absolute MCD population, by decade to 2020, was calculated.

2. Limited projections - Six counties involved were placed in this category because planning boards had made one or possibly two projections for the next 45 years (17-22).

These were not by decade, rather at 15 or 20 year intervals. To utilize these projections, a ratio was made to determine the proportion of the county represented by an MCD at each 10 year interval. This ratio was then extended to 2020. 2.1-20 SGS-UFSAR Revision 6 February 15, 1987

3. No reliable projections - Nine counties were placed in this division. As shown in Table 2.1-9, four of the counties could be projected on the basis of historic trends. Changes in proportions of MCDs to the county for 10 year periods since 1940 were averaged for each MCD. The result was applied to the 1970 MCD proportion to arrive at the 1980 proportion, and so on to the year 2020. Using the county absolute projections, the MCD future populations were derived.

The MCD populations in the other five counties where no reliable projections were available were calculated assuming future population distribution similar to the 1970 distribution. They include Philadelphia City/County which is partially within four sectors and is one MCD. Using census tract data for the city (23), the 1970 population was derived for each sector. The same proportion of census tract to MCD/County was used for future populations. This means that city-wide growth over the next 50 years is assumed to occur in proportions similar to the present population. To determine a more reliable projection \would require a detailed study of the area. However, the 2020 city population total would be the same; only the distribution within the city would alter. Philadelphia is beyond 30 miles of the site and any distribution effect on the sector totals is minimal. Other areas where the 1970 proportions were used throughout the 50 years also fall at the outer edges of the study area and are not divided into many sectors. Sussex County, Delaware, has been restricted since the 1960 census; thus historic trends could not be utilized. Projections for Baltimore County were made to 1985, but were based on 1960 census data and are not reliable. Cecil and Queen Anne's Counties, Maryland, are rural in 2.1-21 SGS-UFSAR Revision 6 February 15, 1987

nature, and the planning boards have not made projections. 2.1.4 Use of Adjacent Land The site is located in the southern region of the Delaware River Valley, which is defined as the area immediately adjacent to the Delaware River and extending from Trenton to Cape May Point, New Jersey on the eastern side; and from Morrisville, Pennsylvania, to Lewes, Delaware, on the western side. The northern region is one of the major commercial, industrial, and residential centers of the nation. Much of the land area is highly industrialized or residential. The southern region is characterized by extensive tidal marshlands and low-lying meadowlands. The major portion of the land in this area is undeveloped. The site, located 15 miles south of the Delaware Memorial Bridge, is isolated from the industrial and population centers of Philadelphia, Wilmington, Camden, and the New York-Washington corridor in general. The Chamber Works, at Deepwater, New Jersey, and at the Carneys Point Works, at Penns Grove, New Jersey, of E. I. DuPont deNemours Company are the southern-most major industrial activities from the Delaware Memorial Bridge, with the exception of the Getty Oil Company refinery across the river near Delaware City, Delaware, about 9 miles north-northwest of the site. The area within a 25 mile radius of the site encompasses the major portion of 5 counties: Cecil in Maryland, New Castle and Kent in Delaware, and Salem and Cumberland in New Jersey. A summary of land use in these counties is presented in Table 2.1-10. As shown in the table, developed urban land constitutes only a small fraction of the available land - about 10 perent on the average for the 5 counties. The remaining 90 percent is used for agriculture (44 percent) or is undeveloped (46 percent). Agriculture statistics are summarized in Table 2.1-12. Crops primarily consist of fruit (apples and peaches), vegetables (snap 2.1-22 SGS-UFSAR Revision 6 February 15, 1987

beans, sweet corn, peppers, and tomatoes), and animal feed products. The Hope Creek Generating Station is located north-northwest of Salem Units 1 and 2. The Hope Creek Generating Station construction area is contiguous to the Salem site (Figure 2.1-5). The remainder is covered with marsh grasses. A strip of land about 1 mile wide to the east of the site extending from Alloways Creek to Hope Creek is owned by the United States Government, and consists entirely of tidal marshes. Most of the diked meadow areas have reverted to tidelands and are not in use. Beyond 3 miles, the land is sufficiently elevated to permit farming and grazing. The Delaware side of the river is similar to the New Jersey site, except that the tidelands and marshes are not as extensive. A great deal of land adjacent to the river on both sides is public land (Federal and state owned), or land planned for future open space projects. In addition, industrial, commercial, or residential growth is limited by recent wetlands and New Jersey Coastal Area Facilities Review Act legislation. Industrial water supplies are obtained directly from the river above the site. Another source of process water is derived from high capacity wells drilled into the excellent aquifers located close to the river which are subsequently recharged by the Delaware River. No industrial installations are located along the river below the site. Because of salt water intrusions, industrial use of the river water below Marcus Hook, some 25 miles upstream of the site, is limited to cooling water applications. Thus radioactive wastes discharged to the river will remain well downstream of any industrial or domestic usage of river water. Potable water supplies in Salem County, New Jersey, are obtained primarily from ground water with some inland areas utilizing surface water sources. All municipalities near the site use deep wells with the exception of the city of Salem, New Jersey, which 2.1-23 SGS-UFSAR Revision 25 October 26, 2010

obtains about two-thirds of its water from surface water supplies from Quinton which is on Alloways Creek about 8 miles northeast of the site. This water supply is a dammed fresh water stream approximately 9 miles upstream of the Delaware River - Alloways Creek confluence. The closest domestic well is a shallow well located about 3 miles from the site. The Delaware Estuary is being studied by a group of Marine Ecologists (Ichthyological Associates). Over 74,000 specimens of 45 species of fish that with environmental were taken in 1,094 trawl hauls. The most prevalent fish species that account for 98.7 percent of the total trawl catch are the bay anchovy, weakfish, white perch, hog choker, alewife, spot, striped bass, blueback herring, and the silver perch. The drifted gill nets revealed that the anadromous American shad tend to migrate to the area west of Reedy Island Dike. During May and June, the greatest catches were in the eastern section of the estuary, and, in September and October, the western section of the estuary was predominant. The largest quantities of specimens and species from both day and night collections were collected in August. It was noted in July 1970 that a large fish kill had occurred somewhere up river. Many thousands of dead fish drifted into the study area after a period of heavy rain and resultant flooding in the watershed area. In most instances, death was attributed to a dissolved oxygen content below the minimum required to sustain fish life. This was caused by dilution of the river with the ground runoff from heavy rainfall. Fishing in the Delaware River Estuary has been reduced markedly since 1900 due to river pollution with only 554,000 pounds (valued at $65,000) landed in 1966. Landings in the Delaware River estuary were comprised of shad, striped bass, white perch, sturgeon, and crab with the latter accounting for 75 percent of the total poundage. In the Delaware Bay at the mouth of the estuary, 2.2 million pounds were landed in 1966 valued at $875,000. Oysters accounted for about one-third of that total with the remaining species including weakfish, shad, striped bass, 2.1-24 SGS-UFSAR Revision 6 February 15, 1987

white perch, and crab. No increase in these values is expected until such time as major water pollution problems are brought under control. The nearest oyster beds are located approximately 4 miles downstream of the Salem site on the New Jersey side of the river. A comparison of current trends (circa 1972) can be made with respect to the 1966 figures from information for landings in Delaware and New Jersey from the Delaware Bay and the Estuary. The total fish catch was 585,600 pounds valued at $99,387; hard crabs landed were 1,245,700 pounds valued at $201,975; and oysters were 814,300 pounds with a value of $588,234 for an overall total of 2,245,600 pounds and a value of $889,596. This compares with the 1966 figures totaled at 2,754,000 pounds with a total value of

 $940,000, and represents no significant difference.

2.1.4.1 Recreational Land Use A description of parks and recreational land use is provided in Section 2.1.3.6.4. The recreational facilities within the 10 mile area around the site are listed in Table 2.1-5. This table lists the recreation areas, populations, and relative position with respect to the site. The location is indicated by compass heading and average distance in miles. 2.1.5 References for Section 2.1

1. U.S. Bureau of Census, 1971, U.S. Census of Population, 1970, "Number of Inhabitants" Final Report: PC (1) A9 Delaware; PC (2) A23 Maryland; PC (1) A32 New Jersey; PC (1) A40 Pennsylvania.

2. U.S. Bureau of Census, 1970, Civil Division Maps for:

Delaware, Maryland, New Jersey, and Pennsylvania

3. U.S. Department of the Interior, Geological Survey Topographic Maps. 7.5-Minute Series: Ben Davis Point, 2.1-25 SGS-UFSAR Revision 6 February 15, 1987

N.J.-Del., 1956; Bennetts Pier, Del., 1956; Bombay Hook, Del.-N.J., 1956; Bridgeton Quad, N.J., 1953; Canton, N.J.-Del., 1948; Cecilton, Del-Md., 1958; Cecilton, Del.-Md., 1944; Claiborne, Md., 1942; Clayton, Del., 1955; Delaware City, Del.-N.J. 1948, 1970; Langford Creek, Md., 1954; Little Creek, Del. (Kent County), 1956; Middletown, Md., 1953; Middletown, Del., 1953; Newark East, Del., 1953, 1970; Saint Georges, Del., 1953; Salem, N.J., (Salem County), 1948, 1970; Smyrna, Del., 1956; Taylors Bridge, Del.-N.J., 1948; and Wilmington South, Del.-N.J., 1967

4. U.S. Department of Agriculture, Aerial photo coverage for area within 5 miles of Salem site, scale of 1 inch to 660 feet, New Jersey portion flown 9/70 and Delaware portion, 5/68.

5. Dresdner Associates, "Distribution of Population within 50 Miles of the Salem Nuclear Generating Station," 1980.

6. 7. U.S. Corps of Engineers, "Waterborne Commerce of the United States," 1970, Part 1: Atlantic Coast, 1971. Communication, Dames & Moore and Michael R. Greenberg, PhD. and Donald A. Drueckeberg, PhD., Associate Professors, Department of Urban Planning and Policy Development, Rutgers University, New Brunswick, N.J.

8. U.S. Bureau of Census, Current Population Reports, Series P-25, No. 470, "Projections of the Population of the United States, by Age and Sex: 1970 to 2020, 11 U.S. Government Printing Office, 1971.

9. U.S. Bureau of Census, Current Population Reports, Series P-25, No. 477, "Preliminary Projections of the Population of States: 1975-1990,u U.S. Government Printing Office, 1972.

SGS-UFSAR 2.1-26 Revision 6 February 15, 1987

10. Delaware Valley Regional Planning Commission, "Preliminary Population Forecasts to the Year 2000," Philadelphia, Pa, 1971.

11. State Planning Board, "Preliminary Employment and Population," Harrisburg, Pa. 1971.

12. N.J. Department of Labor and Industry, "Preliminary Population Projections," Trenton, N.J., 1971.

13. Maryland Department of State Planning, "Preliminary Maryland Population Projections/' 1980-2000, Baltimore, Md., 1971.

14. Delaware State Planning Office, "Final Population Projections," Dover, Del., 1972.

15. York County Planning Commission, 11 Population York County,"

York, Pa., 1971.

16. Salem County Planning Board, "Revised Population Estimates,"

Salem, N.J., 1971.

17. Kent County Regional Planning Commission, "The Comprehensive Plan," Dover, Del., 1971.

18. New Castle County Department of Planning, Population Estimates, Wilmington, Del., 1971.

19. Communication, Dames & Moore and David Cartes, County Administrator, Caroline Co., Md.

20. Harford County Planning and Zoning Commission, "The Comprehensive Plan, 11 Bel Air, Md., 1969.

21. Planning Commission, "The Comprehensive Plan," Kent County, Md., 1968

2.1-27 SGS-UFSAR Revision 6 February 15, 1987

22. Lancaster County Planning Commission, "Sketch Plan, n Vol. 1, Lancaster, Pa., 1970.

23. Philadelphia City Planning Commission, "Population and Housing Statistics for Philadelphia Census Tracts, 1970 Census, 11 Philadelphia, Pa., 1971.

24. Parsons; Brinckerhoff, Quade and Douglas, Inc. , "Evacuation Time Estimates for the Areas Near the Site of Salem and Hope Creek Generating Stations," 1981.

SGS-UFSAR 2.1-28 Revision 6 February 15, 1987

TABLE 2.1-1 NEW JERSEY POPULATION PROJECTIONS TO 2020 NJ as U.S. NJ Proportion Change in Year Population Population of U.S. Proportion 1910 92,228 2,537 0.0275 1920 106,022 3,156 0.0298 +0.0023 1930 123,203 4,041 0.0328 +0.0030 1940 132,165 4,160 0.0315 -0.0013 1950 151,326 4,835 0.0320 +0.0005 1960 179,323 6,067 0.0338 +0.0018 1970 204,800 7,168 0.0350 +0.0012 IC Projections 1980 233,798 8,518 0.0364 +0.0014 1990 269,673 10,152 0.0376 +0.0012 2000 305,111 11,838 0.0388 +0.0012 2010 349,746 13,990 0.0400 +0.0012 2020 397,164 16,363 0.0412 +0.0012 IIIC Projections 1980 233,798 8,144 0.0348 -0.0002 1990 269,673 9,281 0.0344 -0.0004 2000 305,111 10,374 0.0340 -0.0004 2010 349,746 11,751 0.0336 -0.0004 2020 397,164 13,186 0.0332 -0.0004

SGS-UFSAR 1 of 2 Revision 6 February 15, 1987

TABLE 2.1-1 (Cont)

NOTE: 1) Populations shown in 1000 1.s

2) Area under dashed lines represents projections made by continuing the change in proportion in 1990 to 2020 and working left to calculate the state population.

SOURCES: U.S. Bureau of Census, 1971, "Current Population Reports," Series P-25, No. 470. U.S. Bureau of Census, 1972, "Current Population Reports,n Series P-25, No. 477

SGS-UFSAR 2 of 2 Revision 6 February 15, 1987

TABLE 2.1-2 POPULATION PROJECTIONS AVAILABLE FOR MCD'S OF COUNTIES WITHIN 50 MILES OF SALEM No Reliable New Complete Limited Projections State/County Projections(l) Projections(2) Historic(3) 1970(4) Delaware Kent 1990 New Castle 1985 Sussex

Maryland Baltimore County Caroline 2015 Cecil Harford 1985 Kent 1985 Queen Anne's
Talbot *
New Jersey Atlantic Burlington Camden Cape May Cumberland 2000 2000 2000 Gloucester 2000 Salem 2010 Pennsylvania ~'(

Berks 2000 Chester 2000 Delaware 2000 Lancaster 1985,2010 Montgomery Philadelphia 2000 York 2010 NOTES: (I) Projections by decade to date listed (2) Projections only for date listed (3) Historic trends used (4) 1970 proportions of MCDs to Counties used due to redistricting, lack of data, or rural nature of county. 1 of I SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-3 PERSONS PER HOUSEHOLD FACTORS State/County Persons Per Household County Subdivisions Delaware Kent Smyrna East 3. 16 New Castle Middletown - Odessa 3.32 Red Lion 3.27 New Jersey

Cumberland Greenwich Stow Creek Salem 3.16 3.27 Elsinboro 2.85 Lower Alloways Creek 3.28 Mannington 3.18 Pennsville 3.18 Quinton 3.36 Salem 2.99 (1)

(1) 1970 Salem population is within one sector .

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.1-:-4 POPULATION ESTIMATES OF CITIES AND TOWN WITHIN 10 MILES OF THE SITE Population Distance and Town 1970 Location from Site Delaware Bay View Beach 168 3.4 WNW Delaware City 2024 7.5 NNW Middletown 2644 9.5 W Odessa 547 6.5 W Port Penn 369 4.2 NNW St. Georges 358 9.0 NW Townsend 505 9.5 WSW Woodland Beach 100 9.7 SSE New Jersey (1) Canton 350 6.5 E Hancock's Bridge 358 5.0 NE Oakwood Beach 295 6.0 N Quinton 750 8.5 NE (1) 1972 Estimates 1 of 2 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-4 (Cont) Distance and Nearest Location from Major Site City PoEulation Projections to 2020 (Miles) 1970 1980 1990 2000 2010 2020 Salem City 7648 8266 8280 8271 8706 8564 8.0 NNE

SGS-UFSAR 2 of 2 Revision 6 February 15> 1987

TABLE 2.1-5 RESIDENT POPULATION DISTRIBUTION BY ZONE AND SECTOR, 0 TO 10 MILES FROM SGS Sector 0-1 (1) *1-2 (2) 2-3 (3) 3-4 (4) 4-5 (5) 5-6 (6) 6-7 (7) 7-8 (8) 8-9 (9) 9-10 (10) N (A) 0 0 0 0 0 201 200 0 0 44 NNE (B) 0 0 0 0 33 120 120 1,609 5,832 161 NE (C) 0 0 0 9 295 252 285 316 459 366 ENE (D) 0 0 0 35 96 233 255 204 173 153 E (E) 0 0 0 0 0 135 385 153 117 108 ESE (F) 0 0 0 0 0 0 90 42 48 348 SE (G) 0 0 0 0 0 0 0 0 16 50 SSE (H) 0 0 0 0 0 0 0 0 0 82 s (J) 0 0 0 0 16 24 35 16 45 42 SSW (K) 0 0 0 16 11 37 63 74 86 111 SW (L) 0 0 0 0 15 97 104 193 222 291 WSW (M) 0 0 0 0 31 228 259 305 350 381 w (N) 0 0 0 22 54 46 599 61 70 2,721 WNW (P) 0 0 0 113 68 51 102 87 118 154 NW (Q) 0 0 0 108 104 65 101 74 446 110 NNW (R) 0 0 0 10 262 31 35 1,079 1,084 TOTAL 0 0 0 313 985 1,520 2,633 4,213 9,066 5,122 CUM. TOTAL 1,298 22,524 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-6 RESIDENT POPULATION DISTRIBUTION BY ZONE AND SECTOR, 10 TO 50 MILES FROM SGS Sector 10-15 (15) 15-20 (20) 20-25 (25) 25-30 (30) 30-35 (35) 35-40 (40) 40-45 (45) 45-50 (50) N (A) 14,022 67,373 62,976 24,054 43,830 62,400 52,158 69,424 NNE (B) 5,061 11,357 11,972 114,952 422,449 466,085 627,404 431,323 NE (C) 2,013 5,810 3,105 42,450 110,894 350,663 297,887 206,137 ENE (D) 1,776 4,566 4,797 21,646 23,348 21,342 20,832 8,545 E (E) 2,773 19,416 .8,943 48,006 22,550 7,172 . 4,081 13,546 ESE (F) 2,799 9,683 5,014 7,366 11,053 1,417 7,469 12,266 SE (G) 42 17 94 0 0 0 2,356 24,237 SSE (H) 0 40 28 446 374 582 1,470 3,011 s (J) 2,137 7,740 23,781 5,792 9,086 16,173 8,602 9,278 SSW (K) 13,881 3,723 5,276 5,455 5,419 6,998 8,940 5,798 SW (L) 15,072 4,933 3,098 3,820 4,015 4,060 3,877 1,855 WSW (M) 14,232 2,156 2,084 2,797 6,030 1,700 4,018 7,062 w (N) 11,566 4,100 3,284 1,500 26,751 14,448 33,440 75,132 WNW (P) 11,689 4,237 10,492 8,114 16,266 14,454 15,636 16,028 NW (Q) 16,443 23,502 11,736 6,428 12,007 5,460 9,141 17,028 NNW (R) 15,620 27,787 32,941 19,202 9,151 44,094 14,758 13,581 TOTAL 129,084 196,423 189,621 312,028 723,223 1,017,048 1,112,069 914,419 CUM. TOTAL 325,507 501,649 1,740,271 2,026,488 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-7 LAND USE IN FIVE SURROUNDING COUNTIES New Jersey Salem County(l) Cumberland Countr(2) Acres i Acres Developed, urban 16,200 7.4 30,000 9.3 Agriculture 122,000 55.5 80,000 24.9 Undevelo;eed 81,400 37.1 2112500 65.8 Total 219,600 100.0 321,500 100.0 Delaware 3 New Castle County( ) Kent County(3) Acres % Acres % Developed, urban 52,300 18.8 20,500 5.4 Agriculture 94,650 34.0 170,600 45.0 UndeveloEed 131 2 385 47.2 188 2 100 49.6 Total 278,335 100.0 379,200 100.0 Maryland Cecil County(4) Acres % Developed, urban 16,200 7.1 Agriculture 137,000 60.3 Undevelo12ed 73 2 840 32.6 Total 227,040 100.0 Source: (1) Salem County Land Use, 196 7, Salem County Planning Board. (2) Land Use - 1964, Cumberland County Planning Board. (3) Delaware State Development Department - 1964 Delaware State Agriculture Department - 1964 (4) Cecil County Economic Inventory, Maryland Department of Economic Development - 1964 .

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.1-8 AGRICULTURAL STATISTICS (1) County Salem Cumberland New Castle Kent Cecil State N.J. N.J. Del. Del. Md. Total Farm Acreage 122,000 80,000 94)650 170,600 137,000 Percent of County 55.5 24.9 34.0 45.0 60.3 in Farms Number of Farms 796 1035 564 1219 659 Average Farm Size 139 90 215 190 194 (acres) Average Value of Farms $39,000 $37,000 $116,000 $56,000 $84,000 Farm Population 2529 3564 2155 4551 2431 Major Farm Products Dairy Products Poultry Dairy Products Field Crops Dairy Products Vegetables Vegetables Field Crops Dairy Products Field Crops Poultry Poultry Value of Farm Products Sold (millions of dollars) Crops 8.7 17 .o 6.0 10.0 2.4 Livestock and Livestock Products 7.3 8.0 4.0 7.0 5.5 Total 16.0 25.0 10.0 17 .o 7.9 (1) Statistics taken from 11 County and City Data Book," 1967; U.S. Department of Commerce, and Rand.McNally "Commercial Atlas and Marketing Guide," 1967. 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-9 SCHOOLS LOCATED IN EPZ BY EMERGENCY PLANNING AREA ( 1) Name/ Facility School & Location Enrollment & Staff District Lower Alloways Creek School 255 30 Salem County, SD Lower Alloways Creek, N.J. Morris Goodwin School 120 29 Cumberland County, SD Greenwich, N.J. Woodland County Day School 125 25 Stow Creek, N.J. I Stow Creek School 196 27 Cumberland County, SD Stow Creek, N.J. I John Fenwick School 590 35 Salem County, SD Salem City, N.J. Elsinboro School 121 10 Salem County, SD Elsinboro, N.J. Dunnington School 195 21 Salem County, SD Dunnington, N.J. Quinton School 358 26 Salem County, SD Quinton, N.J. Salem Day Care Center 20 8 Salem County, N.J. (1) Reference 24 1 of 3 SGS-UFSAR Revision 25 October 26, 2010

TABLE 2.1-9 (Cont) Name/ Facility School & Location Enrollment & Staff District St. Mary's School 237 11 Non-public Salem City, N.J. Votech Center Complex 203 9 Salem County, SD Mannington,. N.J. Salem Middle School 424 40 Salem County, SD Salem City, N.J. Salem High School 995 68 Salem County, SD Salem City, N.J . Townsend Elementary School 224 29 Appoquinimink, SD Townsend, Delaware St. Andrew!s School 230 140 Non-public Middletown, Delaware (residential) Silver Lake Elementary School 661 56 Appoquinimink, SD Middletown, Delaware Redding Middle School 529 66 Appoquinimink, SD Middletown, Delaware Broadmeadow School 200 40 Non-public Middletown, Delaware Corbit School 142 13 Appoquinimink, SD Odessa, Delaware 2 of 3 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-9 (Cont) Name/ Facility School & Location Enrollment & Staff District Middletown High School 797 79 Appoquinimink, SD Middletown, Delaware Au Clair School 32 29 Non-public School for St. Georges, Delaware (residential) Autistic Children Commodore McDonough School 210 17 New Castle County, St. Georges, Delaware SD-Area 4 Gunning Bedford Middle School 1020 99 New Castle County, St. Georges, Delaware SD-Area 4

Qelaware City School 169 18 New Castle County, Delaware City, Delaware SD-Area 4

SGS-UFSAR 3 of 3 Revision 6 February 15, 1987

TABLE 2.1-10 RECREATIONAL FACILITIES IN SNGS LOCAL AREA Visitors Visitors Peak Peak Day Normal Day Area Name and Address Season Peak Seasons Peak Season Segment Artificial Island Waterfowl 10-15 N 3 Delaware, New Jersey Season Mad Horse Tract Waterfowl 200-250 200 NNE 2,3 Lower Alloways Creek, NJ Season NE 2,3,4 ENE 2,3

                                                                       *E 2-7 ESE   2-8 SE 6,7,8

Maskell's Mill Pond Lower Alloways Creek, NJ Killcohook National Summer Fall 12 25 ENE 8 Outside Wildlife Refuge Pennsville, NJ Woodland Beach Wildlife Fall 100 Hunters SSE 8, 9, 10 Refuge, Delaware S 3-7 (including the area north Swnmer 200-300 200 SW 3 along the shore) Fishermen Fishermen WSW 3 Woodland Beach, Delaware Swnmer 500 SSE 8, 9, 10 Augustine Creek Wildlife Fall 100 WMW 4, 5, 6 Refuge, Delaware NW4, 5, 6 1 of 3 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-10 (Cont) Visitors Visitors Peak Peak Day Normal Day Area Name and Address Season Peak Seasons Peak Season Segment Augustine Beach Swnmer 75 Fishermen NW 4 in boats NNW 4,5 Chesapeake & Delaware Fall 400 NNW 7, 8 Canal, Delaware NW 8, 9, 10 Appoquinimink Wildlife Fall 6 WSW 3, 4 Refuge, Delaware Fort Mott State Park Summer 500 44 N 10 Pennsville, NJ Weekend Fort Delaware Swnmer 1200 310 N 9 Pea Patch Island, NNW 9 Delaware Marlboro Marina Summer 100 marina NNE 9 Salem City NJ 100 in boats Cohansey Marina & Casino Spring/ 100 marina 35 Outside Greenwich, N.J Swnmer 250 casino Hancock Harbor Summer 200 in boats Outside Greenwich, NJ Weekend 300 in restaurant Delaware City Marina Swnmer 100 25 NNW 9 Delaware City, Delaware

SGS-UFSAR 2 of 3 Revision 6 February 15, 1987

TABLE 2.1-10 (Cont) Visitors Visitors Peak Peak Day Normal Day Area Name and Address Season Peak Seasons Peak Season Segment Meadow View Acres Summer 80 E 8, 9 Campground, NJ Visitor's Center Salem 200 Center NGS Holly Mountain Ski Area Winter 225 100 E 9, 10 Weekend

SGS-UFSAR 3 of 3 Revision 6 February 15, 1987
TABLE 2. 1-11 HEALTH CARE FACILITIES(!)

Residents/ Wheel Staff Name/Location Patients Ambulatory Chair Stretcher Day Evening Night Salem Memorial Hospital 142 60 35 47 255 49 29 Mannington, N.J. Salem Nursing and 110 20 85 5 40 17 7 Convalescent Center Mannington, N.J. Association of Retarded 80 (daily) 15 Citizens of Salem County Mannington, N.J. Governor Bacon Health 222 161 46 15 194 66 40 Delaware City, Delaware (1) Reference 24 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.1-12 CORRECTIONAL FACILITIES/JAILS (1) Name/Location Inmates Staff Capacity Average Day Evening Night Salem County Jail 115 75 NA Salem City, New Jersey Delaware Correctional 775 900 175 40 30 Center, New Castle County Delaware (1) Reference 24

SGS-UFSAR 1 of I Revision 6 February 15, 1987

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR REVISION 6 FEBRUARY 15, 1987 General Site Location Figure 2.1-1
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION 6 FEBRUARY 15, 1987 General Site Location 0-60 Miles Updated FSAR Figure 2.1-2
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION 6 FEBRUARY 15, 1987 Site Environs Detail Updated FSAR Figure 2.1-3

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                  ., :i, REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY                Aerial Photograph of Site SALEM NUCLEAR GENERATING STATION Updated FSAR                              Figure 2.1-4

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Area Plot Plan of Site REVISION 6 FEBRUARY 15, 1987 Updated FSAR Figure 2.1-5

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DETAIL "A" ST.ITU'!!: MILE 0 I 2 !I KB Y:

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1970 RESIDENT POPULATION

191!0 IC RESIDENT POPULATION 1,947
1980 IIIC RESIDENT POPULATION RADIUS IN MILES YEAR 0* 1 0*2 0*3 1970 0 0 0

ACCUMULATED 1980- 0 0 0 IC POPULATION 1930- 0 0 me 0 RADIUS IN MILES YEAR O** 0*5 o- 10 1970 :m 2177 251111 ACCUMULATED

                                                                       !980*

IC 367 11111.i 291188 POPULATION 1980* IIIC Jit6 1331 '>OMO R E F £ R [ N C t: THIS ll,UIP Wl'S P~EPARt:0 rAOIII A PORTlON or THE rOL'"" lOIUNG U*S*G*S.* au,.p! WILl*UNQ10Nt DE'LAWA~f, l'JI:£. REVISIONS FEBRUARY 15, 1987 Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY 0-10 Miles SALEM NUCLEAR GENERATING STATION 1970 and 1980 Updated FSAR Figure 2.1*6

I J l r.

0 ST.miff MILE I 2 KEY: 1970 ltBSIDENT POPUL.ATION

~-

I, 729

1990 IC RESIDENT POPUL.ATION 2, 328 " 1990 IIIC RESIDENT POPUL.ATION O*I 0-2 0-3 RADIUS IN MILES YEAR 1970 0 0 0

                                       .I.CCUMUL.ATED 1990-           0            0                 0 IC POPUL.ATION         1990-           0            0                0 me YEAR         0-4         0*5               0

IO RADIUS IN MIL.ES 1970

                                                                              ,:s         1177               2511M ACCUMUL.A TED 199010        665         2551              4111117 POPUL.A TION         199~;...      602          2}12              37130
                                               ~tF"EA[NCE!

nus ~p *AS PR[P ....RED FROM A PORt1.0N OF l'Ht F"OL* LOWtNO u~s .. a,,g., ~9! *iLIUNG10N, 0[l.AWARE'.t 1'966 .. REVISION I FEBRUARY 15, 1187 Resident Population Dlttrlbutlon PUBLIC SERVICE ELECTRIC AND GAS COMPANY 0-10 Mile1 SALEM NUCLEAR GENERATING STATION 1970 and 1990 Updated FSAR

I l DETAIL "A" 0 STATUTE I MILE 2 KEY; 729 1970 RESIDENT POPULATION 2,328 = 2000 IC RESIDENT POPUi,ATION 1,947 " 2000 IIIC RESIDENT POPULATION RADIUS IN MILES YEAR O* I 0* 2 0*3 1970 0 0 0 ACCUMULATED 2000- 0 0 0 IC POPULATION 2000-IIIC 0 0 0 RADIUS IN MILES YEAR 0*i 0*S 0

10 ACCUMULATED 1~70 :m ll77 251111 2000- 893 3294 49962 IC POPULATION 2000-lllC 778 2876 ljJ/jl3 RCFCAENC[:

THIS IMP WAS PREP~RE'O tROM .A PORTION OF THE FOL-

                                       !.OWIMi U*5*G-S*  M.',P! WIJ,.NING'fON, 0ELAWA1=1£,  1()66-.

REVISION 8 FEBRUARY 15, 1987 Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY 0-10Miles 1970 and 2000 SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 2.1-8

I I DETAIL .,.. STATUTE MILE 0 ,I 2 3 KEY: I, 729

1970 RESIDENT POPULATION 2.:1!.!.,
2010 IC RESIDENT POPULATION l,947
2010 IIIC RESIDENT POPULATION RADIUS IN MILES YEAR 0* I o-z 0-3 1970 0 0 0

ACCUMULATED 2010-IC 0 0 0 POPULATION 2010-

                                                                     !IIC        0            0                0 RADIUS IN MILES          YEAR         0-4          0*5             0 - JO 19,0         JB          ll77             ~

ACCUMULATED 2010;-,... 1207 4300 62101 POPULATION 2010-IIIC 1020 3624 IM:t'>'I! FIEF'£11ENCE:: ttHS JilAP WA-S Pftt:PI\R[O FAOM A PORTION Of' Tt-4( FOL*

                                                \,.OWINO u~s-a .. $. YAP~   *h.MIHGTONt OtLAIIARE,   1qee..

REVISION 6 FEBRUARY 15, 1987 Resident Populatlon Distribution PUBLIC SERVICE ELECTRIC ANO GAS COMPANY 0-10 Miles SALEM NUCLEAR GENERATING STATION 1970 and 2010 Updated FSAR Fig11re 2.1-9

N K JfL a:: 665 w 2! 0: DETAIL "A"

                                              ---*  aa**

STATUTE IIIILE 0 I 2 J KEY: 729

I 970 RESIDENT POPULATION

2020 IC Rl!SIDBNT POPULATION I, 9~7 " 2020 IIIC RESID!!NT POPULATION RADIUS IN MILES YEAR 0* l 0*2 0*3 1970 0 0 0 ACCUMULATED 2020* 0 0 0 IC POPULATION zo20-IIIC 0 0 0 RADIUS IN MILES YEAR O*f 0*5 0
10 ACCUMULATED 1970 :m ll77 251114 2020;-r 1600 5622 76617 POPULATION 2020-IIIC 1306 4579 62275 REFtAENtE:

nus )Mp WAS PR£PAltE.O F'JtOU ,,._ PORTION OF lH£ FOL-LOWING U,S*G*S* IAA,p; ;r11.N1NG'rON, OE'LA~~RE, 1%6~ REVISION 6 FEBRUARY 15, 1987 Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY 0-10 Miles SALEM NUCLEAR GENERATING STATION 1970 and 2020 Updated FSAR Figure 2.1-10

                                                               !mTUT'E MILE 10 Ii          i              10 I

2.0 KBY: 1, 630 1970 R~IDENT PORJLATION

                                          ..!:..il2..            1980 IC RESIDENT FOPULA TION 2,140                1980 me RF.SIDENT FOruLA TION

REFERENCE:

RADIUS IN MILBS YEAR 0

20 0
30 0
40 0
50 THIS MAP WAS PREPARED FROM PORTIONS OF ntE FOLLOWINO 1970 378,*589 M0,159 2,603,598 4,744,551 SECT I ONI\L I\ERONI\UT I CAL ACCUMULATED CHART: NEW YORK* 1980- ,,. -415, 169 1,058,119 2,976, 4_78 5,366,006 POPULATlON 1980- 449,887 l, 012,.549 2,979, 952 5,389, 121 REVISION&

1nr FEBRUARY 15, 1987 Regional Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY 10-50 Miles SALEM NUCLEAR GENERATING STATION 1970 and 1980 Updated FSAR Figure 2;1-11

SW.TUTE MILE 10 o* 10 20 KEY: l, 630 1970 RESIDENT POR.ILAnoN

                                           ~                     1990 IC RESIDENT !OPULATION 2,140               1990 me RESIDENT EOIULATION R E F E R E N C El YEAR RADIUS IN MILES                  0 - 2()        0 - 30       0

40 0- 50 THIS MAP WAS PREPAREO FROM PORT! ONS OF THE FOLLOW I NG 1970 3 78, 589 860,159 2,603,598 4,744,551 SECTIONAL AERONAUTICAL A CCU MULA TED CHART; Nl:W YORK, 1990- 595,272 l, 290,071 3,477, 743 6,139,181 lC POPULATION 1990 - 538,736 1,197,383 3, 394, 807 6,074, 100.

REVISIONS me FEBRUARY 15, 1987 Regional Resident Population Distribution PUBLIC SERVICE ELECTRIC ANO GAS COMPANY 10-SOMiles SALEM NUCLEAR GENERATING STATION 1970 and 1990 Updated FSAR Figure 2.1-12

STATUTE MILE 10 0 10 20 KEY: l, 630 1970 RESIDSNT FOPULATION 2,960 2000 IC RESIDENT l'OPULATION 2, 140 2000 me RESIDENT FOPULA TION R E F E R E N C E! o- 30 o- so RADIUS IN MILES YEAR 0 - 20 0 - 40 THIS MAP WAS PREPARED f"ROM PORTIONS OF THE FOU.OWING 1970 378, 589 660,159 2,603,598 4,744,551 SECTIONAL AEAONAUT I CAL ACCUMULATED ClfARl: NEW YORK* 2000- 695,463 1,572,220 3,990,381 6,923,869

                                                  .IC POPULATION     2000-   604,449      I, 367,489   3, 769, 655  6,673,401   REVISION 8 me FEBRUARY 15, 1987 Regional Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY                                              10-50 Miles 1970 and 2000 SALEM NUCLEAR GENERArlNG STATION Updated FSAR                                     Figure 2.1-13

ST4TUT£ MILE 10 0 10 20 KBY: 1,630 1970 RESlDBNT OORJLATION 2,960 2010 JC RESIDENT FOl'ULA TION 2,140 2010 me RBSIOBNT IOIUI..A TION R E F" E R E N C E: RADIUS lN MIL2S Y'SAR o- 20 0- 30 0

40 o- 50 THIS f,IAP WAS PREPARED !'ROI.I PORTI CNS OF TH&" FOLLOW! Ml 1970 378. 589 -660, 1S9 2,603,598 4,744,551 SECTIONAi. AERONAUTICAi.

ACCUMULATBD CHART: NEIii YORI(, 201oic 82S, 722 1,834,465 4,573,681 ,. 864,519 POPULATION 2010~" 697,225 1,569,586 4,249, illt ,, 442, 726 REVISION 8 Ill FEBRUARY 16, 1987 Region.al Resident Population Distribution PUBLIC SERVlCE ELECTRIC AND GAS COMPANY 10-50 Miles SALEM NUCLEAR GENERATING STATION 1970 and 2010 Updated FSAR Figure 2.1-14

STATUTE MILE 10 0 10 20 KEY: 1, 630 1970 RESIDENT !'OPULATION

                                         ~                2020 IC RESIDENT FOPULA TION 2, 140         2020  me RESIDENT POPULATION 0- 50 RADIUS IN MILES   YEAR    0 - :20        O

30 0
40 THIS MAP WAS PREP.-REO F'ROM PORTIONS OF THE F'OLLOWI NG 1970 378,589 1360,159 2,603,598 4,744, 5S1 SECTIONAL AERONAUTICAL ACCUMULATED CHART: NEW YOitt<* 2020- 962,447 2,180,329 5,261,69S 8,924,121 lC POPULATION 2020- REVISION 6 793,595 1,839,747 4, 739, 673 8,209,288 rue FEBRUARY 15, 1987 Regional Resident Population Distribution PUBLIC SERVICE ELECTRIC AND GAS COMPANY 10-50 Miles 1970 and 2020 SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 2, 1*16

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REVISION 8 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY 10 Mlle EPZ Boundary SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 2.1-16

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Population Summary Transient Population Note: 338Sn29 at Salem and Hope Creek Gener.ating Stations REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Evacuation Time Estimates SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 2.1*17

Population Summary

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Evacuation Time Estimates REVISION 6 FEBRUARY 15, 1987 Updated FSAR Figure 2.1-18

2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES The Salem site is located in a rural area consisting of marshes, abandoned meadowland, and some farmland. There are no major manufacturing or chemical plants within 5 miles of the site. All such facilities are beyond 8 miles and would not interfere with the normal operation of the Salem Generating Station (SGS). Due to the lack of plants within 5 miles, our study was extended to 10 miles in order to present an accurate description of the site vicinity. The Delaware River, a major transportation route, represents the only possible hazard to the Salem Station due to the Intracoastal Waterway which passes through the River 1.5 miles west of the site. The freight traffic on the river is described in Section 2.2.3. All features described in this section are shown on Figure 2.1-16. 2.2.1 Location and Routes The location of manufacturing plants, chemical plants, storage facilities, and transportation routes (land and water) and pipe lines are provided on Figure 2.2-1 and listed in Table 2.2-l. 2.2.2 Descriptions 2.2.2.1 Missile Bases or Missile Sites There are no military bases or missile sites within 10 miles of the site. The nearest such facility is the Dover Air Force Base, 20 miles south-southwest, which is capable of handling the C-SA jwnbo jet transport plane. The base has a population of 8,200, which is expected to increase to IO ,000 persons when in full operation (I). Greater Wilmington Airport, 21 miles north-northwest, serves as a station for a combat helicopter squadron and for a C-130 heavy transport wing (2). 2.2-1 SGS-UFSAR Revision 6 February 15, 1987

2.2.2.2 Manufacturing Plants There are no manufacturing plants within 5 miles of the site. There are 11 manufacturing plants within 10 miles of the site producing a variety of goods from canned corn to felt base floor coverings (3,4). The nearest company, Gioia Speciality Foods, Inc., employs less than 25 people and produces canned beans. It is located 6.5 miles west (5). For detailed information about manufacturing plants see Table 2.2-1 and Figure 2.2-1. 2.2.2.3 Chemical Plants and Storage Facilities There are eight chemical companies in a cluster located 10. 0 to 10.5 miles north-northwest. The nearest large operation is the Getty Oil Company Refinery. For more details see Table 2.2-1 and Figure 2.2-1. 2.2.2.4 Oil and Gas Pipelines and Tank Farms There are two pipelines within 10.5 miles of the site, the nearest of which is the Getty Oil Company pipeline which runs within 9 miles of the site. The pipeline serves a large tank farm of approximately 50 tanks. The tank farm is located 10 miles north-northwest of the site. A second pipeline runs from the Pioneer Chloromane plant 2000 feet into the Delaware River at a point 10. 4 miles from the site. These pipelines are shown on Figure 2.2-1. 2.2.2.5 Transportation Complexes (Harbors, Railway Yards, Airports) There are no major harbors, railway yards, or airports within 10 miles of the site. The only "harbor facility 0 of any significance is the Getty Oil Company pipeline terminal in Delaware City, 9 miles north-northwest, used by moderate-size tankers (15 to 30,000 tons). 2.2-2 SGS-UFSAR Revision 6 February 15, 1987

Although there are no harbors located within 10 miles of the site, small craft fishing and pleasure boating is popular in the area. There are two boating access areas within 10 miles, the nearest being Augustine Beach access area, 3.7 miles northwest. Woodland Beach is located 10 miles south-southeast. Al though no boating figures are readily available, it was observed that both areas are used heavily for fishing and pleasure boating during the warmer months and probably for hunting during fall. A list of the Marinas is included in the list of Recreational Facilities in Table 2.1-10. There are no railroads within 5 miles of the site. However, there is a single track serving the chemical complex 10 miles north-northwest. Forty railroad cars were counted within the complex using aerial photos taken in 1968. There are three turf airstrips within 10.5 miles of the site, the nearest of which is Salem Airport, 8 miles north-northeast. 2.2.2.6 Transportation Routes (Highways, Railway, and Waterways) All transportation routes within 10 miles of the site are shown on Figure 2.2-1. The only major route within 5 miles of the site is the Intracoastal Waterway, 1.5 miles west of the island. The Waterway is the main route for barge and freighter traffic from the Atlantic to the Philadelphia Area ports. In 1970, at least 4,700 vessel trips were made past Artificial Island.*

Total traffic (9,858 trips) on the Chesapeake and Delaware Canal, 8 miles north-northwest was subtracted from the total traffic on the Delaware River (14,565 trips) to determine how many vessels passed Artificial Island (6).

2.2-3 SGS-UFSAR Revision 6 February 15, 1987

According to U.S. Corps of Engineer Statistics (6) over 4.5 million passengers traveled from Philadelphia to the sea in 1970. This does not include the 25,000 who traversed the Chesapeake and Delaware Canal. The 4.5 million seems unrealistic and may result from double counting passengers on vessels that made several stops above Salem. The method used in compiling the statistics does not allow for double counting errors. The Delaware River Hydrographic Chart delineates an anchorage zone northwest of Artificial Island. This zone is only for anchorage of vessels carrying explosives. According to Mr. Charles Ide and Lt. Edward Kangeter of the Safety Division, Coast Guard Station at Gloucester City, New Jersey, only one vessel since 1970 has carried explosives up-river past the site. The port of Philadelphia does not accept explosive cargo; all such cargoes, in limited quantities only, must be unloaded at Wilmington, Delaware. It was Lt. Kangeter's opinion that very few vessels with explosive cargoes have passed, or will pass, Artificial Island. He added that, with construction of the Salem Station, the Coast Guard is petitioning for a relocation of the anchorage area. This might increase anchorage by non-hazardous vessels, but Mr. Ide believes this is unlikely due to the distance to port. U.S. Highway 13, 6.5 miles west, is traveled by an average total of 14,560 vehicles daily (7). Trail ways Bus Service runs nine routes daily on this road, and one bus daily through Middletown, 10 miles west on Route 896. Six buses per day run to Salem, New Jersey, 8 miles north-northeast along Route 49 (8). There are no figures for rail traffic within 10 miles of the site, but rail lines in both Delaware and New Jersey handle mainly freight traffic. No commuter train traffic exists in the site vicinity. An in depth analysis of the highway network including local roads is provided in Attachment Il-1 of the Salem Generating Station 2.2-4 SGS-UFSAR Revision 6 February 15, 1987

Emergency Plan. Included in this attachment is the analysis of evacuation times as required by NUREG-0654, FEMA-REP-1: Rev. 1. 2.2.2.7 Petroleum Wells, Mines, or Quarries There are no petroleum wells, mines, or hardrock quarries within 10 miles of the site. The nearest quarrying of any kind is sand and gravel pit activity along Route 49 in Quinton, New Jersey, 9 miles northeast to 10 miles east of the site. 2.2.3 Evaluations 2.2.3.1 Barge Transportation There is no known movement of high explosives in the vicinity of Artificial Island*, the site of SGS. There is barge movement of flammable materials, such as jet fuel and gasoline, but these movements are not likely to pose much of a hazard to the station, since the probability of a runaway barge carrying flammable material striking the intake structure is very remote (i.e.,

            -7 around 1.0      per year).

A quantitative probabilistic analysis indicates that the risk of a runaway barge containing flammable material hitting the intake at Salem and igniting is 5.0 x 10- 9 occurrences per year. Details of the analysis are given below.

This fact has been verified by searching the records kept by the U.S. Army Corps of Engineers and the Bureau of Customs. In addition, the U.S. Coast Guard Offices in Philadelphia and New York also confirmed this fact. Finally, there are no known industrial or military activities in that region which would warrant shipment of high explosives.

2.2-5 SGS-UFSAR Revision 6 February 15, 1987

According to the U.S. Corps of Engineers (9), compilation of traffic between Philadelphia and the sea, there are approximately 550 barge movements of loaded barges past the Salem site per year (1972 data). In addition, some 850 non-seagoing barge movements are needed for lightering purposes (10) and also move past Salem each year. Of this grand total of 1,400 loaded barge movements per year, 850 carry crude oil (essentially non-flammable), 160 carry sulfuric acid, 180 carry jet fuel, and the rest carry "clean" petroleum products such as fuel oil and gasoline (9). Since sulfuric acid and crude oil are not flammable, only 390 barges (1400-850-160) move by Salem each year carrying a flammable cargo. Therefore, the total traffic of concern to this analysis is 390 barge movements per year. Note that these 390 barges all have drafts of less than 16 feet and all can potentially approach the intake and hit it. Based on accident statistics for the years 1968 to 1973 collected by the U. S. Coast Guard (11), the national average accident frequency involving all barges where damage was in excess of $1,500 has been found to be 0.42 accidents per million miles (12). The frequency of all types of barge accidents is therefore 0.42 x

  -6 10     per mile. Runaway barge accidents form a small subset of all accidents since most barge accidents are caused by impact with bridges, weirs, spillways, piers, other barges, etc., but not due to runaways.      A very conservative estimate of number of runaways per total number of accidents is estimated at 0.1 (13).

We calculate the frequency of runaway barge occurrences per year per mile of the Delaware River in the vicinity of Salem, designated by fas: 6 f = number of barge runaways = 390x0.42xl0- xO.l year-mile Delaware

                                                        = l.6xl0- 5 /year-mile We now address the question of how many of these runaways will strike the Salem intake.         As shown below, if a barge can run away 2.2-6 SGS-UFSAR                                                   Revision 6 February 15, 1987

in any direction with equal probability and does not change direction once it has run away, the probability of it striking a target (water intake) of length Q is given by f 2 Consider the collision geometry shown below: Jl Intake Land River d A barge (assumed a point) Bruns away in a random direction at a projected distance x from the center of target (water intakes) of length Q. Once the barge has run away, it is assumed that it does not change directions. In order to hit Q, the barge must be within the section given by angle 0. Since all angles of a runaway are equally likely, the probability of runaway within an angle 0 is 0/2n. The total probability of a barge runaway at any distance x causing a strike at£ is b f £ 0 (x) d 27t X a where a and b represent the two end-points of the barge movements. Since the integrand decreases rapidly with distance away from the intake, the limits can be replaced with+/- m. 2.2-7 SGS-UFSAR Revision 6 February 15, 1987

00 f L.uil dx = -1 x+Jl/2 -1 x-£/2 f 2 TC [tan d - tan d ] dx 00

                                        -1                    £/d
                            ~ f 0

tan 1 + :;.:;X_2___(,._Jl...,/_Z..._)_2 dx d2

                          = f    for the case when d >> Q/2 2    This is true for the Delaware where d = 5000', Q = 100' The above expression represents the probability of a runaway barge (originating anywhere in the river) striking the target£.

Note that the presence of a stream current in real rivers will prevent straight trajectories for runaway barges. However, the effect of slightly curved trajectories is expected to have little effect on the end result. Since the length of the water intake is 110 feet, or O. 02 mile, the probability of a runaway barge striking the intake is: strikes year Not every strike will involve spillage of chemical. On a national average basis, only 45 percent of the barge accidents result in the involvement or release of contents (12) and about 7 percent of the releases result in fire (14). Carrying these frequencies to runaways, the probability of a barge running away, hitting the intake, releasing some of the flammable content and igniting is: I.6xIO-7 x 0.45 x 0.07 = 5.0xlO- 9 occurrences/year This represents an extremely remote event. 2.2-8 SGS-UFSAR Revision 6 February 15, 1987

2.2.3.2 Hazardous Chemicals - Onsite Regulatory Guide 1. 78, Paragraph C. 2 states that hazardous chemicals such as those indicated in Table C-1 of the Guide, must be included in the analysis if they are frequently shipped within a 5 mile radius of the plant. The Guide also defines frequent shipments as being 50 or more trips per year for barge traffic, 10 or more trips per year for truck traffic, and specifies in Paragraph C. l, that chemicals stored or situated at distances greater than 5 miles from the facility need not be considered. Following is the analysis of control room habitability during a postulated hazardous chemical release occurring either on the site or within a 5 mile radius of the plant. As indicated in Section 2.2, the Salem site is located in a rural area with no major manufacturing or chemical plants located within 5 miles of the site. The only major transportation route within 5 miles of the plant is the Delaware River, with the Intracoastal Waterway passing 1 mile west of the site. The SGS uses a sodium hypochlorite biocide system, thus eliminating an onsite chlorine hazard. The control room is equipped with smoke and combustible detectors located in the air conditioning unit ducts, These detectors provide alarms in the control room in the event of smoke or combustible hazards present. The control room is equipped with radiation detectors which provide annunciation, automatically isolate the control room, and initiate emergency ventilation in the pressurized mode. The site was reviewed to identify potentially hazardous chemicals which may impact control room habitability during a postulated release. The site includes the SGS, HCGS, and deliveries to and near the site. Hazardous chemicals which may impact control room habitability are identified as sulfuric acid., nitrogen, ammonium hydroxide, hydrazine, ethanolamine, sodium hydroxide, and helium. Fire fighting agents such as carbon dioxide and halon are discussed later in this section. The basis for identification was the chemical's physical properties, toxicity and/or asphyxiant threshold levels, and storage quantities and locations .

SGS-UFSAR 2.2-9 Revision 23 October 17, 2007

Table 2.2-2 presents the chemicals stored onsite or shipped by the site on the Delaware River which are identified in Regulatory Guide 1.78, Table C-1. Table 6.4-3 in Section 6.4 provides information on the control room ventilation system, as required by Regulatory Guide 1.78, Paragraph C.7. As can be seen from Table 2.2-2, the hazardous chemicals stored onsite are sulfuric acid, nitrogen, ammonium hydroxide, hydrazine, ethanolamine, sodium hydroxide, and helium. Purate is also provided in Table 2.2-2 in addition to the RG 1.78 chemicals. As previously mentioned, several chemicals are stored onsite that are considered hazardous. Sulfuric acid is stored in 4,000 and 2,250 gallon tanks in the SGS Turbine Buildings and it is stored in 16,000 and 6,650 gallon tanks at the HCGS. Calculations indicated that the toxicity limit found in Regulatory Guide 1.78 will not be exceeded in the control rooms during a postulated release at any of the sources. Liquid nitrogen and nitrogen stored as a compressed gas is stored at various locations onsite. According to the criteria contained in Regulatory Guide 1.78, the largest single source should be evaluated for its impact on control room habitability. The sources evaluated at the SGS are the portable nitrogen tube trailers located in various areas throughout the SGS yard area and the (2) liquid nitrogen tanks located behind Unit No. 1 & 2 Auxiliary Buildings which can contain up to 7500 gallons of liquid nitrogen. In addition to these sources, liquid nitrogen is also stored in 9,000 gallon tanks at the HCGS. Calculations indicated that the oxygen depletion is negligible in the control rooms during a postulated release at any of the significant sources. Chemicals used as fire-fighting agents were evaluated. Carbon dioxide is stored on the 84 foot elevation of each of the Auxiliary Buildings. It is also stored at the HCGS. Calculations indicated that the toxicity limit established in Regulatory Guide 1.78 as well as asphyxiation levels would not be exceeded during postulated releases at the significant sources. The Halon storage vessels are relatively small and do not contain the volume of Halon required to cause asphyxiation in the control rooms; therefore, a postulated release will not pose a danger to the control rooms. 2.2-10 SGS-UFSAR Revision 33 October 24, 2022

Ammonium hydroxide is stored in two 350 gallon vessel totes that are connected in series in the SGS Unit No. 1 and SGS Unit No. 2 Turbine Buildings. Evaluations concluded that the control rooms would remain habitable during a postulated release at either of the storage tank locations. The shipments to the site are considered "frequent" and are discussed in Section 2.2.3.3. Hydrazine is stored in a 300 gallon vessel also in the Unit No. 1 side of the SGS Turbine Building. The calculations indicated that the control room concentrations will not exceed toxicity limits established in 29CFR Part 1910.1000, Subpart Z during a postulated release. Ethanolamine is stored in two 350 gallon totes that are connected in series in the SGS Unit 2. The effective volume is 700 gallons. Evaluations concluded that the control rooms would remain habitable during a postulated release at the storage totes. The shipments to the site are considered frequent and are discussed in Section 2.2.3.3. Aqueous sodium hydroxide is stored in various quantities and vessels at both the SGS and HCGS. Upon a release, sodium hydroxide vapors may form locally at the spill, but the physical properties of this chemical preclude the formation of a plume that will travel to the control room air intakes. The vapor pressure of aqueous sodium hydroxide is very low, especially as the concentration is increased. During a postulated release, mostly water will evaporate from the liquid pool, leaving the solid sodium hydroxide behind. The solid form of sodium hydroxide poses no danger to the control room due to its physical properties. Helium is stored in 150 lb cylinders at both the SGS and HCGS. It is much lighter than air and upon a postulated failure of one of the cylinders, the helium would disperse rapidly into the atmosphere and not form a continuing plume. Table 2.2-6 also lists Purate, which was evaluated and does not pose a control room hazard. Chlorine dioxide (dissolved in circulating water) is produced by the Purate system in a structure near the Hope Creek station cooling tower. Evaluations concluded that the control room would remain habitable during a postulated chlorine dioxide release at the Purate System structure. Our analysis of the control room habitability requirements demonstrates that the control room personnel are adequately protected against the effects of accidental release of onsite 2.2-11 SGS-UFSAR Revision 33 October 24, 2022

hazardous chemicals and radioactive gases, and shows that the plant can be safely operated or shut down under design basis accident conditions. Due to the use of sodium hypochlorite, there is no chlorine hazard. 2.2.3.3 Hazardous Chemicals - Offsite Table 2.2-4 provides a tabulation of estimated of hazardous chemicals past Artificial some of which are listed in Table C-1 of Guide 1.78. Guide 1. 78 a control room habitability evaluation for shipments of hazardous chemicals that are considered "frequent" shipments. The frequent criterion for river barges is 50 per year. As seen from Table 2.2-4, none of the hazardous chemicals shipped past the site exceed this criteria, therefore, a control room habitability evaluation is not requ~red. Hazardous chemicals are also delivered to the SGS and the HCGS. Table 2. 2-4 lists the deliveries of hazardous chemicals to the Stations. A review of the deliveries were compared to the criteria as stated in Regulatory Guide 1.78. Aqueous sodium hydroxide, sodium hypochlorite, ammonium bisulfite, and ethanolamine shipments are considered "frequent". As mentioned previously, a release of either sodium hydroxide or sodium hypochlorite will not impact the control rooms due to the physical of these chemicals. Ammonium bisulfite is also characterized a a chemical that will not and form a plume a release due to its very low Therefore, a failure of the tankers these hazardous chemicals onsite will not control room habitability. Ethanolamine (ETA) is shipped frequently to the site in 350 gallon stainless steel totes. ETA is characterized as a chemical that will not readily evaporate and form a plume during a release. Therefore, a catastrophic failure of the truck the totes onsite will not impact control room 2.2-12 SGS-UFSAR Revision 23 October 17, 2007

Ammonium hydroxide, compressed nitrogen and sulfuric acid shipments delivered onsite also require an evaluation of their impact on control room habitability since their deli very schedule exceeds the criteria in Regulatory Guide 1. 78. Calculations conclude that a release of ammonium hydroxide directly from a delivery tanker while onsite may exceed the toxicity limit contained in Guide 1.78; however, administrative controls are in place to prevent the control rooms from the limit. Normal deliveries of ammonium hydroxide will consist of 350 totes constructed of stainless steel. The rupture of this vessel was modeled and shown to be within allowable limits in Regulatory Guide 1. 78. Calculations regarding the portable nitrogen tube trailers and sulfuric acid tankers conclude that the control rooms will not be impacted during a catastrophic release. The station control rooms have separate and ventilation air which are isolable (see Section 9.4.1). The ventilation system uses charcoal filters for the iodine removal in the event of a radiological release. The charcoal provides absorption capability for most of the hazardous chemicals. Protection is further provided through individual emergency breathing apparatus located in or near the control room and by protective clothing which is available in other areas of the plant. 2.2-12a SGS-UFSAR Revision 26 May 21, 2012

THIS PAGE INTENTIONALLY LEFT BLANK 2.2-12b SGS-UFSAR Revision 13 June 12, 1994

Public Service Electric & Gas (PSE&G) has performed detailed studies of the potential hazards of ship transportation of the materials listed in Table 2.2-4, as well as liquified petroleum gas (LPG) and liquified natural gas (LNG). The report, entitled "Analysis of Potential Effects of Waterborne Traffic on the Safety of the Control Room and Water Intakes at Hope Creek Generating Station, " was submitted on October 2, 1974, on the Hope Creek docket (Docket Nos. 50-354 and 50-355). The report provides analyses that support the conclusion that the probability of flammable vapor cloud reaching the nuclear facilities at Artificial Island is sufficiently low such that accidents occurring from the waterborne transportation of hazardous materials need not be considered in the design basis of the nuclear facilities at Artificial Island. PSE&G is required as a condition of the Hope Creek Construction Permit to submit a yearly report updating factors that affect the probability of a flammable vapor cloud reaching Artificial Island. In addition, any significant changes that alter the probability calculations must be reported in a more prompt manner. A condition of the Salem Unit No. 2 Operating License requires that any significant information affecting probabilities reported on the Hope Creek docket must also be reported on the Salem docket. The ability to isolate the ventilation system and recirculate the air, along with the protective breathing equipment, provides sufficient time to bring the plant to a safe shutdown condition from the control room in the event of hazardous chemical release from waterborne traffic. 2.2.4 References for Section 2.2

1. Robert W. O'Brian, Director, "The Comprehensive Plan," Kent County Regional Planning Commission, Dover, Del. 1971 2.2-13 SGS-UFSAR Revision 12 July 22, 1992

2. Communication, Dames and Moore and Greater Wilmington Airport Information Office, Wilmington, Del.

3. Directory of Commerce and Industry, Delaware State Chamber of Commerce, Inc., Wilmington, Del., 1970.

4. Industrial Directory, State of New Jersey, Industrial Directories, Inc., New York, N.Y., 1971.

5. Written Communication, Dames and Moore survey from letter to pertinent industries.

6. Waterborne Commerce of the United States: 1970, Part I, Atlantic Coast, U.S. Army Corps of Engineers, 1971.

7. Communication, Dames and Moore and Rollin Neeman, Bureau of Planning, State of Delaware, Dover, Del.

8. Communication, Dames and Moore and Clerk, Bus Depot, State Road, Del.

9. U.S. Army Corps of Engineers, Waterborne Commerce of the United States, Delaware River, Philadelphia Harbor, 1972.

10. Personal Communications, Mr. Howard Lynch Interstate Oil Transport, Penn Central Plaza, Philadelphia, Pa.

11. U. S. Coast Guard Headquarters, Computer File on all Accidents Involving Damage in Excess of $1,500. Washington, D.C., 1974.

12. U. S. Department of Commerce, A Model Economic and Safety Analysis of the Transportation of Hazardous Materials in Bulk, Report to Office of Domestic Shipping by Arthur D.

Little, Inc., Cambridge, Mass., July 1974.

13. U. S. Coast Guard, "Statistical Summary of Casualties to Commercial Vessels on Western Rivers, "November 1973. In this report, runaway barges are classified as being due to material failure (e.g., a broken tow line) and are found to represent 4 percent of all barge accidents. For the Delaware, a conservative estimate of 10 percent is used.

14. Atomic Energy Commission, "The Probability of Transportation Accidents, "by William A. Brobst. Presented at the 14th Annual Explosives Safety Seminar, New Orleans, Louisiana, November 1972.

15. Waterborne Commerce of the United States, U. S. Army Corps of Engineers.

16. Commodity traffic data for imports and exports collected by the Philadelphia Maritime Exchange.

17. Foreign trade cargo movements collected by the Delaware River Port Authority.

2.2-14 SGS-UFSAR Revision 12 July 22, 1992

18. U. s. Department of Commerce, Census Bureau (handling foreign trade data for custom purposes.

19. Interstate Oil Transport, Inc. (which handles most of the barge operations on the Delaware River).

20. U. s. Coast Guard, Captain of the Port, Philadelphia (who is cognizant of all hazardous materials shipments in the Delaware River).

21. U.S. Coast Guard, Vessel Chemical Traffic Report, "Hazardous Traffic Passing Salem and Hope creek Stations,w Dated July 15, 1993.

2.2-15 SGS-UFSAR Revision 16 January 31, 1998

TABLE 2.2-1 INDUSTRIES WITHIN TEN MILES OF THE SITE Estimated No. of Company Location Employees Product MANUFACTURING PLANTS

1. Gayner Glass 8 miles (NE) 266a Glass Containers

2. Anchor Hocking Co. 8 miles (NE) 1323a Glass Containers

3. Mannington Mills, Inc. 8 miles (NE) 567a Felt Base Floor and Wall Coverings

4. H. J. Heinz Co. 8 miles (NE) 200a Pickled Fruits, Vegetables, Sauces

s. Blue Ridge-Winkler 9.5 miles (W) 26-SOc Garments Textiles

6. Evergreen Acres, Inc. 9.5 miles (W) sob Ornamental Evergreens

7. Gioia Specialty Food 6.5 miles (W) 0-25c Canned Beans Inc .

8. St. Georges Canning Co. 8.8 miles (NW) 51-lOOc Canned Vegetables

9. Tyson F. Sartin, Inc. 8.8 miles (NW) 20b Septic Tanks, Well Rings

10. Globe Union, Inc. 9.5 miles (W) 300b Lead Acid Auto Storage Batteries

11. Delmarva Power & Light 10. 2 miles (NNW) NA Electric Power Co. Fossil Fuel Plant CHEMICAL PLANTS AND STORAGE FACILITIES

12. Getty Oil Co. 10 miles (NNW) 501-lOOOc Petroleum and Petro-chemicals

13. Stauffer Chemical Co. 10 miles (NNW) 51-100c Chemicals, Inorganic Resin

SGS-UFSAR 1 of 2 Revision 6 February 15, 1987

TABLE 2.2-1 (Cont) Estimated No. of Company Location Employees Product MANUFACTURING PLANTS

14. Keysor Chemical Co. 10 miles (NNW) NA Petrochemicals

15. Air Products and 10.5 miles (NNW) 26-50c Hydrogen and Chemicals, Inc. Industrial Gases

16. Standard Chlorine of 10.5 miles (NNW) 26-50c Chlorine, HCL Delaware, Inc.

17. Pioneer Chloromane 10.4 miles (NNW) 28b Chlorine

18. Diamond Shamrock 10.3 miles (NNW) 201-300c Chlorine, Caustic Chemical Co. Soda, Hydrogen

19. Stauffer Hoechst 10.2 miles (NNW) 151-200c Film Polymer Corporation

NA - Not available a- Source: N.J. Industrial Directory, 1971.

b - Source: Questionnaire sent to all Delaware Firms c - Source: Delaware Directory of Commerce and Industry, 1970. (A range in number of employees is given.) Note: Locations shown on Figure 2.5-1.

SGS-UFSAR 2 of 2 Revision 6 February 15, 1987

TABLE 2.2-2 HAZARDOUS CHEMICALS STORED ONSITE Name of Sulfuric Ammonium Sodium Carbon Sodium Chemical Acid Nitrogen Hydroxide Hydrazine Ethanolamine Hydroxide Helium Dioxide Halon Hypochlorite Purate Type of Onsite Onsite Onsite Onsite Onsite Onsite Onsite Onsite Onsite Onsite Onsite Source Human 1.0 N/A 3.5 3.5 N/A N/A N/A N/A N/A N/A Detection Threshold (mg/m3) Maximum 2.0 Asphyxi- 70.0 0.04 2.0 Asphyxi- Asphyxi- Asphyxi- N/A N/A Allowable ant ant ant ant 2-minute Limit (mg/m3) Largest 1) 4000 1) 9000 1) 3000 1) 300 1) 700 1) 4000 1) 150 2) 10 1) 310 1) 88,000 6,650 Single (Unit 1) (Hope (Unit 1) lbs tons lbs (Hope Creek) Container Creek) (Salem) of Chemical 2) 2250 2) 7500 2) 2250 1) 17 tons (gallons) (Unit 2) Salem (Unit 2) (Hope Creek)

3) 6,650 (Hope Creek)

Maximum Approx. Instant- 450 0.53 Approx Instant- Instant- Instant- Approx. Continuous zero aneous zero aneous aneous aneous zero Release Rate (g/s) Vapor Approx. N/A 450@ 35@ 0.3-0.4@ 68F Approx N/A N/A N/A Approx. 50.25 @ 104F Pressure zero 115°F 115°F zero zero (mmHg) Fraction 0% 100% 0% 0% 0% 100% 55% 100% 0% of Chemical Flashed/ Rate of Boiloff when Spilling occurs Closest 1) 290 1) 200 1) 275 1) 375 1) 267 1) 300 1) 325 1) 140 1) <100 1) 575 2400 Distance (portable (Unit 1) (Unit 1) between tube Source 2) 280 trailer) and control 3) 2400 room (ft) 1 of 1 Revision 33 SGS-UFSAR October 24, 2022

TABLE 2.2-3 THIS TABLE DELETED 1 of 1 SGS-UFSAR Revision 16 January 31, 1998

TABLE 2.2-4 ESTIMATES OF HAZARDOUS CHEMICAL TRAFFIC Chemical No. of Vessel Trips/Year¹ Acetone² 2 Ammonia² (incl. anhydrous ammonia) 14 Ammonium Bisulfite 2 Ammonium Hydroxide³ 30 Asphalt 2 Benzene² 5 Butane 1 Caustic Soda 26 Cresylic Caustic 1 Cumene 11 Cyclohexane 3 Ethanol 1 Ethanolamine 6 >10 Heptane 1 Heptene 1 Lube Oil 1 Methyl Alcohol² (methanol) 9 Methyl tertiary butyl ether (MTBE) 32 Napthalene 4 Nitrogen (compressed gas) 2,5 >10 Nonene 8 Paraffin 1 Propane² 2 Propylene 1 1 of 2 SGS-UFSAR Revision 26 May 21, 2012

TABLE 2.2-4 (Cont'd) Chemical No. of Vessel Trips/Year¹ 4 Sodium Hydroxide (mercury cell grade) 6 4 Sodium Hydroxide (diaphragm grade) 26 3,4 Sodium Hydroxide 104 3,4 Sulfuric Acid 110 Toluene 2 VGO 1 Xylene 9 Purate7 12 78% Sulfuric Acid7 12 Notes: (1) Delivery frequencies were provided per Reference 21 except where noted. (2) Chemical is contained in Table C-1 of Regulatory Guide 1.78, or its references. (3) Delivery frequency is based on Salem chemical delivery ordering logs and reflects number of tote deliveries. (4) Delivery frequency is based on Hope Creek bulk chemical delivery ordering logs and reflects tanker truck delivery. (5) Delivery frequencies are based on Salem bulk chemical delivery ordering logs and reflect portable tube trailer delivery. (6) Delivery frequency reflects portable tote truck delivery. (7) Delivery frequency is based on anticipated usage of the Hope Creek Purate Biocide system at the time the new system was installed. 2 of 2 SGS-UFSAR Revision 33 October 24, 2022

K BY: MANUPACW!!.IN(i I'LANTS I, GA YNBR GLASS

2. ANC!Illl. HOCICIOO CO.

3. MANNINGTON MII.LS, INC.

                                   ~-   HEINZ, If. J. CO.

s. IILUERIIXlll*WINKI.J1.R Tl!X11LilS

6. EVERGREEN ACRES. INC.

1. GIOIA SPECIALT\" FOOD, INC.

8. ST. GEORGES CANNING CO.

9. li'SON P. SARnN, INC,

10. 01..082UNION, INC,

11. DELMARVA lOWER AND LIGHT CO.

FOSSIL Plll!L I'LANT CHEMICAL I'LANTS

12. GEITTY 011. CO,

13. STAUFFER CHEMICAL CO.

14, KEYSOR CHEMICAL CO.

15. AIR PRODUCTS AND CHEMICALS, INC.

16. STANDARD CHLORINE OF DBLAWARB, INC,

17. PIONDR cin.o&OMANB

                                 !B. DIAMONDSHAMROCKCO.

19. STAUFFERHOECHSTIOL.YMER CORP, TANK FARM

20. GJ!TTY 011. CO.

21. PIONBBRClfl..OROMANE

22. GETTY OIL CO.

TANKER DOCK

23. GETTY OIL CD.

BOATING ACCESS AREAS 24, AUGUSTINE BEACH 2S. WOODLAND IISACH

26. ALI'HA YACHT CL.UB

                                      !NniiACOASTAL WATl!RWAY n

R£fER£HC£: REVISION I FEBRUARY 11. 1117 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Sit. Vicinity Map Showing Major Fdlties SALEM NUCLEAR GENEilATING STATION Upd*ted FSAR

2.3 METEOROLOGY 2.3.1 Regional Climatology 2.3.1. 1 Data Sources Data sources are listed in the references provided at the end of this section. 2.3.1.2 General Climate Based on the Koeppen climatic classification system, the region intersects two climatic zones. They are humid continental and humid sub-tropical. Both zones have characteristics of warm summers and mild winters (1). Summer maximum average temperatures are near 80 degrees Fahrenheit, and the coldest month is January having an average daily temperature of approximately 32 degrees Fahrenheit. Examining a 30 year mean of precipitation amounts for Wilmington, Delaware, National Weather Service (NWS) station shows that the most rainfall occurs in the summer months, followed by spring, fall, and winter (2). The area of southern New Jersey is frequented by Polar Canadian air masses in the fall and winter and occasionally invaded by Arctic Canadian air late in winter. During the spring and summer, the dominant air mass is Maritime Tropical (1). 2.3.1.2.1 Precipitation The frequency of precipitation events such as rain, snow, ice storms, thunderstorms, and hail are tabulated in Tables 2. 3-1, 2.3-2, and 2.3-3. The data in Table 2.3-1 were obtained from the Revised Uniform Summary of Surface Weather Observations, Dover (Delaware) Air Force Base, 1942-1965. The data presented in Tables 2.3-2 and 2.3-3 were obtained from Philadelphia International Airport and Trenton Airport, respectively. 2.3-1 SGS-UFSAR Revision 6 February 15, 1987

2.3.1.2.2 Humidity, Winds Humidity annually averages 70 percent (3). Prevailing winds on a monthly average during the winter (December to March) are from a northwest direction with a range of speeds from 9 to 13 mph. Average monthly winds for the spring and summer months (April to August) are from a southerly to southwesterly direction at speeds ranging from 7 to 10 mph. Winds during the fall are predominantly from the west-southwest veering to a west-northwest direction by December. The average wind speeds increase as the season progresses (4). 2.3.1.3 Severe Weather The terrain is open and extremely flat which favors a vigorous wind flow. While the area is almost certain to experience hurricane force winds frequently, there is no reason to anticipate fastest mile velocity, reaching 100 miles per hour, more than once in 100 years. Table 2.3-4 lists the distribution of peak winds for Philadelphia International Airport based on a 25-year record. The tornado frequency in this area is reassuringly low; a few small funnels have been observed in southeastern Pennsylvania and southern New Jersey, but it is unlikely that any tornado would affect the site itself more than once in 4300 years. 2.3.2 Local Meteorology Figure 2.3-1 shows the different stations that collect meteorological records. Figures 2.3-2 and 2.3-3 are 2-year wind roses derived from Artificial Island wind data using all hours and only hours with a stable stability, respectively. The wind direction is randomly distributed when stable atmospheric conditions occur, whereas using all hours of data shows a northwest wind direction peak. 2.3-2 SGS-UFSAR Revision 6 February 15, 1987

2.3.3 Onsite Meteorological Measurements Program Meteorological Data Collection Program In order to arrive at atmospheric dispersion factors for use in calculating radiological exposures from both low level normal releases and accidental releases, an extensive data collection program was undertaken at the site. This data collection program is described in detail in the following paragraphs. The present meteorological monitoring program is in conformance with the recommendations of Regulatory Guide 1.23. 2.3.3.1 Preoperational Data Collection Program Data became available from the 300 feet meteorological tower located on Artificial Island in June of 1969. The official preoperational data collection program was terminated at the end of May 1971. The tower was positioned just north of the actual plant site and is shown on Figure 2.3-1. The actual location was 2700 feet north of Unit 2 at a Latitude of 39 degrees 28 minutes 13 seconds north, and a Longitude of 75 degrees, 32 minutes 12 seconds west. A detailed representation of the meteorological facility is not necessary because of the simplicity of the terrain. The tower data used in this study is primarily that from the 33 and 300-foot levels, although some data were obtained at the intermediate 150-foot elevation. The wind instrumentation consisted of Aerovanes, and the temperature-difference measurements were obtained from aspirated resistance thermometers. The usual precipitation, humidity, and solar radiation are on record if they are ever needed for general environmental applications. 2.3-3 SGS-UFSAR Revision 6 February 15, 1987

2.3.3.1.1 Data Summaries and Turbulence Classifications The record of temperature and all other data extends from June 1969. Data are being obtained continuously. A monthly distribution is presented in Table 2.3-5. Table 2. 3-6 shows monthly summaries of precipitation in inches for June 1969 through November 1970. Included with this summary is a range of maximum hourly rates. Table 2.3-7 lists the monthly percentages of hours with fog in 3-hour intervals. October through March have the largest percentage of fog during the hours of 0600 to 0800. During April through September, the largest percentage occurred between the hours of 0300 to 0500. Stability Alternative techniques of estimating the turbulence usually involve one of two methods: approximating it from a combination of lapse rate and 'tJind speed measurements, or from the fluctuations of a standard wind instrument such as an Aerovane. We believe the latter to be more representative of the typical ~Lvu~~m~, and accordingly this is based on wind direction range and data. The rate classification has been used, however, and some of the data are summarized in the report. In this the two techniques are in good Turbulence Classifications The system used for defining the turbulence is that developed originally by Singer and Smith(5) and widely applied in both nuclear and fossil power plant evaluations. The classification is depicted on Figure 2. 3-4, where Classes I and II unstable conditions. 2.3-4 SGS-UFSAR Revision 25 October 26, 2010

Class III is the overcast stormy situation, and Class IV is the stable, inversion flow pattern. In the Preliminary Safety Analysis Report the distribution of turbulence classifications obtained from the Delaware City site 10 miles north-northwest of Salem, was presented as probably typical of the dispersion regimes. In Table 2.3-8 the new Salem data (300-foot level) are compared with the earlier summary from Delaware City, and the agreement is very good despite the fact that the information was obtained in different years. The only notable difference is that Salem showed a more marked tendency toward the neutral Class III turbulence than did Delaware City. This aberration may be real, but it is more likely that the water tower on which the Delaware City instrument was located produced somewhat broader and more turbulent direction traces than the clean installation at Salem. In any case, the difference has no great significance in the dispersion evaluation. At both sites, the distributions seem quite normal for open, mid-latitude locations. The Class II turbulence dominates the distributions, accounting for approximately 60 percent of all hours, and the stable cases are found in roughly 25 percent of the remainder. We had anticipated a noticeable increase in the frequency of Class IV conditions during the late spring and early summer at Salem, because it is directly exposed to over-water flow which might be stable, but apparently the combination of infrequent winds from the 130- to 160-degree sector and the relatively mild bay temperatures did not produce the expected increase. Lapse Rates In Table 2.3-9, the distribution of lapse rates over the year is shown. These data agree well within the indications of the turbulence classification, in that 24 percent of the hours appear to be stable, 14 percent neutral, and the remainder unstable. 2.3-5 SGS-UFSAR Revision 6 February 15, 1987

Another indication that the water influence is fairly small at this site is that the diurnal variation of the lapse rate in June (Figure 2.3-5) does not show any tendency toward stability in the afternoon hours, and, in fact, is quite similar to the December (Figure 2.3-6) pattern. Relation Between Lapse Rates and Turbulence Classes As a final comparison between turbulence classes and the lapse rate data, Table 2.3-10 is presented which clearly shows that the two methods of estimating turbulence are compatible at this site. The vast majority of Class I and Class II turbulence hours are associated with unstable lapse rates, and the Class IV hours are primarily inversion periods as they ought to be. The distributions of lapse rates, winds, and turbulence classes already presented are adequate to define the diffusion meteorology of this site as quite normal and uncomplicated, but it is important to translate the data as accurately as possible into the dispersion parameters actually used in numerical evaluations. Since the experience with the bi-directional wind vane was typically unsuccessful, the measurement of hourly wind direction range was evaluated and used for estimates of ae. These data, separated according to turbulence class, are given for the entire period of observation in Table 2.3-11, and it is apparent that the wind fluctuations at this site are very nearly identical to those at Brookhaven National Laboratory (6) where the turbulence classification was originally developed. It therefore is reasonable to utilize the diffusion parameters developed at that site (7) in this study. One further point is important, and that is to be sure that diffusion with south-southeast winds from the open waters of Delaware Bay is not significantly different from that occurring with other wind directions. Table 2. 3-12 is a replica of Table

2. 3-11, except that only south-southeast winds are represented.

Obviously there is no difference. 2.3-6 SGS-UFSAR Revision 6 February 15, 1987

Hour-by-hour stability frequency tables are presented in Tables 2.3-9 and 2.3-10 . The distribution of wind speeds at the 33- and 300-foot levels as a function of turbulence class are presented in Table 2.3-13, where the most notable feature is the very low frequency of calms. Normally, with an Aerovane as a sensing instrument, calms at the 33-foot elevation are prominent, but the very flat terrain and the air-sea interaction at Salem obviously favor a vigorous wind flow. Also, the percentage of hours having relatively high speeds, reflected in both Tables 2.3~13 and 2.3-14, is quite large, as one would anticipate in this locality. Data recovery percentages for the June 1969 to May 1970, 33-foot and 300-foot wind data, are shown in Table 2.3-15. 2.3.3.2 Operational Data Collection Program The digital Meteorological Data Acquisition Systems provide increased data recovery over traditional systems. The digital Meteorological Data Acquisition systems were designed to meet the intent of Regulatory Guide 1.23. The Salem and Hope Creek Safety Parameter Display System ( SPDS) 1 provides an Artificial Island wide source of 15-minute average meteorological monitoring system parameters, which are read from the two digital data acquisition systems. The parameters available for display are 33-ft wind speed, direction, sigma theta, and horizontal stability class; 150-ft wind speed, direction, sigma theta, and horizontal stability class; 300-ft wind speed, direction, sigma theta, and horizontal stability class; delta temperature between 300 and 33-ft; delta temperature between 150 and 33-ft; vertical stability class for each delta temperature; precipitation; barometric pressure; solar radiation; and ambient and dew point temperatures .

SGS-UFSAR 2.3-7 Revision 23 October 17, 2007

Atmospheric transport and diffusion is calculated by the Meteorological Information and Dose Assessment System (MIDAS) computers installed in both Salem and Hope Creek. A method for determining atmospheric transport and diffusion throughout the plume exposure emergency planning zone during emergency conditions has been developed. The system became operational in April 1976. The location of the 300-foot guy wire supported tower is Latitude 39 degrees, 27 minutes, 48.9 seconds, North and Longitude 75 degrees, 31 minutes, 11.76 seconds, West. The data collection program also includes an additional tower, identified as a backup meteorological tower, consisting of a 10-meter telephone pole. The backup tower is located approximately 500 feet south of the primary meteorological monitoring tower. Backup meteorological data provides wind speed, wind direction, and a computed sigma theta. Wind speed and direction instruments are located at 300-foot, 150-foot, and 33-foot elevations on the primary tower and at the 33-foot elevation on the backup tower. Temperature measurement includes ambient temperature taken at the 33 foot elevation and temperature differences taken between T - T and T - 300 33 150 T levels. Temperature sensors consist of RTDs in a motor aspirated solar 33 radiation shield. The dew-point is measured at the 33-foot level. Rainfall and barometric pressure are measured at approximately 3 and 6 feet, respectively. Figure 2.3-7 depicts the heights of these instruments on the tower. 2.3-8 SGS-UFSAR Revision 33 October 24, 2022

All meteorological parameters are electronically recorded in the Meteorological Instrument Building at the base of the tower. The data acquisition system includes capabilities for remote interrogation in addition to data acquisition. The data acquisition systems consist of primary and backup data acquisition systems (DAS) located at the Meteorological Instrument Building. A diagram of the system configuration is provided on Figure 2.3-8. The rain gauge uses a tipping bucket. The primary and backup DAS are configured with identical hardware. Each DAS is provided with communication ports, including one as a link to the Salem and Hope Creek SPDS. Each DAS provides storage for at least 7 days of 15-minute averages. The primary DAS collects wind speed and direction from the primary tower. The backup DAS collects wind speed and direction from the backup meteorological tower. Each DAS calculates a sigma theta for its respective meteorological tower (each of the three level wind directions on the primary tower, one level on the backup tower). The host computers acquire the meteorological data collected by the data loggers. 2.3-9 SGS-UFSAR Revision 33 October 24, 2022

The calculations of the sigma thetas use samples of horizontal wind direction at each elevation/location. Data interrogation is possible through connection to the digital data acquisition systems, which also provide data to the Salem and Hope Creek SPDS. The SPDS supports display units in the EOF, the Hope Creek Control Point, the Salem and Hope Creek TSCs, the Hope Creek OSC, and the Salem Ops Ready Room. Additional sources of meteorological data to provide a description of airflow trajectories from the site out to a distance of 50 miles include Wilmington and Philadelphia National Weather Service (NWS) stations. Hourly wind, temperature, and cloud cover data are readily available from these NWS stations. 2.3-10 SGS-UFSAR Revision 33 October 24, 2022

2.3.4 Short-Term Diffusion Estimate 2.3.4.1 Objective The objective is to provide conservative and realistic short term estimates of relative concentration (X/Q), at both the site boundary and the o~ter boundary of the low population zone (LPZ) following a hypothetical release of radioactivity from SGS Units 1 or 2. The assessment is based on the results of atmospheric diffusion modeling and onsite meteorological data .

SGS-UFSAR 2.3-11 Revision 18 April 26, 2000

A ground-level accidental radionuclide release from SGS is analyzed at various distances. Conservative and realistic X/Q values at the exclusion area boundary (EAB) are derived for the 0- to 2-hour period following a postulated accident. Conservative and realistic estimates of the X/Q value at the outer boundary of the LPZ are computed for 2, 8, 16, 72, and 624 hours following a postulated accident. For this modeling assessment, the EAB is a distance of 1270 meters in all sectors except the Northeast and East, which were 1391 meters. The LPZ boundary is 5.0 miles (8,047 meters), all sectors. 2.3.4.2 Accident Assessment The short-term, 0- to 2-hour X/Q values for ground-level releases are calculated with the sector dependent model described in Regulatory Guide 1.145, Reference 8. Annual accident X/Q values are also required to derive the intermediate time period X/Q values. These annual accident X/Q values are derived using the long-term diffusion model described in Regulatory Guide 1.111, Revision 1, Reference 9. 2.3.4.2.1 Methodology The procedures used to estimate the X/Q values for the appropriate time periods following a postulated accident are described in Regulatory Guide

1. 145. The diffusion model generates a cumulative frequency distribution of X/Q values for each sector-distance combination representing the first 2 hours after the postulated accident. These 2-hour X/Q values are based on 1-hour averaged data, but are assumed to apply for 2 hours. The frequency distributions are plotted on a log-probability scale for each sector-distance combination, and are then enveloped in accordance with the methodology described by Markee and Levine in Reference 11.

The X/Q value that is equaled or exceeded 0. S% of the time at each sector-distance combination is then determined from the intersection of the envelope and the 0.5'- probability level. The highest sector dependent X/Q value is then compared with the "overall" St accident X/Q value. The highest value represents the conservative 2-hour accident X/Q. The realistic 2-hour accident X/Q is evaluated at the overall 50% probability level. The overall 5% and 50% X/Q values are determined by summing the sixteen sector dependent X/Q distributions for each distance into a curnulati ve frequency distribution representJ.ng all sectors and again enveloping the data points. The 5% and 50% values are determined by the intersection of the envelope with the 5% and 50% probability levels, respectively. 2.3-12 SGS-UFSAR Revision 16 January 31, 1998

The conservative accident X/Q values for time periods of up to 30 following an accident are derived by logarithmic interpolation between the 2-hour 0.5% and the annual accident X/Q value at each sect6r-distance combination. The intermediate time periods for the overall 5% and 50% X/Q are determined by logarithmic *interpolation between the overall 2-hour 5% and 50% X/Q values and the maximum annual X/Q. The maximum conservative X/Q value for a given distance is the maximum sector 0.5% X/Q, or the overall 5% X/Q, whichever is higher, for the conservative assessment. The realistic assessment is based upon the overall X/Q and the overall 50% X/Q. The higher X/Q .value is chosen again. 2.3.4.2.2 Meteorological Data 2.3.4i2.2.1 Representativeness The

Artificial Island meteorological tower data from January 1988 through December 1994 are employed in the accident assessment. The data collected at the tower are representative of the meteorological conditions under which effluents are released, since both are located on the Delaware River shoreline.

Furthermor~, the proximity of the 300-foot tower to SGS ensures .that the data are representative of the conditions used in an accident evaluation . 2.3.4.2.2.2 Joint Frequency Distributions Joint distributions of wind speed and direction by atmospheric stability class are used as input to the diffusion calculations. Wind speed and direction data from the 33-foot level are used in the assessment of diffusion for the ground-level releases. Atmospheric stability is determined for the 33-foot distributions by the difference between the 300- and 33-foot levels. Joint frequency distributions of wind speed and direction by atmospheric stability class are computed for 22.5° sector using the wind speed groups and atmospheric stability classes suggested in Regulatory Guide 1. 23. The 7-year frequency distributions are used in the

2. 3-13 SGS-UFSAR Revision 16 January 31, 1998

With the exception of the calm and 25+ mph wind speed groups, the highest wind in each group is used to represent that group in the diffusion calculations. For conservatism, a wind speed of 0.5 mph is used to represent calms at the 33-foot level. This value represents a conservative threshold wind speed for the 33-foot wind instrumentation. Due to the high wind speeds asso~iated with this site, a wind speed of 30 mph is used to the 25+ mph wind speed group. 2.3.4.3 Atmospheric Diffusion Model The ~eactor building vent is treated as a ground-level source for both short-term and long-term calculations. This implies that no plume rise is calculated and no terrain corrections are applied. A building wake correction factor is used, in accordance with the methodology discussed in Regulatory Guide 1.145 for vent releases. The building wake correction factor takes into account the initial mixing of the plume within the building cavity. The vent release X/Q values are calculated with the following equations from Regulatory Guide 1.145: X/Q = (2.3-2) X/Q (2.3-3) X/Q 1 (2. 3-4) where: 3 X/Q relative concentration, s/m U10 wind speed at the 10 m level, m/s SGS-UFSAR

2. 3-14 Revision 21 December 6, 2004

Sy

lateral plume apeed, m

~ = lateral plume spread with meander and building wake effects, m sz = vertical plume spread, m A = smalleat vertical-plane cross-sectional area of the reactor building, 2 and adjacent structures, m

2 A building wake correction factor of 2430 m is used for calculations of the short-term X/Q.

For neutral or stable conditions combined with wind speeds less than 6.0 mjs, calculations of the X/Q value& are made using Equations 2. 3-2 through 2. 3-4. For all other meteorological conditions, X/Q values are calculated using Equations 2.3-2 and 2.3-3 only. The values computed from Equations 2.3-2 and 2.3-3 are compared, and the higher value is selected. For neutral and stable conditions with a wind &peed less than 6 mjs, the value from Equation 2.3-4 is compared with the value chosen from Equations 2.3-2 and 2.3-3, and the lower value is chosen to represent these conditions. 2.3.4.4 Diffusion Estimates 2.3.4.4.1 Exclusion Area Boundary

                                                                          -4    3 The maximum conservative 2-hour X/Q at the EAB, 0.79 miles, is 1.30 x 10      s/m

This is the maximum overall 0.5% sector dependent value at this distance. This value is larger than the overall 5% X/Q value. The maximum realistic (50\) 2-

                                 -S     3 hour X/Q at the EAB is 3.0 x 10      s/m

This is the overall SO% X/0, value.

Conservative and realistic X/0, values for the EAB (0.79 miles) for all the time periods following the accident are given in Table 2.3-21. 2.3.4.4.2 Low Population zone The maximum conservative and realistic X/0, values, 0.5\ and SO\, respectively, given in Table 2.3-21 represent the maximum X/0, value& (sector value used if greater than the overall value) at the LPZ boundary, 5 miles. 2.3-14a SGS-UFSAR Revision 16 January 31, 1998

2.3.5 Long-Term Diffusion Estimate 2.3.5.1 Objective The objective is to provide realistic estimates of annual average offsite atmospheric dilution factors based on site meteorological data. 2.3.5.2 Calculations Annual X/Q values for sixteen - 22. 5 -degree arcs at sixty distances are presented in Tables 2.3-17 through 2.3-20. The meteorological input data used was the 2-year period, June 1969 2.3-14b SGS-UFSAR Revision 16 January 31, 1998

through May 1971. X/0. estimates are based on the procedures presented in Regulatory Guide 1.111. These values were submitted in July 1976 as part of the Appendix I, 10CFRSO submittal to the NRC. 2.3.6 References for Section 2.3

1. Chritchfield, Boward J. "General Climatology," Englewood Cliffs, N.J.

(Prentice Ball Inc.) pp. 148-151, 1966.

2. Wilmington, Delaware Local Climatological Data, u.s. Department of Commerce, 1980 ad.

3. u.s. Department of Commerce. "Weather Atlas of the united States," pp.

170-175, June 1968.

4. u.s. Department of commerce. "Weather Atlas of the United States," pp.

228-234, June 1968.

5. DELETED

6. DELETED

7. DELETED

8. U.S. NRC "Atmospheric Dispersion Models For Consequence Assessments at Nuclear Power Plants," Regulatory Guide 1.145, Rev 1, Nov. 1982

9. u.s. NRC "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Regulatory Guide 1.111, Rev 1, July 1977

10. Meteorological Evaluation Services Co., Inc., "Accident X/0. Values at the Salam Generating Station Control Room Fresh Air Intakes, Exclusion Area Boundary and Low Population Zone," April 1996

11. Markee, E.B. and J.R. Levine, 1977, "Probabilistic Evaluations of Atmospheric Diffusion Conditions for Nuclear Facility Design and Siting,"

in proceedings of the American Meteorological society conference on Probability and Statistics in Atmospheric Sciences, Las Vegas, Nevada, pp. 146-150 2.3-15 SGS-UFSAR Revision 16 January 31, 1998

TABLE 2.3-1 PERCENTAGE OF DAYS WITH VARIOUS HYDROMETERS DOVER DELAWARE AIR FORCE BASE 1942-1965 Month Fog Snow and/or Sleet Hail Thunderstorms Jan 43.7 4.1 0.4 0.6 Feb 45.0 3.4 0.2 0.9 Mar 48.4 2.7 3.7 Apr 44.4 0.3 . 0.2 8.9 May 49.0 0.9 16.6 Jun 55.3 0.4 17.1 Jul 54.3 0.2 19.6 Aug 66.3. 17.4 Sept 59.0 6.8 Oct 53.8 0.2 3.0 Nov 47.6 0.6 0.2 1.2 Dec 44.5 2.5 0.2 0.5 Annual 51.2 1.2 0.3 8.2

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.3-2 SNOWFALL (inches) PHILADELPHIA INTERNATIONAL AIRPORT Monthly Month Mean Maximum Jan 5.7 19.7 Feb 6.1 18.4 Mar 4.1 13.4 Apr 0.3 4.3 May T T Jun Jul Aug Sept Oct T T Nov 0.8 8.8 Dec 4.6 18.8 Annual 21.6 Length of Record (yr) 28 (T = Trace of precipitation)

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.3-3 SNOWFALL (inches) TRENTON AIRPORT Monthly 24-Hour Month Mean Maximum Maximum Jan 5.8 16.1 10.1 Feb 6.7 23.1 13.0 Mar 4.4 21.5 14.3 Apr 0.4 4.2 4.2 May T T T Jun Jul i. Aug Sept Oct 0.1 1.6 1.6 Nov 1.0 13.0 7.7 Dec 4.9 21.5 16.6 Annual 23.3 Length of Record (yr) 34 (T = Trace of precipitation)

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.3-4 DISTRIBUTION OF PEAK WINDS PHILADELPHIA INTERNATIONAL AIRPORT (25-year record) Fastest Mile Month Speed (mph) Direction Jan 61 NE Feb 59 NW Mar 56 NW Apr 59 sw May 56 sw June 73 w July 67 E Sept NE 49 Oct 66 sw Nov 60 sw Dec 47 NW Fastest Mile Observed in Area: 88 mph, north, July 1931 Estimated Peak Hourly Value: 70 mph

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.3-5 DISTRIBUTION OF HOURLY TEMPERATURES (percent) Temperature Classes (OF) 10 0 +10 +20 +30 +40 +50 +60 +70 +80 +90 ( to to to to to to to to to to to to Month -20 -10 0 +10 +20 +30 +40 +50 +60 +70 +80 +90 +100 Jan 6 19 44 25 6 <1 Feb 6 31 42 17 4 Mar 9 52 35 4 <1 Apr 9 35 38 15 3 <1 May 8 36 34 14 6 2

 *Jun                                                       9   48     36   7  <1
 *Jul                                                       1   28     54  16   1
 *Aug                                                      <1    18    54  24
 *Sep                                                2     15   30     43   8   2
 *Oct                                   <1      6   19     33   34      8  <1
 *Nov                             <1     5     20   42     29     4 Dec                              1    25     59   14      1 Annual                          <1     1     10   18 15         14   17  18   5
 *2 months of data

SGS-UFSAR 1 of 1 Revision 6 February 15,. 1987

TABLE 2.3-6 PRECIPITATION (in water) Month 1969 1970 Range of Maximum Hourlx Rate Jan 0.65 0.01 to 0.10 Feb 1. 70 0.11 to 0.20 Mar 3.03 0.21 to 0.30 Apr 4.54 0.51 to 0.60 May 1.39 0.21 to 0.30 Jun 1.87 3.89 0.51 to 0.60 Jul 7.18 2.82 1.00 Plus Aug 3.75 1.29 o. 71 to 0.80 Sept 2.02 1.47 0.41 to 0.50 Oct 2.92 2.13 0.61 to 0.70 Nov 1.64 5.46 0.51 to 0.60 Dec 6.92 0.51 to 0.60

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 2.3-7 PERCENTAGE OF HOURS WITH FOG Hour 00- 03- 06- 09- 12- 15- 18- 21-Month 02 05 08 11 14 17 20 23 Mean Jan 19.8 22.3 23.8 19.2 13.5 13.8 15.7 17.3 18.2 Feb 21.4 23.3 25.1 18.0 14.2 13.9 16.5 18.2 18.8 Mar 20.3 23.3 24.9 15.8 12.2 12.2 14.9 17.4 17.6 Apr 18.4 24.2 23.2 12.8 8.8 10.1 12.3 14.18 15.6 May 22.7 27.9 22.2 10.1 6.0 5.4 8.6 14.7 14.7 Jun 21.4 37.2 22.9 7.9 4.6 4.0 6.5 11.0 14.4 Jul 22.7 35.8 23.8 5.1 3.6 3.1 4.8 11.7 13.8 Aug 27.6 42.5 31.8 6.8 3.7 3.1 6.3 14.2 17.0 Sept 25.9 37.6 33.9 9.4 5.0 4.8 8.6 16.2 17.7 Oct 23.6 33.5 35.0 11.2 6.6 6.5 9.6 15.0 17.6 Nov 19.4 22.9 27.6 14.9 8.0 8.6 12.3 15.8 16.2 Dec 20.4 21.4 25.5 19.9 14.7 14.9 17.1 18.0 19.0 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.3-8 PERCENTAGE FREQUENCY OF TURBULENCE CLASSES Salem and Delaware City Turbulence Class Month I II III IV Jan 6 (2) 62 (65) 13 (2) 19 (31) Feb 4 (3) 57 (64) 16 (5) 23 (28) Mar 7 (3) 59 (66) 12 (6) 22 (25) Apr 6 (2) 60 (72) 15 (9) 19 (17) May 12 (11) 59 (63) 6 (1) 23 (25)

 *Jun              13    (12)    57     (58)   10  (1)    20     (29)
 *Jul              12      (4)   58     (64)   10  (0)    20     (32)
 *Aug              12      (3)   53     (65)   10  (0)    25     (32)
 *Sep              14      (4)   50     (62)   12  (7)    24     (27)
 *Oct               8      (6)   52     (62)   14  (5)    26     (27)
 *Nov               6      (7)   56     (64)   13 (15)     25    (14)

Dec 4 (8) 72 (51) 12 (12) 12 (29) Annual 8 (6) 58 (62) 12 (5) 22 (27)

       *2 months of data

( ) data for Delaware City

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

                                 -TABLE 2.3-9 PERCENTAGE FREQUENCY OF LAPSE RATES Lapse Rate Group (t300 - t33°F)
                    -1.6 -0.4    +0.6       -1.6 +2.6 +3.6 to   to      to         to   to   to   2:

Month -0.5 +0.5 +2.5 +3.5 +4.5 +4.6

~                   4611                           Feb            18   37   14      10           6   6    3   6 Mar            20   47   14       6           4   3    2   4 Apr             19   45   12       75              6    0   6 May            30   27   10       8           6   7    5   7

Jun 32 40 12 6 4 3 1 2
Jul. 25 45 13 7 5 3 1 1
Aug 30 32 14 8 9 4 2 1
sep 24 32 18 9 7 5 3 2
oc- 19 33 20 10 7 4 25
Nov 13 43 20 8 6 3 3 4 De:: 18 57 15 5 3 1 <1 1 Ar~:;.ua: 22 40 14 8 6 4 2 4
2 mo:-,ths of data

                                       '.l. of 1 SGS-UFSAR                                                        Revision 18 April 26, 2000

TABLE 2.3-10 RELATION BETWEEN LAPSE RATES AND TURBULENCE CLASSES (percent) I Turbulence Temeerature Difference! T300-T33 Ft ((>F)

                     -1.6     -0.4  0.6   1.6   2.6   3.6 to       to   to    to    to   to

S-1.7 -0.5 -0.5 1.5 2.5 3.5 4.5 ~4.6 I 5.6 3.2 0.5 0.1 0.1 0.1 0.1 0.1 II 15.4 26.4 7.3 3.1 1.6 0.9 0.4 0.6 III 0.7 5.9 2.8 1.0 0.6 0.4 0.1 0.2 IV 1.0 3.7 4.5 3.8 3.6 2.7 1.5 2.4 1 of Revision 18 April 26, 2000

TABLE 2. 3-11 AVERAGE HORIZONTAL RANGE (Degrees) Month I II III IV All Jan 60 30 20 <10 25 Feb. 60 30 20 <10 30 Mar 70 30 20 <10 25 Apr 60 30 20 <10 30 May 70 25 20 <10 25

 *Jun         55     25            20         10          25
 *Jul         65      25           15         10          20
 *Aug         65      20           20         10          20
 *Sept        60      25           20         10          25
 *Oct         60      30           20       <10            25
 *Nov         55      30           20        <10           30 Dec        50      30           20        <10           30 Annual     60      30           20        <10 sigma       12      6          3-4        <2
 *2 months of data

SGS-UFSAR 1 of 1 Revision 6 Februa~ 15, 1987

TABLE 2.3-12

- AVERAGE HORIZONTAL RANGE (DEGREE) FOR WIND DIRECTIONS BETWEEN 130 AND 160 DEGREES Turbulence Class Month I II III IV All Jan 90 40 20 <10 10 Feb 80 30 20 <10 10 Mar 60 30 30 <10 10 Apr 50 40 20 <10 40 May 70 30 20 <10 30

  *Jun         70         30           20        10           30
  *Jul         60         30           20         10          20
  *Aug          70        30           30       <10           30
  *Sept         70         30          30       <10           30 Oct         60         30           20      <10            20 Nov         60         30           30      <10            30 Dec         60         30           30                     30 Annual      70         30           20-30     10
   *2 months of data 1 of 1 SGS-UFSAR                                         Revision 6 February 15, 1987

TABLE 2. 3-13 PERCENTAGE FREQUENCY OF WIND SPEED CLASSES 33ft Wind Speed Turbulence Class Calm 2-3 4-7 8-12 13-18 19+ All I 0.6 2.5 4.4 1.7 0.3 0.0 9.5 II 0.7 4.1 20.9 20.0 8.6 1.8 56.1 II! 0.0 0.3 2.6 5.3 2.6 0.7 11.4 IV 1.4 4.2 11.3 5.0 0.9 0.1 22.9 All 2.8 11.1 39.2 32.0 12.3 2.6 100.0 I 0.7 1.9 4. 1 2.1 0.6 0.2 9.6

r 0.2 1.1 7.2 18.0 18.6 11.4 56.5 r:: 0.0 0.0 0.1 0.9 4.8 6.0 11.8

v 0.4 l.O 3.8 ; .1 6.8 3.1 22.2 n..:...:. l~3 4.0 15.2 28.l 30.8 20.8 100.0 1 of 1 SGS-UFSAR Revision 18 April 26, 2000

TABLE 2.3-14 MEAN ANNUAL WIND SPEEDS AT VARIOUS LEVELS (mph) r Turbulence Class 33 ft 300 ft I 5.0 6.0 II 8.0 13 . .() III 10.0 19.0 IV s.o 12.0 All Hours 7.0 13.0 1 of 1 SGS-UFSAR Revision 18 April 26, 2000 f

TABLE 2.3-15 WIND DATA RECOVERY JUNE 1969 - MAY 1970 (percent) I Month 33-ft Level 300-ft Level Jun 1969 85 85 Jul 67 67 Aug 92 85 Sep 64 65 Oct 96 97 Nov 86 96 Dec 93 94 Jar. 1970 - 89 99 Feb 86 86 Ma:::- 78 78 Ap!:" 90 23

. Y

~ ~a 98 84 86 81 1 of 1 SGS-UFSAR Revision 18 April 26, 2000 r

TABLE 2.3-16 METEOROLOGICAL INSTRUMENTATION Height Above Tower Base, ft Sensed Parameter Recorded Parameter 300 Wind speed Wind speed Wind direction Wind direction Temperature{1) Temperature difference Humidity Relative humidity 150 Wind speed Wind speed Wind direction Wind direction Temperature{2) Temperature difference 33 Wind speed Wind speed Wind direction Wind direction Temperature T300 - T33(1) Differential T300 - T33(2) Dew point Dew point Temperature ambient Temperature Humidity Relative humidity 6 Barometric pressure Barometric pressure Solar Radiation Solar radiation 3 Rainfall Rainfall Backup Wind speed Wind speed Tower Wind direction Wind direction 33 (1) Temperature taken as part of temperature differential measurement T300 - T33. (2) Temperature taken as part of temperature differential measurement T150 - T33. 1 of 1 SGS-UFSAR Revision 33 October 24, 2022

TABLE 2.3-16A THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 21 December 6, 2004

VEt~T TABLE 2.3-17 RELEASE .. EXIT VELOCI'l'Y OF 7.2 H/SECOUDS tnJDEPI;TF.D X/g AT GROUND LEVEL APPLICABLE TO LONG TERH {ROUTttm) GASEOUS RELEASES (SECONDS/M3).

SECTOR AUNUAL- X/g AT GROUND LEVEL IUSUHCE SECTOR Bf.AiliNR(DEGREES) *

                                                                                                                                                                .tao,o
                                           .. ..o,.
"llE:l                   22,13*              4!hO                   bl,S                 90,0           .ll~.s           135,0               l!iT,,
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Compass Direc~l.on
sGS-UFSAR REVISION8 FEBRUARY15,1987

TABLE 2.3-18 VEHT RELEASE .. EXIT VELOCITY OF 7. 2 l.f/SECOUDS m~DEPLETED X/g AT GROUND LEVEL APPLICABLE TO LONG TERM ~ROUTINE} GASEOUS RELEASES

                                                                                      . 'SECONDS[Ml2 SECTOR MfNUAL X/Q AT GROUND LEVEL OUTU*CI!                                                                        SEeTo~ 8EAUlNG(O£GRE~9) l4Jl.(S                 i02,5                 ?.cc;,n                  l47,s                  270,0                   292.15               lts.o                 337,5                     :J&n,e
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S,~CI ' , n';)M*

r,d ll,~li'f*OR J,U7U*OR 4,tJ37E ..Oft 3,lni!F.*o& 'i,S14f.to08 7,l72f .. C)~ 7,71U\(*O;

   ~.lttJ     ~. t'tkl-.*nf\     n,lt87f.*OR                l, J liti:.*OR         4. 2llltf *08           J,l\0£ .. 01\        'i,i!&3E*OR            h 1 A4llt: *OA           1 1 ~~~~(*UI SGS-UFSAR.                                                                                                                                                         REVISIONI FEBRUARY15,1917

              **                                                                        TABLE 2.3-19 VEUT llliLEASE - EXIT VELOCITY OF 7. 2 t1/SECOUOS
                                                          *m~DEPLETED X/g ~GROUND LEVEL APPLICABLE TO LONG TERM ~ROUTINE} GASEOUS RELEASES
                                                                                           !SECONDS/M3}

SECTOR MfNUAL X/g AT GROUNO*LEVEL HSTU4CF SfC TfJ!l UE~RlN~(OEGA~fS)

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Ta.3*20 VENT RELEAS~
EKJT VELOCITY OF 7.2 M£SECONDS UNDEPLETED KlO AT GROUND LEVEL ~PPLJCABLE TO LONG TERM ~ROUTINEl GASEOUS RELEASES

                                                   'SECONDStK!

SECTOR ANNUAL X/Q AT GROUND LEVEL SECTOR BEARING (DEGREES) DISTANCE MILES 202.5 225.0 247.5 270.0 292.5 315.0 337.5 360.0 3.70 5.158E*08 4.478E*08 3.164E*08 4.048£-08 3.068E*08 5,030E*08 6,538E-08 7.093E*08 3.80 4.934E-08 4.283E*08 3.026£*08 3.870£*08 2.935E*06 4.813E*08 6.254E*08 6.782E*08 3.90 4.725E*08 4.101E*08 2.897E*08 3.706E*08 2.812E*08 4.610E*08 5.989E*08 6.492E*08 4.00 4.531E*08 3.931E*08 2.777E*08 3.553E*08 2.696E*08 4.420E*08 5.741E*08 6.220E*08 4.10 4.348E*08 3.772E*08 2.665£*08 3.409E*08 2.588E*08 4.243E*08 5,508E*08 5.966E*08 4.20 4.177E*08 3,623£*08 2.559E*08 3.275E*08 2.486E*08 4.076E*08 5.291E*08 5.729E*08 4.30 4.016E*08 3.483E*08 2.460E*08 3.148£*08 2.391E*08 3.919£*08 5.086E*08 5.505E*08 4.40 3,865E*08 3.351E*08 2.367E*08 3,030E*08 2.301E*08 3.nze-os 4.894E*08 5.295£*08 4.50 3.723E*08 3.227E*08 2.280E*08 2.918£*08 2.216£*08 3.634E*08 4.713E*08 5,098E*OB 4.60 3.588£-08 3.111£*08 2.198E*OB 2.813£-08 2.136E*08 3.503!*08 4.542£*08 4.912£*08 4,70 3.462E*08 3.001E*08 2.120E*08 2.714E*OB 2.061E*OB 3.380E*08 4.38ZE*08 4.737E*08 4.80 3.342E*08 2.897E-08 2.047E*08 2.620E*08 1.989E*08 3.264£*08 4.230E*08 4.571£*08 4.90 3.229E*08 2.798E*08 1.977E*08 2.531E*08 1.922E*08 3.153E*08 4.086E*08 4.414E*08 5,00 3.122£*08 2.705E*08 1.912E*08 2.447£*08 1.858£*08 3.049E*OB 3.950E*08 4.266E*08 5.10 3.021E*06 2.617E*08 1.850E*08 2.368E*08 1.797E*08 2.950E*08 3.821£*08 4.125E*08 5,20 2.924E*08 2.534E*08 1.791E*08 2.292E*08 1.739E*08 2.857E*08 3.698E*08 3.992E*08 5.30 2.833E*08 2.454E*08 1.735E*08 2.221E*08 1.685E*08 2.768E*08 3.582£*08 3.866E*08 5.40 2.746E*08 2.379E*08 1.682E*08 2,153E*08 1.633E*08 2.683E*08 3.472E*08 3.746E*08 5.50 2.663E*08 2.307E*08 1.632E*08 2.088E*08 1.583E*08 2.602E*08 3.367E*08 3.631E*08 6.00 2.305E*08 1.996£*08 1.413E*08 1.808E*08 1.367£*08 2.253E*08 2.911£*08 3.136£*08 7.50 1.597E*08 1.384E*08 9,851E*09 1.256E*08 9.372E*09 1.563E*08 2.010£-08 2.15BE*08 10.00 1.005E*08 8,740E*09 6.296E*09 7.964£-09 5.746E*09 9.856E*09 1.257E*08 1.344E*08 15.00 5.303E*09 4.640E*09 3.419E*09 4.256E*09 2.875£*09 5.211E*09 6.550£*09 6.957E*09 20.00 3.366E*09 2.958£*09 2.212E*09 2.726E*09 1.757E*09 3.312E*09 4.123E*09 4.361E*09 25.00 2.356E*09 2.076E*09 1.568E*09 1.919E-09 1.199E*09 2.320£*09 2.870E*09 3,027E*09 30.00 1.754£*09 1.549E*09 1.178E*09 1.435E*09 8.768£*10 1.729£*09 2.129£*09 2.241E*09 35.00 1.364E*09 1.206£*09 9.211E*10 1.119E*09 6.731£-10 1.345E*09 1.651E*09 1. 735E*09 40.00 1.095£*09 9.692E*10 7.427E*10 8.999E*10 5.352£*10 1.081£*09 1.323E*09 1.388E*09 45.00 9.012E*10 7,981E*10 6.132E*10 7.417£*10 4.373E*10 8.896E*10 1.087E*09 1.139E*09 50.00 7.564E*10 6.703E*10 5.160E*10 6.233E*10 3.650E*10 7.469£*10 9.110£*10 9.540E*10 1 of 1 Revision 15 SGS*UFSAR June 12, 1996

TABLE 2.3*21 ACClDENT X/Q ESTIMATES Csec/m3 ) EAB (0.79 Miles) 2 Hours 8 Hours 16 Hours 3 Days 26 Days Annual Conservative Estimate 1.30E*04 6.07E-05 4.15E-05 1.82E*05 5.55E-06 1.30E*06 Realistic Estimate 3.00E*05 1.79E*05 1.38E-05 7.87E*06 3.51E*06 1.31E-06 LPZ (5.0 Miles) Conservative Estimate 1.86E-05 7.76E*06 5.01E*06 1.94E*06 4.96E-07 9.37E*08 Realistic Estimate 2.35E*06 1.38E-06 1.06E-06 5.93E*07 2.59E-07 9.37E-08 1 of 1 Revision 16 SGS*UFSAR January 31, 1998

TABLE 2.3-22 ACCIDENT X/Q VALUES AT LPZ BY SECTOR 3 (HC/ffl) Sector 0.5 percent (2) Annual Bearing X/Q X/Q tiNE 8.20E*06 7.26E*08 NE 9.20E*06 7.7.5£*08 ENE 8.80E-06 6.23E*08 E 7.70E*06 6.041;-08 ESE 7.00£*06 6.11£-08 SE 8.40E-06 8.27E*08 SSE 8.40E*06 7.761:*08 s 1.00E*05 8.02E*08 ssw 1.20£*05 8.93E*08 sw 1.20£-05 8.77f*08 WSW 1.05E*05 6.62E*08 w 9.40£-06 5.42E*08 WNW 9.50E*06 5.10E*08 NW 1.86E-05 (1) 9.37E*08 NNW 1.40E*05 8.59£*08 N 8.50E*06 6.61E-08 OVerat l 5 percent 1.29£-05 (1) 1.86E*05 is the .axinum 0.5 percent X/Q <Conservative at the LPZ) (2) Two Hour value 1 of 1 Revision 16 SGS*UFSAR January 31, 1998

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CLASS 1 LARGE, LAZY CONVECTIVE EDOIES CAUSED BY KEATING AIR CLOSE TO THE GROUND. MOST fREQUENT ON SUMMER MORNINGS .WHEN WINO SPEEDS ARE LIGHT ANO LAKE BREEZES ARE HOT PRESENT. CLASS II TYPICAL DAYTIME TRACE HAVING A HIXTURE Of CON* VECTIVE AND MECHANiCAl TURBULENCE. flUCTUATIONS

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APPENDIX 2.3A THIS APPENDIX INTENTIONALLY DELETED SGS-UFSAR Revision 18 April 26, 2000

2.4 HYDROLOGIC ENGINEERING 2.4.1 Hydrologic Description 2.4.1.1 Site and Facilities The site is located on an irregularly shaped prominence in the Delaware Estuary. It is believed that hydraulic fill, dredged from the Delaware River or Bay was placed on and between two small bars. The preconstruction configuration of the area is shown on Figure 2.4-1, Map of Area. The area was and is quite flat, previously having an average elevation of about 9 feet above sea level. This was raised slightly in the plant area, to Elevation +10.5 Mean Sea Level (MSL) or 99.5 Public Service Datum (PSD). A levee, about 10 feet high, had been constructed around most of the westerly bar. As subsequently discussed, this levee became the basis for a protective sea wall. The predominant form of vegetation is Phragmi tes, a rather tall reed-like grass which is characteristically found in low-lying wetlands in the region. Aside from the access roads and bridges, the only modification to the island and the adjacent river and marsh area is within the station construction area in this area. The site grade has been raised about 1 1/2 feet except for the protective structures at the shoreline. There is a slight gradient toward the Delaware Estuary. The present configuration of the site is shown on Plant Drawing 232091. There was no established systematic surface-drainage system on the site prior to construction. Precipitation either ran off to the Delaware Estuary in a random pattern or collected in puddles where it infiltrated into the ground or evaporated. All surface drainage at the site flowed directly into the Delaware Estuary. 2.4-1 SGS-UFSAR Revision 27 November 25, 2013

The island upon which the site is located is separated from the New Jersey mainland by Hope Creek, a tidal stream which connects Alloways Creek with the Delaware Estuary. Hope Creek drains a rather large marsh, and has undergone some channel dredging and straightening. It is a brackish water stream and is used to a small extent for fishing and hunting. Studies of historical high and low water elevations indicated a maximum high water mark of 8. 5 feet MSL datum, Sandy Hook ( +97. 5 PSD), and minimum water level of -5.9 feet MSL datum (83.1 PSD). Station structures have been designed to not only withstand extreme recorded water levels, but also postulated extreme conditions, as subsequently discussed. Safety-related structures have been designed as follows:

1. The service water pumps can operate to a low water level of 76 feet PSD.

2. The service water structure is shown on Plant Drawing 211612. The portion of the service water intake enclosing the pumps, motors, and vital switchgear is watertight up to Elevation 126 feet PSD with wave runup protection to elevation 128 feet PSD. The service water intake can also withstand the static and dynamic effects of the storm. Each vertical, turbine type service water pump column bowl and suction bell is installed in an individual chamber which is open to the river. The chamber is isolated from the watertight compartments where the pump discharge heads and motors are located.

The pump discharge heads are bolted down to pads at Elevation 92 feet 6 inches. The joint between the pump discharge head and the pad at Elevation 92 feet 6 inches is watertight to prevent leakage of water into the compartments. Provisions have also been made to prevent leakage from the discharge head glands and leakoff connections into the watertight compartments. A sump pump is provided in each 2.4-2 SGS-UFSAR Revision 27 November 25, 2013

compartment to remove any accumulated water in the event a minor leak should occur.

3. All safety-related structures are watertight.

4. The Containment is, by nature, watertight and can withstand the static and dynamic loads associated with a storm producing a stillwater level of 113.8 feet PSD and the corresponding wave runup to 120.4 feet PSD (See Section 2.4.5 for the design storm water levels.)

5. The Auxiliary Building is watertight up to Elevation 115 feet PSD.

All doors in the outer Auxiliary Building walls below Elevation 120.4 feet are watertight. All watertight doors and structural walls can withstand the static and dynamic effects associated with a storm that produces a stillwater level of 113. 8 feet PSD with wave runup to Elevation 120.4 feet. Conduit penetrations above Elevation 115 feet and below Elevation 120.4 feet are packed to eliminate gross inleakage during the design storm. Each residual heat removal pump room, the lowest point in the Auxiliary Building, contains two sump pumps, each adequate to provide the minimum capacity of 50 gpm.

6. The main steam and feedwater pipe penetration area is watertight below Elevation 12 0. 4 feet. The structural walls and watertight doors are also capable of withstanding the static and dynamic effects of the storm 2.4-3 SGS-UFSAR Revision 27 November 25, 2013

which produces a stillwater level of 113.8 feet PSD and wave runup to 120.4 feet PSD. 2.4.1.2 Hydrosphere The station is located on the east shore of the estuarian zone of the Delaware River - Delaware Bay system. Delaware River flow enters the head of Delaware Bay 2 miles downstream of the site. The largest tributaries of the Delaware River are the Schuylkill River in Pennsylvania; the Christina River in Delaware; the Assunpink, Crosswicks, Rancocas, and Salem Rivers; and Big Timbers, Hope, and Alloways Creeks in New Jersey. The head of the Delaware Estuary is at Trenton, New Jersey, about 83 miles upstream of the site. The Chesapeake and Delaware Canal, which connects the Delaware River with Chesapeake Bay, is located about 7 miles north of the Salem site. Figure 2.4-4 presents the site location in relation to the surrounding area. The Delaware River has a drainage area of 12,765 square miles and its average freshwater discharge into the head of the estuary at Trenton is about 12,000 cfs (16,000 cfs at the site). The average tidal flow at Wilmington, Delaware, about 20 miles above the site, has measured at 400,000 cfs. Hence the tidal flow, which greatly exceeds the runoff flow, dominates the flow velocity at the site. The normal daily range in the height of the tide at the site is 5.8 feet. Larger fluctuations have been caused by hurricanes which bring heavy precipitation and may cause storm surges and severe wave action, and by strong northerly winds which push the Delaware River water into Delaware Bay. The highest tide ever recorded in the vicinity of the site (+8.5 feet MSL) occurred in November 1950. The lowest tide likely experienced, based on projections of data recorded at Reedy Point, Delaware, would have occurred on January 25, 1939 (-5.9 feet MSL). Hence, the maximum estimated historical tidal range is about 14.4 feet. 2.4-4 SGS-UFSAR Revision 6 February 15, 1987

The net tidal flow has been estimated at 400,000 cfs, which produces a relatively high current velocity in the station vicinity. Some small dams are in existence well upstream of the site (in New York State). Currently no major dams are planned for the river. As subsequently discussed (Section 2.4.2) the existence of dams upon the Delaware River does not influence the site safety analysis. The nearest public water supply is located about 8 miles northeast of the site. It utilizes both surface water and groundwater. There are five other public water supplies in New Jersey within 25 miles of the site and five in Delaware within 15 miles of the site. All are located upgradient from the site. Private water supplies in the area utilize groundwater as a source of water. The nearest producing well is located more than 2 miles from the site. There are 20 known wells in New Jersey within 4 miles of the site. All are located upgradient from the site. For a more detailed discussion of groundwater supplies, see Section 2.4.13. 2.4.2 Floods The water body to the west of the site is considered to be a tidally affected estuary by the U. s. Geologic survey. As such, water levels are recorded by tidal gauges and no "flood record" is kept. The tidal flow in the site area is estimated to be more than an order of magnitude greater than the average fresh water flow in the site vicinity. Thus, maximum and minimum water levels that may be of concern to plant safety were derived through considerations of coastal environmental conditions rather than riverine conditions. 2.4-5 SGS-UFSAR Revision 6 February 15, 1987

2.4.3 Probable Maximum Flood Not applicable, see Sections 2.4.2 and 2.4.5. 2.4.3.1 Probable Maximum Precipitation The maximum probable rainfall is of consideration only in design of yard drainage facilities and as a possible loading on critical structures, not as it may pertain to river flooding. The Yard Drainage System is designed to pass the drainage associated with a rainfall rate of 4 inches per hour for a period of 20 minutes (based on 90 percent runoff from paved areas and SO percent runoff from graded areas)

This rainfall intensity has a return frequency of 15 years (see Figure 2.4-5) and therefore, an unusually severe storm producing a rainfall rate in excess of 4 inches per hour for time periods of less than 20 minutes can be handled by the system.

In the unlikely event that the Yard Drainage System were to be loaded beyond its capacity, the excess water would accumulate and run off as the storm subsided. All doors and penetrations in the Class I (seismic) buildings are watertight up to Elevation 115 feet (PSD). The interior drains in the Auxiliary and Fuel Handling Buildings are independently piped to the Liquid waste Disposal System and are not connected to the Yard Drainage System. Roof drains are designed to dispose of a maximum rainfall rate of 4 inches per hour for a period of 20 minutes through the Yard Drainage System. Roof slabs are watertight to prevent building interiors from being damaged by severe rainstorms. The slabs are designed to withstand a loading equivalent to a depth of water up to the full height of the building* s parapet or roof curb. In the unlikely event that some of the roof drains become plugged, the backed up water will spill down the outside of the building. Wall penetrations above Elevation 115 feet (PSD) on Class I 2.4-6 SGS-UFSAR Revision 16 January 31, 1998

(seismic) buildings are designed to prevent roof spillage or heavy rain from seeping inside the building . In the event the capacity of the Yard Drainage System were to be exceeded as a result of an unusually severe rainstorm, the excess water would accumulate in puddles in the vicinity of the catch basins and run off. This water would not enter any safety-related structure, since these structures are watertight up to Elevation 115 feet (PSD). Therefore, safety-related equipment would not be adversely affected as a result of a severe rainstorm. 2.4.4 Potential Dam Failures Not applicable, see Sections 2.4.2 and 2.4.5. 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.5.1 Probable Maximum Winds and Associated Meteorological Parameters Probable Maximum Hurricane (PMH) storm surges have been calculated for the site using the bathystropic storm tide theory described by Marinos and Woodward (1968) (1). The hurricane surge was computed at the mouth of Delaware Bay and routed up the bay in accordance with a method described by Bretschneider (1959) (2). Components of the stillwater level are 1) the mean low water depth, 2) the astronomical tide, 3) the rise in water level resulting from the hurricane's atmospheric pressure reduction, 4) the wind stress component perpendicular to the bottom contours (onshore wind components), 5) the wind stress component parallel to the bottom contours which produces a longshore flow that is deflected to the right (in the northern hemisphere) by the Coriolis forces, and 6) the initial surge (a slow general rise in sea level existing before the actual hurricane winds arrive) . 2.4-7 SGS-UFSAR Revision 6 February 15, 1987

The PMH is defined by the U. S. Department of Commerce Report HUR 7-97 (3) as; "A hypothetical hurricane having that combination of characteristics which will make it the most severe that can probably occur in the particular region involved. The hurricane should approach the point under study along a critical path and at an optimum rate of movement." Indices used to calculate maximum storm surge are taken in part from HUR 7-97 where values are grouped according to defined coastal zones and by latitude within each zone. The following parameters and characteristics are based on empirical observations, assumptions, and experience. PMH indices and parameters include:

1. CPI (P ) - The maximum surface pressure in the center of 0

a particular hurricane, in inches of mercury.

2. Asymptotic Pressure (P ) - The surface pressure at the n

outer limits of the hurricane, in inches of mercury.

3. Radius of Maximum Winds (R) - The distance from the storm center to the point of maximum wind velocity in nautical miles.

4. Forward Speed (Vt) - Rate of forward movement of the center of the storm, in knots.

5. Maximum Wind Speed (UMax) - The absolute highest surface wind speed in the belt of maximum winds (measured as a maximum average 10-minute wind at a height of 30 feet above the water) calculated using equations from HUR 7-97.

6. PMH Path - The path selected for the PMH's approach is a critical factor, which in combination with other indices will determine the duration and magnitude of the storm winds over the critical fetch and the resulting peak hurricane surge elevation at the site. The path which produces peak hurricane surge will approach the area of 2.4-8 SGS-UFSAR Revision 6 February 15, 1987

interest normal to the general bottom contours. The hurricane's center will pass to the left (when facing shoreward) of the profile through the bay by a distance that allows the hurricane's maximum winds to pass directly over this profile.

7. Astronomical Tide (Ha) - Data for the predicted high astronomical tides are taken from the National Oceanic and Atmospheric Administration Tide Tables.

8. Initial Surge (Hi) - The initial surge is attributed to a tidal anomaly evaluated on the basis of variations between the observed and predicted tide. Data for initial surge as determined by the Coastal Engineering Research Center (CERC) were used.

9. Bottom Friction Coefficient (k) - The bottom friction coefficient is a function of several variables, among them the slope and width of the Continental Shelf in the area of study.

10. Wind Speed Adjustment Near Shore The computed overwater wind must be adjusted when moving onshore.

The overwater wind field was reduced, from its full value 2 miles offshore to 0.89 of its full value at the shoreline.

11. Wind Stress Factor - The wind stress factor is generally given as a function of wind speed, although other variables enter into its determination. The wind stress factor relationship suggested by CERC was used for the surge computations in this report.

Analyses were undertaken to predict the surge heights at the mouth of Delaware Bay generated by a PMH at latitude 39°N. Maximum surge elevation was calculated by moving the hurricanes across the continental shelf on a track normal to the bathymetric contours. 2.4-9 SGS-UFSAR Revision 6 February 15, 1987

The track of the postulated hurricane is shown on Figure 2.4-6. Two different forward speeds of translation were used to determine the effect that the rate of forward movement of the hurricane would have on the surge elevation. The PMH utilized in the analyses was a large radius, moderate forward speed hurricane which generated the maximum surge on the open coast. The quantitative meteorological parameters describing the PMH are:

1. CPI: 27.09 inches Hg

2. Peripheral Pressure: 30.72 inches Hg

3. Radius of Maximum Winds: 39 nautical miles

4. Maximum Wind Speed: 132 miles per hour

5. Forward Speed: 27 knots A computer program was developed by Dames and Moore using previous work by the Galveston District Corps of Engineers.

The program is described by Marinos and Woodward (1968) (1). Input data to the computer program describing the storm and the bathymetric conditions included the basic parameters of the hurricane, an initial surge of 1 foot, wind friction factor, bottom friction factor (0. 008), wind speed at various radial distances and angles of wind direction relative to the translational velocity vector of the hurricane, bathymetric traverse data and astronomical tide (5.6 feet). Winds which approach the site from a direction off the axis of the bay produce a component which is perpendicular to the axis of the bay. This cross-wind component causes the water surface to be raised on the upwind side of the bay and depressed an equal amount on the downwind side of the bay. 2.4-10 SGS-UFSAR Revision 6 February 15, 1987

As the PMH is moved along its postulated track, wind speed and direction at the site change because of the effects of friction and filling over land and also because of the position of the storm center with respect to the site. The cross-wind effects were calculated for the six wind directions chosen for analysis. The six wind directions or fetches radiate downbay from the site at IS-degree intervals from the east bank of Delaware Bay. The calculations consist of determining the corrected wind speed along the fetch, the cross-wind component of the wind speed, and the resulting cross-wind setup or drawdown. A summary of the calculations for each of the fetches is presented in Table 2.4-1. The wind speed was corrected to include the effect of the fetch distance from the storm center and also for friction and filling overland. The computer maximum surge elevation at the mouth of Delaware Bay was 21.9 feet above mean low water. This surge included the effects of the astronomical high spring tide . The maximum surge of 21.9 feet above mean low water at the mouth of Delaware Bay was routed to the site using the procedure of Bretschneider (1959) (2). The model surge hydrographs for Delaware Bay computed by Bretschneider were then used to determine hurricane surge values at the Salem site (which is within Bretschneider's Section 4) as a function of time. The maximum stillwater elevation at the site is a combination of the storm surge and the crosswind setup or drawdown. Storm surge elevations have been calculated for the six fetches chosen and are presented in Table 2.4-1 with the computed crosswind setup and the maximum stillwater elevation at the site. The six wind fetches radiate downbay from the site at IS-degree intervals from the east bank of the Delaware Bay. Subsequently, site hydrologic design parameters were developed using a maximum surge elevation of

2. 4-11 SGS-UFSAR Revision 6 February 15, 1987

113.8 feet PSD, as recommended by the Nuclear Regulatory Commission consultants. Table 2.4-2 contains a list of agencies and individuals contacted relative to this section. 2.4.5.2 Surge and Seiche History A review of local tidal gage history indicates that the maximum recorded water level was +8.5 feet MSL. It was recorded in November 1950. The lowest recorded level reached -5.9 feet MSL on January 25, 1939. The lowest "historic" water levels at the site that could be postulated from projections of data recorded in Philadelphia (December 31, 1962) (4) is -8.0 feet MSL. 2.4.5.3 Surge and Seiche Sources The most severe storm postulated for the site is the PMH. The PMH indices developed by the U. S. Department of Commerce studies (Memorandum HUR 7-97) (3) and utilized by CERC were described in Section 2. 4.5. I. 2.4.5.4 Wave Action The primary factors influencing the generation of waves will be the maximum wind speed over the water, the effective fetch length, and the average depth of water along the fetch. The values of these parameters used in the computations of wave heights and periods were determined for the fetches analyzed by:

1. Determining the location of the center of the storm required to produce winds along the fetch,

2. Calculating corrected wind speeds to account for friction and filling over the land and distance from the storm center to the fetch center, 2.4-12 SGS-UFSAR Revision 6 February 15, 1987

3. Calculating the still water elevation at the center of the fetch due to storm surge at the time the storm center is located to produce the maximum wind speed along the pre-selected fetch,

4. Computing the average depth along the fetch.

The basic assumptions used in the analyses were:

1. Storm generated waves from the open sea are dissipated at the mouth of Delaware Bay.

2. Steady state waves are generated along each fetch (these waves are independent of time).

3. Only the area northwest of Ben Davis Point generates significant wave energy at the site.

The PMH was located so as to produce maximum waves. In the vicinity of the site, the PMH winds had a maximum sustained wind velocity of 85 miles per hour from the southeast. With the surge level at 113.8 feet PSD, wave runup elevations on safety-related structures inside the sea wall were calculated to be a maximum of 120,4 feet PSD. Maximum wave run up elevation on the service water intake structure was calculated to be 127.3 feet PSD. 2.4.5.5 Resonance As a result of the nature of the estuary upon which the site is located, resonance was not a necessary consideration. 2.4.5.6 Runup As noted in Section 2.4,5.4/ maximum wave runup elevation was calculated to be

 +120.4 feet PSD on critical structures inside the sea wall and 127.3 feet PSD on the service water intake structure. The Sainflou method was used, assuming a minimum sea wall height of Elevation 108 feet PSD in the most critical area.

As discussed in Section 2.4.1, all safety-related structures are protected for water levels to equal or greater elevations. I

2. 4-13 SGS-UFSAR Revision 23 October 17, 2007

Security-Related Information - Witheld Under 10 CFR 2.390 2.4.5.7 Protective Structures

The stability of the dike was checked by Dames and Moore, using a computer program based on the Fellinius method of slices under the effect of the assumed wave forces. Some of the softer soils in the previously existing dike area were replaced with granular fill. 2.4.6 Probable Maximum Tsunami Flooding The occurrence of tsunamis is infrequent in the Atlantic Ocean. Other than the tidal fluctuation recorded on the New Jersey Coast during the Grand Banks earthquake of 1929, there has been no record of tsunamis on the northeastern United States coast. 2.4-14 SGS-UFSAR Revision 23 October 17, 2007

The earthquake of November 18, 1929, on the Grand Banks about ~70 miles south of Newfoundland, resulted in a tsunami which struck the :. south end of Newfoundland about 750 miles northeast of the Massachusetts coast.

The tsunami occurred at a time of abnormally high tide and resulted in some loss of life and destruction of property. The effect of this tsunami was recorded on tide gages along the United States east coast, as far south as Charleston, South Carolina. A tidal fluctuation of approximately nine-tenths of one foot was noted at Atlantic City, New Jersey and Ocean City, Maryland.

The Lisbon earthquake of November 1, 1755, produced great waves, which contributed heavily to the destruction on the coast of Portugal. These waves were noticeable in the West Indies. It had been reported that the Cape Ann, Massachusetts, earthquake of November 18, 1755, caused a tsunami in Saint Martin's Harbor in the West Indies; however, there is no record of a tsunami occurrence along the east coast of the United States at this time and it has since been determined that the Saint Martin's Harbor report actually refers to the tsunami caused by the Lisbon earthquake, which occurred within three weeks of the Cape Ann shock. Some tsunami activity has occasionally followed earthquakes in the Caribbean, but none of these was reported in the United States

There is no evidence of surface rupture in East Coast earthquakes and no history of significant t;sunami activity in the region. Hence, we do not )

believe that the plant site would be subjected to any significant tsunami effect. The maximum expected tsunami would result in only mibor wave action, and the maximum expected storm wave effect is the critical factor in design. 2.4.7 Ice Flooding Ice barriers are provided for the service water intake structuce. Surface ice jams will not exert direct structural loaCV.ng. The barrier wd.ll also enable the intake components to operate normally without the effect of ice.

I 2.4-15 SGS-UFSAR Revision 23 October 17, 2007

2.4.8 Cooling Water Canals and Reservoirs The Delaware Estuary is the cooling water reservoir for the plant. For discussions of the design parameters intended to provide a secure source of water, see Sections 2.4.1.1, 2.4.2, 2.4.3, 2.4.5, 2.4.10, and 2.4.11. 2.4.9 Channel Diversions As the source of cooling water is the Delaware Estuary, no channel diversions need be considered. 2.4.10 Flood Protection Requirements The relationship of hurricane induced surge and wave flooding and the site design parameters are discussed in Sections 2.4.1.1 and

2. 4.5. No other possible sources of flooding are as critical; hence, station design was predicated upon the worst possible meteorological event as previously described (Section 2.4.5).

2.4.11 Low Water Considerations 2.4.11.1 Low Flow in Rivers and Streams Not applicable, see Sections 2.4.2 and 2.4.5. 2.4.11.2 Low Water Resulting from Surges, Seiches, and Tsunamis The anticipated minimum stillwater elevation for the Delaware River Estuary in the vicinity of the Salem Nuclear Generating Station is -10.6 feet MSL. This extreme water level was developed from critically locating a postulated PMH (HUR 7-97) (3). The PMH was located in its more severe position as follows: Latitude of storm center: 39 degrees north. 2.4-16 SGS-UFSAR Revision 6 February 15, 1987

1. CPI: 27.09 inches Hg

2. Peripheral pressure: 30.72 inches Hg

3. Radius of Maximum Winds: 39 nautical miles

4. Forward Speed: 0 knot

5. Maximum Wind Speed: 124 miles per hour The location of the storm center was chosen so that the radius of maximum winds from the northwest would coincide with the axis of the bay between the Salem site and the mouth of the bay. The location of the storm is shown on Figure 2.4-8.

The maximum winds associated with the PMH would be from the northwest (N45°W) along the axis of Delaware Bay when the stillwater level is at the postulated minimum. In the vicinity of the site, the maximum wind velocity would be 85 miles per hour . With the stillwater level at -10.6 feet MSL, the winds would generate waves having a significant wave height and period of 5.0 feet and 4.8 seconds, respectively. This would correspond to a maximum wave height of 8.3 feet. The waves would travel along the axis of Delaware Bay in the most critical condition. Routing these waves to the service water screen well structure, the waves will undergo the effects of refraction, diffraction, and breaking. With the maximum winds of 85 miles per hour from the northwest, local waves trying to refract into this wind would become unstable and break; therefore, the effects of refraction have been ignored. The offshore topography from the service water screen well indicates that during the PMH low water level, there would be exposed shoreline with a northwest alignment, adjacent and to the northwest of the service water screenwell, projecting about 2.4-17 SGS-UFSAR Revision 6 February 15, 1987

150 feet into the Delaware River from the entrance to the service water screen well. Waves coming from the northwest would diffract around this exposed point of land in reaching the screen well entrance. The significant wave height would diffract to 1.5 feet in height while the maximum wave height would first be subjected to breaking due to depth restrictions. A maximum nonbreaking wave of 6.5 feet would diffract to a height of 2.0 feet in reaching the screen well. As the diffracted waves pass the screen well entrance, they will undergo several severe effects causing the wave to become unstable and deformed in shape. Some of these effects are: further diffraction of the waves as they strike the protruding ice barriers and enter the individual service water pump channels, and the reflection of waves in several directions causing a confused sea state at the screen well entrance. To be conservative, the pump channel walls and the ice barriers were treated as a pile array. Using this assumption, the 1. 5 feet and 2. 0 feet wave heights would be reduced to 1.1 feet and 1.5 feet, respectively, as they entered the individual pump chambers. These waves then must travel 50 to 60 feet in reaching the service water pumps, passing through a trash rack, curtain wall, stop log guide, ladders, etc. Therefore, there essentially would be no wave action at the pumps, but only a choppy water level. Water level amplification due to resonance is negligible because the fundamental period of the pump channels is approximately 13 to 16 seconds and the only possible wave excitation would come from a high order harmonic, resulting only in ripples. It is concluded that the highest possible wave at the service water pumps is 0.8 feet to 1.0 feet in height resulting in a water level change of approximately plus or minus 0.5 feet. Therefore, the lowest instantaneous water elevation at the service water pumps is -11.1 feet MSL. 2.4-18 SGS-UFSAR Revision 6 February 15, 1987

2.4.11.3 Historical Low Water See Sections 2.4.2 and 2.4.5. 2.4.11.4 Future Control There are no provisions required for control of the flow in the Delaware Estuary area. 2.4.11.5 Plant Requirements Plant water requirements are predominantly determined by the need for heat dissipation within the plant. The primary heat removal system is the Circulating Water System. The monthly flow is about 10 9.6 x 10 gallons, total for both units. The Service Water System averages approximately 4.3 x 10 7 gallons per month (both units). Requirements in a safe shutdown mode are much less. However, even using operating flow as a criterion, the daily average plant requirement is only about one-eighth of the tidal flow. 2.4.11.6 Heat Sink Dependability Requirements Essentially, the ultimate heat sink is the Atlantic Ocean. The Water Intake System is designed to operate at the lowest postulated water level in the estuary (Elevation -13.1 feet MSL). Also see Sections 2.4.1, 2.4.11.2, and 2.4.11.5. 2.4.12 Environmental Acceptance of Effluents The significance of onsite release of effluent is also discussed in Section 2.4.13.3. Basically, the Delaware River Estuary will be the final recipient of onsite spills or operating discharge. As the water is brackish, there are no public water supplies affected by estuary flows . 2.4-19 SGS-UFSAR Revision 6 February 15, 1987

The Delaware Estuary behaves as a mixed estuary. It is essentially homogeneous vertically; salinity averages 10 to 15 ppt with vertical variations at a given point limited generally to less than 1 ppt. Some variation in salinity is observed across the estuary due to Coriolis Forces which tend to concentrate less-than-average salinities on the west (Delaware shoreline and slightly greater than average salinities on the east (New Jersey shoreline). As a well-mixed estuary, the tidal mixing is sufficiently vigorous to keep the vertical salinity stratification to a low value; thus the dynamic and kinematic processes, which govern salinity, act to produce a relatively one-dimensional salinity distribution until a point is reached in the lower Delaware Bay where the tidal velocities are low enough to permit a degree of vertical stratification to develop. In the lower bay, below the Salem Station, there is an extensive amount of nontidal circulation brought about by the combination of salinity gradients and meteorological conditions. However, above the site the classic salinity profile for the vertically homogeneous estuary is prevalent. The Pritchard-Carpenter Consultants have estimated secondary, or nontidal flow as it can relate to the dispersion of effluent below the Salem Station. Their information indicates that as the observer travels seaward from the upstream freshwater end of the Estuary, there is an increasing amount of nontidal circulation. The relationship of this nontidal circulation to the transport of materials seaward has not been quantitatively established for the Salem Station and is of interest only in a qualitative overview. Based on computations using the vertical salinity measurements taken in conjunction with biological assessments, the net nontidal circulation in the station vicinity due to Coriolis Forces, wind stress, and gravity-induced circulation, produces salinities on the order of one-third of those in the lower bay. Other estimates of nontidal flow as high as six times the net freshwater supply are suggested, but insufficient data are available to assess either the numerical accuracy or the significance of this phenomenon in relation to the dispersion and advection of 2.4-20 SGS-UFSAR Revision 6 February 15, 1987

effluents from the Salem Station. However, it is clear that surface flow at the site is to the Estuary and the Estuary is a well mixed body of water in direct connection with the Atlantic Ocean. 2.4.13 Groundwater 2.4.13.1 Description and Onsite Use On a regional basis, the site is located on the Atlantic Coastal Plain about 18 miles south of the Fall Zone. The aquifers of the Coastal Plain are almost entirely unconsolidated sand and gravel, and water is stored in and transmitted through the primary pore spaces between the sand grains. The most productive aquifers in the region are the Cohansey Sand and the Raritan and the Magothy Formations. Other aquifers include all or portions of the Wenonah and Mount Laurel Sands, the Englishtown Formation and the Vincetown Formation. Sands and gravels of Pleistocene and Recent Age are irregularly distributed throughout the Coastal Plain, but are used as aquifers only in a few areas adjacent to the Delaware River. A summary of the hydrologic characteristics of geologic formations in the regions is presented in Table 2.4-3, Hydrologic Characteristics of Geologic Formations. They are discussed in order of the youngest formation to the oldest. Additional geologic information is given in Section 2.5.1, Geology and Seismology. A total of six production wells have been drilled at the site. They are screened in Wenonah - Mount Laurel and in the Upper and Middle Raritan Formations. Average flow of the wells is 1000 gallons per minute (gpm) with a maximum anticipated requirement of 1400 gpm. The location of these wells is shown on Figure 2.4-9 . 2.4-21 SGS-UFSAR Revision 6 February 15, 1987

2.4.13.2 Sources At the time of the preparation of the original Safety Analysis Report (late 1960s), nearly all water used for consumptive purposes within 25 miles of the site was groundwater. With the exception of the highly industrialized Wilmington, Delaware area, the major use of water is for domestic and agricultural purposes. This situation has not changed significantly in recent times. Public Water Supplies There are six towns in New Jersey within 25 miles of the proposed site that have public water supplies. There are five public water supplies within 15 miles of the site. Data concerning these public water supplies are shown in Table 2. 4-4, Public Water Supplies in the Vicinity of the Site. The locations of these supplies are shown on Figure 2.4-10, Public Water Supplies in the Vicinity of the Site. Wells Nearly all domestic water supplies in this region are obtained from private wells. Most wells are 2 inches in diameter and greater than 75 feet in depth. The aquifer commonly utilized in the vicinity of the site is the Mount Laurel-Wenonah Formation. Information pertaining to these wells is presented in Table 2.4-5, Private Water Wells in Vicinity of Site. The locations of wells in the vicinity of the site are shown on Figure 2.4-11, Known Water Wells in New Jersey in Vicinity of Site. There are no known productive water wells within 2 miles of the site other than those installed by Public Service Electric & Gas (PSE&G) (see Section 2.4.13.1). There are three abandoned wells near the site. The wells are reported to be several hundred feet deep. The location of the offsite wells are shown on Figure 2.4-11; the onsite wells are shown on Figure 2.4-9. 2.4 .. 22 SGS-UFSAR Revision 6 February 15, 1987

The nearest residences to the site are about 3 miles distant. Their water supply is obtained from shallow driven wells, or, in some cases, is carried in along with other provisions. Most water wells inventoried were located 3 to 4 miles from the site. The nearest wells in Delaware are more than 3 miles from the site and were not canvassed since it is believed that they would not be affected by a change in the groundwater regimen at the site because of the intervening Delaware Estuary. Site Groundwater The subsurface soils and groundwater conditions at the site are consistent with the regional picture. The upper soils at the site are dredged fills which were placed there by the United States Army Corps of Engineers around the turn of the century. The fill material apparently came from the channel of the Delaware River. Information obtained from test borings drilled on the site indicates the thickness of the hydraulic fill is generally less than 10 feet. Dames and Moore's report on Foundation Studies for Hope Creek Generating Station states: 11 At the surface, the hydraulic fill extends to a depth of about 30 feet below the present ground surface. The fill deposit is of man-made origin, having been deposited on the site as a result of channel maintenance in nearby areas ... " We have been calling the 30 foot upper layer as hydraulic fill all through the project work, including the correspondence with Nuclear Regulatory Commission. Dames and Moore's site subsurface section designated the upper 30 feet as hydraulic fill also. It is is of the same designation in "Engineering Seismology" (page 2-9) . 2.4-23 SGS-UFSAR Revision 6 February 15, 1987

The fill material is composed of a heterogeneous mixture of silt, silty clay, fine sand, and organic material. Four soil percolation tests were conducted on these materials to measure the absorption rate of the surficial soil. These tests were conducted in accordance with the U. S. Army Corps of Engineers' procedures. The absorption rate ranged from 1 to 4 gallons per day per square foot. The average rate was 2.7 gallons per day per square foot. Water levels are approximately at the level of the adjacent estuary waters. Below the hydraulic fill, a grey sandy and gravelly material, which formally comprised the bed of the Delaware River, was found. This layer varies in thickness from 2 to S feet and is composed of fine-to-coarse sand, a little fine-to-coarse gravel, and a trace of silt. The permeability of the sand, based upon particle size analyses, ranges from about 50 to ISO gallons per day per square foot. The clay facies is essentially impermeable. The lateral extent of this sand member is unknown, but it appears to exist in most of the site area. It is hydraulically connected with the Delaware Estuary, and water levels in this formation change in response to tidal variations. Water levels in this formation are essentially horizontal and although changes in response to tides do occur, the horizontal component of groundwater movement is small. The Kirkwood Formation of Miocene Age underlies the Quaternary soil and extends to about 70 feet in depth. It consists of gray silty clay and is an aquitard. Permeability values are less than SO gallons per day per square foot. The Vincetown Formation is about 45 to 75 feet thick and is encountered at a depth of about 70 feet. It consists of a fine-to-medium-grained sand with occasional gravel and is separated from the Quaternary soils by about 35 feet of impermeable silty clay of the Kirkwood Formation. Grain size analyses of this sand indicate a permeability of about 200 gallons 2.4-24 SGS-UFSAR Revision 6 February 15, 1987

per day per square foot. Water levels in this formation are essentially horizontal with an artesian pressure head just slightly lower than the surficial groundwater table. The horizontal component of groundwater movement in this formation is probably negligible, except for tidal oscillations. Two piezometers were installed about 75 feet from the Delaware Estuary to determine the tidal efficiency of the Vincetown Formation. Water level measurements were made in the estuary from high to low tide and corresponding measurements were made in the piezometers. Total tidal fluctuation amounted to 6. 3 feet, and the maximum variation in the piezometers was 3.9 feet. The time lag between peaks in the estuary and in the piezometers was about 20 minutes. The Vincetown Formation is underlain by the Hornerstown Sand which, according to published information, and information from the borings at the site, is an aquitard. Underlying the Hornerstown is the Navesink and Wenonah-Mount Laurel Sands. The Raritan-Magothy Formation is encountered at a depth of approximately 450 feet at the site. It consists of interbedded clays, gravel, and sands. The sand layers are generally 20 to 30 feet thick and the clay layers on the order of 100 feet in thickness. Fresh water was encountered in the sand layers to a depth of 900 feet at the site. At greater depths, the sands probably contain salt water. Although the site is underlain by sand and gravel formations which are utilized as a source of water supply in the region, these aquifers are separated from the surficial soils by one or more impermeable silty clay beds. Since the hydraulic gradient of these aquifers at the site is too small to measure, it is probable that the only groundwater movement at the site is a result of tidal influences. Except for production wells recently constructed at the site by PSE&G, there are no water wells within 2.4-25 SGS-UFSAR Revision 6 February 15, 1987

2 miles of the site, and the possiblity of offsite wells being affected by changes in the groundwater regimen at the site is remote. 2.4.13.3 Accident Effects In summary, the hydrological conditions at the site are well suited for the operation of the proposed power station. Fluid spills at the surface would be contained within the station drainage system or be drained toward the Delaware Estuary. All public water supplies in the Delaware are upstream of the site. Because of salt water intrusions, industrial use of the river water below Marcus Hook, some 25 miles upstream of the site is limited to cooling water applications. Thus radioactive wastes discharged to the river will remain well downstream of any industrial or domestic usage of river water. Any accidental spills that reached the subsurface would tend to move slowly to the southwest, although short-term reversals occur as a result of tidal fluctuations in the estuary. All water wells in the vicinity of the site are located upgradient. The closest domestic well is a shallow well located about 3 miles from the site. Movement of groundwater through the site is quite low as a result of the comparatively low coefficients of permeability and the low hydraulic gradients. Fluid infiltration in the area surrounding the actual construction site is low as many of the strata are relatively impermeable. Even in the station area, where the Pleistocene-aged and Miocene-aged Kirkwood Formation was removed, infiltration of fluids will be quite slow as the plant structures are founded on a lean concrete fill placed upon the Vincetown soils (which also have low permeabilities as a result of their cemented nature). 2.4-26 SGS-UFSAR Revision 6 February 15, 1987

The Vincetown is a fine to medium-grained calcareous sand, containing variable amounts of cementing material. The groundwater in the Vincetown is artesian and contains chloride concentrations of several thousand parts per million, thus, not suitable for drinking water. Below the Vincetown are the underlying Hornerstown and Navesink Formations which act as confining beds. A groundwater protection program was designed and implemented to provide reasonable assurance that a groundwater leak or spill of radioactive materials should be detected early and effectively remediated well before any potential impact to the offsite public health and safety or onsite workers. 2.4.13.4 Monitoring or Safeguard Requirements Surface and subsurface flow is toward the estuary. In general, infiltration and surface flow are slow. No public water supplies are down-gradient or downstream of the station. Thus, special monitoring or safeguard requirements are not necessary, 2.4.13.5 Consistent with Section 2.4.13.4, no technical have been No emergency plans, other than those presented in Section 13.3 are contemplated. 2.4.14 References for Section 2.4

1. Marinos, G. and Woodward, J. W., "Estimation of Hurricane Surge Hydrographs," American Society of Civil Engineers, Journal of Waterways and Harbors Vol. 94, No. WW2, pp. 189-216, 1968.

2. C. L., "Hurricane Predictions for Delaware and River," Beach Erosion Board, U.S. Army Coastal Engineering Research Center, Misc. Paper No. 4-59, November 1959.

2.4-27 SGS-UFSAR Revision 25 October 26, 2010

3. U.S. Dept. of Commerce, Interim Report Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coast of the United States: Environmental Science Services Administration, Memorandum HUR 7~97, May 7, 1968.

4. Lendo, A.C., "Record Low Tide of December 31, 1962 on the Delaware River, 11 U.S. Geological Survey Water Supply Paper 1586-E, 1966.

5. U.S. Army Coastal Engineering Research Center, Shore Protection Planning and Design, Technical Report No. 4, 3rd Edition, 1966.

2.4.15 Bibliography for Section 2.4 Anon, Geology of Northeast Corridor, USGS Mise. Map 1-814-C. Back, J.L., Hydrochemical Facies and Ground-Water Flow Patterns in the Northern Part of the Atlantic Coastal Plain, U.S.G.S. Professional Paper 498-4, 1966. Barksdale, H.C.; Greenman, D.W.; Land, S.W.; Hilton, G.S.; and Outlaw, D.E., Ground-Water Resources in the Tri-State Region Adjacent to the Lower Delaware River. New Jersey Division of Water Supply and Policy Special Report 13, 1958. Bretschneider, C.L., Storm Surges, Advances in Hydrosciences, Academic Press. Vol. 4, pp. 341-418, 1967. Marine, I. W. and Rasmussen W.C., Preliminary Report on the Geology and Ground-Water Resources of Delaware, Delaware Geological Survey Bulletin No. 4, 1955. Richards, H.G.; Olmstead, F.H.; and Ruhl, J.L., Generalized Structure Contour Maps of the New Jersey Coastal Plain, State of New Jersey, Department of Conservation and Economic Development, 1962. 2.4-28 SGS-UFSAR Revision 6 February IS, 1987

Rima, D.R.; Coskery, O.J.; and Anderson, P.W., Ground-Water Resources of Southern New Castle County, Delaware, Delaware Geological Survey Bulletin No. 11, 1964. Sundstrom, R.W., et al, The Availability of Ground-Water from the Potomac Formation in the Chesapeake and Delaware Canal Area, Delaware, University of Delaware Water Resources Center, 1967. U.S. Army Corps of Engineers, Sewage Treatment Plants, Engineering Manual 345-243, 1959. U.S. Army Corps of Engineers, Report on Hurricane Study, Delaware River and Bay, Pennsylvania, New Jersey and Delaware, Unpublished Report, North Atlantic Division, Philadelphia District Corps of Engineers, September 13, 1963 .

SGS-UFSAR 2.4-29 Revision 6 February 15, 1987

2.4.16 Agencies and Individuals Contacted Agency Location Individual U.S. Geological Survey Trenton, New Jersey Mr. H. Gill Water Resources Division Mr. H. Meisler New Jersey Division of Trenton, New Jersey Mr. J. C. Mearill Water Policy and Supply Coleman Well Drilling Co. Hancocks Bridge, Mr. P. Coleman New Jersey Vicinity of site Numerous local residents SGS-UFSAR 2.4-30 Revision 6 February 15, 1987

TABLE 2.4-1

SUMMARY

OF MAXIMUM STILLWATER ELEVATION DETERMINATIONS Maximum Wind Average Surge Maximum Speed at Angle of Fetch Elevation Stillwater Fetch Wind to Crosswind Depth Crosswind at the Elevation Fetch Center) Bay Axis Component dt Setup Site at the Site Number (mph) (degrees) (mph) (ft) (ft) (ft) (ft) 1 108.6 -13.0 24.4 39.3 0.00 109.2 109.2 2 113.3 2.0 4.0 39.3 0.00 110.9 110.9 3 112.2 17.0 32.8 38.0 0.00 109.2 109.2 4 108.6 32.0 57.5 37.9 0.08 106.5 106.6 5 106.6 47.0 78.0 35.6 0.25 104.3 104.6 6 106.0 62.0 93.5 37.4 0.34 101.8 102.1

1 of 1 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-2 AGENCIES AND INDIVIDUALS CONTACTED Agency I;ocation Individual U.S. Geological Survey Trenton, New Jersey Mr. H. Gill Water Resources Division Mr. H. Meisler New Jersey Division of Trenton, New Jersey Mr. J. C. Mearill Water Policy and Supply Coleman Well Drilling Co. Hancocks Bridge, Mr. P. Coleman New Jersey Vicinity of site Numerous local residents

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.4-3 HYDROLOGIC CHARACTERISTICS OF GEOLOGIC FORMATIONS (Youngest to Oldest Formations) Pleistocene Series: Pleistocene deposits occur in this region as thin discontinuous formations and are not a major source of water. Large capacity wells from these deposits are not feasible; however, infiltration galleries have been used in this formation where hydraulically connected to the Delaware River. Shallow wells draw water from these aquifers for domestic suppliers in some area .

 . Cohansey Sand:    The Cohansey Sand outcrops along a line trending northeast-southwest, about 6 miles east of the site.         The formation dips to the southeast and therefore is not present at the site. It is composed predominantly of well-sorted sand and   gravel,   and   is potentially the most productive aquifer in the Coastal Plain area.

Groundwater in the Cohansey Sand is largely unconfined. There is no significant regional pattern of water movement in the formation. The flow pattern is governed largely by local topography. Kirkwood Formation: The Kirkwood Formation immediately underlies the Pleistocene Soils at the site and dips to the southeast. It is composed of light gray clay with interbedded layers of sand. Domestic and farm water supplies are obtained from wells in the Kirkwood Formation. Yields on the order of 5 to 100 gallons per minute are obtained in the Kirkwood. A few pumping tests have been made in aquifers within the Kirkwood Formation, although none have been documented in the vicinity of the site. The nearest test on record (about 15 1 of 4 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-3 (Cont) miles to the. northeast) indicates a field coefficient of permeability of about 200 gallons per day per square foot. More often than not, the direction of the hydraulic movement in the Kirkwood Formation does not conform with the direction of dip. The major areas of discharge are probably in the permeable parts of the outcrop area where stream channels, swamps, and marshes provide relatively low-elevation discharge areas. A potentially large natural discharge area occurs where the Kirkwood Formation crops out in the Delaware River. This occurs at the site. Vincetown Formation: This formation is a minor but relatively important source of water in New Jersey. It crops out in the vicinity of the site and is composed of a semi-consolidated sand . In the vicinity of Salem, New Jersey, about 8 miles northeast of the site, wells in the Vincetown Formation have been reported to yield as much as 300 gallons per minute. This is in an area where the granular portion of the aquifer is thicker than normal. At the site the Vincetown Formation contains saline water. Navesink Formation: The Navesink Formation is composed of fine to medium-grained sand with some clay. It is not widely used as a source of water supply in the region. Hornerstown Sand: This formation is composed of sand and clay. It is not used as a source of water supply due to its impermeable nature. However, it is not a tight aquiclude and some vertical leakage may occur into or out of the underlying aquifer, depending upon the hydraulic gradient. Production wells tested at the site in 1970 confirmed that vertical 2 of 4 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-3 (Cont.) occurs in some areas due to in hydraulic Wenonah - Mt. Laurel Sands: These formations function hydrological as a unit; the Wenonah sand is composed mainly of fine to coarse-grained sand and is overlain by the Mt. Laurel sand which is characteristically a medium to coarse-grained sand. This unit is well utilized aquifer, used predominantly for domestic purposes. The aquifer recharges from precipitation and discharges predominantly in low outcrop areas. The aquifer outcrops beneath the Delaware a probable area. Since the is confined and withdrawal volumes are small, it is probably that very little water movement occurs. Operation of onsite wells in the Mount Laurel-Wenonah Formation will induce groundwater flow towards the wells. The Marshalltown Formation is composed of clay, is impermeable, and considered to be an aquiclude. This sand formation is not utilized as a source of of the site due to a amount of and silt in the formation. Its increases to the north and east, where it is tapped by wells up to 200 gallons per minute. Merchantville Clay: This formation is characteristically a clay or sandy clay overlain in many areas by the Woodbury clay, of similar characteristics. In combination with the Woodbury clay, it forms an effective aquiclude. 3 of 4 SGS-UFSAR Revision 25 October 26, 2010

TABLE 2.4-3 {Cont) This formation consists of sand with thin beds of silt and I t is a or in much of the area, i t is not utilized south and east of the site due to the high chloride content of its water. Aquifer coefficients, based on pump test data, indicate that the Magothy has a permeability value of about 400 gallons per day per square foot. Its porosity is about 45 percent and the specific yield is about 40 percent. The Potomac consists of an upper (Raritan Formation) and a lower fer (Patuxent Formation) by with sand lenses. The movement of groundwater through this formation is generally downdip, or southeast. This aquifer is not used in the vicinity of the site due to its depth and proximity to the salt water-fresh water interface believed to occur about 5000 feet east of the site. Source: Dames and Moore, 1970. 4 of 4 SGS-UFSAR Revision 25 October 26, 2010

TABLE 2.4-4 PUBLIC WATER SUPPLIES Average Population Output Town Served (mgd)* Source of Water Salem, New Jersey 9,000 1.7 About 2/3 of water consumed is surface water, pumped from the Quinton pumping station about 3 miles east of town and 9 miles northeast of the site. Remainder is obtained from four wells, ranging in depth from 80 to 168 feet, located east of Salem. Pennsville, New Jersey 10,500 Four wells ranging in depth from 105 to 240 feet. The wells are probably completed in the Magothy Formation. PennsGrove, New Jersey 8,000 Two wells, 292 and 360 feet deep. The water probably comes from the Potomac Group. Woodstown, New Jersey 3,000 Eight wells; six are about 100 feet deep and the others are about 300 and 350 feet deep. Elmer, New Jersey 2,500 Three wells; two are 80 feet deep and the third is 500 feet deep. The shallow wells probably tap the Mount Laurel-Wenonah Formation *

SGS-UFSAR 1 of 2 Revision 6 February 15, 1987

TABLE 2.4-4 (Cont) Average Population Output Town Served (mgd)* Source of Water Bridgeton, New Jersey 22,000 A total of 12 wells, some of which are no longer in use, range in depth from 75 feet to 129 feet. They are completed in the Cohansey Sand. Smyrna, Delaware 0.27 Two wells, 20 feet and 95 feet deep supply the town. The shallower well is used for standby purposes. Clayton, Delaware 825 1.2 One well, 272 feet deep, is the source of water supply. Middletown, Delaware 2,000 0.2 Three wells, having depths of 100 feet, 200 feet and 500 feet, supply the town. Delaware City, Delaware 1,500 0.2 Two wells, one 26 feet deep in the Wenonah Formation and the other in the Magothy Formation, supply the town. New Castle, Delaware The town obtains water from a shallow infiltration gallery system located in Pleistocene deposits.

        =millions

mgd of gallons per day 2 of 2 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-5 PRIVATE WATER WELLS IN VICINITY OF THE SITE Static Total Casing Water Well* Depth Dfameter Length Level No. Owner's Name (ft) (in.) (ft) (ft) Yield Remarks 1 Aloes Marina 252 2 220 3~ 2 Dr. Devlin 252 2 210 5 3 Dr. Devlin 252 2 230 2\ 4 Dr. Devlin 252 2 212 4 5 Mr. Henchman 252 2 218 6 6 G. Harbeson 15 42 Dug well 7 G. Harbeso~ 15 42 Dug well 8 F. Harris 12 36 8+ Four wells, Deepest is 32 feet. 9 F. Shimp 90 60+/- 12-13 10 T. Hilliard 90 6 60+/- 12-13 11 Mr. Snideker 10 36 7-8 12 Mr. Snideker 90 4 13 w. Ashlock 252 2 231 8 14 F. Schrier 90 4 60 12-13 15 B. Hendman 89 2 84 15

16. B. Hendman Well filled in.

17 State of N.J. 89 2 84 12 18 2 19 T. Dixon 156 2 147 3 20 Well abandoned. 21 T. Dixon 90 2 12 Well abandoned. 1 of 3 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-5 (Cont) Static Total Casing Water Well Depth Diameter Length Level No. Owner's Name (ft) (in.) (ft) (ft) Yield Remarks 22 . D. Harris 32 2 32 Flowing Well abandoned. 23 Mr. McCray 17 2 17 Flowing Water is salty. 24 Mr. McCray 165 2 147 5 25 J. Pancast 115 2 5-6 26 J. Pancast 89 2 82 4 27 R. Davis 14 36 6 Dug well. 28 w. Hancock 90 4 50 10-12 Iron, bad water. 29 Mr. Ingersol 90 4 50 10-12 30 L . Fonderbank 100 2 86 3 31 0. Ayrs 199 2 189 7 32 Stony Point 315+/- Well abandoned. 33 400+/- 34 900+/- 35 36 165 2 90 37 Eagle Island Gun Club 110 2 103 6 38 J. Dilkes 2 131 8 39 Public Service 298 16 243 20 200 Not in use (Production Well 3) 40 Public Service 284 16 210 200 Not in use (Production Well 4) 2 of 3 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.4-5 (Cont) Static Total Casing Water Well Depth Diameter Length Level

                                      . (ft)

No. Owner's Name (ft) (in.) (ft) Yield Remarks 41 Public Service 300 10 250 200 Intermit (Production Use for Well 1) Construction 42 Public Service 286 16 220 18~ 200 Not in use (Production Well 2) ' SGS-UFSAR 3 of 3 Revision 6 February 15, 1987

PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION UpdatedFSAR Mapof Area REVISION6 FEBRUARY15, 1987 Figure 2.4*1

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2.5 GEOLOGY AND SEISMOLOGY 2.5.1 Basic Geologic and Seismic Information The Salem site is located adjacent to and east of the Delaware River Estuary. It is approximately 19 miles south of Wilmington, Delaware, and 16 miles west of Bridgeton, New Jersey (Figure 2.5-1). The area investigated for the nuclear generating facilities covers approximately 170 acres. Most of the facilities are located in the southern half of this area. The scope of the geologic and seismologic phases of the Salem site was quite broad and encompassed the disciplines of geology, engineering, seismology, geophysics, and soil mechanics. The work was performed by Dames & Moore Consultants in coordination with locally knowledgeable individuals . The research included a review of available pertinent geologic literature and interviews with representatives of state and Federal agencies and individuals possessing local knowledge. A list of the agencies contacted and the publications reviewed to obtain the information contained in this portion of the report is presented in Table 2.5-1 and Section 2.5.6. Seismologic studies included literature research to compile a record of the seismicity of the area, an evaluation of the geologic structure and tectonic history of the region, and analyses to evaluate the response of the foundation materials to earthquake-type loading. Field geophysical studies were performed to aid in evaluating the in-situ dynamic properties of the foundation materials. In addition, a geologic reconnaissance of the site and surrounding area was made by engineering geologists. The site was 2.5-1 SGS-UESAR Revision 6 February 15, 1987

investigated in detail by drilling test borings and performing geophysical explorations. Laboratory tests were performed to aid in evaluating both the static and dynamic properties of the subsurface soils. Physiographically and tectonically, the site lies within the Atlantic Coastal Plain Province. The Coastal Plain has been described as a wedge shaped (thickening to the southeast) series of Cretaceous to Quaternary-aged sediments overlying Precambrian-aged basement rocks (see Figure 2.5-2). The site structures are founded on the Paleocene-Eocene Vincetown Formation) a competent, cemented, granular soil. Below the Vincetown are some 1800 feet of increasingly older sediments. The foundation soils will perform well under the anticipated static and dynamic loadings. The dynamic loads are expected to be low (the largest earthquake experienced in the regions surrounding the site is a Modified Mercalli Intensity VII earthquake). However, the plant is designed to withstand free field ground earthquake acceleration levels of 20 percent of gravity horizontal) and 13.3 percent gravity vertical. 2.5.1.1 Regional Geology 2.5.1.1.1 Physiography The site lies within the Atlantic Coastal Plain Physiographic Province, about 18 miles southeast of the Piedmont Physiographic Province. The Fall Zone marks the contact of the low lying, gently undulating terrain of the Coastal Plain and the higher, more rugged terrain of the Piedmont Province. The relation of the site to these Physiographic Provinces and the Fall Zone is shown on the Regional Physiographic Map (Figure 2.5-1). 2.5-2 SGS-UFSAR Revision 6 February 15, 1987

The Coastal Plain of New Jersey is a low plain rising from sea level to about Elevation +200 feet at the Fall Zone. Ground surface elevations in areas near the larger streams and in the coastal lowlands are generally less than 50 feet above sea level. The regional slope of the ground surface is to the southeast at approximately 1 1/2 feet per mile. The topography is characterized by a series of broad step-like terraces probably formed during Pleistocene time by a fluctuating sea level resulting in alternate deposition and erosion. The terraces are successively less dissected by steam erosion from the Fall Zone to the shoreline. The post-glacial sea level rise inundated the former shore of the Coastal Plain and drowned the lower portions of the streams. Delaware Bay and the estuary of the Delaware River, which extends inland as far as Trenton, have been formed by this sea level rise. It is estimated that the sea has risen approximately 300 feet since the retreat of the glaciers . 2.5.1.1.2 History and Tectonics The record of geologic history of the region starts with the deposition of sediments in Precambrian times. These sediments were metamorphosed subsequently to gneisses and schists. Contemportaneously, granites and other igneous rocks intruded these Precambrian sediments. Subsequently, these intrusives were metamorphosed to gneisses. In late Precambrian to early Paleozoic time (following the completion of the Grenvillian orogenic cycle) the Proto-Atlantic Ocean began to form. The process caused the separation of the North American and African plates. In this initial rifting phase an eastward thickening wedge of clastic sediments, interbedded with volcanic rocks, were deposited unconformably on the Grenvillian basement in water-filled basins within the ancient continental margins . 2.5-3 SGS-UFSAR Revision 6 February 15, 1987

As rifting progressed, the Proto-Atlantic Ocean opened and the previous system of isolated rift basins gave way to a long depositional trough underlain by oceanic crust. The history of the closing of the Proto-Atlantic is reflected in the convergent stage of the Appalachian Orogeny, beginning in Ordovician times. The early phases of this stage are evidenced by a pre-Middle Ordovician unconformity, followed by the influx of detrital sediments over the previous carbonate bank, along the margin of the craton. The close of the Taconic Orogeny marked the destruction of the ancient continental margin and the development of a mature arc-trench-subduction zone. During Triassic time, a series of elongated troughs were formed by faulting between the mountains uplifted by the Appalachian Orogeny, and the remnants of the old mountains to the west. Subsequently, the mountains were eroded and sediments were deposited in these troughs from both east and west. The long period of erosion and deposition extended into early Cretaceous time and reduced the region to a nearly level plain, exposing igneous and metamorphic Precambrian and early Paleozoic rocks with local areas of sedimentary and igneous rocks of Triassic age. The Post-Middle Triassic development of the Orogeny initiated the opening of the North Atlantic Ocean. This structural development (the youngest regionally recognizable diatrophism in the northeastern United States) is characterized by vertical movements and related continental and marine sedimentation, transcurrent faulting along pre-existing zones of crustal weakness and extrusive and intrusive igneous activity. In late Jurassic and early Cretaceous time, the region was downwarped to the east. The downwarping continued intermittently through Cretaceous and Tertiary time resulting in the present-day accumulation of sediments in the Coastal Plain. 2.5-4 SGS-UFSAR Revision 6 February 15, 1987

During the early Pleistocene period, river sand and gravel were deposited by glacial meltwater over much of southern New Jersey . During the last interglacial stage, higher sea levels resulted in the deposition, in the lowland areas, of marine and estuarine sediments. In the higher areas, streams deposited sand and gravel. The last continental glacier which extended into northern New Jersey resulted in the deposition of outwash material in streams and river valleys, such as the Delaware River. The post-glacial sea level rise submerged a large portion of the Coastal Plain marginal lands. Measurements indicate that sea level, on the average, has apparently risen approximately 0.3 cm/yr in the last 100 or so years. The surface exposure of these Precambrian to recent materials is presented on Figure 2.5-3, Regional Geologic Map. 2.5.1.1.3 Stratigraphy The sediments deposited on the downwarped basement in the Coastal Plain range from early Cretaceous to Quaternary in age and consist of inter-bedded silt, clay, sand, and gravel, of both marine and non-marine origin. These strata form a wedge-shaped mass which thickens to the southeast. The strata out crop near the Fall Zone and dip to the southeast as shown on Figure 2. 5-2, Geologic Section and Figure 2.5-3, Regional Geologic Map. Generally, each successively younger formation has a more gentle dip than that lying below, resulting in a decrease in slope upward in the sequence, from a crystalline basement dip of approximately 75 feet per mile, to about 10 feet per mile in the upper Tertiary formations. The decrease in dip is accompanied by gradual thickening of the strata to the southeast. As a result of sea level fluctuations during deposition, the unconsolidated sediments of the Coastal Plain exhibit considerable lateral and vertical variations in lithology and texture . However, since the ocean lay to the east during the accumulation 2.5-5 SGS-DFSAR Revision 6 February 15, 1987

of the sediments, they generally grade finer-grained to the east. Two periods of extreme sea level fluctuations occurred at the end of the Cretaceous and at the end of the Paleocene time. Both were times of widespread erosion and explain the absence of formations in some areas. 2.5.1.1.4 Structure The site is located near the edge of the Chesapeake-Delaware Embayment. To the north and west of the site is the highly folded and faulted Piedmont Province. To the north has been postulated the Cornwall-Kelvin Wrench Fault Zone. The Chesapeake-Delaware Embayment is an area of more extensive downwarping of the Atlantic Coastal Plain. It is marked by the re-entrance of the coastline and a deep accumulation of sediments. The areas of the greatest embayment exist in northern Maryland and in the vicinity of Long Island. Bedrock contours in the vicinity of the site are shown on Figure 2.5-4, Regional Tectonic Map. The present day area of the Chesapeake and Delaware Bays was affected by the formation of the embayment in early Cretaceous or, possibly in some areas, in late Jurassic. From that time on, this region was generally downwarped and accumulated sediments. Local shallow folding has been recognized in some of these sediments, but no faults have been identified within them. The folding may be related to depositional features rather than post-depositional tectonic activity. Eighteen miles northwest of the site, rocks of the basement complex crop out and mark the boundary of the Piedmont Physiographic Province. As discussed in Section 2.5.1.1.2, these rocks were subjected to significant tectonic activity in the form of intense folding, faulting, igneous intrusion, and metamorphism. As would be expected from the regional tectonic history, most of these structural features follow a strong northeast-southeast trend. 2.5-6 SGS-UFSAR Revision 6 February 15, 1987

The basement complex underlying the site is generally similar to the metamorphic and igneous rocks observed in the Piedmont . However, geologic information relative to the basement structure is limited, due to the thick sequence of sediments overlying the basement. Interpretation has been based on geophysical data and the relatively few deep wells penetrating the basement complex in the region. It is probable that there are also faults lying beneath the Coastal Plain sediments, likely following the same regional trend as observed in the Piedmont area. Some minor faulting of this nature was observed in the basement complex underneath approximately 300 feet of sediments in Gibbstown, New Jersey. This is approximately 25 miles north of the site and the closest approach of known faulting to the Salem site. No faulting has been identified in the Cretaceous sediments above the basement complex in this area. Therefore, it is probable that the faulting in the basement is Pre-Cretaceous (more than 135 million years in age) . One feature of interest, approximately paralleling the New Jersey coastline, is revealed only by geophysical data. This feature is the change in the rate of dip of the basement complex from approximately 75 feet per mile to over 200 feet per mile and more. Though the feature is fairly well documented, no really satisfactory explanation of its origin has been proposed for the New Jersey-Delaware area. Some geologists explain it relative to differential peneplanation, some to gradual flexing of the basement, or some due to faulting. The closest approach of this feature is about 55 miles east of the site (see Figure 2.5-4). Approximately 50 to 60 miles north of the Salem site, and transverse to the regional structural trend, is the postulated Cornwall-Kelvin Wrench Fault Zone (1). This zone has been mapped on the basis of subsea topography and geophysical surveys and has been inferred to extend through the Triassic Lowlands of southeastern Pennsylvania. It has been suggested that this fault may be part of a major east-west continental fault which extends 2.5-7 SGS-UFSAR Revision 6 February 15, 1987

from the mid-United States to 300 miles beyond the present Atlantic shoreline. A 94-mile, right lateral offset of sedimentary basins and belts of magnetic anomalies has been determined by oceanographic surveys near the 40th parallel in the ocean basin and onto the continental shelf and slope. However, there is neither geological or geophysical evidence of a continuation at this fault in the continent at the surface, or at depth (2). No disturbance has been observed in the Cretaceous and younger sediments in this zone. Again, it appears that any possible faulting has been Pre-Cretaceous in age. As previously noted, the site is located in the inner plain of the Coastal Plain Physiographic Province. The Coastal Plain Physiographic Province has also been accepted as a tectonic province in accord with definitions in Appendix A to 10CFR100. This physiographic province is bounded on the east by the Atlantic Ocean and on the west by the Fall Zone and the Piedmont Physiographic (and Tectonic) Province. A generalized representation of the subsurface conditions in the site area is shown on Figure 2. 5-5, Geologic Columnar Section - Site Area. Thus, in summary, numerous ancient faults are likely in the basement rock. However, regional diatrophism ceased at least 85 million years ago and only minor fold-like structures appear in the sediments overlying the ancient basement. Considering the lack of Post-Cretaceous tectonic activity along the eastern seaboard of the United States, it is likely that the Post-Cretaceous features are the results of differential compaction over basement relief. 2.5.1.1.5 Groundwater See Section 2. 4. 13. 1 for a discussion of the local hydrologic conditions. 2.5-8 SGS-UFSAR Revision 6 February 15, 1987

2.5.1.2 Site Geology The site is located on the southern tip of what was once a natural bar in the Delaware Estuary, adjacent to the western shore of New Jersey. In the past, the bar and the area between the bar and mainland has been used as a disposal area for material dredged from the Delaware Estuary or River. No additional dredged material has been placed for at least the past 25 years. The subsurface conditions of the site area were investigated by 35 borings to depths of up to 200 feet. The locations of these borings are shown on Figures 2.5-6 and 2.5-7, Boring Plan and Boring Plan - Detail A. Stratigraphy developed from these borings is shown on Section A-A and B-B on Figure 2.5-8, Subsurface Sections. The deepest formation penetrated in the boring program was the top of the Mount Laurel Sand. The sands of this formation and those of the conformably overlying Navesink Formation mark the end of Cretaceous deposition. The top of the Navesink is an unconformity recording a period of widespread erosion. The Red Bank Sand, present in northern New Jersey, and part of the Navesink, were probably removed from southern New Jersey during late Cretaceous or early Tertiary time. During the Paleocene, silty glauconitic sands of the Hornerstown Sand and the clays, silts, and sands of the Vincetown Formation were deposited. The top of the Vincetown again marks a period of erosion during Eocene and Oligocene time. During the Miocene, clays and silts of the Kirkwood Formation were deposited. This formation was encountered in the borings at the site and can be observed in outcrops further north, although it is usually covered by a thin veneer of Quaternary deposits. At the Salem site, the borings encountered Quaternary deposits to an average depth of about 35 feet. These Quaternary soils consist 2.5-9 SGS-UFSAR Revision 6 February 15, 1987

of approximately 25 to 30 feet of hydraulic fill and an alluvium of loose organic silts and clays, and about 5 to 10 feet of coarser sands and gravels at the base. Generalized descriptions of the formations encountered at the Salem site, their physical properties, and their corresponding depths are shown on the Columnar Section - Showing Geophysical Data (Figure 2.5-9). The upper 200 feet of the column are based on the borings at the site, drilled under inspection. The descriptions below 200 feet are based on regional data and deep well information in the vicinity of the site. A 900-foot deep pilot hole, drilled at the site subsequent to the initial investigation, showed generally good correlation with the geologic column. No faulting or folding was observed at the site in a detailed review of all boring data. The Vincetown Formation was determined to be the closest stratum to the ground surface suitable for foundation support. In the Salem Station area the Vincetown is located some 70 feet below grade. The bottom of the base mats of the major Category I structures are located 22 to 46 feet below grade. A lean concrete fill was placed between the Vincetown and the base of the Category I structures. Conventional strength and consolidation tests were performed upon the foundation soils. These laboratory tests confirmed the results of field penetration tests and visual examination of undisturbed samples. The strength of the Vincetown was completely adequate to loads. To evaluate the performance of the Vince town under dynamic earthquake loadings, a study of its liquefaction potential was undertaken. A comparison was made between the subsurface conditions at Salem and the soil conditions at Niigata, Japan, where on June 16, 1964, an earthquake of greater magnitude than that postulated for the site Safe Shutdown Earthquake occurred, causing areas of liquefaction. The standard penetration 2.5-10 SGS-UFSAR Revision 6 February 15, 1987

resistances of the Vincetown soils were compared with those recorded in areas of Niigata where liquefaction both did and did not occur. The penetration resistances of the Vincetown soils were found to be even greater than those in the areas of Niigata where no liquefaction occurred. On the basis of these static and dynamic analyses, the Vincetown was considered to be a suitable foundation medium. All analysis considered the existence of a near surface water table and the artesian head in the Vincetown in accordance with the data presented in Section 2.4.13. 2.5.2 Vibratory Ground Motion 2.5.2.1 Geologic Conditions at Site As described in Sections 2.5.1.1 and 2.5.1.2, the site is underlain by some 1800 feet of Cretaceous~ Tertiary, and Quaternary-aged sediments. Crystalline basement rock outcrops near the Fall Zone, some 18 miles northwest of the site. A graphical representation of the site subsurface conditions is presented on Figures 2.5-5 and 2.5-9, Columnar Sections. Conditions encountered at the site are completely consistent with the known regional picture. 2.5.2.2 Tectonic Conditions The Coastal Plain sediments effectively mask the crystalline basement rock and no significant faulting has been identified in the area. However, based on regional data, the overlying Cretaceous and Tertiary sediments are undeformed. The absence of folding and faulting in the sedimentary strata indicates that, if unknown faults are present in the basement, any displacements along these faults during the last 135 million or so years have been negligible . 2.5-11 SGS-UFSAR Revision 6 February 15, 1987

No known faults exist within the basement rock or sedimentary deposits in the vicinity of the site. The closest known faulting to the site is about 25 miles away. Faults, at this distance, are found in the rocks of the Piedmont west of the Fall Zone; however, a minor fault has been identified east of the Fall Zone, near Gibbstown, New Jersey, about 25 miles northeast of the site. This fault is in the crystalline basement, covered by about 300 feet of Coastal Plain sediments, and apparently parallels the general northeast-southeast trend of the Piedmont. The Piedmont Province consists of igneous and metamorphic rock of Precambrian and early Paleozoic Age, with areas of sedimentary and igneous rocks of Triassic Age. The geologic history of this province is complex (see Section 2.5.1.1.2). Major tectonic activity has occurred in the Piedmont and many zones of major faulting have been identified. Well north of the site there is an inferred east-west trending fault system known as the Cornwall-Kelvin Wrench Fault Zone (see Figure 2.5-4). Regionally developed information (2) indicates that there is neither geological or geophysical evidence of this fault on the continent or at depth. The site lies to the north of the central portion of the Chesapeake-Delaware Embayment. This embayment is a zone of regional downwarping in the Coastal Plain, typical of other areas found extending from the Cape Fear Arch to as far north as the Grand Banks of Newfoundland. It is possible that faulting was associated with the formation of the embayment. 2.5.2.3 Behavior During Prior Earthquakes No major earthquake activity has affected the site area and no record of deleterious behavior of onsite soils (even the poorest surficial materials) is known. 2.5-12 SGS-UFSAR Revision 6 February 15, 1987

2.5.2.4 Geotechnical Properties In summary, the significant soil layers in the site vicinity are from the surface downward:

1. Hydraulic fill

2. River bottom sand

3. Clays of the Kirkwood Formation

4. Basal sand of the Kirkwood Formation

5. Vincetown Formation

6. Various sandy formations (Hornerstown and Navesink Formations)

Support of all Category I structures was provided by a lean concrete fill placed upon the Vincetown Formation. Physical properties developed for use in dynamic design are summarized on Figure 2.5-10. These material properties were developed as described in Section 2.5.1.2. 2.5.2.5 Seismicity The site is situated in a region which has experienced only minor earthquake activity. Only one shock within 50 miles of the site has been large enough to cause even minor structural damage. Since the region has been populated for over 300 years, it is probable that any earthquake of moderate intensity, say VI or

SGS-UFSAR 2.5-13 Revision 6 February 15, 1987

greater on the Modified Mercalli Scale*, would have been reported during this period. It is very likely that all earthquakes within the last 200 years, with intensities greater than V, in the region surrounding the site, have been reported. The first report of significant earthquake occurrence in the general area of the site dates back to 1871. Since then, only 22*'~ earthquakes with epicentral intensities of V or greater on the Modified Mercalli Scale have been reported within about 100 miles of the site (2). None of these shocks was greater than Intensity VII. Few were of high enough intensity to cause any structural damage and only two of these shocks can be considered more than minor disturbances. These were Intensity VII shocks near Wilmington, Delaware and Long Branch/Asbury Park, New Jersey, about 15 and 90 miles from the site, respectively. A list of earthquakes of Intensity V or greater with epicenters located within a distance of about 100 miles of the site is presented in Table 2.5-3, Significant Earthquakes Within 100 Miles of Salem, New Jersey. The locations of these and other earthquakes (through 1970) in the region surrounding the site are shown on Figure 2.5-11, Epicentral Location Map. Most of the reported earthquakes in the region have occurred in the Piedmont Physiographic Province, west of the Fall Zone. The closest approach of the Fall Zone to the site is about 18 miles.

All intensity values in this report refer to the Modified Mercalli Scale as abridged in 1956 by Richter. The intensity scale, a copy of which is presented in Table 2.5-2, is a means of indicating the relative size of an earthquake in terms of its perceptible effect.
- Excluding aftershocks of an event.

2.5-14 SGS-UFSAR Revision 6 February 15, 1987

There have been several large shocks with epicenters in the Coastal Plain, some of which were damaging. The largest are the Charleston, South Carolina, earthquakes of 1886, which are rated as having an epicentral intensity of IX. These two closely spaced (chronologically) earthquakes and other minor earthquakes in the Charleston area are localized in a very limited area. The largest and closest earthquake in the Coastal Plain to be of significance in the current study occurred near the northern New Jersey coast in 1927, about 90 miles northeast of the Salem site. The epicentral intensity of this earthquake was VII. Three shocks were felt over an area of about 3,000 square miles from Sandy Hook to Toms River. Highest intensities were felt from Asbury Park to Long Branch where chimneys fell, plaster cracked, and articles were thrown from shelves. This shock, which is the largest reported earthquake within 100 miles of the site, has not been related to any known geologic features. An Intensity VII earthquake occurred near Wilmington, Delaware, in 1871. It is not possible to precisely locate the epicenter of this shock with the limited data available, but it is probable that the shock occurred along the Fall Zone some 15 to 20 miles north of the Salem site. The epicentral intensity of this shock is rated at VII. At Wilmington, chimneys toppled and windows broke. Damage was also reported at Newport and New Castle, Delaware, and Oxford, Pennsylvania. The earthquake was felt over a relatively small area of northern Delaware, southeastern Pennsylvania and southwestern New Jersey. The shock was probably felt at Salem. Several smaller shocks also have been reported in the Coastal Plain in the region surrounding the site. None of these earthquakes caused any structural damage and they are of interest only in that they indicated the possible presence of unidentified faulting in the basement rock of the Coastal Plain. Nine earthquakes of Intensity V or greater have been reported within about 50 miles of the proposed station site. The largest 2.5-15 SGS-UFSAR Revision 6 February 1St 1987

of these was the aforementioned Intensity VII Wilmington earthquake of 1871. Other shocks occurred in 1879, Modified Mercalli Intensity (MMI) IV to V and in 1973 (MMI V) near the epicenter of the 1871 shock. These shocks were also probably felt at Salem. It is likely that these shocks are related to the Fall Zone or faulting in the vicinity of Wilmington, associated with Piedmont-type geologic structure. The epicenters of two shocks with intensities of IV to V in Harford County, Maryland in 1889, are within the Piedmont and can be related to well documented local structure. Four Intensity V earthquakes (1906 near Seaford, Delaware; 1921 near Moorestown, New Jersey; 1939 in Salem County, New Jersey; and east of Hammonton, New Jersey in 1968), originated within the Coastal Plain. These shocks have not been basement structure, generally similar to that exposed in the Piedmont. Available data regarding these shocks are very limited, and it is impossible to accurately estimate the maximum intensities of these shocks, or to precisely locate their epicenters. It is possible that some reports of older shocks may refer to relatively distant earthquakes which were felt in this area. Other shocks may possibly be attributed to causes other than tectonic activity. 2.5.2.6 Correlation of Epicenters with Geologic Structures In some instances, earthquakes occurring in the eastern United States have been associated with specific geologic structure, or at least some generalized seismogeni tic source area. However, earthquakes occurring within about 200 miles of the site have been small (no greater than MMI VII) and any positive identification with specific fault structure is somewhat tenuous. In general, because of the age of the "larger" shocks and the scatter of both the small, well located shocks and the regional fault systems, earthquakes have been assumed to have an equal possibility of occurrence any place within a tectonic (or seismotectonic) province. As a result, the 1871 and 1927 MMI VII shocks in the regions surrounding the site are of prime significance in selecting the "design" earthquakes for the site although neither have been positively associated with specific geologic structures. 2.5-16 SGS-UFSAR Revision 6 February 15, 1987

As subsequently discussed (Section 2.5.2.9) this lack of specific association requires the conservative use of a "floating" earthquake to define the Safe Shutdown Earthquake for the site. 2.5.2.7 Identification of Active Faults Small earthquakes in the region have been spatially associated with ancient faulting. However, in most instances~ the focal mechanism solution to the shock is not consistent with the stress conditions responsible for the last movements upon the fault in question. In addition, no evidence of surface rupture has been associated with local earthquake activity. Thus, "active 11 faulting, as the term is ordinarily used in connection with active plate margins (e.g. California), is non-existent in the region of the site. 2.5.2.8 See Section 2.5.2.7. 2.5.2.9 Maximum Earthquake The two largest earthquakes nearest the site were the 1871 Wilmington, Delaware and the 1927 Asbury Park, New Jersey shocks (see Table 2.5-3). Both had maximum epicentral intensities of VII. Intensity VII shocks are the largest that have occurred throughout the surrounding regions, and both from a deterministic and p_robabilistic standpoint, appear to be the largest credible earthquake. Therefore, for purposes of seismic design, an Intensity VII shock has been assumed to occur near the site .

SGS-UFSAR 2.5-17 Revision 6 February 15, 1987

The selection of the Operating Basis Earthquake was based upon the assumption of a shock similar to the following:

1. A shock equivalent to the Intensity VII, 1871 Wilmington earthquake occurring as close to the site as its related geologic structure. It is likely that this earthquake was related to the Fall Zone or to faulting in the Piedmont west of the Fall Zone. However, since it is impossible to precisely locate the epicenter of this shock from the limited available data, and since the earthquake was felt in portions of the Coastal Plain, it has been considered that the epicenter of this shock may have been located somewhat east of the Fall Zone and similar geologic structure could be postulated near the site.

2. A shock, equivalent to the Intensity VII northern New Jersey earthquake of 1927 occurring close to the site.

This shock occurred in the Coastal Plain and has not been related to any known geologic structure. Therefore, the conservative assumption has been made that it could occur along a hypothetical geologic structure in the basement rock near the site. Based on the foregoing statements the very conservative assumption has been made that the maximum potential earthquake would be a shock as large as Intensity VII originating in the basement rock close to the site. 2.5.2.10 Safe Shutdown Earthquake For a safe shutdown of the reactor, the facility has been designed using a seismic factor of 20 percent of gravity at foundation level. This level of horizontal ground acceleration is significantly greater than that which would be expected upon the foundation soils at the site if the safe shutdown earthquake (SSE) 2.5-18 SGS-UFSAR Revision 6 February 15, 1987

were to occur. The corresponding vertical ground acceleration is taken as 13.3 percent of gravity . 2.5.2.11 Operating Basis Earthquake On the basis of the seismic history of the area, it does not appear likely that the site will experience any significant earthquake ground motion during the economic life of the proposed facility. However, the proposed nuclear power station has been designed to respond elastically with no loss of function to horizontal earthquake ground accelerations of 10 percent of gravity, and vertical ground accelerations of 6.7 percent of gravity. These values are conservatively greater than the level of ground motion which would be expected at the site during an earthquake similar to any historical event. This ground acceleration is greater than what might be reasonable expected due to an earthquake similar to the 1871 Wilmington shock, Intensity VII, at an epicentral distance of about 15 to 20 miles . 2.5.2.12 Response Spectra Response spectra used in design are presented on Figures 2. 5-12 and 2.5-13, Response Spectra. These spectra conform to the average spectra developed by Dr. G. W. Housner for the frequency range higher than about 0. 33 cycle per second. These average spectra were originally presented in TID-7024. The spectra presented considered Dr. Hausner's latest revisions. The spectra for frequencies lower than about 0.33 cycle per second were prepared utilizing data suggested by Dr. N. M. Newmark. These data are presented in the Proceedings of the International Atomic Energy Agency Panel on Aseismic Design and Testing of Nuclear Facilities (1967). The spectra have been normalized to a horizontal ground acceleration of 20 percent of gravity for the SSE and 10 percent of gravity for the Operating Basis Earthquake . 2.5-19 SGS-UFSAR Revision 6 February 15, 1987

2.5.3 Surface Faulting See Section 2.5.2.7, Identification of Active Faults. 2.5.4 Stability of Subsurface Materials The foundation of the Class I station structures are established directly in the Paleocene silty sands of the Vincetown Formation or upon lean concrete fill extending to this Formation. The Vince town soils are preconsolidated and/ or cemented as a result of its depositional environment and subsequent erosion of younger sediments. Thus, this formation provides excellent foundation support for Salem Generating Station structures. Measurements made throughout plant construction and during initial operation indicated a maximum settlement of only about 0.5 inch. For a further description of the subsurface conditions at the site see Section 2.5.1.2, Site Geology. 2.5.5 Slope Stability At the completion of construction, the only slope of significance across the site is at the sea wall. As discussed in Section 2.4.5. 7, Protective Structures, the sea wall was investigated by conventional engineering procedures and designed to withstand the site maximum environmental loadings. 2.5.6 References for Section 2.5

1. Drake, C.L., and Woodward, H.P., "Appalachian Curvature, Wrench Faulting, and Offshore Structures, tl Trans. New York Academy of Sciences, 1963.

2. King, P.B., "The Tectonics of North America," U.S. Geological Survey Professional Paper 628, 1969.

2.5-20 SGS-UFSAR Revision 6 February 15, 1987

3. Stover, C.W., et al, "Seismicity Map of New Jersey," U.S.

Geological Survey, 1980 .

4. Dombroski, Daniel R., Jr., "Earthquakes in New Jersey," State of New Jersey, Dept. of Environmental Protection, Bureau of Geology and Topography, 1977.

2.5.7 Bibliography for Section 2.5

1. Anderson, J .L., "Cretaceous and Tertiary Subsurface Geology, (Maryland), 11 Dept. of Geology, Mines and Water Resources, State of Maryland, Bulletin No. 2, 1948.

2. Barkdale, H.C.; Greenman, D.W.; Lang, S.M.; Hilton, G. S.; and Outlaw, D.E., "Ground Water Resources in the Tri-State Region Adjacent to the Lower Delaware River," N.J. Dept. of Conservation and Economic Development, Division of Water Policy and Supply, Special Report 13, 1958.

3. Bonini, W.E., nBouguer Gravity Anomaly Map of New Jersey,"

N.J. Geological Survey, Dept. of Conservation and Economic Development, Geologic Report Series, No. 9, 1965.

4. Eardley, A.J., "Structural Geology of North America," 2nd Edition, Harper & Bros., New York, 1962.

5. Ewing, W.M. et al, "Geophysical Investigation in the Emerged and Submerged Atlantic Coastal Plain," Geologic Society of America Bulletin, Volume 61, 1950.

6. Gill, H.E., "Ground Water Resources of Cape May County, N.J.,

Salt Water Invasion of Principal Aquifers," New Jersey - Dept. of Conservation and Economic Development, Division of Water Policy and Supply, Special Report 18, 1962.

7. Hansen, H.J. III, 11 Pleistocene Stratigraphy of the Salisbury Area, Md. and its Relationship to the Lower Eastern Shore-A 2.5-21 SGS-UFSAR Revision 6 February 15, 1987

Subsurface Approach, n Maryland Geological Survey Report of Investigations, No. 2, 1966.

8. Jordan, R.R., "Stratigraphy of the Sedimentary Rocks of Delaware, Delaware Geological Survey, Bulletin No. 9, 1962.

9. Kasabach, H.F. and Scudder, R.J., t!Deep Wells of the New Jersey Coastal Plain," New Jersey Geological Survey, Geologic Report Series No. 3, 1961.

10. Lewis, J.V. and Kummel, H.B., Revised by Kummel, 1931 and Johnson, M.E., "1950 Geologic Map of New Jersey," New Jersey Dept. of Conservation and Economic Development, Atlas Sheet 40, 1910-1912.

11. Marine, I.W. and Rasmussen, W.D., "Preliminary Report of the Geology and Ground Water Resources of Delaware, 11 Delaware Geological Survey Bulletin No. 4, 1955.

12. Minard, J.P., "Geology of the Woodstown Quadrangle, Gloucester and Salem Counties, New Jersey, 11 USGS Geologic Quadrangle Map, GQ-404, 1965.

13. Murray, G. E. , nGeology of the Atlantic and Gulf Coastal Province of North America, 11 Harper and Bros., New York, 1961.

14. Olmsted, F.H.; Parker, G.G.; and Kneighton, W.B., Jr.,

11 Delaware River Basin Report," U.S. Army Engineer District, Philadelphia, Vol. 7, 1959.

15. Pennsylvania State Topographic and Geologic Survey, Geologic Map of Pennsylvania, 1960.

16. Rasmussen, W.C.; Croot, J.J.; Martin, R.O.R.; McCarren, E.F.;

Behn, V.C.; et al, "The Water Resources of Northern Delaware,n Delaware Geological Survey Bulletin No. 6, Vol. 1, 1957 . 2.5-22 SGS-UFSAR Revision 6

                                                        *February 15, 1987

17. Richards, H.G.; Olmsted, F.H.; and Ruhle, J.L., "Generalized Structure Contour Maps of the New Jersey Coastal Plain, 11 N.J .

Dept. of Conservation and Economic Development, Geologic Report Series No. 4, 1962.

18. Rima, D.R.; Coskery, O.J.; and Anderson, P.W., 11 Ground Water Resources of Southern New Castle County Delaware," Delaware Geological Survey Bulletin No. 11, 1964.

19. Spangler, W.B. and Paterson, J.J., 11 Geology of the Atlantic Coastal Plain in New Jersey, Delaware, Maryland and Virginia,"

American Association of Petroleum Geologists Bulletin, Volume 34, No. 1, 1950.

20. Spoljaric, N. and Jordan, R.R., 11 Generalized Geologic Map of Delaware, 11 Delaware Geologic Survey, 1966.

21.* Stover, C. W. et al, 11 Seismicity Map of New Jersey," U.S. Geological Survey, 1981 .

22. Sundstrom, R.W. et al, "The Availability of Ground Water from the Potomac Formation in the Chesapeake and Delaware Canal Area, Delaware," University of Delaware, Water Resources Center, 1967.

23. United States Geological Survey, Tectonic Map of the U.S. ,

United States Geological Survey and the American Association of Petroleum Geologists, 1962.

24. United States Geological Survey, National Atlas - Geology, 1966.

25. United States Geological Survey, Basement Map of North America; American Association of Petroleum Geologists and the U.S.G.S., 1967 .

2.5-23 SGS-UFSAR Revision 6

February 15,
1987.

26. United States Geological Survey, Engineering Geology of the Northeast Corridor, Washington, D.C. to Boston, Massachusetts, Earthquake Epicenters, Geothermal Gradients and Excavations and Borings; Miscellaneous Geologic Investigations, Map 1-514-C, 1967.

27. United States Geological Survey, Engineering Geology of the Northeast Corridor, Washington, D.C. to Boston, Massachusetts, Coastal Plain and Surficial Deposits; Miscellaneous Geologic Investigations, Map 1-514-B, 1967.

28. United States Geological Survey, Engineering Geology of the Northeast Corridor, Washington, D.C. to Boston, Massachusetts, Bedrock Geology; Miscellaneous Geologic Investigations, Map 1-514-A, 1967.

29. Widmar, K., "The Geology and Geography of New Jersey," The New Jersey Historical Series, Volume 19, 1964.

30. Woollard, G.P.; Bonini, W.E.; and Meyer, R.P., "A Seismic Refraction Study of the Subsurface Geology of the Atlantic Coastal Plain and Continental Shelf between Virginia and Florida," University of Wisconsin, Dept. of Geology, Division of Geophysics, 1957.

SGS-UFSAR 2.5-24 Revision 6 February 15, 1987

TABLE 2.5-1 LIST OF REFERENCES Agencies and Individuals Interviewed Agency Location Individual New Jersey Geological Survey Trenton, NJ Mr. F. J. Markewicz Delaware University, Newark, DE Dr. R. R. Jordan Delaware Geological Survey Maryland Geological Survey Baltimore, MD Dr. K. N. Weaver Maryland Geological Survey Baltimore, MD Dr. H. J. Hansen Johns Hopkins University Baltimore, MD Dr. E. Cloos U.S. Corps of Engineers Philadelphia, PA Mr. A. Depman U.S. Corps of Engineers Philadelphia, PA Mr. B. Uibel U.S. Geological Survey Trenton, NJ Mr. H. Meisler U.S. Geological Survey Trenton, NJ Mr. H. Gill Alpine Geophysics Norwood, NJ Dr. C. Frye Lamont Geological Observatory Palisades, NY Dr. C. Drake University of Massachusetts Amherst, MA Dr. R. Bromery

SGS-UFSAR 1 of 1 Revision 6 February 15, 1987

TABLE 2.5-2 MODIFIED MERCALLI INTENSITY (DAMAGE) SCALE OF 1931 (Abridged) I. Not felt except by a very few under especially favorable circumstances. (I Rossi-Forel Scale.) II. Felt only by a few persons at rest, especially on upper floors of buildings. Delicately suspended objects may swing. (I to II Rossi-Forel Scale.) III. Felt quite noticeably indoors, especially on upper floors of buildings, but many people do not recognize it as an earthquake. Standing motorcars may rock slightly. Vibration like passing of truck. Duration estimated. (III Rossi-Forel Scale.) IV. During the day felt indoors by many, outdoors by few. At night some awakened. Dishes, windows, doors disturbed; walls make creaking sound. Sensation like heavy truck striking building. Standing motorcars rocked noticeably. (IV to V Rossi-Forel Scale.)

v. Felt by nearly everyone, many awakened. Some dishes, windows, etc., broken; a few instances of cracked plaster; unstable objects overturned. Disturbances of trees, poles, and other tall objects sometimes noticed

Pendulum clocks may stop. (V to VI Rossi-Forel Scale.)

VI. Felt by all, many frightened and run outdoors. Some heavy furniture moved; a few instances of fallen plaster or damaged chimneys. Damage slight. (VI to VII Rossi-Forel Scale.) VII. Everybody runs outdoors. Damage negligible in buildings of good design and construction; slight to. moderate in well-built ordinary structures; considerabl~ in poorly built or badly designed structures; some chimneys broken. Noticed by persons driving motorcars. (VIII Rossi-Forel Scale.) VIII. Damage slight in specially designed structures; considerable in ordinary substantial buildings with partial collapse; great in poorly built structures. Panel walls thrown out of frame structures. Fall of chimneys, factory stacks, columns, monuments, walls. Heavy furniture overturned. Sand and mud ejected in small amounts. Changes in well water. Persons driving motorcars disturbed. (VIII+ to IX Rossi-Forel Scale.) IX. Damage considerable in specially designed structures; well-designed frame structures thrown out of plumb; great in substantial buildings, with partial co.llapse . 1 of 2 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.5-2 (Cont) Buildings shifted off foundations. Ground cracked conspicuously. Underground pipes broken. (IX+ Rossi-Forel Scale.) X. Some well-built wooden structures destroyed; most masonry and frame structures destroyed with foundations; ground badly cracked. Rails bent. Landslides considerable from river banks and steep slopes. Shifted sand and mud. Water splashed (slopped) over banks. (X Rossi-Forel Scale.) XI. Few, if any, (masonry) structures remain standing. Bridges destroyed. Broad fissures in ground. Underground pipelines completely out of service. Earth slumps and land slips in soft ground. Rails bent greatly. XII. Damage total. Waves seen on ground surfaces. Lines of sight and level distorted. Objects thrown upward into the air .

SGS-UFSAR 2 of 2 Revision 6 February 15, 1987

TABLE 2.5-3 SIGNIFICANT EARTHQUAKES WITHIN 100 MILES OF SALEM, NEW JERSEY

                                         * (Intensity V or Greater)

Distance N. Lat. W. Long. Area Felt From Site Year Date Time Intensity Location (degrees) (degrees) (sq mi) (mi) 1871 Oct. 9 09:40 VII Wilmington, DE 39 3/4 75 1/2 15 1877 Sept. 10 09:59 IV-V Delaware Valley 40.1 74.9 300 60 1879 Mar. 25 19:30 IV-V Delaware River 39 3/4 75 1/2 600 15 1883 Mar. 11 18:57 IV-V Harford County, MD 39.5 76.4 Local* 50 Mar. 12 00:00 IV-V Harford County, MD 39.5 76.4 Local 50 01:00 1884 May 31 v Allentown, PA 40.6 75.5 Local 80 1889 Mar. 8 18.40 VI Southeastern, PA 40 76 3/4 4,000 50 1895 Sept. I 06:09 VI Near High Bridge, NJ 40.7 74.8 35,000 90 1906 May 8 12:41 v Seaford, DE 38.7 75.7 400 50 1908 May 31 i2:42 VI Allentown, PA 40.6 75.5 Local 80 1921 Jan. 26 18:40 v Moorestown, NJ 40.0 75.0 150 45 1927 June 1 07:23 VII New Jersey Coast 40.3 74.0 3,000 90 07:31 07:39 1 of 2 SGS-UFSAR Revision 6 February 15, 1987

TABLE 2.5-3 (Cont) Distance N. Lat. W. Long. Area Felt From Site Year Date Time Intensit}!: Location (degrees) (degrees) (sg mi) (mi) 1933 Jan. 24 21:00 v Central NJ 40.1 74.5 600 60 1938 Aug. 22 22:37 v Central NJ 40.1 74.5 5,000 70 Aug. 23 00:05 02:03 1939 Nov. 14 21:54 v Salem County, NJ 39.6 75.2 6,000 20 1954 Jan. 7 02:25 VI Sinking Spring, PA 40.3 76.0 60 1957 Mar. 23 14:03 VI West-Central, NJ 40.6 74.8 90 1961 Sept. 14 21 ~ 17 v Lehigh Valley, PA 40*.6 75.4 Local 80 1961 Dec. 27 12:06 v PA-NJ Border 40.1 74.8 60 1964 May 12 04:45 VI Cornwall, PA 40.2 76.5 70 1968 Dec. 10 09:12 v Wharton State Forest 39.7 74.6 50 1973 Feb. 28 03:21 v Penns Grove, NJf 39.7 75.4 15,000 20 Wilmington, DE 2 of 2

SGS-UFSAR Revision 6 February 15, 1987

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SECTION 3 TABLE OF CONTENTS DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS section 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1-1 3.1.1 Introduction 3.1-l 3.1.2 Conformance With ABC Proposal General Design Criteria (July 1967) 3.1-2 3.1.3 conformance With ABC General Design Criteria (July 1971) 3.1-40 3.1.4 PSE&G General Criteria 3.1-40a 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.2-1 3.2.1 Seismic classification 3.2-1 3.2.1.1 Definition of Seismic Design Classifications 3.2-l 3.2.1.2 Seismic Classification of Structures, Systems, and Components 3.2-2 3.2.1.3 Seismic Criteria 3.2-7 3.2.2 System Quality Considerations 3.2-7 3.2.2.1 Codes and Standards 3.2-7 3.2.2.2 ANSI 831.7 3.2-7 3.2.2.3 Field-Run Piping 3.2-8 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.3.1 Wind Loadings 3.3-1 3.3.1.1 Design Wind Velocity and Loadings 3.3-1 3.3.2 Tornado Loadings 3.3-l 3.3.2.1 Tornado Parameters 3.3-1 3.3.2.2 Determination of Forces on Structures 3.3-1 3.3.2.3 Interaction of Category I and Non-category I Structures 3.3-2 3-i SGS-UFSAR Revision 12 July 22, 1992

TABLE OF CONTENTS ( COnt) Section 3.4 WATER LBVEL (FLOOD) DESIGN 3.4-l 3.4.1 Flood Elevations 3.4-1 3.4.2 Structural Loadings 3.4-l 3.4.3 Flood Protection 3.4-1 3.4.3.1 Hurricane 3.4-1 3.4.3.2 Precipitation 3.4-5 3.4.4 Protection From Hurricane Drawdown 3.4-6 3.4.5 Reference for Section 3.4 3.4-6 3.5 MISSILE PROTECTION 3.5-1 3.5.1 Internally Generated Missiles 3.5-1 3.5.1.1 Missile Types 3.5-1 3.5.1.2 Missile Protection Methods 3.5-1 3.5.2 Tornado Missiles 3.5-3 3.5.2.1 Critical Missiles Selected for Evaluation 3.5-3 3.5.2.2 Missile Protection Methods 3.5-3 3.5.2.3 Safety Assurance Against Tornado Missile Induced Damages 3.5-5 3.5.2.3.1 Backup Water Sources for Category I Water Storage Tanks 3.5-5 3.5.2.3.2 Safety Evaluation of Loss of Suction to the Auxiliary Feedwater Pumps 3.5-10 3.5.3 Modified Petry Formula 3.5-12 3.5.4 Turbine Missile 3.5-12 3.5.4.1 Turbine Placement and Characteristics 3.5-12 3.5.4.1.1 High Pressure (HP) Turbine 3.5-13 3.5.4.1.2 Low Pressure (LP} Turbine 3.5-15 3.5.4.1.3 Overspeed Protection System Testing 3.5-18 3.5.4.2 Probability of Missile Generation 3.5-18 3.5.4.2.1 High Pressure Turbine Risk Analysis 3.5-19 3.5.4.2.2 Low Pressure Turbine Risk Analysis 3.5-19 3.5.5 References for Section 3.5 3.5-22 3-ii SGS-UFSAR Revision 16 January 31, 1998

TABLE OF CONTENTS (Cont) Section Title 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6-1 3.6.1 Systems in Which Design Basis Piping Breaks Occur 3.6-1 3.6.2 Design Basis Break Criteria 3.6-1 3.6.3 Design Loading Combination 3.6-3 3.6.4 Dynamic Analysis 3.6-3 3.6.4.1 Reactor Coolant Loop System Analysis 3.6-3 3.6.4.2 Reactor Coolant Loop Break Analysis 3.6-5 3.6.4.2.1 Postulated Break Locations 3.6-5 3.6.4.2.2 Mathematical Modeling (Original Design Analysis) 3.6-6 3.6.4.2.3 Solution Method (Original Design Analysis) 3.6-7 3.6.4.2.4 Results of Analysis (Original Design Analysis) 3.6-8 3.6.4.3 Other High Energy Piping Systems Outside of Containment 3.6-8 3.6.4.3.1 Introduction 3.6-8 3.6.4.3.2 Criteria Used for Analysis 3. 6-8 3.6.4.3.3 Analytical Models and Techniques 3.6-10 3.6.5 Protective Measures 3. 6-13 3.6.5.1 Nuclear Components 3.6-13 3.6.5.1.1 Barriers 3.6-13 3.6.5.1.2 Arrangement 3.6-14 3.6.5.1.3 Restraint 3.6-15 3.6.5.1.4 Protective Measures for Unit 1 3.6-16 3.6.5.2 Main Steam System 3.6-16 3.6.5.2.1 Criteria for Determination of Location and Orientation of Postulated Breaks 3.6-19 3.6.5.2.2 Analysis of Postulated Pipe Breaks 3.6-21 3.6.5.2.3 Effects of Postulated Breaks on Safety-Related Equipment 3.6-25 3-iii SGS-UFSAR Revision 24 May 11, 2009

TABLE OF CONTENTS (Cont) 3.6.5.2.4 Analysis of Postulated Break on Unit Shutdown Capability 3.6-28 3.6.5.3 Stearn Generator Feedwater System 3.6-30 3.6.5.3.1 Criteria for Determination of Location and Orientation of Postulated Break 3.6-31 3.6.5.3.2 Analysis of Postulated Pipe Breaks 3.6-32 3.6.5.3.3 Effects of Postulated Breaks on Related Equipment 3.6-32 3.6.5.3.4 Analysis of Postulated Break on Unit Shutdown Capability 3.6-33

3. 6. 5. 4 Chemical and Volume Control Letdown Line 3.6-34 3.6.5.5 Stearn Generator Blowdown System 3.6-35 3.6.5.6 Steam Supply to the Auxiliary Feedwater Pump Turbine 3.6-36 3.6.5.7 Chemical and Volume Control Charging and Reactor Coolant Pump Seal Injection 3.6-37 3.6.5.8 Stearn 3.6-38 3.6.5.9 Water 3.6-39 3.6.5.10 Protection Against Stearn Flooding 3.6-40 3.6.5.11 Design Criteria For Encapsulation Sleeves 3.6-44 3.6.5.12 Moderate Energy Pipe Failure Evaluations- 3.6-45 3.6.5.12.1 Definitions 3. 6-4 6 3.6.5.12.2 Postulated Break Location 3.6-46 3.6.5.12.3 Postulated Crack Size 3.6-47 3.6.5.12.4 Evaluation Procedure 3.6-47 3.6.5.12.5 Results and 3.6-48 Modifications 3.6.5.12.6 Additional Modifications 3.6-53 3.6.5.13 Adjacent Non-Class I, Non-Safety Structure and Equipment Failure Evaluation 3.6-54 3.6.5.14 Electrical Cable Environmental Qualification 3.6-56 3.6.6 References for Section 3.6 3.6-56 App. 3.6A Description of Backdraft Damper 3.6A-1 3-iv SGS-UFSAR Revision 25 October 26, 2010

TABLE OF CONTENTS (Cont) Section 3.7 SEISMIC DESIGN 3. 7-1 3.7.1 Seismic Input 3.7-2 3.7.1.1 Design Response Spectra 3.7-2 3.7.1.2 Design Response Spectra Derivation 3. 7-2 3.7.1.3 Critical Damping Values 3.7-3 3.7.1.4 Bases for Site Dependent Analysis 3.7-4 3.7.2 Seismic System Analysis 3.7-5 3.7.2.1 Seismic Analysis for Structures 3.7-6 3.7.2.1.1 Seismic Analysis for Category I Structures 3.7-6 3.7.2.1.2 Seismic Analysis for Category II and Category III Structures 3.7-7 3.7.2.2 Natural Frequencies and Response Loads 3.7-8 3.7.2.3 Procedures Used to Lump Masses 3.7-8 3.7.2.4 Methods Used to Couple Soil With Seismic-System Structures 3.7-9 3.7.2.5 Development of Floor Response Spectra 3.7-9 3.7.2.6 Differential Seismic Movement of Interconnected Components 3.7-10 3.7.2.7 Combination of Modal Responses 3.7-10 3.7.2.8 Effects of Variations on Floor Response Spectra 3.7-11 3.7.2.9 Method used to Account for Torsional Effects 3.7-12 3.7.2.10 Comparison of Responses 3.7-13 3.7.2.11 Methods to Determine Category I Structure Overturning 3.7-14 3.7.2.12 Analysis Procedure for Damping 3.7-14 3.7.3 Seismic Subsystem Analysis 3.7-15 3.7.3.1 Determination of Number of Earthquake Cycles 3.7-15 3.7.3.2 Basis for Selection of Forcing Frequencies 3. 7-17 3-v SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS (Cont) Section Ti 3.7.3.3 Procedure for Combining Modal Responses 3.7-17 3.7.3.4 Bases for Computing Combined Response 3.7-18 3.7.3.5 Use of Simplified Dynamic Analysis 3.7-19 3.7.3.6 Modal Period Variation 3.7-21

3. 7.3. 7 Torsional Effects of Eccentric Masses 3.7-21 3.7.3.8 Piping Outside Containment Structure 3.7-22 3.7.3.9 Field Location of Supports and Restraints 3.7-23 3.7.3.9.1 General Procedure 3.7-23 3.7.3.9.2 Safety-Related Piping 3.7-23 3.7.4 Seismic Instrumentation Program 3.7-28 3.7.4.1 Comparison With Regulatory Guide 1.12 3.7-28 3.7.4.2 Location and Description of Instrumentation 3.7-28 3.7.4.3 Control Room Operator Notification 3.7-29 3.7.4.4 Comparison of Measured and Predicted Responses 3.7-29 3.7.5 References for Section 3.7 3.7-30 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8-1

3. 8.1 Containment Structure 3.8-1 3.8.1.1 General Description 3.8-1 3.8.1.2 Design Codes 3.8-3 3.8.1.3 Design Loads and Loading Combinat i_ons 3.8-4 3.8.1.4 Design and Analysis Procedures 3.8-15 3.8.1.4.1 Static Analysis 3.8-15 3.8.1.4.2 Dynamic Analysis 3.8-24 3.8.1.5 Structural Design and Acceptance Criteria 3.8-2.6 3.8.1.5.1 Fracture Prevention of Containment Pressure Boundary 3.8-27 3.8.1.6 Materials, Quality Control, and Special Construction Techniques 3.8-27 3.8.1.6.1 Liner Plate 3.8-27 3-vi SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS ( Cont) section 3.8.1.6.2 Base Hat 3.8-30 3.8.1.6.3 Cylinder wall 3.8-31 3.8.1.6.4 Dome 3.8-31 3.8.1.6.5 Penetrations and Openings 3.8-31 3.8.1.6.6 Polar crane 3.8-33 3.8.1.6.7 Missile Protection 3.8-34 3.8.1.6.8 construction Procedures and Practices 3.8-34 3.8.1.6.8.1 COdes of Practice 3.8-34 3.8.1.6.8.2 Concrete 3.8-35 3.8.1.6.8.3 Reinforcing Steel 3.8-47 3.8.1.6.8.4 Waterproofing Membrane 3.8-53 3.8.1.6.8.5 compaction of Fill 3.8-53 3.8.1.6.8.6 Liner Plate 3.8-54 3.8.1.6.8. 7 Construction 3.8-60 3.8.1.6.8.8 Penetrations 3.8-63 3.8.1.6.8.9 Equipment and Personnel Access Hatches 3.8-64

3. 8.1. 6. 8.10 Piping Penetrations 3.8-66

3. 8 .1. 6. 8.11 Electric Penetrations 3.8-68 3.8.1.7 Testing and Inservice Inspection 3.8-68 3.8.2 Steel containment System 3.8-69 3.8.3 Internal Structures 3.8-69 3.8.3.1 General Description 3.8-69 3.8.3.2 Desiqn Codes 3.8-69 3.8.3.3 Loads and Loading Combinations 3.8-70 3.8.3.4 Design and Analysis Provisions 3.8-71 3.8.4 other category I structures 3.8-74 3.8.4.1 summary Description 3.8-74 3.8.4.2 Desiqn Codes 3.8-74 3.8.4.3 Loads and Loading Combinations 3.8-74 3.8.4.4 Design and Analysis Procedures 3.8-77 3.8.4.4.1 HVAC Duct and Support Methodology 3.8-79 3-vii SGS-UFSAR Revision 16 January 31, 1998

TABLE OF CONTENTS (Cont) Section Title

3. 8. 4. s Quality Control, and Construction Techniques 3.8-79 3.8.4.5.1 Masonry Walls 3.8-79 3.8.5 l?ounda t ions 3.8-80 3.8.5.1 3.8-80 3.8.5.2 Codes, and 3.8-81 3.8.5.3 Loads and Load Combinations 3.8-81 3.8.5.4 Design and Analysis Procedures 3.8-81 3.8.6 References for Section 3.8 3.8-82 3.9 MECHANICAL SYS'fEMS AND COMPONENTS 3.9-1 3.9.1 Dynamic Systems J.\nalysis and Testing 3.9-3 3.9.1.1 Vibration Operational Test Program 3.9-3 3.9.1.2 Dynamic Testing Procedures 3.9-5 3.9.1.3 Dynamic Testing and Examination of Reactor Internals 3. 9-6 3.9.1.4 Correlation of Test and Analytical Results 3. 9-11 3.9.1.5 Dynamic Analysis Methods for Reactor Internals 3. 9-12 3.9.1.6 Methods for ASME Code Class I Components 3.9-15 3.9.1.7 Component Supports 3.9-16 3.9.1.8 Dynamic Analysis of the Reactor Coolant 3.9-16 3.9.2 ASME Code Class 2 and 3 Components 3.9-16 3.9.2.1 Analytical and Empirical Methods for Design of Pumps and Valves 3.9-18 3.9.2.2 and Installation Pressure-Relieving Devices 3.9-19 3.9.2.3 Field Run Piping Systems 3.9-20 SGS-UFSAR 3-viii Revision 23 October 17, 2007

TABLE OF CONTENTS (Cont) Section 3.9.3 Seismic Analysis of As-Built Safety-Related Piping 3.9-20 3.9.4 Inservice Testing of Pumps and Valves 3.9-22 3.9.5 References for Section 3.9 3.9-22 App. 3.9A Bolted connections For Linear component Support 3.9A-1 3.9A.1 Introduction 3.9A-1 3.9A.2 Design Approaches 3.9A-l 3.9A.3 Representative Analyses of Typical Supports 3.9A-3 3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.10-1 3.10.1 Seismic Qualification Criteria 3.10-1 3.10.1.1 Qualification Standards 3.10-1 3.10.1.2 Acceptance Criteria 3.10-2 3.10.2 Methods and Procedures For Qualifying Electrical Equipment 3.10-2 3.10.2.1 Seismic Qualification by Type Test 3.10-2 3.10.2.2 Seismic Qualification by Analysis 3.10-3 3.10.2.3 Seismic Qualification by a Combination of Type Test and Analysis 3.10-3 3.10.3 Methods and Procedures for Qualifying Supports of Instrumentation and Electrical Equipment 3.10-4 3.10.4 Results of Tests and Analyses 3.10-4 3.10.5 References for Section 3.10 3.10-4 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT 3.11-1 3.11.1 Equipment Identification and Environmental Conditions 3.11-1 3-ix SGS-UFSAR Revision 6 February 15, 1987

TABLE OF CONTENTS (Cont) Section 3.11.1.1 Equipment Identification 3.11-1 3 .11.1.2 Accident Conditions 3.11-2 3.11.1.3 Normal Operating Environment 3.11-2 3 .11. 2 Qualification Tests and Analyses 3.11-3 3.11.3 Qualification Test Results 3.11-4 3 .11.4 Loss of Ventilation 3.11-4 3.12 CONFORMANCE TO RULES ISSUED AFTER PLANT 3.12-1 LICENSING 3.12.1 NRC Rule on station Blackout 3.12-1 3.12 .1.1 Conformance to NRC Rule on Station Blackout 3.12-1 3.12.2 References 3.12-2 App. JA PSE&G Positions on USNRC Regulatory Guides JA-1 3-x SGS-UFSAR Revision 15 June 12, 1996

LIST OF TABLES Table Title 3.5-1 Mechanical Properties For LP Discs 3.5-2 Mechanical Properties For LP Discs 3.5-3 Mechanical Properties for LP Discs 3.6-1 Postulated Reactor Coolant System Pipe Ruptures 3.6-2 Main Steam Pipe Stress Summary - No. 13 Main Steam Line 3.6-3 Steam Generator Feedwater Pipe Stress Summary - No. 12 Feedwater Line 3.10-1 Summary of Seismic Qualifications for Safety-Related Equipment 3A-l Deleted 3-xi SGS-UFSAR Revision 28 May 22, 2015

LIST OF FIGURES Figure Title 3.3-1 Pressure Distributions 3.3-2 Wind Velocity Distributions for Tornado Conditions 3.3-3 Forces in Containment Vessel due to Tornado Wind Loads 3.3-4 Forces in Containment Vessel due to Tornado Wind Loads 3.3-5 Forces in Containment Vessel due to Tornado Wind Loads 3.4-1 Datum and Water Level Relationships 3.4-2 Dike Elevation "Section A-A" 3.4-3 Dike Elevation "Section B-B" 3.5-1 Deleted: Refer to Plant Drawing 204811 3.5-2 Deleted: Refer to Plant Drawing 232445 3.5-3 Deleted: Refer to Plant Drawing 232444 3.5-4 High Pressure Cylinder-1800 RPM Double-Flow Design 3.5-5 Blade Rings 3.5-6 Low-Pressure Element-1800 RPM Double-Flow Design With 44" Last Row Blades 3.5-6A Unit 2 Low-Pressure Element 1800 RPM Double-Flow Design With 47" Last Row Blades 3-xii SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure Title 3.5-7 Exploded View of Low Pressure Unit 3.6-1 Postulated Break Locations - Reactor Coolant System 3.6-2 Reactor Coolant Loop Model 3.6-3 Steam Generator Upper Support Hodel 3.6-4 SATAN-STHRUST Reactor Coolant Loop Hodel 3.6-5 Reactor Coolant Loop Model Showing Hydraulic Force Locations 3.6-6 Typical Time-History Reactor Coolant Loop Support Forces Following LOCA 3.6-7 Typical Pipewhip Restraint-Reactor Coolant System 3.6-8 Pipewhip Restraint Locations-Typical Steam Generator Feedwater Pipe 3.6-9 Pipewhip Restraint Locations-Typical Main Steam Pipe 3.6-10 Piping Arrangement-Main Steam and Feedwater 3.6-11 Piping Arrangement-North Penetration Area-Plan Views at Elevations 84'-0 11 and 100 1 -0 11 3.6-12 Composite Study-North Penetration Area 3.6-13 Composite Study-South Penetration Area 3.6-14 Typical Pipewhip Restraint-Detail 3-xiii SGS-UFSAR Revision 6 February 15, 1987

LIST OF FIGURES (Cant) Figure Title 3.6-15 Electrical Arrangement-North Penetration Area-Plan ..Niew at Elevation 78'-0" 3.6-16 Electrical Arrangement-North Penetration Area-Plan View at Elevation 100'-0" 3.6-17 Electrical Arrangement-North Penetration Area-Section "B-B" and "C-C" 3.6-18 Electrical Arrangement-North Penetration Area-Section "A-A" 3.6-19 Typical Feedwater Pipewhip Attachment 3.6-20 Isometric View of CVCS Letdown Line Outside of Containment 3.6-21 Letdown Line Encapsulation Details 3.6-22 No. 1 Unit - Containment and Penetration Area - Composite - Elevation 78' 3.6-23 No. 1 Unit - Containment and Penetration Area - Composite - Elevation 100' 3.6-24 Isometric View of Steam Line to the Auxiliary Feedwater Pump Turbine Showing Postulated Break Locations 3.6-25 Typical Pipewhip Restraint and Sleeving for Steam Line to Auxiliary Feedwater Pump Turbine 3-xiv SGS-UFSAR Revision 6 February 15, 1987

LIST OF FIGURES (Cont) Figure Title 3.6-26 Deleted: Refer to Plant Drawing 204805 3.6-27 Deleted: Refer to Plant Drawing 204804 3.6-28 No. 1 Unit Auxiliary Building-Ventilation Design for Pipe Break 3.6-29 Steam Line to Auxiliary Feed Pump and Letdown Line Details 3.6-30 Backdraft Damper Typical Design 3.7-1 Comparison of Seismic Response Spectra O.B.E. (10% g)-2% Damped Structural Response 3.7-2 Comparison of Seismic Response Spectra D. B. E. (20% g)-5% Damped Structural Response 3.7-3 Envelope of Total Accelerations for Shell and Peak Total Accelerations for Mat-D.B.E. 3.7-4 Envelope of Total Accelerations for Shell and Peak Total Accelerations for Mat-O.B.E. 3.7-5 Peak Displacements in Containment Vessel Relative to Foundation Mat At T=2.8 sec. - D.B.E. (0.2 g) 3.7-6 Peak Displacements in Containment Vessel Relative to Foundation Mat At T=2.54 Sec-O.B.E. (0.1 g) 3-xv SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure Title 3.7-7 Envelope of Forces in Containment Vessel Due to Design Basis Earthquake ( 0. 2 g) 3.7-8 Envelope of Forces in Containment Vessel Due to Design Basis Earthquake ( 0. 2 g) 3.7-9 Envelope of Forces in Containment Vessel Due to Design Basis Earthquake ( 0. 2 g) 3.7-10 Envelope of Forces in Containment Vessel Due to Operating Basis Earthquake 3.7-11 Envelope of Forces in Containment Vessel Due to Operating Basis Earthquake ( 0. 1 g) 3.7-12 Envelope of Forces in Containment Vessel Due to Operating Basis Earthquake ( 0. 1 g) 3.7-13 Axisymmetric Finite Element Model - Containment Vessel 3.8-1 Containment Building Cross Section 3.8-2 Deleted: Refer to Plant Drawing 208900 3.8-3 Reactor Containment Cylinder Wall Reinforcement 3.8-4 Deleted: Refer to Plant Drawing 201102 3.8-5 Deleted: Refer to Plant Drawing 201105 3-xvi SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure Title 3.8-6 Deleted: Refer to Plant Drawing 201108 3.8-7 Deleted: Refer to Plant Drawing 201175 3.8-8 Reactor Containment - Cylindrical Wall Liner -Quadrant "A" 3.8-9 Deleted: Refer to Plant Drawing 201181 3.8-10 Deleted: Refer to Plant Drawing 201131

3. 8-11 Design Pressure - Temperature Transient 3.8-12 1.25 Times Design Pressure - Temperature Transient 3.8-13 1.50 Times Design Pressure - Temperature Transient 3.8-14 Designation of Main Reinforcement Pattern for Containment Vessel 3.8-15 Designation of Main Reinforcement Pattern for Containment Vessel 3.8-16 Critical Section for Review at Foundation Mat 3.8-17 Radial Stresses Induced in Mat by Factored Load Combinations 3.8-18 Tangential Stresses Induced in Mat by Factored Load Combinations 3-xvii SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure -**~

3.8-19 Vertical Stresses Induced in Mat by Factored Load Combinations 3.8-20 Shear Stresses Induced in Mat by Factored Load Combinations 3.8-21 Forces in Containment Vessel Due to Dead Load 3.8-22 Forces in Containment Vessel Due to Dead Load 3.8-23 Forces in Containment Vessel Due to Dead Load 3.8-24 Thermal Gradients in Containment Wall Under Operating and Accident Conditions 3.8-25 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Operating Condition 3.8-26 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Operating Condition 3.8-27 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Operating Condition 3.8-28 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Accident Condition 3.8-29 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Accident Condition 3.8-30 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading Under Accident Condition 3-xviii SGS-UFSAR Revision 6 February 15' 1987

LIST OF FIGURES (Cont) Figure Title 3.8-31 Wind Velocity Distributions for Tornado Conditions 3.8-32 Pressure Distributions 3.8-33 Hot Pipe Penetration With Cooling 3.8-34 Redundant Air Supply to Penetration Cooling Coils 3.8-35 Deleted: Refer to VTD 301051 3.8-36 Deleted: Refer to VTD 301075 3.8-37 Pressure Piping - Personnel Lock 3.8-38 Deleted: Refer to VTD 301059 3.8-39 Forces and Moments in Edge Beam of 18' Diameter Opening at Elevation 140' Due to Load Combinations A, B, and C 3.8-40 Forces and Moments in Edge Beam of 18' Diameter Opening at Elevation 140' Due to Load Combinations A, B, and C 3.8-41 Forces and Moments in Edge Beam of 9' Diameter Opening at Elevation 104' Due to Load Combinations A, B, and C 3.8-42 Forces and Moments in Edge Beam of 9' Diameter Opening at Elevation 104' Due to Load Combinations A, B, and C 3-xix SGS-UFSAR Revision 27 November 25, 2013

LIST OF FIGURES (Cont) Figure ._. Forces and Moments in Edge Beam of 9' Diameter Opening at Elevation 104' Due to Load Combinations A, B, and C 3.8-44 Forces and Moments in Edge Beam of 9' Diameter Opening at Elevation 134' Due to Load Combinations A, .B, and C 3.8-45 Force and Moment Contours Around 18' Diameter Opening Due to Critical Load Combination "A* 3.8-46 Force and Moment Contours Around 18' Diameter Opening Due to Critical Load Combination "A" 3.8-47 Force and Moment Contours Around 9' Diameter Opening Due to Critical Load Combination "A" 3.8-48 Force and Moment Contours Around 9' Diameter Opening Due to Critical Load Combination 11 A" Piping Penetration - Hot Pipe 3.8-50 Piping Penetration - Cold Pipe 3.8-50A Service Water Piping Penetration - Cold Pipe 3.8-51 TYPical Low and Medium Voltage Electrical Penetration Assembly 3.8-52 Reactor Cavity - Plan View 3.8-53 Reactor Cavity - Section "A-A" 3.8-54 Reactor Cavity - Section "B-B" 3-xx SGS-UFSAR Revision 13 June 12, 1994

LIST OF FIGURES (Cont) Figure Title 3.8-55 Stresses in a Thick Cylinder 3.8-56 Deleted: Refer to Plant Drawing 201012 3.9.1 Seismic Dynamic Model 3.9-2 Vibration Check-Out Functional Test Inspection Points 3A-1 Effect of Fluence and Copper Content on Shift of RTNDT for Reactor Vessels Exposed to 550°F 3A-2 Building Location Key 3A-3 Deleted 3A-4 Deleted 3-xxi SGS-UFSAR Revision 27 November 25, 2013

SECTION 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH GENERAL DESIGN CRITERIA 3.1.1 Introduction The general design criteria that were followed in the design of this plant are the Atomic Industrial Forum (AIF) version, as published in a letter to the Atomic Energy Commission from E. A. Wiggin, Atomic Industrial Forum, dated October 2, 1967. The criteria were developed as performance criteria which define or describe safety objectives and procedures, and they provide a guide to the type of plant design information which is included in this report. In addition to the AIF General Design Criteria, the Salem Generating Station (SGS) was designed to comply with Public Service Electric & Gas (PSE&G' s) understanding of the intent of the AEC' s proposed General Design Criteria, as published for comment by the AEC in July, 1967. The application of the AEC' s proposed General Design Criteria to the Salem station is discussed in Section 3.1.2. A comparison of the Salem plant design with 10CFRSO, Appendix A (General Design Criteria for Nuclear Power Plants dated July 7, 1971) is provided in Section 3.1.3. PSE&G's general criteria for the Salem plant are discussed in Section 3 .1. 4. 3.1-1 SGS-UFSAR Revision 12 July 22, 1992

3 .1. 2 Conformance with AEC Proposed General Design Criteria (July 1967) Criterion 1 - Quality Standards Those system and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety functions, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures, and inspection acceptance levels used is required. Discussion The systems and components of the facility have been classified according to their importance in the prevention and mitigation of accidents which could cause undue risk to the health and safety of the public. These classifications are described in Section 3.2. A discussion of the codes and standards, test provisions, etc., applying to each system is included in that portion of the report describing that system. The Salem Generating Station Quality Assurance Program is outlined in Section 17. Criterion 2 - Performance Standards Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to mitigation of their consequences shall be designed, fabricated and erected to performance standards 3.1-2 SGS-UFSAR Revision 12 July 22, 1992

that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design. Discussion The systems and components designated Class I are designed to withstand without loss of capability to protect the public, the most severe environmental hazards discussed and analyzed in Sections 2 and 3. The influence of these hazards on various aspects of the plant design is discussed in the sections covering the specific systems and components concerned. Criterion 3 - Fire Protection The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas containing critical portions of the facility such as Containment, control room, and components of engineered safety features. Discussion Primary emphasis is directed at minimizing the risk of fire by use of thermal insulation and adhesives which do not support combustion, flame resistant wiring, adequate overload and short circuit protection, and elimination of combustible trim and furnishings. The facility is equipped with fire protection 3.1-3 SGS-UFSAR Revision 12 July 22,1992

systems for controlling any fires which might originate in plant equipment. See Section 9.5.1 for a description of the Fire Protection System. The Containment and Auxiliary Building Ventilation Systems can be operated from the control room of the corresponding unit as required to limit the potential consequences of fire. Critical areas of the containment and control room and the areas containing components of engineered safety features have detectors to alert the control room of the possibility of fire so that prompt action may be taken to prevent significant damage. Criterion 4 - Sharing of Systems Reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing. Discussion The only systems shared by the two units are Compressed Air, control room area air intake radiation monitoring and parts of the control area ventilation system (see Section 9.4.1 for description of the Control Area Ventilation System), Demineralized Water, Bulk Nitrogen Supply, Hot Shutdown Panel Power Supply (see section 9.5.1.4 for description of the Hot Shutdown Panel), Solid Radwaste Packaging System, portions of the Chemical Volume Control System (see Section 9.3.4 for description of the Chemical Volume Control System Cross-tie) and the Chilled Water System (see Section 9.2.5). There are a minimum of shared components; chemical drain, laundry hot shower tanks and pumps, and the 20,000 barrel Bulk Fuel Oil Storage Tank are the only components in common. The Control Room area radiation monitoring microprocessors 1/2R1B provide redundant functions common to both control area air intake ducts to support operation of the Unit 1 and Unit 2 control area ventilation isolation system. The Hot Shutdown Panel will share electrical power supply during a control room evacuation. Plant safety is not impaired by these instances of system or component sharing. Criterion 5 - Records Requirements Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor. Discussion Public Service or its authorized representatives and Westinghouse 3.1-4 SGS-UFSAR Revision 32 June 17, 2021

Electric Corporation plan to maintain, either in their possession or under their control, a complete set of records of the design, fabrication, construction, and testing of essential plant components throughout the life of plant. This criterion is answered in more detail in Section 17. Criterion 6 - Reactor Core Design The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power. Discussion The ability of the core to function throughout its lifetime without exceeding acceptable fuel damage limits is discussed in Section 4. Detailed information on core design and performance is also included in Section 4. The instrumentation and controls associated with the reactor are described in Section 7 while decay heat removal systems are discussed in Sections 5, 6, and 9. Section 15 demonstrates that adequate fuel integrity is maintained under postulated abnormal situations. Criterion 7

Suppression of Power Oscillations The core design, together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.

3.1-5 SGS-UFSAR Revision 12 July 22, 1992

Discussion The inherent ability of the core to prevent and suppress power oscillations and the instrumentation and controls provided to assist in this function is discussed in Sections 4 and 7, respectively. Criterion 8 - Overall Power Coefficient The reactor shall be designed so that the overall power coefficient in the power operating range shall not be positive. Discussion As discussed in Section 4, the overall power coefficient is negative under normal operating conditions throughout core life. Criterion 9 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime. Discussion As discussed in detail in Section 5, the reactor coolant pressure boundary materials, design, analysis, fabrication, and testing preclude the possibility of gross rupture or significant leakage throughout its design lifetime. The Reactor Coolant System in conjunction with its control and protective provisions is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation or anticipated system interactions, and maintain the stresses within applicable code stress limits. 3.1-6 SGS-UFSAR Revision 12 July 22, 1992

Fabrication of the components which constitute the pressure retaining boundary of the Reactor Coolant System is carried out in strict accordance with the applicable codes. In addition, there are areas where equipment specifications for Reactor Coolant System components go beyond the applicable codes. Details are given in Section 4.5. The materials of construction of the pressure retaining boundary of the Reactor Coolant System are protected by control of coolant chemistry from corrosion phenomena which might otherwise reduce the system structural integrity during its service lifetime, as discussed in Section 9. System conditions resulting from anticipated transients or malfunctions are monitored and appropriate action is automatically initiated to maintain the required cooling capability and to limit system conditions so that continued safe operation is possible, as discussed in Section 7. The system is protected from overpressure by means of pressure relieving devices, as required by Section III of the ASME Boiler and Pressure Vessel Code. Sections of the system which can be isolated are provided with overpressure relieving devices discharging to closed systems such that the system code allowable relief pressure within the protected section is not exceeded. Criterion 10

Containment Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain the functional capability to protect the public for as long as the situation requires.

3.1-7 SGS-UFSAR Revision 12 July 22, 1992

Discussion The design of the containment structure and associated auxiliary systems is described in Section 3. 8. Engineered safety features required to limit pressure inside the containment are described in Section 6. Section 15 demonstrates the adequacy of such systems under various accident conditions including a rupture of the largest reactor coolant pipe. Criterion 11 - Control Room The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10CFR20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause. Discussion Each unit is provided with a control room which contains the controls and instrumentation necessary for operation of each unit's reactor, turbine generator, and auxiliary and emergency systems under normal or accident conditions. The control room is continuously occupied by the operating personnel under all operating conditions. Control room shielding and ventilation are designed such that the occupants of the room shall not receive doses in excess of 5 rem to the whole body, or its equivalent to any part of the body, during the course of a loss-of-coolant accident. This includes doses received during ingress and egress. The Control Room Air Conditioning System is described in Section 9. 3.1-8 Revision 12 July 22, 1992

Hot shutdown control is provided for as discussed in Section 7. Fire hazards in the control room are limited by ita method of construction and outfitting. Criterion 12 - Instrumentation and Control Systems Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges. Discussion As described in detail in Section 7, sufficient instrumentation and controls are provided to monitor and maintain all operationally important reactor operating parameters such as neutron flux, system pressures, flow rates, temperatures, levels, and control rod positions within prescribed operating ranges. The quantities and types of instrumentation provided are adequate for safe and orderly operation of all systems and processes over the full operating range of the plant. Process variables which are required on a continuous basis for the startup, power operation, and shutdown of the plant are indicated in, recorded in, and controlled as necessary from the control room. The operating staff is cognizant and in control of all test, maintenance and calibration work and can fully assess all abnormal plant conditions knowing the extent to which specific and related operating tasks are in process. Additional details on instrumentation and controls are included in sections relating to specific systems and components. Criterion 13 - Fission Process Monitors and Controls Means shall be provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of 3.1-9 SGS-UFSAR Revision 12 July 22, 1992

the core, such as indication of position of control rods and concentration of soluble reactivity control poisons. Discussion The means provided for monitoring the fission process are indicated in Section 7. The means of determining control rod position are described in section 7 while the means of control and determination of boron concentration are detailed in Section 9. Criterion 14 - Core Protection Systems Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits. Discussion The instrumentation and controls provided to prevent or suppress conditions which could result in exceeding acceptable fuel damage limits are described in section

7. This criterion as applied to the Reactor Protection system is discussed more fully under Criterion 26, Protection Systems Fail-Safe Design.

Criterion 15 - Engineered Safety Features Protection Systems Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features. Discussion The facility is provided with adequate instrumentation and controls to sense accident situations and initiate the operation of necessary engineered safeguards systems. These protection systems are described in Sections 6 and 7. 3.1-10 SGS-UFSAR Revision 16 January 31, 1998

Criterion 16 - Monitoring Reactor Coolant Pressure Boundary '-- Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage. Discussion Positive indications in the control room of leakage of coolant from the Reactor Coolant System to the lower containment compartment are provided by equipment which permits continuous monitoring of the lower containment compartment air activity and humidity, and condensate run-off from the fan coolers. This equipment provides indication of normal background which is indicative of a basic level of leakage from primary systems and components. Any increase in the observed parameters will be an indication of change within the lower containment compartment, and the equipment provided is capable of monitoring this change. The basic design criterion is the detection of deviations from normal containment environmental conditions including air particulate activity, radiogas activity, humidity, condensate, and in addition, in the case of gross leakage, the liquid inventory in the process systems and containment sump. Means of detecting leakage from the Reactor Coolant System is also provided by measuring and indicating changes in makeup requirements and containment sump levels. These leakage detection methods are presented in detail in Section 5. Criterion 17 - Monitoring Radioactivity Releases Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. Discussion The facility contains means for monitoring the containment 3.1-11 SGS-UFSAR Revision 12 July 22, 1992

atmosphere, effluent discharge paths, and the facility environs for radioactivity which could be released under any conditions. The details of the effluent discharge path, containment monitoring methods, and environmental radiation monitoring program are described in Section 11. Criterion 18

Monitorin& Fuel and Waste Storage Monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures.

Discussion Sufficient monitoring and alarm instrumentation is provided in waste and fuel storage areas to detect conditions which might contribute to loss of cooling for decay heat removal or abnormal radiation releases. Details of the monitoring systems are included in Sections 9 and 11. Criterion 19 - Protection Systems Reliability Protection systems shall be designed for high functional reliability and in*service testability commensurate with the safety functions to be performed. Discussion All protection systems are designed for the utmost in reliability based on extensive testing in the shop and many years of actual operating experience. Sufficient redundancy of such systems is provided to enable test of instrumentation channels during plant operation without jeopardizing reactor safety. Detailed descriptions of various portions of the systems are included in Sections 6 and 7. 3.1-12 SGS-UFSAR Revision 12 July 22, 1992

Criterion 20 - Protection Systems Redundancy and Independence Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Discussion As detailed in Section 7, sufficient redundancy and independence is designed into the protection systems to assure that no single failure nor removal from service of any component or channels results in loss of the protection function. In addition, the design of the protection systems conform to IEEE Standard 279-1971, "IEEE Standard Criteria for Protection Systems for Nuclear Power Generating Stations," April 5, 1972 of the Institute of Electrical and Electronic Engineers. Criterion 21 - Sin&le Failure Definition Multiple failures resulting from a single event shall be treated as a single failure. Discussion When evaluating the control, protection, engineered safeguards, and other systems of the facility, multiple failures resulting from a single event are treated as a single failure. The ability of each system to perform its function with a single failure is discussed in the sections describing the individual systems. 3.1-13 SGS-UFSAR Revision 12 July 22, 1992

Criterion 22

Separation of Protection and Control Instrumentation Systems Protection systems shall be separated from control instrumentation systems to the extent that failure or removal from service of any control instrumentation system component or channel, or of those common to control instrumentation and protection circuitry, leaves intact a system satisfying all requirements for the protection channels.

Discussion Protection and control channels in the facility protection systems are designed in accordance with the IEEE Standard 279-1971, "IEEE Standard Criteria for Protection Systems for Nuclear Power Generating Stations," April 5, 1972. The coincident trip philosophy is also employed to prevent a single failure from causing a spurious trip or from defeating the function of any channel. In general, reactor trip circuits are designed so that the trip occurs upon deenergization of the circuit; and open circuit or loss of power to a channel will, therefore, result in the channel going into its trip mode. An exception to this is the solid state protection systems' automatic shunt trip which requires 125 V de to operate the shunt trip coil on the reactor trip breakers. Redundancy within each channel provides reliability and independence of operation. Channel independence is carried throughout the system from the sensor to the relay providing the logic. In some cases, however, it is desirable to employ a common sensor for both a control and protection channel. Both functions are fully isolated in the remainder of the channel, control being derived from the primary safety signal path through an isolation amplifier. As such, a failure in the control circuitry does not 3.1-14 SGS-UFSAR Revision 12 July 22, 1992

adversely affect the safety channel. Those reactor trips requiring energy to trip are arranged such that single power supply failures cannot prevent a trip if required. Criterion 23 - Protection A&ainst Multiple Disability For Protection Systems The effects of adverse conditions to which redundant channels or protection systems might be exposed in common, either under normal conditions or those of an accident, shall not result in loss of the protection function. Discussion Protection system components are designed and arranged so that the mechanical and thermal environment accompanying any emergency situation in which the components are required to function does not interfere with that function. Details of this protection are provided in the appropriate portions of Section 7. Criterion 24 - Emergency Power for Protection Systems In the event of loss of all offsite power, sufficient alternate sources of power shall be provided to permit the required functioning of the protection systems. Discussion The facility is supplied with normal and emergency power supplies to provide for the required functioning of the protection systems. Emergency power for each unit is supplied by three emergency diesel-generators, as described in Sections 7 and 8, with two diesels being capable of supplying all the emergency power requirements of one unit. 3.1-15 SGS-UFSAR Revision 12 July 22, 1992

In addition to the emergency diesel*generators, the instrumentation and controls portions of the protection systems may be supplied from the 125 V de station batteries as detailed in Section 8. Therefore, adequate sources of emergency power are available for all protection systems in the event of a loss of offsite power. Criterion 25 - Demonstration of functional Operability of Protection Systems Means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has occurred. Discussion Each protection channel in service at power is capable of being calibrated and tested at power to verify its operation. Details of the means used to test protection system instrumentation are included in Section 7. Criterion 26 - Protection Systems Fail-Safe Design The protection systems shall be designed to fail into a safe state or into a state established as tolerable on a defined basis if conditions such as disconnection of the system, loss of energy (e.g., electrical power, instrument air), or adverse environments (e.g., extreme heat or cold, fire, steam, or water) are experienced. Discussion The details of the design and failure modes of the various protection channels are to be found in portions of Section 7 concerned with those channels. 3.1-16 SGS-UFSAR Revision 12 July 22, 1992

Criterion 27 - Redund&ncy of Reactivity Control At least two independent reactivity control systems, preferably of different principles, shall be provided. Discussion Two independent reactivity control systems, rod cluster control assemblies, and boric acid dissolved in the reactor coolant, are employed in the facility. Details of the construction and operation of the rod cluster control system are included in Sections 4 and 7. Means of controlling the boric acid concentration are included in Section 9. Criterion 28 - Reactivity Hot Shutdown Capability At least two of the reactivity control systems provided shall be independently capable of making and holding the core subcri tical from any hot standby or hot operation condition including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits. Discussion The rod cluster control system is capable of making and holding the core subcritical from all operating and hot shutdown conditions sufficiently fast to prevent exceeding acceptable fuel damage limits. The chemical shim control is also capable of making and holding the core subcritical, but at a slower rate, and is not employed as a means of compensating for rapid reactivity transients. The rod cluster control system is, therefore, used in protecting the core from such transients. Details of the operation and effectiveness of these systems are included in Sections 4 and 9. 3.1-17 SGS-UFSAR Revision 12 July 22, 1992

Criterion 29

Reactivity Shutdown Capability At least one of the reactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when fully withdrawn shall be provided.

Discussion As detailed in Section 4, the reactor may be made subcritical by the rod cluster control system sufficiently fast to prevent exceeding acceptable fuel damage limits, under all anticipated conditions even with the most reactive rod control cluster fully withdrawn. Criterion 30 - Reactivity Holddown Capability At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies. Discussion The facility is provided with the means of making and holding the core subcritical under any anticipated conditions and with appropriate margin for contingencies. These means are discussed in detail in Sections 4 and 9. Combined use of the Rod Cluster Control System and the Chemical Shim Control System permit the necessary shutdown margin to be maintained during long-term xenon decay and plant cooldown. 3.1-18 SGS-UFSAR Revision 12 July 22, 1992

Criterion 31

Reactivity Control Systems' Malfunction The reactivity control systems shall be capable of sustaining any single malfunction, such as, unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.

Discussion The facility reactivity control systems are such that acceptable fuel damage limits (DNBR 2! 1. 3) will not be exceeded even in the event of a single malfunction of either system. An analysis of the effects of postulated malfunctions is presented in Section 15. Criterion 32 - Maximum Reactivity Worth of Control Rods Limits, which include considerable margin, shall be placed on the maximum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structure, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. Discussion The maximum reactivity worth of control rods and the maximum rates of reactivity insertion employing both control rods and boron removal are limited to values which prevent rupture of the coolant pressure boundary or disrupt the core or vessel internals to a degree which could impair the effectiveness of emergency core cooling. Details of rod worths, reactivity insertion rates, and their relationship to plant safety are included in Sections 4 and 15. 3.1-19 SGS*UFSAR Revision 12 July 22, 1992

Criterion 33

Reactor Coolant Pressure Boundary Capability The reactor coolant pressure boundary shall be capable accommodating without rupture, and with only limited allowance for energy absorption through plastic deformation, the static of and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.

Discussion The primary coolant boundary is designed to accommodate static and dynamic loads associated with sudden reactivity insertions {e.g., rod ejection) without failure. Details of the design may be found in Sections 4 and 5 and an analysis of the effects of such incidents as rod ejection is included in Section 15. The operation of the reactor is such that the severity of an ejection accident is inherently limited. Since control rod clusters are used to control load variations only and core depletion is followed with boron dilution, only the rod cluster control assemblies in the controlling groups are inserted in the core at power, and at full power these rods are only partially inserted. A rod insertion limit monitor is provided as an administrative aid to the operator to assure that this condition is met. By using the flexibility in the selection of control rod groupings, radial locations and position as a function of load, the design limits the maximum fuel temperature for the highest worth ejected rod to a value which precludes any resultant damage to the Reactor Coolant System pressure boundary, i.e., gross fuel dispersion in the coolant and possible excessive pressure surges. 3.1-20 SGS-UFSAR Revision 12 July 22, 1992

The failure of a rod mechanism housing causing a rod cluster to be rapidly ejected from the core is evaluated as a theoretical, though not a credible accident. 'While limited fuel damage could result from this hypothetical event, the fission products are confined to the Reactor Coolant System and the reactor containment. The environmental consequences of rod ejection are less severe than from the hypothetical loss-of-coolant, for which public health and safety is shown to be adequately protected. Criterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions. Discussion As detailed in Section 5, the reactor coolant pressure boundary is designed to minimize the probability of rapidly propagating type failures. To fulfill these requirements, the selection of materials for the system and the fabrication of components are closely controlled and inspected. The details of the material selection and inspection procedures are contained in Section 5. The reactor coolant pressure boundary is designed to reduce to an acceptable level the probability of a rapidly propagating type failure. In the core region of the reactor vessel it is expected that the notch toughness of the material will change as a result of fast 3.1-21 SGS-UFSAR Revision 12 July 22, 1992

neutron exposure. This change is evidence as a shift in the nil ductility transition (NOT) temperature which is factored into the operaLing procedures in such a manner that full operating pressure is not obtained until the affected vessel material is above the design transition temperature (DTT) and in the ductile material The pressure during and shutdown at the temperature below NOT is maintained below the threshold of concern for safe operation. The OTT is a minimum of NOT plus 60°F and dictates the procedures to be followed in the hydrostatic test and in station operations to avoid excessive cold stress. The value of the OTT is increased during the life of the as by the shift in NOT, and as confirmed by the data obtained from irradiated of reactor vessel materials the plant lifetime. Further details are given in Section 5. Under conditions where reactor coolant pressure boundary systems' components constructed of ferritic materials may be subjected to such as a reactivity-induced loading, service temperatures shall be at least 120°F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation, or 60°F above the NDT temperature of the component material if the energy release is to be absorbed within the elastic strain energy range. Discussion Sufficient testing and analysis of materials employed in Reactor Coolant Systems' components will be performed to ensure that the required NOT limits specified in the criterion are met. Removable test capsules will be installed in the reactor vessel and removed 3.1-22 SGS-UFSAR Revision 25 October 26, 2010

and tested at various times in the plant lifetime to determine the effects of operation on system materials. Details of the testing

"-"'. and analysis programs are included in Section 5.

Criterion 36 - Reactor Coolant Pressure Boundary Surveillance Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leaktight integrity of the boundary components during service lifetime. For the reactor vessel, a material surveillance program conforming with ASTM-E-185-66 shall be provided. Discussion Provision has been made in the Reactor Coolant System design for adequate inspection testing and surveillance during the facility's service lifetime. The vessel inspection program will conform to ASTM-E-185-66. These provisions are discussed in detail in Section 5. Monitoring of the RTNDT temperature properties of the core region, weldments, and associated heat treated zones are performed in accordance with ASTM E-185-70 (Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels). Samples of reactor vessel plate materials are retained and cataloged in case future engineering development shows the need for further testing. The material properties' surveillance program includes not only the conventional tensile and impact tests, but also fracture mechanics specimens. The fracture mechanics specimens are the Wedge Opening Loading (WOL) type specimens. The observed shifts in RTNDT of the core region materials with irradiation will be used to confirm the calculated limits to start up and shut down transients. 3.1-23 SGS-UFSAR Revision 12 July 22, 1992

Criterion 37

Engineered Safety Features Basis for Design Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends.

Discussion The containment structure, the Containment Ventilation System, the Emergency Core Cooling System, and the Containment Spray System comprise the engineered safety features for the facility. These systems and their supporting systems (Component Cooling System and Service Water System) are designed to cope with any size reactor coolant pressure boundary break up to and including rupture of the largest reactor coolant pipe. The design bases for each system are included in the appropriate portions of Sections 6 and 9. An analysis of the performance of the safeguards is presented in Section 15. Criterion 38 - Reliability and Testability of Engineered Safety Features All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineered safety features, will be influenced by the known and the demonstrated performance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of the plant. SGS-UFSAR Revision 12 July 22, 1992

                                ---------------~-------*~--~--~--~-~~~~-----------------

Discussion All engineered safety features components were tested in the manufacturers' shop and after installation at the facility to demonstrate their reliability. Provision has also been made in the system design for periodic testing of engineered safety features during the plant lifetime. Details of the test to be performed and the basis for the determination of system reliability are included in Section 6 for the Containment, Containment Ventilation System, Containment Isolation Systems, and for the remaining engineered safety features. Criterion 39 - Emergency Power for Engineered Safety Features Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning required of the engineered safety features. As a minimum, the onsite power system and the offsite power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system. Discussion Reliability of electric power supply is ensured through several independent connections to the system grid, and redundant source of emergency power for each unit from three diesel-generators. Power to the engineered safety features is assured even with the failure of a single active component in each system. The facility electrical systems, including network interconnections and the emergency power system, are described in Section 8. 3.1-25 SGS-UFSAR Revision 12 July 22, 1992

Criterion 40 - Missile Protection Protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures. Discussion All engineered safety features are protected against dynamic effects and missiles resulting from equipment failures, to ensure that their safety function is not impaired. The means for accomplishing this protection are described more fully in Section 3.5. Criterion 41 - En~ineered Safety Features Performance Capability Engineered safety features such as Emergency Core Cooling and Containment Heat Removal Systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component. Discussion Sufficient redundancy and duplication is incorporated into the design of the engineered safety features to ensure that they may perform their function adequately even with the loss of a single active component. Details of the capability of these systems under normal and component malfunction conditions are included in Sections 6 and 9. An analysis of the adequacy of these systems to perform their functions is included in Section 15. Criterion 42 - Engineered Safety Features Components Capability Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident. 3.1-26 SGS-UFSAR Revision 12 July 22, 1992

Discussion The design of the engineered safety features, the materials selected for fabrication of these systems, and the layout of the various portions of the systems combine to ensure that the performance of the engineered safety features are not impaired by the effects of a loss-of-coolant accident. Details of the design and construction of the engineered safety features are included in Sections 6 and 9. The ability of these features to perform their functions is analyzed in Section 15. Criterion 43 - Accident A&&ravation Prevention Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse after effects of the loss of normal cooling is avoided. Discussion The operation of the engineered safety features will not accentuate the after effects of a loss-of-coolant accident. These considerations are detailed in Sections 6 and 15. Criterion 44 - Emergency Core Coolin& Systems' Capability At least two emergency core cooling systems, preferably of different design principles, each with a capability of accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shall not share active components and shall not share other features SGS-UFSAR Revision 12 July 22, 1992

or components unless it can be demonstrated that (a) the capability of the shared feature or component to perform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant -- accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a loss-of-coolant accident and is not lost during the entire period this function is required following the accident. Discussion By combining the use of passive accumulators with two independent high pressure pumping systems and two independent low pressure pumping systems, abundant emergency core cooling is provided even if there should be a failure of any component in any system. A description of the system and its operation is contained in Section 6 and an analysis of the operation of the system under accident conditions is included in Section 15. Criterion 45 - Inspection of Emergency Core Cooling Systems Design provisions shall be made to facilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internals and water injection nozzles. Discussion The design of the emergency core cooling system is such that critical portions are accessible for examination by visual, optical, or other nondestructive means. Details of the inspection program for the reactor vessel internals are included in Section 4 while inspection of the remaining portions of the system is discussed in Section 6. 3.1-28 SGS-UFSAR Revision 12 July 22, 1992

Criterion 46 - Testing of Emergency Core Cooling System Components Design provisions shall be made so that active components of the emergency core cooling systems, such as pumps and valves, can be tested periodically for operability and required functional performance. Discussion The Emergency Core Cooling System design permits periodic testing of active components for operability and required functional performance. The test procedures are described in Section 6. Criterion 47 - Testing of Emergency Core Cooling Systems A capability shall be provided to test periodically the delivery capability of the Emergency Core Cooling Systems at a location as close to the core as is practical. Discussion By recirculation to the refueling water storage tank, the Emergency Core Cooling System delivery capability can be tested periodically. The system can be so tested to the last valve before the piping enters the reactor coolant piping. Details of the system tests are included in Section 6. Criterion 48 - Testing of Operational Seguence of Emergency Core Cooling Systems A capability shall be provided to test under conditions as close to design as practical, the full operational sequence that would bring the Emergency Core Cooling Systems into action, including the transfer to alternate power source, SGS-UFSAR Revision 12 July 22, 1992

Piscussion Provision has been made in the Emergency Core Cooling System design for testing the sequence of operation including transfer to alternate power sources. The details of these tests are included in Section 6, and the switching sequence from normal to emergency power is described in Section 7. Criterion 49 - Containment Design Basis The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions, that could occur as a consequence of failure of Emergency Core Cooling Systems. Discussion The containment structure and its contained Heat Removal Systems are designed to accommodate the pressures and temperatures associated with a loss-of-coolant accident without exceeding the design leak rate. A considerable margin for unidentified energy sources has been included in the design. Containment structural design is also based on the following:

1. Leak tightness and testing requirements

2. Seismic requirements

3. Tornado requirements

4. Shielding requirements S. Design basis accident requirements

6. Flood conditions due to maximum probable hurricane

7. Internal missile generation SGS-UFSAR Revision 12 July 22, 1992

The loadings and energy sources considered in the design and the stress and loading criteria are described in Section 3. An analysis of the performance of the containment during a loss-of-coolant accident is included in Section 15. The Heat Removal Systems are described in Sections 5 and 6. Design of the Containment Building is given in Section 6 and Section 3.8. Criterion 50

NOT Requirement for Containment Material Principal load carrying components of ferritic materials exposed to the external environment shall be selected so that their temperatures under normal operating and testing conditions are not less than 30°F above NDT temperature.

Discussion As stated in Section 5, all principal containment load carrying components of ferritic materials exposed to the external environment are selected to ensure that their temperature under normal operating and testing conditions will be at least Jo*r above NOT temperature. Criterion 51 - Reactor Coolant Pressure Boundary Outside Containment If part of the reactor coolant pressure boundary is outside the containment, appropriate features as necessary shall be provided to protect the health and safety of the public in case of an accidental rupture in that part. Determination of the appropriateness of features such as isolation valves and additional containment shall include consideration of the environmental and population conditions surrounding the site. Discussion The reactor coolant pressure boundary is defined as those 3.1-31 SGS-UFSAR Revision 12 July 22, 1992

piping systems and components which contain reactor coolant at design pressure and temperature. With the exception of the reactor coolant sampling lines, the entire reactor coolant pressure boundary, as defined above, is located entirely within the containment structure. All sampling lines are provided with remotely operated valves for isolation in the event of a failure. These valves also close automatically on a containment isolation signal. Sampling lines can be readily isolated. All other piping and components which may contain reactor coolant are low pressure, low temperature systems which would yield minimal environmental doses in the event of failure. The Sampling System and low pressure systems are described in Section 9. Criterion 52 - Containment Heat Removal Systems Where active Heat Removal Systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided. Discussion The containment heat removal system consists of two subsystems, containment spray (two trains) and containment fan cooling units (five cooling coils). The two subsystems are separate, are operated independently, and are of different design principles, but perform a similar containment heat removal function. The containment heat removal system provides adequate margin for maintaining an acceptable post-accident containment atmospheric pressure and thereby meets the intent of the above criterion. In addition, the containment heat removal system meets the requirements of General Design Criteria 38. The design and performance of containment spray and the containment fan cooling units are discussed in sections 6.2.2.1, 6.2.2.2, and 15.4. Criterion 53 - Containment Isolation Valves Penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus. 3.1-32 SGS-OFSAR Revision 20 May 6, 2003

Discussion At least two barriers are provided between the atmosphere outside the containment and the containment atmosphere, the Reactor Coolant System, or closed systems which are assumed vulnerable to accident forces. The valving installed on the various systems penetrating the containment and the other barriers employed in the design are described in Section 5. Criterion 54 - Containment Leakage Rate Testing Containment shall be designed so that integrated leakage rate testing can be conducted at design pressure after completion and installation of all penetrations and the leakage rate measured over a sufficient period of time to verify its conformance with required performance. Discussion Provision is included in the containment design for integrated leak rate testing after completion of construction. The test program and procedures are described in Section 6.2 and are formulated to demonstrate that leakage is below the design value of 0.1 percent per day. Criterion 55 - Containment Periodic Leakage Rate Testing The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime. Discussion Provision for full integrated leakage rate testing is incorporated into the design. The testing program and procedures are described in Section 6.2. 3.1-33 SGS-UFSAR Revision 12 July 22, 1992

Criterion 56 - Provision for Testin& of Penetrations Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leak tightness to be demonstrated at design pressure at any time. Discussion The Containment Penetration Pressurization System provides a means to test the leak tightness of penetrations at any time. This system is described in Section 6.2 and the leak testing program is described in Section 6.3. Criterion 57 - Provision for Testin& of Isolation Valves Capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limits. Discussion Provisions have been made in the plant design for testing the functional operability of containment isolation valves. The Containment Isolation System is described in Section 6.2. Criterion 58 - Inspection of Containment Pressure-Reducing Systems Design provisions shall be made to facilitate the periodic physical inspection of all important components of the containment pressure-reducing systems, such as, pumps, valves, spray nozzles, torus, and sumps. SGS-UFSAR Revision 12 July 22, 1992

Discussion The design of the Containment Ventilation System and the Containment Spray System includes provision for physical inspection of vital components. The inspectability of these systems is discussed in Sections 5 and 6. Criterion 59 - Testing of Containment Pressure-Reducing Systems' Components The containment pressure-reducing systems shall be designed so that active components, such as pumps and valves, can be tested periodically for operability and required functional performance. Discussion Component testing of the Containment Spray System is discussed in detail in Section 6. Containment ventilation testing is discussed in Sections 5 and 6. Criterion 60 - Testing of Containment Spray Systems A capability shall be provided to periodically test the delivery capability of the Containment Spray System at a position as close to the spray nozzles as is practical. Discussion All active portions of the Containment Spray System may be tested. The delivery capacity may be tested up to the last valve before the spray header. Details of the Containment Spray System are included in Section 6. 3.1-35 SGS-UFSAR Revision 12 July 22, 1992

Criterion 61 - Testing of Operational Seguence of Containment Pressure-Reducing Systems A capability shall be provided to test under conditions as close to the design as practical, the full operational sequence that would bring the containment pressure-reducing systems into action, including the transfer to alternate power sources. Discussion Capability for testing of the operational sequence of the Containment Spray System is incorporated into the system design. Provision is also included for testing the Containment Ventilation System. Details of the Containment Spray System are included in Section 6. The switching sequence from normal to emergency power is described in Section 7. The Containment Ventilation System is described in Section 6. Criterion 62 - Inspection of Air Cleanup Systems Design provisions shall be made to facilitate physical inspection of all critical parts of the containment air cleanup systems, such as ducts, filters, fans, and dampers. Discussion The containment ventilation systems, consisting of 5 fan coolers and high-efficiency particulate air filters serves as an air cleanup system for the containment. Section 6.2 discusses the inspection of the Containment Fan Cooler System. 3.1-36 SGS*UFSAR Revision 12 July 22, 1992

Criterion 63 - Testin& of Air Cleanup Systems' Components Design provisions shall be made so that active components of the air cleanup systems, such as fans and damper, can be tested periodically for operability and required functional performance. Discussion Testing of the Containment Fan Cooler and Containment Spray Systems' Components is discussed in detail in Section 6.2. Criterion 64 - Testins of Air Cleanup Systems A capability shall be provided for in situ periodic testing and surveillance of the air cleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits. Discussion Testing of the Containment Fan Cooler System and Containment Spray System is discussed in detail in Section 6.2. Criterion 65 - Testing of Operational Sequence of Air Cleanup Systems A capability shall be provided to test under conditions as close to design as practical, the full operational sequence that would bring the air cleanup systems into action, including the transfer to alternate power sources and the design air flow delivery capability. Discussion A discussion of the operational sequence testing of the Containment Fan Cooler System and Containment Spray System is included in Section 6.2. 3.1-37 SGS-UFSAR Revision 12 July 22, 1992

Criterion 66 - Prevention of fuel Storage Criticality Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls. Discussion Criticality in new and spent fuel storage areas is prevented both by physical separation of new and spent fuel elements and the presence of borated water in the spent fuel storage pool. Criticality prevention in fuel storage areas is discussed in Section 9. Criterion 67 - Fuel and Waste Storage Decay Heat Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs. Discussion The Spent Fuel Pool Cooling System provides decay heat removal for the spent fuel pool. In addition, the water in the pool is sufficient to absorb the decay heat produced from 1 1/3 spent cores. Details of the Spent Fuel Pool Cooling System and fuel handling facilities are described in Section 9. Criterion 68 - fuel and Waste Storaee Radiation Shielding Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10CFR20. 3.1-38 SGS-UFSAR Revision 12 July 22, 1992

Shielding is provided for fuel handling and waste storage areas to lower radiation doses to levels below limits in 10CFR20. for these areas and other shielding requirements and criteria are included in Section 12. Criterion 69 - Protection Against Radioactivity Release From Spent Fuel and Waste Storage Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs. All fuel storage and waste storage facilities are designed to prevent the undue release of radioactivity to the public. Fuel storage facilities are described in Section 9; waste storage facilities are described in Section 11 and analysis of postulated accidents is included in Section 15. The shall include those means necessary to maintain control over the plant radioactive whether gaseous, liquid, or solid. Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR20 requirements for normal operaLions and for any transient situation that might reasonably be anticipated to occur and (b) on the basis of 10CFR50. 67 dosage level for reactor accidents of 3.1-39 SGS-UFSAR Revision 25 October 26, 2010

exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very cities can be affected by the radioactive effluents. Discussion Provision is included in the facility design for storage and processing of radioactive waste and the release of such wastes under controls adequate to prevent exceeding the limits of 10CFR20. The facility also includes provision to prevent radioactivity releases during accidents from exceeding the guidelines of 10CFR50. 67. A description of the Radioactive Waste Disposal System is included in Section 11. The effects of accidents, including a loss-of-coolant accident, are in Section 15. 3.1-39a SGS-UFSAR Revision 25 October 26, 2010

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3.1.3 Conformance with AEC General Design Criteria {July 1971) The Salem Plant design conforms with the intent of "General Design Criteria for Nuclear Power Plants," dated July 7, 1971, with the ~xception of those items listed below. Criterion 4

Environmental and Missile Design Basis The design of Salem Unit 1 complies with General Design Criteria 4 (GDC 4) with respect to protection against the dynamic effects associated with the postulated failure of piping. The PSE&G approach to evaluating high-energy line break consequences is described in Section 3. 6 of the UFSAR and is consistant with the guidance provided by A. Giambusso, Atomic Energy Commission (AEC),

to all licensees in his letter dated December 1972, "General Information Required for Consideration of the Effects of a Piping System Break Outside Containment." For Unit 1, high energy piping systems are those whose temperature exceeds 200°F and whose pressure exceeds 275 psig, coincidentally, during normal operation. Design basis cracks only are postulated for those systems whose pressure is more than 275 psig or whose temperature is more than 200*F. The design of Salem Unit 2 also complies with GDC 4 with respect to protection against the dynamic effects associated with the postulated failure of piping. However, for Unit 2, the criteria are provided by Branch Technical Position APCSB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment." For Unit 2, high-energy piping systems are those whose temperature exceeds 2oo*p or 275 psig during normal operation. This revised criteria resulted in three additional Unit 2 systems requiring analysis as high-energy. Those systems were: CVC charging and Reactor Coolant Pump seal injection, Heating Steam, and Heating Yater. In addition to the revised temperature and pressure criteria, NRC required that a Moderate Energy Break Analysis (MEBA) be performed for Unit 2. 3.1-40 SGS-UFSAR Revision 12 July 22, 1992

Criterion 55 - Reactor Coolant Pressure Boundary Penetrating Containment Criterion 56 - Primary Containment Isolation Criterion 57 - Closed System Isolation Valves Valving arrangements which do not conform with these criteria are discussed in Section 6.2.4. 3.1.4 PSE&G General Criteria Quality and Performance Standards Those features of the reactor facility which are essential to the prevention of accidents which could affect the public health and safety or the mitigation of their consequences are designed, fabricated, and erected to:

1. Quality standards that reflect the importance of the safety function to be performed. Recognized codes and standards are used when appropriate to the application.

2. Performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces imposed by the most severe 3.1-40a SGS-UFSAR Revision 7 July 22, 1987

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earthquakes, flooding conditions, winds, ice, or other natural phenomena characteristic of the Salem site. Features of the facility essential to accident prevention and mitigation are the fuel, reactor coolant, and containment barriers; the controls and emergency cooling system whose function is to maintain the integrity of these three barriers; systems which depressurize and reduce the radioactivity level in the containment; power supplies and essential services to the above features; and the components employed to safely convey and store radioactive wastes and spent reactor fuel. Quality standards for material selection, design, fabrication, and inspection governing the above features conform to the applicable provisions of recognized codes and standards. The concrete structure of the reactor containment conforms to the applicable portions of ACI-318-63. Further elaboration on quality standards for the reactor containment is presented in Section 5. Vessels comply with the ASME Boiler and Pressure Vessel Code under the specific classification dictated by their use, or other appropriate codes. The principles of this code, or equivalent guidelines, are employed where the code is not strictly applicable but where the safety function calls for an equivalent assurance of quality. In the same manner, piping conforms to the requirements of the USA Standard Code for Pressure Piping (ANSI B31.1) and Nuclear Code Cases N-7 and N-10. Particular emphasis is placed on the assurance of quality of the reactor vessel by obtaining material whose properties are uniformly within tolerances appropriate to the application of the design methods of the Code. The fatigue usage factor is less than that at which propagation of material defects would occur. Design margin and material surveillance assure that the vessel will be operated well within the ductile range of temperatures when vessel stresses are above 10,000 psi. The reactor vessel size is within the range of previous experience of the manufacturer and of the nuclear plant designer. Further discussion of quality assurance for the reactor vessel is presented in Section 4. 3.1-41 SGS-UFSAR Revision 6 February 15, 1987

All piping components and supporting structures of the reactor and safety-related systems are designed to withstand any seismic disturbance predictable for the site. The dynamic response of the structure-to-ground acceleration, based on appropriate spectral characteristics of the site foundation and on the damping of the foundation and structure, is included in the design analysis. Structural, equipment, and piping materials used in both the containment and Auxiliary Building have been selected for their compatibility with normal and accident environments. For those items located inside the containment which are required for controlling the design basis accident, the effect of the spray chemical additive (NaOH) has been considered as well as radiation levels, pressure, and temperature. Material compatibility has been discussed in detail in the Indian Point Unit 2 FSAR (Docket No. 50-247). Fire Protection Fire protection facilities are provided in accordance with the recognized guidelines of the Nuclear Energy Property Insurance Association, National Fire Protection Association, and Underwriters Laboratory. Protective features such as fire doors and closed ventilation systems are provided to minimize the possibility of fire or smoke in the control room. The control room is designed and equipped to assure continuous occupancy. Section 9 outlines the basic design and operational features of the plant Fire Protection System. Records Requirements PSE&G (or its authorized representatives) and Westinghouse have retained complete documentation of the design, fabrication, and construction of all essential plant components.

3. 1-42 SGS-UFSAR Revision 6 February 15, 1987

These records are available to verify the high quality and performance standards applicable to all essential plant components. Protection by Multiple Fission Product Barriers Physical barriers are provided by the fuel pellet, fuel cladding, Reactor Coolant System pressure boundary and containment structure to protect the public from the release of fission products produced within the fuel assemblies. The specific details and design basis for each barrier are identified and discussed in Sections 3, 4, and 5. The design of the fuel cladding, core, related structural equipment, and control and protective systems ensures that fuel damage in excess of acceptable limits is not likely, or can be readily suppressed in the unlikely event of its occurrence. The Reactor Coolant System, including the reactor pressure vessel, is designed to accommodate the system pressure and temperatures attained under all expected modes of plant operation, and maintain the stress within applicable code stress limits. Its materials of construction are protected from corrosion phenomena by control of coolant chemistry. It is protected from overpressure by means of relieving devices. High-pressure equipment in the Reactor Coolant System is surrounded by barriers to prevent a missile, generated from the Reactor Coolant System in a loss-of-coolant accident, from reaching either the containment liner or the containment cooling equipment, and from imparing the function of the engineered safety features. The principal missile barriers are the concrete operating floor and the reinforced concrete shield wall enclosing the reactor coolant loops. The pressurizer is protected by a completely enclosed concrete and steel plant compartment constructed above the operating floor. A steel and concrete structure is also provided over the control rod drive mechanisms 3.1-43 SGS-UFSAR Revision 6 February 15, 1987

to block a missile generated from fracture of the mechanism housing. The Reactor Coolant and reactor vessel are enclosed within the containment structure. The containment structure itself is designed to withstand the temperature and pressure conditions associated with the severance of a reactor coolant pipe coincident with a seismic occurrence. Essentially no leakage of radioactive materials to the environment will result under these conditions. Monitoring potentially radioactive areas and operation of the reactor protection and reactor control and turbine is in the control room from which all actions to maintain the safe operational status of the plant are centered. Radiation protection is provided to permit access to equipment in the control room, even under accident conditions, as necessary to shut down and maintain safe control of the facility without radiation exposures to personnel in excess of Code of Federal Regulations' limits. The control room is equipped with those controls which are necessary for monitoring and maintaining control over the fission process and for all conditions that could be to cause variations in core In addition to instrumentation and controls which are required to maintain variables within operating ranges, means are provided to monitor fuel and waste storage handling areas, reactor coolant pressure boundary leakage, containment atmosphere, and all potentially contaminated facility effluent discharge paths. Core protection systems automatically sense accident situations, initiate the operation of necessary engineered safety features that prevent or suppress conditions that could result in 3.1-44 SGS-UFSAR Revision 25 October 26, 2010

fuel damage limits. This combination of monitoring and core protection systems provides the assurance that any radioactive releases are maintained well below established Federal regulatory limits for normal operations, anticipated transients, and possible accident conditions. Positive indications in the control room of leakage of coolant from the Reactor Coolant System to the containment are provided by equipment which permits continuous monitoring of the containment air activity and humidity. The basic design criterion is the detection of deviations from normal containment environmental conditions, including air particulate activity, radiogas activity, humidity, and in addition, in the case of gross leakage, the liquid inventory in the process systems and containment sump. The containment atmosphere, Fuel Handling Building, plant vent, containment fan coolers, steam generator blowdown, the condenser vacuum pump exhaust, and the Waste Disposal System liquid effluent are monitored for radioactivity concentration during all normal operations, anticipated transients, and accident conditions. For the case of leakage from the reactor containment under accident conditions, the plant area Radiation Monitoring System, supplemented by portable survey equipment stored in the control room, provides adequate monitoring of releases during an accident. Monitoring and alarm instrumentation are provided for fuel and waste storage and handling areas to detect inadequate cooling and to detect excessive radiation levels. Radiation monitors are provided to maintain surveillance over the release of radioactive gases and liquids. A controlled ventilation system removes gaseous radioactivity from the atmosphere of the fuel storage and waste treating areas of the Auxiliary Building and discharges it to the atmosphere via the unit vent. Radiation monitors are in continuous service in these

3. 1-45 SGS-UFSAR Revision 6 February 15, 1987

areas to actuate high-activity alarms on the overhead annunciator in the control room, as described in Section 11. Reliability and Testability of Protection Systems Protection systems are designed with a degree of functional reliability and in-service testability which is commensurate with the safety functions to be performed. System design incorporates such features as emergency power availability, preferred failure mode design, redundancy, and isolation between control systems and protective systems. In addition, the protection systems are designed such that no single failure will prevent proper system action when required. For design purposes, multiple failures which result from a single event are considered single failures. The proposed criteria of the Institute of Electrical and Electronic Engineers for nuclear power plant protection (IEEE-279) have been utilized in the design of protective systems. The plant variables monitored and the sensors utilized are identified and discussed at length in Westinghouse proprietary reports submitted in support of this application, and referenced in Section 7. The coincident trip philosophy is carried out to provide a safe and reliable Reactor Protection System since a single failure will not defeat its function nor cause a spurious reactor trip. Channel independence originates at the process sensor and continues back through the field wiring and containment penetrations to the analog protection racks. The power supplies to the protection sets are fed from instrumentation buses which are capable of being powered from the diesel-generators. Two reactor trip breakers are provided to interrupt power to the rod drive mechanisms. The breaker main contacts are connected in series. Opening either breaker will interrupt power to all mechanisms, causing all rods to fall by gravity into the core. Manual trip also actuates the shunt trip breakers. Each 3.1-46 SGS-UFSAR Revision 6 February 15, 1987

protection channel feeds two logic matrices, one for each undervoltage trip circuit. In general reactor trip circuits are designed so that a trip occurs when the circuit is de-energized; an open circuit or loss of channel power therefore would cause the affected circuits to go into a trip mode. Reliability and independence is obtained by redundancy within each channel, except for backup reactor trips such as the reactor coolant pump breaker. Reactor trip is implemented by interrupting power to each control rod drive mechanism, allowing the rod clusters to be inserted by gravity. The protection system is thus inherently safe in the event of a loss of rod control power. Those trips requiring energy to operate are arranged such that single power failures cannot prevent a reactor trip if required. The components of the protection system are designed and laid out so that the mechanical and thermal environment accompanying any emergency situation in which the components are required to function will not interfere with that function. The actuation of the engineered safety features provided for loss-of-coolant accidents, e.g., Emergency Core Cooling System pumps, reactor containment fan coolers, and Containment Spray Systems, is accomplished from redundant signals derived from the Reactor Coolant System, steam flow, and containment instrumentation. Channel independence originates at the process sensor and is carried through the analog protection racks. The initiation signal for containment spray comes from a high-high containment pressure signal. De-energizing a channel will cause that channel to go into its tripped mode, with the exception of containment spray which is energized to actuate. A comprehensive program of plant testing has been formulated for equipment vital to the functioning of engineered safety systems. The program consists of performance tests of individual pieces of equipment in the manufacturer's shop, integrated tests of the 3.1-47 SGS-UFSAR Revision 6 February 15, 1987

system as a whole, and periodic tests of the actuation circuitry and the performance of mechanical components to assure reliable performance upon demand throughout the plant lifetime. The following series of periodic tests and checks can be conducted to assure that the systems can perform their design functions whenever they should be called on during the plant lifetime:

1. Integrated Test Actuation Circuits and Motor-Operated Valves The automatic actuation circuitry, valves, and pump breakers can be checked during integrated system tests performed during each planned cooldown of the Reactor Coolant System for refueling.

2. Accumulator Tanks The accumulator tank pressure and level are continuously monitored during plant operation, and flow from the tanks can be checked at any time using test lines.

3. Safety Injection, Residual Heat Removal, Containment Spray and Charging Pumps The safety injection and containment spray pumps can be tested periodically during plant operation using the minimum flow recirculation lines provided. The residual heat removal pumps are used every time the residual heat removal loop is put into operation and can be tested periodically on minimum flow. The charging pumps are normally run during plant operation. All remote operated valves can be exercised and actuation circuits can be tested periodically during plant operation or routine maintenance.

4. Reactor Containment Fan Coolers 3.1-48 SGS-UFSAR Revision 6 February 15, 1987

The reactor containment fan coolers and the service water pumps operate routinely, and no additional periodic test is required.

5. Boric Acid Concentration in the Accumulators The accumulators and lines are charged with borated water at refueling water concentration of at least 2200 ppm while the plant is in operation. This concentration is checked periodically by sampling.

6. Boron Injection Tank The boron in this tank is maintained at or below the maximum concentration allowable for the RWST (2500 ppm). There is no minimum boric acid concentration. The accident assume 0-ppm boric acid concentration in the BIT.

7. Chemical Concentration in the Spray Additive Tank The concentration of chemical solution in this tank is maintained at approximately 30 weight percent NaOH.

8. Emergency Power Sources The sets can be started from the control room. The of the units to start and accept in 13 seconds is checked.

9. Containment Penetrations Penetrations are designed with double seals and the large access openings such as the equipment hatch and personnel air locks are equipped with double seals. Double seals are so that air leak B) tests can be conducted.

10. Instrumented Protection Channels 3.1-49 SGS-UFSAR Revision 25 October 26, 2010

All reactor protection channels, with the exception of backup reactor trips, are supplied in sets which provide the capability for channel calibration and test. Bypass removal of a trip circuit is used only in 2/4 logic which then becomes 2/3 logic, except for special 1/2 logic such as startup trips which become 1/1 logic. Reactor Protection System protection channels in service at power are capable of being tested to verify operation. The operability of a reactor trip channel can be determined conveniently and without ambiguity. A complete channel test can be performed through and including the final trip breakers, excluding the sensor. Actuation of the engineered safety features including containment isolation also employs coincidence circuits which allow checking of the operability of one channel at a time. Removal or bypass of one signal channel places that circuit in the tripped mode. The normal on-line test procedure (exceptions noted above) consists of tripping the channel downstream of the on-off controller (process control) or superimposing the test signal on the transmitted signal (NIS Power Range). In the process control equipment, the 2/4 logic goes to 1/3 remaining, and 2/3 logic goes to 1/2 remaining. The transmitted signal is disconnected and a precise simulated signal is injected. The trip points are then checked against the precise signal. In the Nuclear Instrumentation System (NIS) power range equipment, a precise signal is superimposed on the exis-ting input signal and the trip point is checked against the combined signal. Sensors are checked by comparing their outputs to each other. 3.1-50 SGS-UFSAR Revision 6 February 15, 1987

Two independent reactivity control systems, of different design in the reactor system These are neutron control rods and chemical of the reactor coolant with boron. The react worth of the worth control rod is less than that to achieve criticality with that rod out of the core and all the remaining control rods fully inserted in the core. The Reactor Coolant System has been designed so that static and dynamic loads on boundary as a result of any inadvertent and sudden release of energy to the coolant will not cause rupture of the pressure boundary. In order to continually guard any weakness the reactor coolant pressure containing components have provisions for inspection and testing to assess the structural and leaktight integrity of the boundary components during their service lifetime. The features provided in this have sufficient redundancy of components and power sources so that even under the conditions of the design basis accident, the systems can, even when with effectiveness, maintain the required integrity of fission product barriers to keep exposure of the public well within the guidelines of 10CFR50.67 (Code of Federal Regulations). A general explanation of each of the engineered safety features follows. Specific details on system design and operation are covered in Section 6.

1. A steel lined concrete containment structure an reliable final barrier the escape of 3.1-51 SGS-UFSAR Revision 25 October 26, 2010

fission products. Containment penetrations, including access openings and ventilation ducts, are provided with double seals. the containment which could become a to the environment following the basis accident are provided with isolation valves.

2. An Core Cooling is to deliver borated water to the core, in the event of a loss-of-coolant accident, in three modes: accumulator active injection, and residual heat removal circulation. The design provides for periodic testing of active components for operability and required functional performance as well as incorporates provisions to facilitate physical inspection of all critical components.

3. Active heat ::::emoval systems are provided within the containment to cool the containment atmosphere under design basis accident conditions. Two systems of different design are the Containment and the Reactor Containment Fan Cooler the worst-case active failure under basis accident conditions, these systems have sufficient redundancy of components to maintain an acceptable post-accident containment pressure.

The Containment: Spray System is also instrumental in the removal of elemental iodine from a post-LOCA atmosphere, thus reducing the concentration of halogen fission products. Further details regarding the iodine scrubbing ability of the containment spray system are discussed in section 5.2.3.1. fuel storage in the fuel and waste facilities are contained and the is such that accidental releases of to the will not exceed the limits of 10CFRlOO. Refer to Section 9. 7. Separate fuel storage facilities are provided for each unit. An Independent Spent Fuel Storage Installation ( ISFSI), which for additional storage of both Salem and Hope Creek spent fuel in a dry configuration, is located on site, north of the Hope Creek Generating Station. See Section 9.1.4.5 for additional infor~ation pertaining to the ISFSI. 3.1-52 SGS-UFSAR Revision 25 October 26, 2010

During the refueling of the reactor, all operations are conducted with the spent fuel under water (see Section 9.7). This provides visual control of the at all times and also maintains low radiation levels. The borated re water assures at all times and also for the fuel during transfer. fuel is taken from the reactor and transferred to the canal and in the Fuel Transfer System. Rod cluster control assembly transfer from a spent fuel assembly to a new fuel assembly is accomplished prior to transferring the spent fuel to a spent fuel storage pool. Each spent fuel storage pool is supplied with a cooling system for the removal of the decay heat of the spent fuel prior to storage at the ISFS: or shipment from the site. Racks are provided to accom..rnodate the of a total of 6 cores. The pools are filled with borated water at a concentration to match that used in the reactor and canal during The spent fuel is stored in a vertical array with sufficient center-to-center distance between assemblies to assure subcriticality (kef£ <0.95) even if unborated water were introduced into the The racks are designed so that it is to insert assemblies in other than the locations. The water level maintained in the pool will sufficient to normal occupancy of the area by Each fuel pool is also with systems to maintain water cleanliness and to indicate pool water level. Gamma radiation is continuously monitored in the Auxiliary Building and a high level is annunciated in the control room. Water removed from the pools must be pumped out as there are no gravity drains. or leakage of any from waste handling facilities go to floor drains which flow to sumps. Postulated accidents the release of from the fuel and waste storage and facilities are shown in Section 14.3 to result in exposures well within the guidelines of 10CFR50.67. The reactor refueling cavity, refueling canals and the spent fuel storage pools are reinforced concrete structures with stainless 3.1-53 SGS-UFSAR Revision 25 October 26, 2010

steel liners. These structures are designed to withstand the anticipated earthquake loadings so that the liner will prevent gross leakage. The transfer tubes which connect the canals and the spent fuel and form part of the reactor containment are provided with valves and blind which close off the transfer tubes when not in use. Effluents Gaseous, liquid, and solid waste disposal faci:ities are designed so that discharge of effluents and offsite shipments shall be in accordance wiLh government regulations. Process and streams are monitored and features are to releases in excess of the limits of 10CFR20. Environmental conditions do not any restrictions on the normal release of operational radioactive effluents to the atmosphere. Radioactive fluids entering the Waste Disposal System are collected in analysis tanks until the course of subsequent treatment is determined. Radioactive gases are pumped by compressors through a manifold to one of the waste gas tanks where are held a sui table of time for decay. Tanks are provided for the normal operations of filling, isolation for decay, and discharge. During normal operation decayed gases are discharged intermittently at a controlled rate from the tanks through the monitored plant vent. All solid wastes are placed in sui table containers and stored onsite until shipment offsite for disposal. wastes are to remove most of the radioactive material. The spent resins from the demineralizers, the filter and the concentrate from the evaporators are and stored onsite until offsite for disposal. The processed water, from which most of the radioactive material has been removed, is discharged through a monitored line into the condenser discharge. 3.1-54 SGS-UFSAR Revision 25 Oc~ober 26, 2010

3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Certain structures, components, and systems of a nuclear power plant are considered important because they perform safety functions required to avoid or mitigate the consequences of abnormal operational transients or accidents. This section discusses the classification of structures, components, and systems according to the importance of the safety function they perform. In addition, design requirements are placed upon such equipment to ensure the proper performance of safety actions when required. 3.2.1 Seismic Classification 3.2.1.1 Definition of Seismic Design Classifications Structures and equipment have been divided into three classes for the purpose of establishing the seismic design requirements. Each structure, system, component, and parts thereof are classified in accordance with the following definitions. Class I Those structures and components, including instruments and controls, whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of excessive amounts of radioactivity. Also, those structures and components vital to safe shutdown and isolation of the reactor. Class II Those structures and components which are important to reactor operation but not essential to safe shutdown and isolation of the reactor, and whose failure could not result in the release of substantial amounts of radioactivity. 3.2-1 . SGS-UFSAR Revision 6 February 15, 1987

Class III Those structures and components which are not related to reactor operation or containment. seismic design measures for these three classifications are described in Sections 3.7 and 5.2. 3.2.1.2 Seismic Classification of Structures, Systems, and Components The classifications which follow are intended, by example, to convey the application of the seismic classification definitions. Class I The following list establishes a general category of Class I items: Structures:

1. Containment (including penetrations and air locks, the concrete shield, and the interior structures)

2. Fuel Handling Building

3. Control Room

4. Auxiliary Building

5. Service Water Intake Structure 3.2-2 SGS-UFSAR Revision 17 October 16, 1998

Equipment, Piping, and Supports:

1. Reactor Protection System, portions of the Radiation Monitoring System System and Process Instrumentation and Controls as required for Class I equipment and systems

2. Reactor Vessel and its supports Vessel Internals Fuel Assemblies RCC assemblies and drive mechanisms Supporting and positioning members In-core instrumentation structure

3. Reactor Coolant System Piping and valves (including safety and relief valves)

Steam Generators Pressurizer Reactor Coolant Pumps Reactor Coolant System supports

4. Chemical and Volume Control System (portions) 3.2-3 SGS-UFSAR Revision 6 February 15, 1987

5. Engineered Safety Features Emergency Core Cooling System (including Safety Injection and Residual Heat Removal Pumps, Refueling Water Storage Tank, Accumulators, Boron Injection Tank, Residual Heat Exchangers, connecting piping and valving)

Containment Spray System (including Spray Pumps, Spray Headers, Spray Additive Tank, and connecting piping and valving) Containment Ventilation System (including fan coolers, distribution ducts, dampers, HEPA filters, and moisture separators)

6. Auxiliary Building Ventilation System (supply and exhaust units)

7. Fuel Handling Building Ventilation System (exhaust units)

8. Auxiliary Feedwater Storage Tanks

9. Residual Heat Removal System

10. Component Cooling System

11. Fuel Transfer Tube

12. Emergency Power Supply Systems Diesel-Generators and associated fuel oil lubricating oil, starting auxiliary systems, fuel storage, and day tanks Diesel-Generator Area Ventilation System 3.2-4 SGS-UFSAR Revision 6 February 15, 1987

DC Power Supply System Power distribution lines to equJ.pment required for emergency transformers and switchgear supplying the Engineered Safety Features Control Boards Motor Control Centers

13. ConLrol equipment, facilities and lines as for the preceding items

14. Containment Polar Crane lS. Auxiliary Feedwater and Service Water (portions)

16. Sampling System Piping (to outermost containment isolation valve)

17. Main Steam System (to isolation valve)

.B. Feedwater (to outermost containment isolation valve) l9. Combustible Gas Control System (partial)

20. Fuel Handllng System

21. Instrumentation and Control Systems required for safe shutdown, sa instrumentation Electrical Cable Tunnels

23. fuel pool cooling and purification system piping (SFPC components have been seismicly evaluated under SQUG GIP methodology) .

QA program controls as identified in Section 17.2 are applied, but not limited

o, the above Class I systems, structures, and components 3.2-5 SGS-UFSAR Revision 18

The following list establishes a general category of Class II items:

1. Pressurizer Relief Tank

2. Sampling System (partial)

3. (This text has been deleted)

4. Holdup Tank Transfer Pumps

5. Evaporator

6. Evaporator Condensate Demineralizers

7. Waste Monitor Tanks

8. Waste Monitor Tank Pumps

9. Primary Water Storage Tanks

10. Concentrates Holdup Tank ll. Waste Gas Disposal System

~lass

he following list establishes a general category of Class III items

Turbine Generator Area Structure Buildings containing conventional facilities

3. Waste Disposal System (partial) 3.2-6 SGS-UFSAR Revision 18 April 26, 2000

4. Chemical Mixing Tank

5. Resin Fill Tank

6. Demineralized Water Storage Tanks

7. Conventional equipment, tanks, and piping other than Classes I and II 3.2.1.3 Seismic Criteria For Class I (seismic) equipment, dynamic methods or conservative static equivalents were used to determine that components and structures will operate or maintain their integrity, as required. For Class .II (seismic) equipment, static methods were used and Class III (sesimic) equipment meets applicable codes.

3.2.2 System Quality Considerations 3.2.2.1 Codes and Standards The codes and standards applied in the design of plant systems are described in the section of the FSAR containing the respective system description. 3.2.2.2 ANSI B31.7 ANSI 831.7 was used for piping design wherever possible. Where not possible to comply with ANSI B31.7, the requirements of ASME III - 1971 were adhered to. The major deviations from ANSI B31.7 requirements involve radiographic inspection technique. The original edition of 831. 7 was deficient in this area as evidenced by the need for ANSI 831 Code Case 72. The ANSI committee recognized that the radiographic requirements of B31.7 were not suitable for field radiography of thin walls and small diameter piping butt welds. The 3.2-7 SGS-UFSAR Revision 6 February 15, 1987

in.CC)r]::)Ora.tion of Code Case 72, without change, into the '1971 edition of Section III of the ASME Boiler and Pressure Vessel Code and its retention up to the present edition demonstrates that it provides a realistic approach to field radiography. In addition, radiographs of Nuclear Class 3 cement lined pipe were difficult to interpret. The 1970 addenda to 831.7 allowed 100 percent magnetic particle inspection in lieu of random radiography. This provision was also incorporated into Section III, 1971 Edition. The Service Water System contains Nuclear Class 3 cement lined pipe for which this alternate inspection method was utillzed. In addition, the weld inspection criteria of later Editions and Addenda of ASME III, as approved by the NRC, can be specified. The use of a later code was restricted to inspection ~nd did not involve any requirements from Section III such as materials, stress calculations, etc, that would modi our original design. Consequently, other from a later Code would not be applicable. Therefore, the integrity of field welds has not been compromised and Public Service Electric and Gas has complied with the comm~tment to use ANSI 831.7 wherever possible. 3.2.2.3 Field- Run Plping Field runnlng of small diameter piping, i.e., complete assembly at the erection without reference to design drawings, is not permitted for essential sys::e~s. Therefore, no special ity assurance measures or performance tests are required. 3.2-8 SGS-UFSAR Revision 18 April 26, 2000

3.3 WIND AND TORNADO LOADINGS 3.3.1 Wind Loadings 3.3.1.1 Design Wind Velocity and Loadings A wind load of 30 pounds per square foot, equivalent to 108 mph, was applied to Category I structures and was found to be less critical than the operating basis earthquake load. 3.3.2 Tornado Loadings 3.3.2.1 Tornado Parameters The Reactor Containment, Fuel Handling, and Auxiliary Buildings have been checked against a tornado loading based on a peripheral wind velocity of 300 mph and a translational velocity of 60 mph. Simultaneous with wind loading, an atmospheric pressure drop of 3 psig has been considered . 3.3.2.2 Determination of Forces on Structures The three tornado wind distributions shown on Figure 3. 3-1 were investigated in the Category I structural design. In combination with the static forces produced by the 360 mph maximum wind, a 3 psig atmospheric pressure drop was specified for the containment structure. The shape factor, C, for the dome was taken as 0.4 and for the cylinder, 0.5. No gust factor was applied. The experimental pressure distribution curves by Hoerner and Born were considered by Conrad Associates to be representative for the size of the containment cylinder. To obtain conservative membrane shear stresses, an upper-bound pressure distribution function minimizing the windward side suction component was adopted as shown by the solid line on Figure 3.3-2 . 3.3-1 SGS-UFSAR Revision 6 February 15, 1987

A bursting pressure of 3 psig was also applied to the containment structure as part of the ultimate loading combination. The tornado loading analysis indicates that the peak membrane shear (which is the only significant stress component) induced at the base is 61.5 K/ft in comparison with 152.2 K/ft and 88.7 K/ft for the design basis earthquake and operating basis earthquake, respectively. It is further concluded from the investigation by Conrad Associates that a maximum wind velocity of 475 mph could be applied to the containment before any of the reinforcing steel reaches its yield strength. Forces, moments, and shears in the containment structure due to tornado wind load are shown on Figures 3.3-3, 3.3-4, and 3.3-5. 3.3.2.*3 Interaction of Category I and Non-Category I Structures Non-Category I structures adjacent to Category I structures are heavily braced to withstand tornado wind forces such that they will not collapse on the Category I structures. Metal siding on the Turbine Building is designed to be blown out to relieve tornado-generated differential pressure. SGS-UFSAR 3.3-2 Revision 6 February 15, 1987

60 MPH CASE* 1 CASE II CASE m
PUBUC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Pressure Distributions REVISION6 FEBRUARY15,1987 Updated FSAR FIG. 3.3-1

Cp 2.0 1.5

                                 = 0.42 + 1.1 cos 29 <o ~ e !! tr;'2  >

Cp : -0.68 <""'2 t e f'lf l 180

X PRESSURE DISTRIBUTIONAROUND CIRCULAR CYLINDERS Valu.. art In unitsof paf PRESSURE DISTRIBUTIONAPPLIED IN TORNADO ANALYSISOF CONTAINMENTVESSELREVISIONS FEBRUARY15,1987 WindVelocity Distributions PUBliCSERVICEELECTRICAND GAS COMPANY for TomadoConditions SALEMNUCLEARGENERATING STATION UpdatedFSAR FIG. 3.3-2

MERIDIONAL FORC (KIP/FT) Meridional Force~-- 0 50 100 I I I I I I I K/FT ELE*V. 76' REVISION8

                ....,.__..__R_A_D_I_A__L_S_TR_E_S_S____;.,(K_S.;...F...:)-+-V.:...;E:..:R.:..T.:...;I..:.C-.A:=L_:::_S~~R_£_S_S_(K~      FEBRUARY15, 1987 PUBLICSERVICEELECTRICAND GAS COMPANY                                   Forces in Containment               VesseldueotoTornadoWind loads SALEMNUCLEARGENERATING               STATION UpdatedFSAR                                                             Figure 3.3-3

MOM C:rcumfortfttlo Moment

                                                           +/- 21.0 0          50       100 I I I I I  I         I KP'T/FT ELEV. 76' TANGENTIALSTRES                    REVISION6 FEBRUARY15,1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Forces in Containment Vessel due to Tornado Wind loads SALEM NUCLEAR GENERATING STATION Updated FSAR                                      figure 3.3-4

0 50 100 I I I I I I I K/Ft ELEV. 7e' REVISION8 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Forces in Containment Vessel due to Tornado Wind Loads SALEM NUCLEAR GENERATING STATION Updated FSAR Figure3.3-5

3.4 WATER LEVEL (FLOOD) DESIGN ~ 3.4.1 Flood Elevations Figure 3.4-1 displays the relationship between the Mean Sea Level (MSL) datum, the Public Service Datum (PSD), the plant grade level and water levels for various conditions. The Probable Maximum Hurricane (PMH) surge level for the site is 113.8 feet PSD as estimated by the Coastal Engineering Research Center. The highest recorded water level at the site was +8.5 feet MSL in November 1950. This elevation is referred to as high-high water (HHW). The site grade was established by fill at an elevation of +10.5 feet MSL, or 2 feet above HHW. 3.4.2 Structural Loadings Load combinations and calculations for Category I structures are described in Section 3.8. 3.4.3 Flood Protection 3.4.3.1 Hurricane Safety-related equipment required for cold shutdown are located inside the containment, service water intake, Auxiliary Building, and main steam and feedwater pipe penetration areas. The containment is watertight and can withstand the static and dynamic loads associated with a storm producing stillwater level of 113.8 feet PSD and the corresponding wave runup to 120.4 feet PSD. The portion of the service water intake enclosing the pumps, motors, and vital switchgear is watertight up to Elevation 126.0 feet PSD with wave runup protection to Elevation 128.0 feet PSD. The service water intake can also withstand the static and dynamic effects of the storm. Each vertical, turbine type service water pump column bowl and suction bell is installed in an individual chamber which is open to the river. The chamber

SGS-UFSAR 3.4-1 Revision 15 June 12, 1996

is isolated from the watertight compartments where the pump discharge heads and motors are located. The pump discharge heads are bolted down to pads to Elevation 92 feet-6 inches PSD. The joint between the pump discharge head and the pad at Elevation 92 feet-6 inches PSD is watertight to prevent leakage of water into the compartments. Provisions have also been made to prevent leakage from the discharge head glands and leakoff connections into the watertight compartments. A sump pump is provided in each compartment to remove any accumulated water in the event a minor leak should occur. The Auxiliary Building is watertight up to Elevation 115 feet PSD. All doors in the outer Auxiliary Building walls below Elevation 120.4 feet PSD are watertight. All watertight doors and structural walls can withstand the static and dynamic effects associated with a storm that produces a stillwater level of Elevation 113.8 feet PSD with wave runup to Elevation 120.4 feet PSD. Conduit penetrations above Elevation 115 feet PSD and below Elevation 120.4 feet PSD will be packed to eliminate gross inleakage during the storm. Each residual heat removal pump room, the lowest point in the Auxiliary Building, contains two sump pumps, each adequate to provide the minimum capacity of 50 gpm. The main steam and feedwater pipe penetration area will be watertight below Elevation 120.4 feet PSD. The structural walls and watertight doors will also be capable of withstanding the static and dynamic effects of the storm which produces a stillwater level of Elevation 113. 8 feet PSD and wave runup to Elevation 120.4 feet PSD. 3.4-2 SGS-UFSAR Revision 27 November 25, 2013

Security-Related Information - Witheld Under 10 CFR 2.390 Electrical equipment not housed in structures designed to provide flood protection is located at a minimum elevation of 3.0 feet above HHW (or Elevation 100.5 feet PSD). The Turbine Building and Auxiliary Building floors are constructed at 2.5 feet above HHW (or Elevation 100 feet PSD). Turbine Building sump pumps and available portable pumps can be used to dispose of accumulated water resulting from leakage into the Turbine Building. The existing earthen dike was replaced by a protective rockfill dike along the portion of the Delaware estuary subjected to maximum wind wave forces, to protect the safety related structures and equipment. The following describes the Public Service implementation of the recommendations of the Dames and Moore reports (References 1, 2, and 3) as well as modifications made necessary by an increase in the estimated PMH surge level from Elevation 110.9 feet PSD to Elevation 113.8 feet PSD. The protective dikes were provided South of the power block between the Salem barge slip and the Salem circulating water intake structure, between the Salem circulating water and service water intake structures, and North of the Salem service water intake structure.

A dike stability analysis was also performed by Dames and Moore, using a computer program based on the Fellinius method of slices. 3.4-3 SGS-UFSAR Revision 23 October 17, 2007

Based upon Dames and Moore's recommendation, all recent alluvial deposits and dredge soil were removed and replaced with granular fill. The fill consisted of a rock toe dike and a general fill consisting of clean granular soils. In the vicinity of the barge slip, unsuitable soils were excavated directly beneath the proposed dike down to the old river bottom on slopes as steep as possible and replaced with clean soils. The barge slip provides adequate toe protection for stability and against erosion. The configuration of the toe dike is based on a conservative value of angle of internal friction of the sand fill. If a better grade material is available, the toe dike could be reduced, or possibly eliminated. In the area of the existing fuel oil tank, the excavation slope was steepened sufficiently to provide a minimum distance of 10 feet between the pile cap and the top of the slope. The excavation was backfilled as soon as possible to minimize local slope failures. Tests and analyses of borings in the dike area confirmed the validity of these design features. The shoreline protection and dike system will be inspected by station operating personnel prior to storms and hurricanes and following the passage of such storms and hurricanes. Additionally, a more complete annual inspection will be conducted both by boat and from the dike itself. The station security forces also make regular patrols of these areas as part of their surveillance duties, and are instructed to report any abnormalities observed in the structure.

3. 4-4 SGS-UFSAR Revision 23 October 17, 2007

In the event of rising water levels, all watertight doors will be closed to maintain watertight integrity. The decision to shut down the plant in the event of rising water levels will be made by the Station Manager. Accordingly, the Technical Specifications specify the flood levels at which (1) watertight integrity will be established (at which time flood protection procedures will be initiated on a site-wide basis to protect the plant from flood waters) and (2) plant shutdown will be initiated. 3.4.3.2 Precipitation The Yard Drainage system is designed to pass the drainage associated with a rainfall rate of 4 inches per hour for a period of 20 minutes (baaed on 90 percent runoff from paved areas and 50 percent runoff from graded areas). Interior drains in the Auxiliary and Fuel Handling Buildings are independently piped to the Liquid Waste Disposal System and are not connected to the Yard Drainage System. Roof drains are designed to dispose of a maximum rainfall rate of 4 inches per hour for a period of 20 minutes through the Yard Drainage System. Roof slabs are watertight to prevent buildinq interiors from being damaged by severe rainstorms. The slabs are designed to withstand a loading equivalent to a depth of water up to the full height of the building' s parapet or roof curb. In the unlikely event that some of the roof drains become plugqed up, the backed up water will spill down the outside of the building. Wall penetrations above Elevation 115 feet PSD on Class I (seismic) buildings are designed to prevent roof spillage or heavy rain from seeping inside the building. In the event the capacity of the Yard Drainage system were to be exceeded as a result of an unusually severe rainstorm, the excess water would accumulate in puddles in the vicinity of the catch basins (see Figure 3.4-4) and run off. This water would not enter 3.4-5 SGS-UFSAR Revision 16 January 31, 1998

any safety-related structure, since these structures are watertight up to at least Elevation 115 feet PSD. Therefore, safety-related equipment would not be adversely affected as a result of a severe rainstorm. 3.4.4 Protection From Hurricane Drawdown As discussed in Section 2. 4, the lowest instantaneous water elevation at the service water pumps is 11.1 feet MSL. This could occur during the passage of a PMH oriented to produce drawdown at the site. In order to ensure operation of the service water pumps during this the minimum submergence level is specified to be -13.0 feet MSL. 3.4.5 Reference for Section 3.4

1. Shoreline Investigation and Oceanographic Study, Proposed Nuclear Generating Station, Salem, New Dames and Moore, November 20, 1970.

2. Stability of Protective Dikes, Salem Nuclear Generating Station, Salem, New Jersey, Dames and Moore, April 2, 1973.

3. Final Report, Stability of Protective Dikes, Salem Nuclear Station, New Jersey, October 8 1 1973.

4. 0. S. Army Coastal Engineering Research Center, "Shore Protection Planning and Design,n Technical Report No. 4, 3~ Edition, 1966.

3.4-6 SGS-UFSAR Revision 23 October 17, 2007

MSL PSD* FLOOD PROTECTION LEVEL FOR 31.4 120.4

SAFEGUARDS S't!TEMS I

22.9' PUBLIC SEFWICE E LEVATION IOCY 11.0 100.0 <TURBINE AND AUXILIARY BUILDING FLOOR1 PRESENT GRADE 2.5' 2.0' ORIGINAL GRADE o.'s* 8.5 97.5 HIGH HIGH WATEB (NOVEMBER. 1950J 8.5'

MEAN HIGH WATER
3.2' MEAN SEA LEVEL 0 89.0 (SAND'( HOOK 1929 ADJUSTMENT)

U.S. COAST AND GEODETIC SURVEY DATUM 2.6' 2.9'

  -2.6    86.

MEAN LOW WATER U . S . ARM'( CORPS OF ENGINEERS DATUM 8.0' s.g- 13.o*

  -5.9    83A

LOW LOW WATER -RECORDED (JANUARY 25. 1939)

 - 8.0    81.0                                        LOW LOW WATER- PROJECTED COECEMBER 31, 1962J
 -13.0 76.0                                           LOW LOW WATER- DESIGN.

.PUBLIC SERVICE DATUM REVISION 8 FEBRUARY 15.1987 Datum and Water Level Relationships PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR FIG. 3.4-1

                     \.._.\

SECURITY -RELATED INFORMATION- WITHHELD UNDER 10 CFR 2.390 REVISION 23, OCTOBER 17, 2007 Salem Nuclear Generating Station PSEG Nuclear, LLC PROTECTIVE DIKE CROSS SECTION (EAST OF CIRCULATING WATER INTAKE STRUCTURE) SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.4-2 CD 2000 PSEG Nuclear, LLC. All Rights Reserved.

SECURITY -RELATED INFORMATION-WITHHELD UNDER 10 CFR 2.390 I REVISION 23, OCTOBER 17, 2007 Salem Nuclear Generating Station PSEG Nuclear, LLC PROTECTIVE DIKE CROSS SECTION (BETWEEN SERVICE & CIRCULATING WATER INTAKE STRUCTURE) SALEM NUCLEAR GENERATING STATION Updated FSAR Fiqure 3.4-3 CD 2000 PSEG Nuclear, LLC. All Riohts Reserved.

FIGURE 3.4-4 intentionally deleted. Refer to Figure 2.4-2 . REVISION 8 FEBRUARY 15, 1987

3.5 MISSILE PROTECTION structures, shields, and barriers are provided as protection against the effects of both internally and externally generated missiles. For turbine generated missiles, additional factors such as safe inspection and re-inspection schedules for low pressure turbine discs are included in the risk analysis. The discussion on turbine missiles is given in Section 3.5.4. The following section covers internally generated missiles. 3.5.1 Internally Generated Missiles A loss-of-coolant accident (LOCA) or other plant equipment failure might result in internally generated missiles. For such engineered safety features as are required to assure safety in the event of such an accident or equipment failure, protection from these missiles is considered in the layout of plant equipment and missile barriers. 3.5.1.1 Missile Types The types of missiles for which protection is provided include:

1. Valve stems

2. Valves

3. Instrument thimbles

4. Various types and sizes of nuts and bolts

5. Complete control rod drive mechanism, or parts thereof, except the MG set flywheels, which will not produce missiles under any anticipated accident condition.

3.5.1.2 Missile Protection Methods High pressure Reactor Coolant System equipment which could be the source of missiles is suitably shielded either by the concrete 3.5-1 SGS-UFSAR Revision 14 December 29, 1995

shield wall enclosing the reactor coolant and pressurizer loops or by the concrete operating floor to block any passage of missiles to the containment walls even though such postulated missiles are deemed most improbable. A structure is provided over the control rod drives to block any missiles generated from a fracture of equipment. Missile protection for the plant is provided to comply with the following criteria:

1. To protect the containment and lines from loss-of-function due to damage by such missiles as might be generated in a LOCA for break sizes up to and including the double-ended severance of a reactor coolant pipe.

2. To protect the engineered safety features 1 systems and components required to maintain containment integrity against loss-of-function due to damage by the missiles defined in Section 3.5.1.1.

The following considerations are included in the design for missile protection to meet the above criteria:

1. The Reactor Coolant System is supported by steel structures designed to withstand the forces associated with a double-ended rupture of a reactor coolant pipe and shielded by concrete walls designed to stop the missiles.

2. The structural design of the missile shielding takes into account both static and impact loads.

3. Components of the Reactor Coolant System are examined to identify and to classify missiles according to size, shape, and kinetic energy for purposes of analyzing their effects.

The Petry formula as described in Section 3. 5. 3 is used to check the missile penetration. The energy approach is used to determine the equivalent static load from the missile impact. The original missile shields located above the control rod drive mechanisms have been removed due to installation of the integrated head assembly (IHAl . The IHA utilizes an integrated steel missile shield that provides the necessary blockage of any missile that could be generated by the control rod drive mechanisms. To ensure the IHA missile shield is capable of withstanding a missile impact and adequately performing its design function, a penetration evaluation, as well as a finite element strain evaluation, were performed. 3.5-2 SGS-UFSAR Revision 22 May 5, 2006

The penetration evaluation uses the USNRC Standard Review Plan Section 3.5.3, Barrier Design Procedures, for guidance in determining the minimum required thickness of the IHA missile shield to ensure it can absorb the impact energy without perforation. A non-linear transient finite-element analysis evaluated the missile shield for the overall effects expected from the impact of a CRDM missile. 3.5.2 Tornado Generated Missiles Salem 1 and 2 have been evaluated against General Design Criteria 2 and 4 of February 1971. All Category 1 equipment is protected from the effects of tornado generated missiles such that the equipments safety function can be fully deployed. Category I structures including Reactor Containments, Auxiliary Building, Service Water Intake Structures, and Fuel Handling Buildings are designed to withstand tornado missiles. These structures are also designed such that under the impact of the most damaging tornado missile, they will not create a secondary missile of enough mass or velocity to penetrate any adjacent Category I structure. The missile utilized as the critical object for penetration analysis of all surfaces of the designated structures, of both units, is a wooden utility pole with dimensions 40 feet long, 12 inches in diameter, weighing 50 pounds per cubic foot, traveling 150 mph. The missile is assumed to impact the surface at normal incidence, such that its kinetic energy at the point of impact is maximized. A design basis tornado is assumed to cause a loss of offsite power; however, the assumption of an additional single failure is not postulated, per Regulatory Guide 1.117 Rev. 0 (See Appendix 3A). 3.5.2.1 Critical Missiles Selected for Evaluation The missiles below are those that are assumed to be generated by a design basis tornado. Salem has certified that the exterior walls of the safety-related plant structures, which have a minimum thickness of 18 inches, have been evaluated against the following two missiles that are identified as Missiles C and F of the Standard Review Plan 3.5.1.4, Rev. 0:

1. Steel rod, one inch diameter, three feet long, weight eight pounds, traveling horizontally at 316 feet per second and vertically at 252 feet per second, at all elevations, and

2. Utility pole, 13-1/2 inches diameter 35 feet long, weight 1490 pounds, traveling horizontally at 211 feet per second and vertically at 169 feet per second, at all elevations less than 30 feet above grade within 1/2 mile of the facility structures.

The velocities delineated above are the fraction of total tornado velocity designated in the SRP where the total velocity, 360 mph, is taken from Regulatory Guide 1.76, Rev. 0. Vertical velocities are 80% of horizontal velocities in accordance with SRP 3.5.1.4 Rev. 0. The maximum postulated velocity for a tornado generated missile, and the velocity of normal incidence assumed for all safety related structures, is the horizontal velocity. A passenger car missile is also used in the evaluation. The elevation, below which the generation of a utility pole missile by a tornado is assumed, is computed to be 140.7 ft. Public Service Datum (Ref. 3.5.5.12). By virtue of similarity of construction, this section is applicable to both Salem 1 and Salem 2. 3.5-3 SGS-UFSAR Revision 32 June 17, 2021

Security-Related Information - Witheld Under 10 CFR 2.390 3.5.2.2 Missile Protection Methods

Penetrations into steel barriers are calculated using equations developed based on tests conducted at the Stanford Research Institute, summarized in Reference 7. The equation for penetration can be written in the form e 3 B e 0.0452D 02 ( )2 + ( )( ) = d 128 d d Ss 3.5-4 SGS-UFSAR Revision 32 June 17, 2021

where e = perforation thickness (in) d = effective projectile diameter (in) B = width of plate between rigid supports (in) D = w/d3 where w = missile weight in pounds 0 = missile striking velocity (ft/sec) Ss = ultimate tensile strength of steel target (psi) The value of e calculated using the above equation is multiplied by 1.25 when designing steel barriers td = 1.25 e Steel missile barriers are designed using the following allowable ductility ratios: Flexural members d < 10.0 Columns with slenderness ratio (l/r) less than or equal d < 1.3 to 20 Columns with slenderness ratio (l/r) greater than 20 d < 1.0 Members subjected to tension d < 0.5 eu ey where eu = ultimate strain ey yield strain 3.5.2.3 Safety Assurance Against Tornado Missile Induced Damages In addition to the Category I systems discussed below, elements of certain Category I components are not designed or constructed to be totally within the envelope of a Category I structure; for example, the Turbine Driven Auxiliary Feed Pump exhaust stacks, Main Steam Isolation Valve pilot valve exhaust stacks, Emergency Diesel Generator exhaust stacks and ventilation housings, etc. These components are evaluated utilizing the Tornado Missile Risk Evaluator (TMRE) method, Reference 3.5.13, an NRC approved Method of Evaluation (Reference 3.5.14), and have an acceptable level of risk, such that they do not require further deterministic protection beyond their original construction. TMRE does not provide a basis for modifications to remove existing tornado missile protection or to omit protection for new configurations that otherwise require tornado missile protection according to Salems licensing basis. Use of the methodology is subject to the restrictions and limitations contained in the NRC Safety Evaluation (SE) for Grand Gulf Nuclear Station (Reference 3.5.15). Category I systems outside the Category I buildings that may be damaged by tornado or secondary missiles are the refueling water storage and auxiliary feedwater storage tanks and systems. Sufficient precautions have been taken to assure a safe shutdown of the reactor and to maintain it in a safe condition. 3.5-5 SGS-UFSAR Revision 32 June 17, 2021

3.5.2.3.1 Backup Water Sources for Category I Water Storage Tanks In the event that the integrity of Category I storage tanks are violated due to tornado induced missiles, sufficient backup water sources are available to assure a safe shutdown of the reactor. Primary system makeup requirements during cooldown are normally accomplished by the Chemical and Volume Control System (CVCS}, providing makeup water from the primary water storage tank. In the event that the primary water storage tank is the affected tank, primary system makeup can be provided by transferring water to the Reactor Coolant System from the three 8500 cubic foot capacity holdup tanks or the spent fuel pool via the eves. Approximately 24 I 000 gallons of water are required to maintain the level in the Reactor Coolant System in the transition from full power operation to cold shutdown. Removable borated water (31,000 gallons) is available in the spent fuel pool above the spent fuel pool pump suction line. In addition, another 13,000 gallons of borated water is available above the spent fuel pool pump suction line in the fuel transfer pool. Feedwater system makeup requirements during cooldown are normally supplied from the Main Feedwater System, providing makeup water from the condenser. In the event the Main Feedwater System becomes inoperable, the Auxiliary Feedwater system is placed in operation and feedwater makeup is supplied from the auxiliary feed storage tank. This Category I tank is adequately protected from the effects of earthquakes, tornado, wind loads, and floods. Backup water sources for the auxiliary feedwater pumps are the two demineralized water storage tanks (500,000 gallon capacity each), the two fire protection and domestic water storage tanks (350,000 gallon capacity each), and the station Service Water System. Because of the chloride content of the water (approximately 6000 ppm), the Service Water System would be used for this purpose only as a last resort in the event that other water sources become unavailable. The Service Water System will be used for decay heat removal as necessary, only until the normal or backup water sources are again made available. Backup water supply from the demineralized water storage tanks is readily available, upon filling and venting the associated suction piping. The fill process is accomplished using local/manual valves. Automatic vent valves are provided to ensure the line is properly filled. Once filled the alternate suction stops, AFS2's are opened from the Main Control Room to place this backup source of water in service. The fire protection and domestic water storage tanks can be used as a backup water source in the event that the demineralized water storage tanks become unavailable. Fire protection and domestic water can be provided to the auxiliary feed pumps only when a spool piece has been connected by station personnel. The Service Water System can be used as a backup water I SGS-UFSAR 3.5-6 Revision 17 October 16, 1998

source in the event that the other water sources become unavailable. Service Water can be provided to the Auxiliary Feed Pumps with a spectacle flange installed between the auxiliary feedwater and service water piping. Service water can be provided to the auxiliary feed pumps by rotating the spectacle flange into the full bore position. The time required to connect the two systems is less than 30 minutes. Operations involving alignment of normal and alternate sources of water to the auxiliary feedwater pumps are covered by station procedure. The backup water sources from the CVCS holdup tanks, the spent fuel pool, and the Service Water System are located in buildings/structures designed to withstand tornado induced missiles. The water storage tanks are located in different areas of the station site (refer to the arrangement drawings in Section 5) Although not specifically designed to withstand tornado induced missiles, the separated locations of the various tanks preclude the possibility that all tanks would be rendered unavailable due to tornado induced missiles. The systems which are required to bring the unit to a safe shutdown are enclosed in tornado protected buildings/structures and capable of being powered by the standby AC power systems. Plant Shutdown Operating Procedures In the unlikely event that a breach of any of the primary water sources were to occur (refueling water storage tank, auxiliary feedwater storage tank, or primary water storage tank), station operating personnel would start to shut down and cool the reactor below 350°F, using normal plant shutdown operating procedures. These procedures initiate immediate decay heat removal by steam dump to the condenser. As feedwater in the secondary plant cycle absorbs heat from the reactor coolant through the steam generators, it is converted to steam, giving up heat to the circulating water in the condenser. Hot shutdown can be maintained with Auxiliary Feedwater System operation and main steam atmospheric relief valves. Boration can be accomplished via the charging pumps, boric acid tanks, and boric acid trans fer pumps. In the event the Auxiliary Feed Storage Tank is unavailable, transfer to one of the backup water sources (previously discussed in this section) is accomplished by station procedure. No automatic water supply switchover is provided. The backup water sources for the Auxiliary Feedwater System water sources are addressed in the Technical Specifications (except for the service water backup source) . The unit would be maintained in a hot shutdown condition until station operating personnel have assessed operating conditions, damage, and availability of water sources to attain cold shutdown. Necessary repairs would be made and alternate water supplies provided as necessary for the CVCS to maintain Reactor Coolant System inventory and proceed to bring the unit to a cold shutdown condition. 3.5-7 SGS-UFSAR Revision 28 May 22, 2015

Shutdown procedures would be accomplished in as normal a fashion as operating conditions permit utilizing alternate water sources as necessary. The methodology would be dependent upon the extent of tornado missile damage as determined by operating personnel in their assessment of plant conditions. The arrangement of the service intake structure, service water pumps and piping is discussed in Section 9. 2. The service water pumps are located within the service intake structure, which is seismic Category I and designed to be tornado missile proof. The piping is buried below grade, thereby protected from tornado missiles. Indication is provided in the control room for all outdoor water storage tanks. Upon indication of loss of water from all outdoor water storage tanks such that they become unavailable for auxiliary feedwater supply (tornado induced failure), personnel would be dispatched to install the connection between the Service Water and Auxiliary Feedwater Systems. It has been determined by actual demonstration that two people can rotate the spectacle flange in 15 minutes. The installation requires no special tools. The spectacle flange bolts will be removed, the spectacle flange will be rotated, and then the bolts will be reinstalled to provide the required flow. An analysis was performed to determine the time period following a loss of ac power and main and auxiliary feedwater flow before the core becomes uncovered. The pertinent assumptions used in the analysis are as follows:

1. All ac power lost at time of incident

2. Rods assumed to begin dropping into core 2 seconds following incident

3. ANS standard decay heat curve assumed

4. Pressurizer relief and safety valves operative

5. Initial power is 1.006 times rated power

6. Loss of all main and auxiliary feedwater following incident.

The calculation was subdivided into the following areas:

1. Heat required to raise primary to saturation temperature 3.5-8 SGS-UFSAR Revision 28 May 22, 2015

2. Heat required to uncover core after primary saturation temperature is reached

3. Secondary heat sink available Primary Saturation The amount of heat to *raise the Unit 1 primary inventory from the initial average temperature to the saturation to the pressurizer power operated relief valve setpoint was calculated to be approximately 16 full power seconds. The primary side inventory increases by 3.4% as a result of steam generator replacement thus; this value increases to approximately 16.5 full power seconds for Unit 2.

The quantity of heat required to boil sufficient Unit 1 primary inventory in order to begin uncovering the core was found to be approximately 5.8 full power seconds. The primary side inventory increases by 3. 4% as a result of steam generator. replacement thus; this value increases to approximately 6. 0 full power seconds for Unit 2. Based upon nominal initial mass, the heat to the Unit 1 secondary inventory via the steam generator safety valves was calculated to be equivalent to 7 4. 8 full power seconds. The secondary side inventory decreases by 6.3% as a result of steam generator replacement thus; this value is reduced to approximately 70.1 full power seconds for Unit 2. The total heat generation, which results in uncovering the core, is then the sum of the three heat sinks itemized above. These components yield a total of 96.6 full power seconds for Unit 1 and 92.6 full power seconds for Unit 2. Assuming a standard ANS decay heat curve, the period of time following initiation of the incident was calculated to be approximately 4200 seconds or 70 minutes for Unit 1 and approximately 65 minutes for Unit 2. With this analysis taken into consideration, a conservative estimate of time history to initiate feedwater following a loss of outdoor water storage tanks follows: 3.5-9 SGS-UFSAR Revision 24 May 11, 2009

Time (Min.) Event 0 Tornado induces failure in all outdoor water storage tanks. (Low water indication in control room) 30 Acknowledgement by control room, dispatch personnel to rotate spectacle flange. 40 Spectacle flange rotation commences 55 Service water connection complete; auxiliary feedwater initiated. This discussion demonstrates that sufficient water sources are available to bring the unit to a safe shutdown condition in the event of a breach of a primary water source due to a tornado induced missile. 3.5.2.3.2 Safety Evaluation of Loss of Suction to the Auxiliary Feedwater Pumps The design of the Auxiliary Feedwater (AFW) System was evaluated with regard to tornado-induced loss of suction to the AFW pumps. The Auxiliary Feedwater Storage Tank (AFST) and AFW pump suction piping have been designed to withstand the design basis tornado wind loadings as described in Section 3.3. The following discussion with respect to the evaluation therefore considers only tornado missiles. In the evaluation it was conservatively assumed that loss of pump suction pressure would result in instantaneous damage to the AFW pumps. The design basis tornado generated missile is a wooden utility pole described in Section 3. 5. 2. 1. The probability of a missile impact on the AFST outlet piping has been determined to be less 3.5-10 SGS-UFSAR Revision 28 May 22, 2015

       -7 than 10    per year and thus no further   evaluat~on  has been performed  ~n th~s regard.

W~th regard to the APST, it has been determined that the design basis utility pole striking the tank at its base would conservatively result in a 490 square inch opening. Such an opening would result in the AFST draining in approximately 5 minutes. Continuous AFST level indication is provided in the control room, as well as low level (100,000 gallon) and low low-level (30,000 gallon) audible and visual alarms. Additional audible and visual alarms are being provided to detect deviation from the technical specification minimum volume (200,000 gallons). The above indications and alarms prov~de prompt indicat~on of tornado-~nduced damage to the AFST. The potential effects of such a loss of water to the damaged AFST on various modes of plant operation were also evaluated. In the event of a tornado forecast, the shift crew would operate the unit with an increased sensitivity toward the potential effects of a tornado, and thus, it was concluded that sufficient time is available to recognize and assess the damage, trip the AFW pumps from the control room, and change over to an alternate suction path without damage to the AFW pumps. In order to further enhance the overall design of the AFW System, a safety grade automatic low suction pressure trip for each AFW pump has been incorporated. To preclude inadvertent actuation, this modification is designed such that it can be made operable only during periods when a ~tornado warning~ has been put in effect by the National Weather Service. Also, in accordance with I&E Bulletin 80-11, a wire mesh net protects the Unit 1 RWST and AFW tank in the event of adjacent wall collapse due to seismic loading.

3. 5-11 SGS-UFSAR Revision 7 July 22, 1987

3.5.3 Modified Petry Formula The penetration of the missile into concrete has been calculated by the Modified Petry formula: D 1 + A 215,000 Where: D penetration in feet

                                                -3    3 K     coefficient equal to 4.76 x 10           ft /lb for reinforced concrete with a crushing strength of 3200 psi, and
                     -3     3 2.82 x 10     ft /lb for reinforced concrete with a crushing strength of 5700 psi.

W missile weight in pounds V striking velocity in feet per second A missile frontal area in square feet 3.5.4 Turbine Missile 3.5.4.1 Turbine Placement and Characteristics Plan and elevation views of the turbine building are provided on Plant Drawings 204811, 232445 and 232444. As described in Section 10, each unit consists of one tandem compound, six flow, four casing, condensing, 1800 RPM turbine. The Unit 1 low pressure 2 element is designated as a Siemens Westinghouse 13.9 m retrofit turbine. The Unit 2 low pressure element is designated as a Westinghouse building block 81R. 3.5-12 SGS-UFSAR Revision 27 November 25, 2013

3.5.4.1.1 High Pressure (HP) Turbine

             'i'he HP turbine element,     as shown on Figure 3. 5-4,      is of a double flow design and is inherently thrust balanced.           Steam from the four control valves enters at the denter. of the turbine element through four inlet pipes, two in the base and two in the cover. These feed an admission ring in the inner casing connected to the turbine casing.      Steam  leaves   the   admission  ring  and    flows    through   the reaction blading.        The  reaction blading is mounted in           the  inner casing and the guide blade carriers shown on                  3.5-5 which are mounted in the turbine casing.

The HP rotor is made of NiCrMoV alloy steel. The fled minimum mechanical properties are as follows: Tensile Strength, N/nw2 (psi) :S:820 (.S:l1B931) 2 Yield Strength, N/rnm (psi) 1 580-680(84122-98626) (0.2 percent offset) Elongation (1 =5d), percent ~16 0 Reduction of Area, percent ~50 Impact Strength, Charpy V-Notch, J 100 (min at room temperature) 50 Fracture Appearance Transition :S:-22 Temperature, °F, max The main body of the rotor weighs approximately 110,000 lb. The approximate values of the transverse centerline diameter, the maximum diameter, and the main body length are 4 5 inches, 64 inches, and 138 inches, respectively. The casing cover and base are made of carbon steel castings. The specified minimum mechanical properties are as follows: Tensile Strength, psi, min 70,000 3.5-13 SGS-UFSAR Revision 21 December 6, 2004

Yield Strength. psi, min 36,000 Elongation in 2 inches, percent, min 22 Reduction of Area, percent, min 35 The ibend test specimen is capable of being bent cold through an angle of 90 degrees and around a pin 1 inch in diameter without cracking on. the outside of the bent portion. The blade and the inner casings are made of alloyed steel castings. The specified minimum mechanical properties are as follows: Tensile Strength, N/mm2 (psi} 540-690 (78320~100076) Yield Strength, N/mm2 (psi), ~355 (~51488) (0.2 percent offset) Elongation (1 =5d), percent ~18 0 Reduction of Area, ~45 Impact Strength, Charpy V-Notch, J ~40 The approximate weights of the inner casing, two blade the casing cover, and the casing base are 46,000 lb, 26,500 lb, 115,000 lb, and 115,000 lb, respectively. The casing cover and base are tied together by means of more than 100 studs . The stud material is an alloy steel having the following mechanical properties: 2-1/2 Inches Over /2 Over 4 and Less To 4 Inches Tensile Strength, psi, 125,000 115,000 110,000 min. Yield Strength, psi, min. (0.2 percent 105,000 95,000 85,000 offset) Elongation in 2 inches, percent, min. 16 16 16

Reduction of Area, percent min. 50 so 45 3.5-14 SGS-UFSAR Revision 21 December 6, 2004

The studs have from 17 to 66 inches and diameters ranging from 2.75 inches to 4.5 inches. About 90 percent of them have diameters ranging between 2.5 and 4 inches. The total stud cross-sectional area is 2 approximately 900 in and the total stud free-length volume is 3 approximately 36,000 in 3.5.4.1.2 Low Pressure (LP) Turbine

1. Unit 1 The double flow LP turbine, shown on Figure 3. 5-6, high efficiency blading, diffuser type exhaust, and incorporates liberal exhaust hood design. The last stage rotating blade length is 4 6". The LP turbine cylinder is fabricated from steel plate to provide uniform wall thickness thus reducing thermal distortion to a minimum. The entire outer casing is subjected to low temperature exhaust steam.

The LP turbines include an inner casing with the installed stationary blading and external heat shields. The heat shield mitigates heat losses by radiation from the inner to the condenser and also reduces thermal stressing of the inner casing components. Figure 3.5-7 shows the welded inner casing with its blade carriers and blade rings. The upgrade also features compartments for the inlet steam and steam extractions, which are designed to safely handle maximum pressure differences at steady state and transient operating conditions. The upgraded inner casing features optimal moisture removal at the circumferential outer diameter of the flow path. In addition, the last stage hollow stationary blades feature moisture removal, slots to take moisture from the blade surface and discharge it through the slots into the hollow blade and from there into the condenser. The fabricated inner is supported by the outer casing at the horizontal centerline and is fixed transversely at the top and bottom and axially at the centerline of the steam inlets, thus allowing freedom of expansion independent of the outer 3.5-15 SGS-UFSAR Revision 21 December 6, 2004

The outer cylinder and the inner cylinder are fabricated mainly of ASTM 515-GR65 material. The minimum specified properties are as follows . Tensile Strength, min 65,000 Yield Strength. psi, min 35,000 Elongation in 8 inch percent min 19 Elongation in 2 inch percent min 23 The LP rotors and discs are made of NiCrMoV alloy steel. The specified minimum mechanical properties are as follows: Tensile Strength, N/mm 2 (psi) ::::;820 ($118931) 2 Yield Strength, N/mm (psi), 580-680{84122-98626) (0.2 percent offset) Elongation (1 =5d), percent ;;;:16 0 Reduction of Area, percent 250 Charpy V-Notch, J 100 (min at room temperature) 50 Percent Fracture Appearance Transition S-22 Temperature, °F 1 max There are discs shrunk per shaft with three per flow. 3.5-16 SGS-PFSAR Revision 21 December 6, 2004

2. Unit 2 The Unit 2 LP turbine is a double flow design consisting of three elements. The original LP turbine rotors were of the built-up type consisting of individual discs shrunk on and keyed to a shaft. To optimize thermal performance and provide long term reliability, the turbine has been retrofitted with a monoblock rotor forging with fully integral discs and couplings. This design incorporates blading with integral shrouds. The last stage rotating blade length is 47", resulting in a large annuluf area.

Figure 3.5-6A depicts the LP turbine. Steam will enter the LP turbine inlet at the cross-over pipe tee connection. Inlet flow guides will direct the steam in both directions of this double flow element. There are nine stages of controlled reaction blading. The first four stages on each end are housed in the first blade ring. The next two stages on each end are housed in the second blade ring. Both blade rings on each end are separately supported and made from stainless steel to minimize erosion/corrosion concerns. The last three stages on each end are of segmental assembly construction that are caulked into the inner cylinder. Advanced techniques were used to design the one-piece inner cylinder, resulting in minimized steam leakage and structural streamlining which minimizes blade path differential thermal expansion. The entire outer casing is subjected to low temperature exhaust steam. The outer and inner cylinders are fabricated mainly of ASTM 515-GR65 material. The minimum specified properties are as follows: Tensile Strength, psi, min. 65,000 Yield Strength, psi, min. 35,000 Elongation in 8 in., percent, min. 19 Elongation in 2 in., percent, min. 23 3.5-17 SGS-UFSAR Revision 16 January 31, 1998

The LP rotors are made of 3. 5 NiCrMoV alloy steel. The specified minimum mechanical properties are as follows: Tensile Strength, psi, min. 115,000 Yield Strength, psi, min. 100,000 Elongation in 2 in., percent, min.* 17 Reduction of Area, percent, min. 50 Impact Strength, Charpy V-Notch, 120 ft-lb, min. at room temp. 50% fracture Appearance 35 Transition Temp. °F, max. The first five stages of the rotors use a one-piece integral shroud blade. The first three rows are made of 12% Cr material. The next two are made of 17-4 PH stainless steel material. The last three rows of the rotors use shot peened profiled free-standing blades also made of 17-4 PH stainless steel. 3.5.4.1.3 Overspeed Protection System Testing The turbine overspeed protection system is tested by accelerating the turbine from. 1800 rpm until i t trips. Should the turbine fail to trip automatically, it is manually tripped prior to exceeding 12 percent overspeed. The turbine stop and gdvernor valves are tested by actually cycling the valves. For Salem

Unit-13 1 & 2, the Turbine Overspeed Protection requirements are discussed in 10.2.2.6, "Turbine Overspeed Protection".

3.5.4.2 of Missile Present manufacturing and inspection techniques for turbine rotor and disc forgings make the of an undetected flaw and subsequent catastrophic structural failure extremely remote. Forgings are subject to inspection and testing both at the forging suppliers and at Westinghouse. Current design procedures are well established and conservative, and analytical tools .such as finite element and fracture mechanics techniques allow in depth analysis of any potential trouble spots such as areas of stress concentration or inclusions which could give rise to crack propagation. SWPC has pr~pared missile probabilities methodology that has been approved by the NRC, in NRC SER titled, "Siemens Westinghouse Topical Report, le Analysis Methodology General Electric (GE) Nuclear Steam Turbine Rotors by Siemens Westinghouse Power Corporation (SWPC), Project No. 721" 3.5-18 Revision 21 December 6, 2004

(reference 10). Although the title of this document references GE turbine rotors, the conclusions provide the following statement: The approval of the Siemens methodology includes the use of specified values for the use of the PPDBURST and PDMISSILE computer programs, and the use of specified values for some key input and built-in parameters for those two programs for future plant-specific turbine missile probability analyses for GE and Siemens rotors. The results of SWPC missile analysis is documented in SWPC Technical Report, CT-27336, Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1 (reference 8). This report is in full compliance with the requirements of the SER and meets the specific analysis requirements established by the SER conclusion section. The conclusion of the report states that the missile probabilities are within the limits required by Regulatory Guide 1.115, Protection Against Low-Trajectory Turbine Missiles, Rev. 1, (1E-5 per year) for an unfavorably oriented low pressure turbine based on a six month turbine valve testing frequency for the high pressure turbine stop and governor valves. The hot reheat stop and hot reheat intercept valves will be tested on a twelve month interval with a 100,000 equivalent operating hours inspection frequency based on an Engineering Technical Evaluation documented under 70214643-0010. Siemens further clarifies that the inspection intervals will be limited to 87,600 equivalent operating hours inspection frequency based on current NRC approved inspection frequency. 3.5.4.2.1 High Pressure Turbine Risk Analysis The design and fabrication of the HP rotor is such that the vendor does not require periodic inspection of the rotor to address the risk of missile generation. Ductile burst is unlikely, since it would require a rotational speed beyond the terminal speed of typical units and failure from this mechanism need not be considered. Failures due to high cycle fatigue fracture have not occurred in the past and the retrofit rotors have improved design safety factors making this mechanism unlikely. Failures due to low cycle fatigue are unlikely since LCF life is significantly greater than 10,000 start cycles for the original and retrofit rotors. Based on these factors, there are no requirements for periodic in-service inspections to address the risk of missile generation. For more details, refer to Reference 9. 3.5.4.2.2 Low Pressure Turbine Risk Analysis

1. Unit 1 The Salem Unit 1 low pressure turbines will be inspected on at least a 10 year (87,600 equivalent operating hours) frequency to ensure that the probability of turbine generated missile remains within the requirements of Regulatory Guide 1.115. This inspection frequency requires that the high pressure turbine stop and governor valves be tested at a six month frequency and the hot reheat stop and hot reheat intercept valves be tested at a twelve month frequency.

3.5-19 SGS-UFSAR Revision 32 June 17, 2021

Basically, the approach used is to make a conservative prediction of how a presumed or actual crack will grow and then schedule an inspection to the time the crack grows large enough to be of concern. Analytical components of this approach are Load, Crack Branching Fracture Toughness, Yield Strength, SCC Growth Rate, and Initial Crack Size. SWPC Technical Report, CT-27336, "Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1" {reference 8) provides a discussion of these parameters and values utilized in the development of the Salem Unit 1 Low Pressure Turbine Generated Missile Probabilities.

a. Load SWPC determines load for each disc through the use of finite element analysis. They assume that the analysis provides accurate results within 5% of tolerance due to the uncertainties in geometry as well as thermal and mechanical loads. A normal distribution is assumed. The mean values for the disc are:

Disc #l - 498 MPa Disc #2 - 521 MPa Disc #3 - 535 MPa

b. Crack Branching Factor The branching factor k is assumed to be normally distributed with a mean of 0.65 and a standard deviation of 0.175, whereby j 0.65 {/' C'rack Depth s Jin k =l. 1 other1rise

c. Fracture Toughness The normal distribution has been used in describing scatter in fracture toughness data with a mean of 219.8 MPa~ and standard deviation of 10% of the mean value.

    , d. Yield Strength The yield strength values are assumed to be distributed normally with mean and standard deviation values on internal investigation data:

Disc #1 815 MPa and std. deviation ~ 30 MPa Discs #2 & #3 855 MPa and std. deviation = 30 MPa

    ; e. SCC Growth Rate The stress corrosion (SCC) rate is assumed to be independent of the stress intensity level.       The main parameters influencing the SCC rate are temperature, material yield                  and water chemistry.      The empirical 3.5-20 SGS-!p'FSAR                                                               Revision 21 December 6, 2004

equations for sec rates were developed based on SWPC field measurements and laboratory test data. The sec rate is given in inches/hour, temperature in °F 1 and the material strength in ksi. da . 7302 dt =axp(-4.968- T +460+ 0.0278 *O'y) The log normal distribution sec rate with a standard deviation of 0.578 is assumed.

f. Initial Crack Size The initial crack size is assumed to be a non-varying variable with a value equal to 3 mm.

2. Unit 2 The Salem Unit 2 LP turbines are equipped with fully integral rotors which will be inspected at suitable intervals to ensure the probability of rotor burst is acceptably low .

The LP turbine inspection schedule is based on the probability of rotor failure and subsequent missile generation. Inspections are scheduled before the probability of a failure increases beyond the threshold level. There are several mechanisms 1 which may result in rotor failure (Reference 5):

1. Ductile burst resulting from extreme rotor over speed.

2. Fracture resulting from high cycle fatigue cracking.

3. Fracture resulting from low cycle fatigue cracking.

4. Fracture resulting from stress corrosion cracking (SCC).

The probability of a turbine missile, assuming that destructive overspeed occurs from simultaneous failure of two valves on the same main steam line to close, has been analyzed in WCAP-11525 and WCAP-16054-P. The frequency of turbine valve testing has been used to maintain the probability of a 5 turbine missile at or below 1 x 10- per year. The missile probability is a combination of the probability that the turbine valves remain open when the control system trips them and the probability of the failure of the control system to trip the valves. The failure of the turbine valves themselves is the most important component of that failure. The simultaneous failure of two valves on the same main steam line to close is the most likely event leading to a turbine missile (6). Scenarios where additional valves fail to close have lower probabilities than the failure of the two valves, since more simultaneous failures are required

3.5-21 SGS-UFSAR Revision 23 October 17, 2007

For the Salem fully integrated rotor, the assumptions used in WCAP-11525 and WCJ\P-16054-P, that one steam path remaining open through the HP turbine will cause a missile, become invalid. The probabilities calculated for a missile in WCAP-11525 and WCAP-16054-P include a small probability of a turbine missile at running speed and at design overspeed. However, in the updates, a destructive overspeed probability is calculated. This destructive overspeed probability now becomes the probability of reaching equilibrium overspeed. Reference 6 describes the calculation of the destructive and equilibrium overspeeds and the relationship between them. Of the four mechanisms listed above, the potential for stress corrosion cracking has the greatest influence on rotor integrity and therefore forms the basis for setting inspection schedules. The approach is to make a conservative prediction of how a presumed or actual crack will grow and then schedule an inspection prior to the time that the crack grows large enough to be of concern. Analytic components of this approach are:

a. Probability of Crack Initiation

b. Crack Growth Rate

c. Critical Crack Size

'rhe Westinghouse criterion as given in Reference 5 for establishing each of these factors is as follows:

a. Probability of Crack Initiation Westinghouse has used inspection data from bu:i.l t-up type rotors to calculate the probability of crack initiation for each disc number in the fully integral rotors. This approach results in conservative estimates since the built-up rotors have stresses and yield strengths which are significantly higher than the fully integral rotors.

b. Crack Growth Rate Westinghouse has performed statistical studies using the field data on crack sizes and shapes as related to temperature of operation, crack location, material strength, and environment. They have used parameters for the crack growth rate model which are the same as those used for keyway stress corrosion crack growth rate in built-up rotors. This approach results in conservative estimates since the built-up rotors have stresses and yield strengths which are significantly higher than the fully integral rotors.

SGS-OFSAR 3.5-21a Revision 23 October 17, 2007

c. Critical Crack Size Westinghouse used the between fracture toughness and stress intensity to calculate critical crack size. Finite element analysis was used to determine crack tip stress intensity factors for a single thru thickness radial crack emanating from the rim of a disc. The most severe thermal stress encountered a transient condition was alao considered in these calculations.

The analysis determined that, at running speed, the stress intensity for all crack depths less than the total depth of the disc was well below the fracture toughness, where the total depth of the disc is defined as the distance from the disc rim to the point where the disc blends into the main body of the rotor. Westinghouse i::onservati vely concluded that the critical crack size is equivalent to the total depth of the disc. A limit load analysis confirmed that ductile fracture of the rotor would not occur under these conditions. Considering the factors above, combined with the number of discs and the proQability of the unit reaching a overspeed condition, analyses were made to determine the probability of rotor bursting. Results indicate that an inspection interval of 30 years will result in a probability of missile generation which is below the threshold value, However, as recommended by Westinghouse, an inspection schedule with shorter inspection intervals will be maintained. With the increased LP rotor component reliability (elimination of shrouds, rivets, and lashing wires), improved LP blade path seating, improved to withstand abnormal operating conditions and the one piece inner cylinder design (minimizes the effects of erosion/corrosion and less susceptible to

              ,   Westinghouse recommends a 10 year inspection interval for the LP This recommendation is based on the expected need for maintenance dtie to normal "wear and tear," and is not driven by nuclear safety concerns.

I 3.5-2lb SGS-UFSAR Revision 21 December 6, 2004

3.5.5 References for Section 3.5

1. DELETED

2. DELETED

3. DELETED

4. CT-25243, "Turbine Missile Report (Heavy Disc - Keyplate Design LP's)"

Revision 0, November 1985.

5. Westinghouse Report WSTG-4-P, "Analysis of the Probability of the Generation of Missiles from Fully Integral Nuclear Low Pressure Rotors,"

October, 1984.

6. Westinghouse Technical Report TM-95185, "Overspeed Analysis for Public Service Electric and Gas - Salem 2 Fully Integral Rotors Dated August 15, 1995"

7. W. B. Cottrell and A. W. Savolainen, "U.S. Reactor Containment Technology,"

ORNL-NSIC-5, Vol. 1, Chapter 6, Oak Ridge National Laboratory.

8. Siemens Technical Report CT-27336, Missile Probability Analysis PSEG Nuclear LLC Salem Unit 1, Revision 1, dated November 5, 2003.

9. EC-02262, Missile Generation Risk Assessment for Original and Retrofit Nuclear HP Rotors, Siemens Westinghouse Power corporation, December 17, 2002.

10. Siemens Westinghouse Topical Report, Missile Analysis Methodology for General Electric (GE) Nuclear Steam rotors by Siemens Westinghouse Power Corporation (SWPC), Project No. 721, dated May 16, 2002.

11. WCAP-16054-P, Probabilistic Analysis Of Reduction in Turbine Valve Test Frequency For Nuclear Plants With Siemens-Westinghouse BB-95/96 Turbines, dated April 2003.

12. PSEG Engineering Document 6S0-1772, Design of Missile Barriers for Diesel Intake and Exhaust Openings. Rev. 0, March 25, 2005.

3.5-22 SGS-UFSAR Revision 32 June 17, 2021

13. Nuclear Energy Institute (NEI) Technical Report NEI 17-02, Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document, Rev. 1B

14. Ho K. Nieh, Director Office of Nuclear Reactor Regulation, Timely Resolution Of Issues Related To Tornado-Missile ProtectionSupplemental Information, Letters to Nuclear Energy Institute, Entergy Nuclear, NextEra Energy, Dated 02/07/2020 (ADAMS Accession No. ML20015A299)

15. Grand Gulf Nuclear Station, Unit 1, Issuance of Amendment No. 220 Related to Request to Incorporate the Tornado Missile Risk Evaluator into Licensing Basis, Dated 06/18/2019 (ADAMS Accession No. ML19123A014) 3.5-23 SGS-UFSAR Revision 32 June 17, 2021

TABLE 3.5-1 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 21 December 6, 2004

TABLE 3.5-2 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-'OFSAR Revision 21 December 6, 2004

TABLE 3.5-3 THIS TABLE HAS BEEN DELETED 1 of 1 SGS-UFSAR Revision 21 December 6, 2004

Figure F3.5-1 intentionally deleted. Refer to plant drawing 204811 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

Figure F3.5-2 intentionally deleted. Refer to plant drawing 232445 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

Figure F3.5-3 intentionally deleted. Refer to plant drawing 232444 in DCRMS SGS-UFSAR Revision 27 November 25, 2013

I II i Ii . _.......,

                             -*-*-,-*-*-*        M!--1-N""...l i                        ~~

I I t Rev1s1on21, Dec. 6, 2004 Salem NuclearGeneratingStation PSEG Nuclear,LLC HIGH PRESSURE CYLINDER 1800 RPM DOUBLE-FLOW DESIGN SALEM NUCLEAR GENERATINGSTATION UpdatedFSAR Figure3.5-4 Iear*LLC. All RightsReserved.

View of turbinr: cylinder IJ.'Id blade ring, showing method ot supporting and locking lower blade ring in position.

Blade rin$s of large high-pressure, high tempr:rarure ~~~rbinr:,

with statiOnary blades in place . Features r) Centedine supporting block insures center alignment while allowing differ-ential expansion between blade ring and cylinder.

z.) Blades are inserted in blade ring halves.

3) Tongue and groove holds blade ring in position.

4) Metallic seals between blade rings and cylinder prevent leakage of steam in support grooves.

5) Upper plate, in cylinder cover, prevents any "riding-up" of the blade ring.

REVISION8 FEBRUARY15,1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Blade Rings SALEM NUCLEAR GENERATING STATION Updated FSAR Figure3.5-5

s..... u w (j). (J) (J) 0 0: u w z co 0:

J w

0:

J (J)

(J) w 0: a:. PSEG Nuclear,LLC SALEM NUCLEAR GENERATINGSTATION

1) ln.oer eyli.ader, supported &.tthe hori:z.oatd ceaterl.i.ne 1.0d fixed tra.a.svc:ady at the top andboctom by dowelpins, .Uows freec!om of expaa.sloa ~t of the c:IUtc:r eylizldet'.

2) Eatircouter eylindcr is at exhaust steam t.emp<:n.tur'c.

3) Fxb*u:Sf hood of labon..toty"1'1'0Vcd design minimires hood los:;.

4) Provision for extn.dloa r.oaes: with molsturc rcmovd.

S) lcnet A!ld outer cylinders of fabriea.tod steel ooast.rue(joo with improved one--piece izmer eylic.der design.

6) Fliminotioa of shrouds, rivets, and luhi.ag wires.

7) Reduced stresscorrosion cnd:ing.

8) Reduced high eycle fatigue.

9) Heat rat.:: improvel.l'.lellt.

11) Op<:ntioa &.t up to eight (8) inches of ba.c-l:pn::ssu.R:.

12) <>P<::ntioa at +/- 1.5 Hz off-frequency.

13) Oae plccc rotor forging.

14) No separate discs.

IS) Integral couplic.gs. Unit2 Low-Pressure Element 1800RPM Double-Flow Designwith47"LastRowBlades PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION Updated FSAR Figure3.5-6A Revision16 January 31, 1998

Location No. o1 l.PINLET IN 1 EXiAACTJON A4 1 E.XTRAC.TJON A3 2 EXTRACnON A2 2 EXTR.ACliON A1 4 EXHAUST ex 2 i

                                                  *-*+*--*

i EndVIew Rev1s1on21, Dec. 6, 2~~4 Solem NuclearGeneratingStation PSEG Nuclear,LLC EXPLODED VIEW OF LOW PRESSURE UNIT SALEM NUCLEAR GENERATINGSTATION UpdatedFSAR

3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 Systems In Which Design Basis Piping Breaks Occur Design basis piping breaks are postulated to occur in the following systems:

1. Reactor Coolant Loop, as limited by the NRCs approval of Leak-Before-Break (Reference 6 in Section 3.6.6).

2. Pressurizer Surge Line, as limited by NRC approval of Leak-Before-Break (Reference 9 in Section 3.6.6. See Section 3.6.4.2.1 for additional clarification on affected systems.)

3. Main Steam System

4. Feedwater System

5. CVC Letdown Line

6. Steam Generator Blowdown Line

7. Steam Supply to Auxiliary FW Pump Turbine

8. CVC Charging and RC Pump Seal Injection (Unit 2 Only)

9. Heating Steam (Unit 2 Only)

10. Heating Water (Unit 2 Only) 3.6.2 Design Basis Piping Break Criteria The criteria for postulating design basis piping breaks are discussed for each system in Sections 3.6.4 and 3.6.5 where the analysis and protection of the various systems are presented.

3.6.2.1 Criteria for Determination of Location and Orientation of Postulated Breaks Inside Containment - Main Steam, Main Feedwater, and Steam Generator Blowdown Salem Unit 1 was designed and constructed in the time frame of September, 1968 through August, 1976. The NRCs Giambusso letter, issued in 1972, imposed HELB 3.6-1 SGS-UFSAR Revision 33 October 24, 2022

criteria on systems outside containment after the plant had already been designed, and major rework was required to meet these new (and changing) criteria. The guidance and criteria noted are those that evolved over time. The Main Steam (MS) and the Main Feedwater (MFW) high energy line breaks (HELBs} discussed in UFSAR Section 3.6 are for breaks containment, but rather containment. Given the above background information, explicit Salem criteria for HELBs inside containment were determined 1 including the Steam Generator Blowdown (SGBD) lines. Based on this information, the criteria used for Salem MS 1 MFW, and SGBD HELBs inside containment are given below. The current NRC guidance for HELBs is provided in Generic Letter 87-11 and in the latest revision to BTP MEB 3-1, attached to that Generic Letter. During the Steam Generator Replacement Project (SGRP), which the original Westinghouse Model 51 steam (SGs) with Model F SGs, the current NRC guidance was used to postulate breaks inside containment for the portions of piping modified by the SGRP. As evidenced by Generic Letter 87-11 and as noted in the "Background" discussions in MEB 3-1, the latest NRC positions are intended to utilize the available piping design information by postulating pipe ruptures at locations having relatively higher potential for failure, such that an adequate and practical level of protection may be achieved. The current HELB of the MS, MFW, and SGBD piping and the criteria used for HELBs inside containment are below. These Positions the current and basis for the Salem MS, MFW 1 and SGBD lines inside containment replaced or modified by the SGRP. Note that any corresponding conformance to the BTP MEB 3-1 Sections related to pipe breaks/cracks does not constitute a commitment to conform to all the positions of the Branch Technical Position. Position I:

1) Circumferential breaks are postulated in piping exceeding a nominal pipe size of 1 inch. This conforms to BTP MEB Section B.3.a(l) .Refer to Positions III and IV.

3.6-la SGS-UFSAR Revision 24 May 11, 2009

2) Longitudinal (slot) breaks are postulated in piping 4 inches nominal pipe size and larger. This conforms to BTP MEB Section B.3.b(1). Also see Position V.

3) Leakage cracks are postulated in piping exceeding a nominal pipe size of 1 inch. This conforms to BTP MEB 3-1, Section 8. 3. c ( 1) . Refer to Position VII.

Position II: A reconciliation, of the differences between the original design of the MS piping system and the as-designed for construction configuration, has been prepared in accordance with NCIG-05, Revision 1 (Guidelines for Piping System Reconciliation). Based on this reconciliation, there is no new break postulated for the MS lines inside containment. Based on these evaluations, it is concluded that no changes are required to any of the existing in-situ protective features provided for protection from a MS pipe break inside containment. Position III: Circumferential Breaks at SG Nozzle (for breaks, see Position V) Circumferential breaks are postulated at the steam generator nozzle for the MS, MFW, and SGBD (i.e., blowdown and drain) piping modified by the SGRP. This conforms to BTP MEB 3-1, Section B.l.c(2) (a). Also see the discussions in Position VI. Position IV: 0.8 (Sh~al (for longitudinal breaks see Position V)

1) Circumferential breaks are postulated to occur at locations where the stresses exceed 0.8 (Sh + Sa). This corresponds to BTP MEB 3-1, Section B.l. c (2) (b) (ii) .

2) The stresses in the modified MFW/SGBD lines inside containment have been confirmed and documented to be below the 0.8 criterion. A reconciliation of the MS line inside containment (see Position II) has verified that the stresses are below the 0.8 criterion. Therefore, circumferential breaks in the MS, MFW, and SGBD lines (except at the SG nozzles) are not postulated to occur. Thus there are no MS, MFW, or SGBD intermediate breaks postulated, based on the 0.8 stress criterion.

3.6-lb SGS-UFSAR Revision 18 April 26, 2000

3) Based on the NRC Generic Letter 87-11, additional arbitrary (non-stress based) intermediate pipe breaks are NOT postulated to occur in any of the high-energy piping (MS, MFW, SGBD).

Position V:

1) Longitudinal (slot) breaks in the SGBD piping are not assumed to occur, since this piping is less than 4 inches nominal pipe size. This is consistent with Position I and with BTP MEB 3-1, Section B.3.b(1).

2) Longitudinal breaks in the MS and MFW piping are not assumed to occur at the terminal ends, consistent with BTP MEB 3-1, Section B.3.b(2).

3) Other longitudinal breaks in the MS or the MFW piping are not assumed to occur, except where other circumferential breaks are assumed. This is consistent with BTP MEB 3-1, Section B.3.b, which states that:

             "The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of the circumferential breaks specified in B.3.a .... "

(BTP MEB 3-1, Section B.3.a discusses locations of circumferential breaks. The circumferential break locations for Salem MS, MFW, and SGBD are those discussed in Positions III and IV, above. Since MS and MFW circumferential breaks are only required to be postulated at the steam generator nozzles, this removes longitudinal breaks from further consideration, and is consistent with NRC BTP MEB 3-1.) Position VI:

1) The MS piping modified by the SGRP is provided as a like-for-like modification in form, fit and function.

The SGRP MS piping modification does not affect the existing in-situ provisions at the MS containment penetration. 3.6-lc SGS-UFSAR Revision 24 May 11, 2009

2! MFW line inside containment: The containmAnt penetration calculations were revised considering pipe loads from the revised stress analysis calculations affected by-the SGRP, as applicable. A re-analysis of the stresses for MFW demonstrates that stresses at the containment pene-tration anchor point inside containment do not exceed 0.25 (Sh +Sa)*

3) SGBD line inside containment: A seismic anchor point is located in the existing piping between the modified blowdown piping and the containment penetration. Thus the SGRP SGBD line rnodifi~ation does not affect the existing stress analysis for the containment penetration.

Position VII: Leakage ("Critical") Cracks

1) Leakage cracks are postulated at axial locations where the calculated stresses exceed 0.4 (Sh + Sal. This is consistent with NRC BTP MEB 3-1, Section B.1.e(2). (Note that Equations (9) and (10) of NC/ND-3653 are equivalent to Equations 12a and 13a of ANSI B31.1, 1977 Edition.)

2) The fluid flow from a leakage crack is based on a circular opening of area equal to that of a rectangle one-half pipe diameter (OD) in length and one-half pipe wall thickness in width. This conforms to BTP MEB 3-1, Section 8.3.c(3).

3) Leakage cracks are not postulated to occur in the MS piping modified by the SGRP, since a reconciliation of the portion of the MS line inside containment modified by the SGRP has verified that the stresses are below the 0.4 criterion. For the rest of the MS piping inside containment (see Position II), the SGRP did not modify the adequacy of the in-situ design features previously provided.

4) The effects of leakage cracks in the MFW piping installed by the SGRP, where the stresses exceed 0. 4 !Sh .,. Sa), have been evaluated and the appropriate mitigating features have been provided, where required.

5) The effects of leakage cracks of the SGBD piping installed by the SGRP, where the stresses exceed 0. 4 (Sh + Sa), have been evaluated and the appropriate mitigating features have been provided, where required .

SGS-UFSAR 3.6-ld Revision 18 April 26, 2000

3.6.2.2 Unit 2 Steam Generator Replacement Activities associated with NRC Generic Letter 87-11 Modifications to the Main Steam, Feedwater and Blowdown piping were performed in order to replace the Unit 2 steam generators in 2008. The effects of these modifications upon these systems with regard to NRC Generic Letter 87-11 and NRC Branch Technical Position MEB 3-1 (Reference 4, as cited in UFSAR Section 3.6.6), are as follows: Main Steam In reattaching the main steam piping to the steam generators, the main steam limit stop (MSLS-1) tie rods attached to each of the main steam pipe 90° elbows were not restored. With the application of NRC Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, this is permitted. Following the installation of the RSG, main steam restraint MSR-2, located closest to the RSG main steam nozzle, was restored to the original design condition. Therefore, the main steam piping is maintained functional by restraining the pipe whip loads from a postulated main steam nozzle circumferential break (terminal end break) without compromising the function of essential equipment. Feedwater The reattachment of the Feedwater piping to the RSGs resulted in no changes to the piping geometry, materials, or existing MFW rupture restraints. Therefore, the feedwater piping is maintained functional by restraining the pipe whip loads from a postulated feedwater nozzle circumferential break (terminal end break) without compromising the function of essential equipment. Blowdown Due to the relocation of the blowdown nozzles on the RSGs, the terminal ends of the blowdown piping were correspondingly relocated. No safety related equipment is located in the vicinity of the relocated RSG blowdown nozzles; therefore, no rupture restraints are required for the terminal end breaks of the re-routed blowdown piping. 3.6-2 SGS-UFSAR Revision 24 May 11, 2009

3.6.3 Design Loading Combination Loadings on protective structures, pipe whip restraints, and other components subsequent to a postulated piping break include pipe pipe whip, jet impingement, pressurization of compartments, water flooding, and steam flooding. They are combined directly as appropriate for each case. Their combination with other loadings generally categorized as faulted condition are included in other appropriate sections in this report; for instance, for Class 2 and 3 components, in Section 3.9.2; and for Class 1 components in Section 5.2. 3.6.4 Dynamic Analysis I Dynamic analyses of the Reactor Coolant System (RCS) and other high-energy fluid systems are described below. 3.6.4.1 Reactor Coolant Loop System Analysis The reactor coolant loop analysis is described in Section 3.9.1.8. I 3.6-3 SGS-UFSAR Revision 24 May 11, 2009

(THIS PAGE IS INTENTIONALL BLANK} 3.6-4 SGS-UFSAR Revision 20 May 6, 2003

3.6.4.2 Reactor Coolant Loop Break Analysis In order to assure the safe shutdown capability of the Salem units in the unlikely event of an RCS pipe rupture, a detailed dynamic analysis of the RCS piping and equipment supports was performed. For the original design analysis, a large number of discrete break locations were postulated in the piping in order to determine the design adequacy and structural integrity under the most critical loading condition. Time dependent loadings resulting from each of the postulated breaks were calculated and applied to lumped mass models of the piping and equipment supports. All piping and structures analyzed were shown to remain stable and functional. 3.6.4.2.1 Postulated Break Locations Postulated break locations are listed in Table 3.6-1 and shown on Figure 3.6-1. The current design basis break locations postulated for the reactor coolant loop utilize the application of leak-before-break (Reference 6,8-14). Circumferential breaks are postulated at the loop nozzles of the accumulator lines, RHR lines, and the pressurizer surge line. The breaks were conservatively assumed to have a one-millisecond opening time in order to analytically describe an instantaneous break. In addition to the large primary loop RCS piping, the NRC approved a license amendment request (Reference 15) to apply Leak-Before-Break (LBB) methodology to piping systems six-inches and larger connected to the Reactor Coolant Loop in both Units 1 and 2, and to the pressurizer surge line. This approval allowed excluding from the Licensing Basis the dynamic effects of postulated ruptures of the following lines:

1. 6-inch Safety Injection Lines - Inclusive of piping from the hot leg (or RHR line for Loop 1) to the SJ156 check valve.

2. 10-inch Emergency Core Cooling System (ECCS) Accumulator Lines -

Inclusive of piping from the cold leg piping of Loop 1, Loop 2, Loop 3, and Loop 4 to the SJ56 check valves near the Accumulator tanks.

3. 14-inch Residual Heat Removal (RHR) Lines - Inclusive of piping from the hot leg piping of Loop 1 up to the RH1 valve.

4. 14-inch Pressurizer Surge Lines - Inclusive of the entire pressurizer surge line from the primary loop nozzle junction to the pressurizer nozzle junction.

3.6-5 SGS-UFSAR Revision 33 October 24, 2022

The supporting analyses (References 10-14) are included with the license amendment request (Reference 15). The LBB concept is based on calculations and experimental data demonstrating that certain pipe materials have sufficient fracture toughness to prevent a small through-wall crack from propagating rapidly and unstably to a catastrophic pipe rupture and to ensure that the probability of a pipe rupture is extremely low. LBB is used to demonstrate that small through-wall cracks in the pipe are stable, and the associated leakage will be detected by the RCS leakage detection systems promptly so that operators can shut down the reactor and take corrective actions before the pipe rupture. See Section 5.2.7 for further discussion of RCS leakage detection. 3.6.4.2.2 Mathematical Modeling (Original Design Analysis) Several mathematical models were used in order to properly assess the consequences of the pipe ruptures described above. The reactor coolant loop piping and primary equipment were modeled for the Westinghouse computer program, WESTDYN. The reactor coolant loop model describes the spatial geometry, lumped mass locations, and other node points as shown on Figure 3.6-2. Stiffness characteristics of the equipment support structures were incorporated as elastic restraints in the reactor coolant loop model. The WESTDYN program computes internal member forces, support structure reactions, nodal point deflections, and stresses, and also determines natural frequencies and normal modes. Supports for the steam generator and the reactor coolant pump were modeled for the STRUDL or STASYS computer programs. Geometry and topology of the structures, properties, and orientation of all members (including appropriate constraint releases), and all support conditions were accurately represented. A typical model for an equipment support is shown on Figure 3.6-3. For blowdown analysis, the entire primary system is represented by a two-loop SATAN-STHRUST reactor coolant loop hydraulic model as shown on Figure 3.6-4. In this model, one loop represents the broken loop and the other represents the three unbroken loops combined into one with appropriate scaling factors. The hydraulic model is used to define the space-time dependent hydraulic forcing functions generated by the fluid in the primary coolant loops during a design basis LOCA. Hydraulic forcing functions are defined throughout the reactor coolant loop at locations where there is a change in either direction or area of flow as shown on Figure 3.6-5. Typical plots of forcing function time histories are shown on Figure 3.6-6. 3.6-6 SGS-UFSAR Revision 33 October 24, 2022

3.6.4.2.3 Solution Method (Original Design Analysis) The time-history dynamic structural of the reactor coolant loop was performed in the following manner:

1. The initial deflected position of the reactor coolant loop model was defined by using pressure analysis.

2. Natural and normal modes of the loop were determined using the WESTDYN computer program.

3. The initial deflection natural frequencies, normal modes, and time-history forcing functions were input into the Westinghouse proprietary computer program, FIXFM, to determine the time-history dynamic deflection response of the lumped mass representation of the reactor coolant loop.

4. The forces imposed upon the supports by the were obtained by multiplying the support stiffness matrix and the time-history of the displacement vector at the support point .

5. The computer program, WESTDYN-2, treated the time-history dynamic deflections at mass as an imposed deflection condition on the reactor coolant loop model and computed internal forces, deflections, and stresses at each end of the members of the reactor coolant loop piping system.

For the support structures modeled, stiffness analyses were performed using the STRUDL and STASYS computer programs to obtain support stiffness matrices. These support stiffness matrices were included in the reactor coolant loop model to represent their effects on the supported system .

SGS-UFSAR 3.6-7 Revision 23 October 17 1 2007

Membc-1r influenCE) coefficients obtained from STRUDL, member properties, and support plane loads obtained from the reactor coolant loop analysis were used in the THESSE computer program to obtain maximum internal forces and stresses in each member of the support system. 3.6.4.2.4 Results of AnaJ.ysis (Original The resulting stresses from the above outlined were compared with the stress/strain limits for the 1.\CS piping and equipment supports. In all cases studied, .stresses in critical support: elements were within yield limits afte.r* load redistribution so that the supported equipment and piping was maintained within the Faulted Condition limits. 3.6.4.3 3.6.4.3.1 Introduction An has been completed which assesses the consequences of postulated brc-~aks :Ln high <-me:rgy f:lu.i.d piping systnrns. /-\lthough such breaks are considered highly improbable, they were postulated to occur at selected locations in the high energy piping systems outside of the conta:Lnment. 3.6.4.3.2 Criteria Used for Analysis The analysis was based upon the guidelines recommended in the document "General Information Required for Consideration of Effects of a Piping System Break Outside of Containment" ( 1) as clarified by Reference 3. Unless otherwise noted, break location criteria recommended in Reference 2 were used where applicable to the system being evaluated. Where the aforementioned criteria were not applicable, other criteria were used which resulted in an equivalent degree of SGS-OF'SAR 3.6-8 Revision 23 October 17, 2007

Some additional assumptions and definitions beyond those provided by Reference 1 were required to define the bounds of the analysis. These and a sum.rnary of the criteria recommended by Reference 1 are as follows:

1. High energy piping systems for Unit 1 are those whose temperature whose pressure exceeds 275 psig during normal reactor operation. For Uni c: 2, high energy piping systems are those whose exceeds 200°F or whose pressure exceeds 275 psig during normal reactor

2. Except as discussed in Sections 3. 6. 5. 2 and 3. 6. 5. 3, high energy lines were assumed to fail circumferentially or longitudinally at the following locations:

a. Terminal ends of piping runs

b. All locations where the summation of stresses due to listed in Reference 1 or circumferential) calculated on an elastic basis exceeded 0. 8 ( Sh S ), or the a

expansion stresses exceeded 0. 8 S . For piping which did not a receive seismic the thermal expansion criteria alone was applied

c. At the two most highly stressed points between terminal ends (if two such points were not required by b above)

Subsequent to the original and construction of the NRC Generic Letter 87-11, Relaxation in Intermediate was issued. The revised BTP 3-1 eliminated to consider the dynamic effects and the environmental effects from arbitrary intermediate pipe ruptures that were in the original plant design. 3.6-9 SGS-UFSAR Revision 25 October 26, 2010

Design basis cracks were postulated to occur in heating steam and heating water high energy lines at selected locations having !tlaximum stress levels. At these locations of maximum stress, the piping was encapsulated and vented outside the Auxiliary Building to from the adverse environmental effects due to steam escaping from these of postulated failure. For Cnit 1 only, basis cracks were for those systems whose pressure is !tlore than 275 psig and whose temperature is more than 200°F.

3. A single component failure was assumed to occur within the combined to attain cold shutdown for those accidents that cause protection system actuation. An exception is that for a main steam line break in the exclusion zone of the penetration area, i.e., from the containment to the MSIV, a single active failure is not judged to be credible. The active or passive nature of the single failure assumed is subject to the stipulation that during recirculation phase, one active or passive failure must be accommodated, but not in addition to a failure in the ection (an design basis).

Per the Salem basis, safe shutdown is hot standby. The Salem design is not required to meet the 36 hour cool down to cold shutdown requirements of NRC BTP RSB 5-l as detailed in Section 5.5.7.3.4. The plant can be maintained in a safe hot standby condition while manual actions are taken to permit achieving cold shutdown conditions. Alignment for initiating RHR decay heat removal to commence cool down c.o cold shutdown conditions can be achieved in 36 to 4 8 hours post accident. For fire events, cold safe shutdown is within 72 hours. 3.6-9a SGS-UFSAR Revision 25 October 26, 2010

THIS PAGE INTENTIONALLY BLANK 3.6-9b SGS-UFSAR Revision 7 July 22, 1987

4. The analysis of the effects of pipe failure performed on each postulated rupture includes the following:

a. Pipe whip

b. Jet impingement and thrust reaction

c. Water flooding

d. Steam flooding

e. Compartment pressurization

5. No other accident was assumed to occur concurrently with the pipe failure outside of containment. For high energy lines as defined in this section under item 2, analyses have been made to assure that occurrence of design basis cracks would not result in loss of plant safety functions.

3.6.4.3.3 Analytical Models and Techniques Several analytical models and techniques were utilized for the c analyses. The principal models used were:

1. Pipe Stress Analysis Stress analysis for selection of pipe break locations was performed using the Franklin Institute Research Laboratory's PIPDYN program. The program provides automatic summation of longitudinal stresses due to dead weight, internal pressure, thermal expansion, and earthquake (where applicable) at all points of interest. Stress intensification and flexibility 3.6 10 SGS-UFSAR Revision 7 July 22, 1987

effects were included in accordance with the provisions of ANSI B31.1.

2. Piping Slowdown Models Mass and ene:r:gy flow :r:atea fo:r: pure steam and eatu:r:ated water piping ruptures were calculated using the Moody blowdown model. For fluids aubcooled in excess of 30 atu/Lbm, isentropic incompressible fluid flow theory was utilized.

A discharge coefficient of 1.0 was used for circumferential and full longitudinal breaks, and a discharge coefficient of 0.6 was used for crack-1 ike openinqs. Resultant environmental conditione were calculated assuming adiabatic expansions from the source pipes. Ho credit for local heat losses was taken.

3. Pipe Whip Model Pipe whip momentum thrust forces were calculated using the relationship:

Fj "' where: Fj - jet reaction force (lbf) A

                    -     break area 2

(in ) K thrust coefficient

                    =     1.26 for saturated ateam or saturated water
                    =     2.0 for suhcooled water c
                    -     discharge coefficient 3.6-11 SCS-UI'SAR                                                     Revision 7 July 22, 1987

1.0 for circumferential or full longitudinal breaks

                   ... 0.6 for crack-like breaks p2   ... pressure of fluid in pipe, psiq p

a = local atmospheric pressure

4. Jet Impingement Model Jet impingement forces were calculated using the relationship:

FI = F. CD X

where: FI = Impingement force on the target F. J

                   =    total jet force calculated in 3 above CD   =    drag coefficient, a function of target shape.

X = fraction of jet intercepted by target, assuming a 10 degree jet divergence half-angle for gases, 0 degree divergence half-angle for liquids.

s. Compartment Pressurization Analysis Model Steady state compartment pressures were calculated using the piping blowdown models described in 2 (above), for defining fluid influx rates, and isentropic flow models to define gas efflux rates.

3.6-12 SGS-UFSAR Revision 16 January 31, 1998

3.6.5 Protective Measures 3.6.5.1 Nuclear Components The arrangement of nuclea~ safety-related equipment for this plant was developed in accordance with Preliminary Safety Analysis Report commitments made at the time of the receipt of the Construction Permit. Even though there were no industry wide or governmental standards in effect at the time, due consideration was given to providing assurance that engineered safety features within the containment are adequately protected in the unlikely event of an accident. This protection is in the form of barriers, supports, restraints, and/or physical separation, as necessary. 3.6.5.1.1 Barriers The polar crane wall and the refueling floor serve as barriers between the reactor coolant loops and the containment liner. The 3 foot thick wall, which extends from Elevation 81 feet to 130 feet, acts as a barrier between the containment liner and the sources of jet forces, pipe whip, and missiles associated with a failure of the RCS. All "essential components" (safety-related components in the containment which are required for operation during an accident) are located behind these missile barriers and therefore, are not subject to damage resulting from the dynamic effects associated with a LOCA. The annulus between the reactor cavity and the missile barrier is subdivided into two compartments by the refueling cavity, one for each pair of reactor coolant loops, to prevent the proliferation of dynamic effects extending beyond a two loop compartment. The steam and feedwater lines are protected from LOCA associated dynamic effects throughout their length. Above Elevation 130 feet these lines are protected by their structural steel restraints. Below Elevation 130 feet the steam and feedwater lines are routed where possible behind barriers which protect the containment liner from the steam or feedwater lines as well as the steam or 3.6-13 SGS-UFSAR Revision 7 July 22, 1987

feedwater line from the reactor coolant loops. The steam and feedwater lines associated with a given steam generator are separated. A missile barrier has been installed around that portion of the pressurizer that extends above the loop compartment. This barrier will prevent potential missiles from reaching the containment liner, engineered safeguard pipes, or essential equipment which is located outside the reactor compartments. 3.6.5.1.2 Arrangement All safety injection lines have check valves located as close as possible to the reactor coolant loop connection. The check valves effectively shorten the reactor coolant pressure boundary and minimize pipe whip. All safety injection piping with the exception of the individual branch lines which feed a given reactor coolant loop is located outside the containment missile barrier. Check valves +on the Residual Heat Removal (RHR) pump discharge ( lines and the normally closed motor operated gate valve on the RHR pump suction line have been located as close as possible to the reactor coolant loop connections thereby shortening the length of pipe containing pressurized water. A check valve is located close to the connection to the main pressurizer spray to minimize length of pipe whose rupture would result in a LOCA. Size of line and routing prevent damage from pipe whipping. Two check valves on each reactor coolant pump seal water injection line are located inside the missile barrier close to the pump connections to minimize pipe whip. 3.6-14 SGS-UFSAR Revision 7 July 22, 1987

All sample lines have normally closed valves, located inside the missile barrier. Whipping could result if a rupture occurred between the reactor coolant loop and the closed valves. The size of the sample lines and the pipe routing are such that subsequent damage to other piping would not occur even if whipping occurred. Equipment located inside the containment which is required for operation during a LOCA1 such as the containment spray piping, containment isolation valves, containment fan cooling units, essential electrical cables and instrumentation are all located between the missile barriers. 3.6.5.1.3 Restraint Where physical arrangement and barriers were ascertained to be impractical and/or to offer insufficient protection, physical restraint of piping was provided on reactor coolant, main steam, and feedwater lines. In the original design, circumferential and longitudinal ruptures of the reactor coolant loop and surge line were postulated at of change in direction, and restraint has been provided to pipe whip from occurring (see Figure 3. 6-7). Note that for the reanalysis of the RCL, as a result of the application of leak-before-break, the breaks in the primary piping are eliminated. Breaks postulated for the RCL analysis are the accumulator line at the loop nozzle, RHR line at the loop nozzle, and the pressurizer surge line at the loop nozzle (see Table 3. 6-1). The reactor vessel and steam generator supports have been designed to prevent an initial rupture of a reactor coolant loop from causing the failure of another reactor coolant loop, or a steam or feedwater line. Also the steam support system prevents a reactor coolant hot leg rupture from causing the failure of the cold leg and vice-versa as well as a steam or feedwater rupture from damaging a reactor coolant line. All lines connected to the reactor coolant loop, which penetrate the containment wall, have been anchored to the crane wall. Each anchor has been designed to be stronger than the pipe. Should a reactor coolant loop rupture occur, the resulting jet force will therefore not be transferred to the containment wall through any branch lines. 3.6-15 SGS-UFSAR Revision 24 May 11, 2009

The charging and letdown lines are normally opened to the reac;:or and their rupture could result in whipping within a loop compartment. The lines are small, however, and therefore unable 1:0 damage the large reactor coolant lines. Both lines are anchored to the missile barrier to prevent containment damage from RCS break forces being transferred through the charging or letdown lines. Safety injection piping which penetrates the containment has either been anchored at the missile barrier wall to prevent a reactor coolant loop rupture from breaching the containment through the safety injection line or has been routed to assure that loop rupture forces cannot transfer sufficient loads back to the containment wall through the safety injection line to cause a breacr.. All RHR lines penetrating the missile barrier have been anchored to the missile barrier wall to prevent reactor coolant pipe rupture forces from being transferred ~o the containment through the RHR branches. The portions of the steam and feedwater lines not routed within barriers are restrained to prevent pipe whip. Main steam and feedwater lines have been anchored at and restrained outside the containment so that rupture will not affect containment integrity. (See figures 3.6-8, 3.6-9, and 3.6-10.) 3.6.5.1.4 Protective Measures Evaluated for Salem Unit 1 Main Stearn, Main Feedwater, and Steam Generator Blowdown as a Result of the Steam ( Generator Replacement Project ll A deficiency has been resolved to include the effects of the MS terminal end break at the steam generator nozzle on the MS/MFW pipe whip restraint structure. 2! A MFW pipe whip restraint has been designed, analyzed, and provided to mitigate a circumferential rupture of the MFW line at the steam generator nozzle, including the effects of the jet impingement on the MFW pipe whip restraint.

3) The effects of a circumferential rupture of the MFW line at the steam generator nozzle have been evaluated, and the MS/MFW pipe whip rescrainc structure has been modified to mitigate this break.

4) The SGBD line has been evaluated to assure that, for* a circumferential rupture of the SGBD line at the steam generator nozzle, an unrestrained SGBD l1ne does not affect the safe shutdown capability of the unit for tha: scenar~o, assuming a loss of offsite power.

C... 3.6-16 SGS-UFSAR Revision 18 April 26, 2000

3.6.5.2 Main Steam System The Main Steam System is illustrated on Figures 10.3-lA and B and is described in Section 10.3. The piping arrangement is shown on Figures 3. 6-10 through

3. 6-13. Each of the four 30-inch OD x 1.007-inch wall main steam pipes runs from the nozzle on its respective steam generator inside the containment (Elevation 172 feet) vertically downward and then out of the containment through a separate penetration which also forms an anchor point in the system (Elevation 108 feet}. A 16-inch ID flow restrictor is located in the vertical run of each main steam line inside the containment at Elevation 135 feet. Two steam lines penetrate the containment and enter an enclosed penetration on the north side of the containment, while the two remaining steam lines penetrate the containment into an enclosed penetration area on the south side. Each steam line then changes to 32-inch OD x 1.5-inch wall thickness and is routed toward the turbine in a horizontal run. This horizontal* section within each penetration area contains the 34-inch 00 x 3.314-inch thick safety valve manifold and the 24-inch ID main steam isolation valves (MSIVs) .

Downstream of the MSIVs, the steam lines were originally designed to return to 32-inch OD x 1.073-inch wall thickness, run vertically to Elevation 125 feet or 13 feet, exit the penetration through wall sleeves, and run horizontally to the mixing bottle, the four steam lines are manifolded to two pipes which then run to the high pressure turbine. The Main Stearn System has been conservatively designed in order to provide the u~timate in reliability of operation under normal conditions and improbability of failure under abnormal conditions. The piping, as originally designed, has been manufactured with a conservative wall thickness of 2.11 inches inside the penetration* wall sleeve, 1. 50 inches from the sleeve to the main steam safety valve manifold and 3.314 inches at the safety valve manifold, even though the code required wall thickness is less than 1 inch. This extra wall thickness serves the dual purpose of eliminating any "high stress" areas at the penetration anchor (thus making the anchor point no more likely to fail than any ather portion of the line) and providing added assurance of preservation of containment integri~y and main s~eam isolability. To further assure 1solability, maintenance of containment integrity and continued shutdown capability, an extensive network of steel limit stop columns and supports has been provided in the penetration area. This steel has been designed to withstand the fluid dynamic forces I 3.6-17 SGS-UFSAR Revision 18 April 26, 2000

associated with the postulated pipe breaks and thus prevents the imposition of excessive strains upon the piping upstream of the MSIVs or upon the containment pressure boundary as a result of such breaks. / Mechanical restraints have been installed at, and adjacent to, the MSIVs due to their criticality and proximity to the containment. Several types of restraints have been used, each serving a distinct function. As shown on Figure 3. 6-9, restraints are provided on the steam lines adjacent to the MSIVs. These restraints serve to limit pipe motion following a postulated rupture such that neither excessive moments nor physical impact can damage the containment, containment penetrations, or MSIVs. Supports are provided on the valve body to aid in supporting valve weight and supply restraint against undesired motion. Where necessary, the valve operator has also been restrained to limit seismically induced motions. The entire Main Steam Piping System outside the containment has been designed in accordance with the ANSI B31.1, Code for Pressure Piping Power Piping. The piping was purchased in accordance with ASTM Specification A155 Grade KC70 Class I, A155 Grade KCF70 Class I, AI06 Grade B, A106 Grade c, or substitutable chrome alloy, or stainless steel material depending upon the specific piece in question. The ASTM Al55 piping was subjected to supplementary tests Sl (check analysis on each length of pipe), S2 (Tension and Bend Tests), S~ (Hardness Tests), and 54 (Magnetic Particle Examination) ( in addition to the 100 percent ultrasonic examination of the unformed plate required by the base ASTM Plate Specification. The ASTM Al06 piping was subjected to supplementary tests 52 (check analysis on each length of pipe), 54 (flattening test on each length of pipe) 1 and 55 and 56 (etching test on each length of pipe). Radiographic examination has been performed on joints in piping of 3/4-inch wall thickness and over, and nozzle welds were subjected to magnetic particle examination. Welded joints in the main header downstream of the MSIVs and welded joints in all main steam piping upstream of the MSIVs were made utilizing the gas tungsten 3.6-18 SGS-UFSAR Revision 16 January 31, 1998

arc welding p~ocess during the first pass, either with the consumable insert or open root with hand-fed filler metal technique. Additionally, that portion of the piping which runs outside the containment from the containment wall penetration to the MSIVa has been constructed in accordance with the materials, fabrication, fabrication inspection, and quality control requirements of ANSI B3l.7 for Nuclear Class 1 Piping. 3.6.5.2.1 Criteria for Determination of Location and Orientation of Postulated Breaks Critical evaluation of the criteria for postulation of pipe breaks presented in paragraph 2 of Reference 1 has shown them not to be applicable, in toto, to the main ste&Jll (or feedwater) piping in the Salem plant. This is a result of differences between the design concept utilized for this piping and the design concept utilized for the piping upon which the criteria were based. Reference l suggests three location criteria for postulation of pipe breaks:

1. At terminal points (anchor points)

2. At locations where stresses due to normal operational loadings are in excess of 0.80 of code-allowable stress

3. At a minimum of two locations between anchor points These locations appear to be baaed upon the technically justifiable premise that the probability of failure of a pressure boundary component is directly related to the stress level in that component. A well known and established theory of mechanical metallurgy, the Griffith Theory 2, has shown that propagation of a crack in metal is related to the square of the stress level in the metal. Hence, the square of the ratio of two given stress levels yields the relative probability of the propagation of a crack at those stress levels.

3.6-19 SGS-UFSAR Revision 16 January 31, 1998

Based on the above criterion, Reference 2 has proposed a "probabLlity of faLlure 2 ratio"* of 0.64 (0.64 = 0.80 ) as a critical value, above which a localized pipe break must be considered "possible." Accepting the apparently arbitrary selection of the 0. 64 probability ratio as an engineering judgment, the criterion can be considered universally applicable in that it is using actual stress levels in the component under examination as the basLs for postulating failures of that component. No such universal applicability can be attributed to criterion 1, however. The suggestion of postulating pipe breaks at tenuinal points {criterion 1) appears to be based upon statistical evidence which indicates that, for piping systems designed to the minimum requirements of the applicable code and designed without special provisions for equalizing stresses within the system, there is a sLqnificantly higher stress level at tenuLnal points ( l .39 times that encountered in straight runs), and hence a higher relative probability of failure 2 at terminal points ( l .39 = 1. 93 times that in straight runs)

If, however, the above phenomenon was taken into account during the design phase of a system and the piping was specifically designed such that stresses at terminal points are not significantly greater than stresses in straight runs, the increased probability of failure at terminal points does not exist, and applicatLon of criterion 1 is not justifiable in light of the minimal increase in factor of safety that could be achieved. Such is the case for the main steam (and feedwater) piping at the Salem plant. Any justification for the postulation of a tenuinal point break in the penetration area baaed upon amplification of stresses due to pipe support errors is also untenable, since the steam line is rigidly supported, not spring
7.'he "probability of failure ratio'* is defined as the ratio of the probability of failure under stresses associated with normal operational loadings to the probability of failure permitted by code allowable stress levels.

3.6-20 SGS-UFSAR Revision 6 February 15, 1987

hung, adjacent to the penetration anchor (due to the absence of vertical thermal motion in this area) . Based upon the actual low stress levels present in the main steam (and feedwater) piping systems (see Tables 3.6-2 and 3.6-3) and their conservative anchor design, breaks have not been arbitrarily postulated at terminal points. Additionally, Reference 4 specifies that breaks and cracks need* not be postulated for the piping from the containment wall up to and including the isolation valves provided that break exclusion are satisfied. The main steam piping between the containment penetrations and main steam isolation valves was demonstrated to satisfy the following stress criteria for break in Reference 4:

1. The sum of stresses due to sustained loads, occasional loads and thermal expansion including a DBE do not exceed 0.8 (1.8 Sn ~Sal.

2. The stresses due to a break outside of the break exclusion area do not exceed the lesser of 2.25 Sh and 1.8 Sy.

For those of high-energy fluid piping, preservice and subsequent inservice examinations are performed in accordance with the requirements specified in ASME Section XI. During each inspection interval, as defined in IWA-2400, an ISI is performed on all ASME Code Section XI circumferential and longitudinal welds contained within the break exclusion region for high-energy fluid system. piping as required per the approved risk informed break exclusion region (RI-BER) process. Breaks have been postulated, however, at locations where relative stress levels indicate them to be appropriate for protection. In order to provide a degree of protection equivalent to the recommended Refe:rence 1 criteria, the following were implemented:

1. The wall thickness of the main steam piping within the areas and in that portion that runs adjacent to the Auxiliary Building (Control Room) wall was increased to 1.5 inches.

2. During the initial plant design stresses in the main steam piping runs described in 1 above were limited to 0.25 + for the normal operational plus OBE ( 1/2 SSE) loading cases (or plus safety valve discharge loading cases) .

3.6-21 SGS-UFSAR Revision 21 December 6, 2004

3. Circumferential and longitudinal breaks were postulated at high stress locations in the steam line (point A and on Figure 3.6-10).

4. Critical cracks (as defined in Reference l) were postulated in the steam line both inside and outside the penetration areas.

5. For the break exclusion piping in the penetration areas, the following assumptions were made for performing equipment environmental qualifications:

a. The largest main steam line break size between the containment penetrations and main steam isolation valves was assumed to be 1.0 ft 2

This assumption satisfies the requirements of Reference 5 with regards to main steam and feedwater piping that meet break exclusion requirements, The purpose of this assumption is to ensure that, even though a break is not required to be postulated per Reference 4, essential equipment is qualified for the environmental effects of a nonmechanistic longitudinal break having a cross sectional area of at least one square foot. This provides added protection for essential equipment located in the penetration area.

(

b. No single active failure was postulated to occur s~ultaneously with~the break. This assumption is applicable since no break is postulated per the break exclusion requirements and the sole purpose of the assumed nonmechanistic break is to provide protection of essential equipment.

3.6.5.2.2 Analysis of Postulated Pipe Breaks As stated in Section 3.6.5.2.1, main steam line pipe breaks were postulated at point A and on Figure 3.6-10 in both the circumferential and longitudinal directions. 3.6-2la SGS-OFSAR Revision 16 January 31, 1998

THIS PAGE INTENTIONALLY BLANK 3.6-2lb SGS-UFSAR Revision 16 January 31, 1998

The forcing function used for design of restraints was assumed to be a step function of magnitude 1. 26 PA, where P is the initial pressure in the pipe (conservatively assumed to be hot standby pressure) and A is the cross sectional flow area of the ruptured pipe. The 1.26 thrust factor is theoretically derivable from ideal gas laws, its conservatism is assured in that the Moody correlation a saturated steam thrust factor closer to

1. 20. Dynamic load factors were used where applicable to account for energy loadings in the static analysis of structural components.

I In addition, the main steam (MS) lines of the Salem steam generators are subjected to postulated guillotine breaks. The pipe break points are near nozzle weld sections. The time history analysis is performed to determine the pipe whip restraint reaction forces and the the break point movements of the MS line. The pipelines from the nozzle break points to the wall anchorage points are modeled with plastic pipe and elbow elements. The pipe whip restraints are modeled with springs. Each directional restraint gap is considered. Pipe tie rods are modeled as slacks. The gap-spring combination element is used for the restraints. Pipe break force time histories are used for the break points and elbow tangent points. Unit area thick plates are used to carry the follower loads. The beam elements under the thick are used to stabilize the All of the and beams have no masses. The fontal solver of ANSYS Version 5.2 is used for dynamic analysis. The pipeline and its restraints are properly modeled in the finite element mode. Damping coefficients are considered proportional to the mass and stiffness. Seven (7) percent of critical damping is used for frequencies of 10 Hz and 100 Hz. The damping is lower between these frequencies and higher for other frequencies. The restraints had no damping associated with stiffness. The restraints are considered to have no masses; hence no damping is associated with masses either. restraint condition is to the anchorage sleeve for the cut-off point boundary conditions. The analysis results are time history plots and listings of designated response items. The extreme response values and time are also determined. The following items are both important to safety and are located near the postulated break locations in the main steam and feedwater lines:

1. The wall adjacent to the Control Room.

2. The wall.

3. The penetration area roof beneath the steam lines.

3.6-22 SGS-UFSAR Revision 24 May 11, 2009

4. The ventilation intake on the penetration area roof.

5. The inboard. penetration area north - south wall between columns CC and.

DD at Elevation 100 feet. Calculation of the energy loading imposed upon the Auxiliary Building wall adjacent to the Control Room and. upon the penetration area wall (items 1 and. 5 above) by the postulated. ruptured linea showed that it was doubtful that the walls could remain abaolutely leaktight and impervious to steam after impact of the pipe. To preclude unacceptable damage, structural members were built up behind the main steam and feedwater pipes, as shown on Figures 3.6-10, 3.6-11, and 3.6-12. The design basis for these structural members, which were designed using a dynamic load. factor of 2, was to be able to withstand. the calculated. pipe break forces and. thus prevent pipe impact with the above mentioned walla following the postulated breaks. Fluid impingement due to design basis cracks will not compromise the integrity of the walls. The total resultant forces and unit impingement pressures can be easily withstood by the walla, and apalling due to induced thermal gradients will be minimal. The pipe support structure in the yard was reinforced to prevent pipe whip impact with the ventilation intake structure. The intake itself was raiaed to Elevation 170 feet and steel plate was added to the pipe support structure in order to prevent steam ingestion by the ventilation system. Pipe whip energy loadings on the containment wall were computed by assuming the pipe would undergo rigid body rotational motion about the nearest credible plastic hinge. consideration of jet force, pipe linear mass density, and rotational inertia then permits calculation of rotational velocity and linear energy density loadings on the building. For impact loading, the containment was checked for penetration depth under a low mass, high velocity impact, and deflection under 3.6-23 SGS-OFSAR Revision 16 January 31, 1998

a large mass, low velocity impact. Equivalent static loading was obtained from a pseudo-stress method by equating the product of load and displacement to the kinetic enerqy. The method of plastic collision was also used to account for enerqy absorption in the momentum transfer. The ductility factor of the structure was checked and held within limits to maintain structural stability. The wall was shown to remain intact, leaktight, and completely functional after impact. Breach of containment integrity by overstressing or distorting wall penetrations is precluded by providing reinforced piping at penetrations and local pipe restraints in the penetration area. No moments or forces of unacceptable magnitude resulting from the postulated main steam pipe breaks can be imposed upon the penetration assemblies. Details of the penetration area restraint steel are shown on Figures 3.6-11 through 3.6-13 and a typical restraint is shown on Figure 3.6-14. Barrier protection is provided for safety-related items to withstand the loadings caused by jet forces and all credible missiles resulting from the critical crack. Full jet impingement forces were taken to be 1.26 PA and 0.6 x 1.26 PA for open ended and crack-like breaks, respectively. Where barriers are not immediately adjacent to jet sources, a jet divergence half-angle of 10 degrees is used to calculate the portion of the jet intercepted. The Class I (seismic) structures housing the main steam and feedwater lines were ( designed baaed on ACI 318-63 "Working Stress Design* for normal operating load plus Operating Basia Earthquake, and *ultimate Strength Deaiqn" for normal loads plus Design Basis Earthquake or Tornado. In the working stress design under Operating Basis Earthquake loads, the allowable stresses are one-third above the normal applicable code working stresses. Wind stresses have been found to be less critical than those generated by an operating Basis Earthquake. Load factors of unity have been used in the ultimate design under Design Basis Earthquake or 3.6-24 SGS-UFSAR Reviaion 16 January 31, 1998

tornado loading. The design lilnit for tension members (i.e., the capacity required for the design load) is based upon the yield stress of the reinforcing steel. The stress in reinforcing steel for ultilnate strength design has been kept below 0.9 Py. The capacity reduction factor "e" for concrete stress is applicable to all Class I (seismic) structures. A coefficient "k" of 0.85 for 3500 psi concrete has been used in addition to "e" for equivalent rectangular concrete stress distribution. The capacity reduction factor *e* is provided for the possibility that small adverse variations in material strengths, workmanship, dimensions, and control (while individually within required tolerances and the limits of good practice) occasionally may combine to result in under-capacity. For tension members, the factor "e" is established as 0.95. The factor "e" is 0.90 for flexture and 0.85 for diagonal tension, bond, and anchorage. Steel members inside the Class I (seismic) structures are designed in accordance with the AISC Manual of Steel Construction (Sixth Edition) except for the main steam and feedwater limit atop steel and the containment liner which are designed with *e*

0.95 for tension and 0.90 for compression and shear.

The possibility of all load reversals was considered and analyzed. The design was predicated on the combination of loadings that caused the most severe conditions. 3.6.5.2~3 Effects-of Postulated Breaks on Safety-Related Equipment The Class I (seismic) structures have been reviewed for their adequacy based on the postulated critical cracks of the main steam and feedwater lines. The walls and floors in the penetration areas have been designed for a dead load of 325 paf and a live load of 250 psf. The 3.6-25 SGS-tJFSAR Revision 6 February 15, 1987

equipment loads are included in the dead loads. The above loadings are based I on the Working Stress Method. The floors and walls in this area are of withstanding a 4. 0 pressure buildup plus a 4. 0 foot head of water on the Elevation 100 feet slab and the resultant water pressure distribution on the walls, using the ultimate stress method previously described. The use of reinforcing steel stress limits of yield strength and ultimate tensile strength as determined from mill test report data wi thm:t capacity reduction factors gives calculated internal pressure capabilities of 5.1 psi and 7.6 psi, respectively, thus assuring continued structural stability of the building during the postulated occurrence. Where must be sealed to prevent fluid flow between acent areas, steel dogged doors are used. These doors are capable of remaining leak-tight against the pressure loadings previously described. Roof-hatch covers in this area are steel encased, reinforced concrete "boxes" approximately 3 feet thick, bolted down to withstand tornado loadings. As such, they are unsuitable for use as pressure relief openings and no credit has been taken fur them. Failure of other non-Class I structures will in no way the Class I (seismic) structures. electrical and control systems have been engineered and designed to meet the Single Failure Criteria as required by IEEE Standard 27 9-1968, "Proposed, IEEE Criteria for Nuclear Plant Protection Systems." The following features are provided in the safety-related electrical and control systems:

1. Safety-related equipment is actuated from redundant protection system trains.

2. Power for have been from redundant vital buses.

3.6-26 SGS-UFSAR Revision 25 October 26, 2010

3. Air supply for pneumatic actuators is supplied through redundant air headers backed up by the two emergency air compressors.

4. Redundant safety-related equipment has been spatially isolated and mounted in separate seismically qualified enclosures.

safety-related equipment located in the penetration areas is shown on Figures 3.6-15 through 3.6-18 and includes the steam generator pressure transmitters and the solenoid valves required for actuation of the main steam isolation valve - vent valves. These instruments are mounted in individual NEMA IV, Class I (seismic) qualified enclosures. The enclosures, with equipment installed, have been environmentally qualified for operation under the most 2 adverse ambient conditions poatulated for an assumed nonmechanistic 1.0 ft longitudinal crack in a main steam line or a postulated crack in a feedwater line within the penetration area. In addition, the individual enclosures have been arranged for spatial separation within the penetration area to maintain the channel separation requirements and to meet single failure criteria. A Leak Detection System has been provided for the penetration area to supplement the steam break and feedwater break protection systems, which now include:

1. High steam line differential pressure

2. High steam flow

3. Low steam line pressure

4. Steam generator low water level

s. Low feedwater flow

6. Steam flow/feedwater flow mismatch 3.6-27 SGS-UFSAR Revision 16 January 31, 1998

The penetration area Leak Detection Syatem includes monitoring of the ambient temperature in the penetration area by nonindicating temperature switches. A aet of four temperature monitors will alarm on temperature rise indicative of steam or feedwater leakage. 3.6.5.2.4 Analyais of Postulated Break on Unit Shutdown Capability Postulated main ateam line and feedwater line breaks outaide of the penetration area have been analyzed for adverse effects that would jeopardize the capability to effectively and aafaly ahutdown the reactor. The postulated steam line and feedwater line breaks cauae a transient which results in a number of automatic operations that trip the reactor, isolate the main steam and feedwater lines and bring the reactor to a shutdown condition. Subsequent cooldown of the RCS is accomplished according to plant emergency operating procedures. Detailed analyses of steam pipe ruptures and loss of normal feedwater are discussed in Section 15. These analyaea demonstrate that no significant effects result from the postulated line breaka which would prevent the cold shutdown of the reactor and protection of the core against cladding damage. A break of adequate aize in any of the four main steam lines is detected by steam line flow and pressure instrumentation. The main steam lines are automatically isolated upon receipt of high steam line flow signals from two of the four main steam lines (one out of two per line), in coincidence with either low main steam line pressure signals (two out of four linea) or low-low RCS average temperature (two out of four l6ops). small breaks outside containment may not receive an automatic isolation aiqnal and therefore require manual isolation by the operator. Isolation is assumed to occur 10 minutes after the steam line break. The po*tulated break reaulta in the lowering of the RCS temperature and pressure and aubsequently low pressurizer pressure and level. The Safety Injection System (SIS) is automatically actuated upon receipt of either coincident low pressurizer pressure and low preaaurizer level aiqnala {one of three), a steam line isolation signal, steam line differential pressure siqnala between a steam line and the remaining lines. The safety injection siqnal will also trip the reactor and main 3.6-28 SGS-UFSAR Revision 16 January 31, 1998

feedwater pumps, initiate feedwater isolation and start the auxiliary feedwater pumps. These automatic actions will shut down the reactor and mitigate the consequences of a steam line break. Operating personnel will verify that the preceding automatic actions have occurred and then proceed according to recovery procedures, which include:

1. Identifying and isolating the affected steam generator

2. Regulating auxiliary feedwater flow to the unaffected steam generators to maintain an indicated water level

3. Stopping the safety injection pumps once pressurizer level is restored and activating the pressurizer heaters to maintain a steam bubble in the pressurizer

4. Stabilizing reactor coolant temperature and steam generator pressure and level by steam dump through atmospheric relief valves When the RCS temperature has stabilized, operating personnel will commence boration of the RCS as necessary for cold shutdown. Cooldown will proceed within design limits according to plant operating procedures. When the RCS reaches the required temperature and pressure, RHR System operation will be initiated to bring the reactor to the cold shutdown condition.

The electrical, control, and instrumentation equipment necessary for the automatic and manual operations described above are located in the containment, penetration area, and Auxiliary Building. Postulated steam line breaks outside the containment will not affect redundant safety-related controls and equipment required to bring the reactor to a cold shutdown condition. The control room will be available for the recovery and subsequent shutdown of the reactor following the postulated event. Access to the control room will be available and the equipment in the 3.6-29 SGS-t1FSAR Revision 6 February 15, 1987

control room will remain functional following a steam line or feedwater line rupture. Details of the Control Room design and ventilation system design are discussed in Sections 7 and 9, respectively. It is concluded that the capability to bring the unit to a cold shutdown condition is not jeopardized by the postulated main steam line rupture. 3.6.5.3 Steam Generator Feedwater System The Steam Generator Feedwater System is shown on Plant Drawings 205202 and 205302 and described in Section 10.4. The piping arrangement is shown on Figures 3.6-10 through 3.6-13. Each of the four feedwater pipes runs from its respective steam generator nozzle inside containment (Elevation 144 feet) vertically downward and then out of the containment through a separate penetration which also forms an anchor point in the system. Two feed lines penetrate the containment into an enclosed penetration area on the south side. Each line then turns toward the turbine (east) in a horizontal run. Downstream of the feedwater check valves, each line then rises to Elevation 130 feet (or 132 feet) before exiting the penetration area. The design philosophy and conservatism utilized in the design of the main steam piping (refer to Section 3. 6. 5. 2) has also been utilized for the feedwater piping. The increased wall thickness of the piping within the penetration sleeves (2.065 inches) and up to the isolation valves (Schedule 120) assures elimination of high stresses at the anchor, and preservation of containment integrity. Extensive restraint steel has been provided for the feedwater lines, as was done for the steam lines. The entire Feedwater System from the feed pumps to the steam generators has been designed in accordance with ANSI B31. 1. The piping was purchased in accordance with ASTM Specification A106 Grade B, A106 Grade C, A335 Grade P22 or substitutable stainless steel material, depending upon the specific piece in question. Mill inspection and test reports were obtained for this feedwater piping, and a check analysis and flattening test of one length of each pipe lot was required. For sections of this piping which are 8 inches nominal size and larger, supplementary etching tests were required. Welded joints in main headers were made using the consumable insert tungsten inert gas first pass technique. Welded joints for piping replaced under the Steam Generator Replacement Project were made using the open root, manual gas tungsten arc process for the root pass, and the automatic gas tungsten arc process for the balance of the weld. 3.6-30 SGS-UFSAR Revision 27 November 25, 2013

Radiographic examination or Ultrasonic Testing has been performed on welded joints in piping of 3/4-inch wall thickness and over, and nozzle welds have been subjected to magnetic particle examination. Additionally, piping from the steam generator inlet feedwater control valves to the feedwater isolation valves has been constructed in accordance with Section I of the ASME Boiler and Pressure Vessel Code, and piping from the feedwater isolation valves to the steam generators has been constructed in accordance with the materials, fabrication, fabrication inspection, and quality control of ANSI B31.7 for Nuclear Class 1 piping. To eliminate potential cracking and mitigate erosion corrosion concerns in the Unit 1 steam generator inlet feedwater piping, steam generator feedwater nozzle-piping transition forging pieces have been installed. The material for these forgings was purchased in accordance with ASME Section II, Part A, Specification SA-508, Class 2 quenched and tempered low alloy forging material. This forging material meets all ASME Section III, Class 1 fracture toughness requirements. Although installed in feedwater piping classified as Class 2 piping, these forging pieces are fabricated, inspected and installed to Nuclear Class 1 requirements. Additionally, these transition forging pieces are analyzed to meet the stress limits and requirements for cyclic operation of the ASME Code Section III, Sub-article NB-3200, for Class 1 piping. Feedwater piping installation welding, and examination involved in installing the Unit 2 replacement Steam Generators utilized ASME Section XI (1998 Edition with 2000 Addenda) and ASME Section III, Subsection NC (1995 Edition with 1996 Addenda). Both of these later codes are NRC-endorsed per 10CFR50.55a and were reconciled to the original construction codes. 3.6.5.3.1 Criteria for Determination of Location and Orientation of Postulated Break The Reference 1 criteria are not totally applicable to the feedwater piping for the same reasons discussed in Section 3.6.5.2.1 for the main steam piping. In order to provide a degree of protection equivalent to the recommended Reference 1 criteria, however, the following were implemented:

1. The wall thickness of the feedwater piping within the penetration areas and in the portion that runs adjacent to the Auxiliary Building (Control Room) wall were increased to Schedule 120.

3.6-31 SGS-UFSAR Revision 24 May 11, 2009

2. Stress in the feedwater piping runs described in (1) above was limited to 0.25 (Sh + ) for the normal operational plus OBE (1/2 SSE) loading cases.

3. Circumferential and longitudinal breaks were postulated at high stress locations in the feedwater line (points C, D, and F on Figure 3.6-10).

4. Critical cracks (as defined in Reference 1) were postulated in the feedwater line both inside and outside the penetration areas.

3.6.5.3.2 of Postulated Pipe Breaks As stated in Section 3.6.5.3.1 feedwater piping breaks were postulated at points C, D, and F on Figure 3. 6-10 in both the longitudinal and circumferential directions. The forcing function for design of restraints was assumed to be a step function of magnitude 2.0 PA, where P and A are as defined in Section 3. 6. 5. 2. 2. The 2. 0 thrust factor is theoretically derivable for nonflashing liquids, and is known to be conservative for saturated and subcooled liquids, whose thrust factor is closer to 1.20. Dynamic load factors were used where The consequences of feedwater pipe breaks are identical to those for steam line breaks in Section 3. 6. 5. 2. 2. Pipe whip restraints similar to those provided for the main steam lines were provided for the feedwater piping to prevent damage to the Auxiliary Building and penetration area walls discussed in Section 3.6.5.2.2. These restraints are shown on Figure 3.6-19. The discussion presented in Section 3. 6. 5. 2. 2 regarding the design of the restraints, fluid impingement effects, containment and is also to the feedwater line breaks. In addition, the Feedwater (FW) lines of the Salem steam generators are subjected to postulated guillotine pipe breaks. The pipe break points are near nozzle weld sections. The time history analysis similar to that described in Section 3.6.5.2.2 for the main steam line breaks is performed to determine the pipe whip restraint reaction forces and the pipe line displacements. 3.6.5.3.3 Effects of Postulated Breaks on Equipment The discussion in Section 3.6.5.2.3 also to the Steam Generator Feedwater 3.6-32 SGS-OFSAR Revision 24 May 11, 2009

3.6.5.3.4 Analysis of Postulated Break on Unit S~utdown Capabqity Analysis of postulated main feedwater line breaks outside of containment demonstrates that the required equipment for safe shutdown of the reactor will be available and not damaged by such breaks. The postulated event will require automatic and manual operations in accordance with plant emergency operating procedures to bring the unit to a shutdown condition and then subsequent cool down. No significant effects result from the postulated breaks that would prevent cold shutdown of the reactor and protection of the core. A break in any of the four main feedwater lines results in loss of feedwater to the affected steam generator and causes the water level in the steam generator to drop rapidly. The three intact steam generators will also see a drop in water level. A reactor trip is initiated upon receipt of 2 out of 3 low water level signals from any steam generator or a coincident low feedwater flow signal (one out of two) and low steam generator water level signal (one out of two) from any steam generator. As the affected steam generator empties, the SIS will be actuated by steam line differential pressure signals between the steam line of the affected steam generator and the remaining steam lines. The safety injection signal will also trip the main feedwater pumps, initiate feedwater isolation, and start the auxiliary feedwater pumps. These automatic actions will shut down the reactor and mitigate the consequences of the feedwater line break. Once the operating personnel have verified that the preceding automatic actions have occurred, they will then initiate recovery operations. Recovery will be essentially the same as that described for the postulated steam line break in Section 3.6.5.2.4 and includes: 3.6-33 SGS-UFSAR Revision 6 February 15, 1987

1. Isolating the affected steam generator by closure of the steam line, feedwater line, and blowdown line isolation valves.

2. Directing auxiliary feedwater flow to the unaffected steam generators and regulating flow to maintain an indicated water level.

3. Stopping the safety injection pumps once pressurizer level is restored.

4. Stabilizing reactor coolant temperature and steam generator pressure and level in the remaining steam generators by steam dump to the condensers or through the atmospheric relief valves.

Boration of the RCS as necessary for cold shutdown and subsequent cooldown will proceed according to plant operating procedures as described for the postulated steam line break in Section 3.6.5.2.4. Postulated feedwater line breaks outside of the containment will not affect redundant safety-related control and electrical equipment necessary to bring the reactor to a cold shutdown condition. The postulated events will not affect remote operation of the required equipment from the control room nor access to the control room. The capability to bring the unit to a cold shutdown condition is not jeopardized by the postulated main feedwater line ruptures. 3.6.5.4 Chemical and Volume Control Letdown Line The letdown line of the Chemical and Volume Control System (CVCS) is described in Section 9.3 and is shown on Plant Drawings 205228 and 205328. An isometric view of the piping arrangement is provided in Figure 3.6-20. The letdown line carries 300 psig, 300°F reactor coolant from the regenerative heat exchanger inside the containment through a containment penetration in the Elevation 78 foot piping penetration area. The pipe runs across the penetration area into the Elevation 84 foot pipe alley, where it bifurcates. One branch goes to normally closed valve 1CV8 in the safety injection pump room, while the other branch runs along the pipe alley wall until it turns into the letdown heat exchanger room. After it exits the letdown heat exchanger, it is no longer considered to be a high energy line. 3.6-34 SGS-UFSAR Revision 27 November 25, 2013

The letdown line is constructed to Nuclear Class 2, Seismic Class II standards, from ASTM-A312 TP 304 stainless steel pipe. Postulated break locations have been sleeved and restrained in accordance with the design criteria of Section 3.6.5.11. The section of piping located in the safety injection pump room has been totally encapsulated, as shown on Figure 3.6-21. The pipe alley and letdown heat exchanger room are fitted with watertight steel doors to contain the energy released by a postulated break. The letdown heat exchanger room is vented through the normal exhaust ducting. The pipe alley is vented to the mechanical penetration area and then through a vent penthouse to the atmosphere. With these modifications, postulated ruptures of this piping will have no detrimental effect on plant shutdown capability or safety systems. Emergency procedures in the event of rupture of this line consist of (a) identification of the break, and (b) remote manual isolation of the line. 3.6.5.5 Steam Generator Blowdown System The Steam Generator Blowdown System is described in Section 10.4 and is shown on Plant Drawings 205225 and 205325. The piping arrangement is shown on Figures 3.6-22 and 3.6-23. The steam generator blowdown piping which is encompassed by the definition of high energy lines contains 814 psig (Unit 1) and 885* psig (Unit 2) saturated water from the secondary side of the steam generators. The four 3-inch pipes, one from each steam generator, penetrate the containment wall into the Elevation 78 foot piping penetration area, then run immediately upward into the Elevation 100 foot piping penetration area. Here, each of the lines bifurcates with one branch of each line going to each of the No. 11 and 12 steam generator blowdown tanks. The steam generator blowdown lines are constructed to Nuclear Class 2, Seismic Class I standards upstream of the containment isolation valves, and non-nuclear, Seismic Class III standards downstream of the isolation valves. Postulated break locations have been restrained to prevent pipe whip damage to vi tal equipment. With these modifications, postulated ruptures will have no detrimental effect on plant shutdown capability or safety systems. Emergency procedures in the event of a rupture of these lines consist of (a) identification of the break, and (b) remote manual isolation of the line. Primary flow 82,500 gpm, no fouling, 0% tube plugging, Tave 577.9°F, full power 867.75 MWt. 3.6-35 SGS-UFSAR Revision 27 November 25, 2013

Modifications to the blowdown piping were performed in order to replace the Unit 2 steam generators in 2008. With regard to NRC Generic Letter 87-11 and NRC Branch Technical Position MEB 3-1 (Reference 4, as cited in UFSAR Section 3.6.6), the relocation of the blowdown nozzles on the RSGs resulted in the terminal ends of the blowdown piping being correspondingly relocated. However, since no safety related equipment is located in the vicinity of the relocated RSG blowdown nozzles, no rupture restraints are required for the terminal end breaks of the rerouted blowdown piping. The new blowdown pipe routing has been analyzed for intermediate breaks. 3.6.5.6 Steam Supply To The Auxiliary Feedwater Pump Turbine The steam lines to the auxiliary feedwater pump turbine are described in Section 10.3 and shown on Plant Drawings 205203 and 205303. An isometric view of the piping is shown on Figure 3.6-24. The two pipes carry 814 psig (Unit 1) and 885* psig (Unit 2) saturated main steam from the main steam lines in the Elevation 100 foot piping penetration area along the Auxiliary Building wall, joining together into one line before entering the Auxiliary Building. The steam line enters the Auxiliary Building in the waste evaporator room on Elevation 100 foot. It then passes down into the pipe alley and enters the auxiliary feedwater pump turbine room on Elevation 84 feet. The steam lines are constructed to Nuclear Class 2, Seismic Class I standards from ASTM-A106 Grade B carbon steel pipe or substitutable chrome alloy or stainless steel material. The piping was subjected to Supplementary S2 (check analysis), S4 (flattening test on each length of pipe), and S5 - S6 (etching test on each length of pipe) . Welds are made utilizing the consumable insert - tungsten inert gas first pass technique. The pipe alley and auxiliary feedwater pump turbine room are provided with watertight steel doors to contain the energy released by a postulated break. The compartments are connected by two blowout panels to allow venting through a common discharge path. Postulated break locations shown on Figure 3.6-24 have been sleeved and restrained to limit pipe break mass flow rates and preclude pipe whip damage to vital equipment. Typical sleeve and restraint designs for this line are shown on Figure 3.6-25. Primary flow 82,500 gpm, no fouling, 0% tube plugging, Tave power 867.75 MWt. 3.6-36 SGS-UFSAR Revision 27 November 25, 2013

Emergency procedures in the event of rupture of this line consists of (a) identification of the break, (b) manual isolation of the affected line(s) if the valves are accessible, and normal plant shutdown, irrespective of the results of (b). With these modifications, the only anticipated consequence is the possible loss of one of three auxiliary feedwater pumps, which does not affect unit shutdown capability. 3.6.5.7 Chemical And Volume Control Char&ing And Reactor Coolant Pump Seal Injection - Unit 2 Only Piping for the above system is located in the following Auxiliary Building areas:

1. Primary, Auxiliary Feed, and Refueling Water Tank Heater Area (Elevation 84 feet)

2. South Penetration Area (Elevation 78 feet)

3. Pip~ Alley (Elevation 84 feet)

4. Safety Injection Pump Room (Elevation 84 feet)

5. Access Corridor (Elevation 84 feet)

6. Aisle No. 2 {Elevation 84 feet)

7. Access Corridor (Elevation 100 feet)

8. Boric Acid Evaporator Room (Elevation 100 feet)

9. Waste Evaporator Room (Elevation 100 feet)

10. Refueling Yater Purification Filter Room (Elevation 100 feet) 3.6-37 SGS-UFSAR Revision 7 July 22, 1987

11. Boric Acid Transfer Pump Room (Elevation 100 feet)

12. Emergency Diesel Rooms (Elevation 100 feet)

13. Control Room Air Conditioning Equipment Room (Elevation 122 feet)

14. Control Equipment Room (Elevation 122 feet)

15. Auxiliary Building Air Conditioning (Elevation 122 feet)

16. Boric Acid Batch Tank Area (Elevation 122 feet)

Normal pressure for the charging and seal injection lines is approximately 2520 psig at 127F. Design pressure for the piping is 2825 psig. The system is constructed to Nuclear Class 2, Seismic Class I standards with austenitic Type 316, Schedule 160 material. In the unlikely event of a postulated failure of the piping, water jet impingement or pipe whip could adversely affect some safety-related components in the Elevation 84 foot pipe alley and south penetration area at Elevation 78 feet. Damage potential from jet impingement has been eliminated by use of pipe shrouds or impingement baffles at required points along the path of the piping. Damage potential from pipe whip has been resolved by the use of pipe restraints. At the first refueling outage, 33 pipe whip restraints were installed inside the containment and 21 pipe whip restraints were installed outside the containment. 3.6.5.8 Heating Steam System - Unit 2 Only Piping for the Steam Heating System is located in the same areas as described in Section 3.6.5.7. Steam for the system is supplied from an auxiliary boiler in the yard or from main turbine extractions. The heating steam system is considered to be a high energy piping system for Unit 2 but not for Unit 1 because the design pressure and temperature are 185 psig and 382 °F. Normal pressure in the system varies between 150 psig and 50 psig at saturated or slightly superheated conditions. Its purpose in the Auxiliary Building is to supply a heating medium for the following equipment. 3.6-38 SGS-UFSAR Revision 29 January 30, 2017

1. Boric Acid Evaporator and Feed Preheater

2. Waste

3. Boric Acid Batching Tank Heating steam piping is constructed in accordance with the ANSI B31.l, Power Piping Code. Material used is Schedule 40 carbon steel or substitutable chrome alloy or stainless steel material.

The 316 stainless steel flex hose connections between the enclosed encapsulations and the vent piping are classified as related and Seismic Category I. The heating steam system encapsulations are rated for 174 psig and 378 °F design conditions. In the unlikely event of a postulated failure of this piping, ambient temperatures above normal, caused by steam escape, could possibly adversely affect: controls and electrical equipment in the of the break. To preclude this the following modifications were implemented:

1. Provided enclosed of the at selected of postulated failure in order to restrict steam flooding of the 84 ft elevation corridor and adjacent switchgear rooms.

2. The enclosed encapsulations are connected via flex hose connections to 1-1/2 inch piping that will vent any steam escaping from the points of postulated failure. The vent piping exhausts to atmosphere outside of the Building on the roof of the Main Steam whip restraints, sleeves, and a 1 1/2 inch vent line in the Auxiliary Building were installed at the first outage.

3.6.5.9 Hea1~Q~Water System- Unit 2 Only Piping for the Heating Water System is located in the same areas as described in Section 3.6.5.7. The Heating Water System provides the heating medium for the purposes: 3.6-39 SGS-UFSAR Revision 25 October 26, 2010

1. Room space heating

2. Primary, Auxiliary Feedwater, and Refueling Water Storage Tank Heating

3. Control Room and Auxiliary Building Vent Duct Heating This system operates as a pumped closed loop with each of the users having both a supply and return line. Piping is constructed in accordance with ANSI B31.1, Power Piping Code. Material used is Schedule 40 carbon steel or substitutable chrome alloy or stainless steel material.

The heating water system is considered a high energy system for Unit 2 but not for Unit 1 because the design pressure and temperature for the system are 125 psig and 300 °F for the system supply side and 125 psig and 200 °F for the system return side. The normal pressure and temperature are the same as the design pressure and temperature ratings. Unlikely postulated failures of this piping present conditions similar to those specified for the Heating Steam System, except that the quanti ties of steam involved are less. To preclude the possibility of adverse effects of failures in this piping, enclosed vented encapsulations at selected points of postulated failure have been provided, the same as for the Heating Steam System. Pipe whip restraints, encapsulation sleeves, and vent lines in the Auxiliary Building were installed at the first refueling outage. Design basis cracks were not postulated at arbitrary locations in the Heating Steam and Heating Water Systems. Instead, specific failure points were determined based on maximum stress locations for which vented to atmosphere, enclosed encapsulations were provided to preclude steam flooding in the Auxiliary Building corridor and adjacent switchgear rooms. 3.6.5.10 Additional Protection Against Steam Flooding in the Auxiliary Building As shown on Plant Drawings 204805 and 204804, of items 3 thru 7 in Section 3.6.1, the following are plant areas where rupture of the piping could conceivably be undesirable:

1. Piping penetration area (Elevation 100 feet)

2. Piping penetration area (Elevation 78 feet) 3.6-40 SGS-UFSAR Revision 27 November 25, 2013

3. Waste evaporator room (Elevation 100 feet)

4. Pipe alley (Elevation 84 feet)

5. Safety injection pump room (Elevation 84 feet)

6. Auxiliary feedpump turbine room (Elevation 84 feet)

1. Letdown heat exchanger room (Elevation 84 feet)

In order to preclude undeairable effects due to steam flooding, these seven areas were transformed into one environmentally isolated contiguous zone which is vented to the atmosphere via the penetration area pipe break vent penthouse. 'l'he bounds of this zone are ahown by shading on Figure 3.6-28. Normal ventilation of this zone is provided by ducting with backdraft-type dampers. These dampers are designed such that those pipe break induced steam flows which are significantly in excess of normal ventilation airflows will close the dampers, thus preventing steam flow into adjacent vital areas either through supply or exhaust ducting. The closure of the backdraft dampers will then force the pipe break induced steam flow to be vented to the atmosphere via the penetration area vent penthouse. Where steam flows were found to be in excess of the capabilities of available vent area, sleeving of selected break locations was employed to reduce pipe break mass flow rates to acceptable values. Normal and post-pipe break flow paths, as well as backclraft damper locations are shown on Figure 3.6-28. The Letdown Beat Exchanger Room, located on Elevation 84 feet, contains a portion of the high energy eves letdown line which ia described in Section 3.6.5.4. Pipe break locations, selected on the basis of the criteria in section 3.6.4.3.2, are sleeved and restrained in order to prevent whip and to limit the two phase water/steam mass flow rates from the postulated breaks. A leaktight door with an integral backdraft damper was added to the room in 3.6-41 SGS-UFSAR Revision 16 January 31, 1998

order to provide an inlet path for ventilation supply, but still prevent break induced steam flow from reaching vital equipment on Elevation 84 feat. calculation of steam and water mass flow rates showed that the normal ventilation exhaust ducting in the room ia adequate to handle the steam without significantly disrupting ventilation system balance, and that the floor drains already provided are adequate to prevent excessive water buildup in the room. Cable in the room was rerouted in conduit to protect vital cables from direct impingement and to permit leaktight seals at wall penetrations. The Auxiliary Faedwatar Pump Room, located on Elevation 84 feet, contains a portion of the high energy steam line to the No. 13 auxiliary feedwater pump turbine described in Section 3.6.5.6. Pipe break locations, selected on one basis of the criteria presented in Section 3.6.4.3.2, were sleeved and restrained in order to prevent pipe whip and limit the steam mass flow rates from the postulated breaks. In order to preclude steam induced environmental damage to the No. 11 and 12 auxiliary feedwater pump motors and adjacent vital control centers, the No. 13 auxiliary feedpump and its turbine were enclosed in a steel plate subroom with a drop ceiling as shown on Figure 3.6-29. The steam line was lowered several inches in order to facilitate installation of the new ceiling below presently installed vital piping and equipment. Normal ventilation and cooling of the subroom is accomplished through the Auxiliary Building Ventilation system and a pump room cooler. Two pressure relief panels in the wall opening located adjacent to the pipe alley provide a defined path to the atmosphere for steam flows resulting from postulated pipe ruptures. ( The Safety Injection Pump Room, located on Elevation 84 feet, contains a portion of the high energy eves letdown line which is described in Section 3.6.5.4. In order to protect adjacent safety-related piping from pipe whip impact and to protect the 3.6-42 SGS-UFSAR Revision 16 January 31, 1998

safety injection pump motors from the steam environment following a postulated rupture of this pipe, the pipe is entirely sleeved within this room. The sleeving design is shown on Figures 3.6-29 and 3.6-21. No modifications were required, other than the letdown line several inches to sleeve interference. The Waste Evaporator Room, located on Elevation 100 feet, contains a portion of the high energy steam line to the auxiliary feedwater pump turbine which is described in Section 3.6.5.6. Pipe break locations, selected on the basis of the criteria presented in Section 3. 6. 4. 3. 2, are sleeved and restrained in order to prevent whip and limit the steam mass flow rates from the postulated breaks. In order to preclude the detrimental effects of steam and jet a steel enclosure is provided around this where it passes through this room. A new vent opening in the floor of the room within the enclosure was provided to vent the steam flow from the breaks to the pipe alley below. These modifications are shown on Figures 3.6-28 and 3.6-29. The Pipe Alley, located on Elevation 84 feet, contains portions of both the high energy eves letdown lines and the steam line to the auxiliary feedwater pump turbine. Pipe break locations, selected on the basis of the criteria presented in Section 3.6.4.3.2, are restrained to prevent whip. Although the pipe does not contain any equipment, (or backdraft is to steam from entering any areas not specifically designed to accommodate the steam environment. A wall opening to the penetration area provides atmospheric venting via the penetration area vent penthouse. The Piping Penetration Areas (Elevations 78 feet and 100 feet) contain portions of all of the high energy systems listed in Section 3.6.1. Pipe break locations, selected on the basis of the criteria presented in Section 3.6.4.3.2 were restrained as necessary to prevent pipe whip and fluid impingement. In order to maintain the internal pressure of the areas within limits following the postulated main steam and feedwater system ruptures, the inboard and outboard penetration area venting panels were increased from 100 to 650 by raising their roofs to Elevation 141 feet. 3.6-43 SGS-UFSAR Revision 25 October 26, 2010

However, certain venting panels have obstructions and are not fully operational. Considering all obstructions, the net available venting areas are: 337 ft 2 for the inboard penetration area, and 391 ft 2 for the outboard penetration area. calculations have confirmed that the net available area maintains the internal pressure of the areas within previously established limits as discussed in section 3.6.5.2.2. Certain Elevation 100-foot and 78-foot area walls and floor sections were sealed in order to confine the environmental effects of the postulated breaks, and some localized impingement protection was added in order to accommodate the postulated breaks. 3.6.5.11 In some instances, of of the high energy systems was used to limit break mass flow rates and, hence, preclude any effects of pipe rupture. As stated in Section 3.6.5.10, where steam flows were found to be in excess of the capabilities of the available vent area, sleeving of selected break locations was employed to reduce pipe break mass flow rates to acceptable values. Encapsulation sleeving was installed in the Letdown Heat Exchanger Room, the Auxiliary Feedwater Pump Room, the Safety Injection Pump Room and the Waste Evaporator Room. In these cases, the encapsulation sleeves are designed and installed in accordance with the following criteria:

1. The sleeves were and in a manner which will not introduce significant strain concentrations on the encapsulated section of piping.

I 2. The piping beyond the encapsulation sleeves was restraints or anchors which restrict its axial displacement and motion provided with within the sleeves following a postulated circumferential pipe break.

3. The encapsulation sleeves were (a) to withstand the dynamic forces of internal from the escape of high energy fluid at the postulated pipe break location, assuming complete pipe severance and axial to the extent by the pipe restraints, and (b) to restrict the flow at the open ends of the sleeve to a level required to preclude compartment pressurization beyond the allowable 3.6-44 SGS-UFSAR Revision 25 October 26, 2010

structure design limits or beyond the of features to accommodate resultant environmental effects. Some of t:he encapsulations are sealed closed at both ends and are vented via piping to the roof outside of the Auxiliary Building.

4. The stresses imposed on the encapsulation sleeve during dynamic pressurization are limited to the design limits associated with "emergency condition" as permitted by ASME Section III, Nuclear Power Plant Components Code, for Class 2 Components.

5. The encapsulation sleeves were constructed in accordance with the current revision of the ANSI Standard Code for Pressure Piping, ANSI B31.1. Material inspection, fabrication, quality control, and applicable installation conformed with the requirements of the current revision of the ANSI Standard Code for Pressure Piping, Nuclear Power Piping, ANSI B31.7, for Class II piping, with the provision that each layer of the final assembly weld shall be nondestructively examined by surface examination techniques (i.e., liquid penetrant or magnetic

                ) rather than radiography.

6. The non-vented encapsulation sleeves are provided with open nipples, which extend beyond the pipe insulation as a means of monitoring the encapsulated pipe section for any leaks, which might develop in service.

7. The design of the encapsulation sleeves permits removal by machinery or flame cutting techniques or the replacement of the encapsulated pipe sections in the event leaks develop which require repair or replacement of the piping.

3.6.5.12 Moderate Energy Pipe Failure Evaluations- Unit 2 Only Moderate energy fluid systems outside of containment as defined in Section 3.6.5.12.1, have been evaluated for the consequences of

3. 6-45 SGS-OFSAR Revision 24 May 11, 2009

through-wall leakage cracks. Components required for the safe shutdown of the reactor were evaluated and have been provided, as necessary, with measures to ensure operability. 3.6.5.12.1 Definitions Moderate Energy Lines (MEL) Moderate energy piping includes those systems where both of the following conditions are met:

1. The maximum is 200°F or less

2. The maximum operating pressure is 275 psig or less Hazard For purposes of this evaluation, postulated leakage shall be considered for the effects of resulting flooding or liquid spray on components required for safe unit shutdown.

3.6.5.12.2 Postulated Break Location Moderate energy piping that is located in areas containing systems and components important to safety were postulated to develop a through-wall leakage crack at the most adverse location to determine protection needed to withstand the effects of the resulting liquid spray and floodircg. Moderate-energy piping that is located in areas that communicate, either through a door, curb, hatch, sleeve or drain, with those areas containing systems and components to were to a crack at the most adverse location to determine the if any, to withstand the effects of the resulting flooding. Piping systems that are isolated and physically separated from systems and components important to safe~y by plant arrangement and layout were not considered for postulated leakage cracks. Moderate energy piping that is located in the same area as high energy fluid systems considered for breaks was not considered for cracks. 3.6-46 SGS-UFSAR Revision 25 October 26, 2010

3.6.5.12.3 Postulated Crack Size Through-wall leakage cracks were postulated in runs and branches over 1-inch nominal size. Crack size was assumed to be the diameter in length by 1/2 the pipe wall thickness in width. 3.6.5.12.4 Evaluation Procedure A review of the Auxiliary Building was made to determine those compartments or areas with components required for safe reactor shutdown. Each of the compartments or areas containing safe shutdown equipment was evaluated to determine the effects of flood and spray from cracks within the area and the effects from from areas that could co~~unicate with it. Crack postulated flow rates of the largest MEL in the given space were l/2 estimated on the basis of the Bernoulli Q=KA (2gh) where K, the orifice coefficient, was assumed to be 0. 6. An accumulation rate or flood level was then estimated based on a comparison between floor and the postulated leakage rate. If liquid accu~ulation a flood threat to components within a compartment, an evaluation was made to determine the possibility of damaging each component and the acceptability of such damage. If necessary, modification to the existing design was appraised and performed to correct any conditions found to be adverse. Fluid spray consequences were evaluated on the basis of between components in the area or space and the pressure or most unfavorably oriented MEL. If it was determined that liquid spray in a given compartment or space could interact with componem:s within that space, evaluation was made to determine acceptability of such interaction and, if necessary, modification to exlsting design was determined to correct the condition. 3.6-47 SGS-UFSAR Revision 25 October 26, 2010

3.6.5.12.5 Inspection Results and Required Modifications Flooding in the RHR pump rooms can occur as a result of either MEL breaks inside or outside the rooms. Breaks in the pipe alley on Elevation 84 feet can communicate with RHR pump rooms on Elevation 45 feet via a pipe chase. Flooding to the RHR pump rooms could also occur as a result of MEL fluid from breaks on upper elevations running down staircases and conceivably into both RHR pump rooms. To prevent this from occurring, curbs were installed on Elevation 55 feet above the RHR pump rooms such that fluid flow from MEL failures on elevations above can only flow to one RHR pump room, no~ both rooms. Alarms in the control room from high RHR pump room sump level will alert the control room who in turn will terminate flow by remote or manual valve realignment. Water spray from component cool:.ng lines in the RHR pump rooms could affect safety-related equipment in those areas. A single postulated MEL failure however will only involve one of two redundant RHR pump trains because of the wall between the rooms. spray failures in this area do not j ze safe shutdown of the Piping located in this area includes an auxiliary feed suction line, a fire hose station, and a preaction sprinkler system. Preaction sprinkler systems replaced the o:d C02 fire suppression systems in the 4160 v and 460 V switchgear rooms and the electrical penetration area. The feed suction line is empty and is not considered for cracks. The fire hose station has been provided with a shroud to prevent spray on empty, is an electric/electric interlock system that uses a detection system and air in the piping. The of both a head and both a thermal and a smoke detector are required to open the deluge supply valve. The piping in the room is designed to Seismic I requirements and the sprinkler heads are also seismically qualified. In the event of a fire, one or two sprinkler heads are expected to open to provide water to U:e effected area. The room has two 4-inch floor drains that are piped to the sump tank on the 5l ft elevation. 3.6-48 SGS-UFSAR Revision 25 October 26, 2010

For the worst case scenario, six sprinkler heads are assumed to supply approximately 393 gpm to suppress the fire. Flow restrictors (orifices) were added to the two drains in the room to drain flow from ~he sump tank overflow line thus of other rooms on the 64 ft elevation due to backflow the interconnected floor drains. Area 3 - Electrical Penetration Area Elevation 78 Feet This area contains only the fire protection preaction sprinkler piping which is similar to tha~ installed in the 64 ft switchgear room that replaced the C02 fire suppression system. In a fire, the zed air that is in the line between the valve and the closed heads is released by the of a fusible link that opens the valve when other electric alarm have been received. When the deluge valve opens the dry preaction sprinkler piping is charged with water. The backpressure provided by the 20 psig air that is maintained in the dry preaction system piping helps prevent the occurrence of water hammer in the preaction systen when the deluge valve opens. Under the assumed worst case co:1di tions, the preaction system can supply approximately 382 gpm flowing to six sprinkler heads to suppress a fire. Because the 4-inch drain in the room is due to HELB considerations, a new 4-inch drain was added that to the RHR valve room on the 55 ft elevation. Check valves in the drain lines back£ low interaction with the two new drains added to 84 ft switchgear room that also empty to the RHR valve room. The drainage from the 84 ft switchgear room and the 78 ft elevation electrical penetration room combine and empty into the RHR valve room and from there drain to one RHR punp room on the 45 ft elevation. at access into these areas has been to prevent from adjacent areas. 'l'he 4 60 V room on the 84 ft elevation also the C02 fire system with an automatically interlocked, preaction sprinkler system that was designed and installed per the requirements in NFPA 13 (2002 Edition). The preaction sprinkler system ~n the 4 60 V switchgear room operated the same as those in the 4160 V S1*'l'i tcr.gear Room and in the Electrical Penetration Room. 3.6-49 SGS-UFSAR Revision 25 October 26, 2010

The fire protection piping within the 460 V switchgear room meets Seismic Category I requirements. The fire protection tie-ins to the 6-inch fire protection header and isolation valves located outside the room in the hallway on the 84 ft elevation meet Seismic II/I requirements. There are two existing 4-inch floor drains in the hallway that drain to the waste holdup tanks on the 64 ft elevation. Two new 4-inch drains were added to remove water from the switchgear room should the sprinkler system discharge in the event of a fire. Assuming the worst-case scenario, six sprinkler heads in a tight area grouping would discharge 379 gpm in the event of a fire in the room. The drain lines, which are open-ended and equipped with check valves to prevent backup, empty to the RHR valve room on the 55 ft elevation. From there, the water drains via an existing floor drain to a RHR pump room sump pit on the 45 ft elevation. With flooding of one RHR pump room that could potentially incapacitate one RHR pump, the other RHR pump located in the adjacent, separate, non-flooded RHR pump room would be available. No other design basis accidents are postulated to occur coincidental with a fire. However, switchgear room fires may result in the loss of both onsite and offsite power to the vital buses, which could result in a total loss of the RHR system, making it temporarily unavailable for providing decay heat removal. Safe shutdown for Salem is defined as hot standby. Existing procedures identify repairs to the components that are required for establishing one RHR loop as necessary for achieving cold safe shutdown. In addition, Appendix R provides for remote cabling of either RHR pump in the event of a fire. However, the maximum flood level in the RHR pump room 30 minutes after actuation of one of the preaction sprinkler systems added to the switchgear rooms and the electrical penetration room is calculated to be less than 13 inches, which is well below the elevation of an RHR pump. Area 7 - Safety Injection Pump Room - Elevation 84 Feet This area contains MEL piping. Floor drainage capacity however is adequate to prevent flooding of the compartment. Water spray from service water or demineralized water piping could affect safety injection pump motors. The safety injection pump motors have been protected from overhead spray by means of a protective shroud. Area 8 - Component Cooling Heating Exchanger Rooms -Elevation 84 Feet This area contains service water and fire protection MEL piping. Floor drainage capacity in the area is adequate to prevent flooding. Water spray from service water pipe cracks could affect 22 or 23 component cooling pump motors and associated controls, dependent on the crack location. 3.6-50 SGS-UFSAR Revision 31 December 5, 2019

Component cooling pump 21, however, is isolated in another compartment and would not be affected by this fault. Therefore, spray failures in this area do not affect safe shutdown of the Area 9 - Auxiliary Feed Pump Room - Elevation 84 Feet This area contains service water, fire protection, component cooling, demineralized water, and refueling water storage tank piping. Floor drain capacity in the area as well as drain capacity in the corridor to which this area is open, is adequate to prevent local flooding. Water spray from the MEL piping in the area can affect safety- related motor control centers 2C west and 2A west as well as control panels 205, 206, 2C7, and 213. In addition, water spray can affect the auxiliary feed pump motors. To prevent water spray to these vital components, motor control centers and control have been to withstand the effects of spray. feed pump motors have been with a shroud to prevent spray These rooms contain fuel oil MEL piping. from fuel oil cracks could affect fuel oil transfer pump motors. However, 21 and 22 fuel oil transfer pumps are physically isolated from each other and local MEL failures in one room will not affect the other room. No modifications are required. The 10 Ten C02 Room, the Diesel Fuel Storage Tank areas, and the entrance to these areas contain service water, fuel oil, and carbon dioxide fire protection MEL piping in addition to diesel fuel oil storage and transfer system equipment. The areas do not contain floor drainage; hence service water or fuel oil piping failure could cause flooding of the areas and the adjoining fuel transfer pump rooms. Therefore, flooding of diesel fuel oil storage and transfer system equipment can occur. Spray damage can also occur. However, because loss of off-site power coincident with service water or fuel oil piping failure is not a design basis condition at Salem Generating Station, diesel fuel oil storage and transfer equipment are not required to function to mitigate the consequences of such piping failure. Protection against the effects of service water or fuel oil piping failure is not required for diesel fuel oil storage and transfer system equipment. 3.6-51 SGS-UFSAR Revision 25 October 26, 2010

This area contains service water, component cooling, and spent fuel cooling MEL piping. is adequate to prevent flooding in the area. Portions of the MEL piping have been provided with a baffle to prevent spray damage to the charging pump motors. Area 12 - Containment Spray Pump Area - Elevation 84 Feet This area contains refueling water storage MEL piping. Drainage in the area is adequate to prevent flooding. Shrouds over containment spray pump mo~ors have been provided to prevent water spray damage from MEL piping. This area does not contain any MEL piping. No modifications are required. Area 14 - Electrical Penetration Area - Elevation 100 Feet This area contains service water MEL piping to the chiller condensers. Drainage capacity is adequate to prevent flooding from postulated cracks. Water spray in this area does not any safe shutdown hazards. No modifications are Area 15 - Emergency Diesel Rooms - Elevation 100 Feet These rooms contain service water, demineralized water, and fuel oil MEL piping. The rooms do not contain floor drainage, hence a postulated service water line failure could conceivably cause flooding in a single room. The individual emergency diesel engines however, are physically isolated from each other and hence local MEL failures in one room will not affect the other rooms. No modifications are in this area. Area 16 Control Room Elevation 122 Feet This area contains service water, chilled water, and heating wa~er piping. A pressure tight steel enclosure is provided to encase all the piping. A drainage path is provided to remove any liquid from the enclosure. This area contains chilled water and service water MEL piping. in the area is to prevent flooding. Water spray does not present any safe shutdown hazards. No modifications are required in this area. 3.6-52 SGS-OFSAR Revision 25 October 26, 2010

Postulated of piping in this area is discussed in Section 9. 2. No modifications are in this area. Switchgear Room A 12-inch Class I (Seismic) demineralized water line passes through the switchgear room at Elevation 64 feet. This line is a nonessential backup water supply to the Auxiliary Feedwater System and will remain dry during normal plant operation, thereby precluding any potential for accidents after intrusion into vital electrical areas. Equipment arrangement and floor drainage systems are adequate to prevent flooding serious enough to impair the operation of equipment necessary for safe shutdown. In the event of any Class I (Seismic) line failure of "critical crack" size on floor elevations 84 ft and above, the discharged effluent would likely spread out over a large floor area and be carried away via existing floor drains to the waste hold-up tanks or the RHR pump room sumps. Floor drains on the 64 ft elevation are piped to the sump tank in each Unit and from there are either pumped to the waste holdup tanks or overflow to an RHR pump room sump. In addition, from MEL cracks on the 64 ft elevation and below can gravity drain via stairwells and pipe chases to an RHR pump room sump. The RHR sumps each have two Class lE powered pumps that discharge to the waste hold up tanks. Curbs are installed between the RHR valve rooms on the 55 ft elevation to prevent fluid flow from MEL cracks on elevations above from flowing into both RHR pump rooms. RHR sump level alarms would alert the to take the necessary action to maintain the in a safe condition. To minimize effluent from tanks into the Building, a "critical crack" in a Class I (Seismic) pipe, steps could be taken in the yard area to reduce tank inventories or divert the inventory to other storage facilities. 3.6.5.12.6 Additional Modifications In addition to the installation of the required modifications identified in Section 3.6.5.12.5, the floor in the switchgear rooms between Elevation 100 feet and 84 feet and 84 feet 3.6-53 SGS-UFSAR Revision 25 October 26, 2010

and 64 feet were sealed so that liquid blowdown would be directed to the floor drains. 3.6.5.13 Section 1 contains elevation and arrangement drawings which show the Containment, Auxiliary, and Fuel Handling Buildings. The Turbine Building, designed for Class III (seismic) requirements, is a separate building from all Class I (seismic) structures. The has a membrane which will tend to any from piping, tanks, or equipment within the building confines. Failure of any piping or tanks within the Turbine Building would not hamper the operation of any safeguards systems. Investigations have been made as to the capability of operation of related equipment in the event of a of non-seismic Class I tanks which could cause extensive flooding in the A seismic verification using the SQUG GIP methodology of the large non-seismic tanks in the Auxiliary Building determined that none were likely to rupture due to a seismic event. The sudden catastrophic failure of these tanks that are constructed to industrial standards is not credible. Many of the large tanks are contained within berms to prevent the spread of their contents over a broad area in the Auxiliary Building should operator error cause them to be drained. Those without berms have wide bases and will not over a seismic event. The tanks in the Building, except for error, are not considered credible sources of flooding. The study indicated that rupture of any tank in these buildings would not interfere with operation of the reactor or safe shutdown of the reactor. Floor drains are provided in the vicinity of all tanks. The floor drains in the vicinity of tanks having a volume of 1,000 gallons, or less, are more than to provide in the event of rupture of a tank. 3.6-54 SGS-UFSAR Revision 25 October 26, 2010

In the unlikely event that one of the larger tanks without a dike or berm were to drain its contents, most likely due to operator error, the resulting flood would spread out over an extensive floor area in the Auxiliary Building which would limit the flood height and preclude damaging safe shutdown equipment. The waste holdup, waste monitor-holdup, and evaporator bottoms storage tanks are diked to contain the volumes within the tank area. The monitor tanks are not diked, but the failure of any of these tanks would not cause flooding serious enough to prevent Class I (seismic) safety-related equipment from operating satisfactorily. Aside from floor drainage systems, stairwells, and floor openings would prevent water from rising to levels that could be termed critical. A similar investigation showed that equipment arrangement and floor drainage systems design are adequate to prevent flooding in the event of a non-Class I (seismic) pipe rupture serious enough to prevent safeguards systems from operating satisfactorily. Fire Protection pipe systems have been demonstrated to be adequately supported to withstand seismic events without structural pipe failure. Nitrogen and hydrogen storage cylinders are located in the Auxiliary Building. Ruptures will not jeopardize the required operation of a Class I (seismic) system, since the tanks, located at Elevation 122 feet in corridors to the north and south of the drumming and baling area, are isolated from Class I (seismic) equipment by virtue of their location, as well as by concrete walls. Supplementing the Public Service Electric & Gas (PSE&G) letter of November 2, 1972 (response to Mr. R. C. DeYoung's letter of September 26, 1972), the failure of carbon dioxide fire protection equipment will not affect operation of safeguards systems. Manual systems are provided in the diesel-generator areas. Manual carbon dioxide fire protection equipment is provided in the control and relay room areas. 3.6-55 SGS-UFSAR Revision 31 December 5, 2019

3.6.5.14 All electrical cable types which are used for equipment in areas subject to adverse environmental conditions from pipe ruptures have been qualified for continued operation in these environments. Qualification tests consisted of exposure of the cable samples to thermal aging (e.g. 250°F for 7 days) radiation exposure (e.g. 100 x 1 R equivalent air dose with a Co 60 source) , and cyclic steam and chemical spray (e.g. 34 0°F, 105 psig steam, Boric Acid and Sodium Hydroxide, cycled for 14 ). after exposure showed no significant detrimental in insulation insulation dielectric breakdown capability, or cable and parameters. 3.6.6 References for Section 3.6

1. Letter, A. Giambusso (AEC) to F. W. Schneider (PSE&G), dated December 18, 1972, with attachment "General Information Required for Consideration of the Effects of a System Break Outside "Letter, D. B. Vassallo (AEC) to F. W. Schneider (PSE&G),

dated January 31, 1973, with attachment "Errata Sheet for 'General Information for Consideration of the Effects of a Piping System Break Outside Containment.'"

2. A. A., "The Phenomena of Rupture and Flow Solids,"

Philosophic Transactions of the Royal Society of London, Vol. 221, pp. 163-198, 1920.

3. Letter, R. C. De Young (AEC) to F. W. Schneider (PSE&G), dated May 21, 1973.

4. Branch Technical Position MEB 3-1, "Postulated Locations ir.

Fluid System Piping Inside and Outside Containment," attached to SRP Section 3.6.2, Rev. 2, June 1987.

5. Branch Technical Position SPLB 3-1, "Protection Postulated Piping Failures in Fluid Systems Outside Cor.tainment," attached to SRP Section 3.6.1, Rev. 2, October 1990.

6. Letter from Mr. James C. Stone, NRC, to Mr. Steven E. Miltenberger, PSE&G, dated May 25, 1994, "Leak-Before-Break Evaluation of Primary Loop Piping, Salem Nuclear Generating Station, Units 1 and 2 11

7. EPRI TR-1006937 "Extension of the EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs," April 4, 2002.

3.6-56 SGS-UFSAR Revision 25 October 26, 2010

8. WCAP-13659, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Salem Generating Station Units 1 and 2, May, 1993. [Proprietary]

9. Letter from Mr. James S. Kim, NRC, to Mr. Eric Carr, PSEG, dated February 23, 2021, Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 336 and 317 Re: Leak-Before-Break for Accumulator, Residual Heat Removal. Safety Injection, and Pressurizer Surge Lines (EPID L-2020-LLA-0088)

10. WCAP-18248, Revision 0, Technical Justification for Eliminating Safety Injection Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology

11. WCAP-18249, Revision 0, Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology

12. WCAP-18253, Revision 0, Technical Justification for Eliminating Residual Heat Removal Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology

13. WCAP-18261, Revision 0, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology

14. WCAP-18516, Revision 1, Fatigue Crack Growth Evaluations of Salem Units 1 and 2 Accumulator, RHR, Pressurizer Surge and Safety Injection Lines Supporting Expanded Scope Leak-Before-Break

15. Letter from Charles V. McFeaters, PSEG, to NRC, dated April 24, 2020, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology 3.6-57 SGS-UFSAR Revision 33 October 24, 2022

TABLE 3.6-1 POSTULATED REACTOR COOLANT SYSTEM PIPE RUPTURES Designation Description 1

1. ACCUM DEC( ) in accumulator line at RCL nozzle

2. RHR DEC in RHR line at RCL nozzle

3. PSL DEC in pressurizer surge line at RCL nozzle 1

Notes: ( )DEC: double-ended circumferential rupture

2. See Section 3.6.4.2.1 and reference to License Amendment Nos. 336 and 317 for updated LBB piping segments, excluding the dynamic effects of postulated pipe rupture in specific portions of systems attached to the Reactor Coolant System.

1 of 1 SGS-UFSAR Revision 33 October 24, 2022

TABLE 3.6-2 MAIN STEAM PIPE STRESS

SUMMARY

No. 13 Main Steam Line (1) 0.8(1.8 Sh + Sa) (3) foint 121 St£@§!,3J AlJ,owabJ,e RATIO 1 6657 46200 2 0.14 14604 46200 0.32 3 13403 46200 4 0.29 15244 46200 0.33 s 17436 46200 0.38 13 14826 46200 14 0.32 14626 46200 0.32 21 15017 46200 22 0.33 13807 46200 0.30 23 11035 46200 27 0.24 11678 46200 0.25 28 18513 46200 32 0.40 11683 46200 0.25 39 12866 46200 40 0.28 12419 46200 0.27 41 9295 46200 45 (Anchor) 0.20 3384 46200 0.07 Notes: (1) Stresses tabulated for Nos. 11, 12 and 14 steam lines at geometrically similar locations are of similar magnitude. (2) Stress point locations are shown on Figure 3.6-10. (3) Stresses and allowables from calculation 1SC-110. Minor changes in pipe stresses due to revisions of associated pipe stress calculations will be documented within 1SC-110 and will not require update of this OFSAR table. 1 of 1 SGS-OFSAR Revision 16 January 31, 1998

TABLE 3.6-3 STEAM GENERATORFEEDWATERPIPE STRESS

SUMMARY

No. 12 Feedwater Line (1) Stress s a + sh (Sa + Sh) Point (2) Stress, psi (3) 1 1,187 43,750 0.027 2 2,567 43,750 0.058 3 5,951 43,750 0.136 5 6,201 43,750 0.141 6 6,365 43,750 0.145 11 6,669 43,750 0.152 13 8,199 43,750 0.187 15 7,938 43,750 0.181 21 7,857 43,750 0.179 23 8,140 43,750 0.186 26 6,220 43,750 0.142 Notes: (1) Stresses tabulated for Nos. 11, 12, and 14 steam lines at geometrically similar locations are of similar magnitude. (2) Stress point locations are shown on Figure 3.6-11. (3) Stresses tabulated are the swnmation of longitudinal pressure, thermal, dead weight, and OBE (1/2 SSE) . 1 of 1 SGS-UFSAR Revision 6 February 15, 1987

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+---------- .,._____..,._____........ REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Typical Feedwater Pipewhip Attachment SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.6-19
CONT,iµllifCNr. .Af'.NETfATfCW It ,\HC'Pi:, REVISION 6 FEBRUARY 15, 1987 Isometric View of CVCS Letdown Line PUBLIC SERVICE ELECTRIC AND GAS COMPANY Outside of Containment SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.6.20
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F3.6-26 y Re r to plant 204805 in DCRMS SGS-UFSAR Revision 27 November 25, 20
F3.6-27 y Re r to plant 204804 in DCRMS SGS-UFSAR Revision 27 November 25, 20
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APPENDIX 3.6A DESCRIPTION OF BACKDRAFT DAMPER I SGS-UFSAR Revision 6 February 15, 1987
Appendix 3.6A DESCRIPTION OF BACKDRAFT DAMPER A. The backdraft dampers, which isolate and/or direct the postulated steam flow to an acceptable area, are of the hinged parallel blade design with inteconnecting linkage to enable the blades to operate in unison. Gasketing is provided along the blade edges to limit blade leakage to the design limits. Figure 3.6-30 shows the typical design detail and control logic for the backdraft damper operation. Normal operation essentially consists of a differential pressure transmitter actuating a solenoid valve that provides an air supply to operate the damper mechanism. Fail-safe logic is designed into the components by means of an internal spring in the damper 1 s drive mechanism which makes the damper go to the fail-safe position should there be a loss-of-air supply. B. The backdraft dampers are considered an integral part of the equipment and hardware provided to protect against the unlikely consequences of the postulated break in the piping systems. As such, the dampers are designed to the existing Seismic Class I criteria for all the plant ventilation dampers which included actual prototype shake table testing. In addition to the dampers, equipment such as pipe encapsulation sleeves, pipe restraints, and impingement baffles were also employed where the preferred physical separation approach was not feasible in implementing regulatory criteria. For the dampers in question and because there is no additional space available to install a third level of protection against the postulated event(s), a periodic inspection testing program will be developed used to assure continuous functionality of the dampers.
c. Pressure differential transmitters have been provided with trip points to assure a safe minimum trip point. Steam leakage causing conditions below the trip points is within the ventilation system capa.city so as to provide conditions for a safe cold shutdown should it be required in the event of a small steam leak *
SGS-UFSAR 3.6A-l Revision 15 June 12, 1996
3.7 SEISMIC DESIGN Plant structures and components are designed according to the criteria described herein to resist the dynamic forces resulting from the earthquake conditions postulated for the site. These structures and components are analyzed either dynamically or statically for the respective equivalent loadings according to their classification. Two types of seismic loadings are considered: Operational Basis Earthquake (OBE) and Design Basis Earthquake (DBE) . For the OBE loading condition, the nuclear steam supply system is designed to be capable of continued safe operation. Therefore, for this loading condition critical structures and equipment needed for this purpose are required to operate within design limits. The seismic design for the DBE is intended to provide a margin in design that assures capability to shut down and maintain the nuclear facility in a safe condition. In this case, it is only necessary to ensure that required critical structures and components do not lose their capability to perform their safety function. This has come to be referred to as the "no-loss-of-function" criteria and the loading condition as the "design basis earthquake" loading condition . Not all critical components have the same functional requirements for safety. For example, the reactor containment must retain capability to restrict leakage to an acceptable level. Therefore, based on present practice, general elastic behavior of this structure under the "Design Basis Earthquake" loading condition must be ensured. On the other hand, many components can experience significant permanent deformation without loss of function. Piping and vessels are examples of the latter where the principal requirement is that they retain their contents and allow fluid flow. What follows in Section 3. 7 is a description of the original design basis I seismic analyses performed for Salem Units 1 and 2. The reanalysis of the Unit 1 and Unit 2 reactor coolant loops, which addresses the elimination of snubbers from the steam generator upper supports, is described in Section 3.9.1.8 .
SGS-UFSAR 3.7-1 Revision 23 October 17, 2007
3.7.1 Seismic Input 3.7.1.1 Design Response Spectra The El Centro ground motion of May 18, 1940, was recommended by Darne.s and Moore as the most appropriate motion for the site. Its peak horizontal acceleration was normalized to O.lOg and 0.20g for Operating and Design Basis Earthquake, respectively. Two-thirds of the above mentioned values are used for vertical ground motions, and they are considered to be acting simultaneously with the horizontal ground motion. Modified Hausner's average response spectra as shown on Figures 3.7-1 and 3.7-2 are used for manual modal analysis. The Salem ground response spectra are generally lower than those normalized from Regulatory Guide 1.60. However, the conservative damping values used in the Salem analyses compensate for the differences. Furthermore, time history input was used for Category I structure seismic analyses. The normalized El Centro 1940 ground response N-S components, as shown on Figures 3.7-1 and 3.7-2, are considerably higher than the Salem ground response spectra. 3.7.1.2 Design Response Spectra Derivation The tabulation which follows provides a comparison of the damping values used in the seismic analysis with those identified in Regulatory Guide (RG) 1.61. It can be seen that the damping values used in the Salem analysis are consistently more conservative than the Regulatory Guide recommended values. 3.7-2 SGS-UFSAR Revision 6 Februarv lS 1QR7
Component Damping Values Salem RG 1. 61 Salem Large Diameter Piping 1.0 3.0 0.5 2.0 Systems Greater than 12" Small Diameter Piping 0.5 2.0 0.5 1.0 Systems 12" or less Concrete Structures 5.0 7.0 2.0 4.0 Bolted or Riveted Steel 5.0 8.0 2.5 4.0 Welded Steel 3.0 4.0 1.0 2.0 In the seismic analysis of mechanical equipment (Westinghouse supplied) and Category I structures the method of combining responses is to add absolutely the results of the vertical and the worst of the two horizontal earthquake components. 3.7.1.3 Critical Damping Values The following damping values are used in the design: Component Percent of Critical Damping
1. Concrete Structures 2 (OBE) 5 {DBE)
SGS-UFSAR 3.7-3 Revision 23 October 17 1 2007
2. Structural Steel Bolted or Riveted 2.5 (OBE) 5 (DBE)
Welded 1.0 (OBE) 3 (DBE)
3. Vital Piping System 0.5 (OBE) 0.5 (DBE)
For the DBE, a damping factor of 5 percent of critical damping is used for analysis for structure and soil. Similarly, for the OBE, a damping factor of 2 percent is applied for both structure and soil. 3.7.1.4 Bases for Site Dependent Analysis The soft soil properties, as determined by Dames and Moore, used in the finite element seismic analysis will filter out some high frequency waves. However, the input response spectra, as shown on Figure 3. 7-2, indicate that the structure will experience very little or no amplification in the high frequency range. Various modal analyses, both with or without soil interaction, indicate that the fundamental mode and other modes with large participation factors are not in the high frequency range. Thus high frequency filtering will not significantly affect the structural response. Conrad Associates has also performed two separate studies, with and without soil backfill around the Containment Building, to find the most critical building response. They have concluded that the inclusion of the soil backfill in the finite element model yields a more critical condition for the Containment Building. The result of the plots is shown on Figure 2-63 of the Conrad Associates Report (1). A nonlinear stress-strain relationship for soil properties has been recognized in our modal analyses. Due to the different strain levels in the soil under the OBE and DBE) the secant modulus of the soil has different values. Our soils' consultant, Dames and Moore, established the values for the secant modulus 3.7-4 SGS-UFSAR Revision 6 February 15, 1987
through dynamic triaxial tests of soil samples which imposed strain levels on the soil equivalent to those levels which are predicted to develop in the soil at the Salem site under the two different earthquake magnitudes. The range of the secant modulus of the different layers of soil at the Salem site under OBE and DBE motions is listed in Table 2-3 of Reference 1. The free field response spectra at Elevation 30 feet, scaled down from El Centro N-S components, are shown on Fiqures 3.7-1 and 3.7-2. Both of the free field response spectra envelop the smooth curve site design spectra except at frequencies smaller than 0.25 cps, which is beyond the range of all structural modal frequencies. Although the soil properties were not varied in their courses of motion under varying strains for each separate OBE and DBE seismic analysis, this effect has been taken into account in the parametric studies for establishing the most severe design load. The analytical model for the most critical case is that which has compacted fill completely surrounding the containment Building, while in the actual case, the compacted sand encompasses only a portion of the periphery of the containment wall. In this conservative model, the soil rocking spring was increased more than the equivalent rise of the soil property by 30 percent. Reduction of the soil modulus of elasticity, or neglecting the backfill completely, results in a lese critical case in the structural response. Thus, the most conservative response from soil interaction has been used in the design of the Category I (seismic) structures. 3.7.2 seismic system Analysis This information is printed in detail in Reference 1. 3.7-5 SGS-UFSAR Revision 6 February 15, 1987
3.7.2.1. Seiamic Analysis for Structures 3.7.2.1.1 Seismic Analysis for Category I Structures For seismic analysis of structures, two separate model analyses, horizontal and vertical motions, were performed and their results superimposed. The acceleration time histories from the result of the structural seismic analysis were used for the generation of horizontal and vertical response spectra at specified floors or locations for equipment of seismic design. They are kept on file by Public Service Electric & Gas (PSE&G) (1 and 3). Total accelerations, peak displacements, and the envelope of forces in the containment structure under DBE and OBE conditions are shown on Figures 3.7-3 through 3.7-12. Clearances between category I buildings and adjacent structures were checked based on the relative displacements at various building elevations under seismic and design basis accident loadings to assure that the required separations are maintained. Seismic design for category I structures is based on dynamic analysis. The mathematical model of an equivalent system has been used to aimulate the response. The system consists of lumped masses at floor levels supported by weightless springs. Floor rotations are neglected. Both shear stiffness and flexure stiffness are included in the spring constants. The lumped mass system is considered resting on the rigid base mat. To solve the modal frequencies and mode shapes, the Stodola-Vianello Procedure and the Modified Rayleigh Method were used in manual analysis. The independent computer analysis sets up the dynamic equilibrium equations in the matrix form and uncouple& the response of the structure through the computer program. 3.7-6 -* SGS-UFSAR Revision 16 January 31, 1998
Average response spectra were used in the manual analysis while time history inputs were used in the computer analysis. Rocking of a structure due to yielding of the subgrade under the lean concrete fill was considered. Response of the structure under the combined rocking and translational modes as well as the separated independent modes were evaluated. For soil structure interaction in the computer analysis, a series of studies were conducted in order to establish the free field soil boundaries. With the boundaries established, the finite element model, including the surrounding soil mass, can be set up more accurately. With the exception of Auxiliary Building horizontal seismic responstS, the containment structure seismic analysis was performed through (1) lumped mass model manual analysis, using average response spectra ground input, and (2) a finite element model ana lysis, using time history ground input. The detailed report from Conrad Associates (1) and the independent manual calculations are kept on file by PSE&G. The computer analysis yielded a slightly higher result in accelerations, shears, and moments in comparison "ith the manual analysis. The most conservative results were used in design. The seismic analysis of the containment structure by the finite element method was performed by computer using a step-by-step direct integration procedure. Studies were made to establish free field soil boundary condition. The model used in the analysis is shown on Figure 3.7-13. 3.7.2. 1.2 Seismic Analysis for Category II and Category III Structures There are no Category II structures at the Salem Generating Station. 3.7-7 SGS-UFSAR Revision 6 February 15, 1987
Category III structures are designed for loadings commonly used in the design of conventional power plants. In the State of New Jersey, no seismic analysis is required for conventional power plants, or non-nuclear elements of nuclear power plants. Extra precautions have been taken to brace the Category III structures in the direction of Category I structures so that the possibility of failure of the Category III structure affecting the integrity of the Category I structure is eliminated. 3.7.2.2 Natural Frequencies and Response Loads Containment Building mode frequencies, shapes, and participation factors are shown on Figures 2-29 through 2-46 in the Conrad Associates' report. The floor response spectra are shown on Figures 2-68 through 2-91 in that report. 3.7.2.3 Procedures Used to Lump Masses Masses were lumped at points of mass concentration, such as building floors, or selected points so that the displacement of these points gave a good representation of the distortion of the structure. The ratio of floor mass to the supported equipment is indeed very large. Thus, the equipment compliance will not alter the building of floor response. A finite element model was used for seismic analysis containment structures at the Salem station. A total of 190 elements were used in discretizing the structure. Lumped mass models were used for the seismic analysis of the Auxiliary Building and Fuel Handling Building. The points of mass concentration of these buildings are most apparent at the roof, floor, and foundations. Heavy equipment and subsystems in the buildings are rigidly attached to the floors. Therefore, the masses of the analytical models were logically 1 umped at these levels. 3.7-8 SGS-UFSAR Revision 6 February 15, 1987
Analytical models have been reviewed and it is concluded that the degrees of freedom used are adequate. Additional degrees of freedom in these models will not result in more than 10 percent increase in structural responses. The mass ratios of subsystems to the supporting structures are less than 0. 1 and, therefore, the subsystems are not included in the structural model. Based on the above, the Salem design is in compliance with the modeling criteria defined in Section 3.7.2 of the Standard Review Plan.
3. 7. 2. 4 Methods Used to Couple Soil with Sei smi_c- Sys tern Structures Studies were undertaken to examine the influence of the soil boundary location on the dynamic behavior of the Containment Building. The soil-structure continuum boundary was established through five finite elementary models. With the free field soil boundary determined, the finite element model was set up for seismic analysis. The seismic analysis of the Containment Build:ing by the finite element method was performed by computer using a step-by-step direct integration procedure.
3.7.2.5 Development of Floor Response Spectra Response spectra at different floors in the structures were derived for use in the design of Category I (seismic) mechanical and electrical equipment, piping, and their supports. These spectra were obtained as follows: The building was subjected to the input ground acceleration time history and the corresponding output acceleration time histories at the floors of interest were determined. The acceleration time histories were then used to derive single-degree-of-freedom system response spectra, which are the floor response spectra, for each floor of interest. 3.7-9 SGS-UFSAR Revision 6 February 15, 1987
3.7.2.6 Differential Seismic Movement of Interconnected Components In most cases, the seismic analysis of Category I (seismic) piping systems was performed using the modal response spectrum method. In order to account for the effects of the relative displacement between piping support points in this analysis, the following procedures were used:
1. For any given system under analysis, define piping support locations which could transmit significant forces to piping due to relative building and/or equipment displacements.

2. For analysis of building and/or equipment vibrational characteristics, establish an upper bound on the magnitude of relative displacements for locations defined in 1., above.

3. Apply significant relative displacements found in 2. ,
above to the piping system as static anchor (or support) displacements. 3.7.2.7 Combination of Modal Responses In the response spectra analysis, the total response for the structures and components is the square root of the sum of the square of all maximum values from all modes, provided there are three modes or more. In cases where there are only two modes contributing significantly or when there are three contributing modes with only one dominant, the absolute sum of the maximum values was used for design. In the time history analysis, the corresponding response in each vibrating mode is calculated as a function of time. The total response for any desired instant is evaluated by summing the response of all significant modes. 3.7-10 SGS-UFSAR Revision 6 February 15, 1987
l f modal frequencies are closely spaced, the total response for structural items is computed as the absolute of the modal responses. 3.7.2.8 Effects of Variations on In equipment seismic design using floor response spectra as input, PSE&G requested that suppliers ensure that the period falls outside of the peak regions in order to avoid a resonance effect and sensitive variation in response under a small shifting of period. On other steep portions of curves, conservative response readings were made to account for the inaccuracy of periods of vibration due to various unaccountable effects in the modal analysis. A minimum of ! 10 percent shifting of period coordinates was provided in obtaining the system's conservative response from the floor response spectra. This allows for the possible period inaccuracy due to the variation of structural properties, damping values, soil properties, and soil structural interaction. The intent in qualifying component seismic capability was that the component should be analyzed with mode frequencies apart from the floor response peaks to avoid the effect of damaging resonance. Vendors were instructed to stiffen the component to shift its model frequencies outside the spike area. In other more flat areas of the response spectra, a 10 percent shifting of period coordination has been provided in obtaining the system's conservative response. Regulatory Guide 1.122 (issued on September 1976) requires that the peaks of the response spectra be broadened by + 10 percent fj if computed value is less than that amount ~ 15 percent fj if the variation value is not computed. The Salem floor response spectra was developed before the issuance of this Regulatory Guide. At the time of the Salem design, 10 percent shifting was considered sufficient. 3.7-11 SGS-UFSAR Revision 6 February 15, 1987
As noted above, the containment floor response spectra were broadened by :!:_10 percent. The prominent spike, however, is at 1 cps, a very flexible frequency. None of the equipment falls in this area. In the other Category I buildings, there is more than one spike in the floor response spectra. In order to avoid undesirable resonance effects, components were stiffened and frequencies shifted outside the sharp spiked areas, thereby eliminating the broadening considerations. In other flat areas of the response spectra, a 10 percent shifting of frequency coordinates was applied to obtain the equipment response. Since these areas are flat, an additional 5 percent shifting would not cause an appreciable change to equipment response. 3.7.2.9 Method Used to Account for Torsional Effects PSE&G's evaluation in response to the Nuclear Regulatory Commission (NRC) request to add a 5 percent accidental eccentricity in the containment seismic design is as follows:
1. Torsional shears as a result of 5 percent eccentricity in the thick containment shell wall and interior crane wall are very small. Under DBE loading, the containment shell wall will have torsional shear of 7 psi while in the crane wall, only 3 psi.

2. Designed to be resisted solely by the 45° diagonal reinforcement. The concrete and liner plate were neglected in taking any portion of the shear load. The reserve capacities in the concrete and liner can easily handle the extremely small torsional shear stress.

3. If one attempts to avoid seismic shear in the concrete and liner, then the hoop reinforcement in the cylinder will act as torsional stirrups. The torsional shear will merely increase the rebar stress by 0.7 ksi. In our loading combinations (B)' (C)' and (E) when the 3.7-12 SGS-UFSAR Revision 6 February 15, 1987
torsional moment is present, the stresses in the hoop reinforcement in the lower portion of the cylinder are under 50 percent of their capacity.
4. The reserve capacity of the concrete in the crane wall, assisted by the many interior walls framed into the crane wall, will easily take care of the even smaller torsional shear.
The Containment Building is basically axisymmetric; therefore the torsional mode is not prominent. In designing the Auxiliary and Fuel Handling Buildings, the torsional moments at each floor level due to the distance between the center of rigidity and the center of mass were taken into account. The floors are very strong and rigid and the supporting walls would deflect equally, provided the center of mass coincides with the center of rigidity. The horizontal shear would be distributed to the cross walls in proportion to their stiffness. However, since the lateral shear is applied not at the center of rigidity, but at the center of mass, an adjustment for shear distribution must be made to compensate for the torsional effect. The shear resisting capacity of walls which resist the torsional moment have therefore been increased. For conservatism, other wall shears were not reduced where the torsional moment becomes beneficial to them. 3.7.2.10 Comparison of Responses The finite-element model was only applied to the seismic analysis of the containment structure. The Auxiliary Building and the Fuel Handling Building modal analyses were based on lumped mass models. Two independent seismic analyses were performed for the containment structure, namely a finite-element time history analysis and a lumped mass response spectrum analysis. The comparison of response at selected points in the containment structure from these two analyses is as follows: 3.7-13 SGS-UFSAR Revision 6 February 15, 1987
Acceleration Displacement Time History 0.60g 1.88 inches 0.28g 0.53 inch
Response
Spectrum 0.59g 1. 96 inches 0.29g 0.69 inch 3.7.2.11 Methods to Determine Category I Structure Overturning Linear modal analysis was performed using time history and ground response spectra as input. Modal response was determined through both computer and manual analyses. From the displacement responses, the maximum values of the shear and overturning moment were determined by applying the proper spring constant and lever arms at different floor levels. As part of the seismic analysis of the finite element model by Conrad Associates, soil reactions in the compacted sand fills were computed for both the DBE and OBE conditions. Vertical seismic analysis for the Containment Building was also performed. The vertical response was considered to be either upward or downward so that it would increase the soil pressure when it was downward and decrease the stability factor against overturning when it was upward. 3.7.2.12 Analysis Procedure for Dampirrg Major Category I structures are essentially concrete. Steel equipment supports and platforms represent a very small percentage of the total mass of the entire structure. In the building modal analysis, the equipment and platform masses were included in the lumped mass system and a uniform damping value was used. The 3.7-14 SGS-UFSAR Revision 6 February 15, 1987
floor response spectra were thus developed to be used as the input loading to the equipment and support structures, where the appropriate damping value for the material was assigned in the analysis. 3.7.3 Seismic Subsystem Analysis 3.7.3.1 Determination of Number of Earthqua__k_e___ ~---- The containment liner is the most critical structural component that may be subjected to fatigue stress under cyclic loading. Tests* have indicated that plates stressed under reverse stresses of 20 ksi tension and compression produced failure after 180,000 cycles. The containment liner plate stresses under seismic load alone are below that range. Thus, the number of seismic loading cycles indicated above is not at all damaging. Category I (seismic) piping, with the exception of the primary coolant loop, has been designed to the stress criteria indicated in Section 3. 9. 2. The number of maximum amplitude loading cycles due to seismic events has been considered in the design criteria; however, it has been found unnecessary to apply correction to allowable stresses to account for fatigue phenomena in the design of this piping. Experience dictates that the fundamental mode of vibration of piping systems varies between 0.1 and 1 second. Earthquake strong motion can be conservatively estimated as 10 seconds in duration. Postulating five OBE events and one DBE event over the life of the plant, the number of seismically induced stress cycles can be calculated as
~Highway Research Record No. 176, "Fatigue Tests of Plates and Beams" 3.7-15 SGS-UFSAR Revision 6 February 15, 1987
Nx Seismic Strong Motion Duration N s = ea T where: Ns = number of seismically induced stress cycles Nea = number of seismic events T = fundamental piping period, seconds Therefore, for the OBE: N = (5 events)(10 seconds/event) s (0.1 seconds/cycle)
= 500 cycles and, for the DBE:
N = (1 event)(lO seconds/event) s (0.1 seconds/cycle) 100 cycles Although the allowable design stresses are greater than the endurance limits of the materials used, it can be shown that the usage factor, U, defined as: u = Number of Specified Stress Cycles Number of Permitted Stress Cycles is sufficiently small to provide an adequate margin of safety. Westinghouse supplied Category I (seismic) systems, components, and equipment requiring a fatigue evaluation by the appropriate 3.7-16 SGS-UFSAR Revision 6 February 15, 1987
codes and standards are evaluated according to a criteria somewhat different than that used by PSE&G. Five OBE events and one DBE event are postulated during the life of the The actual number of maximum response cycles in a response history varies with several parameters such as damping of the equipment and the support structure, ground time history, etc. The response cycles of a magnitude less than the maximum response may be converted to an equivalent number of maximum response cycles by using the fatigue curves given in the ASME Boiler and Pressure Vessel Code, Section III. The Areva NP Unit 2 RSG and Westinghouse supplied Category I (seismic) systems, components, and equipment analyzed for fatigue are evaluated for an of 10 maximum response cycles per event. 3.7.3.2 Basis for Selection of Forcing Frequencies Components and equipment supports are required to have their fundamental frequency apart from the building frequencies to avoid resonance. When their fundamental were found to be in the range of the resonance peak area, they were stiffened or otherwise modified to shift their mode frequencies outside of that region. 3.7.3.3 Procedure for Combining Modal Responses If modal frequencies are closely spaced, the total response for structural items is computed as the absolute sum of the modal responses. For Category I (seismic) piping (except the primary coolant loop), where the modal response spectrum technique is utilized, the combined total response for each earthquake is taken as the square-root-of-the-sum-of-the-squares of the modal responses for 3.7-17 SGS-UFSAR Revision 24 May 11, 2009
any response parameter considered. The use of this criterion for combining modal responses in the response spectrum method of analysis may not be valid, however, in combining closely-spaced in-phase modes of vibration. This is accomplished by computing modal responses and then using both the square-root-of-the-sum-of-the-squares criteria and the absolute sum criteria in combining modes. In many locations in a complex model, both criteria give nearly equal results, which means a single mode is contributing to the response. If the absolute sum and the square-root-of-the-sum-of-the-squares combinations are different, the modes which contribute are checked. If contributing modes are closely-spaced in-phase modes, they are combined using the absolute sum criteria and treated as a single mode when combined with the rest of the modes using the "root-mean-square (RMS)" criteria. For Westinghouse-supplied equipment, the combined total seismic response is also obtained by adding the individual modal responses, utilizing the square-root-of-the-sum-of-the-squares method. Combined total response for closely spaced modal frequencies whose eigenvectors are perpendicular are handled in the above mentioned manner. In the rare event when two significantly closely spaced in-phase modes occur, the combined total response is obtained by adding the square-root-of-the-sum-of the-squares of all other modes to the absolute value of one of the closely spaced modes. 3.7.3.4 Bases for Computing Combined Response For Seismic Category I piping (except primary coolant loop piping) receiving modal analyses, the combined responses to horizontal and vertical seismic excitation are computed as follows:
1. The moments (and forces) in each of three orthogonal directions e.g. (F t F , and F ) are computed separately X y Z for each mode and for both horizontal and vertical excitation.
3.7-18 SGS-UFSAR Revision 6 February 15, 1987
2. The algebraic sum of moments and forces generated by the vertical and horizontal excitations is computed in each axial direction for each mode.

3. The square-root-of-the-sum-of-the-squares of each modal algebraic sum (from 2., above) is found for each orthogonal direction. This results in "RMS" values for F , F , and F due to the simultaneous action of X y Z vertical and horizontal excitations.
In the seismic analysis of primary loop piping, the results for the vertical direction are added absolutely to the results of the worst of those for the north-south and east-west directions. Safety-related instrumentation, except for the trip breakers, is tested by the methods recommended in IEEE Standard 344-1971, "IEEE Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations." The trip breakers have been shock tested instead of sine beat tested because IEEE Standard 344-1971 was not applicable at the time the equipment was purchased. 3.7.3.5 Use of Simplified Dynamic Analysis Most components and systems can be simplified as a single degree of freedom system for which the period of vibration can be readily obtained. Even in a multi-degree system the fundamental period can be arrived at by the Modified Rayleigh Method with relative ease. From the resulting period of the component, we can compare that with the input period to assure there wi l1 be no resonance effect. Seismic Category I p1p1ng is divided into two categories: rigid and flexible. Rigid category piping is defined as that piping whose lowest natural mode of vibration has a frequency over 20 Hz, while flexible category piping has a frequency less than 20 Hz. 3.7-19 SGS-UFSAR Revision 6 February 15, 1987
Seismic Category I piping which falls into the flexible category generally receives a modal dynamic analysis (with either response spectra or time history inputs). Seismic Category I piping which falls into the rigid category generally receives a simplified or "static equivalent" analysis. This procedure is justifiable because, for rigid systems, essentially no amplification of support accelerations occurs. In these cases, horizontal and vertical support accelerations are imposed upon the piping as static loads and the resulting stresses computed. The "cutoff frequency" (or "cutoff point") is defined as that frequency at which the spectral response curve at a given location in a given building or structure "flattens-out"; i.e., that frequency above which the response of a single degree of freedom oscillator is independent of its natural frequency and only a function of location (elevation) in the building. Elastic vibration theory predicts that if a modal analysis is conducted on a system subjected to an excitation which produces a flat response curve, the resultant loadings will be the same as those obtained by applying a "static" loading equivalent in magnitude to the flat portion of the response curve. Thus, for vibrating systems whose first or lowest normal mode of vibration is above the "cutoff frequency" described above, i t is justifiable to perform a "static equivalent" or "simplified" analysis. The fundamental frequencies of key subsystems were considered in relation to the dominant frequencies of their supporting systems. The key subsystems have been determined to be adequately designed for the applicable loads. Elimination of resonance was one of the principles of design. Various methods for seismic qualification were employed for key subsystems. In most cases, the key subsystems were considered to 3.7-20 SGS-UFSAR Revision 6 February 15, 1987
be very flexible and were analyzed/tested as a decoupled system from the supporting system. Additional information is contained in Sections 3.7.3.3, 3.7.3.4, 3.9.1.2, and 3.10. These subsystems were analyzed/tested as a decoupled system from the supporting system, because the mass ratio of the subsystem to that of the supporting system is less than 1 percent. 3.7.3.6 Modal Period Va The mass and spring constants due to variation in materials have been properly accounted for in the dynamic equations. The damping factor is the only remaining factor. Refer also to Section 3.7.2.9. The fundamental equation of motion is: M x" + C x' + Kx :::: F(t) In which the spring constant, K, is a function of the modulus of elasticity, E, and the modulus of rigidity, G, of the material of the structural component. The proper E and G, with respect to various materials, are used for the stiffness of the weightless spring in the dynamic equations. A mass matrix in the above equation is also established by applying the proper mass density for the various materials. A conservative damping coefficient, C, is used in the input spectrum of time history. Modal periods are obtained by solving the dynamic equations. 3.7.3.7 Torsional Effects of Eccentric Masses The torsional effects of valves and other eccentric masses are accounted for in the seismic piping analysis. 3.7-21 SGS-UFSAR Revision 6 February 15, 1987
For rigid components (Section 3.7.3.5), this is accomplished by modeling these components as a concentrated (point) mass, equal in magnitude to the component mass, located at the center of gravity of the component. The mass is then modeled as being connected to the piping model by a massless rigid member. For flexible components, the torsional effects are included by modeling as a lumped mass-massless spring system coupled to the piping system model. 3.7.3.8 Piping Outside Containment Structure Seismic Category I piping located outside the containment structure is either located in (or supported from) Seismic Category I buildings or structures, or is buried in the ground. For that Seismic Category I piping supported from Seismic Category I structures, analytical procedures and design criteria are the same as for piping within the containment. For buried steel piping outside the containment, flexible joints have been provided at wall sleeves where necessary to accommodate the maximum expected differential ground motion without over stressing piping or structural components. Also, some piping has been surrounded by friable material to reduce bearing stresses due to differential motion at points of entrance to buildings. Seismic Category I buried concrete piping sections have been provided with specially designed slip joints. These joints permit seismically induced angular deflections and axial strains without overstress or loss of system integrity. An access manway for personnel egress and isolation valve is installed in the buried SW discharge headers near the tie-in to the CW piping allowing for the headers to be dewatered for inspections. Concrete missile shield vaults with steel covers are provided to protect the access manway and valve operator reach rod assembly from tornado-generated missiles. These missile shields meet the requirements in Section 3.5.2.2 to withstand applicable tornado missiles as defined in Section 3.5.2.1 of the UFSAR. 3.7-22 SGS-UFSAR Revision 30 May 11, 2018
3.7.3.9 Field Location of Supports and Restraints The locations of seismic supports for Seismic Category I piping is determined by analysis during the design of a system. These locations and orientations are then shown on detailed piping arrangement drawings and "hanger detai 1 11 drawings. The seismic supports are then installed according to the above drawings, i.e., their location is not a field decision. 3.7.3.9.1 General Procedure During the erection of piping systems for Unit 1, a Stress Analysis Task Force was formed whose function was to assure, by inspection, that safety-related piping conformed to stress isometric drawings. This included incorporating the "as-buil t 11 conditions. These "as-built 11 conditions were reviewed by Stress Engineering. If the Stress Engineer determined that the revision was questionable, the calculation was rerun and the stress isometric updated. If the condition was a minor deviation and no adverse effect would occur to the stress levels, this condition was accepted "as-built" and noted on the stress isometric. 3.7.3.9.2 Safety-Related Piping Safety-Related Systems There are 15 systems involved in the safety-related piping investigation to confirm conditions of actual configuration. 3.7-23 SGS-UFSAR Revision 6 February 15, 1987
The following is a list of the systems which are inspected:
1. Reactor Coolant System

2. Safety Injection System

3. Steam Generator Feed System

4. Component Cooling System

5. Service Water - Nuclear System

6. Auxiliary Feedwater System

7. Containment Spray B. Main Steam System

9. Chilled Water System

10. Chemical and Volume Control System

11. Control Air System

12. Diesel Generator Starting Air and Fuel System

13. Steam Generator Blowdown System

14. Spent Fuel Cooling System

15. Residual Heat Removal System The stress analysis calculation numbers for each system for all safety-related piping and the calculations that have been stress walked are given in Reference 2. The completed field walk covered the inaccessible and the reactor coolant pressure boundary calculations.
3.7-24 SGS-UFSAR Revision 6 February 15, 1987
The results of the field walk were evaluated by PSE&G Stress engineers. The inspection walk results obtained can be examined at PSE&G Newark and Salem offices. As requested by NRC IE Bulletin 79-14, the safety-related piping systems were field walked. This added assurance to identify nonconformances and confirm that seismic input information conforms to the "as-built" conditions. In each system PSE&G identified the analytical isometrics which represent the calculated piping. The isometrics incorporate the following information:
1. Arrangement drawing numbers - to obtain piping geometry.

2. Insulation drawing numbers - type, weight, and location of insulation.

3. Hanger detail drawing numbers specific location and type of support.

4. Equipment manufacturer's print numbers - the location of nozzles, weights of valves, centers of gravity, and dimensions.
In responding to NRC Bulletin 79-07, PSE&G agreed to perform a stress walk with the NRC. That sample walk took place in the inaccessible areas of the containment by the Resident NRC Inspector, together with the PSE&G Stress Engineer. The procedure that was followed is described below. This same procedure was followed during the implementation of Item No. 2 of the subject Bulletin 79-14. The results of these walks confirmed our previous assertions that the actual configuration conforms to the stress isometric drawings. 3.7-25 SGS-UFSAR Revision 6 February 15, 1987
Procedure for Stress Isometric Verification The purpose of the stress walk was to give reasonable assurance that Unit 1, Salem Generating Station, as constructed, was represented on the isometric drawings (ISO) and, therefore, our stress calculations are valid. The Stress ISO Verification Program was initiated prior to Unit 1 11 Hot Functional Testing. 11 The stress isometric was used to perform this function because it incorporates other drawings into one composite. It is not drawn to scale, but shows dimensional piping configuration. It also includes:
1. Pipe material, size, and wall thickness

2. Allowable SA values

3. Operating temperature

4. Specifications

5. Support location and type

6. Insulation information Verification of the stress isometrics involved in "Hot Functional Testing" was done by checking the following:

1. Piping conforms to isometric configuration.

2. No obstruction impedes thermal pipe growth:

a. Engineering thermal calculations were used to determine necessary clearance in sleeves.

b. Carefully checked location of first guide after a change in direction of pipe run.
3.7-26 SGS-UFSAR Revision 6 February 15, 1987
c. Supports were described correctly on ISO and erected accordingly.

d. Supports found to restrict or. allow growth in the proper direction.

3. Verified that a movement chart existed on the isometric where pipe was connected to equipment.

4. Verified that bends and fittings were properly defined.
On 2- inch diameter and smaller pipe, verified socket welded fittings.
5. Control valves noted for center of gravity and if valve was tilted, it was noted as such.
Upon completion of the "Hot Functional Testing" Stress Isometric Verification Field Walk, all discrepancies were reviewed and the following steps were taken for achieving corrective status:
1. Calculations were rerun as necessary, using "as-built" information on the ISO.

2. Unacceptable calculations which occurred due to "as-built" revisions necessitated revisions by the field forces to correct "as-built" conditions to agree with the original acceptable stress isometric.

3. These stress isometrics were then rewalked to verify corrections.
Field Procedures 719 and 720 used by United Engineers and Constructors, Inc. for "Hanger and Hanger Support Shop Fabrication Mechanical" and "Mechanical Pipe Hanger Fabrication and Installation' 1 are contained in Reference 2. These procedures were also used when verifying the "as-built" conditions. 3.7-27 SGS-UFSAR Revision 6 February 15, 1987
3.7.4 Seismic Instrumentation Program
3. 7 .4.1 The seismic instrumentation provided follows closely the guidelines of Regulatory Guide 1. 12. It is comprised of the following:

1. A strong motion accelerograph system of a centrally located recording and printout device and three triaxial sensors with triggers. One of the triaxial sensors will be on the mat of the containment foundation and one each of the remaining two triaxial sensors will be at a higher elevation in the Containment Building (approximately Elevation 130 feet) and the Auxiliary Building (approximately Elevation 122 feet).
Installation of the triaxial sensors will be according to Item C.2 in Regulatory Guide 1.12.
2. Peak recording accelerographs will be located on selected Category I (seismic) structures and components in the Containment, the Fuel Handling, and the Auxiliary Buildings.
3.7.4.2 Location and Description of Instrumentation The number and location of peak recording accelerographs to be installed on Category I (seismic) components is under development. It is planned to locate these devices such that readings from the peaking recording instruments can be used to verify the seismic results derived analytically from the traces recorded in the strong motion accelerographs and also the dynamic model analysis. 3.7-28 SGS-UFSAR Revision 6 February 15, 1987
3.7.4.3 Control Room operator Notification Whenever the accelerograph recorder is triggered to record in any or all the channels, an alarm will be initiated in the control room so as to enable the operator to collect the traces from the accelerograph. The recording for the printout device is located near the control room area for immediate access. Starting of the recording device can be initiated for all nine channels of the three triaxial sensors by the triaxial starter unit. The signal from each axis of each triaxial sensor is recorded in one channel of the recorder. Peak level indication will be available to the operator within a few minutes. 3.7.4.4 Comparison of Measured and Predicted Responses The DBE level is 20 percent *g*, and the triggering level is set at 2 percent
g **
The Salem site possesses a common underlying soil condition, and the previously described design provides adequate measures to include all the pertinent considerations of different dynamic response of different Category I structures. Four peak recording accelerographs will be installed on selected category I (seismic) structures. They will be located so that the recorded accelerations can be used to verify the analytically derived seismic response. Two accelerographs will be located in the Containment Building, one at Elevation 81 feet and the other at Elevation 130 feet. one will be in the Auxiliary Building at Elevation 122 feet and the remaining one in the Fuel Handling Building at Elevation 130 feet. The instruments are designed to perform their function over the normal range of environmental conditions such as temperature, humidity, pressure, and vibration. 3.7-29 SGS-UFSAR Revision 6 February 15, 1987
Measured responses at various locations in Category I (seismic) structures or components are collected to ascertain that the severity of the peak responses does not equal or e.zceed the maximum on response for the various locations. In the event that the maximum OBE responses are equaled or exceeded, a detailed inspection will be conducted to determine whether there are any major cracks in the structural components or any equipment has undergone permanent deformation resulting from the earthquake motion. Additionally, the time history record from the acceleroqraph that has equaled or e.x.ceeded the OBB response will be analyzed to determine the ground motion response spectrum so that a comparison with the input design spectrum can be made. The effect of the earthquake on plant operation may thus be determined and any necessary atepa can be taken to assure the continued safe operation of the plant. 3.7.5 References for Section 3.7
l. *structural Analysis of containment Vessel, Salem Nuclear Generating station," Conrad Associates (submitted with FSAR Amendment 13, July 31, 1972).

2. Letter dated 9/14/79 Schneider to Grier, "NRC, IE Bulletin No. 79-14, Revision 1, Salem Generation Station,* Unit No. 1.
I
3. VTD 320237-01, "Design Basis Response Analysis of the Salem Nuclear Generating station structures,* EQE Final Report, January, 1995.
3.7-30 SGS-OFSAR Revision 16 January 31, 1998
VELOCITY IN INCHES / &!C.O ...OS PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION Comparison of Seismic (10%G) 2% DAMPEDSTRUCTURAL UpdatedFSAR REVISION8 FEBRUARY15,1987 ResponseSpectra O.B.E.
RESPONSE
Figure3.7*1
*0 IC,)
0 r*
~
V'6L.OCIT'VIN lt.&CMIS PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION
/SeCONDS Comparison (20%G) 5% DampedStructural UpdatedFSAR REVISION6 FEBRUARY 15,1987 of SeismicResponseSpectra D.B.E.
Response
Figure3.7-2
~-----------------------------------L-----------------------------------------~
TOTAl. VERTICAl. ACCELERATIONS TOTAL HOIIIUZONTAl. ACCELERATION 0.2718 G
EL. Zt'
'IERTICAL ACCELERATIONS IN T::2.4 SEC . HORIZONTAL ACCEL!ftATIONS IN MAt T:2.1 SEC.
REVISION 8 FEBRUARY 15, 1987 Envelope of Total Acceleratio.ns for Shell PUBLIC SERVICE ELECTRIC AND GAS COMPANY and Peak Total Accelerations for Mat*D.B.E. SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.7*3
VERTICAL RATIONS TOTAL HORIZONTAL ACCELERATIONS -
EL,7e' .*.. ...
-~-*
EL.ZI' VERTICAL ACCELERATIONS IN MAT, T::4.6 SEC. HORIZONTAL ACCELERATIONS IN MAT, T:r2.6 SEC. REVISION6 FEBRUARY15,1987 Envelope of Total Accelerations for Shell PUBLIC SERVICE ELECTRIC AND GAS COMPANY and Peak Total Accelerations for Mat-O.B.E. SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.7*4
-ORIGINALPOSITION
- - - DISP\.ACED POSITION u = I.88 11 U : HORIZONTALCOMPONENT v =o.o" V : VERTICALCOMPONENT ..,.,... / """"' . //"EL.264t II = 1.73" = *0.!51" I . ! EL. 218+ -'-
I u : 1.32 11 I V: -0.67 .. I 1 1 I EL.I80t. uv:= 1.0!5" 11
-0.66 I EL. ISO I U: 0.82n f
v: -0.6!5 .. I ----, J u: 0.!53 11 v = -0.63 11 J EL.IOO I u: 0.40.. I L -~,._--..-._ _...._ v = -0.6 .. Lu y
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION UpdatedFSAR REVISION6 FEBRUARY15,1987 Figure3.7-5
u Y ORIGINALPOSITION
---DISPLACEDPOSITION = HORIZONTALCOMPONENT = VERTICALCOMPONENT v =o.o" \ u = 0.93" \Y =-0.20 \
u = 0.72" v =. . Q.26 u :: 0.5 7 11 11 y:: -0.25 u =0.44" v =-0.25" f I I u =0.33" v =-0. 24..EL. 100 I I I ___ -_-_-_-_-..;-;;...;;:-;.:a.-+-~~~------",.__
- .................. ........_.__ ____ ...J u =0.05 11
v =-0.22 REVISION6 FEBRUARY15,1987 PeakDisplacements in Containment Vessel PUBLICSERVICEELECTRICAND GAS COMPANY Relative to Foundation Mat SALEMNUCLEARGENERATING STATION AtT = 2.54Sec.* O.B.E. (0.1 G) UpdatedFSAR Figure3.7-6
ME (KIP/FTI (KIP/FT) Merldronot Force _ _ .,.
+/-47.0 0 100 200 I I I ELEV. n* K/fT 40 VERTICALSTRESS (KSF)
REVISION8 FEBRUARY15,1987 Envelope of Forces in Containment Vessel Due to PUBLIC SERVICE ELECTRIC AND GAS COMPANY Design Basis Earthquake (0.2g) SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.7*7
MOMENT MERIDIONAL MOMENT (KIP FT/FT) C.rcumf.,.ntlal Moment
+/-52.5 100 200 0
I I EL£V. 7&' K.FT/fT
(KSF)
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION TANGENTIAL STRESS(KSF) Envelope of Forcesin Containment DesignBasisEarthquake REVISION6 FEBRUARY15. 1987 VesselDue to (0.2g) UpdatedFSAR Figure3.7*8
---------.&.tl2.1 0 100 200 I I I ELE" 76'
PUBLIC SERVICE ElECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR REVISION6 FEBR UARY15,1987 Envelope of Forces in Containment Vessel Due to Design Basis Earthquake (0.2g)
Figure3.7-9
MERIOIONA (KIP/FT) Mtrlcllonal Force---
!:2&
.,..;;;.~.......~~14!5..0
_1WW~~U6.3 0 100 200 huo I I K/FT ELEV. 76' RADIAL STRESS (KSF) PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION VERTICALSTRESS (KSF) Envelopeof Forcesin Containment Operating BasisEarthquake REVISION8 FEBRUARY15, 1987 VesselDueto UpdatedFSAR Figure 3.7*10
0 50 100 I I I I I I I ELE\t 76' K/trT
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATtON REVISION6 FEBRUARY15,1987 Envelope of Forces in Containment Vessel Due to Operating Basis Earthquake (O.lg)
Updated FSAR Figure 3.7-12
..i z*.s TOP C/1 1
J
' IHELL ELEMENT I -.*--**-*--*--::11.
IL. 101 * .,.. Dni._A. CRANE...., CONTAINMENT VESSEL *. t:L. l:tO'
. . . /\SOLID ELEMENTS\ !t MOUND SURFACE EL.
100 100
~~ ,_ \ ~* \ Jl.
lAND ... l
*. 1 ....... ~
II. ... HYIIIUULIC fiLL .............'; ........... IF~ . . .... .. Ill**.
.../. ~-** ~ !j I *. ".* ** * .~:~~ ~~
J. ~ ICIIUCWOOD ,ORMATION ~ ~,t:AOtfllll CONCftETI J~
,. 'A;.*
I'* "LL ,~.:
!0 I!I, I~~~ J~
0 YINCENTOWN FCHIMATION ~ .s J. ~
*40 .40 HOIINEfttTOWII IMD ~ L NAY!II. . P'OftiiATION -'1'5 .75 ... L IIICUtT LAUttEL IMD -110 --- - --- - -- -- -- J .ItO Ill 0 53' 104' IH' t1S' tlO' 175' *550' 440' REVISION 6 Axisymmetric Finite Element Model-containment Vessel FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUClEAR GENERATING STATION Updated FSAR Figure 3.7-13
Security-Related Information - Witheld Under 10 CFR 2.390 3.8 DESIGN OF CATEGORY I STRUCTURES 3.8.1 Containment Structure 3.8.1.1 General Description For arrangement of containment structures, the patterns of reinforcements, and the layout for liner, see Figures 3.8-1, 3.8-3, 3.8-8 and Plant Drawings 208900, 201102, 201105, 201108, 201175, 201181 and 201131. The reactor containment structure is a reinforced concrete vertical right cylinder with a flat base and a hemispherical dome. A welded steel liner with a minimum thickness of 1/4 inch is attached to the inside face of the concrete shell to ensure a high degree of leak tightness. The design objective of the containment structure is to contain all radioactive material which might be released from the core following a loss-of-coolant accident (LOCA). The structure serves as both a biological shield and a pressure container.
3.8-1 SGS-UFSAR Revision 27 November 25, 2013
The underground portion of the containment structure is waterproofed in order to avoid seepage of ground water through cracks in the concrete. The waterproofing consists of an impervious membrane which is placed under the mat and on the outside of the walls. The Ethylene Propylene Diene Monomers (by Uniroyal, Inc.) membrane will not tear in handling or placing of backfill against it. The installation of the membrane is described in Section 3.8.2.6.8.4. The basic structural elements considered in the design of the containment structure are the base slab, side walls, and dome acting as one structure under all possible loading conditions. The liner is anchored to the concrete shell by means of anchors so that it forms an integral part of the entire composite structure under all loadings. The reinforcing in the structure will have an elastic response to all loads with limited maximum strains to ensure the integrity of the steel liner. The lower portions of the cylindrical liner are insulated to avoid buckling of the liner due to restricted radial growth when subjected to a rise in temperature. The reinforcement patterns of the base mat are shown on Plant Drawings 201102 and 201105. The reinforcement patterns of the cylindrical wall are shown on Figure 3. 8-3. The reinforcement patterns of the dome are shown on Plant Drawing 201108. The containment structure is inherently safe with regard to common hazards such as fire, flood, and electric storm. The thick concrete walls are invulnerable to fire and only an insignificant amount of combustible material, such as lubricating oil in pump and motor bearings, is present in the containment. A lightning protection system is installed on the containment dome to protect against electrical storm damage. The dead weight of the structure is a minimum of 3. 0 times the buoyancy force that may be exerted on the structure if the ground water level is considered to be at a grade which is 3.5 feet higher than the normal ground water table. In case of a hypothetical hurricane 3.8-2 SGS-UFSAR Revision 27 November 25, 2013
flooding to a height of 20.9 feet above grade, the dead weight will be a minimum of 1.6 times the buoyant force. Therefore, the highest water conditions in the river will present no hazard to the flotation of the containment. Internal structures consist of equipment supports, polar crane gantry, shielding, reactor cavity and canal for fuel transfer, miscellaneous concrete and steel for floors and stairs. A 3-foot thick concrete ring wall serving as a partial radiation shield surrounds the Reactor Coolant System (RCS) components and supports the polar-type reactor containment crane. A 3 to 5-foot thick reinforced concrete floor covers the RCS compartments. Removable concrete plugs are provided to permit crane access to the reactor coolant pumps. The four steam generators, pressurizer, and various pipes penetrate the floor. Stairs provide access to the areas below the floor. The refueling canal connects the reactor cavity with the fuel transport tube to the spent fuel pool. The floor and walls of the canal are concrete, with walls and shielding water providing the equivalent of 6 feet of concrete. The floor is 4.5-feet thick. The concrete walls and floor are lined with 1/4-inch thick stainless steel plate. The linings provide a membrane that is resistant to abrasion and damage during fuel handling operations. The containment characteristics used to determine the containment structural heat sinks considered in the containment accident analysis are shown it Tables 15.4-20 and 15.4-21. 3.8.1.2 Design Codes The Containment Building has been designed under the following codes:
1. Building Code Requirements for Reinforced Concrete, ACI 318-63.

2. AISC Manual of Steel Construction, 6th Edition or later edition, as applicable.
3.8-3 SGS-UFSAR Revision 29 January 30, 2017
3. ASME Boiler and Pressure Vessel Code, section III, section VIII, and Section IX (Applicable portions) - 1968.
3.8.1.3 Design Loads and Loading Combinations The following loads are considered to act upon the containment structure creating stresses within the component parts:
1. Dead load The dead load consists of the weight of the complete structure as shown in the construction drawing. To provide for variations in the assumed dead load, the coefficient for dead load components is adjusted by +/-5 percent as indicated in the various cases of loading combinations.

2. Live load Live load consists of snow or construction loads on the dome and also the weight of major components or equipment in the containment. A construction load of 50 pounds per square foot, which is more severe than the snow load, is used in dome design.

3. Internal Pressure The internal pressure transient used for the containment design and its variation with time is shown on the pressure-temperature transient curve, Figure 3.8-11. For the free volume of 2,620,000 cubic feet within the containment, the design pressure is 47 psig. This pressure transient is more severe than those calculated for various LOCAs and Main Steam Line Breaks (MSLB) which are presented in Section 15.
3.8-4 SGS-UFSAR Revision 17 October 16, 1998
4. Thermal Thermal expansion stresses due to an internal temperature increase caused by a LOCA have been considered. This temperature and its variation with time is shown on the pressure-temperature transient curve, Figure 3.8-11. The maximum temperature at the uninsulated section of the liner under accident conditions is 246°F. For the 1.25 times and 1.50 times design pressure loading conditions given in Section 3.8.1.4.1, the corresponding liner temperature will be 285°F and 306°F, respectively. The pressure-temperature transient curves for these loading conditions are shown on Figures 3.8-12 and 3.8-13, respectively. The maximum operating temperature is 120°F.
For the Main Steam Line Breaks (MSLB), Figure 15.4-100 provides the containment pressure and temperature transients for the limiting temperature case. The governing peak temperature is 351.3°F.
5. Buoyancy Uplift due to buoyant forces created by the displacement of ground water by the structure has been considered. Computations are based on normal ground water being at grade level and flood water at 20.9 feet above grade during a hypothetical hurricane.

6. Seismic Load The site seismology and ground response spectra are described in Section 2. Seismic design criteria for structures and equipment are described in Sections 3.7 and 3.8.1.4.2.
3.8-5 SGS-UFSAR Revision 17 October 16, 1998
7. Wind Load A wind load of 30 pounds per square foot, equivalent to 108 mph, was applied to structures and found to be less critical than the operational Basis Earthquake (OBE) load.

8. Tornado The Reactor Containment, Fuel Handling, and Auxiliary Buildings have been checked to withstand a tornado loading based on a peripheral wind velocity of 300 miles per hour and a translational velocity of 60 mph.
Simultaneous with wind loading, an atmospheric pressure drop of 3 psig for all Class I structures has been considered. The shape factor, c, for the dome is 0.4 and for the cylinder, 0.5. No gust factor is applied. For additional information on tornado loadings, see Section 3.3.2.
9. Test Pressure The test pressure for the containment structure is 115 percent of the design pressure or 54 psig.

10. Negative Pressure Loading from an internal negative pressure of 3. 5 psig has been considered. A pressure of this magnitude would result from the combined effects of: cooling of the containment volume 70°F below the temperature at which 3.8-6
. SGS-UFSAR Revision 6 February 15, 1987
the containment was sealed, a rise in external barometric pressure of 1 psi, and burning up of -- hydrogen evolved in an accident conditi.on. The load combinations utilized to determine the required limiting capacity of any structural element in the containment structure have been computed as follows: Case A Operating plus DBA c~l.OD + O.OSD + 1.5P + 1.0 (T + TL) + l.OB Case B Operating plus DBA plus OBE c~I.OD + O.OSD + 1.25P + 1.0 (T' + TL')
+ 1. 25E + 1 . OB Case C Operating plus DBA plus DBE C+ 1. OD + 0. OSD + 1. OP + 1.0 (T" + TL") + l.OE I + 1.0B Case D Operating plus Tornado c~I.OD + O.OSD + l.lOW + l.OB + l.OPb - t Case E Operating plus DBE c~I.OD + o.osD + I.OT + t.OE' + I.OB Case F Testing C~l.OD + O.OSD + 1.15P + l.OB Symbols used in these formulae are defined as follows:
c Required load capacity of section. D Dead load of structure and equipment loads p ~ Accident pressure load as shown on pressure-temperature transient curves. 3.8-7 SGS-UFSAR Revision 6 February 15) 1987
T = Load due to maximum temperature gradient through the concrete shell and mat, based upon temperatures associated with 1.5 times accident pressure. TL = Load exerted by the liner based upon temperatures associated with 1.5 times accident pressure. T' = Load due to maximum temperature gradient through the concrete shell and mat based upon temperatures associated with 1.25 times accident pressure. TL' = Load exerted by the liner based upon temperatures associated with 1.25 times accident pressure. T' = Load due to maximum temperature gradient through the concrete shell and mat based upon temperature associated with the accident pressure. TL" = Load exerted by the liner based upon temperature associated with the accident pressure. T = Load due to operating temperature gradient through the steel liner, concrete shell, and mat. E = Load resulting from assumed OBE or wind, whichever is greater. E' = Load resulting from assumed Design Basis Earthquake (DBE) B = Load resulting from buoyancy effect of ground water.
= Wind load due to tornado. = Bursting pressure loading associated with a tornado.
3.8-8 SGS-UFSAR Revision 6 February 15' 1987
The load factor approach is being used in this design as a means of making a rational evaluation of the isolated factors which must be considered in assuring an adequate safety margin for the structure. This approach permits the designer to place the greatest conservatism on those loads most subject to variation and which most directly control the overall safety of the structure. In the case of the containment structure, therefore, this approach places minimum emphasis on the fixed gravity loads and maximum emphasis on accident and earthquake or wind loads. The extent to which equilibrium checks of external loads against internal stresses have been made are as follows: Equilibrium checks of external loads against internal stresses have been conducted with a finite element computer program developed specifically for axisymmetric structures under non-symmetric loading by Conrad Associates. The required ultimate load capacity for any structural component of the Containment Building was established by utilizing the following load combination relationship: (a) C ~ l.OD + 0.05D + l.SP + l.OT + l.OB (b) C ~ l.OD + O.OSD + 1.25P + l.OT' + 1.25E + l.OB (c) c ~ 1. OD + 0 . OSD + 1. OP + 1. OT" + 1. OE I + 1. OB (d) C = l.OD + O.OSD + l.lOWt + l.OPb + I.OB (e) C = l.OD + O.OSD + l.OT + l.OE' + l.OB Symbols used in these formulae are defined as follows: c ~ Required load capacity of section. D = Dead load of structure and equipment loads. 3.8-9 SGS-UFSAR Revision 6 February 15, 1987
p ::: Accident pressure loads as shown on pressure temperature transient curves. T = Load due to maximum temperature gradient through the steel liner, concrete shell, and mat, based upon temperatures associated with 1.5 times accident pressure. T' = Load due to maximum temperature gradient through the steel liner, concrete shell, and mat, based upon temperature associated with 1.25 times accident pressure. T" = Load due to maximum temperature gradient through the steel liner, concrete shell, and mat, based upon temperature associated with the accident pressure. T I = Load due to operating temperature gradient through the steel liner, concrete shell, and mat. E = Load resulting from OBE or wind, whichever is the greater. E' = Load resulting from DBE.
= Wind load due to tornado. = Bursting pressure associated with a tornado.
B = Load resulting from buoyancy effect of ground water. Load combination a assumed that the containment will have the capacity to withstand loadings at least 50 percent greater than that calculated for the postulated LOCA alone. 3.8-10 SGS-UFSAR Revision 6 February 15, 1987
Load combination b assumed that the containment will have the capacity to withstand loadings at least 25 percent greater than that calculated for the postulated LOCA with a coincident OBE. Load combination c assumed that the containment will have the capacity to withstand loadings at least as great as those calculated for the postulated LOCA with a coincident assumed DBE. Load combination d assumed that the containment will have the capacity to withstand tornado winds and associated external pressure drop loadings. Load combination e combines the thermal gradient associated with normal operating conditions with the DBE. The resulting combination produces the maximum compressive stresses in the liner. The horizontal and vertical components of earthquake loads are considered to act simultaneously on the Containment Building. Resultant stresses from both components of loading are added directly with the other loads in the combination. Since the horizontal component of earthquake loading is non-symmetrical, producing tension on one side of the containment vessel and compression on the other, both the positive and negative values of earthquake stress resultants were considered in the load combinations. The combination producing the most critical stress was used in the design. The tornado and tornado generated missile analyses are provided in Section 3.8.1.4. The load combination Case D specifies tornado loads combined with operating loads. The tornado load (Wt) includes the static forces produced by the 360 mph maximum wind velocity t a 3 psi negative pressure and the structural response to the missile impact.
3. 8-11 SGS-UFSAR Revision 6 February 15, 1987
The stresses on any structural member produced by the effective pressure transformed from the tornado wind, the impact of the missile, and also differential pressure were superimposed to obtain the most critical total stress, provided the induced stress from these three components are in the same direction. When one of the components induced an opposite stress, thereby reducing the total stress in the member, it was neglected. In other words, all six loading combinations listed in the Standard Review Plan (SRP) have been considered with factors of 1 instead of 0.5 for Wp in combinations iv and vi and also have taken into account stress directions as stated previously. Hydrostatic loadings from the hurricane condition were applied to the structures to check their stability. The procedures used by our consultant (Dames and Moore) for transferring the static and dynamic flood effects to load were as delineated in the U. S. Army Coastal Engineering Research Center Technical Report No. 4. Total head, including wave effects, was considered to investigate the lateral and overturning effects. Containment flooding for fuel recovery was not a design consideration. The load combinations utilized in the design of the containment and other Category I structures were equivalent to or more ~onservative than those outlined in the SRP. The following tabulations provide a comparison of load combinations utilized with the SRP criteria. 3.8-12 SGS-UFSAR Revision 6 February 15, 1987
CONCRETE CONTAINMENT STRUCTURE Test SRP (D+L) + Pt + Tt Salem *D +/- O.OSD + Pt + Tt Construction Not Critical Normal Not Critical Extreme SRP (D+L) + T0 + Wt + R0 + Pv Environmental (1) Salem
D +/- O.OSD + l.lWt + B + Pb Extreme SRP (D+L) + T + E + R + P 0 0 v Environmental (2) Salem

D +/- O.OSD + T' I I + E' + B Abnormal SRP (D+L) + l.SP +T+R aaa Salem * (D +/- O.OSD + 1.5P + (T a
+ TL) + B Abnormal/Service SRP (D+L) + 1.25P + T + 1.25E + R +y +y aa a rm Environmental Salem *D +/- O.OSD + 1.25P + (T 1 + TL') +L25E+B Abnormal/Extreme SRP (D+L) + P + T + E' + R + Y + Y aaa r m Environmental Salem
D +/- O.OSD + P + (T 1 1
+ TL' ')+E'+B
See preceding pages of this Section for identification of Salem symbols. See SRP Section 3. 8. 3 for SRP symbols.
Although the R and Y forces are not listed in the overall structural analysis load combination formulae, the local effects under piping load, jet load, and missile impingement were taken into account. 3.8-13 SGS-UFSAR Revision 6 February 15, 1987
INTERNAL CONCRETE STRUCTURES (2) SRP 1.4D + 1.71 = 1.9E Salem Not Critical (2b) SRP 0.75 (1.4D + 1.7L + 1.9E = 1.7 T0 + 1.7R) 0 Salem Less Critical Than (5) (3) SRP D + L + T0 + R0 + E' Salem Less Critical Than (6) ( 4) SRP D + L + Ta a+ R + 1.5P a Salem D + L + Ta + Ra + l.SP (5) SRP D + L + T + R + 1. 25P + (Yr + Yj + Ym) a a a
+ 1. 25E Salem D + L + T + Ra + 1.25P + (Yr + Yj + Ym) + 1.25E a (6) SRP D+ L + T + R + P + (Yr + Yj + Ym) + E' aaa Salem D + L + T + R + P + (Yr + Yj + Ym) + E' aa The buoyancy effect of ground water has been included in the assessment of the sliding and overturning potential of the Containment Building and all other Category I structures. The buoyancy effect will reduce the dead weight and thus reduce the factors of safety against sliding and overturning. To include the buoyancy effect in assessing the sliding and overturning potential is the more conservative and correct approach. However, the maximum hurricane, flood, and earthquake are not postulated to occur simultaneously.
The safety against sliding, overturning, and flotation for the Containment Building and all other Category I structures under all loading combinations are within the limits set by SRP 3.8.5. 3.8-14 SGS-UFSAR Revision 6 February 15, 1987
3.8.1.4 Design and Analysis Procedures The containment structure has been analyzed to determine stresses, moments, shear, and deflections due to the static and dynamic loads. 3.8.1.4.1 Static Analysis The containment structure has been analyzed and designed for all loading conditions combined with load factors as outlined in Section 3.8.1.3. Mathematically, the dome and cylinder are treated as thin-walled shell structures which result in a membrane analysis. Since the thickness of the dome and cylinder is small in comparison with the radius of curvature (cylinder 1/15.5, dome 1/20), the stress due to pressure and wind or earthquake can be calculated by assuming that they are uniformly distributed across the thickness. In general, membrane stresses are carried by the reinforcement. Some are carried by the steel liner, but none by the concrete unless they are compressive stresses. Manual analysis of the containment structure, based on "Theory of Plates and shells," by Timoshenko and Woinowsky-Krieger (1) and "Theory of Elasticity," by Timoshenko and Goodier (2), have been performed to obtain shears, moments, and stresses within the structure as the basis of our preliminary design for reinforcements and _J.iner plate. An independent three-dimensional axisymmetric modal analysis using the finite element method was made by Conrad Associates (3) to ascertain that the design of the containment structure was adequate. 3.8-15 SGS-UFSAR Revision 6 February 15, 1987
The manual shell analyses calculations and the "Conrad Associates" design review report are submitted separately. The design includes the consideration of both primary and secondary stresses. The design limit for tension members (i.e., the capacity required for the design load) is based upon the yield stress of the reinforcing steel. The load factors used in the design primarily provide for a safety margin on the load assumptions. The capacity reduction factor "0" is provided for the possibility that small adverse variations in material strengths, workmanship, dimensions, and control, while individually within required tolerances and the limits of good practice, occasionally may combine to result in under capacity. For tension members, the factor "0" is established as 0.95. The factor "0" is 0.90 for flexure and 0.85 for diagonal tension, bond, and anchorage. For the liner steel the factor "0" is 0.95 for tension, 0.90 for compression and shear. The detailed design has been reviewed by Conrad Associates' finite element computer program to verify its safety. Stress values for rebars and liner plates at various locations for all loading combinations involving LOCA are given in Tables 3.8-1 through 3.8-10. The designation of main reinforcement pattern for the containment structure is shown on Figures 3.8-14 and 3.8-15. seismic reinforcing consists of diagonal bars at 45° to the horizontal plane each way, extended from mat to the lower portion of the dome. They are designed to resist the lateral shear under earthquake such that the horizontal component per foot of diagonals will be equal to the maximum value of the shear flow. Although, in the cylinder, the liner and the concrete have some capacity available to resist the seismic shears, no credit was taken for the capacity. Dowel action of the main bars was also neglected. The containment structure has also been evaluated for increase in design loads due to the postulated MSLBs. The evaluation shows that for the design of the containment structures LOCA is the governing condition. 3.8-16 SGS-UFSAR Revision 17 October 16, 1998
Wall Stresses The stresses in the wall reinforcement from the independent check listed in Tables 3.8-1 through 3.8-8 are all under yield pointt except the only location where the diagonal bars are critically stressed is at Elevation 84 feet under load combination (c). Howevert as stated by Conrad Associates (3), the stress indicated at that location as 60.79 ksi was obtained neglecting all contributions of the main meridional and hoop reinforcement to the seismic shear-resisting capacity of the containment wall. An inspection of the stresses incurred by the main reinforcement as a result of forces other than the seismic shear indicates that these bars are markedly understressed in this zone of the containment shell. Thus the stress value of 60.79 ksi in the diagonal reinforcing bars resulting for seismic shear is overestimated. The discontinuity stresses are accounted for in the design. The moments and shears are computed by equating the deformations and angular rotations of the two parts of the structure at the point of juncture and solving for the resulted discontinuity stresses. The total stresses are obtained by adding the discontinuity stresses to the membrane stresses. The moments and shears at the base of the containment wall are determined on the basis of the rigidity of the resulting cracked section, with the steel on the inner face in tension and concrete on the outer face in compression. The compressive stress in the concrete is checked to ascertain that it is less than a. 75 fc. The tension bars are checked to ascertain that the stresses are not more than 0.90 fy. The shears are carried by hooked diagonal radial bars and no reliance is made on the concrete. Additional diagonal radial bars inclined in a direction normal to the shear diagonal bars will be placed in the wall to take care of diagonal tension. In the stress analysis, uncracked section for concrete is found to be more critical in creating secondary bending stresses in the areas of discontinuity. This conservative assumption was used by Conrad Associates to check the design in such areas.
3. 8-17 SGS-UFSAR Revision 6 February 15, 1987
The deformation of the containment is larger if cracked section property is employed. The values obtained from this approach are being used in calculating the relative displacement between the buildings for clearance, assuring that adequate clearance has been provided. The working stress check under operating conditions has been found to be at very low level. The maximum concrete compressive stress under dead load, operating thermal load, and OBE is 835 psi, while the maximum stress in the reinforcement for the same loading combination is 6540 psi. The concentric dome ring was conservatively designed as a tension ring subjected to uniform pull around the periphery. Two sets of l-inch diaphragm plates are used to transfer the tension through the ring to the meridional reinforcements merging at the peak of the dome. The stress level in the cylindrical tube is minimal. The ring plate is made of ASTM A516 Grade-70 pressure vessel quality, ultrasonic testing per ASTM A-435 except with 100 percent coverage. All w~lding conforms to Section III of the ASME Boiler and Pressure Vessel Code. The ring is physically connected to the meridional reinforcement and the liner. I Liner Plate The maximum tensile stress in the liner plate under the test condition is 30.9 ksi, below the minimum yield point of 32 ksi. This is the preoperational artificial pressure test without the accompanied temperature rise. This case induces the higher tensile stress in the liner plate than the design basis accident 3.8-18 SGS-UFSAR Revision 20 May 6, 2003
condition. Under other cases of critical loading combinations involving LOCA the maximum tensile stress in the liner plate is 27.5 ksi, with 14 percent extra safety margin. The maximum interaction coefficient for biaxial compression and shear in the liner plate under critical load combinations involving LOCA is 0.902 with approximately 10 percent extra safety margin. The listed stresses in Tables 3.8-1 through 3.8-9 have already taken account of the capacity reduction factors, 0. In other words, the stresses have been divided by the appropriate 0, 0.95 for tension, 0.90 for flexure, and 0.85 for shear, etc. The combined biaxial compression and shear in the liner plate have been examined by the following interaction formula: where: a , o hoop and meridional stresses in liner plates X Z oxo' azO = maximum allowable stress in hoop and meridional direction (critical buckling stress or the yield stress) T shear stress in the liner plate T maximum allowable shear stress 0 The resulting interaction coefficients for Operating, LOCA, and Test conditions 1 are listed in Table 3.8-10. The containment liner has also been evaluated for the increased containment temperature of 351.3°F and the concurrent pressure due to the postulated MSLBs. The evaluation shows that the liner in the uninsulated portion tends to yield locally at EL.l20 '-0"; however, the total design forces at this local section can 3.8-19 SGS-UFSAR Revision 17 october 16, 1998
be carried by the containment reinforcing steel alone, without using the liner as a strength element. The corresponding strains in the liner at this section are low relative to the allowable liner strain values specified in Table CC-3720-1 of 1995 ASME B&PV Code, Section III, Division 2 (Reference 6) for maintaining leaktight integrity of the liner. Thus, both strength and leaktight integrity of the containment are assured.
s. B. Batdorf and M. Stein in their paper "Critical Combinations of Shear and Direct Stress for Simply Supported Rectangular Flat Plates" (NACA Technical Note 1223, 1947), obtained the critical stress combination for the case of shear and simultaneous uniaxial compressive stresses as:
(t/tO) 2 + a/aO = 1 For biaxial compression, Timoshenko and Gere in their "Theory of Elastic Stability" defined the allowable biaxial compression in the form of: Modifying The Batdorf and Stein expression to include the biaxial effect we have used the following equation to check the interaction stability: 2 (x + y) (t/t0) + 1 (xO +yO) The edge condition was assumed to be simple supported which is more conservative. 3.8-20 SGS-UFSAR Revision 17 October 16, 1998
Base Mat In designing the base mat, the slab is considered to be a circular plate of constant thickness, t. The loads are imposed upon the slab by the exterior cylinder wall, the central circular crane wall and, to a lesser degree, by the equipment. The soil reaction pressure was found in a conventional manner by treating the slab, which is 16-feet thick, as a rigid mat. The mat is then analyzed as a plate subjected to soil pressures and supported by a circular wall symmetrical with respect to the center of the mat. The supporting walls are considered as either simply supporting the mat or partially fixed. The exterior cylinder wall has been considered partially fixing the mat; the crane wall is a simple support. The containment base mat is analyzed as a rigid circular plate subjected to loadings from the axisymmetric exterior cylinder wall, crane wall, interior walls, and equipment acting around an equivalent circle. The soil pressure is found in a conventional manner without the benefit of its elastic deformation. Manual analysis was based on the AC! Paper, Title No. 63-63, "Analysis of Circular and Annular Slab for Chimney Foundation," by Kuang-Han Chu and omar F. Afandi. A finite element program was used to check the rebar under five loading combinations. Since the mat is covered by a 2 to 5-foot thick concrete slab, and also the lower 34-feet of cylinder liner is insulated, the thermal effect on the mat has been neglected. The design of the base mat reinforcement has been reviewed for five load combinations at three different mat sections. The maximum radial, tangential, vertical, and shear stresses at these sections are shown on Figures 3. 8-16 through 3.8-20. The stresses shown in these figures are integrated over the thickness of the slab and transformed to forces per unit length of circumference. These forces are then distributed to the top and bottom reinforcing bars at the section under investigation. The resulting stresses in the bars are all under 30 ksi. The maximum tangential shear under DBE for the interior structure at top of reactor pit is 7600k.. The shear is transmitted through the pit wall at Elevation 76 feet and then bearing against the base mat. The unit shear is 73 psi and bearing is 42 psi, both well within allowable values. 3.8-21 SGS-UFSAR Revision 17 October 16, 1998
Five static load analyses consisting of dead load, buoyancy, internal pressure, thermal, and tornado loadings have been performed for the containment structure. The complete report by Conrad Associates {3) and manual design calculations are kept on file by Public Service Electric & Gas (PSE&G). They are summarized as follows. Dead Load Analysis Finite element model is used to perform the dead load analysis. Static secant moduli are used in representing the soil stiffness under dead load. For the vertical load, only horizontal restraints are imposed at side boundaries of the soil system. Stresses, moments, and shears at containment wall and mat are shown on Figures 3.8-21, 3.8-22, and 3.8-23. Buoyancy Analysis Normal ground water table for the site is at Elevation 96 feet. For design purpose it is considered to be at 6 inches above the plant grade level, Elevation 99 feet-6 inches. Under hurricane condition the water level could be expected to rise to Elevation 120.4 feet; however, since the direction of the hydrostatic pressure is so small it does not create a critical loading combination. The result of the buoyancy-induced stresses in the containment vessel are very small and are confined to the lower portion of the structure. No plot is given because it does not affect the design. Internal Pressure Analysis The internal pressure transients used for the containment design and its variation with time are shown on Figures 3.8-11 through 3.8-13. For the free volume of 2,620,000 cubic feet within the containment, the design pressure is 47 psig. The maximum temperature at the uninsulated section of the liner under the accident condition is 246°F. For 1.25 times and 1.5 times design pressure loading conditions, the corresponding liner temperatures are 285°F and 306°F, respectively. Static pressure loads are used in design, since the pressure increase is very gradual from the transient curve. I SGS-UFSAR 3.8-22 Revision 17 October 16, 1998
Thermal Analysis The thermal gradients in the containment wall under operating and accident conditions are shown on Figure 3. 8-24. Both loadings are analyzed for the containment structure. The analytical model employed by Conrad Associates ( 3) for finite element thermal analysis is an axisymmetric assemblage of solids of revolution. Each segment across the containment wall consists of ten elements to represent the thermal gradient through the wall thickness. Orthotropic material properties are used to represent the variable shell area in the hoop and meridional directions. Due to the one-dimensional nature of the reinforcing bars, Poisson's ratio was set equal to zero in the plane of the equivalent steel shell. For accident loading, the concrete is assumed to be totally cracked in the hoop and meridional directions, but uncracked in the radial direction. For operating loading, concrete is assumed to be uncracked. The liner plate is modeled as a thin isotropic steel shell with an elastic modulus of 28,000 ksi and a Poisson's ratio of 0.3. Between the liner plate and the concrete containment shell a thin element, 0.01-feet thick, is introduced to facilitate modeling of the discontinuity in temperature occurring at the liner-to-concrete interface under accident conditions. A fixed boundary is introduced at the foundation mat. Thermal stresses and strains are not likely to develop in the thick mat which has excellent insulating properties. 3.8-23 SGS-UFSAR Revision 6 February 15, 1987
Stresses under operating and accidental thermal loadings involving LOCA are shown on Figures 3.8-25 through 3.8-30. Tornado and Tornado Generated Missile Analysis Three tornado wind distributions were investigated in the Category I structural design as shown on Figure 3. 8-31. In combination with the static forces produced by the 360 mph maximum wind, a 3 psig atmospheric pressure drop was specified for the containment structure. Evaluations of structural adequacy against tornado wind loads and tornado missiles are given in sections 3.3.2 and 3.5.2, respectively. 3.8.1.4.2 Dynamic Analysis The containment structure seismic analysis was performed through (a) lumped mass model manual analysis, using average response spectra ground input, and (b) a finite element modal analysis, using time history ground input. The detailed report from conrad Associates (3) and the independent manual calculations are kept on file by PSE&G. The computer analysis yields a slightly higher result in accelerations, shears, and moments in comparison with the manual analysis. The most conservative results are used in design. The seismic analysis of the containment structure by the finite element method is performed by computer using a step-by-step direction integration procedure. Studies have been made to establish free field soil boundary condition. The model used in the analysis is shown on Figure 3.7-13. The El Centro ground motion of May 18, 1940, was recommended by Dames and Moore as the most appropriate motion for the site. Its 3.8-24 SGS-UFSAR Revision 17 October 16, 1998
peak horizontal acceleration was normalized to 0.10 g and 0.20 g for OBE and DBE, respectively. Two-thirds of the above-mentioned values are used for vertical ground motions, and they are considered to be acting simultaneously with the horizontal ground motion. Modified Hausner's average response spectra, as shown on Figures 3.7-1 and 3.7-2, are used for normal modal analysis. Seismic design criteria and procedures for structures are described in Section 3.7. For the DBE, a damping factor of 5 percent of critical damping is used for analysis for structure and soil. Similarly for the OBE, a damping factor of 2 percent is applied for both structure and soil. Two separate modal analyses, horizontal and vertical motions, are performed and their results superimposed. The acceleration time histories from the result of the structural seismic analysis are used for the generation of horizontal and vertical response spectra at specified floors or locations for equipment of seismic design. They are presented in the Conrad Associates' report and kept on file by PSE&G. Total accelerations, peak displacements, and the envelope of forces in the containment structure under DBE and OBE conditions are shown on Figures 3.7-3 through 3.7-12. Clearances between Category I buildings and adjacent structures are checked based on the relative displacement at various building elevations under seismic and design basis accident loadings to assure that the required separations are maintained. 3.8-25 SGS-UFSAR Revision 6 February 15, 1987
3.8.1.5 Structural Design and Acceptance Criteria The containment structure is designed to meet the following design criteria stated in the "General Design Criteria for Nuclear Power Plant Construction Permits." Reactor containment shall be provided. The containment structure shall be designed (a) to sustain without undue risk to the health and safety of the public, the initial effects of gross equipment failures such as a large reactor coolant pipe break, without loss of required integrity and (b) together with other engineered safety features as may be necessary, to retain for as long as the situation requires, the functional capability of the containment to the extent necessary to avoid undue risk to the health and safety of the public. The reactor containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the leakage of radioactive materials from the containment structure under conditions of pressure and temperature resulting from the largest credible energy release following a LOCA, including the calculated energy from metal-water or other chemical reactions that could occur as a consequence of failure of any single active component in the Emergency Core Cooling System (ECCS), will not result in undue risk to the health and safety of the public. The containment structure design parameters are based on the following:
1. Leak tightness and testing requirements

2. Seismic requirements

3. Tornado requirements 3.8-26 SGS-UFSAR Revision 6 February 15, 1987

4. Shielding requirements

s. Design basis accident requirements

6. Flood conditions due to maximum probable hurricane

7. Internal missile generation The stresses of concrete, reinforcing steel, and liner plate under various loading combinations are as described in Section 3.8.1.4.
The containment integrity evaluation, including the containment pressure transients and safety margin, are presented in Section 15. 3.8.1.5.1 Fracture Prevention of Containment Pressure Boundary The containment pressure boundary parts, which do not rely on concrete structures to provide the pressure retaining capability, are constructed in accordance with the material, design, fabrication, and installation requirements of the ASME Code, Section III, 1968 Edition. The Code requirements took into consideration procedures for prevention of brittle failures and fracture propagations in containment pressure boundaries. These procedures include Charpy V notch tests of plate materials, sufficient margins in the design allowables, preheat of steel plates, and postweld heat treatment of penetration assemblies. 3.8.1.6 Materials. Quality Control, and Special Construction Techniques 3.8.1.6.1 Liner Plate A welded steel liner of thicknesses varying from 1/4 inch to 1/2 inch is anchored to the inside face of the concrete shell with 3.8-27 . SGS-UFSAR Revision 6 February 15, 1987
1/2-inch diameter studs to ensure containment leak tightness. This containment liner is designed to carry a portion of the membrane force from the different combinations of loading; however, for conservatism it is not counted on in the resistance to lateral shear. The out-of-roundness tolerance of the liner shall not exceed plus or minus 2 inches from the true diameter of 140 feet. The lower 34 feet of cylinder liner is insulated, except locally around liner penetrations and around interferences with other commodities, to prevent buckling of the liner due to restricted growth under a rise in temperature. The membrane tension and the combined stress of biaxial compression and shear in the liner plate are described in Section 3.8.1.4.1. Our computations for the liner plate indicate that there would be no inelastic buckling of the plates. Under stress, the variation in plate thickness would cause small differential movements between the liner and the concrete. Also, the shrinkage cracks in the concrete would have the same result. Soft corks are placed around the studs adjoining the liner plate to allow differential movement between the liner and the concrete. The stud anchors are designed such that their failure in shear or tension will not break the leak tight integrity of the liner plate. Tests will be made to verify this criterion. Even if stud failure developed, it would be random in nature. This would not impair the liner integrity, nor would it cause progressive failure. The design load per anchor is low, and if an anchor should fail, the load it would have carried would be easily distributed to the adjacent anchor. Tensile and shear tests were conducted on the liner plate studs. Three tensile and three shear test assemblies approximating as 3.8-28 SGS-UFSAR Revision 17 October 16, 1998
close as possible the welded studs in service, were fabricated as test specimens. The results of the tests indicate that the studs pulled away from the liner plate at a tensile stress of between 74,500 psi and 80,600 psi. Under shear loading the studs sheared off between 62,600 psi and 67,000 psi. In both failure modes the leak tight integrity of the containment liner plate was not affected. Each liner plate splice in the dome, cylinder, and mat is covered by a steel channel. The steel channels are embedded in the concrete mat. To prevent any possible shearing of the channels from the differential movement between the liner plate and the inner concrete slab, they are isolated from the concrete by 1/4 inch of asphalt impregnated expansion material, and Styrofoam all around. Where there are a large number of penetrations in one area, the thickness of the liner plate is increased from 3/8 inch to 3/4 inch for reinforcement. The original intent of the steel channels was for leak testing the liner welds. However, leak testing will be performed in accordance with 10CFR50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," instead of pressurizing the liner weld channels. The leak testing program and procedures are described in Section 6.2.1. The 3/4-inch knuckle plate connects the cylinder liner to the base liner. The thicker plate is used to resist buckling due to concentrated loadings from liner anchors in the base mat and also to take care of the warped surface created by the double curvature at the junction. The detail of the anchor plate is shown in Section "1-1" of Plant Drawing 201175. Tension anchors to transfer the uplift force for essential pieces of equipment to the mat are also shown in Sections "X-X" and "5-5" of Plant Drawing 201175. Where there is a shear load in combination with the tensile load, as there would be in the case of an earthquake, the shear load will be transmitted into the 2-foot thick or 5-foot 3.8-29 SGS-UFSAR Revision 27 November 25, 2013
thick concrete slab located above the liner plate by shear lugs attached to the equipment base plates. The transfer of shear load from inner structures through the bottom liner plate of the containment is by means of the reactor well acting as a key. The inside surface of the liner plate in the cylin<.;:ler and dome is painted with catalyzed epoxy paint. Surfaces treated in this manner can be easily washed for decontamination. Keeler and Long No. 7475 and No. 7844 epoxy was specified as the protective coating based on their test-proven ability to withstand design basis accident and washdown conditions. 'l'he coating, during test, had been exposed to the pressure~temperature-time cycles for design basis accident condition with satisfactory results. The spray solutions used in the tests are similar to the chemical compositions proposed for Salem but of higher concentration. The protective coating also behaves well in the steam environment, as reported by the Franklin Institute Research Laboratory. The direct jet impingement with jet I temperature above 300°F will cause disintegration of the coating in that local area, which can be repaired after the accident. A train of strainer modules has been connected to the sump at the bottom of the containment for retaining any disintegrated particles and to eliminate any possibility of flow blockage . Fouling of engineered safety features as a result of the local failure of the coating is not considered possible. 3.8,1.6.2 Base Mat The design of the base mat is described in Section 3.8.1.4.1. The base mat was poured on top of lean concrete fill in circular segments, as described in Section 3.8.1.6.8.7, with only vertical SGS-UFSAR 3.8-30 Revision 23 October 17, 2007
construction joints. The base mat is poured to a level 6 inches below the final elevation of the bottom liner plate. The backing tees are then positioned and concrete poured to a level flush with the top of the backing tees.
Steam generators and reactor coolant pumps are supported by heavy welded steel frames embedded in the concrete and tied down deep into base mat by 6-inch diameter and 4-inch diameter bolts, 18 feet-6 inches long to prevent the tremendous uplift during pipe rupture accident. (See Figure 3.8-32.) 3.8.1.6.3 Cylinder Wall The design of the cylindrical wall is described in Section 3.8.1.4.1. The wall pours are made in lifts 4 to 5 feet in height. 3.8.1.6.4 Dome The design of the dome is described in Section 3.8.1.4.1. The lifts in the dome are approximately 3 to 5 feet in height and each lift is poured continuously with no joints parallel to the liner plate allowed. Near the top of the dome, terminations of the lifts are horizontal rather than normal to the liner plate. For arrangement of the dome line and the top enclosure ring detail, see Plant Drawing 201181. 3.8.1.6.5 Penetrations and Openings For a description of various containment wall penetrations, hatch openings, their details and basis for design and analysis, see Section 3.8.1.6.8.8. 3.8-31 SGS-UFSAR Revision 27 November 25, 2013
The piping penetration sleeves were fabricated to applicable portions of the ASME Boiler and Pressure Vessel Code, Section III, Nuclear Vessels, 1968, with the exception that there was no requirement made to stamp the pipe with the "N" symbol. Inspection of the piping penetrations was in accordance with the above indicated code except that hydr-ostatic test of rolled and welded pipe to ASTM AISS was not performed at the mill. The plate for this pipe, however, was ultrasonically examined and welds were completely radiographed. In addition, the sleeves were pressure tested with the containment as well as pneumatically leak tested internally. Cooling, by both free and forced convection, is provided where necessary to maintain concrete temperatures adjacent to hot pipe penetrations below 150°F. A hot pipe passing through the containment wall can transfer heat to the wall via any or all of three paths. These paths, shown on Figure 3.8-33, are:
1. Radial conduction in the pipe cap and longitudinal conduction through the expansion joint and along the penetration sleeve outside the containment (Path A).

2. Radial conduction in the pipe cap and longitudinal conduction along the penetration sleeve inside the containment (Path B).

3. Radial conduction through the insulation within the penetration (Path C).
The quantity of heat transferred via Path A is inconsequential due to the high thermal resistance presented by the thin cross section of the expansion bellows. 3.8-32 SGS-UFSAR Revision 6 February 15, 1987
The quantity of heat which could be transferred via Path B is significant for some penetrations, i.e. it could cause localized containment concrete temperatures to rise above acceptable limits. As such, annular heat transfer fins (extended surfaces) are provided where necessary. These fins serve to dissipate sufficient heat to the containment atmosphere by natural convection to maintain acceptable temperature in the wall. The fins are designed to dissipate the Path B heat load without the aid of any other cooling mode. The potential heat transfer via Path C can also be significant. As the magnitude of natural heat dissipation in the containment wall is not sufficient to cause a large enough steady state temperature drop in the insulation within the penetration assembly, other means are required to remove the heat and maintain the desired concrete temperature. The heat is removed by compressed air flow in plate-type heat exchangers (coolers) installed within the penetration sleeves. Protection against loss of cooling capability is provided by both the inherent "reliability" of the free convection mode and by redundant compressed air supply lines as shown on Figure 3.8-34. It has been shown that for constant exposure of concrete to temperatures up to 150°F, the loss in strength is quite small; and for temperatures as high as 500°F to 600°F, the deterioration in structural properties is tolerable. Considering the redundancy in air supply lines, the only cause of loss of penetration cooling would be complete loss of the station air compressors, a condition which would not be permitted to persi!il long enough to cause significant localized concrete deterioration. 3.8.1.6.6 Polar Crane The polar crane is described in Section 9. 1. 3.8-33 SGS-UFSAR Revision 6 February 15, 1987
3.8.1.6.7 Missile Protection High pressure RCS equipment which could be the source of missiles is suitably shielded either by the concrete shield wall enclosing the reactor coolant and pressurizer loops or by the concrete operating floor to block any passage of missiles to the containment walls even though such postulated missiles are deemed most improbable. Protection against internally generated missiles is described in Section 3. 5. I. 3.8.1.6.8 Construction Procedures and Practices 3.8.1.6.8.1 Codes of Practice Materials and workmanship conformed to the following codes and specifications: ACI 318-63 "Building Code Requirements for Reinforced Concrete" ACI 301-66 "Specification for Structural Concrete for Buildings" ACI 613-54 "Reconunended Practice for Selecting Proportions for Concrete" ACI 614-59 "Reconunended Practice for Measuring, Mixing and Placing Concrete" ACI 347-63 "Reconunended Practice for Concrete Formwork" ACI "Manual of Concrete Inspection" - 1957 3.8-34 SGS-UFSAR Revision 6 February 15, 1987
Security-Related Information - Witheld Under 10 CFR 2.390 ASME Boiler and Pressure Vessel Code, 1968: Section III "Requirements for Class B Vessels" (penetrations and hatches only) Section VIII "Requirements Pertaining to Methods of Fabrication of Unfired Pressure Vessels" Section IX "Welding Qualifications" AISC "Manual of Steel Construction," 6th Edition or later edition, as applicable ACI 301-66, "Specifications for Structural Concrete for Buildings," together with ACI 318-63 "Building Code Requirements for Reinforced Concrete," form the basis for the PSE&G concrete specifications. 3.8.1.6.8.2 Concrete
3.8-35 SGS-UFSAR Revision 6 February 15, 1987
Preliminary Tests The PSE&G Testing Laboratory obtained samples of the aggregates to be used in the concrete for preliminary testing and approvaL Testing methods and acceptance standards were as follows: Acceptance Test Method Standards Sampling ASTM 075-59 ASTM C-33 Gradation - Sand ASTM Cl36-63 ASTM C-33 Gradation - Stone ASTM C136-63 ASTM C-33 Sodium Sulfate ASTM C88-63 ASTM C-33 Soundness Loss Angeles Abrasion - ASTM C131-66 ASTM C-33 Stone Material Finer than No. 200 Sieve ASTM C117-66 ASTM C-33 Organic Impurities - ASTM C40-66 ASTM C-33 Sand Potential Reactivity - Chemical Method ASTM C289-6S ASTM C-33 In addition, the following tests were performed to give necessary information concerning the aggregates. Test Method Fineness Modulus ASTM C125-66 Unit Weight ASTM C29-60 Specific Gravity ASTM CI27-59 Absorption ASTM C128-59 3.8-36 SGS-UFSAR Revision 6 February 15, 1987
The coarse aggregate selected and used on the Salem Project was quarried stone, crushed and graded to meet the detail specifications. The stone, commonly known as traprock, was a basic igneous rock consisting of diabase and basalt. The quarries and crushers were located in Lamberville, Pennington, and Kingston, New Jersey. The fine aggregate selected was known locally as Dorchester sand. It was a silica sand found in bank run deposits. The sand was dredged, washed, and then graded to meet project deta i 1 specifications. The Portland Cement (Type II) used conformed to ASTM Specification C-150, latest edition. Flyash was used as an admixture in the majority of the concrete and conformed to ASTM Specification C-350-65T, except that the fineness of the flyash was in accordance with the ASTM Specification C-618-68T, which has not replaced ASTM C-350. A retarding densifier was also used as an admixture which conformed to ASTM Specification C-494, Type D. The retarder was a water reducing admixture of the hydraxylated carbolic acid type and contained no calcium chloride. Trial mixes were made by the PSE&G Testing Laboratory with the above ingredients in accordance with ACI 301-66, Section 308 - Method 2. Proportions of ingredients were determined and tests conducted in accordance with ACI 613-54, "Recommended Practice for Selecting Proportions for Concrete. 11 The concrete mixes used for construction were approved by the PSE&G Structural Engineering Division and specified in the project detail specification for concrete. The dry density of the concrete mixes used for construction exceeded 144 lbs/cu ft. All concrete mixes used in the work were fully documented. For structural concrete, the maximum allowable slump for concrete placed was 4 inches. In areas with closely spaced reinforcing bars, the detail 3.8-37 SGS-UFSAR Revision 6 February 15, 1987
specification allowed the use of a concrete mix with a coarse aggregate of 3/8 inch and a maximum slump of 5 inches. For the reactor containment wall. in the area adjacent to the equipment and personnel hatches where additional reinforcing steel was specified, a more plastic mix was designed for adequate concrete placement. In this case, the slump was increased to be between 6 and 7 inches. For fill concrete, one batch in ten could have a slump of up to 5 inches with the majority of concrete placed at 4 inches slump or less. Seven-inch slump concrete was used only in the area of the equipment and personnel hatches where reinforcing steels are crowded. The water-cement ratio was 6.25 gallons per bag; 7.5 bags of cement were used per cubic yard of concrete. One hundred pounds of flyash per cubic yard of mix was added. It is believed that the erosion resistance of this specific mix is as good as the regular low slump mix. After the form was removed, the surface of the concrete appeared to be smoother and without visible cracks, due to the workability of the mix. The flyash contains no calcium chloride to cause corrosion. The 1970 edition of "Concrete Industries Year Book" states that concrete made with flyash is more resistant to weak acids and sulfates, which cause corrosion. Batch Plant The bulk of the concrete for the project was supplied from a batch plant at the site operated by United Engineers and Constructors, Inc. Technical details of this plant are as follows:
1. Eric Strayer central mix concrete plant t rated at 240 cubic yards/hourt although the maximum rate of concrete produced was 180 cubic yards/hour 3.8-38 SGS-UFSAR Revision 6 February 15, 1987

2. Four compartment aggregate bin

3. Nine cubic yard aggregate hatcher weighing aggregates cumulatively and automatically with a dial scale

4. 1,500 barrel bin divided into two compartments for flyash and cement

5. Cement and flyash hatcher weighing cumulatively and automatically with a dial scale

6. Water and ice hatcher weighing cumulatively and automatically with a dial scale The plant provided fully automatic wet hatching for the various mixes required. The operator inserted the proper card for the mix required, set a dial for the quantity of concrete desired and the machine measured out the ingredients automatically and recorded the weight automatically. Weight measurements were also visually observed by the Quality Control Inspector at the control console on three separate 2-foot diameter indicating dials to check and confirm the recording tapes. The hatching accuracy of the weighing equipment was within +/-1 percent of the true values. The weighing equipment was calibrated prior to initial use. Standard weights were used for periodic calibrations. The calibrations were made quarterly or every 40,000 cubic yards poured, whichever occurred first. Moisture probes embedded in the aggregate binds determined moisture content and compensations were made to maintain the proper water-cement ratio.
During cold weather, the temperature of the concrete was controlled by heating the mixing water and heating the aggregate bins. During hot weather, the temperature of the concrete was controlled by cooling the mixing water. During extremely hot weather, flaked ice was added to the mix. The flaked ice and water was weighed separately, but cumulatively in a compartmented weigh hopper. 3.8-39 SGS-UFSAR Revision 6 February 15, 1987
During hot weather the temperature of the concrete as placed was not more than 80°F. During cold weather, when the mean daily temperature fell below 40°F, the temperature of the concrete as placed was not less than 50°F. These procedures were in accordance with the detail concrete specification. During concrete operations, the batch plant inspector verified the mix proportions of each batch of concrete and ascertained that samples were taken and tests were made of the concrete ingredients. The batch plant inspector verified that the mixed proportions complied with those of the design mixes with the water content modified as required by measurement of surface moisture on the aggregates. The batch plant inspector also prepared a daily report to document for each batch the following: mix number, mixer cycle time, weight of each ingredient (including ice), and the batch number. Truck dispatch tickets for each batch showing the time of discharge from mixer, concrete mix number, load number, total water content, and location where used, were prepared by the inspector. Placement Distribution The majority of the fill concrete was distributed directly from the concrete batch plant to the point of placement via conveyors. The longest mn was approximately 1, 000 feet and consisted of two 250 foot belts; four 100 foot belts, and two 50 foot belts. During adverse weather conditions the conveyor belts were covered with metal hoods. To ascertain the concrete integrity from the batch plant to the point of placement, slump tests, temperature measurements, and test cylinders were made from the same batch of concrete at the beginning of the belt and at the end of the belt. These tests showed no changes in strength and no significant change in slump 3.8-40 SGS-UFSAR Revision 6 February 15, 1987
and temperature. The fill concrete was then slumped at the beginning of the conveyor belts and test cylinders with accompanying slump tests made at the beginning of the belt. The distribution point inspector did the following:
1. Visually checked each batch and estimated the slump

2. Notified pour site inspector by field telephone of the quantity of concrete conveyed from the batch plant.
This was done so the pour site inspector would know when to make concrete test cylinders and perform other associated tests. For the majority of the structural concrete pours, the concrete was distributed with standard transit mix trucks which served only to transport and agitate the concrete to keep it plastic. The trucks were loaded at the batch plant from a holding hopper via a short conveyor. The distribution inspector visually checked each batch on the conveyor and estimated the slump. He also prepared a truck batch ticket showing the batch number, the time, location to be used, concrete mix code, and the total amount of water in mix. For structural pours, slump tests and test cylinders were made at the location of the pour by PSE&G Test Laboratory personnel. The PSE&G Testing Laboratory Inspector performed the following for all concrete poured:
1. When necessary to add water to truck-delivered concrete for workability, the batch ticket was checked to determine how much water (if any) could be added and assure that water was added in accordance with the following limitations:

a. Maximum slump was not exceeded 3.8-41 SGS-UFSAR Revision 6 February 15, 1987

b. Total water (including that added at the pour site) was not exceeded by more than 1 gallon per yard, the amount specified in the design mix to produce a maximum allowable slump. Mixer was rotated at least 30 revolutions (at mixing speed) after addition of water. However, in no case did the total revolutions of the mixer exceed 300.

c. The total water in the mix (including water added to the truck at the pour site) was shown on the Slump Test Report

d. Water added at the pour site was added within 45 minutes after hatching

2. Every 20 cubic yards of concrete at each pour location was checked for slump following the procedures of ASTM Cl43-66 and results recorded. Loads with higher than allowable slump were rejected.

3. Determined the temperature of the concrete each time a slump test was made and recorded the results. Loads were rejected when temperatures required by the detail specifications were not met.

4. Made one set of concrete test cylinders for curing (6 inches by 12 inches) per ASTM C31-66 daily for each 100 cubic yards or portion thereof placed per class of concrete. A set of cylinders consists of 6 cylinders.
Concrete cylinders were cured initially in accordance with Section 9 (a) of ASTM C31-66. Concrete cylinder molds conformed to the requirements of ASTM C470-65T.
5. From each load of concrete sampled for the preparation of concrete cylinders, a slump test and a temperature check was made.
3.8-42 SGS-UFSAR Revision 6 February IS, 1987
6. Prepared daily reports of field concrete poured which contained the following information:

a. Date

b. Location of pour (portion of structure)

c. Class and quantity of concrete placed

d. Number and identification of test cylinders made

e. List of concrete batches tested with the time, water added at pour site, if any, slump and concrete temperature The concrete cylinders made by the PSE&G Testing Laboratory Inspector, after sufficient field curing, were transported to the Salem Job Laboratory for stripping, curing, and capping in accordance with ASTM Cl92-66. Two cylinders from each set were tested at age 7 days; three at age 28 days. If the results of the 7 and 28 day tests caused concern that the concrete did not meet specification requirements, the remaining cylinder was saved and tested at age 90 days or as directed by the Structural Engineer.
Otherwise, it was discarded. Compression tests of concrete cylinders were made in accordance with ASTM C39. In addition to the compression tests, the density of the concrete was measured from the test cylinders and recorded. Concrete strength tests were evaluated by the PSE&G Structural Division, Electric Engineering Department, in accordance with ACI 214-65 and ACI 301-66, Chapter 17. If any tests for individual cylinders or group cylinders failed to reach the specified compressive strength of the concrete, the Structural Engineer was immediately notified to determine if further action would be required. 3.8-43 SGS-UFSAR Revision 6 February 15, 1987
Statistical quality control of the concrete was maintained by a computer program. This program analyzed compression test results in accordance with methods required by ACI 214, "Recommended Practices for Evaluation of Compression Test Results of Concrete." The computer results of the data analyzed included normal frequency distribution curves, standard deviations, and coefficients of variation. Placing of concrete was by bottom dump buckets, concrete pumps, or by conveyor belts. Bottom dump buckets did not exceed 3 cubic yards in size. The discharge of concrete was controlled so that concrete could be effectively consolidated around embedded items and near the forms. Vertical drops greater than 5 feet for any concrete were not permitted except where suitable equipment was provided to prevent segregation. All concrete placing equipment and methods were subjected to the approval of the Resident Structural Engineer. The surface of all construction joints were thoroughly treated to remove all laitance and loose aggregate. The construction joint surfaces in the reactor containment vessel, including all the exterior walls, were roughened to expose the coarse aggregate by cutting the surface with stiff brooms or by cutting with an air-water jet after the initial concrete set had occurred, but before the concrete had reached its final set. After cutting, the surface was washed and rinsed. Where in the opinion of the Resident Structural Engineer, the use of air-water jet or brooming as above was not advisable in a specific instance, that surface was roughened by using either hand tools or other satisfactory means to produce the requisite surface. Before placing subsequent concrete lifts, the surfaces of all construction joints were thoroughly cleaned and wetted and all 3.8-44 SGS-UFSAR Revision 6 February 15, 1987
excess water was removed. Horizontal joints were then covered with a minimum of 1/4-inch thick sand/cement grout and new concrete was then placed immediately against the fresh grout. The water-cement ratio of the grout did not exceed that of the concrete itself. Where grouting was not feasible in some areas, a bonding compound such as Colma-Fix, manufactured by Sika Chemical Corp., was used instead, on top of dry cleaned concrete. Vertical joints were wetted and slushed with a coat of neat cement immediately prior to placing the next pour. Curing and protection of freshly deposited concrete conformed to ACI 301, Chapter 12, using an absorptive material with a waterproof covering and sprinkling at intervals necessary to prevent drying for 3 days. The waterproof coverings remained in place for 7 days after the pour. Also, curing compounds, conforming to ASTM C-309, were used as required. The following select sampling and testing was made of the concrete ingredients:
1. Cement was sampled from each silo used to ascertain conformance to ASTM C-150-67 for Type II cement. The cement manufacturer also supplied certified mill reports for each silo of cement used. The storage environment effects were tested in accordance with ASTM-C-0109 and C-266. Vicat apparatus (ASTM C-191) was specified to determine the time of setting of hydraulic cement in the Preliminary Safety Analysis Report. However, it is believed that for our purpose, the Gillmore Test (ASTM C-266) is a more stringent test, in that an approximation of both initial and final set is obtained, while the Vicat Test is only addressed to initial set.
For that reason, the Gillmore Test was actually conducted in the field. 3.8-45 SGS-UFSAR Revision 6 February 15, 1987
2. All concrete aggregates were delivered to the site by truck and each load was visually inspected. Also, every 250 tons received was tested for gradation and determination of fineness modulus. In addition, for sand, an organic impurity test was conducted with the gradation test. These tests were conducted per test methods and acceptance standards to ASTM C-33 with the exception that for sand gradation, the requirement for the percentage of fines was decreased.

3. Flyash from each storage bin at the source was sampled and tested in accordance with C-350-65T using acceptance standards of ASTM 618-68T, which now replaces C-350-6ST, with sampling and testing frequency as follows:

a. Weekly - three composites from daily samples were checked for a carbon and surface area.

b. Monthly -a composite was taken from the weekly composite or a completed chemical or physical analysis.

4. Mixing water (including ice) was checked monthly to assure that it did not contain more than 100 ppm each of chlorides, sulfides, and nitrates and that the turbidity did not exceed 2, 000 ppm of suspended solids content or 25 Formaxine Turbidity Units.
Due to economic and efficiency considerations, the relatively small amounts of concrete necessary to complete the Salem Station may have been obtained from the batch plant at the Hope Creek Generating Station site. 3.8-46 SGS-UFSAR Revision 6 February 15, 1987
Security-Related Information - Witheld Under 10 CFR 2.390 The mixes selected for use at Salem are approved for use in Category I (seismic) structures at Hope Creek and meet the following minimum requirements:
3.8.1.6.8.3 Reinforcing Steel Material Reinforcing steel was required by specification to conform to the following for testing methods and acceptance standards. ASTM A-432-65 "Standard Specification for Deformed Billet Steel Bars for Concrete Reinforcement," with a Minimum Yield Strength of 60,000 psi and a Minimum Tensile Strength of 90,000 psi. ASTM A-408-65 Standard Specification for Special Large Size Deformed Billet Steel Bars for Concrete Reinforcement," with a Minimum Yield Strength of 40,000 psi. ASTM A-615 (Grade 40) "Standard Specifications for Deformed and Plain Billet-Steel Bars for Concrete Reinforcement," with a Minimum Yield Strength of 40,000 psi and a Minimum Tensile Strength of 70,000 psi. Reinforcement bars of the above ASTM designations also conformed to ASTM A-305-65 "Minimum Requirements for the Deformations of Deformed Steel Bars for Concrete Reinforcement." In addition to the ASTM requirement as to chemical composition of A-432 bars, the specification required that the carbon and manganese content did not exceed 0.45 percent and 1.30 percent, 3.8-47 SGS-UFSAR Revision 6 February 15, 1987
respectively, in order to assure better bending properties. Also, the specification required that 148 and 188 bars be subjected to 90 degree bend tests using a pin with a diameter eight times the diameter of the bar being bent, to check ductility. Certification of physical properties and chemical content of each heat of reinforcing steel delivered to the jobsite was required from the steel supplier. In addition "users' tests" were performed by a testing laboratory to confirm compliance with physical requirements and verification of mill test results. Two specimens were taken for each 25 tons or less of the full heat of steel. No sample was selected from the end 12 inches of any bar. The test was performed to determine yield point, ultimate strength, and percentage elongation. If test results did not meet specification requirements, the heat of steel was resamp1ed, this time selecting four specimens instead of the two required originally. If any specimen of the second sampling failed to meet the requirements of the specification, the entire heat was rejected. At the jobsite, reinforcing steel was kept separated by size, heat, and area to be used. At the fabricator's shop and storage area, the reinforcing steel was kept identified by size and heat. Also, when loaded for shipment from the mill, the bars were properly bundled by size, heat, and tagged with the manufacturer's identification number. Reinforcing steel for the dome, cylindrical walls, and base mat of the containment was high-strength deformed billet steel bars conforming to A8TM A-432-65. For the internal concrete of the Reactor Containment vessel, the majority of the reinforcing steel required was ASTM A-15-65, and for the large bars, ASTM A-408-65. In isolated cases, the drawings called for ASTM-A-432-65 for the internal structure. 3.8-48 8G8-UF8AR Revision 6 February 15, 1987
Placing Placing of reinforcing steel conformed to the requirements of Chapter 5 of ACI 301, "Structural Concrete for Buildings, 11 and -- Chapter 8 of ACI 318, "Building Code Requirements for Reinforced Concrete." No tack welding to A-432 reinforcing bars was allowed. Splices All splices of main load carrying reinforcing steel in the Reactor Containment shell were made by the cadweld process using type "T" sleeves to develop the minimum ultimate tensile strength specified by the ASTM for the grade of the bar being spliced. To ensure the integrity of the cadweld splices, the detail specification required random sampling of splices in the field. The selected splices were removed and tested to the minimum tensile strength of the bar being spliced. In some cases, the drawings required bar sizes No. 11 and smaller to be spliced by the cadweld process. In a few instances, the drawings specified other than type "T" sleeves which were required for the splicing of reinforcing to special sections. A type "B" sleeve was used to join main load carrying reinforcing bars to structural steel in order to develop the same minimum ultimate tensile strength of the bar. The detail specification required the average value of all cadweld splices tested to equal or exceed the specified minimum ultimate tensile strength of the ASTM grade of bar being spliced. In addition, no more than 5 percent of the splices tested had an ultimate strength less than 85 percent of that specified by the ASTM for the grade of bar being spliced. If any of the foregoing requirements were not satisfied, production was halted until the cause and extent of the defective splices was determined. 3.8-49 SGS-UFSAR Revision 6 February 15, 1987
Quality control of the splices was maintained by three independent procedures as follows:
1. Each crew in a program including which was by the Erica Products of Ohio. Prior to the splicing of the reinforcing bars, each operator or crew three splices for each of the positions used in production \..Jork.
These samples were then tested to assure conformance with the specifications.
2. Visual inspection of every splice was made by a Quality Control The inspectors to this job attended the same progr.am as the An manual containing the recommendations of the manufacturer was issued to guide the inspector in his judgment of a satisfactory Any splices judged to be in doubt as to integrity were cut out and replaced.

3. Test splices were made by having 3-foot splices produced in sequence with the production bars. These splices were tensile tested for each crew as follows: one of the first 10 , three of the next 100
~~~~~~~, and two of the next and units of 100 In one was randomly cut out and tested for every 100 production made by each crew.
Should any splice tested fail at a value less than the tensile strength required for the bar, then the splice made by the same splicer immediately preceding or following the substandard splice was cut out and tested. If this second test splice did not meet the requirements all work by this splicer was stopped and five adjacent made by the splicer were cut out and tested. If any of these an evaluation was made during which time the crew discontinued 3.8-50 SGS-UFSAR Revision 25 October 26, 2010
Also, the man who made these splices was required to requalify before performing any further production splices. Should the five splices meet the test requirements, the process was considered to be in control. In addition to the above requirements, the following procedures were used to assure acceptable splices:
1. The splice sleeve, powder, and mo] ds were stored in a clean dry area with adequate protection from the elements to prevent absorption of moisture.

2. Each splice sleeve was visually examined immediately prior to use to ensure the absence of rust and other foreign material on the inside surface.

3. The molds were preheated to drive off moisture when the molds were cold.

4. Bar ends to be spliced were previously square cut. The ends of the bars were brushed to remove mill scale, rust, and other foreign material to ensure cleanliness, and then heated.

5. A permanent point was marked from the end of each bar for a reference point to confirm that the bar ends were properly centered in the splice sleeve.

6. Before the splice sleeve was placed into final position, the bar ends were examined to ensure that the surface was free from moisture. If moisture was present, the bar ends were heated until dry.

7. Special attention was given to maintaining the alignment of sleeve and guide tube to ensure a proper fill.
3.8-51 SGS-UFSAR Revision 6 February 15, 1987
8. The splice sleeve was preheated after the materials and equipment were in position.

9. Completed splices were visually inspected to assure that fill was within acceptable limits at both ends of the splice sleeve and at the top hole in the center of the splice.
The following documentation and records were maintained:
1. Mill test reports of the material furnished

2. Record of cadwelder qualification

3. Record of visual inspection of splices

4. Drawings showing splice locations

5. Record of tensile tests of splices Where accessibility of limited space precluded the use of the cadweld processes, the specifications permitted splicing by butt welding. These cases constituted less than 1 percent of the total number of splices made. Welding was performed in accordance with AWS Specification D-12.1, with double "V" groove butt joints in the horizontal position and single "V" groove butt joints in the vertical position. Welding was performed by the shielded arc processes using low hydrogen stick electrodes.
Qualification of the welding procedures was made in accordance with the philosophy and intent of Section IX of the ASHE code. Full section tests were made on Grade 60, 188 reinforcing steel bars. These weld tests indicated tensile, yield, and elongation values in excess of minimum requirements of the bar material. The welding procedures required pre-heated temperatures to 325°F +/-25°F and interpassed temperatures of 300°F to 500°F. The filler 3.8-52 SGS-UFSAR Revision 6 February 15, 1987
material conformed to AWS Specification No. A 5.5-69, E-100 18-02 and certified mill test reports were required. The completed welds were post heat treated at 1050°F to 1100°F for 15 minutes and then were wrapped with a protective blanket of insulating material to avoid rapid cooling. All welds were 100 percent visually and radiographically examined. Radiographic examination was in accordance with UW 51 of the ASHE Boiler and Pressure Vessel Code, Section VIII. 3.8.1.6.8.4 Waterproofing Membrane To waterproof the subgrade exterior walls and foundations, a rubber waterproof membrane was installed under all foundations and was extended vertically up to 6 inches below yard grade. The horizontal waterproofing membrane was 1/16-inch thick Ethylene Propylene Diene Monomers (EPDM rubber). The waterproofing membrane used on vertical surfaces was 3/64-inch thick nylon reinforced Ethylene Propylene Diene Monomers (Nylon Fabric Inserted EPDM). The specification detailed the properties of the material and the ASTM test methods. The specification required that mill tests by the manufacturer for conformance be witnessed by PSE&G Testing Laboratory personnel. In addition, the manufacturer was required to supply certificates of compliance of the material with the specification. Also, users' tests were made by PSE&G on a random basis. 3.8.1.6.8.5 Compaction of Fill The detail specification required compacted fill be installed under the Class I storage tanks to 98 percent and area adjacent to Class I structure to 95 percent of the maximum dry density attainable by the AASHO Specification T-180-61 method of compaction. The specification also required that all fill material conform to Type I (all classes) or Type 4 (Classes A and B) of the New Jersey State Highway Department's "Standard 3.8-53 SGS-UFSAR Revision 6 February 15, 1987
Specification for Road and Bridge Construction, except that the amount of fines (particles passing the No. 200 sieve) could not exceed 15 percent by weight. The suitability of materials for use in the construction of compacted fill was made by a representative of PSE&G' s soils consultant (Dames and Moore). Before the installation of any compacted fill material was started and during the first week of work, a Dames and Moore representative supervised and reviewed the compaction process and the testing performed by the PSE&G Laboratory. The specification required that the fill material be compacted in 8-inch layers, and, before subsequent lifts were placed, an inplace density test was made by the PSE&G Testing Laboratory personnel in accordance with AASHO T191-64 (ASME Specification D1566-64, Density of Soil in Place by the Sand Cone Method). Where test results indicated that the required density was not attained, the materials were reconditioned and recompacted to the required density. The preparation of the Optimum Moisture - Dry Density Curves was done by the PSE&G Testing Laboratory and Dames and Moore in accordance with AASHO T180-61 (ASTM Specification D1557-64). Also, particle size analysis of soils was in accordance with AASHO T-88 (ASTM D422-63). United Engineers and Constructor's Quality Control inspected the compaction operation and coordinated the testing requirements. All reports, tests, and other pertinent information were documented at the site by United Engineers and Constructor's Quality Control. 3.8. 1.6.8.6 Liner Plate General The fabrication and erection of the Reactor Containment liner, personnel locks, equipment hatch, and liner plate attachments were 3.8-54 SGS-UFSAR Revision 6 February 15, 1987
performed by CB and I. PSE&G or its agent reviewed all test results and monitored all work. Material The detail specification for the Reactor Containment* liner required that the steel for the main shell, including the dome, cylindrical walls, and the bottom, be low carbon/high manganese steel with fine grain structure, meeting. ASTM Specification A442-66, Grade 60. In addition, the liner material was* impact tested in accordance with the 1968 edition, ASME Boiler and Pressure Vessel Code, Section Ill, Paragraph Nl2ll, at a temperature JO*F below the minimum service temperature of so*F. Mill test reports certifying the physical and chemical properties of the plate delivered to the jobsite were submitted. The Reactor Containment liner was fabricated and erected in accordance with Part UY, "Requirements for Unfired Pressure Vessels Fabricated By Welding," Section VIII of the ASME Boiler and Pressure *. <assel Code, 1968 edition, and PSE&G Detail Specification No. 68-7123. Where any conflict was evident, the PSE&G Specification was followed. The qualification of all welders and welding procedures was performed in accordance with Part A, Section IX, of the ASME Boiler and Pressure Vessel Code, 1968 edition. To assure good welds the following were required as a minimum:
1. Thorough cleaning of weld preparations

2. Removal of slag from previous passes

3. Proper control of welding current and polarity
SGS*UFSAR 3.8-55 Revision 13 June 12, 1994
4. Make certain welding materials and base materials were dry before welding. Heating of base material in the vicinity of weld when temperature wa$ below 70*F

5. Shielding welding arcs from winds and drafts Evaluation of porosity in spot radiography was .in accordance with.
the standards of Appendix 4 of Section VIII, ASME Boiler and Pressure Vessel Code, 1968 edition. Standards for field welding. were in accordance with the requirements of Section VIII of the ASME Boiler and Pressure Vessel Code, 1968 edition. The liquid penetrant inspection of the liner plate welds was in accordance with Appendix 8, "Methods of Liquid Penetrant Examination," Section VIII of the ASME Boiler and Pressure Vessel Code, 1968 edition. Inspection of the liner seam welds was accomplished as follows:
1. A trained inspector responsible for welding Quality Control inspected every weld.

2. For the bottom liner plates, liquid penetrant and/or magnetic particle inspections of 2 percent of the weld seams was performed. In addition, the first 10 feet of weld made by each welder was also liquid penetrant and/
or magnetic particle inspected.
3. All the liner bottom plate welds were 100 percent vacuum box tested to 5 psi pressure differential with atmospheric pressure.

4. The liner plate seam welds in the cylindrical walls and dome were 2 percent radiographed for each welder and positioned in accordance with UW52. In addition, the 3.8-56 SGS-UFSAR Revision 6 February 15, 1987
first 10 feet of weld made by each welder was 100 percent radiographed. The following preliminary tests were made during the liner erection using the test channels:
1. All welds were covered by channels and zoned after which a strength test was performed by applying 54 psig air pressure to the channels in a zone for a period of 15 minutes. The exposed welded joints were given a soap test for leaks. If bubbles indicated a leak, the leak was repaired and the zone retested.

2. The zones of channels were then retested to a pressure of 4 7 psig with a 20-percent, by weight, freon-air mixture. The entire run of the channel to plate weld was traversed* with a halogen leak detector. If a leak was detected, repairs were made and a retest performed.

3. In addition, the zone of channels was held at the 47 psig air pressure for a period of 2 hours. When pressure drop exceeded the standards, zones were repaired and retested.

1. Overall out of roundness below El. 130 feet: +/- 2 inches
- above El. 130 feet: +/- 4 inches
2. Overall deviations out of plurnbness - l/500 of the height.
3.8-57 SGS-UFSAR Revision 6 February 15, 1987
Equipment Hatches, Personnel Locks, and Penetration Sleeves The detail fication required that the equipment hatches, except as noted '-' for the outage equipment hatch, and personnel locks be made from ASTM A-516, Grade 60, conforming to ASTM A-300 requirements. In addition, the steel was normalized by heating to 1,700°F (+/- 50°) and cooling in still air and "V" notch tested to a minimum of 15 foot pounds at -4 0°F in accordance with ASTM fication A-370-65. For the material on the interior bulkhead which is not subjected to low temperatures, the material was required to pass a Charpy "V" Notch test of 15 foot pounds at +20°F, in accordance with ASTM Specification A-370-65. The OEH conforms to ASME Code, Section VIII material requirements. Access hatches, except as noted, were fabricated and tested at the fabrication The personnel locks and any portion of the equipment access door extending beyond the concrete shell conforms to the requirements of the 1968 edition of the ASME, Section III for Class B Vessels. The chambers were completely fabricated in the shop and the weldments were inspected by the Hartford Insurance Group. The chambers were not stamped as a Class B pressure vessel, but were covered by an equivalent to a Manufacturer Partial Data Form N-lA for the personnel locks and covered by a Manufacturer Partial Data Form for the equipment hatches. The user and his authorized representative rr.oni tared test at the fabrication shop and audited all records. ~he OEH conforms to the requirements of ASME code, Section VIII. The piping penetration sleeves are carbon steel and are described in Section 3.8.i.6.8.10. The electrical penetration sleeves are of Schedule 80 caroon steel. The piping and electrical penetration sleeves were welded to the liner assemblies at the fabrication shop in accordance with ASME, Section III, Class B vessels, except for stamping. The weldments were inspected in the The Hartford Insurance Group and are covered by a Manufacturer Partial Data Form. Strengtn and leak tests were performed in the shop on all access hatches. Leak test of the penetration sleeve welded channel was also performed in the shop. 3.8-58 SGS-UFSAR Revision 18 April 26, 2000
The shop leak test procedures were conducted in accordance with Appendix A of ANS 7. 60 "Proposed Standard for I,eak Rate Testing of Reactor Containment Structure for Nuclear Reactor." Proof tests were applied to chambers to pressurize the necessary areas to 54 psig. The pressure was maintained a sufficient time to allow soap bubbles and Freon sniff tests of all welds and mating surfaces. Any leaks found were repaired and retested. The repair of defective welds was in accordance with paragraph N-528 of Section III, "Nuclear Vessels t" of the ASME Boiler and Pressure Vessel Code, 1968 edition. Attachments to Steel Liner Nelson Studs Nelson studs were welded to the plating as shown on the drawings. Each welder, at the beginning of each day, attached at least one test stud which was tested by bending the stud approximately 45 degrees toward the plate to demonstrate the integrity of the weld. If failure occurred in the weld, the welding procedure or technique was correctedt and two successive studs successfully welded and tested before further studs were attached to the liner plate. These test studs were allowed to remain in place but were not considered as part of the regular stud pattern required by the design. All studs on which a full 360-degree weld was not obtained were removed and replaced with a new stud. Before welding a new stud, where a defective one han been removed, the area was ground flush and smooth. Stiffeners were welded to the exterior of the dome. These stiffeners conformed to ASTM A-36 material. 3.8-59 SGS-UFSAR Revision 6 February 15, 1987
3.8.1.6.8.7 Construction Construction at the site started with clearing and the installation of the first stage of a Dewatering System. The first stage consisted of installing a Peripheral Ejector System at existing grade (Elevation 99+ 1, PS datum) utilizing approximately 230 ejector wells, each 90-feet deep and double 12-inch header system connecting the wells to three diesel driven pumps. Piezometers were installed adjacent to the area to be excavated to determine the ground water level at all times. Daily readings were taken and recorded to ascertain that the ground water level remained well below the excavation level. When the Dewatering System started drawing the ground water down, an open excavation sufficiently large to accommodate all major structures was started. At an approximate depth of 23 feet below existing grade, a cellular cofferdam was constructed which encircled the excavation for all the major or Class I building structures. The cofferdam consists of 24 circular cells, 60 feet in diameter, with connecting arcs, and "toed" into the Vincetown Strata approximately 10 feet. The area within the cofferdam was then excavated to Elevation 43 feet (PS datum) and a second stage and separate Dewatering System was installed. This system consisted of approximately 140 well points, each 24-feet deep with an 8-inch header connecting the wells to one self-priming centrifugal pump. Later, this piping system was encased in lean concrete fill, abandoned in place and filled under pressure with grout. The total Dewatering Systems pumped between 1, 000 to 1,400 gpm continuously. Then the final stage of excavation within the cofferdam area proceeded to the Vincetown Strata. During this final stage of excavation, a Consulting Soils Engineer was at the site on a continuous basis. The Soils Engineer, the Resident Engineer, and a Quality Control Supervisor visually inspected the bottom of the excavation to verify that the excavation had reached the top of Vincetown Formation prior to 3.8-60 SGS-UFSAR Revision 6 February 15, 1987
placing any lean concrete. Prior to the completion of the excavation, at approximately Elevation 45 feet, 15 exploratory borings were drilled through the remaining Kirkwood Formation and into the underlying Vincetown to verify the original study borings. These additional borings showed no measurable differences from the study borings. In addition, after the Vincetown Strata had been exposed, six test borings were drilled in the excavated area into the underlying Vincetown Strata to verify and ensure that the foundation mat was, in fact, directly supported on the Vincetown Formation. Four of these borings were drilled under the Unit 2 Reactor Containment; two were drilled under the Unit 1 Reactor Containment; all borings penetrated a minimum of 20 feet into the underlying Vincetown Formation. During cold weather the exposed Vincetown was protected from freezing with insulated blankets; also, in some areas a 2 to 3-foot thick earth "blanket" was used for protection from both frost and construction traffic. The lean concrete poured over the Vincetown Strata had an overall average strength of 3509 psi with a 16.45-percent coefficient of variation as verified statistically by computer. The concrete was placed via conveyors directly from the batch plant to the point of deposit, under the quality control procedures stated in previous sections. After the lean concrete was poured and screened to the bottom level of the foundation mat, a 1/16-inch thick rubber, ethylene Propylene Diene Monomer waterproofing membrane was installed. A 1/8-inch thick hard board was installed over the membrane and then a 3-inch thick concrete protection course was installed. Later, the waterproofing membrane was extended vertically up the foundation walls with 3/64-inch thick nylon reinforced rubber and protected with 1/8-inch thick hardboard. The reactor containment base mats for Units 1 and 2 were poured in 6 segments and 8 segments, respectively. Vertical construction joints were constructed with expanded wire mesh. No horizontal joints were permitted. 3.8-61 SGS-UFSAR Revision 6 February 15, 1987
The base mat was poured to a level 6 inches below the final elevation of the bottom liner plate. The backing "T's" were then leveled to final position and concrete was poured flush with the top of the backing "T' s". Then the bottom liner plate was installed and the welds over the backing 11 T' s" made. The knuckle plate was installed and the cylindrical portion of the liner was erected to Elevation 120 feet (PS datum). Reinforcing steel for the outer walls were then placed. The exterior concrete containment wall was constructed in 5-foot lifts during the construction of the interior concrete. All welds were checked for compliance with the approved weld inspection and test requirements. The reactor containment interior concrete was built on the mat liner. On the completion of the interior concrete structure to Elevation 130 feet (PS datum), the polar crane was then erected. Concrete in the exterior wall was poured in uniform 5-foot lifts around the entire circumference. The completed steel wall liner was braced internally and locally with temporary bracing to prevent distortion during concrete placement. On completion of the concrete exterior walls to Elevation 100 feet (PS datum), a rubber waterproof membrane was attached to the exterior concrete surface with adhesives and a 1/8-inch protection board installed to protect the membrane. The space between the excavation and the reactor containment structure was then backfilled and compacted in 8-inch layers to 95 percent of the maximum dry density in accordance with procedures. The liner was completed, finishing with the construction of the 1/2-inch thick steel dome, with all welds constructed and tested in accordance with procedures previously outlined. The steel dome liner was supported during erection from a steel tower which was erected at Elevation 130 feet. Sections of the 3.8-62 SGS-UFSAR Revision 6 February 15, 1987
steel dome liner were lifted into place, braced with wind girders, and welded into their final locations. 3.8.1.6.8.8 Penetrations In general, a penetration consists of a sleeve embedded in the concrete wall and welded to the containment liner. The weld to the liner is shrouded by a continuous channel which is test pressurized to demonstrate the integrity of the penetration-to-liner weld joint. The pipe, electrical conductor, duct, or equipment access hatch passes through the embedded sleeve and the end.of the resulting annulus is closed off, either by welded end plates, bolted flanges, or a combination of these. Provision has been made for differential expansion and misalignment between each pipe and sleeve. No piping loads are imposed on the liner. Pressurizing connections are provided to demonstrate the integrity of the penetration assemblies. There are three large openings that significantly perturb the reinforcing pattern. One is the equipment hatch with an 18-foot diameter outer barrel; the others are two personnel hatches with 9 foot-9 inch diameter outer barrels. As a rule of thumb, only openings having diameters greater than 2 1/2 times the wall thickness are considered to require detailed analysis. The design and analysis for these three large openings are described in Section 3.8.1.6.8.9. The main wall reinforcing, consisting of vertical and horizontal reinforcing bars, is bent around all the openlngs. Contlnui ty of shell reinforcement is therefore maintained. For large openings, in addl.tion to these bars, circular reinforcing bars have been provided to take care of axial thrust and principal moments around the opening. Radial stirrups have been provided to take care of the torsion and shear. This combination of reinforcing bars takes care of all primary and secondary stresses. 3.8-63 SGS-UFSAR Revision 6 February 15, 198?
3.8.1.6.8.9 Equipment and Personnel Access Hatches Equipment and personnel access hatches, except as noted for the outage equipment hatch, OEH, are fabricated from A516, Grade 60 steel normalized to A300 requirements. All personnel locks and the portion of the equipment access hatch extending inside the containment structure beyond the concrete shell are designed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Class B. The Code was used as a guide; therefore the N Stamp requirement is waived. The equipment and personnel hatch details are as shown on VTDs 301051 and 301075, respectively. The OEH is designed in accordance with ASME VIII. The hatch barrel is embedded in the containment wall and welded to the liner. Provision is made to test pressurize the space between the double gaskets of the door flanges and the weld seam channels at the liner joint, hatch flanges, and dished door. The personnel hatches will be double door, mechanically latched, welded steel assemblies. A quick-acting type equalizing valve connects the personnel hatch with the interior of the containment vessel for the purpose of equalizing pressure in the two systems when entering or leaving the containment. The personnel hatch doors are interlocked to prevent both being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened. Remote indicating lights and annunciators situated in the control room indicate the door operational status. Provision is made to permit bypassing the door interlocking system to allow doors to be left open during plant cold shutdown. Each door lock hinge is designed to be capable of independent three-dimensional adjustment to assist proper seating. An Emergency Lighting and Communication System powered from an external emergency power supply are provided in the lock interior. Emergency access to either the inner door, from the containment interior; or to the outer door, from outside, is possible by the use of special door unlatching tools. Plan and elevation drawings of the personnel air lock with all electrical and piping penetrations identified are provided on Figure 3. 8-37 and VTDs 30107 5 and 301059. 3.8-64 SGS-UFSAR Revision 27 November 25, 2013
Pressure and monitoring taps are provided to pressure test the double gaskets on each door to a "between the seals test pressure" of 10 psig. When testing for seal leakage during periods of when the air locks are frequently opened. Test connections are also provided to permit pressurization of the entire air lock. Leakage rate testing of the airlocks is described in Section 6.2.1. Tie-downs are used to prevent the inner door from becoming unseated during pressure tests. The yoked ends of the tie-downs are pin connected to the horizontal stiffeners at the door, and the threaded ends of the tie-downs are slipped through the holes of the tie-down beams and secured with nuts. The instruction manual for the personnel air locks requires tightening the nuts to draw the door flange to approximately 1/16 inch from the bulkhead flange before pressurization. There is no monitoring device to read the force exerted on the door. This mechanism cannot be operated from within the air lock. Around these hatch openings, thickened concrete edge beams are provided to take care of the high membrane and bending stresses in the opening areas. Loading combinations 1, 2, and 3 as listed in section 3.8.1.3 were used in the large opening analysis. Tornado stresses were found to be far less critical than stresses resulting from accident pressure or earthquake. Manual design utilized the elastic center method for ring beam analyses. The method is described and illustrated in "Statically Indeterminate Structures" by L. c. Maugh (4). The ring beams are designed to resist biaxial bending moments, axial tension, torsion, and biaxial shear. conrad Associates* finite element analysis (3) for the three large openings is based on the procedure formulated for linearly elastic, thin shells of arbitrary geometry which uses a fully compatible plate bending element to obtain the bending stiffness, and a constant strain triangle element to determine the membrane 3.8-65 SGS-UFSAR Revision 16 January 31, 1998
stiffness. The results of these analyses are shown on Figures 3.8-39 through 3.8-48. From the more refined computer analysis, modifications have been made in the preliminary penetration details to strengthen the reinforcement arrangement in areas where higher stresses have resulted from the independent check. Details for personnel and equipment hatch reinforcements are as shown on Plant Drawing 201131. 3.8.1.6.8.10 Piping Penetrations High integrity piping penetrations are provided for all piping passing through the containment. Figures 3. 8-49 and 3. 8-50 show typical cold and hot pipe penetrations respectively. The pipe is centered in the embedded sleeve which is welded to the containment liner. Seal plates are welded to the pipe at both ends of the sleeve. In some instances several pipes pass through the same embedded sleeve to minimize the number of penetrations required. In such cases, each pipe is welded to the inside seal plate and to the expansion bellows which is, in turn, welded to the outside seal plate. Large single pipe containment penetrations were installed with expansion test bellows, attaching the process piping to the penetration sleeves, which allowed for Appendix J type "B" pressure testing of the compartment formed between the process piping and the embedded sleeve, via a test connection on the bellows. Containment piping penetrations designed for Salem are not required to be type "B" tested for 10CFR50 Appendix J (Ref. Safety Evaluation S-C-R700-MSE-0253 Rev. 0). The type "B" test is applicable to piping penetrations that utilize expansion bellows as the leakage limiting boundary. The piping penetrations at Salem rely on partial/full penetration seal welds inside containment as the leakage limiting boundary, which are leak rate tested as part of the Appendix J type "A" containment Integrated Leak Rate Test (ILRT). 3.8-66 SGS-UFSAR Revision 27 November 25, 2013
Therefore, for containment p!~ing penetrations, leak rate testing of separate penetrations (type .i)" testing) has been replaced by the containment integrated leak rate test (type "A" testing) as allowed by 10CFRSO Appendix "J". As a result, the abandoned expansion test bellows on the (10) service water lines penetrating the containment have been eliminateo as part of the service water system piping upgrade program. Figure 3.8-SOA shows a typical service water piping containment penetration detail. In the case of piping carrying hot fluid, the pipe is insulated and cooling is provided to limit the concrete temperature adjacent to the embedded sleeve to lSO*F. For the larger hot pipe penetrations, strong anchoring is necessary and is provided as shown. The anchors engage a large segment of the wall to adequately resist thrusts. Should a piping failure occur within the containment, the additional loading imposed upon the penetration is transmitted through the anchor to the containment structure. Therefore no permanent deformation of the pen.,tration will be realized. Moment eliminators are installed outside of the containment structure. Hangers and limit stops assist in supporting and reducing any moment loading of a free-hanging pipe. 3.8-66a Revision 13 June 12, 1994
THIS PAGE INTENTIONALLY LEFT BLANK 3.8-66b SGS-UFSAR Revision 13 June 12, 1994
A multiple guide arrangement is used to limit the high frequency vibrations which could be imposed on the piping penetrations and surrounding liner regions by vibrating pipes which pass through the containment structure. Further, the penetration sleeve and containment liner are anchored in the concrete so that the liner does not participate in vibration. The design of the equipment, pipes, support structures, and penetrations is in accordance with the stress limitations of Section III, ASHE Nuclear Vessel Code, and the ANSI B31.1, Code for Pressure Piping. A thermally induced loading is considered to produce uniform deformation at the liner-sleeve interface. The average liner stress was calculated in both the horizontal and vertical directions and the maximum moment and maximum hoop stress in the sleeve was computed. The thickness of the s] eeve was chosen so that the stresses do not exceed the allowable stresses. The material used for penetrations is carbon steel and conforms with the requirements of the ASHE Nuclear Vessel Code. As required by the Nuclear Vessel Code, the penetration material meets the necessary Charpy V-Notch impact values at a temperature of 20°F. The penetrations are not exposed to the elements and, therefore, are not subject to 0°F outside temperature. Penetration expansion bellows are suitably protected against field damage; such protection remains a part of the permanent installation. The originally specified material for all penetration sleeves was ASTM A-155-KC 70, Class I. Due to design modifications, A106 Grade B, A106 Grade C, Al55-KC 60, and A333 Grade I material has been utilized for some sleeves. In such cases, seamless pipe was used where possible. Where it was required, a heavier wall thickness than that specified for the origjnal Al55-KC 70 Class I was used, so that in all these cases, sleeves of equal or greater strength than the original design have been used. 3.8-67 SGS-UFSAR Revision 6 February 15, 1987
3.8.1.6.8.11 Electric Penetrations Power, control, fiber optic, and shielded conductors are assembled in canisters which have been inserted in and welded to nozzles in the field. Figure 3.8-51 shows typical electrical penetrations. A prototype of each type of penetration has been factory tested at 271F and 62 psig in a steam chamber. Tests prove the ability of prototypes to function properly, electrically and mechanically, before, during, and after subjection to these conditions. Each penetration is factory tested before shipment to verify that the leakage rate does not exceed
-6 1 x 10 cc/sec at one atmosphere differential when tested with dry helium.
There are 56 electrical penetrations per unit. The penetration sleeves to accommodate the electrical penetration assemblies are Schedule 80 carbon steel, except where otherwise noted. 3.8.1.7 Testing and Inservice Inspection Shop leak testing procedures have been conducted in accordance with Appendix A of ANS 7.60, "Proposed Standard for Leak Rate Testing of Containment Structures for Nuclear Reactor." A proof test is applied to each penetration which pressurizes the necessary areas to 54 psig. This pressure is maintained a sufficient time to allow soap bubble tests of all welds and mating surfaces. It is retested to a pressure of 47 psig for leakage rate. All penetrations, the personnel locks, and the equipment hatches are designed with double seals which are pressure tested at 54 psig. Individual penetration internals and closures and sleeve weld channels are leak tested at 47 psig after installation. 3.8-68 SGS-UFSAR Revision 29 January 30, 2017
Information and inservice testing and additional information on pre-service testing is contained in Section 6.2.1. 3.8.2 Steel Containment System Since the containment is a reinforced concrete structure lined with only a thin steel plate leak-tight barrier, this section does not apply. For equipment hatch, personnel access hatches, and penetrations, see Section
3. 8. 1.
3.8.3 Internal Structures 3.8.3.1 General Description For arrangement of the walls, elevated slabs, and components of the internal structures see Figure 3.8-1 and Plant Drawing 208900. For further description, see Sections 3.8.1.1 and 3.8.3.4. 3.8.3.2 Design Codes The internal structures have been designed under the following codes:
1. Building Code Requirements for Reinforced Concrete, ACI 318-63.

2. AISC Manual of Steel Construction, 6th Edition or later edition, as applicable.
3.8-69 SGS-UFSAR Revision 27 November 25, 2013
3.8.3.3 Loads and Loading Combinations The following design load criteria were used in the design of all compartment structures. (a) C - 1.0 D +/- 0.05 D + 1.5 P + 1.0 T (b) C - 1.0 D +/- 0.05 D + 1.25 P + 1.0 T + 1.25 E (c) C - 1.0 D +/- 0.05 D + 1.0 P + 1.0 T + 1.0 E' + J (d) C- 1.0 D +/- 0.05 D + 1.0 T + 1.0 E' + J' where: c the capacity of the structure D the dead load p the pressure differential of 15 psi T the thermal effect E the OBE E' the DBE J the greater of 950 kips jet force or pipe rupture load transmitted by restraints J' the greater of 1500 kips jet force or pipe rupture load transmitted by restraints For additional information see Internal Pressure Analysis in Section 3.8.1.4. The load combinations utilized in the design of the internal structures were either equivalent to or more conservative than those outlined in the SRP. The following tabulations provide a comparison of load combinations utilized with the SRP criteria. 3.8-70 SGS-UFSAR Revision 13 June 12, 1994
INTERNAL CONCRETE STRUCTURES (2) SRP 1. 4D + 1. 7E 1. 9E Salem Not Critical (2b) SRP 0.75 (1.4D + 1.71 + 1.9E = 1.7 T0 + 1.7R)0 Salem Less Critical Than (5) (3) SRP D + L + T + R + E' 0 0 Salem Less Critical Than (6) (4) SRP D + L + T + R + l.SP aa a Salem D + L + T + R + l.SP a a (5) SRP D + L + T + R + I. 25P + (Yr + Yj + Ym) a a a
+ 1. 25E Salem D + L + T + R + 1.25P + (Yr + Yj + Ym) + 1.25E aa (6) SRP D + L + T + R + P + (Yr + Yj + Ym) + E' aaa Salem D + L + T + R + P + (Yr + Yj + Ym) + E' a a See FSAR Section 3. 8. 1. 3 for identification of symbols. Although the R and Y forces are not listed in the overall structural analysis load combination formulae, the local effects under piping load, jet load, and missile impingement were taken into account.
3.8.3.4 Design and Analysis Provisions The reactor vessel is surrounded and supported by a massive circular wall known as the primary shield, which also supports the central portion of the refueling canal walls above i t and the operating deck which rests on the top of the refueling canal walls.
3. 8-71 SGS-UFSAR Revision 6 February 15, 1987
The reactor vessel cavity wall, known as the "Primary Shield," was designed as a thick cylinder subjected to internal pressure, dead load, design and operating basis earthquakes, jet forces, and guillotine breaks of the reactor coolant loop piping entering and leaving the reactor and entering the steam generator. Cognizance was taken of the discontinuity moments and shears at the base of the cylindrical wall. Figures 3. 8-52 through 3. 8-55 show the plan and cross sections of the vessel cavity with pressure and forces indicated,* and a sketch showing thick wall stress analysis. These pressures and forces were used as conservative desi.gn criteria. The values resulting from the final analysis are well below these magnitudes. The reactor pressure vessel cavity extends vertically from Elevation 54 feet to Elevation 104 feet, Above Elevation 104 feet, the area becomes part of the refueling canal. Below Elevation 81 feet, the cavity walls are inherently safe from any disaster as they are embedded in the massive concrete foundation of the building. I As of the original design of the primary shield, in the event of a failure of a reactor pressure vessel nozzle, pressure relief is provided by B blowout plugs, approximately 10 square feet in area each, venting to Elevation 104 feet. Additionally, there is always a 2-inch air gap between the vessel flange and the wall at Elevation 104 feet, except during refueling. A pressure of 900 psi was used for the design of the nozzle penetrations through the shield wall to withstand a longitudinal break in the nozzle extensions. The reactor cavity is designed for a pressure of 175 psi acting upon the walls following a break in the vessel nozzle. A guillotine break of the nozzle was also considered. The reaction in this case is taken by the reactor supports, which are integral with the reactor cavity walls. Subsequent to the original design, leak-before-break (LBB) was approved for the primary loop piping {see Section 3. 6. 4). LBB allows the elimination of the dynamic effects of pipe rupture, including sub-compartment pressuriZcation. Thus, the dynamic pressure loads resulting from a break in the primary piping can be eliminated from the design of the primary shield, also eliminating the need for the pressure relief function of the blowout plugs. In order to reduce personnel radiation exposure and critical path time during refueling outages, the blowout plug covers that were installed and removed during every refueling are now left in place auring normal operation. The existing analysis of the primary shield, which includes breaks in the primary piping with pressure relief through the blowout plugs, is conservatively retained. 3.8-72 SGS-UFSAR Revision 23 October 17, 2007
During seismic loading the entire internal structure could be subjected to torsional stresses resulting from a 5-percent accidental eccentricity. For information concerning the extent to which the internal structure and its walls are capable of withstanding the stresses resulting from the accidental eccentricity, see Section 3.7. The following structures are removable; provisions have been made to prevent them from becoming missiles. Hatch covers at Elevation 130 feet They are 4-foot thick massive concrete blocks bolted down to the slab to withstand any uplift pressure. Missile shield over Reactor Vessel The missile shield over the reactor vessel consists of a 181-inch diameter, 2-inch thick steel plate that is permanently attached to the Integrated Head Assembly (IHA). It is secured to prevent it from becoming a missile. See UFSAR Section 5.5.14.1 for a description of the IHA . Lead blocks at Elevation 100 feet in fuel transfer canal area and at Elevation 89 feet-6 inches to 95 feet-10 inches transfer chamber They are for radiation protection for the gap between containment exterior wall and interior structure. Blocks are held in place by angle frames and steel plates to hold them during earthquake motion. Hatch cover over fuel transfer chamber at Elevation 100 feet It is a steel enclosure filled with poured lead. It is laterally
SGS-UFSAR 3.8-73 Revision 23 October 17, 2007
restrained, subject to no jet force and could never become a missile. 3.8.4 Other Category I Structures 3.8.4.1 Summary Description The Category I structures other than the containment structures are listed in Section 3. 2. They are Auxiliary Building, Fuel Handling Buildings, Service Water Intake Structure, Class I water tank foundations, and the Class I equipment supports. The orientation of the principal structures is shown on Plant Drawing 201012. The general arrangements of Containment Buildings, Auxiliary Building, and Fuel Handling Buildings are shown in Section 5. 3.8.4.2 Design Codes Category I structures were designed under the following codes:
1. Building Code Requirements for Reinforced Concrete, ACI 318-63.

2. AISC Manual of Steel Construction, 6th Edition or later edition, as applicable.
3.8.4.3 Loads and Loading Combinations Loads and load combinations for Category I structures under SSE or tornado conditions are similar to those of the containment except for LOCA pressure and temperature loadings. The working stress design method was used for operating and OBE conditions. (No load factor was used for OBE loadings.) For combination of containment loadings and load symbols, see Section 3.8.1.3. 3.8-74 SGS-UFSAR Revision 27 November 25, 2013
The tornado and tornado generated missile analyses for Category I structures are provided in Section 3.8.1.4. The load combination Case D specified tornado loads combined with operating loads. The tornado load (W ) includes the static t forces produced by the 360 mph maximum wind velocity, a 3 psi negative pressure, and the structural response to the missile impact. The stresses on any structural member produced by the effective pressure transformed from the tornado wind, the impact of the missile, and also differential pressure were superimposed to obtain the most critical total stress, provided the induced stress from these three components are in the same direction. When one of the components induced an opposite stress, thereby reducing the total stress in the member, it was neglected. In other words, all six loading combinations listed in the SRP have been considered with factors of one instead of 0.5 for Wp in combinations iv and vi and a1so have taken into account stress directions as stated above. Category I water storage tanks are not designed to withstand tornado-induced missiles. Sufficient backup water sources are available to assure a safe shutdown of the reactor and the ability to maintain the reactor in a safe condition. Venting of structures under tornado generated differential pressure was not adopted as a design criterion for Category I structures. These structures are capable of sustaining the differential pressure generated by a tornado. Metal siding on the Turbine Generator Building (non-Category I), however, was designed to be blown out to relieve tornado generated differential pressure. 3.8-75 SGS-UFSAR Revision 32 June 17, 2021
category I structures are not additionally loaded or affected by the adjacent non-category I structures since the non-Category I structures (adjacent to Category I structures) are heavily braced to withstand tornado wind forces such that they will not collapse on the Category I structures. -- Five heavy bays of steel bracing system have been installed in the turbine-generator area structure to withstand the static horizontal shear force under earthquake condition. With this bracing system, the Turbine Building will not collapse on the adjacent Category I structures under OBI or DBB loads. The load combinations utilized in the design of Category I structures were either equivalent to or more conaervative than those outlined in the SRP. The working stress design method waa used for the category I concrete structures under operating and OBB loadings. The Salem design is in conformance with load combination a(i)2a in SRP 3.8.4. The strength design method was used for the Category I structures under SSE/tornado and DBA loadings. The Salem design is in conformance with load combination (6) and (8) in SRP 3.8.4. Both load combinations represent the moat severe cases. Hydrostatic loadings from the hurricane condition were applied to the structures to check their stability. The procedures used by our consultant (Dames and Moore) for transferring the static and dynamic floor effects to load were as delineated in the u. s. Army coastal Engineering Research center Technical Report No. 4. Total head, including wave effects, was considered to investigate the lateral and overturning effects. The buoyancy effect of groundwater was included in the assessment of the sliding and overturning potential of all Category I structures. The buoyancy effect will reduce the dead weight and thus reduce the factors of safety against sliding and overturning. To include the buoyancy effect in assessing the sliding and overturning potential is the more conservative and correct approach. 3.8-76 SGS-UFSAR Revision 16 January 31, 1998
The safety against sliding, overturning, and flotation for all Category I structures under all loading combinations are within the limits set by the SRP 3.8.5. Masonry Walls For the loading criteria for non-structural masonry walls see Section 3.8.4.5.1. 3.8.4.4 and Procedures The structures have been on ACI 318-63 "Working for normal OBE and "Ultimate Design," for normal loads DBE or tornado. In the under OBE the allowable stresses are one-third above the normal code working stresses. Wind stresses are found to be less critical than those for an OBE. Load factors of have been used in the ultimate design under DBE or tornado loading. The stress of reinforcing steel under ultimate has been under 0. 9 Fy. The reduction factor "0" as described in Section 3.8.1.4.1 for concrete stress is for all Category I structures. A coefficient "k" of 0.85 for 3500 psi concrete has been used in addition to "0" for equivalent rectangular concrete stress distribution. During the design phase of the re-racking which was implemented in 1994, the Spent Fuel Pool was reanalyzed for increased fuel storage capacity. The following is the list of items incorporated in the analysis:
1) The "ultimate strength" design method based on NUREG-0800, Standard Review Plan 3.8.4, Rev. 1, 1981 was used.

2) The plant design spectra, given in Salem UFSAR, for DBE and OBE events were broadened, per provisions of Reg. Guide 1.122 and used.

3) The response spectrum method was used to determine the self-excitation loading on the pool structure.

4) The pool structure was modeled in three dimensions via a 3-D finite element model.

5) The thermal across the slab and the pool walls was computed using finite element method. Thus, the effect of interaction between the ambient, pool water, and is fully incorporated in the 3.8-77 SGS-UFSAR Revision 25 October 26, 2010

6) The pressure on the lower portion of the wall during a seismic event undergoes a cyclic pulsation due to the hydrodynamic coupling between racks and the pool walls. This loading was quantified using Whole Pool Multi-Rack analysis. This loading was included in the analysis. ~

7) Analyses have been performed that evaluate the spent fuel pool structure (reinforced concrete as well as the stainless steel liner and its anchors) for a boiling pool condition. The acceptance criteria used for these evaluations was ACI 359 (ASME Code Section III Division 2, Reference 7). The spent fuel pool structure, including the liner and its anchors, was shown to meet all requirements of this code for a boiling pool condition. Analyses were also performed for the spent fuel pool liner and its anchors for a maximum normal pool temperature of 150°F. The liner and its anchors were shown to meet the acceptance criteria of ACI 359. Thermal cycling of the liner and its anchors was evaluated and shown to not be a concern.
Steel members inside the Category I structures are designed in accordance with the AISC Manual of Steel Construction (Sixth Edition or later edition, as applicable). Seismic design criteria are described in Section 3. 7. Tornado and tornado generated missile design is described in Sections 3.3.2 and 3.5.2. Four independent seismic analyses, similar to those for the Containment Building, have been performed for the 1) Auxiliary Building, 2) Fuel Handling Building,
3) Service Water Intake Structure, and 4) Outer Penetration Building.
Conservative results have been utilized for the building design. The time history computer analysis calculations are kept on file with PSE&G. Specifically, the information for the Auxiliary Building and the Fuel Handling Building has been provided in the Reference 3 report, while the information for the Service Water Intake Structure and Outer Penetration Building has been provided in the Reference 5 report. The loading combinations used for Category I (seismic) steel structures other than containment are as follows: 3.8-78 SGS-UFSAR Revision 17 october 16, 1998
Working stress design
1. D + L + I + H Allowable stresses in accordance with AISC Manual of Steel Construction

2. D T 1 + I + H + E Allowable stresses are one-third above the normal allowable stresses.
Ultimate
1. D + L + I + H + E'

2. D + L + I + H+ T The stress in the ultimate strength design has been kept under 0.9 Fy 3.8-78a SGS-UFSAR Revision 25 October 26, 2010
THIS PAGE INTENTIONALLY BLANK ISGS-UFSAR 3.8-78b Revision 17 October 16, 1998
where: D "' Dead load L = Live load I "" Impact load where moving load is present H = Thermal load E = Operating Basis Earthquake E' "" Design Basis Earthquake T = Tornado loading Protection against tornado wind loads and tornado missiles is discussed in Sections 3.3.2 and 3.5.2. Protection against turbine disc rupture and missile generation is described in Section 3.5.4. Masonry Walls For analysis of non-structural masonry walla see section 3.8.4.5.1. 3.8.4.4.1 HVAC Duct and Support Methodology Original Seismic Design Methodology The original seismic design assessed selected members (i.e., those with potential nonductile failure modes) for compliance with the stress limits. In addition, limiting support dimensions and tolerances, configurations, spans and member sizes were provided which ensured that the duct systems and supports' frequencies are rigid. The effective sheet metal structural properties are based on the 1969 version of the American Iron and Steel Institute (AISI) derivation of effective width for thin sections with stiffened compression elements subject to local buckling. HVAC Duct Seismic Adequacy Verification Methodology An alternate approach has been developed to provide consistent criteria and methodology for functional and seismic qualification of HVAC duct systems and supports. This methodology utilizes an alternative approach based on review of industry standard codes and practices, past earthquake performance data and shake table test results. All known credible failure modes for HVAC duct systems and supports, when subjected to earthquake loadings, were specifically documented. Then, engineering efforts were focused onto these credible failure modes and specific guidance for elimination of the failure mode or criteria for maintaining 3.8-79 SGS-UFSAR Revision 16 January 31, 1998
a suitable margin of safety against the failure mode were developed. The methodology anaures that the resulting margins of aafaty for aeiamic loadings and documentation requirement& are conaiatent with those of USI A-46 program requirement& par the SQUG GIP. All applicable seismic desiqn basis data and criteria from the SQOG GIP are adopted, and supplemented by teat results and stress requirements, as follows:
1. The GIP criteria for fans, air handlers and dampers that are directly applicable to seismic evaluation of BVAC duct systems are used.

2. The GIP criteria for expansion anchors and welded anchorage& are ulled directly. The higher margine of aafety for new inatallationa aa specified in the GIP are also used directly.

3. The SQUG GIP includes detailed guideline& and criteria for evaluation of raceway system supports, including limited analytical reviews with equivalent static load factors. The BVAC duct methodology adopts a similar approach, revising the load factors to account for differences in damping between raceways (as evidenced by dynamic testa) and HVAC ducts (limiting damping to St of critical for Design Basis Earthquakes as stated in the UFSAR for bolted steel structures).

4. The SQUG GIP includes detailed caveat& and inclusion rules based on earthquake experience and dynamic teat programs for each equipment class and raceways. Similar caveats and inclusion rules, based on review of the same earthquake experience data base and other industry HVAC duct system teat programs, are included in the methodology.

s. The GIP criteria for spatial seismic interaction evaluations are used directly.
The acceptable stress limits for Design Basis Earthquake loading are baaed on the industry standard working level streas allowablea, increased by standard factors used for nuclear seismic designs. For pressure loading, stress limits for the ducting sheet metal and duct stiffeners are in accordance with the requirements of industry standard codes (SMACNA). Implementation of this methodology demonstrates that the existing installations are adequate for seismic and pressure loadings. These implementation results exemplify that the new methodology baa margins of safety consistent with or exceeding the original design basis. 3.8-79a SGS-UFSAR Revision 16 January 31, 1998
3.8.4.5 Materials, Quality Control, and Special Construction Techniques 3.8.4.5.1 Masonry Walls There is no masonry block construction in the Containment and Fuel Handling Building. In the Auxiliary Building and penetration area, removable block walls are reinforced with steel bars and also anchored to the slab. These provisions are used to prevent the wall from collapsing under earthquake forcesi however, they are not considered as major shear walls to carry the lateral forces for the building. 3.8-79b SGS-UFSAR Revision 16 January 31, 1998
All of the masonry walls that have been installed within (and between) Category I structures and adjacent to Category I tanks have been re-evaluated for seismic loadings and are found to be within the following two groups:
1. Those walls whose collapse would endanger or affect in any way the safety of any Category I structure

2. Those walls whose collapse would not affect the safety of any Category I structure A structural analysis has been performed on each of the walls whose collapse would affect any of the Category I structures to determine the shear and bending stresses to assess their margin of safety.
For the re-evaluation and analysis, the masonry wall field testing and inspection, the design for the corrective action, the structural steel reinforcing, and the drawings, see "Report on Re-evaluation of Masonry Walls," dated November 28, 1980, which was submitted to the NRC on December 10, 1980. Corrective actions for those masonry walls which do not meet the NRC criterion (1.33 times allowable ACI shear or tensile stress for mortar when the wall is subjected to out of plane bending during an SSE) are detailed in a PSE&G letter (Liden to Varga) dated December 8, 1982. 3.8.5 Foundations 3.8.5.1 Description All major Category I structure foundations, except the tank foundations, are built on top of lean concrete fill, which in turn bears on the Vincetown formation at approximately Elevation 30 feet. The actual elevation of top Vincetown was established by visual inspection and additional borings after the excavation was completed. The Service Water Intake Structure was constructed within a 3.8-80 SGS-UFSAR Revision 17 October 16, 1998
steel sheet cofferdam. The material within the cofferdam has been removed down to the Vincetown formation and all the less compact materials on top of the formation were removed and replaced with tremie concrete to the bottom of the mat. The Category I water storage tank foundation is a 3-foot concrete mat serving as the top of a pipe trench. The trench foundation rests on compacted backfill brought up from the top of the lean concrete fill at Elevation 79 feet. The profile of the principal plant structures, cofferdams, and subsurface formations are shown on Plant Drawing 201012. A summary of foundations for plant structures is given in Table 3.8-11. Seismic separation joints for building foundation mats adjacent to each other are provided to allow independent motion of each building under earthquake conditions. 3.8.5.2 Applicable Codes, Standards, and Specifications All foundations were designed according to all of the applicable sections of the same codes, standards, and specifications as the buildings and structures which they support. These are listed in Sections 3.8.1.2, 3.8.3.2, and
3. 8. 4. 2.
3.8.5.3 Loads and Load Combinations All foundations were designed according to all of the applicable loads and load combinations as the buildings and structures which they support. These are listed in Sections 3.8.1.3, 3.8.3.3, and 3.8.4.3. 3.8.5.4 Design and Analysis Procedures The containment base mat is analyzed as a rigid circular plate subjected to loadings from the axisymmetric exterior cylinder 3.8-81 SGS-UFSAR Revision 27 November 25, 2013
wall, crane wall, interior walls, and equipment acting around an equivalent circle. The soil pressure is found in a conventional manner without the benefit of its elastic formation. Our manual analysis was based on the ACI Paper, Title No. 63-63, "Analysis of Circular and Annular Slab for Chimney Foundation," by Kuang-Han Chu and omar F. Afandi. A finite element program was used to check the rebar under five loading combinations. Since the mat is covered by 2 to 5-feet thick of concrete slab and also the lower 34 feet of cylinder liner is insulated, the thermal effect on the mat has been neglected. The Service Water Intake Structure and Category I water tank foundation seismic design has been based on the manual dynamic model analyses using the average response spectra as the ground motion input. Other category I structure foundation mats were designed for all loading combinations as described in Section 3.8.4.3. 3.8.6 References for Section 3.8
1. Timoshenko and Woinowsky-Kriegar, "Theory of Plates and Shells," McGraw-Hill, 1959.

2. Timoshenko and Goodier, "Theory of Elasticity," McGraw-Hill, 1951.

3. "Structural Analysis of Containment Vessel - Salem Nuclear Generating Station," Conrad Associates, van Nuys, california, 1970.

4. Maugh, L. c., "Statically Indeterminate Structures," John Wiley and Sons, New York, New York, 1946.

5. VTD 32023 7-01, "Design Basis Response Analysis of the Salem Nuclear Generating Station Structures," EQE Final Report, January, 1995.

6. 1995 ASME Boiler & Pressure Vessel Code, Section III, Division 2, Code for Concrete Reactor Pressure Vessels and Containments.

7. ACI 359-95 (ASME Boiler and Pressure Vessel Code section III Division 2),
"Code for Concrete Reactor Vessels and Containments."
3.8-82 SGS-UFSAR Revision 17 October 16, 1998
TABLE 3.8-1 STRESSES IN HOOP REINFORCING BAR NO. 1 (KIPS/IN. 2) Load Combination El. Critical (feet) A E Case 289.71 288.66 285.42 280.14 272.96 264.12 253.87 242.54 230.46 227.38 224.28 221.15 218.00 51.75 46.36 38.80 4.48 -5.06 A 215.00 55.45 49.58 41.41 4.54 -5.21 A 210.00 56.46 50.60 42.35 4.13 -5.35 A 200.00 55.27 49.69 41.52 3.26 -5.48 A 190.00 56.07 50.72 42.63 4.24 -5.62 A 180.00 56.19 51.29 43.42 4.27 -5.74 A 170.00 55.87 51.56 43.94 3.30 -5.83 A 160.00 55.54 51.87 44.55 3.57 -5.96 A 150.00 55.38 52.34 45.21 3.82 -6.12 A 140.00 58.50 56.00 48.51 4.36 -6.42 A 130.00 57.35 55.65 48.27 4.45 -6.91 A 120.00 54.42 53.80 46.99 4.47 -7.80 A 110.00 48.15 49.17 44.28 4.72 -7.03 B 100.00 37.46 40.20 34.44 3.44 -3.16 B 92.00 24.04 28.78 24.78 -3.12 -1.93 B 84.00 12.58 20.30 20.91 -4.62 -3.77 c 76.00 4.59 10.83 6.05 -3.45 -2.06 B
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-2 STRESSES IN MERIDIONAL REINFORCING BAR NO. 1 2 (KIPS/IN. ) Load Combination El. Critical (feet) AB c D E Case 289.71 54.12 47.53 38.58 -1.77 -4.27 A 288.66 56.63 49.10 39.69 -2.67 -4.28 A 285.42 49.56 43.25 35.13 -2.51 -4.24 A 280.14 55.96 48.80 39.65 -3.37 -4.59 A 272.96 50.16 44.15 36.02 -2.80 -4.96 A 264.12 52.09 45.89 37.48 -3.28 -4.98 A 253.87 52.12 46.02 37.66 -3.77 -4.56 A 242.54 52.25 46.35 38.05 -4.34 -4.42 A 230.46 52.53 46.83 38.56 -4.88 -4.55 A 227.38 52.29 46.80 38.62 -5.21 -4.65 A 224.28 52.39 47.02 38.86 -5.42 -4.61 A 221.15 48.97 44.02 36.44 -5.65 -4.24 A 218.00 42.67 38.40 31.87 -5.89 -3.47 A 215.00 43.69 39.47 32.87 -7.02 -3.14 A 210.00 43.02 39.26 32.87 -7.49 -3.33 A 200.00 42.28 39.24 33.15 -8.19 -3.58 A 190.00 41.39 39.24 33.51 -9.08 -3.83 A 180.00 40.59 39.41 34.03 -10.09 -4.10 A 170.00 39.79 39.63 34.63 -11.19 -4.39 A 160.00 38.96 39.89 35.29 -12.46 -4.69 B 150.00 38.61 40.64 36.37 -13.72 -5.14 B 140.00 38.51 41.67 37.70 -14.96 -5.60 B 130.00 37.97 42.34 38.76 -16.19 -5.91 B 120.00 37.79 43.39 40.13 -17.40 -6.13 B 110.00 34.68 41.78 39.32 -18.60 -5.90 B 100.00 26.86 35.52 34.84 -20.03 -5.53 B 92.00 14.23 19.96 18.76 -13.83 -3.86 B 84.00 18.79 25.71 25.82 -15.56 -4.65 c 76.00 19.29 19.49 15.59 -11.99 -3.21 B
l;GS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-3 STRESSES IN HOOP REINFORCING BAR NO. 2 (KIPS/IN. 2 ) Load Combination El. Critical (feet) A Case 289.71 54.76 47.56 38.41 -2.25 -4.36 A 288.66 53.77 46.83 37.94 -2.31 -4.35 A 285.42 53.24 46.52 37.85 -2.47 -4.33 A 280.14 53.59 47.00 38.38 2.31 -4.29 A 272.96 53.89 47.46 38.92 2.81 -4.22 A 264.12 53.95 47.74 39.37 3.43 -4.11 A 253.87 53.89 47.98 39.83 4.25 -3.99 A 242.54 47.61 42.75 35.69 4.00 -3.93 A 230.46 49.54 44.51 37.26 4.54 -4.01 A 227.38 50.42 45.32 37.99 4. 72 -4.04 A 224.28 52.06 46.73 39.16 4.88 -4.48 A 221.15 53.91 48.32 40.49 4.99 -4.78 A 218.00 51.75 46.36 38.80 4.48 -5.06 A 215.00 55.50 49.64 41.49 4.60 -5.22 A 210.00 56.54 50.67 42.42 4.24 -5.36 A 200.00 55.34 49.75 41.61 3.40 -5.50 A 190.00 56.12 50.76 42.68 4.07 -5.63 A 180.00 56.21 51.31 43.44 4.11 -5.75 A 170.00 55.87 51.56 43.95 3.44 -5.83 A 160.00 55.53 51.87 44.55 3.71 -5.96 A 150.00 55.37 52.34 45.22 3.97 -6.13 A 140.00 58.49 56.01 48.61 4.55 -6.44 A 130.00 57.39 55.70 48.50 4.69 -6.97 A 120.00 54.54 53.93 47.33 4.76 -7.86 A 110.00 48.38 49.38 44.49 4.92 -7.04 B 100.00 37.71 40.53 35.59 3.81 -3.42 B 92.00 24.17 29.01 26.11 -3.49 -2.31 B 84.00 12.46 20.27 21.30 -4.87 -3.93 c 76.00 3.82 10.52 8.07 -3.70 -2.35 B
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-4 STRESSES IN MERIDIONAL REINFORCING BAR NO. 2 (KIPS/IN. 2 ) Load Combination El. Critical (feet) D E Case 289.71 288.66 285.42 280.14 272.96 264. 12 253.87 242.54 230.46 227.38 224.28 221.15 218.00 215.00 210.00 200.00 190.00 180.00 170.00 160.00 150.00 140.00 130.00 120.00 110.00 100.00 92.00 15.70 22.42 22.37 -15 .OS -4.55 B 84.00 17.44 25.36 25.84 -16.61 -4.99 c 76.00 10.97 16.11 16.15 -10.89 -2.31 c
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-5 STRESSES IN HOOP REINFORCING BAR NO. 4 (KIPS/IN. 2 ) Load Combination El. Critical (feet) A B c D E Case 289.71 288.66 285.42 280.14 272.96 264.12 253.87 242.54 46.98 42.17 35.21 4.00 4.64 A 230.46 48.93 43.95 36.80 4.54 4.79 A 227.38 49.78 44.73 37.50 4. 72 4.83 A 224.28 51.36 46.09 38.64 4.88 4.86 A 221.15 53.15 47.62 39.91 4.99 4.79 A 218.00 50.93 45.61 38.19 4.48 4.63 A 215.00 55.09 49.37 41.51 5.14 4.65 A 210.00 56.41 50.53 42.39 5.11 4.60 A 200.00 55.21 49.58 41.79 4.54 4.67 A 190.00 55.83 50.49 42.63 2.74 4.79 A 180.00 55.70 50.87 43.12 2.87 4.96 A 170.00 55.24 51.04 43.58 4.60 5.18 A 160.00 54.84 51.29 44.09 4.83 5.37 A 150.00 54.65 51.74 44.88 5.12 5.55 A 140.00 57.87 55.53 48.93 6.01 5. 71 A 130.00 57.15 55.63 49.90 6.59 5.65 A 120.00 55.09 54.55 49.63 7.03 5.21 A 110.00 49.88 50.75 45.91 6.44 4.45 c 100.00 39.52 42.91 44.42 6.78 5.43 c 92.00 25.09 30.73 36.48 -6.37 5.27 c 84.00 11.48 20.02 24.32 -6.84 -3.65 B 76.00 -2.36 -11.62 -26.43 6.76 6.26 c
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-6 STRESSES IN HOOP REINFORCING BAR NO. 5 (KIPS/IN. 2) Load Combination El. Critical (feet) D E Case 289.71 48.60 41.93 33.75 -2.25 3.68 A 288.66 51.25 44.52 36.03 -2.31 3.73 A 285.42 51.85 45.25 36.80 -2.47 3.80 A 280.14 52.57 46.07 37.62 2.31 3.90 A 272.96 53.04 46.68 38.28 2.81 4.02 A 264.12 53.19 47.05 38.79 3.43 4.21 A 253.87 53.19 47.35 39.30 4.25 4.45 A 242.54 46.98 42.17 35.21 4.00 4.64 A 230.46 48.93 43.95 36.80 4.54 4.79 A 227.38 49.78 44.73 37.50 4.72 4.83 A 224.28 51.36 46.09 38.64 4.88 4.86 A 221.15 53.15 47.62 39.91 4.99 4.79 A 218.00 50.93 45.61 38.19 4.48 4.63 A 215.00 55.14 49.42 41.59 5.21 4.66 A 210.00 56.48 50.60 42.46 5.22 4.60 A 200.00 55.28 49.64 41.88 4.68 4.68 A 190.00 55.88 50.53 42.69 2.57 4.79 A 180.00 55.72 50.89 43.14 2. 71 4.96 A 170.00 55.24 51.05 43.59 4.75 5.18 A 160.00 54.83 51.29 44.09 4.97 5.37 A 150.00 54.64 51.74 44.90 5.27 5.56 A 140.00 57.86 55.54 49.02 6.20 5.74 A 130.00 57.19 55.68 50.13 6.83 5.71 A 120.00 55.21 54.68 49.97 7.31 5.28 A 110.00 50.12 50.96 46.12 6.64 4.47 B 100.00 39.77 43.24 45.57 7.16 5.70 c 92.00 25.22 30.96 37.81 -6.74 5.62 c 84.00 11.36 19.99 24.72 -7.09 -3.80 c 76.00 -3.13 -12.52 -29.32 7.37 7.03 c
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-7 STRESSES IN MERIDIONAL REINFORCING BAR NO. 5 (KIPS/IN. 2 ) Load Combination Critical El.(feet) A B c E Case 289.71 47.15 41.15 33.31 -1.77 4. 77 A 288.66 56.19 48.70 39.36 -2.67 3.60 A 285.42 49.27 42.99 34.91 -2.51 3.57 A 280.14 55.62 48.48 39.39 -3.37 3.88 A 272.96 49.83 43.84 35.76 -2.80 4.24 A 264.12 52.08 45.88 37.48 -3.28 4.18 A 253.87 52.32 46.20 37.80 -3.77 3.67 A 242.54 52.36 46.45 38.13 -4.34 3.44 A 230.46 53.19 47.43 39.06 -4.88 3.50 A 227.38 52.88 47.34 39.07 -5.21 3.52 A 224.28 51.51 46.21 38.19 -5.42 3.31 A 221.15 46.82 42.05 34.81 -5.65 2.75 A 218.00 40.54 36.46 30.26 -5.89 1.84 A 215.00 42.17 38.08 31.71 -7.02 1.43 A 210.00 41.96 38.29 32.07 -7.49 1.52 A 200.00 41.49 38.52 32.55 -8.19 1.59 A 190.00 40.63 38.55 32.93 -9.08 1.65 A 180.00 39.96 38.83 33.55 -10.09 1. 71 A 170.00 39.33 39.21 34.28 -11.19 1. 79 A 160.00 38.74 39.69 35.12 -12.46 1.90 B 150.00 38.52 40.56 36.30 -13.72 2.12 B 140.00 38.54 41.70 37.72 -14.96 2.39 B 130.00 38.12 42.48 38.87 -16.19 2.57 B 120.00 37.69 43.29 40.05 -17.40 2.73 B 110.00 34.39 41.52 39.10 -18.60 -2.47 B 100.00 26.80 35.47 34.80 -20.03 2.65 B 92.00 15.15 22.35 22.93 -15.52 -3.41 c 84.00 13.98 22.53 23.62 -17.02 -4.99 c 76.00 5.90 12.27 13.60 -10.68 -3.71 c
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-8 STRESSES IN DIAGONAL SEISMIC REINFORCING BAR NO. 3 (KIPS/IN. 2 ) Load Combination Critical El. (feet) AB c D E Case 289.71 288.06 285.42 280.14 272.96 264.12 253.87 242.54 29.70 38.30 -21.94 38.30 C&E 230.46 35.58 45.85 -24.89 45.85 C&E 227.38 38.83 50.01 -26.03 50.01 C&E 224.28 41.03 52.83 -26.95 52.83 C&E 221.15 43.43 55.89 -27.90 55.89 C&E 218.00 45.97 59.14 -28.85 59.14 C&E 215.00 31.16 40.06 -18.94 40.06 C&E 210.00 33.87 43.49 -19.44 43.49 C&E 200.00 38.08 48.79 -20.24 48.79 C&E 190.00 42.00 53.67 -21.20 53.67 C&E 180.00 31.71 40.40 -16.04 40.40 C&E 170.00 34.01 43.17 -17.29 43.17 C&E 160.00 36.16 45.72 -20.43 45.72 C&E 150.00 38.19 48.07 -23.69 48.07 C&E 140.00 40.10 50.23 -26.98 50.23 C&E 130.00 41.84 52.25 -30.16 52.25 C&E 120.00 43.36 54.20 -32.78 54.20 C&E 110.00 44.50 56.15 -34.11 56.15 C&E 100.00 36.82 46.63 -27.02 46.63 C&E 92.00 39.96 51.33 -27.02 51.33 C&E 84.00 44.26 60.79 -27.02 60. 79. C&E 76.00 31.59 44.44 -27.02 44.44 C&E
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-9 MAXIMUM TENSILE STRESSES IN LINEAR PLATE (KIPS/IN. 2) Load Combination Test Critical El. (feet) c Pressure Case 289.71 22.8 Test Pressure 288.66 24.6 Test Pressure 285.42 20.5 Test Pressure 280.14 23.2 Test Pressure 272.96 21.0 Test Pressure 264.12 21.6 Test Pressure 253.87 22.3 Test Pressure 242.54 20.3 Test Pressure 230.46 20.1 Test Pressure 227.38 21.6 Test Pressure 224.28 23.3 Test Pressure 221.15 25.2 Test Pressure 218.00 24.2 Test Pressure 215.00 26.4 Test Pressure 210.00 27.6 Test Pressure 200.00 27.6 Test Pressure 190.00 28.8 Test Pressure 180.00 28.3 Test Pressure 170.00 28.3 Test Pressure 160.00 28.2 Test Pressure 150.00 28.4 Test Pressure 140.00 30.8 Test Pressure 130.00 30.9 Test Pressure 120.00 30.1 Test Pressure 110.00 27.3 Test Pressure 100.00 22.6 26.4 25.6 20.7 B 92.00 22.4 27.5 27.3 13.4 B 84.00 17.2 24.8 26.9 10.9 c 76.00 26.0 24.1 22.6 21.6 A
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 3.8-10 INTERACTION COEFFICIENT FOR LINER PLATE IN COMPRESSION Load Combination Critical El.(feet) A B c D E Case 289.71 .163 .225 .213 .026 .282 E 288.66 .019 .108 .123 .029 .283 E 285.42 .128 .194 .190 .023 .284 E 280.14 .063 .143 .151 .018 .285 E 272.96 .191 .251 .237 .010 .287 E 264.12 .153 .223 .216 .005 .290 E 253.87 .093 .175 .181 .000 .297 E 242.54 .144 .227 .230 .009 .311 E 230.46 .127 .223 .237 .015 .333 E 227.38 .116 .223 .243 .023 .346 E 224.28 .048 .177 .210 .025 .345 E 221.15 .038 .170 .210 .027 .349 E 218.00 .106 .235 .269 .035 .358 E 215.00 .029 .204 .272 .048 .409 E 210.00 .055 .245 .318 .054 .435 E 200.00 .107 .317 .400 .074 .475 E 190.00 .103 .345 .447 .095 .516 E 180.00 .101 .372 .492 .117 .556 E 170.00 .113 .412 .545 .139 .595 E 160.00 .131 .455 .601 .172 .634 E 150.00 .158 .507 .662 .208 .673 E 140.00 .135 .516 .688 .247 .715 E 130.00 .180 .582 .759 .292 . 759 E 120.00 .325 .732 .902 .339 .814 c 110.00 .136 .569 .791 .381 .787 c 100.00
* .059 .419 .415 .683 E 92.00 ** .186 .276 .517 A 84.00 76.00 * .051 .473 .157 .293 .344 .723 .454 E
E
*Liner is in tension
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
Security-Related Information - Witheld Under 10 CFR 2.390 TABLE 3.8-11
SUMMARY
OF FOUNDATIONS
1 of 2 SGS-UFSAR Revision 6 February 15, 1987
Security-Related Information - Witheld Under 10 CFR 2.390 TABLE 3.8-11 (Cont.)
2 of 2 SGS-UFSAR Revision 6 February 15, 1987
SECURITY -RELATED INFORMATION-WITHHELD UNDER 10 CFR 2.390 REVISION 6, FEBUARY 15, 1987 PSEG NUCLEAR, L.L. C. CONTAINMENT BUILDING CROSS SECTION SALEM NUCLEAR GENERATING STAT ION Updated FSAR Fiqure 3.8-1
© 2000 PSEG Nuclear, LLC. All Riohts Reserved.
Figure F3.8-2 intentionally deleted. Refer to plant drawing 208900 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
SECURITY-RELATED INFORMATION-WITHHELD UNDER 10 CFR 2.390 REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Reactor Containment Cylinder Wall Reinforcement SALEM NUCLEAR GENERATING STATION Updated UFSAR Figure 3.8-3
Figure F3. 8-4 intentionally deleted. Refer to plant drawing 201102 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
Figure F3.8-5 intentionally deleted. Refer to plant drawing 201105 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
Figure F3.8-6 intentionally deleted. Refer to plant drawing 201108 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
Figure F3.8-7 intentionally deleted. Refer to plant drawing 201175 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
SECURITY-RELATED INFORMATION-WITHHELD UNDER 10 CFR 2.390 REVISION 6 FEBRUARY 15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Reactor Containment Cylindrical Wall Liner - SALEM NUCLEAR GENERATING STATION Quadrant A Updated UFSAR Figure 3.8-8
Figure F3.8-9 intentionally deleted. Refer to plant drawing 201181 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
Figure F3.8-10 intentionally deleted. Refer to plant drawing 201131 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
LINER TEMPERATURE °F 0 0 U) 0 0 U) 8.... 0 0 (") 8
/ 8.... / / /
I I I I I I f0 I ,..: I <~;!' II 0 I Q. 8.... I ....<=! I
\ \ \ \ \ ), ~
Q.
\ 2 <=! \ 8~
0 0 1- \ 1.1.1 a: -rn u.i wZ (!)
\ ;:)
U) 2 a:c \ j::: U) 1.1.1
>z a:
\ Q. !;(~
a:rn ww \ .... 0
a. a:

ea: \
wo 1-U
\ \ \ ....0 \ \ \ \ \ \
I
~----------------.--------r--------.--------r--------~------~-
PRESSURE PSIG REVISION6 FEBRUARY15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Design Pressure-Temperature Transient SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8-11
LINER TEMPERATURE °F 8 C> It) 8N ... C> It) 8 i.... C") N
/
I'
/ /
I I I I I f I co I f8 I I C> I 8... I I I
'\ \ \ \ \ \ \ \ (I) 0 z
I\ g8
\ .... w w (I) cr

: w*
~ \ :::>
(I) :E 0 \ (I) w cr:: j::
<:J \ Cl.
w! \ X cr::o It)
;:)Z "!
r-o
<(o.
C::CI.I ww o.cr::
EC::
\
wo \ 1-0 \ C>
\ \ \ \ \ ~--------~--~r----.----~--~----~ ....
g i i C>
.... REVISION8 PRESSURE PSIG FEBRUARY 15*1987 .1.25 Times Design Pressure-PUBLIC SERVICE ELECTRIC AND GAS COMPANY Temperature Transient SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8-12
LINERTEMPERATURE °F 8 ~ 8N ~ It) 8...
§ (1') N J' ... / /
I I I I I I i? I I I 0 I 8.... I
'\ \ \ \ \ \ \ /\ CoO 0
2 0 c.. 00
\ ow ~ \ w .-c,o u.i 0 \
a: 1- :::> ::E CJ \ CoO CoO j:::
\ w w! a:
a:o c..
>z X 1-0 etc..
a:c.o ww I ...
~
o..a: I
ea: I wo \
1-0
\ \ \
0
' '\ \ \ \ \ \ \
L-----------------.--------r--------~------~~------~-------+- 8.... PRESSUREPSIG REVISION6 FEBRUARY15,1987
*1.50 TimesDesignPressure*
PUBLICSERVICEELECTRICAND GAS COMPANY TemperatureTransient SALEMNUCLEARGENERATINGSTATION Updated FSAR Figure3.8-13
MERIDIONAL REBAR #1 \ MERIDIONAL REBAR #2 - \ \
\
MERIDIONAL REBAR f/3 (SEISMIC) MERIDIONAL REBAR #5 TYPICAL QUADRANT HOOP REBAR #5 HOOP REBAR #4
HOOP REBAR #3 (SEISMIC)
HOOP REBAR f/2 , HOOP REBAR f/1 ._/ / LINER PLATE-_/ t=4.5' PLAN VIEW
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISIONS FEBRUARY 15, 1987 Designation of Main Reinforcement Pattern for Containment Vessel Updated FSAR Figure3.8-14

MERIDIONAL REBAR #2-----.
MERIDIONAL AND HOOP REBAR #3 MERIDIONAl REBAR #1-----.. ,-----MERIDIONAL REBAR 15 HOOP REBAR #1 HOOP REBAR 15 HOOP REBAR 12 HOOP REBAR 114 LINER PLATE
R = 70'-0" R = 70'-5-1/4" R = 70'*7*1/2"
.I R = 70'-9*3/4" R = 71'-2" R = 72'-3-1 /8" R = 73'*9" R = 73' 1/4" R = 74'*1*1/2" SECTION A-A
PUBliC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION 8 FEBRUARY 15,1987 Designation of Main Reinforcement Pattern for Containment Vessel Updated FSAR

Figure 3.8-15
*~ *.., ~~
1ft 0
~
_j ..J
*w .., ..J I
I I I I I I I L ___ l I I Fe r----_J I I car I ,CD I I I I I Fo REVISION6 FEBRUARY15,1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Critical Section for Review at Foundation Mat SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8*16
LOAD COMBINATIONS A B c 0 E SECTION A-A
-14.1 -14.1 -42.8 -13.0 ::::
SECTION B-B 5.5
-I SECTION c-c
( NOTE: ALL UNITSARE KSF ) REVISION 8 FEBRUARY 15, 1987 Radial Stresses Induced in Mat by Factored PUBLIC SERVICE ELECTRIC AND GAS COMPANY Load Combinations SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8*17
LOAD COMBINATIONS A 8 c D E 5 9 0 -I -11.8 8.3 -10.1 12.8 11.3 9.4 -8.3 SECTION A-A
-11.2~~...
SECTION 8-8
-5.3 .... -22.4 -14.9 SECTION c-c (NOTE: ALL UNITSARE KSF)
REVISION8 FEBRUARY15,1887 Tangential Stresses Induced in Mat by Factored PUBLIC SERVICE ELECTRIC AND GAS COMPANY Load Combinations SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8*18
LOAD COMBINATIONS A B c 0 E
-12.3 SECTION A-A -12.3 -11.7 -4.1 -5.8 SECTION B-B -28.3 SECTION c-c (NOTE a ALL UNITSARE KSF)
REVISIONS FEBRUARY15, 1987 Vertical StressesInduced in Matby Factored PUBLICSERVICEELECTRICAND GAS COMPANY Load Combinations SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3.8-19
LOAD COMBINATIONS A B c 0 E -1.1 -3.2 Q3 1.8 1.0 _ _._...;:0.8 ....._ 1.6 _ _......... a3 -2.7 SECTION A-A
~l..
_4.5 7.8 ~~w 4.3 15
-f!i'lll;...._.ol\ I ~ 2.5 ~m SECTION B-B SECTION (NOTE
ALL UNITSARE KSF )
REVISIONI FEBRUARY15,1987 ShearStresses Induced in Matby Factored PUBLICSERVICEELECTRICAND GAS COMPANY LoadCombinations SALEMNUCLEARGENERATINGSTATION UpdatedFSAR Figure3.8-20
FORCE (KIP/FT) FOtce- -
11.5 4 0 100 200 I I. I EL.EV. 71' K/fT
+I +I 0 -*o * -*o RADIALSTRESS (KSF) VERTICALSTRESS (KSF)
REVISION8 FEBRUARY15, 1987 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Forces in.Containment Vessel Due to Dead Load SALEM NUCLEAR GENERATING STATION
~---*----------------------------------------~
Updated FSAR Figure3.8-21
14..1 100 100 I I K.FT/P'T
SHEAR STRESS (KSF)
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION TANGENTIALSl'RESS O<SF) REVISIONS FEBRUARY15,1987 Forces in*containment Vessel Due to Dead load UpdatedFSAR Figure3.8-22
(KIP /FT)
.-----------------*-.------------------~---*--
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY15,1987 Forces in Containment Vessel Due to Dead Load Updated FSAR Figure 3.8-23
~---------------------~---
LINER PLATE J CONTAINMENT WALL OPERATING CONDITIONS NON INSULATEDLINER PLATE}
(INSULATEDLINER PLATE}
CONTAINMENT WALL LINER PLATE_) ACCIDENT CONDITIONS
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY15, 1987 Thermal Gradients in Containment Wall under Operating and Accident Conditions Updated FSAR Figure 3.8-24
.. HOOP STRESSES (KSI) MERIDIONAL STRESSES (KSI) ..,
-*** -1.5
-*** -t.O
-10.5 ELEV. 110-~.......,.- -*** ~
ELE~78--~- -~----------------+-----------------~ LINER
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY 15,1987 Stresses in Liner Plate and Reinforcing Steel Due to Thermal loading under Operating Condition Updated FSAR figure 3.8*25
... HOOP STRESSES (KSI) MERIDIONAL STRESSES (K.Sl)
+
I
-4 .* -z.*
-4.1 -2.1
-2.4 INNER REBAR
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY 15, 1987 Stresses in Liner Plate and Reinforcing Steel Due to Thermal Loading under Operating Condition Updated FSAR Figure 3.8-26

HOOP STRESSES (KSI) MERIDIONALSTRESSES (KSI) .
S.t I.I
4.1 I.e 1.1 2.0 OUTER REBAR

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION8 FEBRUARY15,1987 Stresses. in liner Plate and Reinforcing Steel Due to Thermal Loading under Operating Condition Updated FSAR Figure 3.8*27
.. HOOP STRESSES (KSI) MERIDIONALSTRESSES (KSI) ...
-29.11-----\
- 33.1 t------t t - - - - - 26.9
-3 7. 9 t------t 1----1 -28.6 ELEV. 110 -~--r--
fi a ELEV.78 _ ~..~.--
........ --l1L---------+----------
LINER
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY15, 1987 Stresses in liner Plate and Reinforcing Steel Due to Thermal loading under Accident Condition Updated FSAR Figure 3.8-28

HOOP STRESSES (KStl MERIDIONAL STRESSES (KSI)

11.2 8.4 7.5 9.1 INNER REBAR

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY 15, 1987 Stresses in liner Plate and Reinforcing Steel Due to Thermal Loading under Accident Condition Updated FSAR Figure 3.8*29
... HOOP STRESSES (KSI) MERIDIONALSTRESSES (KSI) ...
14.1
10.7 1.1 7.1 OUTER REBAR

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION8 FEBRUARY15,1987 Stresses in liner Plate and Reinforcing Steel Due to Thermal Loading under Accident Condition Updated FSAR Figure3.8-30

300 MPH 80 MPH CASE* 1 CASE n m
CASE REVISION6 FEBRUARY1$,1987 WindVelocity Distributions PUBLICSERVICEELECTRICAND GAS COMPANY forTornadoConditions SALEMNUCLEARGENERATINGSTATION UpdatedFSAR Figure3.8*31
: 0.42*1.1 COS 28 (0:!9
- o.6e <"2.
f 11i2)
e :£1f, I
X I
\ !
X
'-.___ X V A'-- Hoerne~
PRESSURE DISTRIBUTIONAROUND CIRCULAR CYLINDERS
-22.7.1 Value* ore In unitaof ptf PRESSURE DISTRIBUTIONAPPLIED IN TORNADO ANALYSISOF CONTAINMENTVESSEl REVISIONS FEBRUARY15.1987 Pressure Distributions PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure 3.8-32
q* 0 0 0
'0 0
0 0 0
.0 0. ~
(j
.-o 0
0. fP
<J 011) * .,J. 0 d>*
0~ 0
........ 0 CliO c:::>
0 0 0 0
<(
0 0
<I 3: 0 1-o::
oo Zw r::J 0 0
.* ... *co* .,*
_.~....,..<f. 0 wz *
p 0
5-z-1 (!* 0 0
<(Iii ~
1-z 0 0 0
*~*
0 (1
0 6 *
(.)
.~'&:f. oo*
0
<Oo r;J * **
0 0
** t.*. ... 0
0 (I *
<1.1 ~
u. a:
<[
z z<(
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUClEARGENERATINGSTATION Hot Pipe Penetration With Cooling REVISIONS FEBRUARY15,1987 UpdatedFSAR Figure 3.8-33
VENT
"\ ...J ...J
0. 0.

0. 0.
)

::;)
CoO CoO
0. ...J
)
~ J :E
() a: a:J 0 2
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION 8 FEBRUARY 15,1987 Redundant AirSupply to Penetration Cooling Coils Updated FSAR

Figure 3.8-34
Figure F3.8-35 intentionally deleted. Refer to VTD 301051 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
Figure F3.8-36 intentionally deleted. Refer to VTD 301075 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
SECURITY-RELATED INFORMATION-WITHHELD UNDER 10 CFR 2.390 REVISION 17 OCTOBER 16, 1998 PUBLIC SERVICE ELECTRIC AND GAS COMPANY PRESSURE PIPING - PERSONNEL LOCK SALEM NUCLEAR GENERATING STATION Updated UFSAR Figure 3.8-37
Figure F3.8-38 intentionally deleted. Refer to VTD 301059 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
v LOAD COMBINATIONS=
A B -*-*-- C -----------------..- ~- 22,937. 19,936. -4.915.
~--+-- -3,947.
16,307.
~-r---Ji-_ -3,343.
616. 7.204. 763. 9p&3.
+
901. 9,260.
+ -978.
16,878 20.541 23,581~~ RADIAL FORCE 11100.0 KIP TANG.ENTIAL FORCE 1*1= 20,000 KIP SHEAR FORCE 1'1.: 2POO KIP
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION REVISION6 FEBRUARY 1S. 1987 Forces and Moments in Edge Beam of 18 Feet Diameter Opening at Elevation 140 Feet Due to loadCombination A, B, and C UpdatedFSAR Figure 3.8-39

T LOAD COMBINATIONS*
A 8 -*-* *~ c --.----------
-658. -811.
+ +
-969. ....70. ~*-2.270. -3_o46. -16.159. -19,872. -23,058.~
RADIAL MOMENT t"*IOOO* KIP- FT. TANGENTIAL MOMENT l"aiO,OOOKIP-FT. TORSIONAL MOMENT I"*WOO KIP-FT.
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION REVISION6 FEBRUARY15,1987 Forces and Moments in Edge Beam of 18 Feet Diameter Opening at Elevation 140 Feet Due to Load Combinations A, B, and C Updated FSAR Figure 3.8-40

v LOAD C0~.1BINATION Sr A
B -*-*-- C ~ ----~- ....... - ... ---
-2,859. ~--\----1,967. -1,793.
589. 719.
+ +
830. RADIAL FORCE 1'b 1000KIP TANGENTIALFORCE 1..= 20,000KIP SHEAR FORCE 1'b 2t)OOKIP REVISION6 FEBRUARY 15, 1987 Forces and Moments in Edge Beam of PUBLIC SERVICE ELECTRIC AND GAS COMPANY 9 Feet Diameter Opening at Elevation 104 Feet SALEM NUCLEAR GENERATING STATION Due to Load Combinations A. B and C Updated FSAR Figure 3.8-41
T LOAD COMBINATIONSz A
B -*-*-* c ---------- 3,919. 5,519.
+ +
4,961.
-974. -1,208. -1,377.
RADIAL MOMENT 1\ 1000KIP-FT. TANGENTIAL MOMENT 1'!:10,000 KIP.. FT. TORSIONALMOMENTI"*2000 KIP-FT. REVISION 6 FEBRUARY 15, 1987 Forces and Moments in Edge Beam of PUBLIC SERVICE ELECTRIC AND GAS COMPANY 9 Feet Diameter Opening at Elevation 104 Feet Due to load Combinations A, B. and C SALEM NUCLEAR GENERATING STATION Updated FSAR Figure 3.8-42
v LOAD COMBINATIONS:
A B ~*-*-- C ......... ~._,-~ .... ----- 14,264.-- 12,456.--
-3,142.
9,972.--......_
)It"---\-- - 2, 161. -1,778.
704. 883.
+ 989. +
7,835.---- 9,750.----- 10,853.-----' 1'= 20,000KIP SHEAR FORCE I'~ 2$)00KIP 1 RADIAL FORCE I"=1000KIP TANGENTIAL FORCE REVISION6 FEBRUARY15, 1987 ForcesandMoments in EdgeBeamof PUBLICSERVICEELECTRICAND GAS COMPANY 9 FeetDiameterOpeningatElevation 104 Feet SALEMNUCLEARGENERATING STATION Dueto Load Combinations A, 8, andC Updated FSAR Figure3.8-43
T LOAD COMBINATIONS:
A B -*--*~* C ---~-,_, __ _.,._._ 4,877. 6,225.
+ +
6,700.
-1,169. -1,482 ----- ~---1,643.
RADIAL MOMENT 1=1000KIP-FT. TANGENTIAL MOMENT 1..=IO,OOOKIP-FT. TORSIONAL MOMENT 1"*2DOOKIP-FT. 11 REVISION 6 FEBRUARY 15, 1987 Forces and Moments in Edge Beam of PUBLIC SERVICE ELECTRIC AND GAS COMPANY 9 Feet Diameter Opening at Elevation 134 Feet Due to Load Combinations A, 8, and C SALEM NUCLEAR GENERATING STATION Updated FSAR Figure3.844
I I
/
HOOP FORCE (KIP/FT.) MEMBRANE SHEAR (KIP/FT.) MERIDIONAL FORCE (KIP/FT.) REVISION8 FEBRUARY15,1987 ForceandMoment Contours Around18 FeetDiameter PUBLICSERVICEELECTRICAND GAS COMPANY OpeningDuetoCritical LoadCombination "A" SALEMNUCLEARGENERATINGSTATION UpdatedFSAR Figure3.8-45
0 0 MERIDIONAL MOMENT (KIP-FT./FT.) CIRCUMFERENTIALMOMENT TORSIONAL MOMENT {I<IP-FT./FT.) (KIP-FT./FT.) REVISION6 FEBRUARY15,1987 ForceandMoment Contours Around18 FeetDiameter PUBLIC SERVICE ELECTRIC AND GAS COMPANY OpeningDuetoCritical Load Combination"A" SALEM NUCLEAR GENERATING STATION Updated FSAR Figure3.8-46
=...., -1 2000 2100 !000 ...l..-- * ~
400
~OUTER BOUNDARY OF EDGE BEAM MERIDIONAL FORCE (I<IP/Ft} HOOP FORCE (KIP/FT.) MEMBRANE SHEAR (KIP/FT)
REVISION6 FEBRUARY15,1987 ForceandMomentContours Around9 FeetDiameter OpeningDueto PUBLICSERVICEELECTRICAND GAS COMPANY Critical LoadCombination "A" SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3.8-47
0 0
-----*oo~ ..--o'----- ..,Oo MERIDIONAL MOMENT (KtP...FT./Ft, CIRCUMFERENTIAL MOMENT TORSIONAL MOMENT (KIP-Ft/FT.)
( KIP-Ft/FT.) REVISION6 FEBRUARY15,1987 ForceandMomentContours PUBLICSERVICEELECTRICAND GAS COMPANY Around9 FeetDiameter OpeningDueto Critical LoadCombination "A" SALEMNUCLEARGENERATINGSTATION UpdatedFSAR Figure3.8-48
CONT-"NMIINT INNER WAri.LL STII&L.L.INaR AIR -raST CONN. INSIO& CONTA.INMIE:NT-+----1--------1._....
-.NC:HOR.
INSULATION W&LO JOINT 5~0-=' W&LO FULL. PENETI:rATIO Wf:LO PLAN COOLI~ COIL. PIPIN6 (4) G.UIOES COOLINGCOIL. AIR.TliSTCONN. PIPING SECTIONB-6 REVISION6 SECTIONA-A FEBRUARY15,1987 Piping Penetration PUBLICSERVICEELECTRICAND GAS COMPANY Hot Pipe SALEMNUCLEARGENERATING STATION Updated FSAR Figure3.8-49
I C.ONCr:iETE: FIELD WELD AIQ. TeST CONN.---- STEEL. JR-- IEL..O WELD TVFtiC:AL INSlOE CONTAINMENT I SI-IOP WELD FULL. PE-NETRATION TYPICAL Fle.L-0 X.PANStON ~OINT ASSEME!.L..Y GUICE.-- I REVISION6 FEBRUARY15, 1987 Piping Penetration PUBLICSERVICEELECTRICAND GAS COMPANY Cold Pipe SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3.8-50
~ 4'-6' ,...
CONCRETE
~ ~ ~
AIRTEST CONN.
~~ {1 f
INSIDE CONTAINMENT
~ ~~
PARTIALPENETRA~ FIELDWELD (TYPICAL) ~
~ ... ANCHOR ~ ~GUIDE PUBliCSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION SERVICE WATER PIPING PENETRATION COLD PIPE UpdatedFSAR Sheet1of 1 Revision13,June12,1994 Figure3.8-SM
CRBLF -INLINt CO~NtCib!t L, Cbl l>\le.l~ I VDLW,t REVISION8 FEBRUARY15,1987 TypicalLowandMediumVoltage PUBLICSERVICEELECTRICAND GAS COMPANY Electrical Penetration Assembly SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3.8-51
N..-.--
<£ SLDt.f. /"'EL, /03 ~I~
cL.1o4 '-1!,..
/tCC,JDeNT J.1 ok li!. /.. 6'~0 ItT ltiiY TWC.
SlJPPtJII.r {t!J)I:-L,4S') I ,, I 6 sJ-D
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Reactor Cavity - Plan View REVISION8 FEBRUARY15,1987 Updated FSAR Figure3.8*52
.. -~-
//
SC../1/..G' /~ : II J-I 0
PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Reactor Cavity

Section "A*A" REVISION 6 FEBRUARY 15, 1987 Updated FSAR Figure3.8-53
t.U
....,)
AJ
~ ~ ---~_.;,...:.-1--1-*-t' ---------!--
el,t1/3 '-11 + 3
~__,
~~-
SUL~ I " '
'b-::.1-0
PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATINGSTATION Reactor REVISION6 FEBRUARY15,1987 Cavity*Section"B*B" UpdatedFSAR Figure3.8*54

sr

PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION REVISION6 FEBRUARY 15, 1987 Stresses in a Thick Cylinder UpdatedFSAR Figure3.8-55
Figure F3.8-56 intentionally deleted. Refer to plant drawing 201012 in DCRMS SGS-UFSAR Revision 27 November 25, 2013
3.9 MECHANICAL SYSTEMS AND COMPONENTS The original design basis analysis of the Unit 1 and 2 Reactor Coolant Loop/Supports System was conducted based on an integrated analytical model which included the effects of the supports and the supported equipment. A three-dimensional, multi-mass, elastic-dynamic model was constructed to represent the Reactor Coolant Loop/Supports System (The model is as shown on Figure 3. 6-2) . The seismic floor spectrum at the internal concrete to support interface, obtained from an elastic-dynamic model of the reactor containment internal structure, was used as input to the piping analysis. The dynamic analysis employed displacement method, lumped parameter, stiffness matrix formulations, and assumptions that all components behave in a linearly elastic manner. The proprietary computer code WESTDYN was used in this analysis. I The reanalysis of the reactor coolant loop, which addresses the elimination of snubbers from the steam generator upper supports, is described in Section 3.9.1.8. Normal Operating Loads System design operating parameters were used as the basis for the analysis of equipment, coolant piping, and equipment support structures for normal operating loads. The analysis was performed using a static model to predict deformation and stresses in the system under normal operating conditions. The analysis with respect to the piping and vessels was in accordance with the provisions of USAS B31.1 and ASME Section III. Results of the analysis gave six generalized force components, three bending moments, and three forces. These moments and forces were resolved into stresses in the pipe in accordance with the applicable codes. Stresses in the structural supports were determined by the material and section properties assuming linearly elastic small deformation theory.
SGS-OFSAR 3.9-1 Revision 23 October 17, 2007
Seismic Loads I Analysis for seismic loads was based on dynamic analyses. The appropriate time history motions at the basemat elevation were used as input forcing functions to the detailed dynamic models. The loads developed from the dynamic models were incorporated into a detailed support model to determine the support member stresses. Analysis of blowdown loads resulting from a loss-of-coolant accident (LOCA) was based on the time response of simultaneously applied blowdown forcing functions on a single broken loop dynamic model. The forcing functions are defined at points in the system loop where changes in cross section or direction of flow occur such that differential loads are generated during the blowdown t*ransient. The loads developed from the dynamic model were incorporated into a detailed support model to determine the equipment support member stresses. The stresses in components result from normal loads and the worst case blowdown analysis *were combined with the worst case seismic analysis to determine the maximum stress for the combined loading case. This is considered a very cons*ervative method since i t is highly improbable that both maxima will occur at the s'ame i.nstant. These stresses were combined to determine that the Reactor Coolant I.oop/Supports System will not lose its intended functions under this highly improbable situation to assure the station capability for safe shutdown. The limiting stress criteria to be used in the analysis are defined in Section $.2. 3.9-2 SGS-UI?SAR Revision 23 October 17, 2007
3.9.1 Dynamic Systems Analysis and Testing 3.9.1.1 Vibration Operational Test Program Despite the lack of formal codes pertaining to vibration testing, it has long been considered good engineering practice by Public Service Electric and Gas (PSE&G) observe for unanticipated transients during system testing. The observations include, but are not necessarily limited to, vibration, thermal growth, hydraulic transients, leakage, and the like. Directives to this effect are made part of general test documents rather than of each specific test procedure. The net result is that observations for excessive unanticipated transients will be made during operational and startup testing. Additionally, confirmatory component vibration testing will be performed as part of the preoperational test program in the form of the specific transients listed below. These tests should confirm that the piping and piping restraints within the reactor coolant pressure boundary have been designed to withstand the dynamic loadings from operational transient conditions that will be encountered during service. In addition to visually checking all systems for excessive vibration during the normal course of the preoperational test program, specific attention will be directed to evaluating possible vibration problems during the performance of the specific transients listed below: fie Trans
1. Reactor Coolant System Operational Tests of Heat up Centrifugal Charging Pump (Step Changes)
Reactor Coolant Pump Start Operation of Pressurizer 3.9-3 SGS-UFSAR Revision 6 February 15, 1987
Power-Operated Relief Valves Operation of Pressurizer Spray Valves Operation of Letdown Isolation Valves
2. Reactor Coolant System Operation of Pressurizer at Temperature Power-Operated Relief Valve Reactor Coolant Pumps (Stopping and Starting)

3. Reactor Coolant System Initiation of Residual Heat Cooldown Removal

4. Emergency Core Cooling Initiation and Termination of Full Flow Test the Following:
A. Safety Injection (Safety Injection Pumps) B. Boron Injection (Centrifugal Charging Pumps with Primary Water) C. Safety Injection (Residual Heat Removal Pumps)
5. Chemical and Volume Operational Tests of Positive Control System Test Displacement Charging Pump (Stop and Start)
Systems and components will be visually examined for the following types of deficiencies which are indicative of possible vibration problems:
1. Cracks in the grouting of equipment foundations

2. Leaking gaskets in piping systems and pump seals

3. Leaks from flanged connections in piping systems 3.9-4 SGS-UFSAR Revision 6 February 15, 1987

4. Metal-to-metal contact indications on piping system restraints

5. Water hammer "noises" during transient operations If any of the above types of indications are observed, or the response to any transient is deemed excessive by the cognizant test engineer, the response will be measured, evaluated, and documented. Methods available to measure the magnitude of piping system vibrational responses include: direct observation and measurement by scale for low frequency, large amplitude responses, and the use of stroboscopic equipment and inertial vibration sensors for high frequency, low amplitude responses.
No speci fie "go-no go" criteria have been established for determining the acceptance of piping sys terns or components in terms of vibration requirements. Confirmation of structural integrity of vibration systems will be based upon favorable comparison of vibration-induced and code allowable stresses. In the event any structural or piping modifications become necessary as a result of excessive vibration responses, the specific transient(s) will be repeated to assure vibrational acceptability. 3.9.1.2 ~ynamic Testing Procedures Safety-related mechanical equipment was analyzed for seismic and accident loading conditions. However, seismic qualification type testing as recommended in WCAP - 7558, "Seismic Vibration Testing With Sine Beats," by A. Morrone, October 1971, is performed for equipment which is too complex for analysis. The integrity of the RCS for normal vibratory loading conditions is verified by preoperational flow tests. Quality Assurance Engineers perform examinations of selected components. Particular attention is given to selected areas of the reactor vessel internals. The operating mode is evaluated in the analysis or testing of 3.9-5 SGS-UFSAR Revision 6 February 15, 1987
safety-related mechanical equipment whenever the component is required to function following the vibratory loadings. Safety-related air-operated valves supplied by PSE&G were either seismically tested or analyzed. Those valves which were vibration tested were shown to be operable during a seismic event through cycling of the valve. These tests demonstrated operability of the actuator and other vital appurtenances physically attached to the valve which are necessary for its proper operation. Those valves which were analyzed were demonstrated to have natural frequencies above at least 30 Hz and were therefore considered to be rigid components without amplification due to a seismic event. Motors associated with safety-related pumps provided by PSE&G have been tested or analyzed for operability under postulated seismic conditions as specified in Section 3.10. For pumps and valves in the Westinghouse Scope of Supply, vital appurtenances which are attached to an active pump and motor or an active valve operator were designed and installed using the applicable standards and engineering practice in use at the time of ordering equipment. The system containing the active equipment will undergo cold hydro testing and hot functional testing prior to startup. After startup, the components will be inservice inspected and tested as required at regular specified intervals to assure continued functional availability. I 3.9.1.3 Dynamic Testing and Examination of Reactor Internals A program consisting of extensive design analysis, model testing, and post hot functional inspection was used to establish internals integrity. Additionally, Westinghouse has instrumented full size reactors to measure the dynamic behavior of the first~of-a-kind of each size plant and has compared measurements with predicted values. This program was instituted as part of a basic philosophy of westinghouse to instrument the internals of the first-of-a-kind of 3.9-6 SGS-UFSAR Revision 20 May 6, 2003
the current nuclear steam supply system design. The magnitude of this test program was much greater than the intent of the philosophy, and was established as part of an extensive plan to develop theories and basic concepts related to internals vibration under various operating conditions. Thus, not only is added assurance obtained that all of the hardware will operate in the manner for which it was designed, but these data also assist in the development of increased capability for the prediction of the dynamic behavior of Pressurized Water Reactor (PWR) internals. The previous "first-of-a-kind 11 plants that were instrumented were Zorita, one loop; Ginna, two loop; and Robinson, three loop. Indian Point 2, a four loop plant, has been the most thoroughly instrumented plant to date, for the reasons presented above. The Indian Point 2 reactor was established as the prototype for the Westinghouse four loop plant internals verification program. Subsequent four loop plants are similar in design. Past experience with other reactors indicates that plants of similar designs behave in a similar manner. For these reasons a comprehensive instrumentation program was conducted on the Indian Point 2 Plant to confirm the behavior of the reactor components. The main objectives of this test were to increase confidence in the adequacy of the internals by determining stress or deflection levels at key locations and to obtain data that can be used to develop improved analytical tools for prediction of internals vibration. In the final analysis, the proof that the internals are adequate, free from harmful vibrations, and perform as intended is through component observations and examinations during service. With this thought, Indian Point 2 (the 5 loop prototype) was subjected to a thorough visual and dye penetrant examination by a qualified Westinghouse Quality Assurance Engineer before and after the hot functional test. This inspection was in addition to the normal inspection of the internals in the shop, and before and after shipment. 3.9-7 SGS-UFSAR Revision 6 February 15, 1987
Safety Guide 20, Paragraph D, "Regulations for Reactor Internals Similar to the Prototype Design," is satisfied for subsequent four loop plants in the following manner~ The internals were subjected to a thorough examination prior to preoperational flow tests. This examination included the 35 points shown on Figure 3.9-2 (Sheet 2). These 35 points included the following:
1. All major load bearing elements of the reactor internals were relied upon to retain the core structure in place

2. The lateral, vertical, and torsional restraints were provided within the vessel

3. Those locking and bolting devices whose failure could adversely affect the structural integrity of the internals.

4. Those other locations on the reactor internal components which were examined on the Prototype Indian Point design.
The interior of the reactor vessel was also examined for evidence of loose parts or foreign material. Specifically, the inside of the vessel was inspected before and after the hot functional test, with all the internals removed, to verify that there were no loose parts or foreign material. Lower Internals A particularly close inspection was made on the following items or areas, using a SX or lOX magnifying glass or PT where applicable. The locations of these areas are shown on Figure 3.9-2 (Sheet 1).
1. Upper Barrel Flange and Girth Weld 3.9-8 SGS-UFSAR Revision 6 February 15, 1987

2. Upper Barrel to Lower Barrel Girth Weld

3. Upper Core Plate Aligning Pin. Examine for any shadow marks, burnishing, buffing, or scoring. Check for the soundness of lockwelds.

4. Irradiation Specimen Basket Welds

5. Baffle Assembly Locking Devices. Check for lockweld integrity.

6. Lower Barrel to Core Support Girth Weld

7. The Flexible Tie Connections (Flexures) at the Lower End of the Thermal Shield

8. Radial Support Key Welds to Barrel

9. Insert Locking Devices. Examine soundness of lockwelds.

10. Core Support Columns and Instrumentation Guide Tubes.
Check all the joints for tightness and soundness of the locking devices.
11. Secondary Core Support Assembly Welds

12. Insert Locking Devices. Examine soundness of lockwelds.

13. Lower Radial Support Lugs and Inserts. Examine for any shadow marks, burnishing, buffing, or scoring. Check the integrity of the lockwelds. These members supply the radial and torsion constraint of the internals at the bottom relative to the reactor vessel while permitting axial growth between the two. One would expect to see, on the bearing surfaces of the key and keyway, burnishing, buffing, or shadowing marks which would indicate pressure loading and relative motion 3.9-9 SGS-UFSAR Revision 6 February 15, 1987
between the two parts. Some scoring of engaging surfaces is also possible and acceptable.
14. Bearing Surfaces of Upper Core Plate Radial Support Key

15. Mounting Blocks Thermal Shielding to Core Barrel.
Examine the connections for evidence of change in tightness or lockweld integrity.
16. Gaps at Baffle Joints. Check for gaps between baffle and top former and at baffle to baffle joints.
Upper Internals A particularly close inspection was made on the following items or areas, using a magnifying glass of SX or lOX magnification where necessary. The locations of these areas are shown on Figure 3.9-2 (Sheet 1).
1. Thermocouple Conduits, Clamps, and Couplings

2. Guide Tube, Support Column, and Thermocouple Column Assembly Locking Devices

3. Support Column and Conduit Assembly Clamp Welds

4. Radial Support Keys and Inserts Between the Upper Core Plate and Upper Core Barrel. Examine for any shadow marks, burnishing, buffing or scoring. Check the integrity of lockwelds.

5. Connections of the Support Columns and Guide Tubes to the Upper Core Plate. Check for tightness.

6. Thermocouple Conduit Gusset and Clamp Welds 3.9-10 SGS-UFSAR Revision 6 February IS, 1987

7. Thermocouple End-Plugs. Check for tightness.

8. Guide tube closure welds, tube-transition plate welds, and cadwelds Acceptance standards were the same as required in the shop by the original design drawings and specifications.
During the hot functional test, the internals will be subjected to a total operating time at greater than normal full flow conditions (four pumps operating) of at least 10 days or 240 hours. This provides a cyclic loading of greater than one million cycles on the main structural elements of the internals. In addition, there will be some operating time with only one, two, and three pumps operating. Therefore, when no signs of abnormal wear are found or of harmful vibration being present in the core support structures, and with no apparent structural changes taking place, the four loop core support structures are considered adequate. They perform their function as intended and are free from harmful vibrations. This program of pre- and post-hot functional examination is identical to that which will be used for all subsequent four loop core support structures. 3.9.1.4 Correlation of Test and Analytical Results Results of measurements taken in models and prototypes are used to refine the methods of analysis and to establish inputs to the LOCA computations. Measurements taken when performing shaker tests on the prototype at the shop and results obtained during hot functional tests ensure that the natural frequencies, normal modes~ and damping values used in the analysis can be justified in a realistic manner. Specifically, this was done to confirm the analysis results for the guide tube beam modes, and barrel beam and shell modes. By comparing the results obtained in water and 3.9-11 SGS-UFSAR Revision 6 February 15, 1987
air, the effect of immersion is taken into consideration. Results of the prototype tests also allow the effect of the inertial mass of water to be incorporated into the analysis. The distribution of the mass of the water on the internals components is found from the results of the behavior of the I components during vibration measurements. 3.9.1.5 Dynamic Analysis Methods for Reactor Internals A 0.10 inch clearance is intentionally provided between the upper core plate and guide tube as specified in paragraph 3.2.1.4 of WCAP-7332-L. The highest vertical loads acting on the internals during a hot leg break are due to impacting as the internals are deflected by the transient hydraulic forces. Table 1-1 of WCAP-7332-L shows the upper core plate maximum deflection under these loads to be 0.014 inch; well below the 0.10 inch allowable limit. Many subsequent analyses on other internals show no deflection greater than o. 025 inch. Since a clearance is always present, no significant direct axial loads are induced in the guide tube. An elastic system dynamic analysis and an elastic component analysis were used for the reactor internals analysis under blowdown and seismic excitation. 3.9-12 SGS-UFSAR Revision 20 May 6, 2003
(THIS PAGE IS IN'I'ENTIONALL BLANK) 3.9-13 SGS-UFSAR Revision 20 May 6, 2003
The reactor internals are analyzed for postulated pipe rupture of the main coolant loop (DBA), as limited by the NRCs approval of Leak-Before-Break (Reference 6 in Section 3.9.5). For vertical excitation, the reactor internals are represented by a multi-mass system connected with springs and dashpots simulating the elastic response and the viscous damping of the components. The effects of clearances between various internals, snubbing action caused by solid impact, and preloads in hold down springs have been incorporated in the analytical model. Various reactor internal components are also subject to transverse excitation during blowdown. Specifically, the barrel, guide tubes, and upper support columns are analyzed to determine their response to this excitation. A blowdown digital computer program which is developed for the purpose of calculating local fluid pressure, flow and density transients that occur during a LOCA is applied to the subcooled, transition and saturated two-phase blowdown regimes. This code is based on the method of characteristics wherein the result of ordinary differential equations, obtained from the laws of conservation of mass, momentum, and energy are solved numerically using a fixed mesh in both space and time. Predictions of this code have been compared with numerous test data and the results show good agreement in both the subcooled and the saturated blowdown regimes. The appropriate dynamic differential equations for the models of the internals are formulated and the responses are determined by performing a time-history analysis using the forcing functions from the aforementioned pressure and velocity transients. The results obtained from the linear analysis of the vessel internals indicate that during blowdown, the relative displacement between the components will close the gaps and consequently the structures will impinge on each other, making the linear analysis unrealistic and forcing the application of nonlinear methods to study the problem. It is clear that linear analysis will not provide information about the impact forces generated when 3.9-14 SGS-UFSAR Revision 33 October 24, 2022
components impinge each other, but can, and is, applied prior to gap closure. The effects of the gaps that would exist between vessel and barrel, between fuel assemblies and baffle plates, and between the control rods and their guide paths are considered in the analysis. Further details of the method of analysis are given in "Reactor Internals Response Under a Blowdown Accident," First International Conference on Structural Mechanics in Reactor Technology, Berlin, September 20-24, 1971 by G. J. Bohm and J. P. LaFaille. 3.9.1.6 Analytical Methods for ASME Code Class 1 Components The ASME Section III Nuclear Power Plant Components Code is inapplicable to the Salem Station; hence, the normal, upset 1 emergency, and faulted conditions terminology does not apply. However, since the RCL loop vessels (reactor vessel, pressurizer, and steam generators) are basically standard components, analysis of these vessels with the more recent ASME Code conditions (normal, upset, emergency 1 and faulted) have been performed with the .load combinations and associated stress limits for ASME Code Class 1 components and supports given in Section 5.2. Load combinations and stress limits for ASME Code Class 2 and 3 and balance-of-plant components and piping are combined as discussed in Section 3.9.2. As-built safety-related piping system stress analysis calculations and pipe support re-evaluation is discussed in Section 3.9.3. For the RCS, the square-root-sum-of-the-squares (SRSS) method of load combination has been utilized to combine the effects of LOCA and design basis earthquake. For the integrated head assembly, SRSS was utilized for combining the LOCA and design basis earthquake loads. Justification of the SRSS method *I is provided in Westinghouse Topical Reports 1 WCAP-9283 and WCAP-9279. A plastic instability analysis of the support and supported system was not needed since the adequacy was proved by elastic analysis. 3.9.1.7 Component Supports Information on component supports is included in Section 3.9.3 and in applicable sections covering the component and its support as parts of an integral system in terms of design and analysis. Design information concerning bolted connections for linear component supports is presented in Appendix 3.9A. 3.9-15 SGS-UFSAR Revision 22 May 5, 2006

~--------~------------------------------------

The report, "Evaluation of the Reactor Coolant System for Salem Unit No. 1 and 2," as amended by PSE&G letter dated March 6, 1979, (R. L. Mittl too. D. Pan) provides the buckling design criteria which was used for all ASME Class 1 component supports subjected to faulted condition loading combinations and justification if criteria exceeds the limits of Paragraph F-1370(c) of the ASME Section I I ICode, Appendix F. The PSE&G report, "Evaluation of the Reactor Coolant System Considering Subcompartment Pressurization Following a LOCA for Salem Units 1 and 2," was submitted on March 6, 1979, and revised on March 29, 1979. Section 6.1 of this report states that the computer program WESAN was used to analyze the steam generator and reactor coolant pump supports for the effects of asymmetric pressure loads combined directly with LOCA loop depressurization loads. The design control measures as required by Appendix B of 10CFR50, that were used to demonstrate the applicability and validity of the WESAN program are given in the I?SE&G submittal entitled "Explanation and Verification of the Computer Program WESAN," dated January 8, 1980. 3.9.1.8 Dynamic Analysis of the Reactor Coolant Loop In order to eliminate the maintenance and testing required for steam generator snubbers, all of the steam generator snubbers at Salem Unit 1 and Unit 2 have been deactivated. The two snubbers on the reactor side of each steam generator have been removed. Each of the two backside snubbers on each steam generator has been converted to a rigid, single-acting compression strut, via the addition of a compression collar clamped to the snubber body. A reanalysis of the reactor coolant loop (RCL) and primary equipment supports was performed to evaluate the revised configuration of the steam generator upper supports. This reanalysis includes the parameters and characteristics of the Model F (Onit 1) and Model 61/19T (Unit 2) steam generators to demonstrate that the structural criteria will be met with the respective steam generators in place. A non-linear, time history analysis was performed to evaluate earthquake loadings. Time history analysis has also been performed for the analysis of postulated pipe breaks. Revised loads and stresses were evaluated for the loop piping 1 primary equipment supports, primary equipment nozzles, main steam and feedwater line piping/ auxiliary piping attached to the RCL, and building structure embedments. The application of leak-before-break was also verified to remain applicable to the primary loop piping. 3.9-16 SGS-UFSAR Revision 24 May 11, 2009
The model used in the original seismic analysis was modified to include all four primary loops, the primary (reactor vessel, steam and reactor coolant pumps) and the primary equipment supports. The steam generators and reactor coolant pumps are represented in the model as discreet mass models. A simplified model of the concrete internal structures of the containment building was generated and is coupled to the RCL/supports model. The reactor vessel is also represented by a discreet mass model with the masses lumped at various locations along the length of the vessel and along the of the of the core internaJ.s. The primary supports are included as stiffness matrices. The component upper and lower J.ateral supports are inactive plant cool down, and normal conditions. However, these restraints become active under the rapid motions of the reactor coolant components that occur from the dynamic loadings of postulated earthquake and pipe break events, and are represented by stiffness matrices and/or individual tension or compression spring members in the dynamic model. The analyses are performed at the full power condition. The time history analysis is performed using the WECAN computer code (Reference 4), and employs non-linear modal super-position techniques. From the mathematical description of the system, the overall stiffness matrix is developed from the individ~al element stiffness matrices, from which a reduced stiffness matrix associated with mass of freedom is From the mass matrix and the reduced stiffness matrix, the natural and the normal modes are determined. Time history input motions are at the basemat elevation for input to the model. used in the model is consistent with the values in Section 3.7. The modal superposition method is then used to generate a time history solution for the response of the reactor coolant loop. Three time history motions are applied individually at the basemat elevation representing the north-south, east-west, and vertical earthquakes, for both the OBE and SSE. The results are then combined by adding the vertical earthquake response absolutely with the worst of the two horizontal earthquake responses. 3.9-16a SGS-UFSAR Revision 20 May 6, 2003
The mathematical model used in the static is modified to the tine history pipe break analyses. The natural and are determined fran this loop model. The time history hydraulic forces at the node points are combined to obtain the forces and moments acting at the corresponding structural lumped-mass node points. The dynamic structural solutions are obtained by using a modified predictor-corrector-integration technique and normal mode theory. The postulated pipe breaks included in the take of the of leak-before-break in the primary (Section 3.6.4.2), as well as the elimination of intermediate breaks (Section 3.6.2.1). The breaks considered in the of the RCL are breaks at the loop nozzles of the accumulator lines, RHR lines, and the pressurizer surge line. Also, breaks at the steam generator nozzles of the main steam line and feedwater line are evaluated. In order to determine the thrust and reactive force loads to be applied to the reactor coolant loop during the postulated LOCA, it is necessary to have a detailed description of the t:ydraulic transient. Hydraulic forcing functions are calculated for the ruptured and intact reactor coolant loops as a result o: a postulated LOCA. These forces result from the transient flow and pressure histories in the reactor coolant system. The calculation is in two steps. The first step is to calculate the transient pressure, mass flow rates, and as a function of time. The second uses the results obtained from the hydraulic with of areas and direction coordinates, and calculates the time history of forces at appropriate locations (e.g., elbows) in the reactor coolant loop. The model the behavior of the coolant fluid within the entire RCS. Key parameters calculated by the hydraulic model are pressure, mass flow rate, and density. These are input to the thrust calculation, together with plant layout information, to determine the time-dependent loads exerted by the fluid on the loops. In evaluating the hydraulic forcing functions during a postulated LOCA, the pressure and momentum flux terms are dominant. The inertia and terms are taken into account in evaluation of the local fluid conditio~s in the model. 3.9-16b SGS-UFSAR Revision 25 October 26, 2010
The blowdown hydraulic analysis, performed with the MULTIPLEX computer code, provides the basic information concerning the dynamic behavior of the reactor core environment for the loop forces, including predictions of the flow, quality, and pressure of the fluid throughout the reactor system. The hydraulic analysis considers a coupled fluid-structure interaction by accounting for the deflection of the core support barrel. The depressurization of the system is calculated, using the method of characteristics applicable to transient flow of a homogeneous fluid in thermal equilibrium. The system geometry is represented by a network of one-dimensional flow passages. In the second step, the transient (blowdown) hydraulic loads resulting from a LOCA are calculated using the THRUST code. In the model used to compute forcing functions, the reactor coolant loop system is represented by a model similar to that employed in the blowdown analysis. Each node is fully described by: 1) blowdown hydraulic information, and 2) the orientation of the streamlines of the force nodes in the system, which includes flow areas and projection coefficients along the three axes of the global coordinate system. Each node is modeled as a separate control volume, with one or two flow apertures associated with it. Two apertures are used to simulate a change in flow direction and area. The force components are then summed over the total number of apertures in any one node to give the total forces. These thrust forces provide the input to the RCL piping/support dynamic analysis. When elements of the system can be represented as single acting members (tension or compression members) , they are considered as nonlinear elements, which are represented mathematically by the combination of a gap, a spring, and a viscous damper. The force in this nonlinear element is treated as an externally applied force in the overall normal mode solution. The time history solution is performed using the WESTDYN computer program (Reference 5)
The time history displacement solution of all dynamic degrees of freedom is obtained based on 4-percent critical damping. The time history displacement response of the loop is used in computing support loads and in performing stress evaluation of the reactor coolant loop piping. The support loads are computed by multiplying the support stiffness matrix and the displacement vector at the support point. The support loads are used in the evaluation of the supports. The time history displacements are used to 3.9-160 SGS-UFSAR Revision 20 May 6, 2003
determine the internal forces, deflections, and stresses at each end of the piping elements. For this calculation, the displacements are treated as imposed deflections on the reactor coolant loop masses. The results of this solution are used in the piping stress evaluation. in the evaluation of the RCL piping is the same as that presented in Table 5.2-13. The stress limits for normal and upset condition loadings are taken from the 1967 Edition of the USAS B31.1 Power Piping Code. The loading combinations applied are presented in Table 5.2-12. In the original design, the faulted loading conditions used the Design Limit Curves discussed in Table 5.2-13. For the steam generator snubber elimination analysis, the faulted condition stress criteria has been modified to be more consistent w:Lth industry standards. A 1970 interpretation for the 831.7 piping code {ANSI 831.7, Case 70) established that the faulted allowable stresses should be limited to 2 times l.2S, or 2. 48, where S is the allowable stress defined in 831.1. 'J'his allowable value is in line with industry practice and has been used in the reactor coolant loop reanalysis. The structural integrity of the RCL supports is verified using the load combinations defined in 'I'able 5. 5-3. The stresses are compared to allowable values, also defined in Table 5.5-3, which are based on the American Institute of Steel Construction (AISC) Specification. The supports evaluated in the reanalysis include the steam generator upper and lower supports, reactor coolant pump supports, and reactor vessel supports. Also included is an analysis of the converted backside snubbers. The compression collars are analyzed to demonstrate that stresses are within allowable limits for all applied loads. The snubber bodies are also reanalyzed to address the modified load path resulting from the addition of the compression collar. Results of Analyses The analysis results for the reactor coolant loop demonstrates that the piping stresses remain within the allowable limits. In all support member stresses have been shown to remain within the appropriate allowable values for each loading combination. Also, the primary equipment nozzle loads have been shown to be less than the allowable nozzle loads as specified in the appropriate design specifications. SGS-UFSAR 3.9-16d Revision 23 October 17, 2007
Evaluations of the nozzle displacements resulting from the RCL analysis for the main steam and feedwater lines, and the auxiliary piping attached to the reactor coolant loop, and the assessment of the embedment loads at the primary equipment support attachment locations show that the stresses in these lines and the embedment loads remain within their respective allowable limits.
3.9.2 ASME Code Class 2 and 3 Components Analyses of Westinghouse supplied equipment were performed in accordance with the codes and standards requirements in effect at equipment order date. The codes used for the components of the RCS and Emergency Core Cooling System (ECCS) are given in Sections 5.2 and 6.3, respectively. unlike the 1971 Edition of the ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components Code," these codes did not have design condition categories of "normal, 11 "upset," "emergency," and "faulted 11 conditions and associated stress limits. However, seismic analyses performed for selected Seismic category I components using response spectra that envelop the vertical and horizontal floor response spectra show that the stress for the Operating Basis and Design Basis Earthquake are generally a small fraction of the upset limits of ASME section III, 1971 Edition, or the faulted condition limits of the soon to be published Appendix F of the same code. Similarly, analysis of the Reactor Coolant Loop/Support System shows that the stresses for the combined loading of Design Basis Earthquake (DBE) and DBA are within the Faulted Condition limits given in Section 5.2. All nuclear piping comparable to ASME Code Classes 2 and 3 and the then current Atomic Energy commission Quality Group Classifications was designed to stress criteria conforming to the design philosophy of ANSI B31.1.0, Code for Pressure Piping, Power Piping. As this piping was designed and purchased at a later point in time *than the aforementioned westinghouse-supplied equipment, it was, however, possible to extend the basic design philosophy and stress criteria of ANSI 831.1.0 in a manner paralleling the newer codes. The results of classifying the design stress limits (using ANSI B31.1. o terminology) into 11 Design Condition Categories" (using present day ASME Section III terminology) are given as follows: 3.9-17 SGS-UFSAR Revision 20 May 6, 2003
Design Design Design Loading Condition Stress Component Combinations Category Limit class 2 and Normal pressure, and normal Pc.:::_Sh 3 piping weight and external Pl.:::_Sh loadings Maximum (short time) upset P .:::_l.2S e h pressure and weight p .:::_1.2S 1 h and external loadings and OBE Maximum (short time) faulted p ..:::_1.28
..t c II pressure and weight p ..:::_1.5(1.2)8 =
1 h and external loadings and DBE Where: OBE = Operating basis earthquake DBE = Design basis earthquake Sh "" Allowable "hot" stress from ANSI Bll.l.0-1967 P Circumferential (hoop) stress due to loadings listed c P = Longitudinal stresses due to internal pressure and various external 1 loadings listed 3.9.2.1 Analytical and Emperical Methods for Design of Pumps and Valves The terminology "active components"* and "faulted conditions" is not applicable to the Salem station. This terminology was introduced in codes (ASME Section III, Summer 1968 Addenda) and standards after the code applicability date for the Salem Station. Valves were designed in accordance with USAS Bl6. 5, in general, and ASME Section VIII for flange connections. Valves were
Active components of fluid system (e.g., valves, pumps) are those whose operability is relied upon to perform a safety function such as safe shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the reactor coolant pressure boundary.
I 3.9-18 SGS-UFSAR Revision 20 May 6, 2003
hydrostatically tested to USAS B16.5 rating table with the hydrostatic test followed by a seat leak test to MSS-SP-61 criteria, except that the seat leakage criterion was as specified in the Preliminary Safety Analysis Report. Leak testing is performed to assure that no gross deformation was caused by the hydrostatic test. The components of the reactor coolant boundary which are required to function following a postulated pipe rupture (DBA) are the valves which isolate the normally high pressure systems from the low pressure safety systems. These components are generally designed for deflection limitation and for stress under normal operating conditions which is generally low. The differential pressure loadings on these valves decreases following a postulated pipe rupture of the RCS and, hence, the loadings on the valves due to system depressurization will decrease. The stresses will likewise decrease. The allowable stress that was used for the design of valves was 7000 psi as required by USAS-B16 .5. ASHE Section VIII was employed in the design of gaskets for flange connections only as required by USAS-Bl6.5. 3.9.2.2 Design and Installation Criteria, Pressure-Relieving Devices The design criteria applicable to mounting of nuclear class pressure-relieving devices provides for the system to withstand valve reaction forces due to the concurrent discharge of all valves on any given header in combination with pressure, weight, and seismic forces. The primary longitudinal stresses developed as a result of these combined loads are limited to the values listed in Section 3.9.2. Design procedures include an analysis of the safety and relief valve systems to determine dynamic fluid reaction thrusts. The combination of reaction thrust with pressure, weight, and seismic forces are then used to compute the stresses imposed on valves and piping in accordance with the design philosophy of ANSI B31. 1.0. 3.9-19 SGS-UFSAR Revision 6 February 15, 1987
The computed stresses are compared with the design criteria limits to determine acceptability of the design. Installation provisions - to meet the aforementioned stress criteria include the addition of piping reinforcement, piping restraints, or combinations of both. The design criteria for safety and relief valves in the primary coolant pressure boundary provide for the system to withstand maximum loads due to combinations of valve reaction, weight, and seismic force. Maximum load, as determined in the design analysis of the piping and valves, is absorbed by a system of braces and shell girth bands rigidly connected to the pressurizer shell. The relief valve loads are actually transmitted to this system via a rigid connection on the relief valve body. 3.9.2.3 Field Run Piping Systems Field running of small diameter piping, i.e., complete assembly at the erection point without reference to design drawings, was not permitted for essential systems. Therefore, no special quality assurance measures or performance tests were required. 3.9.3 Seismic Analysis of As-Built Safety-Related Piping Public Service Electric and Gas has conducted the re-evaluation of safety-related piping required by IE Bulletins 79-14 and 79-07. Modifications shown to be necessary by this evaluation have been made and the capability for safe shutdown in a seismic event has been demonstrated. The re-evaluation of pipe stresses and supports was based on current standards and was performed subsequent to verification of "as-built" piping using the piping isometric drawings. 3.9-20 SGS-UFSAR Revision 6 February 15, 1987
The 15 safety-related systems that encompass the total scope of the evaluation are listed in Section 3.7.3.9.1. These systems are those required for safe shutdown of the plant. The re-evaluation of pipe stresses was performed in accordance with the criteria previously submitted to the NRC as part of the responses for Salem 1. This included additional information supplied as a response to IE Bulletin 79-07 (1), a guideline for implementing IE Bulletin 79-14 (2) and a description of the program for the reanalysis and modification of supports (3). The results indicated there are not overstressed pipes. The reanalysis method used to determine pipe support capabilities was based on seismic loads resulting from the three-dimensional SRSS method. This was performed in accordance with procedures submitted to the NRC as part of the r-esponses for Salem Unit 1. The reanalysis resulted in certain modifications of supports. The reanalysis of pipe stresses 1n the 15 safety-related systems indicated no overstressed conditions. However, the pipe supports within those stress calculations resulted in several modifications which have been completed. The majority of the modifications were to the U-bolt and strap anchor supports. These were basically designed to function as a six-way restraint and the pipe stress calculations were performed considering these conditions. However, when individual hanger details were being designed, fabricated, and installed, they did not fulfill the requirements of a rigid anchor as was assumed for pipe str-ess calculations. Hence, theoretically, they did not function as was originally intended. These U-bolt type anchors and strap anchors were redesigned to comply with the original intended functions. 3.9-21 SGS-UFSAR Revision 6 February 15, 1987
3.9.4 Inservice Testing of Pumps and Valves Inservice testing of pumps and valves is performed in accordance with the program originally transmitted to the NRC on April 29, 1981, and as subsequently amended. 3.9.5 References for Section 3.9
1. Letter from Schneider to Grier, "NRC IE Bulletin No. 79-07, Supplemental Response, No. 1 Unit, Salem Generating Station," Sept. 21, 1979.

2. Letter from Schneider to Grier, "NRC IE Bulletin No. 79-14, Supplement 2, No. 1 Unit, Salem Generating Station," Oct. 30, 1979.

3. Letter from Librizzi to Schwencer, "NRC IE Bulletin No. 79-07, Supplemental Response, No. 1 Unit, Salem Generating Station, Docket No.
50-272," Oct. 18, 1979.
4. WCAP-8929, Benchmark Problem Solutions Employed for Verification of the WECAN Computer Program, April 1977.

5. WCAP-8252, Revision 1, Documentation of Selected Westinghouse Structural Analysis Computer Codes, May 1977.

6. Letter from Mr. James C. Stone, NRC, to Mr. Steven E. Miltenberger, PSE&G, dated May 25, 1994, Leak-Before-Break Evaluation of Primary Loop Piping, Salem Nuclear Generating Station, Units 1 and 2.

7. WCAP-13659, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Salem Generating Station Units 1 and 2, May 1993. [Proprietary]
3.9-22 SGS-UFSAR Revision 33 October 24, 2022
THIS FIGL!,fi!E HAS B~EN DELETED R8VlSlOn 18, A ~pr1 126, 2~00 Solem NuclearGenerotin8Station PSEG Nuclear,LLC SEISMIC DYNAMICU DEL SALEM NUCLEAR GENERATINGSTATION UpdatedFSM Figure3.9-1
<D 2000PstG rltle<r,LLC. All Ri ts Res~rved.
I REVISION6 FEBRUARY15,1987 VibrationCheck*OutFunctional Test PUBLICSERVICEELECTRICAND GAS COMPANY Inspection Points SALEMNUCLEARGENERATINGSTATION UpdatedFSAR
Sheet1 of2 Figure3.9-2
FEATURESTO BE EXAMINED I THERMOCOUPLECONOUITCLAMPSINSIDE THE THERMOC<XJPlECOLUMN. 2 CLAMPARRAN(.[MENTSAT THE M<XJNTII\IG BRACKETLOCATIONS. J PLUG TO CONDUIT WELD AT THE FIVE SUPPORT COLUMNS ADJACENTTO THE THERMOCOUPLECOLUMNS. . 4 ACCESS18lEANGLECONDUITCLAMPSINSIDE THE UPPER SUPPORT COLUMNS. 5 ACCESSIBlEWELD JOINTSAT THE THERMOCOUPLESTOP FOR THE SELF INSTRUMENTEDCOLUMIIS. G .WELD JOINTSON ACCESSIBlESUPPORTCOLUMNANO MIXINGDEVICE GUSSETS. (THERMOCOJPL£ SUPPORTHARDwARE)
,.,.~IGIOIT'l' OF EXPOSEDPORTIONOF THE_§,MOC<XJPLECONCUITRUNS.AT ACCESSIBLELOCATIONS.
7 IJNSIOESUPPORTCOlUMNS-LOW£R ENOJ t a: 8 , RIGIDNESSOF TH£ ACCESSIBLEPROTRUDING THERMOCOUPLETIPS
~,--------------------------------------------------------~
9 THERMOCOUPLE COLUMN AND GUIDE TUBE SCREW LOCKINGDEVICES. I 1 i t' ACCESSISt.£~IPPORT COLUI.IIIMIXINGDEVICE.ORIFICE PLATE.AND CORE P\.ATE
'IO, INSERTSCREW LOCKINGDEVICES.
I'. ~ UPPER COOf PlATE INSERTS.
! 121 Ca.tCUITCOI'.NECTORFITTINGSAND CROSS RUN CLAMP ARRANGEMENTS. \ 13! DEEP SEAM WELDS AT THE SKIRT AND AT THE OUTER HOLLOW ROUNDS. -f--1'4! ACCESSIBlE GUIDE TUBE wf..LOS.
151 UPPER BARRELTO FLANGE GIRTH WElD. l r6j UPPER BARREL TO LOWER BARREL GIRTH WELD. 17 LOWER BARREL TO.CORE SUPPORTGIRTHWELD* 18~ UP?ER CORE PLATE ALIGNINGPIN WELDS ANO BEARINGSURFACES. 19 OUTLETNOZZLE INTERFACESURFACEC()IIOITION. 1
.20! THERMALSHIHD FLEXURE ARM. ATTACHMENTSTO BARREL. AND wELD TO THE
THERMALSHIELD. DYE PENETRAII4TINSPECT ALL SIX.
<n ...J 21! THERMALSHIELD INTERFACt::AT THE HANGOFF PAOS. ~
w ~2~
....g. IRRADIATIONSPECIMEN BASKETWELDS.
0:
! SAFFLE ASSEMSI..YSCREW LOCKINGARRANG.EMENTSAT THE TWO TOP AND THE 23 TWO BOTTOMFORMER ELEVATIONS.
w
~_, 1241 COR£ SUPPORT COLUMNTO LOWER CORE PLATE SCREW LOCKING*DEVICES.(24 RANDOMLYCHOSEN iz5j CORE SUPPORT COLUMN ADJUSTINGSLEEVES.
26J ACCESS18LE(2) INSTRUMENTATION GUIDE COLUMN LOCKINGCOLLARSNEAREST THE MA'NWAY, 27 LOCKINGDEVICESOF THE BOTTOM INSTRUMENTATION GUIDE COWMNS. 128 LOCKINGDEVICESOF THE SECONDARY COR£ SUPPORT.
?O ACCESSIBLELOCKINGOE:_VICES OF THE OFF-SET INSTRUMENTATION COLUMN* ..,. QJPPERAND LOWER ENOS) .30' RADIALSUPPORTKEY LOCKING ARRANGEMENTSAND SEARING SURFACES. ~I HEAD AND VESSEL ALIGNINGPIN SCREW LOCKINGDEVICES AND BEARING SURFACES.
132 CONTACTAT INTERFACEOF THE ACCESSISLEINSTRUMENTATION GUIDE COLUMNS. 33 CONTACTAT INTERFACEOF THE ACCESSIBLECORE SUPPORTCOLUMNNUTS.
~~ 34 VESSEL CLEVISLOCKINGARRANGEMENTSAND BEARINGSURFACES.
H - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 1REVISION6
~~~~v-~ ___L_N_o_z_zL_E_r_N_TE_R_.F._AC_E_su
__R_F~ __E_C_ON_~_T_I_ON_.________________________________ ~FEBRUARY15,1987 Vibration Check- Out Functional Test Inspection Points PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Updated FSAR
Sheet 2 of 2 Figure 3.9-2
APPENDIX 3.9A BOLTED CONNECTIONS FOR LINEAR COMPONENT SUPPORT
SGS-UFSAR Revision 6 February 15, 1987
APPENDIX 3.9A ~ BOLTED CONNECTIONS FOR LINEAR COMPONENT SUPPORT 3.9A.l Introduction Appendix XVII-2461 of the ASME Code Section III requires that bolt loads in bolted connections for linear component supports include prying effects due to the flexibility of the connection. The material in this Appendix responds to an NRC request that PSE&G:
1. Provide confirmation that the loads in bolted connections for linear component supports were determined by considering the deformation of the connection and tension-shear interaction for the bolts.
For connections of supports which are anchored to a concrete structure, provide in addition:
a. The type of anchor bolt
~ b. The factors of safety (and their bases) against pullout under static, repeated, and transient loading
2. Provide complete analytical or experimental justification where any connection was assumed to be rigid.
3.9A.2 Design Approaches
1. Tension and shear interactions were considered in developing designs for bolted component supports for piping. The design conservatism on structural members is considered sufficient such that deformation of the connection does not adversely affect the capacity of connections to withstand design loadings.
~ 3.9A-l SGS-UFSAR Revision 6 February 15, 1987
2. Types of anchor bolts used for the various bolted connections in the plant are as follows:

a. The majority of safety-related supports employ connections bolted to concrete inserts which derive their strength from an integral steel coil embedded in the concrete at the time of structure forming

b. The other type of anchor bolt used employs an expandable wedge piece inserted in a prebored hole in the concrete

3. Loads applied to these anchor bolts are within manufacturer 1 s specified limits. A representative analysis of typical standard supports follows in Section 3.9A.3 of this Appendix.

4. The assumption of rigidity for bolted linear support connections, where applicable, is made on the basis that the applied loads to the supports have been determined by analytical methods to be adequate. Refer to Section 3.9A.3 of this Appendix.

5. Large diameter pipes which exert sizable loads on supports are of welded construction. Bolted connections are not used on large diameter pipes in major systems (i.e., Main Steam, Feedwater, Residual Heat Removal, etc.). The concerns with regard to prying effects on bolted linear supports, therefore, do not apply to such piping.

6. The bolted components shown in the analysis of Section 3.9A.3 do not require preload.

7. High strength bolts are not used for the bolted components in question.
3.9A-2 SGS-UFSAR Revision 6 February 15, 1987
8. Piping systems subject to significant transient loads are dynamically analyzed. Support components are designed to handle loads based on results of this analysis.

9. Equipment support bases are supplied by the equipment manufacturers. Design stiffness of these supports for weight and vibration considerations preclude undesirable effects from prying.

10. Piping systems subject to cyclic loading are provided with clamping and/or additional support components to preclude undesirable effects of cyclic loading.

11. A supplemental analysis which considers shear loading is presented in Section 3.9A.3 of this Appendix. The values of length used for determining concrete and anchor bolt strain were made equivalent in accordance with the assumption of linearity shown for the strain diagram used in developing the correlations. It is seen that the correlations yield prying loads approximately the same as originally determined (values varied from 1 percent to 7 percent greater). Hence, the evaluation of the adequacy of the bolted connections to withstand prying effects remains unchanged.
3.9A.3 Representative Analyses of Typical Supports Pages 3.9A-4 through 3.9A-12 present a representative analysis of typical standard supports. Pages 3. 9A-13 through 3. 9A-19 present a supplemental analysis which considers shear loading .
SGS-UFSAR 3.9A-3 Revision 6 February 15, 1987
CONCRETEANCHOR U/S CONC. ASTMA*36 (TYP) I MEMBER { YOUNGSMODULUS-t-E1 I MOMENTOF INERTIA--t-1 I
- T ASTM A30.,..../T~ L I ~--6-----------------------------------------------~*~
(TYP) I
. TYPE "G"FRAMING I CONCRETEANCHOR 0 b UIS CONCRETE TUBEb xdx t ASTMA*36 MEMBER { YOUNGSMODULUS-t- E1 MOMENTOF INERTIA-1 ~-.- - - L--~,.1 SGS-UFSAR TYPE"E"OR "F"FRAMING 3.9A-il REVISION8 FEBRUARY15,1987
/
INPUT PARAMETERS: B* ~(in case of circular I.'s or round tubes -~"8
RADIUS D *%(in case of circular R.'s or round tubes -D

RADIUS A1

TENSILE CROSS-SECTIONAL AREA OF A. BOLT L

SPAN LENGTH h

DEPTH OF MAIN MEMBER I

MOMENT OF INERTIA OF MAIN MEMBER(S)
ADDITIONAL REACTION IN CONCRETE ANCHOR DUE TO PRYING ACTION (~R): 2 r AR* PL afo -~)(L +!!!!
\' 3 Zc Z +.!!)
5 2 U2 L/2 nA 5~ ( BB
- +20-nA1 L .1-kd
SGS-UFSAR 3.9A-5 REVISION6 FEBRUARY15, 1987*
1- EXAMPLE n=8, 2 B=~=2in D=~=2in As =0.61in 2 (1""'BOLT}
,2' 'I' L =45 in, t = 17.8in4 = > 2 C5 x 9, h = 5 kd._ 810 611 + [(810 611) 2 + 2(21(8~(0.611] 112 =1.52 ;n 2 2 1 52 J.d = 2 -3=
1.49.m Zc= 1/2(2) (2) (1.49)= 2.98in3 zs =2(0.61)(2) 5.
{1.49)= 0 73. 3 m 2
.6R = 45 p =0 9P 811.49)(45 + 2(8)117.8)+ 2(17.8))
2.98 0.73 TOTALREACTIONFOR ONE END= R + .6R = 0.5P+ 0. = 1.4P CAPACITYOF RICHMONDTYPE EC INSERT:
T = 10000lbs (SAFEWORKINGLOAD) MAXIMUMHANGERCAPACITY=1.4P= 10000 P = 7143lbs HANGERA2 - SWH - 49 (DWG 236838D4253) COMBINEDLOAD= 4745lb SGS-UFSAR REVISION6 FEBRUARY15,1987
2- EXAMPLE(LOADACTINGAT MIDSPAN): n
8; 8

s-;s

2.62 in; D

30 2,

1.5 in; A1

0.61 in2 (1 .. ~BOLT)
L
43 in; h

4.0 in; I""9.18 in4 C4 x 7.25 kd**~ + a*o.61 z 2*1 5 ~
2.62 2.62 +
2.62 kd * -1.86 + .J3.47 + 5.59 * - 1.86 + 3.01

1.15 in jd* 1.5-1
*~ 5 *1.12 in z
2. 0.61. 1.5. 1.12

0 51 . 3 s 4.0

lA TOTALREACTIONFOR ONE END..,.R + .O.R

0.50 P + 1.41 P

1.91 CAPACITYOF "RICHMOND" TYPE EC - 2W 1" CONC. INSERT:
T
10000 lbs (SAFEWORKINGLOAD)
MAXIMUM HANGERLOAD: 1.91P
10000 ..... pmax

5235 Jbs HANGERA2 - SWH - 50 (DWG. 23683804253)

SGS-UFSAR REVISION6 FEBRUARY15, 1987
GENERALEXAMPLE
~R 1 (11 COMPUTE"11R"AS THOUGH 1
1 THE LOAD"P" WAS ACTING 6R AT MIDSPAN. RL (2) COMPUTERL & RR, DIS* tRR REGARDTHE SMALLER 11 11 ONE OF THESEVALUES:
'p - X y L
(31 COMPUTEMAX.REACTION FOR ONE SIDE: Rmax= 11R + RL < T fsafeworkingloadofCONC. ANCH.) SGS-UFSAR REVISION6 FEBRUARY15,1987
DERIVATION r
------...... -...... :) ....... M L .I THE FOLLOWING EQUATION Will YIELD "M.. :
a*6-~ ADDITIONAL FORCE IN BOLT DUE TO PRYING ACTION: l.\R *.!, jd INTERNAL LEVER ARM -id* 0- ~d FOR "kc:J" SEE NEXT PAGE
SG$.-UFSAR 3.9A-!' REVISION8 FEBRUARY15,1987
ASSUMPTIONS FOR CONCRETEANCHORAGE: (1) BEHAVIORSAMEAS FOR WORKINGSTRESSMETHOD, (21 STEEL,PLATEIS RIGID.THISASSUMPTIONIS CONSERVATIVE FOR THE CALCULATION OF THE ROTATIONAL ANGLE"Ct.". 1 C = -f Bkd 2 c D D- kd E5 T =A n - - f wheren = - s kd c Ec T STRESSES 1 D- kd I:F =0 -+C = T -+-Bkd= A n - - 2 s kd E~ kd D- kd I:M= 0 -+M "'Cjd= !t Bkdjd 2 c M fc=-,-- jBkdjd STRAINS PL L kd
!::. = - =6--+.t::. = f - wherekd""'AFFECTEDCONCRETE AE E c c Ec DEPTHSUBJECTTO AXIALCOMPRESSION .!::. .. f h/2 whereh = DEPTHOF MAINMEMBER.
ss E s 2h = BOLTLENGTHCONSIDERED. SGS-UFSAR 3.91-10 REVISION6 FEBRUARY15, 1987
GENERALLYKNOWNEQUATIONS
2 M * -PL

_ _ _._::...:_:.
ZcEcZs_ _ __ 8 LZcEcZs+ 21E5 Z1 + 2ZcEc PL2 M*------ SL +16nl +.!!!. Zc Zs M PL2 b.R * - * ---~=----- jd slo-~)IL+2nl
\' 3 ~ +!!)
zc z*
SGS-UFSAR 3.9A-11 REVISION6 FEBRUARY15, 1987
Product: 1" diameter Richmond E.C. Type Insert with machine thread coil pulled from 15" x 18" x 6" concrete slab by means of 1" x 36" Anchor Stud bolt with nuts. The insert was made of
.4425 wire, and its setback in the concrete was 1/8'.The concrete slab was reinforced with a .442 wire mat, 6" x 6" center opening, located at mid-depth of the slab. The strength of the concrete was 2850 p.s.i., and the slip dial indicator was zeroed in at a load of 2000 lbs.
Failure occurred in both specimens by the insert pulling out of the concrete slab. Six cracks emanated from the insert on the tot) of the slab and extended down on four side surfaces to the reinforcement. The first crack appeared with a load of about 14000 lbs. on both specimens. Detail: Anchor Stud Insert Specimen No. 1 Specimen No. 2 Load, kips Slip, in. Load, kips Slip, in. 2 0 2 0 4 0.021 4 0.008 6 0.036 6 0.015 8 0.048 8 0.023 10 0.066 10 0.044 12 0.080 12 0.058 14 0.092 14 0.072 16 0.141 16 0.089 18 0.162 18 0.107 20 0.182 20 0.127 22 0.205 22 0.149 24 0.245 24 0.180 Ultimate Load = 25500 lbs. Ultimate Load = 24600 lbs. SGS-UFSAR 3.9A-12 REVISION6 FEBRUARY15, 1987
CONCRETEANCHOR bxdx~
I { YOUNG'SMODULUS-+ E1 MEMBER I MOMENTOF INERTIA-+I 1 I
;= =;= ~ *I L
TYPE"G"FRAMING CONCRETEANCHOR U/S CONCRETE TUBEb x d xt YOUNG'SMODULUS'""".Es { MOMENTOF INERTIA-+ I (. L ,., TYPE"E"OR "F"FRAMING FOR INFORMATION NOTSHOWNSEE DWG. 3068604073. SGS-UFSI\R REVISION6 J.9A-13 FEBRUARY15, 1987
INPUTPARAMETERS:

B ='i (in case of circular R. or round tube ~B =RADIUS) 0 =%(in case of circular R. or round tube 4 D =RADIUS)
A5 =CROSS-SECTIONAL AREAOF A. BOLTAT ROOT OF THREAD. L =SPANLENGTHOF MAINMEMBER. I =MOMENTOF INERTIAOF MAINMEMBER. t 6R r ADDITIONAL REACTION"6R"IN tR CONCRETEANCHORDUE TO PRYINGACTION:. D. L/2 L/2 iPB (kdl3 -~PBD (kd)2 L B(kd)2 + 2nA5(kd)- 2nA50 jd=D-~ 3 SOLUTION: SELECT"kd"SO THATM1 = M2;THENCOMPUTE"D.R". NOTETHATkd<DI SGS-UFSAR 3.9A-14 REVISION6 FEBRUARY15, 1987
GENERALEXAMPLE (11 COMPUTE"l>R"AS THOUGH THE LOAD"P"WAS ACTING AT MIDSPAN.
(2) COMPUTERL & RR, DIS* REGARDTHE SMALLER ONE OF THESEVALUES: RL
PLv (GOVERNS) X y L
{3) COMPUTEMAX.REACTIONFOR ONE SIDE: Rmax "'l>R+ RL < T (safeworking loadof CONC.ANCH.)
SGS-UFSAR 3.9A-15 REVISION8 FEBRUARY15, 1987
DERIVATION U/S CONCRETE r ( A1- ---------,:) M M I. L .I THE FOLLOWINGEQUATIONWILLYIELD"M": ADDITIONALFORCE IN BOLTDUE TO PRYINGACTION:' llR =~ . INTERNALLEVERARM -+jd= D- ~d FOR "kd"SEE NEXTPAGES SGS-UFSAR 3.9A-16 REVISION6 FEBRUARY15, 1987 *
.\SSUMPTIONS FOR CONCRETEANCHORAGE:
(11 BEHAVIORSAMEAS FOR WORKINGSTRESSMETHOD, (2) STEELPLATEIS RIGID. THISASSUMPTION CONSERVATIVE FOR THE CALCULATION OF IS THE ROTATIONAL ANGLE"IX'. C .. .!.F B kd 2 c c jd D I:MT""'0 -M .. Cjd.. ..!f 2 cc Bkdjd-+f .. __M 1 __ F 2Bfckdjd STRESSES PL L kd . t_.:z-,..6--/::i. * - f -With) AE E c Ec c kd"=o' ASSUMED CONC.DEPTH SUBJECTTO COMPRESSION. kd
!:Mc = 0-M*(T-Fljd=Afjd-Fjd-+-withF*f.
s$ 2 D D-k 1 kd "=o' ASSUMEDBOLTLENGTHSUBJECT STRAINS TO TENSION(TO BE THE SAMEAS CONC.DEPTHSUBJECTTO COMPRESSION).
!:F .. o-c .. T- F .....M.A jd s n~*~- F kd Bkdjd MB (kdl2 ... A5 n (D - kd)
2M - FB (kdl2jd SGS-UFSAR REVISION6 3.9J1.-17 FEBRUARY15, 1987
e
~ MB(kdt2 = 2nA5 MD- 2nA5 M (kdl- FB (kdt2(D -kd) 3 i= -i <t £ FBD (kd)2 - FB (kd)3 + MB(kd)2+ 2nAsM (kd)- 2nAsMD = 0 \ _ M [ B11<<11 2 + 2n A, lkdl - 2n A, D] * ~ FB (kd) 3 - FBD (kdl 2 TWO UNKNOWNIN ABOVEEQUATION-~oM & kd.
A SECONDEQUATIONWITHSAMEUNKNOWNVALUESFOLLOWS: ML tP = 2E I GENERALLYKNOWNEQUATIONS
-~s~ ~- 81jd p = LZcEcZs+ 2Es1Z5 +i ZcEclM ~6Z 5 E 5 1 2E5 1ZcEcZs (Z5 L2 - 81jd)P M=*-------
2 1LZ5 + 2nI s + 21)8 zc SGS-UFSAR REVISION 6 3.9J1.-l8 FEBRUARY15. 1987
BACK-SUBSTITUTING (A DL2 s kd VALUESZc&Z1:
J!!. -81jd)P (A5DL2 -81kd) Pjd M= A (LA5Djd + 4nl Bs + 21kd)8 M AR
jd SOLUTIONBY ITERATION:
(1) ASSUMEA VALUEFOR "kd"AND COMPUTE..M., (SHEET#7) EQUATIONQ) (2) WITHASSUMED"kd.. ALSOCOMPUTE"M"FROM EQUATION~ (3) IF M-VALUESFROMSTEPS1 & 2 ARE NOTTHE SAMEADJUST "kd"AND STARTALLOVER AGAINWITHSTEPS1 & 2 (4) COMPUTE"AR"IF M*VALUESFROMSTEPS1 & 2 ARE EQUAL
SGS-UFSAR 3.9A-19 REVISION8 FEBRUARY15, 1987
3.10 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT This section presents information to demonstrate that instrumentation and electrical equipment classified as seismic Category I is capable of performing safety-related functions in the event of an earthquake. Seismic Category I instrumentation and electrical equipment is identified in Table 3. 10-1. Electrical equipment that is not designed to seismic Category I criteria and whose structural failure could affect the operation of seismic Category I equipment is located, physically restrained, or structurally designed, such that its postulated structural failure during seismic conditions does not prevent Category I electrical equipment from performing its safety function. 3.10.1 Seismic Qualification Criteria 3.10.1.1 Qualification Standards The methods of meeting the general requirements for seismic qualification of Category I instruments and electrical equipment as described by General Design Criteria (GDC) 1, 2, and 23 are described in Section 3.1. The general methods of implementing the requirements of Appendix B to 10CFR Part 50 are described in Section 17. The seismic Category I instrumentation and electrical equipment and their supports as listed in Table 3.10-1 are qualified in accordance with the methods described in IEEE Standard 344-1971. The seismic acceleration levels used in the seismic qualification tests and analyses are selected to envelope the plant specific levels defined in Section 3.7 . 3.10-1 SGS-UFSAR Revision 6 February 15, 1987
3.10.1.2 Acceptance Criteria Seismic qualification must demonstrate that Category I instrumentation and electrical equipment is capable of performing designated safety-related functions during and after an earthquake of magnitude up to and including the Safe Shutdown Earthquake (SSE). The qualification must also demonstrate the structural integrity of mechanical supports and structures at the Operating Basis Earthquake (OBE) level. Some permanent mechanical deformation of supports and structures is acceptable at the SSE level provided that the ability to perform the designated safety-related functions is not impaired. 3.10.2 Methods and Procedures for Qualifying Electrical Equipment The seismic qualification of Seismic Category I instrumentation and electrical equipment is demonstrated by testing, analysis, or a combination of these methods in accordance with IEEE Standard 344-1971. The choice of qualification method employed for a particular item of equipment is based upon many factors including: practicability, complexity of equipment, function of equipment, and availability of previous seismic qualification. The qualification method employed for a particular item of equipment is identified in Table 3.10-1. 3.10.2.1 Seismic Qualification by TyPe Test From 1969 to mid-1974, Westinghouse seismic test procedures employed single axis sine beat inputs in accordance with IEEE Standard 344-1971 to seismically qualify equipment. The input form selected by Westinghouse was chosen following an investigation of building responses to seismic events (1). In addition, Westinghouse has conducted seismic retesting of certain items of equipment as part of the Supplemental Qualification Program (2). This retesting was performed at the request of the NRC Staff on agreed selected items of equipment employing 3.10-2 SGS-UFSAR Revision 6 February 15, 1987
multi-frequency, multi-axis test inputs (3) to demonstrate the conservatism of the original sine beat test method with respect to the modified methods of testing for complex equipment recommended by IEEE Standard 344-1975. The original single axis sine beat testing (4) and the additional retesting completed under the Supplemental Test Program has been the subject of generic review by the NRC Staff. Balance-of-plant (BOP) equipment was tested in accordance with IEEE Standard 344-1971. 3.10.2.2 Seismic Qualification by Analysis Employing motors as an example, the structural integrity of safety-related motors is demonstrated by a static seismic analysis in accordance with IEEE Standard 344-1971. Motor operability during a seismic event is demonstrated by calculating critical deflections, loads, and stresses under various combinations of seismic, gravitational, and operational loads. The worst case (maximum) calculated values are tabulated against the allowable values. On combining these stresses, the most unfavorable possibilities are considered in the following areas: 1) maximum rotor deflection~ 2) maximum shaft stresses, 3) maximum bearing load and shaft slope at the bearings, 4) maximum stresses in the stator core to frame welds, 5) maximum stresses in the motor mounting bolts and, 6) maximum stresses in the motor feet. The analytical models employed and the results of the analysis are described in the qualification references. 3.10.2.3 Seismic Qualification by a Combination of Type Test and Analysis A combination of test and analysis is employed in the qualification of equipment such as a cabinet that may house several different configurations of devices. This method is also used for multiple joined equipment. A test is performed on the 3.10-3 SGS-UFSAR Revision 6 February 15, 1987
device supporting structure (e.g., cabinet) and an analysis model is then developed to perform a structural evaluation of different configurations. The test is utilized in the development and refinement of the analytical model and provides a verification of the analytical results. When needed, tests are performed on devices with applicable device seismic environments determined by the analytical model. 3.10.3 Methods and Procedures for Qualifying Supports of Instrumentation and Electrical Equipment The seismic qualification of the supports for Category I instrumentation and electrical equipment is demonstrated by testing, analysis, or a combination of these methods. The preferred method of qualification for these supports is to test the support with the equipment as described above. When testing is not practical, qualification of supports is by static and/or dynamic analysis procedures; the possible amplified design loads for vendor supplied equipment are considered as follows:
1. The support is tested with the actual components mounted or with the component loads simulated.

2. Analysis of the support includes the component loads.
3.10.4 Results of Tests and Analyses The results of the seismic tests and analyses that ensure the criteria established in Section 3.10.1 have been satisfied, employing the qualification methods described in Sections 3.10 .1 and 3.10.3, are provided in individual seismic qualification reports. These reports are referenced in Table 3.10-1. 3.10.5 References for Section 3.10
1. Morrone, A. , "Seismic Vibration Testing with Sine Beats,"
WCAP-7558, October 1971. 3.10-4 SGS-UFSAR Revision 6 February 15, 1987
2. NS-CE-692, letter from C. Eicheldinger (Westinghouse) to D.
B. Vasello (NRC), July 10, 1975 .
3. Jarecki, S. J., uGeneral Method of Developing Multi-frequency Biaxial Test Inputs for Bistables, n WCAP-8624 (Proprietary),
. September 1975 and WCAP-8695 (Non-Proprietary), September 1975.
4. Vogeding, E. L., et al, "Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)." WCAP-7397-L (Proprietary) and WCAP-7817 (Non-Proprietary), December 1971, plus Supplements 1-8 .
SGS-UFSAR 3.10-5 Revision 6 February 15, 1987
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATIONS FOR SAFETY-RELATED EQUIPMENT Equipment Method Results 1 Control Console Test & Dynamic Simultaneous Time History Test Analysis producing accelerations greater than design basis earthquake (DBE). Test results were acceptable. Accelerations at the device location were determined by dynamic analysis. Nuclear Instrumenta- Test Sine Beat Test with electrical functions of the tion System Cabinet; Single Axis equipment monitored. Process Control Equipment Cabinets; Solid State Protection Actuation Cabinet 12 KVA (Unit 1) Triaxial Single axis Sine Sweep Resonance search test in 10 KVA (Unit 2) Vital Multifrequency front-to-back, side-to-side, and vertical axes Bus UPS Random Motion followed by 30 second duration Triaxial Test Multifrequency Random Motion test. The specimen was subjected to 5 Operating Basis Earthquake (OBE) tests and one Design Basis Earthquake (DBE) test. Auxiliary Control Test Single axis Sine Sweep Resonance search test in System Terminal and Single Axis front-to-back, side-to-side, and vertical axes Relay Cabinets followed by 30 second duration Triaxial Multifrequency Random Motion test. The equipment functioned satisfactorily. 1 of 9 SGS-UFSAR Revision 32 June 17, 2021
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATIONS FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Relay Racks Dynamic Analysis and The finite element dynamic Test Multiple Axis analysis shows that the stress in the racks are within acceptable limits. The acceleration at different locations of the rack were also determined by analysis. The relays were qalified by test to a level higher than the acceleration level of the rack. 1 125V and 28V de Test Simultaneous Time History Test was Distribution Biaxial performed. The distribution cabinet Cabinets components functioned properly. Terminal Cabinets Static Analysis Structural integrity of the cabinets was Multiple Axis justified. Diesel Control Test Simultaneous Time History Test1
It was Cabinets Biaxial demonstrated that the specimens possessed sufficient integrity to withstand, without compromise of structures or electrical function, the prescribed simulated seismic environments.

Control Room Recorder Dynamic Analysis A finite element computer analysis Panels Multiple Axis was performed. It was demonstrated that the panel possessed sufficient integrity to withstand, without compromise of structural integrity, the prescribed simulated seismic environment.
2 of 9 SGS-UFSAR Revision 13 June 12, 1994
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATIONS FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Solid State Protection Test Simultaneous Time History Test 1 Output Test and Biaxial with the functionality monitored. Interface Cabinets Radiation Monitoring Test Random multifrequency test with Cabinets and Components Biaxial functionality monitored. Safeguards Equipment Test It is demonstrated, thru a Control Single Axis combination of Analysis Dynamic Analysis and Dynamic Testing Multiple Axis of similar equipment, that the equipment will perform its function under a design basis earthquake event . Replacement CEU's .Dynamic Analysis Each of the three existing Control Multiple Frequency Test Electronics Units (CEU's) in the Multiple Axis SEC's are replaced with a new CEU, a new test panel, and various other more minor modifications (switches, pushbuttons, etc.). Analysis of the reconfigured cabinet(s) confirmed seismic structural integrity, anchorage integrity, and II/I interaction acceptability; and produced Required Response Spectrum (RRS) for the CEU location in the cabinet(s). Seismic testing, employing simultaneous three axis,* multi-frequency random motion, was acceptably completed in conformance with the requirements of IEEE-344. 3 of 9 SGS-UFSAR Revision 13 June 12, 1994
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATION FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Control Centers Test The MCC was tested in the Single Axis three orthogonal directions and the transmissibility was determined. The individual components were tested separately. Unit Substations and de Test The unit was shock tested at Switchgear Single Axis accelerations greater than the design basis earthquake (DBE). Test results are acceptable. 5 kV Switchgear Static Analysis It is justified thru analysis that Multiple Axis the equipment maintains its structural integrity when subjected to Design Basis Earthquake (DBE) event. Diesel Generator and Static Analysis It is justified thru analysis that Accessories Single Axis the equipment maintains its structural integrity when subjected to Design Basis Earthquake {DBE) event. Tray and Hangers Dynamic Analysis It is justified thru analysis that Multiple Axis the equipment maintains its structural integrity when subjected to Design Basis Earthquake (DBE) event. 4 .of 9 SGS-UFSAR Revision 13 June 12, 1994
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATION FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Battery Charger 125V Test A resonant search was performed. Biaxial The unit was then subjected to a resultant acceleration greater than the Design Basis Earthquake (DBE) at the determined natural frequencies. Test results were acceptable. Battery Charger 28V Test Resonance search from 1-35Hz Single Axis and a random multi-frequency test with functionality monitored. The test results were acceptable. Batteries 28V & 125V .Test Random multifrequency testing Biaxial for 5 OBE and 1 DBE with functionality monitored. Test results were acceptable. Battery Racks Static Analysis Qualified on the basis of dynamic Dynamic Test testing and testing and static analysis. Electrical Penetrations Static Analysis Stresses were determined using Multiple Axis maximum g level obtained from response spectra. Results of the analysis were acceptable. 5 of 9 SGS-UFSAR Revision 13 June 12, 1994
TABLE 3.10-1 FOR SAFETY-RELATED EQUIPMENT Aux. Bldg. Vent. Fan Static Analysis The analysis substantiates that Motors the motor will function both mE~crlal1l'ca~~ and under DBE conditions. Service Water Pump Static Analysis The analysis substantiates that Motors the motor will function both mechanically and electrically under DBE conditions. Aux. Feedwater Pump Static Analysis The substantiates that Motors the motor will function both mE~crlalll,Ca.LL and under DBE conditions. Engineered Safeguards Analysis Pump motors supplied by Westinghouse with seismic Motors 6 of 9 SGS-UFSAR Revision 25 October 26, 2010
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATION FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Radiation Monitoring Test Subjected to a biaxial random Transformers Bi-axial multifrequency test with functionality monitored during and after the test. Radiation Monitoring Test The unit was vibrated in system circuit Breaker Single Axis all of the three orthogonal Panel boards Single Frequency axes and the functionality monitored. Pressure Transmitters, Test RMF tests were performed in Differential, Absolute, Multiple Axis three mutually perpendicular axes. and Gauge During testing the transmitters maintained structural integrity and met functional requirements. The tests establish the functional adequacy of the transmitters to a certain input acceleration level. Analysis of the supporting structure ensures that the actual seismic loading is within the qualified level. 7 of 9 SGS-UFSAR Revision 14 December 29, 1995
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATION FOR SAFETY*RELATED EQUIPMENT Equipment Method Results Limit Switches Test Seismic tests performed on Single Axis limit switches qualify the switches for a certain acceleration limit. Analysis of the supporting structure ensures that the postulated seismic loading is within the qualified level. ASCO Solenoid Test The seismic test establish Valves Single Axis the valve adequacy to a certain input level. Analysis of the supporting structures ensures that the actual seismic loading is within the qualified levels. Instrument Panels, Test Some of the panels are Cabinets, NEMA 12 Dynamic Analysis qualified by a finite elem~nt Enclosures, Racks Multiple Axis analysis. Some panels were tested, with devices mounted on them. 8 of 9 SGS-UFSAR Revision 13 June 12, 1994
TABLE 3.10-1
SUMMARY
OF SEISMIC QUALIFICATION FOR SAFETY-RELATED EQUIPMENT Equipment Method Results Pressure Switches Test Tests were performed in three Multiple Axis mutually perpendicular axes. The switches met their structural and functional requirements. Instrumentation Test Tests were performed in three Axis mutually axes. I - Controllers The equipment met their respective - Thermostats structural and functional E/P Converters Power - Indicating Stations - Terminal Blocks Stations - Timers - Pneumatic Controllers - Switches Indicators Pressure 1 All time history tests were performed using the time history response spectra of floor Elevation 122' in the Auxiliary Building. 9 of 9 SGS-UFSAR Revision 25 October 26, 2010
3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT The electrical portions of the Engineered Safety Features and the Reactor Protection Systems are designed to remain functional in all abnormal environments anticipated under normal, test, and design basis accident conditions. This section presents information on the design basis and qualification verifications for mechanical and electrical equipment for these systems. Section 3,7 presents the seismic design requirements and Section 3.10 presents the seismic qualification of electrical equipment. On May 23, 1980, the NRC Commissioners issued Memorandum and Order CLI-80-21 which stated that the DOR guidelines and NUREG-0588 set the requirements that Licensees and Applicants must meet regarding the environmental qualif~cation of safety-related electrical equipment to satisfy 10CFRSO, Appendix A, General Design Criteria (GDC) 4. This evaluation was conducted and the required information is detailed in the docketed "Salem Generating station, Environmental Qualification Review Report," Volumes 1 and 2, transmitted to the NRC .on December 12, 1980, and in subsequent revisions. This information was reviewed by the NRC staff and their consultant, the Franklin Research Institute. Safety Evaluation Reports {SERs) for both units were issued initially in mid-1981 and in January 1993. Public Service Electric & Gas (PSE&G) has responded, as required, to the nopen Itemsn in these SERs. 3.11.1 Equipment Identification and Environmental Conditions 3.11.1.1 Equipment Identification Class lE equipment that is located in a harsh environment and must operate to mitigate the effects of an accident and maintain the plant in a safe condition are identified in the SAP database .
SGS-UFSAR 3.11-1 Revision 21 December 6, 2004
3.11.1.2 Plant environments in various plant zones for an array of accident conditions are also contained in the 11 Salem Generating Station, Environmental Design Criteria, document S-C~ZZ-SCD-1419. 3.11.1.3 Normal Operating Environment Temperature in the control room and adjoining equipment room is maintained for personal comfort at 70°F + 15°. Protective equipment in this space is designed to operate within design tolerance over this temperature range. Design specifications for this equipment specify no loss of protective function over the temperature range from 40°F to ll0°F. Thus there is a wide margin between design limits and the normal operating environment for control room equipment . Within containment, the normal operating temperature for protective equipment, except out-of-core neutron detectors and equipment inside the Pressurizer Enclosure, will be maintained below l20°F. Protective instrumentation is designed for continuous operation within design tolerance in this environment. Out-of-core neutron detectors are designed for continuous operation at l35°F, and the normal operating temperature will be maintained below this value. The detectors will withstand operation at l75°F for short durations (8 hours). Process instrumentation in containment which is vital to plant protection is designed to survive the post-accident environment long enough to perform the required protective function. Pressurizer Enclosure equipment has been evaluated and qualified to higher temperatures. SGS-UFSAR 3.11-2 Revision 21 December 6, 2004
Qualification testing has been performed on various safety systems such as Process Instrumentation, Nuclear Instrumentation, and Relay Racks. This testing involved demonstrating operation of safety functions at elevated ambient temperatures to 120°F for control room equipment and in full post-accident environment for required equipment in containment. Detailed results of some of these tests are proprietary to the suppliers, but are on file at the suppliers and available for audit by qualified parties.
The initial qualification test of individual components, an4 the integrated tests of the systems as a whole, complement each other to assure performance of the system as designed to prove proper operation of the actuation circuitry. For engineered safety features inside the containment, qualification testing and/or analysis is performed under the effects of the conservative post accident environmental parameters where applicable. 3.11.2 Qualification Tests and Analyses For Class lE equipment, Salem Station meets the Institute of Electrical and Electronics Engineers (IEEE) Standard 323-1971, "IEEE Standard for Qualifying lE Equipment for Nuclear Power Generating Stations." Containment fan cooler motors have been tested according to IEEE Standard 334-1971, "Trail use Guide For Type Tests of Continuous Duty Class 1 Motors Installed Inside Containment of Nuclear Power Generating Stations." Comprehensive testing and/or analysis is conducted for that .Class lE electrical equipment and components which are required to function during and subsequent to any of the design basis accidents. The program consists of performing tests of individual pieces of equipment in the manufacturer's shop, integrated tests of the system as a whole in the field, and periodic inspection and tests of the activation circuitry and mechanical components to assure reliable performance upon demand, throughout the 'plant lifetime . 3.11-3 SGS-UFSAR Revision 6 February 15, 1987
3.11.3 Qualification Test Results The results of qualification tests for all safety-related equipment that is required to operate in a harsh environment and mitigate the effects of an .;tccident are containedin the Sa,lem Equipment Qualification Binders. 3.11.4 Loss of ventilation The control room, equipment rooms, relay room, and switchgear rooms will not experience a harsh environment caused by a high energy break analysis, main steam line break, loss-of-coolant accident, or recirculated fluids. These areas are E,lerved by Class lE redundant ventilation equipment which are also not subject to harsh environments and are located in benign areas for radiation doses following an accident. 3 .11-4 SGS-UFSAR Revision 21 December 6, 2004
3.12 CONFORMANCE TO RULES ISSUED AFTER PLANT LICENSING 3.12.1 NRC Rule On Station Blackout On July 21, 1988, the Code of Federal Regulations, Title 10, Part 50 was amended to include a new Section 50.63, "Loss of All Alternating current Power, " (Station Blackout)
The Station Blackout (SBO) rule requires that each light-water cooled nuclear power plant licensed to operate must be able to withstand and recover from an SBO. An sao is defined in 10CFR50.2 as the complete loss of alternating current (AC) electric power to the essential and non-essential switchgear busses (i.e., loss of offsite power concurrent with a turbine trip and unavailability of the onsite emergency ac power system). SBO does not include loss of station batteries or loss of AC power from station batteries through inverters, nor does it assume a concurrent single failure or design basis accident of the affected Unit.
The NRC issued Regulator Guide (RG) 1.155 in August of 1988, to provide the industry with guidance that was acceptable for meeting the requirement of 10CFR50.63. In RG 1.155, the NRC states that NUMARC 87-00 (Reference 1) also provides guidance acceptable for meeting the requirements of 10CFRS0.63, except when RG 1.155 takes precedence over NUMARC 87-00 as indicated in Table 1 of RG 1.155. 3.12.1.1 Conformance to NRC Rule on Station Blackout An SBO coping analysis was performed to determine SGS's coping duration and ability to cope with an sao. This coping duration was based on:
a. Offsite Power Design Characteristic

b. Emergency AC Power Supply System Configuration

c. Calculated EDG Reliability; and

d. Allowed EDG Target Reliability The coping duration for SGS was calculated as four hours in accordance with NUMARC 87-00, Section 3.0 with the exception of the frequency of Loss of Offsite Power events due to severe weather (SW) and Extremely Severe Weather (ESW). Site-specific weather data was used to determine the SW and ESW frequency as detailed in Report No. NUS-5175, Rev. 1 (Reference 2).
3.12-1 SGS-Uli'SAR Revision 15 June 12, 1996
The ability to cope with an SBO event is based on the ability to maintain "appropriate containment integrity" (as defined in RG 1.155), provide adequate condensate inventory for decay heat removal, provide adequate class lE battery capacity and compressed air capacity for the duration and evaluate equipment operability due to loss of ventilation. The ability to cope with an SBO event is described in programmatic standard SC. DE-PS. ZZ-0040 (Q), "Salem Station Blackout Program." The Station Blackout Program was reviewed to determine the impact of the Unit 2 Model 61/19T steam generators. The aspects of the SBO determined to be potentially affected by the replacement steam generator design are those that relate to the RCS inventory and condensate inventory. The RCS inventory increases with the Model 61/19T steam generators, which has a beneficial impact in that with the assumed RCP there is a longer period of time to core uncovery. The Unit 2 steam generators have a amount of Sensible heat than the original Series 51 steam generators, but the condensate to cope with a SBO will decrease with the RSGs. The review concluded that the AFST has adequate capacity. Therefore, there is no adverse impact on the SBO Program (Reference 3). 3.12.2 References
1. NUMARC 87-00, "Guidelines and Technical Bases for Initiatives Station Blackout at Light Water Reactors," Rev. 1, 1991.

2. Halliburton NUS Environmental Corporation, NUS-517 5, Rev. 1, "Estimated Frequency of Loss of Off-Site Power due to Extremely Severe Weather (ESW) and Severe Weather (SW) for Salem and Hope Creek Generating Stations",
March 1992.
3. PSEG VTD 328299, Areva NP Document No. 51-5049368-04, Salem 2 NSSS/BOP Review.
3.12-2 SGS-UFSAR Revision 24 May 11, 2009
APPENDIX 3A PSE&G POSITIONS ON USNRC REGULATORY GUIDES SGS-UFSAR Revision 6 February 15, 1987
APPENDIX 3A PSE&G POSITIONS ON USNRC REGULATORY GUIDES (*) Regulatory Guide 1.1 -NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS Regulatory Guide 1. 1 requires that the Emergency Core Cooling and Containment Heat Removal Systems be designed so that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss-of-coolant accidents (LOCAs). The Westinghouse design of the Emergency Core Cooling System and Containment Spray System provides adequate NPSH to all system pumps. The NPSH for all of the pumps is evaluated for both the injection and recirculation modes of operation following a LOCA, except for the containment spray pumps, which are only used during the injection mode (see Section 6. 2) The evaluation has shown that the end of the injection mode of operation gives the limiting NPSH available for the centrifugal charging and safety injection pumps. At the end of the injection mode, the suction for these pumps is being provided from the refueling water storage tank. The NPSH available at this time is determined from the elevation head and vapor pressure of the water in the refueling water storage tank, which is at atmospheric pressure, and the pressure drop in the suction piping from the tank to the pumps. The NPSH evaluation is based upon all pumps operating at the design flow rates. The recirculation mode of operation gives the limiting NPSH requirement for the residual heat removal pumps, and the minimum NPSH available is determined from the following calculation: NPSH =(h) -(h) + (h) - (h) available containment vapor static loss pressure pressure head
The original nomenclature "Safety Guide" was changed to "Regulatory Guide" in December 1972.
3A-1 SGS-UFSAR Revision 24 May 11, 2009
The containment pressure value will be equal to the initial air pressure in containment prior to the LOCA (i.e., the pre-accident partial air pressure in containment). However, when the containment sump vapor pressure exceeds the containment initial pressure then the following is assumed: (h) . =(h) contalnment vapor pressure pressure The containment air pressure value used in the NPSHa calculation is based on the containment conditions prior to the accident only and does not include any credit for accident pressure conditions is conservatively determined based on minimum containment initial pressure, and maximum temperature and relative humidity conditions. The calculation also accounts for further reduction of this initial air pressure based on possible maximum cooldown of the containment environment post-LOCA. The vapor pressure term used in the NPSHa for the sump water being pumped is based on the highest temperature of the sump fluid for the condition being evaluated. The static head term in the NPSHa is calculated using the minimum available water inventory in the containment for recirculation operations. This minimum water inventory ensures that the containment sump strainers are fully submerged prior to initiation of recirculation phase. It is believed that the methods utilized in calculating NPSH meet the intent of the Regulatory Guide, of ensuring adequate NPSH with adequate margin for the centrifugal charging, safety injection, residual heat removal, and containment spray pumps. Regulatory Guide 1.2 - THERMAL SHOCK TO REACTOR PRESSURE VESSELS Although NRC Regulatory Guide 1.2 was withdrawn by the NRC on July 31, 1991, SGS commitments, as stated below, are not affected by this withdrawal. Current Westinghouse research programs and pressure vessel design conform with the intent of the Regulatory Guide. 3A-2 SGS-UFSAR Revision 24 May 11, 2009
Westinghouse is continuing to obtain fracture toughness data through participation in the HSST Program at the Oak Ridge National Laboratory. The fracture toughness data recently obtained include tests on un-irradiated material using specimens up to 12 inches thick. In addition, new testing techniques have evolved which allow the measurement of valid fracture toughness data with much smaller specimens than have been used in the past. These un-irradiated data correspond to startup or beginning-of-life of a plant. Post-irradiation data were obtained from 2-inch thick specimens in 1970. Post-irradiation data on 4-inch thick fracture mechanics specimens are available from the HSST Program in 1974. Elastic plastic test procedure should greatly simplify the problem of obtaining irradiated fracture toughness data because of the associated reduction in required specimen size. Westinghouse is also engaged in an extensive materials irradiation surveillance program from which irradiated fracture toughness data are obtained for actual vessel material. The present data were used in a rigorous linear-elastic fracture mechanics analysis of the reactor vessel thermal shock problem. The results of this analysis show that under the postulated accident conditions, the integrity of the reactor vessel would be maintained throughout the life of a plant. Westinghouse's continuing participation in the HSST Program will yield confirmatory in formation of material properties and fracture mechanics analytical methods. If additional margin against brittle fracture is required, or if the remaining data from the HSST Program does not confirm the present analysis, the reactor vessel can be annealed at any point in its service life. Westinghouse is currently engaged in a research program to determine the optimum annealing time and temperature. No hardware for vessel annealing has yet been designed, but appropriately designed space heaters could be utilized as one conceivable method of annealing. The design of Westinghouse reactor vessels does not preclude post-irradiation heat treatment. 3A-3 SGS-UFSAR Revision 24 May 11, 2009
Regulatory Guide 1.3 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1.4 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS The Salem Station complies with Regulatory Guide 1.183 instead. Regulatory Guide 1 . 5 - _A_S_S_U_M_P_T_I_O_N_S_ _U_S_E_D___F_O_R___E_V_A_L_U_A_T_I_N_G___T_H_E___P_O_T_E_N_T_I_A_L RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1. 6 - INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS It is believed that the Salem Station design conforms with the intent of the Regulatory Guide. Regulatory Guide 1. 7 - CONTROL OF COMBUSTIBLE GAS CONCENTRATION INCONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT The Salem Station design conforms to the intent of the Regulatory Guide as described in Section 6.2. 3A-4 SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.8 - QUALIFICATION AND TRAINING OF PERSONNEL FOR NUCLEAR POWER PLANTS, revision 2, April 1987 Salem complies with Regulatory Guide 1.8, except as noted below. The Operations Director shall either hold an SRO license or have held an SRO license for a similar unit (PWR) or have been certified at an appropriate simulator for equipment senior operator knowledge. Licensed Operator qualifications and training shall be in accordance with 10CFR55. The Radiation Protection Manager shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The Director-Nuclear Oversight (NOS), and the Engineering Manager positions under the Site Engineering Director, which correspond to the Engineer in Charge, must meet or exceed the qualifications of ANSI/ANS 3.1-1981. Qualification requirements for the Nuclear Safety Review Board personnel performing the offsite independent review function and PORC members are described in their associated program documents. See Section 13 for further discussion of staffing of plant personnel. Regulatory Guide 1.9 - SELECTION, DESIGN, AND QUALIFICATION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES The Salem Station design conforms to the intent of the Regulatory Guide as indicated in Section 8. Regulatory Guide 1.10 - MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES Although NRC Regulatory Guide 1.10 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station design conforms to the intent of the Regulatory Guide as described in Section 3.8. Regulatory Guide 1.11 - INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT The Salem Station design conforms with the intent of Regulatory Guide 1.11 and General Design Criterion (GDC) 55 for instrument lines. Both containment pressure and RVLIS isolation inside containment is provided by a sealed bellows arrangement. The containment pressure bellows are located immediately adjacent to the inside containment wall. The RVLIS bellows are located near the process pressure sources at the RCS hot legs, in the seal table room and in the reactor cavity. Outside containment isolation for containment pressure is provided by the diaphragm in the pressure transmitter connected to the bellows by a sealed, fluid filled tube. Outside containment isolation for RVLIS is provided by sealed, fluid filled isolators that convey RCS pressure to DP transmitters. The justification for these special arrangements results from the importance of containment pressure and RVLIS indication during accident conditions. 3A-5 SGS-UFSAR Revision 29 January 30, 2017
Regulatory Guide 1.12 - INSTRUMENTATION FOR EARTHQUAKES The Salem Station design conforms with the intent of the Regulatory Guide as described in Section 3.7. Regulatory Guide 1.13 - SPENT FUEL STORAGE FACILITY DESIGN BASIS The Spent Fuel Cooling System design conforms with the intent of the Regulatory Guide as discussed in Section 9. 1. The design of the Fuel Handling System conforms to the recommendations of Regulatory Guide 1.13. Regulatory Guide 1.14 - REACTOR COOLANT PUMP FLYWHEEL INTEGRITY The Salem Station design conforms with the intent of the Regulatory Guide. Regulatory Guide 1.15 -TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES Although NRC Regulatory Guide 1.15 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station design generally conforms to the intent of the Regulatory Guide as discussed in Section 3. 8. However, instead of one full diameter specimen from each bar size tested for each 50 tons, or fraction thereof, of rebar produced from each heat as required by the Regulatory Guide, two specimens for each 25 tons or less of heat have been taken for testing. If any of the four specimens failed to meet the specification, the entire heat was rejected. It is believed that this procedure is as conservative as that of the Regulatory Guide. Regulatory Guide 1.16- REPORTING OF OPERATING INFORMATION Information will be reported as indicated in the Regulatory Guide, with the exception of the information provided in the Monthly Operating Report. The Monthly Operating Report information will be reported as indicated in Generic Letter 97-02. 3A-6 SGS-UFSAR Revision 17 October 16, 1998
Regulatory Guide 1.17 - PROTECTION AGAINST INDUSTRIAL SABOTAGE, 6/73 (endorses N18.17) Although NRC Regulatory Guide 1.17 was withdrawn by the NRC on July 5, 1991, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station security plan will conform with the intent of the Guide. Regulatory Guide 1.18 - STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS Although NRC Regulatory Guide 1.18 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station containment structural acceptance test will conform with the intent of the Regulatory Guide as described in Section 6.2. Regulatory Guide 1.19 -NONDESTRUCTIVE EXAMINATION OF PRIMARY CONTAINMENT LINER WELDS Although NRC Regulatory Guide 1.19 was withdrawn by the NRC on July 21, 1981, SGS commitments as stated below, are not affected by this withdrawal. The Salem Station containment liner examinations conform with the intent of the Regulatory Guide. Regulatory Guide 1.20 -VIBRATION MEASUREMENTS ON REACTOR INTERNALS Westinghouse will comply with the requirements of the Regulatory Guide. If, for some overriding reason, deviations from this guide are followed, non-compliance will be justified. 3A-7 SGS-UFSAR Revision 13 June 12, 1994
For each prototype reactor internals design, a program of vibration analysis, measurement, and inspection will be developed and reviewed by the NRC prior to the performance of the scheduled preoperational functional test. Westinghouse has prepared the vibrational analysis and test programs for prototype 2-, 3-, and 4-loop plants. Regulatory Guide 1.21 Revision 2 - MEASURING AND REPORTING OF EFFLUENTS FROM NUCLEAR PLANTS Salem Stations conformance to RG 1.21 is controlled via the Salem Offsite Dose Calculation Manual (ODCM) and its implementing procedures. Routine reporting of effluents is controlled by the ODCM. Emergency effluent reporting is controlled by the Emergency Plan. Regulatory Guide 1.22 - PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS The Salem Station Protection System is designed in accordance with IEEE Standard 279-1971. Safety actuation circuitry is provided with a capability for testing with the reactor at power. The protection system design complies with the Regulatory Guide. Under the present design, there are protection functions which are not tested at power. These are described in Section 7. Additionally, the following manual functions are not tested at power:
1. Generation of a reactor trip by tripping the reactor coolant pump breakers

2. Generation of reactor trip by tripping the turbine 3A-8 SGS-UFSAR Revision 32 June 17, 2021

3. Generation of a reactor trip by use of the manual trip switch

4. Generation of a reactor trip by manually actuating the Safety Injection System

5. Generation of a safety injection signal by use of the manual safety injection switch

6. Generation of a containment spray signal by use of the manual spray actuation switch Exception is taken to testing the devices listed above, as allowed by the Regulatory Guide, where it has been determined that:

1. "There is no practicable system design that would permit operation of the equipment without adversely affecting the safety or operability of the plant."
The present position is that it is not a "practicable system design" to provide equipment to bypass a device such as a reactor coolant pump breaker or a MSIV solely to test the device. In the case of manual initiation switches, the design for test capability would require that switches be provided on a train or sequential basis. This increases the operator action required to manually actuate the function.
2. "The probability that the protection system will fail to initiate the operation of the actuated equipment is, and can be maintained, acceptably low without testing the actuated equipment during reactor operation."
The probability of failure of the above devices is considered to be very low. 3A-9 SGS-UFSAR Revision 16 January 31, 1998
3. "The actuated equipment can routinely be tested when the reactor is shut down."
In all the cases discussed above, it is only the device function which is not tested. The logic associated with the devices has the capability for testing at power. Regulatory Guide 1.23 Revision 1 - ONSITE METEOROLOGICAL PROGRAMS The Salem Station meteorological program will conform with the intent of the Regulatory Guide. Regulatory Guide 1.24 - ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PWR RADIOACTIVE GAS STORAGE TANK FAILURE The assumptions used are in agreement with the Regulatory Guide, as described in Section 15. Regulatory Guide 1.26 - QUALITY GROUP CLASSIFICATIONS AND STANDARDS The Salem Station design meets the intent of the Regulatory Guide in that appropriate quality levels have been assigned to components, structures, and systems relative to the importance of their safety functions. Systems, structures, and components were designed in accordance with the codes and standards that were in effect at the time of design and at the equipment order
dates, 3A-10 SGS-UFSAR Revision 32 June 17, 2021
which were prior to the inception of the NRC's quality group classification system. The Regulatory Guide was not issued until March 1972, at which time construction was well underway. The codes and standards which were used are presented in the appropriate sections of the FSAR. Regulatory Guide 1.27 -ULTIMATE HEAT SINK (Revision 2) The Salem Station design generally conforms with the intent of the Regulatory Guide (FSAR Section 9.2). Regulatory Guide 1.28 - QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION) Salem Generating Station is committed to the requirements of NQA-1-1994 for Quality Assurance Program requirements. Regulatory Guide 1.29 - SEISMIC DESIGN CLASSIFICATION, 8/73 The Salem Station design conforms to the intent of the Regulatory Guide. Previously, the only area of non-conformance with the Regulatory Guide was in the classification of the Spent Fuel Pool Cooling (SFPC) System. SFPC piping and pipe supports are analyzed as seismic class I. SFPC components have been seismically evaluated under SQUG GIP methodology. The basis for this classification is provided in Section 9.1. Regulatory Guide 1.30 - QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING INSTRUMENTATION AND ELECTRIC EQUIPMENT, 8/72 (endorses N45.2.4) The Salem Station design conforms with the intent of the Regulatory Guide. 3A-11 SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.31 - CONTROL OF FERRITE CONTENT IN STAINLESS STEEL WELD METAL PRIOR TO REVISION 4 The Regulatory Guide states that weld deposits should contain between 5 and 12 to 15 percent delta ferrite. It is not practical to specify "absolute minimum" or even maximum delta ferrite limits as a basis for acceptance or rejection of otherwise acceptable austenitic stainless steel welds. Westinghouse places control on the actual wire analysis for inert gas welding processes and on the final weld deposit for the fluxing weld process. In the case of the bare wire, when used with inert gas processes, although the wire may contain 5 percent ferrite, only about 1 or 2 percent ferrite will be developed in the resultant weld deposits. This is not the case in fluxing processes such as when using coated arc electrodes or submerged arc, since the flux is enriched with additional ferrite formers resulting in higher ferrite contents in the resultant weld deposits. Similarly, the amount of ferrite that may exist in any given weld will vary across the width of the weld deposit depending upon the base materials being joined. For example, when fully austenitic wrought product is welded, the interface regions will be practically zero percent ferrite because of the resultant base metal dilution, but it will progressively increase toward the weld centerline. Conversely, when a two-phase (austenitic + ferrite) cast product which normally contains over 15 percent ferrite is welded, the interface region will be high in delta ferrite content depending upon the amount of delta ferrite available and diluted from the casting base material. The ferrite distribution in a weld will also vary depending upon the weld position. That is: in areas of the downhand and horizontal position, weld deposit ferrite will be the highest; whereas, in the vertical and overhead position, weld deposit 3A-12 SGS-UFSAR Revision 31 December 5, 2019
ferrite will be the lowest in a given weld because of different welder manipulations necessary to overcome effects of gravity. In addition, types 310 and 330 weld materials are always fully austenitic, yet sound welds are being made every day with these alloys using fine tuned welding procedures. Also, welds are being made without the use of filler metal, such as electron beam welds and autogeneous gas shielded tungsten arc welds. Furthermore, the limits as set are arbitrary because various methods used to measure the percentage of delta ferrite yield widely differing results. The Welding Research Council has recognized this situation and have an organized approach which may result in an acceptable solution. The basis for classifying the low, medium, and high energy input ranges is not given in the Regulatory Guide. Using the Westinghouse conservative welding procedure parameters, the following energy inputs are being applied to produce high quality welds. They are:
1. SMAW 15.4 to 95 kJ/in. using 1/16 to 3/16 dia electrodes

2. GTAW 2.16 to 32.5 kJ/in. using .03 to 1/8 dia wires

3. GMAW 46 to 55 kJ/in. using .03 to 1/16 dia wires

4. SAW 74 to 79 kJ/in. using .09 to 1/8 dia wires Westinghouse has a large amount of evidence showing that the above energy input ranges produce fissure-free weldments in both shop and onsite welding.
Westinghouse does not require in-process delta ferrite determination. When the welding material is tested (in accordance with the requirements of ASME Section III, NB2430, and includes delta ferrite determinations), sound welds displaying more than one 3A-13 SGS-UFSAR Revision 6 February 15, 1987
percent average delta ferrite content by any agreed method of determination will be considered unquestionable. All other sound welds which display less than 1 percent average delta ferrite will be considered acceptable provided there is no evidence of malpractice or deviation from procedure parameters. If evidence of the latter prevails, sampling will be required to determine the acceptability of the welds. The sample size shall be 10 percent of the welds in the system or component. If any of these weld samples are defective, that is, fail to pass bend tests as described by ASME, Section IX, all remaining welds shall be sampled and all defective welds shall be removed and replaced. Field welding of the nuclear steam supply system and other nuclear class components is performed using Public Service Electric and Gas (PSE&G) welding procedures. In some areas of austenitic stainless steel welding, these procedures call for use of the 16-8-2 electrode. This particular electrode composition was developed to provide fissure-free welds in austenitic systems without reliance on ferrite content, which is generally limited to 3 percent, and frequently the amount is less than 1 percent. Therefore, ferrite control and determination, which comprise the bulk of the Regulatory Guide, are not considered applicable to the 16-8-2 welding electrode. The 16-8-2 welding electrode was initially developed for service temperatures where delta ferrite exhibits a tendency to transform into the sigma phase, and embrittling condition in austenitic stainless steel. Service temperatures at the Salem Station are too low to support a need for this type of protection, but PSE&G's long service history with this welding composition (since 1955) in steam piping systems has provided a level of confidence and expertise which overrides the consideration of alternate materials. Service and inspection records show that numerous welds have been performed satisfactorily in high pressure steam service temperatures up to 1100F for operating times exceeding 150,000 hours in the PSE&G generating systems. Regulatory Guide 1.31 - CONTROL OF FERRITE CONTENT IN STAINLESS STEEL WELD METAL REVISION 4, OCTOBER 2013 Salem Generating Station adheres to Revision 4 of Regulatory Guide 1.31. Per this revision of the regulatory guide, ferrite content in the weld metal as depicted by a ferrite number (FN) of weld metal used for welds in austenitic stainless steel core support structures, reactor internals, and class 1, 2 and 3 components should be between 5 and 20. The lower limit provides sufficient ferrite to avoid microfissuring in welds, whereas the upper limit provides a ferrite content adequate to offset dilution and reduce thermal aging effects. 3A-14 SGS-UFSAR Revision 31 December 5, 2019
PSE&G welding and inspection practices comply with the intent of the Regulatory Guide and Appendices A and B to 10CFR50 in the following manner:
1. Strict control is maintained over electrode chemistry and identification for procedure qualification, welder qualification, and production welding. This is accomplished through purchase specifications, certified mill test reports, segregation of untested lots from approved lots, locked storage of welding supplies on site, recorded allocation of electrodes to welders, and maintenance of lot identity from site receiving to completed weld joint.

2. The weld procedure qualification demonstrates the capability of producing welds free from unacceptable fissuring. This includes visual examination of procedure qualification bend bars and macrotech specimens with the unaided eye and under 10 power magnification.

3. Welder performance qualification bend bars, when made, are examined in the same manner to verify that the welder's technique maintains freedom from unacceptable fissures.

4. Welds for nuclear class systems are subjected to a liquid penetrant and radiographic examination where required. Heavy wall welds, such as in the reactor coolant piping, are subjected to in-process examinations by a liquid penetrant and radiography at one or more intermediate stages in the welding out of the groove.

5. Ferrite content for each lot of austenitic stainless steel electrode is qualified by magnegage measurements of a test weld pad. For nuclear plant welds, ferrite 3A-15 SGS-UFSAR Revision 6 February 15, 1987
outside the range of 5FN to 20FN for E-308, E-309, and E-316 is considered rejectable.
6. Production welding parameters are monitored on a spot-check basis by the field welding supervision and the Field Quality Control Group.
Regulatory Guide 1.32 - USE OF IEEE STANDARD 308-1971, "CRITERIA FOR CLASS 1E ELECTRIC SYSTEMS FOR NUCLEAR POWER GENERATING STATIONS" The Salem Station design satisfies the requirements of IEEE Standard 308-1971, with the exception that Class 1 diesel fuel oil storage capacity provides less than seven days of diesel operation under worst case loading. See Section 9.5.4 for a description of how long term Emergency Diesel generator fuel oil storage requirements are met. Regulatory Guide 1.33 - QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION), 2/78 (endorses N18.7-1976/ANS 3.2) The Salem Generating Station is committed to the requirements of NQA-1-1994. See the Quality Assurance Topical Report, Appendix C, Section 1.3.2.3 for further discussion. Regulatory Guide 1.34 - CONTROL OF ELECTROSLAG WELD PROPERTIES Electroslag welding of Nuclear Classes 1 and 2 components is confined to the area of reactor coolant piping elbows. These are made from cast clamshells of ASTM A351 Gr. CF-8M joined together on longitudinal seams by the electroslag process. Welding of these components was performed under specified weld procedure control monitored by Westinghouse. PSE&G also established that the shop production welds were in conformance to the procedure qualification. 3A-16 SGS-UFSAR Revision 31 December 5, 2019
Regulatory Guide 1.35 - _I_N_S_E_R_V_I_C_E_____S_U_R_V_E_I_L __LAN ___C_E____O __F____U_N_G_R_O_U __T_E_D___T_E_N__ D_O_N_S___I__ N PRESTRESSED CONCRETE CONTAINMENT STRUCTURES This Regulatory Guide is not applicable to the Salem Station containment structures. Regulatory Guide 1.36 - NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL The Salem Station design conforms with the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.37 - QUALITY ASSURANCE REQUIREMENT FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS, 3/73 The Salem Station program for cleaning of fluid systems and associated components conforms to NQA-1-1994 and the intent of the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.38 - QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR PLANTS, 10/76 Quality Assurance requirements comply with the requirements of NQA-1-1994. Regulatory Guide 1.39 - HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS, 3/73 Salem Generating Station complies with the requirements of NQA-1-1994. 3A-17 SGS-UFSAR Revision 24 May 11, 2009
Regulatory Guide 1.40 - QUALIFICATION TESTS OF CONTINUOUS - DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS The Salem Station design conforms to the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.41 - PREOPERATIONAL TESTING OF REDUNDANT ON-SITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS The preoperational testing program for the onsi te electric power system will verify the independence of each redundant power source and its ability to supply its associated load group. The actual test methods being developed may not be identical to those specified in the Regulatory Guide for the following reasons. The Salem plant design includes the safeguards equipment controller which incorporates several testing features to assure that loading of the diesel-generator will be accomplished in accordance with the design requirements. It is expected that a series of overlapping tests will adequately substitute for the tests specified in the Regulatory Guide. The Unit 2 initial preoperational test program is in full conformance with the Regulatory Guide which functionally demonstrates the independence among redundant onsite power sources and their load groups. This is accomplished by the performance of the Integrated Safeguards Test. As stipulated in part c.1 of the Guide, isolation from the offsite transmission network will be accomplished by the direct actuation of the undervoltage sensing relays (opening the 4 kV ac undervol tage relay knife switches) . All loads off the Unit 2 group buses not required to maintain necessary and independent construction and testing activities, as well as backup power to Unit 1, will be de-energized to the maximum extent practical. The functional testing requirements covered under c.2 and c.3 of this guide are performed as part of the Integrated Safeguards Test. 3A-18 SGS-UFSAR Revision 6 February 15, 1987
Regulatory Guide 1.42 - INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT-WATER-COOLED NUCLEAR POWER REACTORS Although NRC Regulatory Guide 1.42 was withdrawn by the NRC on March 22, 1976, SGS commitments, as stated below, are not affected by this withdrawal. Gaseous radioiodine releases from the Salem Station will conform with the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.43 - CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW ALLOY STEEL COMPONENTS The Salem Units 1 and 2 reactor vessel flange, shell course, head and nozzle surfaces in contact with primary coolant were clad with stainless steel weld metal. The original reactor vessel heads and shell courses were constructed of ASTM A-302 Grade B for the dome and peel sections, and ASTM A-508 Class 2 for the flange (Unit 1), and ASTM A-533 Grade B Class 1 for the dome and peel sections, and ASTM A-508 Class 2 for the flange (Unit 2) plate material made to fine grain practice, and clad by the 3-wire submerged arc process, which is a low heat-input process. This material and cladding process are not restricted by Regulatory Guide 1. 43 and consequently these portions of the vessels comply directly with the guide. The original head and vessel flanges and the primary nozzles for both units were constructed of A-508 Class 2 material, and clad by the manual metal arc (MMA) and manual inert gas (MIG) processes. In addition, the 3-wire submerged arc process was also used on the flanges for both units. The 3-wire submerged arc and MMA processes are low heat-input process to the same degree as MMA and submerged arc processes. The replacement RVCH on both units are constructed from a monoblock forging of SA-508 Gr. 3 Cl. 1 steel rather than by welding formed plates and a flange. The surfaces in contact with reactor coolant were clad with stainless steel weld overlay using either the submerged arc welding process or the shielded metal ark welding process following strict preheat and temperature requirements. Therefore, the Salem Units 1 and 2 replacement reactor vessel closure heads comply with Regulatory Guide 1.43. 3A-19 SGS-UFSAR Revision 22 May 5, 2006
However, as recognized by the Regulatory Guide, and as shown by the extensive metallurgical examinations and fracture mechanics evaluation performed by Westinghouse ( 1), underclad cracking, if present, would be of high integrity and would have no detrimental effect on the structural integrity of the affected components. Thus, the vessels are suitable for the use intended. For Unit 1 steam generators, stainless steel weld cladding is applied to the steam generator channel heads in contact with primary coolant. The heads, including the nozzles and manway openings, are cast ASTM A-216 grade WCC material. The head hemispherical surfaces are clad by the two-wire series submerged arc process with controlled dilution of the deposit, and the channel head nozzles and manway openings are clad by the oscillating submerged arc single wire technique. Both processes are low-heat input techniques. The material and the weld processes are not restricted by the Guide. For the Unit 2 steam generators, the heads, including the tubesheet, nozzles and manway openings are forgings, SA-508, Gr3, Cl.2. The ferritic base metals that are clad are procured to fine grain practice and are not considered susceptible to underclad cracking. Weld procedure qualification is performed on material of the same specification (or equivalent) as used in production. For the Unit 2 steam generators, all primary side ferri tic steel surfaces (primary side of the tubesheet and inside surfaces of the primary head) are clad to prevent corrosion. The tube sheet is clad with Inconels 82 and 182. Due to its sensi ti vi ty to reheat cracking, Inconel 52 is not used for the tubesheet cladding on which the tube-to-tubesheet welding is performed. The most stressed areas of the channel head in contact with the primary coolant are clad (buttering of the tubesheet on which the divider plate is welded) or welded with Inconel 152. These areas are associated with the divider plate to primary head and tubesheet junctions. The primary head is clad with Type 308L and 309L stainless steel. Stainless steel weld cladding is applied to the pressurizer shell courses, heads, spray nozzle, and manway opening surfaces in contact with primary coolant. The shell courses are constructed of SA-533 Class 1 plate, use of which is not restricted by the guide. (1) Westinghouse Nuclear Energy Systems Report WCAP-7733, "Reactor Vessels Weld Cladding- Base Metal Interaction," T. R. Mager, et al., April 1971. 3A-20 SGS-UFSAR Revision 24 May 11, 2009
In addition, the shell courses are clad by the two-wire series submerged arc process with controlled dilution of the deposit, which is a low-heat input technique. The pressurizer heads are cast SA-216 grade WCC material, applied by a low-heat input process, namely the two-wire series submerged arc process with controlled dilution of the deposit. Because of the materials and/or the low-heat input processes used for stainless steel weld cladding of the pressurizer and steam generators, no underclad cracking is expected, and the intent of the Regulatory Guide is met for these components. Regulatory Guide 1.44 - CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL Treatment of sensitized stainless steel components of the Nuclear Steam Supply System (NSSS), particularly the reactor vessel nozzle safe ends, has been covered in detail in the Westinghouse Topical Report, WCAP-7477-L, "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," March 1970. This report indicates that, where applicable, Westinghouse and their subcontractors have complied with the intent of the Regulatory Guide. Field erection procedures under the direct supervision of PSE&G and United Engineers and Constructors (UE&C) will comply with the regulatory positions wherever possible. Exposure to sensitizing temperatures in field welding operations will be monitored by testing weld procedure qualification specimens. Regulatory Guide 1.45 - REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS The Salem Station design conforms to the intent of the Regulatory Guide. Regulatory Guide 1.46 - PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT Although NRC Regulatory Guide 1.46 was withdrawn by the NRC on March 11, 1985, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station is protected against pipe whip inside containment as described in Section 3.6. 3A-21 SGS-UFSAR Revision 24 May 11, 2009
Regulatory Guide 1.47 - BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR PLANT SAFETY SYSTEMS The Salem Station design meets the requirements of the Regulatory Guide with the exception of regulatory position C. 4. The Salem Station design does not provide for manual initiation of individual alarms. The Overhead Alarm System has the capability to manually initiate all alarms for testing purposes. It is believed that this arrangement is a suitable alternative to regulatory position C.4. Regulatory Guide 1.48 - DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS Although NRC Regulatory Guide 1.48 was withdrawn by the NRC on March 11, 1985, SGS commitments, as stated below, are not affected by this withdrawal. As stated in Section 3. 9, the piping design commitment for the Salem Station was to utilize the design philosophy of ANSI B31. 1, but with allowances of 1.2 Shand 1.8 Sh for the summation of primary type longitudinal stresses under the OBE and DBE earthquake loadings, respectively. Although the design commitment was made early in the design process, these values conform closely with values published in paragraphs NC-3611.1 (b) (4) (c) (1) and (2) of the Winter 1972 Addendum to ASME Section III. Thus, the suggested stress limits for piping in the Regulatory Guide appear to be satisfied. Despite the above stress limit commitments, in a number of cases actual stresses encountered were below Sh, the code limit for the "normal" condition. It is believed that this fact provides a reasonable assurance of the continued operability of equipment connected to this piping, and, as such, satisfies the intent of the Regulatory Guide. 3A-22 SGS-UFSAR Revision 13 June 12, 1994
Regulatory Guide 1.49 -POWER LEVELS OF WATER-COOLED NUCLEAR POWER PLANTS The license application power levels and ultimate power levels of the Salem units are below those maximum power levels set forth in the Regulatory Guide. Regulatory Guide 1.50 - CONTROL OF PREHEAT TEMPERATURES FOR LOW-ALLOY STEEL WELDING Field welding of low-alloy steel on Nuclear Class systems conforms with the regulatory positions set forth in the Regulatory Guide. Regulatory Guide 1. 51 - INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR PLANT COMPONENTS NRC Regulatory Guide 1.51 was withdrawn by the NRC on July 15, 1975. 10CFR50.55 a (g) (4) addresses the design, access considerations and Pre-Service Inspection exam requirements, for ASME Section XI Codes that become effective subsequent to the construction of those applicable components. Later editions of the Code have incorporated inspection requirements for Class 2 and 3 components. 3A-23 SGS-UFSAR Revision 16 January 31, 1998
Regulatory Guide 1.52 - DESIGN, TESTING AND MAINTENANCE CRITERIA FOR ENGINEERED SAFETY FEATURE ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ABSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (Revision 1, 7/76) The Salem Station atmosphere cleanup systems which fall within the scope of the Regulatory Guide are as follows: 3A-23a SGS-UFSAR Revision 30 May 11, 2018
THIS PAGE INTENTIONALLY LEFT BLANK 3A-23b SGS-UFSAR Revision 13 June 12, 1994
Primary Systems: 1. Containment Fan Cooler Units Secondary Systems: 1. Control Room Emergency Filtration Unit
2. Auxiliary Building Exhaust Units

3. Fuel Handling Building Exhaust Units The Fuel Handling Ventilation (FHV) system was originally considered an Engineered Safety Feature ventilation system since the original fuel handling accident (FHA) dose analysis credited the filtration system for removal of radioactive material. In Amendments 251/232, the Salem fuel handling accident dose analysis was converted from the original TID-14844 method to alternate source term (10CFR50.67). The revised FHA dose analysis removed the credit for the FHV system filtration and the HEPA and Charcoal filter requirements. In Amendments 263/245, the requirements for FHV charcoal and HEPA filtration were removed from the Technical Specifications. Subsequent TS amendments 271 (Unit

2) and 289/273 revised the fuel decay time used in the FHA analysis. The revised FHA analysis continued to not credit the HEPA and Charcoal filtration but also evaluated a case were the radiological dose limits would be met without the operation of the fuel handling building ventilation system.
Amendments 334/315 removed the FHV system from the Salem Technical Specifications. With these changes to the dose analysis, the FHV is no longer considered an ESF ventilation system as described in RG 1.52. Periodic testing of the FHV system flowpath and filtration will be controlled in accordance with the preventative maintenance program. All of these systems conform to the intent of the Regulatory Guide in many respects. The areas where the systems are at variance with the intent of the regulatory positions follow: Regulatory Position C.2.a. In the secondary systems, those portions of the systems designed for mitigation of accident doses are standby units and are not redundant. These filter banks are used in conjunction with other normally operating components which are maintained operable or fail-safe under an accident condition. High-efficiency particulate air (HEPA) or other types of after-filters are not provided downstream of charcoal filter banks. The use of such after-filters has not previously been practiced in Pressurized Water Reactor (PWR) design. The Salem charcoal filters have been designed for relatively low levels of iodine deposition and the charcoal cells are to be blown free of "fines" during manufacture. Regulatory Position C.2.j. The primary and secondary systems are not "intact" units. Individual components will be maintained or replaced individually. It would be impractical to fabricate, ship, or install "intact" units of the size used in the Salem Station design. These units generally consist of modular sections that are maintained 3A-24 SGS-UFSAR Revision 32 June 17, 2021
individually. Exposure of workers during maintenance is taken into consideration. Regulatory Position C.2.1. Although the leakage rate from the housings in the primary system is designed to be 1 percent of the design flow rate, system ductwork leakage is permitted to be as much as 5 percent of the design flow rate. Because the primary system is a recirculation system entirely within the containment, it is unnecessary to require a low leakage rate from the ductwork. The 5-percent leakage rate allowed is reasonable and practical for sheet metal construction. Regulatory Position C.3.c. Prefilter materials have not been restricted to those on the UL Building Material List. Flammability, corrosion, erosion, radiation resistance, and other aspects of materials generally available have been considered. In general, the Westinghouse guidelines for materials acceptable within containment have been used, as well as recommendations appearing in ORNL-NSIC-65 ( 1) . Regulatory Position C.3.e. Filter and absorber mounting frames will be painted. An epoxy base paint will be used which provides resistance to general rusting and chemical attack, and allows for easy decontamination when necessary. Regulatory Position C.3.h. The use of zinc coated (galvanized) steel has not been discouraged in the Salem Station systems. It is primarily used for ductwork and is not discouraged for such use in ORNL-NSIC-65. 3A-25 SGS-UFSAR Revision 6 February 15, 1987
Regulatory Position C.4.a Vacuum breakers are not provided on the access doors of the filter trains, since it is not intended to allow any person to enter the housings while they are in use, except during performance of certain tests under controlled conditions. Regulatory Position C.5.a Since the ABVS exhaust air filtration system is not considered an ESF, testing following painting, a fire, or a chemical release in any ventilation zone communicating with the system would only be required if an evaluation concluded that the painting, a fire, or a chemical release could contaminate, or adversely impact, the charcoal adsorbers (and HEPA filters) from the fumes, chemicals, and foreign materials. Regulatory Position C.5.b. The planned DOP testing for HEPA filters is:
1. For normally operating units, initially and upon replacement (estimated to be after 9 to 18 months' use)

2. For standby units, initially, 18 months, and upon replacement.
Filter manufacturers have indicated that HEPA filters exposed to a relatively clean atmosphere can be expected to last for more than 3 years. Intermittent DOP testing of these long life filters will not necessarily detect an adverse change in filter efficiency. The use of silicone sealants is not prohibited in the atmosphere cleanup systems. ORNL-NSIC-65 suggests the use of silicone rubber as a general caulking or sealing compound (1). Regulatory Position C.5.c. Since the ABVS exhaust air filtration system is not considered an ESF, testing following painting, a fire, or a chemical release in any ventilation zone communicating with the system would only be required if an evaluation concluded that the painting, a fire, or a chemical release could contaminate, or adversely impact, the charcoal adsorbers (and HEPA filters) from the fumes, chemicals, and foreign materials. (1) Burchsted, C. A.; and Fuller, A. B., "Design, Construction and Testing of High-Efficiency Air Filtration Systems for Nuclear Application," ORNL-NS1C-65, Oak Ridge National Laboratory, January 1970. 3A-26 SGS-UFSAR Revision 30 May 11, 2018
The acceptance criteria for in-place DOP testing of HEPA filters is penetration less than 1% as per the guidance of Generic Letter 83-13 for assumed charcoal adsorber efficiency of at least 90% and assumed HEPA filter efficiency of at least 99%. Regulatory Position C.5.d. The acceptance criteria for leak testing of activated charcoal adsorber sections is bypass leakage less than 1% as per the guidance of Generic Letter 83-13 for assumed charcoal adsorber efficiency of at least 90% and assumed HEPA filter efficiency of at least 99%. Regulatory Guide 1.53 - APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS The Salem Station Protection System design, in general, meets the requirements of the Regulatory Guide. The Salem Station protection system is designed in accordance with the requirements of IEEE Standard 279-1971 which requires that any single failure within the Protection System shall not prevent proper protective action at the system level. The design of the Protection System included the application of the single failure criterion to the logic and actuators. Testing provisions for the Protection System which assure the operability of equipment needed to perform a protective function include continuity checks for those cases, as described in Regulatory Guide 1.22, where there is no practicable system design that would permit testing of actuated equipment during power operation. The testing provisions assure that failures within the Protection System are detectable. Regulatory Guide 1.54 - QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS, 6/73 (endorses N101.4) ANSI N101.4 is being used to provide guidelines for Quality Assurance requirements for safety-related coatings. Regulatory Guide 1.55 - CONCRETE PLACEMENT IN CATEGORY I STRUCTURES Although NRC Regulatory Guide 1.55 was withdrawn by the NRC on July 21, 1981, SGS commitments as stated below are not affected by this withdrawal. 3A-27 SGS-UFSAR Revision 30 May 11, 2018
The concrete placement in Salem Category I structures generally conforms with the intent of the Regulatory Guide. ACI 318 as well as ACI 301 are followed as the basis of the concrete specification. Concrete compression tests are evaluated in accordance with ACI 301-66, Chapter 17. During concrete operations, an inspector was at the batch plant to certify the mix proportions for each batch of concrete. The Inspector took samples of the ingredients and ran tests periodically to determine moisture content of aggregates, gradation of aggregates, and temperatures of materials when applicable. An inspector was also at the construction site to inspect reinforcing and form placements, make slump tests, and take concrete test cylinders in accordance with specified procedures. The project specification for concrete also requires construction joints to be shown on the design drawings. Once the construction joints were established, concrete was poured in accordance with the drawings. Placement of concrete during cold and hot weather is as described in Section 3. 8 of the FSAR. Contact between designer and constructor was well maintained. Experienced field personnel supervised the concrete placement in accordance with the specifications and good practice. Regulatory Guide 1.56 - MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1.57 - DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS This Regulatory Guide is not applicable to the Salem Station which has a reinforced concrete containment with a steel liner. 3A-28 SGS-UFSAR Revision 13 June 12, 1994
Regulatory Guide 1.58 - QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL, 9/80 NRC Regulatory Guide 1. 58 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994. 3A-29 SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.59 - DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS (Revision 2) The design basis floods and the protection requirements for the Salem Station are discussed in Sections 2.4 and 3.4. Regulatory Guide 1.60 - DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS Normalized Housner' s average ground response spectra were used for the Salem Station design. The design response spectra presented in the Regulatory Guide are based on three reference papers published in 1973, and are not applicable to the Salem Station design. Regulatory Guide 1.61 - DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS The damping values used for seismic design of the Salem Station are more conservative than those presented in the Regulatory Guide. 3A-30 SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.62 - MANUAL INITIATION OF PROTECTIVE ACTIONS The design of the protection systems for the Salem Station conforms with the intent of the Regulatory Guide. Regulatory Guide 1.63 - ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER-COOLED NUCLEAR POWER PLANTS The Salem Station design conforms with the intent of the Regulatory Guide (See Section 8.1.5). Regulatory Guide 1.64 - QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS, 10/73 SGS is committed to the requirements of NQA-1-1994. Regulatory Guide 1.65 - MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS The Regulatory Guide was published after procurement of the reactor vessel bolting material for the Salem Units 1 and 2. However, this material meets the intent of the Regulatory Guide as follows: The reactor vessel closure stud bolts for both units were machined from bars of SA 540 Grade B24 material. The closure nuts and washers were machined from tubes of SA 540 Grade B23 material. The bolting material qualification tests were performed per the ASME Section III Code and Addenda ( 1965 Summer Addenda for Unit 1 and 1966 Summer Addenda for Unit 2) in effect at the 3A-31 SGS-UFSAR Revision 25 October 26, 2010
time of the bolt procurement, which required meeting an average of 35 ft-lbs energy with no lateral expansion tests required and no maximum tensile strength limitation. All bolting material for both units met the ASME code requirements. Charpy tests were performed at 10°F on bolting material bar and tube specimens (three impact tests per bar or tube end) as required by the ASME Code. For the bar material from which the Unit 1 studs were made, all data from the bars tested were in excess of 45 ft-lbs, which conforms directly with the impact energy requirements of the guide. For Unit 2 material, six out of nine bars tested showed impact values of 45 ft-lbs or greater at both ends. Of the remaining three bars, the one with the lowest impact energy (at one end) showed values of 40, 43, and 45 ft-lbs. For the tube material from which the Unit 1 nuts and washers were made, none of the tubes that were tested showed impact energy values of 45 ft-lbs or greater on both ends of the tube. The lowest energy values, obtained on one end of one tube, were 38, 38, and 38 ft-lbs. For Unit 2, seven of eleven tubes tested showed impact energy values at both ends of 45 ft-lbs or greater. Of the remaining tubes, the one with the lowest impact data showed values of 40, 42, and 45 ft-lbs (at one end) . For the bars and tubes showing 10°F impact data averaging below 45 ft-lbs, Westinghouse believes that the intent of the guide is met, inasmuch as sufficient fracture toughness is expected at the preload temperature or at lowest service temperature, both of which are significantly above the 10°F Charpy test temperature. Also, the impact energies at the preload or lowest service temperatures will be higher than was obtained at the lower, actual Charpy test temperature. 3A-32 SGS-UFSAR Revision 13 June 12, 1994
The bolting materials for both units were inspected to the requirements of the ASME Code and Addenda requirements in effect at the time of bolt procurement. A radial scan covering 100 percent of the circumferential surface of the bars, based on a standard back reflection, was performed. The requirements for protection of the stud bolts and stud bolt holes against corrosion and in-service inspection requirements for the bolting material are discussed in Section 5. Regulatory Guide 1.66- NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS Although NRC Regulatory Guide 1. 66 was withdrawn by the NRC on September 28, 1977, SGS commitments, as stated below, are not affected by this withdrawal. Tubular products and fittings have been nondestructively examined in accordance with the requirements of the applicable paragraphs of ANSI B31.7. Regulatory Guide 1.67 - INSTALLATION OF OVERPRESSURE PROTECTION DEVICES Although NRC Regulatory Guide 1.67 was withdrawn by the NRC on April 15, 1983, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station design conforms to the intent of the Regulatory Guide. Any deviations from the design recommendations of the Regulatory Guide have been analytically justified. Regulatory Guide 1.68 - PREOPERATIONAL AND INITIAL STARTUP TEST PROGRAMS FOR WATER-COOLED POWER REACTORS The Salem Station Preoperational and Initial Startup Test Program 3A-32a SGS-UFSAR Revision 13 June 12, 1994
conforms with the intent of Revision 0 of the Regulatory Guide with the following exceptions: Appendix A A.1.c- Vibration Tests -Tests will be performed as indicated in Section 14.4, Exception 3, and Section 3.9. 3A-32b SGS-UFSAR Revision 26 May 21, 2012
A.4.a - Power Conversion System - System operability tests will be through performed; expansion and restraint tests A.4.h are not planned. A. 5.p - Leak Detection Systems - Extensive testing on an individual basis of sumps, drainage system, and instrumentation will be accomplished during the testing program. A.6.e- Emergency Power Systems - Tests of this system as presently planned do not provide for the carrying of all required loads for several hours. A.l3 - Radioactive Waste Systems - No tests are planned at this time to verify the amount of plateout in sample system piping. This is a long-range program which is not adaptable to the preoperational and initial startup phases of testing. B.l.j - Vibration Monitoring - Reference FSAR Section 14.4, Exception 3, and Section 3.9. D.l.a- Natural Circulation Tests Reference Section 14. 4, I Exception 1. D.l.n- Dropped Rod- Reference Section 14.4, Exception 7. D.l.p- Vibration Monitoring - Reference Section 14.4, Exception 3, and Section 3. 9. The Unit 2 initial test program is in conformance with the Regula tory Guide with the following exceptions: Paragraph 2a - The shutdown margin shall be verified by boron analysis only to ensure the boron concentration is as required by the Technical Specifications. 3A-33 SGS-UFSAR Revision 10 July 22, 1990
Paragraph 2f - Vibration levels and piping reaction to transient conditions are evaluated prior to fuel loading as discussed in Sections 3.9 and 5.2. Differential pressure instrumentation is not part of the Salem Station design for monitoring the differential pressure across the reactor vessel and steam generators. The cold and hot leg temperatures are used for monitoring fouling of the major components. Pump flow measurements are calculated during the test program which will verify that the differential pressure across the major components is not excessive. Paragraph 4a - The boron coefficient will be determined at the normal hot zero power temperature. The boron coefficient over the temperature range allowed by the Technical Specifications for criticality does not vary significantly to require additional measurements to be made. Paragraph 4u - See Exception 6a, Section 14.4. Paragraph 5c - See Exception 1, Section 14.4. Paragraph 5c - This requirement is not applicable to the Salem Station design. Paragraph 5d- This requirement is only applicable to BWRs. Paragraph 5e - See Exception 5, Section 14.4. 3A-34 SGS-UFSAR Revision 10 July 22, 1990
Paragraph 5g - Adequately covered prior to Power Ascension Program plus Normal Technical Specification and/or operational surveillance. Demonstration at specified power levels provides no additional information. Paragraph 5i - The system used for determining the individual rod positions is the Rod Position Indication System. The Technical Specifications require that the system be operable to detect control rod position within the Technical Specification limits. Paragraph 5j -A specific test to verify that the reactivity control system functions in accordance with design will not be included in the test program. These systems are continuously being tested during the test program whenever power level changes are made. In addition, response to 5g applies. Paragraph 5m - No additional testing of the Reactor Coolant System is planned subsequent to initial criticality. See response to 5g. Paragraph 5q - Failed fuel detectors are calibrated, alarms verified operable, and are continuously monitored while in service. No special tests are required. Paragraph 5r See Exceptions, Section 14.4. In addition, response to 4o applies. Paragraph 5s - See response to 5g. Paragraph 5t - Relieving capacities not verified. Additionally, see response to 5g. Paragraph 5u - See response to 5g. 3A-35 SGS-UFSAR Revision 25 October 26, 2010
Paragraph Sv - Accomplished during various other power ascension procedures, but not necessarily conducted at designated power levels. Additionally, see response to Sg. Paragraph Sw - Station design does not permit measurement of component temperatures. Shielding and penetration cooling systems are demonstrated operable during preoperational testing. Paragraph Sx - Room coolers, Service Water System pumps, Core Cooling System, and integrated safeguards verify minimum operating components available. Engineering design assures adequate heat removal capacities at minimum specified flow rate. See response to Sg. Paragraph Sec - See response to Sg. Paragraph Sdd- See response to Regulatory Guide 1.68.2. Paragraph See - See response to Sg. Paragraph Sff - See response to Sg. Paragraph Sgg The demonstration of the operability of equipment for anticipated transients without a reactor trip is still under discussion between the NRC and Westinghouse and other NSSS vendors. This test requirement will not be performed until agreement is reached, if required. Paragraph Sii - See Exceptions, Section 14.4, and response to Sg. Paragraph Skk - No plans for this demonstration. The response of the station to bypassing a single feedwater heater that results in the most severe credible case of feedwater temperature reduction is analyzed in Section 1S. The consequences of this incident were shown to be more moderate than those considered for the Excessive Load Increase Accident, which is equivalent to a 10-percent step change in station power. The 10-percent step changes are 3A-36 SGS-UFSAR Revision 10 July 22, 1990
performed during the Phase III test program and are considered adequate to check station responses. Paragraph 511 - See Exception 6, Section 14.4. Paragraph 5mm The main steam isolation valves will close during power operation in response to any one of the following signals being generated by the station reactor protection system or operator initiation:
1. 2/4 high steam line flow in coincidence with 2/4 low-low Tavg or I 2/4 low steam line pressure.

2. Hi-Hi containment pressure.

3. Manual actuation from the control room console (4 separate pushbuttons only.)
Signal 1 will also generate a safety injection signal which initiates a reactor trip. Signal 2 is generated after a Hi containment pressure signal is generated which will initiate a safety injection and reactor trip. For Signals 1 and 2, a reactor trip will precede the closing of the main steam isolation valves. Signal 3 will only be initiated when the operator manually initiates each signal to each main steam isolation valve from the control room console. This is done by procedure, in which case a reactor trip will have occurred. 3A-37 SGS-UFSAR Revision 16 January 31, 1998
The closing of the main steam isolation valves while at power will not be performed because in all cases a reactor trip will precede the closing of the main steam isolation valves during plant operation. Paragraph 5oo - See response to 2f. Regulatory Guide 1.68.2 - INITIAL STARTUP TEST PROGRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILITY FOR WATER-COOLED NUCLEAR POWER PLANT The initial startup test program conforms to Objectives C.1.a and C.1.b of the Regulatory Guide. The design of the plant, however, does not permit Objective C.1.c, verification of cold shutdown capability, to be performed. The Salem Station was designed for remote hot shutdown from outside the control room. This is described in Section 7.7 of the FSAR. Our capability to go to a cold shutdown condition through the use of procedures and temporary modification is described in Section 7. General Design Criterion 19 of 10CFR50 Appendix A requires a design capability for remote hot shutdown with a potential capability for subsequent cold shutdown through suitable procedures. A detailed procedure was written explaining the actions to be taken to bring the station from a hot shutdown to a cold shutdown condition from outside the control room. In order to demonstrate that the actions described in the operating procedure could be performed, a walk through encompassed a visual inspection of the areas that are required to be manned by the operators during various phases of remote cold shutdown. It demonstrated the availability and access to the equipment and also the communication required. No physical operation or changes will be made. The major steps for remote cold shutdown are the following: 3A-38 SGS-UFSAR Revision 6 February 15, 1987
1. Trip the reactor (this may be accomplished from the control room or locally - the reactor trip will also cause a turbine trip).

2. Reactor Coolant System temperature will decrease to the no-load value through automatic operation of the Steam Dump System or main steam atmospheric relief. Hot standby temperatures will be maintained by automatic system operation and can also be maintained by local manual control of the main steam atmospheric relief valves with termination of steam dump (closure of dump valves or MSIVs).

3. The Main Feedwater System isolation valves will close when Reactor Coolant System temperature reaches 554°F. Steam generator levels will be maintained by local manual operation of the Auxiliary Feedwater System.

4. Maintain station status from the remote hot shutdown panels.

5. Borate Reactor Coolant System to cold shutdown condition manually.

6. Take manual actions necessary to prevent inadvertent Safety Injection System operation as required in accordance with normal shutdown procedures.

7. Cooldown within appropriate limits through manual control of main steam atmospheric relief valves.

8. Maintain steam generator levels and pressurizer levels during cool down through manual control of auxiliary feedwater and charging.
3A-39 SGS-UFSAR Revision 16 January 31, 1998
9. Reduce pressurizer temperature and pressure through manual control of pressurizer spray valves within appropriate limits.

10. When Reactor Coolant System pressure reaches :c:; 1000 psig, manually close accumulator isolation valves.

11. Initiate Residual Heat Removal System operation locally when Reactor Coolant System temperature is below 350°F and pressure is below 375 psig.

12. Take actions necessary to arm Pressurizer Overpressure Protection System when Reactor Coolant System temperature is 312°F and pressure is less than 375 psig.

13. Bring unit to cold condition and maintain via local control.
Regulatory Guide 1.69 - CONCRETE RADIATION SHIELD FOR NUCLEAR POWER PLANTS The Salem Station design and construction of the concrete radiation shields conforms with the intent of the Regulatory Guide. Regulatory Guide 1.70- STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS The Salem FSAR was submitted prior to publication of this Regulatory Guide, and was not, therefore, prepared in the standard format suggested by the guide. The FSAR does, however, contain essentially all of the information suggested in the Regulatory Guide, either in the text or in the form of responses to specific questions from the regulatory staff. However, PSEG will utilize Regulatory Guide 1.181 in conjunction with Revision 1 of NEI 98-03, Guidelines for Updating Final Safety Analysis Reports, as guidance for maintaining the UFSAR in accordance with 10 CFR 50.71(e). 3A-40 SGS-UFSAR Revision 28 May 22, 2015
Regulatory Guide 1.71- WELDER QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY The welding supervisor's knowledge of the area and his assignment of welders to the job was relied upon to assure that satisfactory welds are made in locations of limited or restricted accessibility. Welds with limited accessibility may require radiography beyond Code requirements at the discretion of the site welding department. Site welding department personnel audited the welding parameters on production welds to assure that they were within the parameters established in the "Welding Procedure Specifications." Regulatory Guide 1.72- SPRAY POND PLASTIC PIPING This Regulatory Guide is not applicable to the Salem Station. Re gu 1 at or y Guide 1 . 7 3 - -"Q'-'-U-=-A=L:. :I:. :F:. . :I=-C:. :A_:_T=-=-I-'-0-=-N----=T-=E:..:S-=T'-'-S'--_O:_:F=----E=L=E--'-C-=T-'-R-"'I'--'C'---V'-"A-=L=-V-'--E=---O-=-=-P=E-=-RA=-=-=T-=O:..::R_:c:_S INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS Environmental qualification of motor operated valves located inside the containment is discussed in Section 3.11 and Section 7. Regulatory Guide 1.74- QUALITY ASSURANCE TERMS AND DEFINITIONS, 2/74 NRC Regulatory Guide 1.74 was withdrawn by the NRC. SGS is committed to the requirements of NQA-1-1994. Regulatory Guide 1.75- PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS 3A-41 SGS-UFSAR Revision 23 October 17, 2007
As discussed in Section 7. 5. 3 .1, the Salem Station electric systems do not conform to the recommendations in Regulatory Guide 1.75, since this was not an original design criterion. New equipment will be integrated into our existing separation provisions. The Salem separation criteria have been approved by the NRC staff as described in Section 7. 8 of the original and Supplement 1 and Section 8.4.5 of Supplements 3 and 4 of the Safety Evaluation Report. Regulatory Guide 1.78, Revision 1, December, 2001- Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release The Salem Generating Station design conforms to the intent of the Regulatory Guide. 3A-42 SGS-UFSAR Revision 26 May 21, 2012
Regulatory Guide 1.79- PREOPERATIONAL TESTING OF ECCS FOR PWRs This guide is addressed in Section 6.3. Regulatory Guide 1. 82 - SUMPS FOR EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS This guide is addressed in Section 6.3. Regulatory Guide 1.83 - INSERVICE INSPECTION OF PWR STEAM GENERATOR TUBES (Revision 1) Inservice inspection of steam generator tubing will be performed in accordance with the Technical Specifications, which are in general conformity with Regulatory Guide 1.83. Regulatory Guide 1.88 - COLLECTION, STORAGE, AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS, 10/76 NRC Regulatory Guide 1. 88 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994, Supplement 175-1, Section 4, Storage, Preservation, and Safekeeping, with the following specific exceptions for the Records Storage Room No. 145 in the Nuclear Administration Building:
1. Per NUGEG-0800, Records Storage Room No. 145 was built to comply with option ( 3) "a 2 hour rated fire resistant file room meeting NFPA 232 ... ".
Regulatory Guide 1.88 endorses NFPA 232-1975 and NQA-1-1994 endorses NFPA 232-1986; however, during construction, NFPA 232-1991 was utilized to provide an acceptable level of record protection, 3A-42a SGS-UFSAR Revision 24 May 11, 2009
2. A cable tray which passes through the room is enclosed with a 3 hour rated symmetrical wrap system to assure its presence will not effect the room contents or fire protection features, and

3. The ceiling is pierced by several miscellaneous drainage lines and two ventilation ducts. A drip pan, with discharge outside the room, is provided for the miscellaneous drainage plumbing to minimize the potential for inadvertent wetting of records and fire dampers are installed in the ventilation ducts.
Regulatory Guide 1.91 - EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANTS This subject is discussed in Section 2. Regulatory Guide 1.94 - QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS, 4/76 Major modifications made to the Salem Station will comply with NQA-1-1994. 3A-42b SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.97 - INSTRUMENTATION OF LIGHT-WATER COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT Implementation of this Regulatory Guide is described in Section 7. Regulatory Guide 1.99 - EFFECTS OF RESIDUAL ELEMENTS ON PREDICTED RADIATION DAMAGE TO REACTOR VESSEL MATERIALS The basis as well as the scope of the Regulatory Guide for predicting adjustment of reference temperature are inappropriate since the data base used was incomplete and included some data which were not applicable. The primary data were obtained from the 6-inch A302B ASTM reference correlation monitor material reported to be irradiated at 550°F and do not appear to include significant operating plant data which have been collected by industry. Specifically, it appears from our evaluation of the Regulatory Guide that data obtained from various Westinghouse surveillance capsules which were irradiated 19 2 at low fluences (1x 10 n/ cm ) were not included in the development of the slope of the curve. Also, data from the Yankee Rowe surveillance capsules, which were irradiated at 450°F to 525°F (1) were inappropriately included in the development of the slope for the ASTM reference material. As indicated in the Regulatory Guide itself, irradiation at 450°F has been shown to cause twice the adjustment of reference temperature when compared to irradiation at 550°F. Inclusion of the low fluence surveillance capsule data and detection of the high fluence (low temperature) Yankee Rowe capsule data in the development of the curve results in a curve which is significantly different in slope than the curve in the Guide which results in higher predicted adjustment of reference 3A-43 SGS-UFSAR Revision 6 February 15, 1987
temperature at high fluences (2). Preliminary analyses of irradiation data being performed by the Metal Properties Council Task Group, which has been assigned the task of developing radiation damage curves, confirm the Westinghouse contention that the slope of the curves presented in the Regulatory Guide would result in unrealistic adjustments in reference temperature. With reference to the Guide Position controlling residual elements to levels that result in a predicted adjusted reference temperature of less than 200°F at end-of-life, Westinghouse contends that the stresses in the vessel can be limited during operation in order to comply with the requirements of Appendix G to 10CFR50 even though the end-of-life adjusted reference temperature may exceed 2 0 0°F. By applying the procedures of Appendix G to ASME Section III, the stress limits including appropriate Code safety margin can be met. Westinghouse believes that Figure 2 of the Regulatory Guide is incorrect since upper shelf energy for the 6-inch thick ASTM A302B reference correlation monitor material reported by Hawthorne indicates essentially a constant upper 19 2 shelf as fluences above approximately 1.0 x 10 n/cm (3). Concerning the Guide definition of upper shelf energy, Westinghouse believes, on the basis of extensive experience, that a curve should be fit to the existing data using best engineering judgment. Normally, at least three specimens would be included above the upper end of transition region; additional specimens would be included when the shelf level appears to be marginal. As an alternative to Regulatory Guide 1. 99, operating limits for Unit 2 have been determined by using the current radiation damage curves as shown on Figure 3A-1. These limits are provided in the Technical Specifications. 3A-44 SGS-UFSAR Revision 6 February 15, 1987
References
1. Letter to the Secretary of the Commission by C. E. Eicheldinger, NS-CE-748, Figure 1, September 22, 1975.

2. C.Z. Serpan, Jr., and Hawthorne, J. R., "Yankee Reactor Pressure Vessel Surveillance Notch Ductility Performance of Vessel Steel and Maximum Service Fluence Determined from Exposure During Cores II, III, and IV,"
NRL Report 6616, September 29, 1967.
3. Hawthorne, J. R., "Radiation Effects Information Generated on the ASTM Reference Correlation- Monitor Steels," to be published.
Regulatory Guide 1.100 - SEISMIC QUALIFICATION OF ELECTRICAL EQUIPMENT FOR NUCLEAR POWER PLANTS The Unit 2 design does not fully comply with the Regulatory Guide which endorses IEEE Standard 344-1975, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations." Safety-related electric equipment for Salem Unit 2 (other than Westinghouse supplied NSSS electric equipment) was seismically tested or analyzed based on IEEE Standard 344-1971. Regulatory Guide 1.101 - EMERGENCY PLANNING AND PREPAREDNESS FOR NUCLEAR POWER PLANTS Salem conforms to Regulatory Guide 1.101, Revision 3, August 1992, and used as the planning basis "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", NUREG-0654/FEMA REP.-1, Rev. 1 (November 1980); and "Methodology for Development of Emergency Action Levels", NUMARC/NESP-007. The Emergency Plan Manuals, as revised, describe the total emergency program as described in Section 13.3. 3A-45 SGS-UFSAR Revision 16 January 31, 1998
THIS PAGE INTENTIONALLY LEFT BLANK 3A-46 SGS-UFSAR Revision 16 January 31, 1998
Regulatory Guide 1.102 - FLOOD PROTECTION FOR NUCLEAR POWER PLANTS (Rev. 1) Flood protection, as described in Section 3.4, conforms to this Regulatory Guide. Regulatory Guide 1.105 - INSTRUMENT SETPOINTS The instrument range, accuracy, and set points for the protection system at Salem Station were determined on the bases of system design, accident analyses, and Technical Specification requirements. Unit 2 meets the intent and is in general conformance with the Regulatory Guide subject to the following:
1. The Technical Specifications establish a setpoint and an allowable value providing a set margin allowance for inaccuracy of the instrument, calibration uncertainties, and instrument drift. Periodic functional tests and checks specified in the Technical Specifications are expected to detect any instrument drift exceeding the Technical Specification margins that may occur during the interval between required calibrations.

2. Instrument range is based on the span necessary for the instrument to perform its intended function.

3. The necessary qualifications testing of selected instruments to verify instrument performance and accuracy under adverse conditions was performed in general agreement with the requirements of IEEE Standard 323-1971, "IEEE Trial-Use Standard Guide for Qualifying Class 1 Electric Equipment for Nuclear Power Generating Stations," instead of IEEE Standard 323-1974 as recommended in Regulatory Guide 1.89.

4. Control of setpoints is provided through a combination of any of the following: administrative procedures, cabinet/panel locks, and setpoint locking devices.

5. The setpoints and the accuracy of the instruments are adequate to meet the assumptions used in the safety analyses. The Technical Specifications provide sui table margins to account for the i terns in Regulatory Position 1. Documentation of drift rates and their relationship to testing intervals is not provided in the setpoint analyses.
3A-47 SGS-UFSAR Revision 26 May 21, 2012
Regulatory Guide 1.106- THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR-OPERATED VALVES The regulatory position requires ensuring that safety-related motor operated valves whose motors are equipped with thermal overload protection devices integral with the motor starter will perform their function. Two options are provided under regulatory position 1 and regulatory position 2. Regulatory Position 1 provides the option to bypass the thermal overload protection (a) continuously (except during testing) or (b) under accident conditions. This option can be implemented provided that the completion of the safety function is not jeopardized or that other safety systems are not degraded. Regulatory Position 2 provides the option to keep the thermal overload in service provided the trip setpoint of the thermal overload protection device is established with all uncertainties resolved in favor of completing the safety-related action and the thermal overload protection device should be periodically tested. The MOV control circuit original design bypassed the Thermal Overload Relay (TOR) contacts in order to comply with Regulatory Guide 1.106, position 1. In the mid 1990's, modifications were performed to the MOV control circuits to re-establish the TOR. The sizing of the Thermal Overload Heater (TOH) for the safety related MOVs is based on IEEE standard 741 guidance. Periodic testing of the TOR and TOH has been established. Current thermal overload protection for safety-related motors at Salem is in accordance with RG 1. 106 Regulatory Position 2. Regulatory Guide 1.108 - PERIODIC TESTING OF DIESEL GENERATORS USED AS ONSITE (Revision 1) ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS The Unit 2 design meets the intent of the Regulatory Guide. Periodic testing requirements for the diesel-generators will be identified in the Technical Specifications. 3A-48 SGS-UFSAR Revision 26 May 21, 2012
The Unit 2 initial preoperational test program is in conformance with Regulatory Guide 1.108 with the following exceptions: Paragraph c2.a(4) - Compliance with the section will be by tripping the diesel output breaker at 2750 kW (2000 hour rating) verifying that the voltage regulation is maintained within the acceptable limits and the allowable overspeed does not trip out the diesel. We feel this transient is more severe than the load shedding requirements identified in the Regulatory Guide. Paragraph c. 2a ( 5) - The test described in this section will be performed, but due to the sequence of testing it may not be immediately after the test described in c.2.a(3) The generator systems will, however, be at full load temperatures. Paragraph c.2.a(6) - The station is not designed to perform the test described in this section. Paragraph c.2.a(9) To accomplish this reliability demonstration, the frequency of surveillance testing will be increased to acquire the 23 consecutive successful valid starts per diesel prior to proceeding beyond the Zero Power Physics Test Program. Credit will be taken for those diesel starts accomplished to date or scheduled during Integrated Safeguards Testing, as long as the diesels are loaded to a minimum of approximately 25 percent and the run durations are approximately 30 minutes or more. All additional starts will comply with the Regulatory Guide criteria for valid tests. Regulatory Guide 1.114 - GUIDANCE ON BEING OPERATOR AT THE CONTROLS OF (Revision 2) CONTROLS OF NUCLEAR POWER PLANT Salem Station conforms with the intent of this Regulatory Guide. 3A-49 SGS-UFSAR Revision 6 February 15, 1987
Regulatory Guide 1.115 - PROTECTION AGAINST LOW TRAJECTORY TURBINE MISSILES (Revision 1) The Unit 2 design conforms with the intent of the Regulatory Guide. Regulatory Guide 1.117 -TORNADO DESIGN CLASSIFICATION The Unit 2 design conforms to the requirements of the Regulatory Guide. Specific details of the missile protection methods used for various Unit 1 and Unit 2 components are discussed in Section 3.5.2.2. Regulatory Guide 1.118 - PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS Periodic testing requirements are identified in the Technical Specifications. Regulatory Guide 1.121 - BASIS OF PLUGGING DEGRADED PWR STEAM GENERATOR TUBES The Salem Station design conforms to the intent of this Regulatory Guide. Regulatory Guide 1.124 - DESIGN LIMITS, LOADING COMBINATIONS, AND (Revision 1) SUPPLEMENTARY CRITERIA FOR CLASS I LINEAR-TYPE COMPONENT SUPPORTS Component support design conforms to the intent of the Regulatory Guide. Specific provisions to determine allowable stress increase factors for faulted conditions are used. Regulatory Guide 1.127 - INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS This Regulatory Guide does not apply to the Salem design. 3A-50 SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 1.130 - DESIGN LIMITS AND LOADING COMBINATIONS FOR CLASS I PLATE-AND-SHELL-TYPE COMPONENT SUPPORTS The Unit 2 methodology and criteria for the design of the Reactor Coolant System piping and equipment supports is discussed in Sections 3. 7, 3. 9, and
5. 5.
Regulatory Guide 1.137 - FUEL-OIL SYSTEMS FOR STANDBY DIESEL GENERATORS, 10/79 Diesel fuel oil sampling is performed as follows:
a. A fuel oil sample is taken from each truck delivering fuel oil to Salem whenever possible. However, if several trucks arrive at once, a minimum of 1 in 4 trucks is sampled.

b. All newly received fuel oil is pumped into the 20,000 barrel Fuel Oil Storage Tank. Fuel oil in this tank is sampled at least once every 30 days.

c. A small percentage of the fuel oil in the 20,000 barrel tank is introduced into the diesel fuel oil storage system as necessary. This small percentage is added infrequently to the four 30,000 gallon Diesel Fuel Oil Storage Tanks (two for each unit) as necessary to maintain the minimum level above the 23,000 gallon limit in each Diesel Fuel Oil Storage Tank as specified by the Salem Technical Specifications.

d. Fuel oil in the four 30,000 gallon Diesel Fuel Oil Storage Tanks is sampled as required by the Salem Technical Specifications.
3A-51 SGS-UFSAR Revision 16 January 31, 1998
e. All fuel oil samples taken in actions a through d are sent to an independent lab ora tory within 48 hours of the time the sample is taken. The analysis performed is consistent with Regulatory Guide 1.137 and ASTM D975-77 and the analysis report is submitted to the Salem Station within 30 days of receipt of the sample at the laboratory.

f. All fuel oil deliveries, samples taken, and related analysis reports are logged at the station. When reports indicate that fuel oil quality is not within acceptable limits, station management will take appropriate action to restore it to within acceptable limits.

g. Actions a through f are subject to verification during routine monitoring and audits of the fuel oil program and procedures conducted by NQA personnel.
Regulatory Guide 1.141 - CONTAINMENT ISOLATION PROVISIONS FOR FLUID SYSTEMS The Salem Station design utilizes the Regulatory Guide for guidance in acceptable containment isolation design configurations as detailed in Section 3.0 in ANSI Standard N271-1976. Specific guidance is provided in this Section for those penetration designs which fall under the heading of "Other Defined Basis" as detailed in ANSI Standard N271-1976. Regulatory Guide 1.143 - DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT - WATER - COOLED NUCLEAR POWER PLANTS In accordance with guidance provided in this Regulatory Guide, the Contaminated Floor and Equipment Drain Systems and small portions of the Liquid Waste Disposal System that are designated with Piping Schedule 53D (Piping Specification SPS53) have been reclassified to be Non-Nuclear (Quality Group D). The Salem Station design meets the intent of the Regulatory Guide. Augmented quality assurance requirements have been imposed to ensure that the quality level recommended in the Regulatory Guide is maintained. 3A-51a SGS-UFSAR Revision 16 January 31, 1998
Regulatory Guide 1.144 - "AUDITING QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS", 9/80 (endorses N45.2.12) NRC Regulatory Guide 1.144 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994. Regulatory Guide 1.145 - "ATMOSPHERIC DISPERSION MODELS FOR POTENTIAL ACCIDENT CONSEQUENCE ASSESSMENTS AT NUCLEAR POWER PLANTS", 11/82 (reissued 2/83) The atmospheric dispersion calculations are consistent with the Regulatory Guide, as described in Section 2.3. Regulatory Guide 1.146- "QUALIFICATION OF QUALITY ASSURANCE PROGRAM AUDIT PERSONNEL FOR NUCLEAR POWER PLANTS", 8/80 (endorses N45.2.23) NRC Regulatory Guide 1.146 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994. Branch Technical Position APCSB 9.5-1, Appendix A, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976". The QA Program is applied to the Fire Protection Program to an extent consistent with the requirements of Section C of Appendix A to Branch Technical Position APCSB 9.5-1. Regulatory Guide 1.155 - "Station Blackout", August 1988. Salem Station complies with Regulatory Guide 1.155 as described in Section 3.12. 3A-51b SGS-UFSAR Revision 23 October 17, 2007
Regulatory Guide 8.8 - INFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT NUCLEAR POWER STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) It is the philosophy of management that in the design and operation of the Salem Station, personnel exposure to radiation be kept to ALARA. Regulatory Guide 8. 8 will be followed when applicable and practical as judged by the personnel responsible for the ALARA program. The Salem Station design is in compliance with Regulatory Guide 8. 8 to the extent that a great deal of consideration was given to reducing radiation exposures and levels to ALARA by design, layout, and selection of materials. Extensive radiation shielding has been installed and a detailed radiation contamination survey program has been established to verify that the radiation levels are within the specified design limits. Periodic radiation surveys are taken to assure that the radiation levels are within the specified limits and to identify potential problem areas. Maximum equipment and compartment dose rates have been calculated and shielding installed to reduce radiation levels to the designated zone limits for the particular area. Separation of components by shielding, the use of reach rods, and separate valve compartments have been used to reduce exposures for operation and maintenance of the unit. Labyrinth compartment entrances and locked wire mesh doors are used in addition to administrative procedures to minimize exposure from high radiation sources. Pipes carrying radioactive fluids are routed through shielded pipe chases or hot equipment compartments to maintain low radiation levels in corridors and aisles. Instrument transmitters and direct readout radiation monitors are generally located in low radiation areas. A balanced ventilation system and floor and equipment drains are used to handle airborne contamination and radioactive liquids, respectively. Other design features employed to maintain ALARA dose rates and exposures include shielded remote filter handling equipment, decontamination room, back flush connection on major radioactive system piping, and automatic resin and sludge drumming and solidification. Local sample connections for most samples have been moved from hot compartments to lower radiation areas. Local portable shielding is available for areas where extended maintenance may have to be performed in high radiation areas. In addition to these design features, administrative procedures are employed for personnel exposure monitoring and reducing or maintaining low in-plant radiation exposures. 3A-52 SGS-UFSAR Revision 16 January 31, 1998
The Plant Manager has responsibility at the station level for the establishment and execution of the station Radiological Protection Program. He implements the Program to meet the requirements of 10CFR20, Regulatory Guides, and the Technical Specifications. The Radiation Protection Manager (RPM) is responsible for implementation of the Nuclear Business Unit Radiation Protection Program at the Salem and Hope Creek generating stations. The RPM reports to the Plant Manager in matters of radiation protection and ALARA. The RPM maintains liaison with other NBU departments and upper management regarding the program, and directs the Radiation Protection Program through the Radiation Protection Superintendents. The RPM is the member of the NBU management who is responsible for the ALARA program, and reviews the program to determine how personnel dose may be reduced. The Radiation Protection Supervisors are responsible to the RPM for the day-to-day functions in the ALARA, Operations and Support areas. The RPM (or his designate):
1. Ensures ALARA pre-job planning of major work items that are estimated to give large man-rem exposures

2. Establishes an effective Radiation Work Fermi t Program to control individual exposures

3. Ensures that adequate controls are in place to prevent the uncontrolled release of radioactive materials

4. Performs periodic formal assessments of the content and implementation of the NBU ALARA program including recommendations for improvements.

5. Ensures that ALARA requirements and concepts are appropriately included in the NBU ALARA program.
3A-53 SGS-UFSAR Revision 21 December 6, 2004
Design and equipment changes are reviewed by the Plant Operations Review Committee (PORC). A Radiation Protection Program has been established and implemented for both Salem and Hope Creek. Procedures are written and approved covering such subjects as access control, radioactive material control, dosimetry, surveys, and radiation exposure permits. Through the use of these established procedures, and with guidance and supervision by the Radiation Protection Superintendents and Supervisors, work will be performed in accordance with ALARA philosophy. The typical portable radiation survey instruments are described in Section 12. The criteria for selection of the portable instruments include:
1. Ability of instrument to perform in its intended use with reliability and accuracy

2. Ease of calibration and repair

3. Interchangeability of components 3A-54 SGS-UFSAR Revision 24 May 11, 2009

4. Weight and size for user acceptance

5. Standard readouts and controls/adjustments to simplify training of users The installed radiation monitoring instrumentation and locations are described in Sections 11 and 12.
The access control for the areas where radiation and contamination are possible is described in Section 12. The single point of access to and exit from the controlled area is located at Elevation 100 feet (ground level) in the Service Building. The traffic pattern for male and female workers is basically the same with the exception of the locker room/change area. The facilities available and used to provide proper radioactive material and contamination control are established and are common to both units. REGULATORY GUIDE 1.181 - Content Of The Updated Final Safety Analysis Report in Accordance With 10 CFR 50.71(e) The PSEG Nuclear procedures for updating the UFSAR are based on NEI 98-03, Revision 1, which is endorsed by Regulatory Guide 1.181. The purpose of NEI 98-03 is to provide licensees with guidance for updating their FSARs consistent with the requirements of 10CFR50.71(e). Guidance is also provided for making voluntary modifications to UFSARs (i.e., removal, reformatting and simplification of information, as appropriate) to improve their focus, clarity and maintainability. REGULATORY GUIDE 1.183 -ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS The assumptions used are in agreement with the Regulatory Guide, as described in Section 15. 3A-55 SGS-UFSAR Revision 28 May 22, 2015
TABLE 3A-l This Table intentionally deleted 1 of 1 SGS-UFSAR Revision 28 May 22, 2015
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1: 'i (* I' I* 10 B 6 1019 2 10 zo FLUENCE(N/CM2>1 MEV) REVISION8 FEBRUARY15,1117 Effect of Fluence and Copper Content on Shift PUBLIC SERVICE ELECTRIC AND GAS COMPANY of RT NOT for Reactor Vessel Exposed to 550°F SALEM NUCLEAR GENERATING STATION Updated FSAR Figure3A*1
Adrnfnistrntion Builc.ling Turbine - Cenerator Building Service Building Unit 2 Unit 1 Containment Containment Auxiliary Building Controlled Facilities Building cC=1 REVISION I FEBRUARY15. 1187 PUBLIC SERVICE ELECTRIC AND GAS COMPANY Building Location Key SALEM NUCLEAR GENERATING STATION Updated FSAR figure 3A*2
Figurehasbeendeleted Auxiliary Building PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3A-3 Revision16 January31, 1998
Figurehasbeendeleted Controlled Facilities Building Elevation 113 PUBLICSERVICEELECTRICAND GAS COMPANY SALEMNUCLEARGENERATING STATION UpdatedFSAR Figure3A--4 Revision16 January 31, 1998
SECTION 4 REACTOR TABLE OF CONTENTS Section Title Page 4.1
SUMMARY
DESCRIPTION 4.1-1 4.1.1 Reference for Section 4.1 4.1-4 4.2 MECHANICAL DESIGN 4.2~1 4.2.1 Fuel 4.2-2 4.2.1.1 Design Bases 4.2-2 4.2.1.1.1 Fuel Rods 4.2-3 4.2.1.1.2 Fuel Assembly Structure 4.2-5 4.2.1.2 Design Description 4.2-8 4.2.1.2.1 Fuel Rods 4.2-10 4.2.1.2.2 Fuel Assembly Structure 4.2-11 4.2.1.3 Design Evaluation 4.2-14 4.2.1.3.1 Fuel Rods 4.2-14 4.2.1.3.2 Fuel Assembly Structure 4.2-24 4.2.1.3.3 Operational Experience 4.2-28 4.2.1.3.4 Test Rod and Test Assembly Experience 4.2-28a 4.2.1.4 Testing and Inspection Plan 4.2-28a 4.2.1.4.1 Quality Assurance Program 4.2-28a 4.2.1.4.2 Quality Control 4.2-29 4.2.1.4.3 Onsite Inspection 4.2-32 4.2.2 Reactor Vessel Internals 4.2-33 4.2.2.1 Design Bases 4.2-33 4.2.2.2 Description and Drawings 4.2-34 4.2.2.3 Design Loading Conditions 4.2-40a 4.2.2.4 Design Loading Categories 4.2-42 4.2.2.5 Design Criteria Basis 4.2-43 4.2.3 Reactivity Control System 4.2-44 4-i SGS-UFSAR Revision 11 July 22, 1991
TABLE OF CONTENTS (Cont) Section 4.2.3.1 Design Bases 4.2-44 4.2.3.1.1 Design Stresses 4.2-44 4.2.3.1.2 Material Compatibility 4.2-45 4.2.3.1.3 Reactivity Control Components 4.2-45 4.2.3.1.4 Control Rod Drive Mechanisms 4.2-48 4.2.3.2 Design Description 4.2-49 4.2.3.2.1 Reactivity Control Components 4.2-51 4.2.3.2.2 Control Rod Drive Mechanism 4.2-56 4.2.3.3 Design Evaluation 4.2-63 4.2.3.3.1 Reactivity Control Components 4.2-63 4.2.3.3.2 Control Rod Drive Mechanism 4.2-73 4.2.3.4 Tests, Verification, and Inspections 4.2-77 4.2.3.4.1 Reactivity Control Components 4.2-77 4.2.3.4.2 Control Rod Drive Mechanism 4.2-79 4.2.4 References for Section 4.2 4.2-81 4.3 NUCLEAR DESIGN 4.3-1 4.3.1 Design Bases 4.3-1 4.3.1.1 Fuel Burnup 4.3-2 4.3.1.2 Negative Reactivity Feedbacks (Reactivity Coefficient) 4.3-3 4.3.1.3 Control of Power Distribution 4.3-4 4.3.1.4 Maximum Controlled Reactivity Insertion Rate 4.3-6 4.3.1.5 Shutdown Margins 4.3-7 4.3.1.6 Stability 4.3-8 4.3.1. 7 Anticipated Transients Without Trip 4.3-9 4.3.2 Description 4.3-10 4.3.2.1 Nuclear Design Description 4.3-10 4.3.2.2 Power Distribution 4.3-13 4.3.2.2.1 Definitions 4.3-13 4.3.2.2.2 Radial Power Distribution 4.3-16 4*11 SGS-UFS.AR Revision 6 February 15, 1987
TABLE OF CONTENTS (Cont) Section 4.3.2.2.3 Assembly Power Distribution 4.3-17 4.3.2.2.4 Axial Power Distribution 4.3-17 4.3.2.2.5 Deleted 4.3-18 4.3.2.2.6 Limiting Power Distribution 4.3-20 4.3.2.2.7 Experimental Verification of Power Distribution Analysis 4.3-26 4.3.2.2.8 Testing 4.3-29 4.3.2.2.9 Monitoring Instrumentation 4.3-29 4.3.2.3 Reactivity coefficients 4.3-30 4.3.2.3.1 Fuel Temperature (Doppler) Coefficient 4.3-30 4.3.2.3.2 Moderator coefficients 4.3-31 4.3.2.3.3 Power Coefficient 4.3-34 4.3.2.3.4 comparison of Calculated and Experimental Reactivity Coefficients 4.3-34 4.3.2.3.5 Reactivity Coefficients Used in Transient Analysis 4.3-35 4.3.2.4 Control Requirements 4.3-35 4.3.2.4.1 Doppler 4.3-36 4.3.2.4.2 variable Average Moderator Temperature 4.3-36 4.3.2.4.3 Redistribution 4.3-37 4.3.2.4.4 Void content 4.3-37 4.3.2.4.5 Rod Insertion Allowance 4.3-37 4.3.2.4.6 Burn up 4.3-38 4.3.2.4.7 Xenon and Samarium Poisoning 4.3-38 4.3.2.4.8 pH Effects 4.3-38 4.3.2.4.9 Experimental Confirmation 4.3-39 4.3.2.5 control 4.3-39 4.3.2.5.1 Chemical Poison 4.3-39 4.3.2.5.2 Rod Cluster Control Assemblies 4.3-40 4.3.2.5.3 Burnable Absorbers 4.3-41 4.3.2.5.4 Peak xenon Startup 4.3-41 4.3.2.5.5 Load Follow Control and Xenon control 4.3-42
SGS-UFSAR 4-iii Revision 17 october 16, 1998
TABLE OF CONTENTS (Cont) Section Title Page 4.3.2.5.6 Burnup 4.3-42 4.3.2.6 Control Rod Patterns and Reactivity Worth 4.3-42 4.3.2.7 Criticality of Fuel Assemblies 4.3-45 4.3.2.8 Stability 4.3-46 4.3.2.8.1 Introduction 4.3-46 4.3.2.8.2 Stability Index 4.3-46 4.3.2.8.3 Prediction of the Core Stability 4.3-47 4.3.2.8.4 Stability Measurements 4.3-48 4.3.2.8.5 Comparison of Calculations with Measurements 4.3-50 4.3.2.8.6 Stability Control and Protection 4.3-51 4.3.2.9 Vessel Irradiation 4.3-52 4.3.3 Analytical Methods 4.3-53 4.3.3.1 Fuel Temperature (Doppler) Calculations 4.3-54 4.3.3.2 Macroscopic Group Constants 4.3-55 4.3.3.3 Spatial Few-Group Diffusion Calculations 4.3-58 4.3.3.4 Pin Power Reconstruction 4.3-60 4.3.4 References for Section 4.3 4.3-60 4.4 THERMAL AND HYDRAULIC DESIGN 4.4-1 4.4.1 Design Bases 4.4-1 4.4.1.1 Departure From Nucleate Boiling Design Basis 4.4-2 4.4.1.2 Fuel Temperature Design Basis 4.4-2a 4.4.1.3 Core Flow Design Basis 4.4-3 4.4.1.4 Hydrodynamic Stability Design Bases 4.4-4 4.4.1.5 Other Considerations 4.4-4 4.4.2 Description 4.4-5 4.4.2.1 Summary Comparison 4.4-5 4.4.2.2 Fuel Cladding Temperatures (Including Densification) 4.4-7 4.4.2.2.1 Uranium Dioxide Thermal Conductivity 4.4-9 4-iv SGS-UFSAR Revision 31 December 5, 2019
TABLE OF CONTENTS (Cent) Section Title 4.4.2.2.2 Radial Power Distribution in UO Fuel Rods 4.4-10 4.4.2.2.3 Gap Conductance 4.4-10 4.4.2.2.4 Surface Heat Transfer Coefficients 4.4-12 4.4.2.2.5 Fuel Clad Temperatures 4.4-12
.4. 4. 2. 3 Critical Heat Flux Ratio or Departure from Nucleate Boiling Ratio and Mixing Technology 4.4-13 4.4.2.3.1 Departure from Nucleate Boiling Technology 4.4-14 4.4.2.3.2 Definition of Departure from Nucleate Boiling Ratio 4.4-lSa 4.4.2.3.3 Mixing Technology 4. 4-17a 4.4.2.3.4 Engineering Hot-Channel Factors 4.4-19 4.4.2.3.5 Effects of Rod Bow on DNBR 4.4-21a 4.4.2.3.6 Transition Core DNB Methodology 4.4-21a 4.4.2.4 Flux Tilt Considerations 4.4-22 4.4.2.5 Void Fraction Distribution 4.4-22
4. 4. 2. 6 Core Coolant Flow Distribution 4.4-23

4. 4. 2. 7 Core Pressure Drops and Hydraulic Loads 4.4-23 4.4.2.7.1 Core Pressure Drops 4.4-23 4.4.2.7.2 Hydraulic Loads 4.4-24

4. 4. 2. 8 Correlation and Physical Data 4.4-25 4.4.2.8.1 Surface Heat Transfer Coefficients 4.4-25 4.4.2.8.2 Total Core apd Vessel Pressure Drop .4. 4-26 4.4.2.8.3 Void Fraction Correlation 4.4-27

4. 4. 2. 9 Thermal Effects of Operational Transients 4. 4-28 4.4.2.10 Uncertainties in Estimates 4.4-29 4.4.2.10.1 Uncertainties in Fuel and Clad Temperatures 4.4-29 4.4.2.10.2 Uncertainties in Pressure Drops 4.4-30 4.4.2.10.3 Uncertainties Due to Inlet Flow Maldistribution 4.4-30 4.4.2.10.4 Uncertainty in DNB Correlation 4.4-30
SGS-UFSAR 4-v Revision 18 April 26, 2000
TABLE OF CONTENTS (Cant) Section Title 4.4.2.10.5 Uncertainties in DNBR Calculations 4.4-30 4.4.2.10.6 Uncertainties in Flow Rates 4.4-31 4.4.2.10.7 Uncertainties in Hydraulic Loads 4.4-31 4.4.2.10.6 Uncertainty in Mixing Coefficients 4.4-32 4.4.2.11 Plant Configuration Data 4.4-32 . 4. 4.3 Evaluation 4.4-33 4.4.3.1 Core Hydraulics 4.4-33 4.4.3.1.1 Flow Paths Considered in Core Pressure Drop and Thermal Design 4.4-33 4.4.3.1.2 Inlet Flow Distributions 4.4-34 4.4.3.1.3 Emprical Friction Factor Correlations 4.4-35 4.4.3.2 Influence of Power Distribution 4.4-36 4.4.3.2.1 Nuclear Enthalpy Rise Hot Channel Factor 4.4-37 4.4.3.2.2 Axial Heat Flux Distributions 4.4-38 4.4.3.3 Core Ther.mal Response 4.4-38 4.4.3.4 Analytical Techniques 4.4-39 4.4.3.4.1 Core Analysis 4.4-39 4.4.3.4.2 Fuel Temperatures 4.4-47 4.4.3.4.3 Hydrodynamdc Instability 4.4-47 4.4.3.5 Hydrodynamic and Flow Power Coupled Instability 4.4-47 4.4.3.6 Temperature Transient Effects Analysis 4.4-50 4.4.3.7 Potentially Damaging Temperature Effects During Transients 4.4-51 4.4.3.8 Energy Release During Fuel Element Burnout 4.4-52 4.4.3.9 Energy Release or Rupture of Water-logged Fuel Elements 4.4-53 4.4.3.10 Fuel Rod Behavior Effects from Coolant Flow Blockage 4.4-53 4.4.4 Testing and Verification 4.4-55 4.4.4.1 Tests Prior to Initial Criticality 4.4-55 SGS-UFSAA 4-vi Revision 6 February 15, 1987
TABLE OF CONTENTS (Cont) Section Title 4.4.4.2 Initial Power and Plant Operation 4.4-55 4.4.4.3 Component and Fuel Inspections 4.4-56 4.4.4.4 Auq.mented Startup Test Program 4.4-56 4.4.5 References for Section 4.4 4.4-56 4.5 RELOAD ANALYSIS 4.5-1 4.5.1 References for Section 4.5 4.5-2
SGS-UFSAR 4-vii Revision 17 October 16, 1998
LIST OF TABLE:S 4.1-1 Thermal and Hydraulic Design 4.1-2 Analytic Techniques Incore Design 4.1-3 Design Loading Conditions for Reactor Core Components 4.2-l Maximum Deflections Allowed for Reactor Internal Support Structures 4.2-2 Comparison of Single and Double Encapsulated Secondary Source Designs 4.3-1 Reactor Core Description 4.3-2 Nuclear Design Parameters 4.3-3 Reactivity Requirements for Rod Cluster Control Assemblies
4. 3-4 Axial Stability Index-PWR Core With a 12-Foot Height 4.3-5 Typical Neutron Flux Levels at Full Power
~.3-6 Comparison of Measured and Calculated Doppler Defects 4.3-7 Benchmark Critical Experiments SGS-UFSAR 4-viii Revision 18 April 26, 2000
LIST OF TABLES (Cont) 4.3-8 Saxton Core II Isotopics, Rod MY, Axial Zone 6 4.3-9 Critical Boron Concentrations, BOL 4.3-10 Comparison of Measured and Calculated Rod Worth
4. 3-11 Comparison of Measured and Calculated Moderator Coefficients at HZP, BOL 4.4-1 Reactor Thermal and Hydraulic Design Parameters 4.4-2 (This text has been deleted) 4.4-3 Void Fractions at Nominal Reactor Conditions With Design Hot Channel Factors 4.4-4 Comparison of THINC-IV and THINC-I Predictions With Data From Representative Westinghouse Two and Three Loop Reactors 4.5-1 (This text has been deleted) 4.5-2 (This text has been deleted)
SGS-UFSAR 4-ix Revision 23 October 17 1 2007
LIST OF FIGURES Figure 4.2-1 Fuel Assembly Cross Section - 17 x 17 4.2-2 Standard Fuel Assembly Outline - 17 x 17 4.2-2A 17 x 17 Vantage+/Vantage SH Fuel Assembly Comparison 4.2-28 17 x 17 Standard Robust Fuel Assembly Outline I 4.2-2C 17 x 17 RFA ZIRLOl'M+2 Outline 4.2-3 Standard Fuel Rod Schematic 4.2-3A 17 x 17 Vantage+/Vantage 5H Fuel Rod Comparison 4.2-38 17 x 17 Standard RFA Fuel Rod Schematic I 4.2-3C 4.2-4 17 x 17 RFA ZIRL0'+2 Fuel Rod Schematic Typical Clad and Pellet Dimensions as a Function of Exposure 4.2-5 Plan View 4.2-6 Representative Fuel Rod Internal Pressure and Linear Power Density for the Lead Burnup Rod as a Function of Time 4.2-7 Top Grid to Nozzle Attachment 4.2-8 Lower Core Support Assembly 4.2-9 Elevation View, Grid to Thimble Attachment 4.2-10 Upper Core Support Structure
4. 2-11 Guide Thimble to Bottom Nozzle Joint 4.2-12 Plan View of Upper Core Support Structure 4.2-13 Full Length Rod Cluster Control and Drive Rod Assembly With Interfacing Components SGS-OFSAR 4-x Revision 19 November 19, 2001
LIST OF FIGURES (Cant) Title 4.2-14 Full Length Rod Cluster Control Assembly Outline 4.2-15 Full Length Absorber Rod 4.2-16 Burnable Absorber Assembly 4.2-17 Pyrex Burnable Poison Rod Cross Section 4.2-17A WABA Rod Cross Section 4.2-18 Primary Source Assembly 4.2-19 Single Encapsulated Secondary Source 4.2-20 Thimble Plug Assembly (Optional Usage} 4.2-21 Full Length Control Rod Drive Mechanism 4.2-22 Full Length Control Rod Drive Mechanism Schematic 4.2-23 Nominal Latch Clearance at Minimum and Maximum Temperature 4.2-24 Control Rod Drive Mechanism Latch Clearance Thermal Effect 4.2-25 Schematic Representation of Reactor Core Model 4.3-1 Production and Consumption of Higher Isotopes 4.3-2 Boron Concentration vs Cycle Burnup (Typical) 4.3-3 Normalized Power Density Distribution At BOL, Unrodded Core, HFP, No Xenon (Typical) 4.3-4 Normalized Power Density, Distribution Near BOL, Unrodded Core, HFP, Equilibrium Xenon (Typical) 4-xi SGS-UFSAR Revision 18 April 26, 2000
LIST OF FIGURES (Cont) Figure 4.3-5 Normalized Power Density Distribution Near BOL, Group D Inserted, HFP, Equilibrium Xenon (Typical) 4.3-6 Normalized Power Density Distribution Near Middle of Life, Unrodded Core, HFP, Equilibrium Xenon (Typical) 4.3-7 Normalized Power Density Distribution Near EOL, Unrodded Core, HFP, Equilibrium Xenon (Typical) 4.3-8 Typical Axial Power Shapes Occurring at Beginning of Life 4.3-9 Typical Axial Power Shapes Occurring at Middle of Life 4.3-10 Typical Axial Power Shapes Occurring at End of Life
4. 3-11 Maximum F0 -Power vs Axial Height During Normal Operation 4.3-12 Peak Power Density During Control Rod Malfunction Overpower Transients 4.3-13 Peak Linear Power During Boration/Dilution Overpower Transients

4. 3-14 Comparison Between Calculated and Measured Relative Fuel Assembly Power Distribution 4.3-15 Comparison of Calculated and Measured Axial Shape 4.3-16 Measured Values of FQ for Full Power Rod Configuration 4.3-17 Doppler Temperature Coefficient at BOL and EOL, (Typical) 4.3-18 Doppler Power Coefficient-SOL, MOL, EOL, (Typical) 4.3-19 Doppler Power Defect-BOL, MOL, EOL, (Typical) 4.3-20 Moderator Temperature Coefficient-SOL, ARO (Typical) 4-xii SGS-UFSAR Revision 17 October 16, 1998
LIST OF FIGURES (Cont) Figure Title 4.3-21 Moderator Temperature Coefficient-EOL, ARO (Typical) 4.3-22 Moderator Temperature Coefficient as a Function of Boron Concentration, BOL, ARO (Typical) 4.3-23 Hot Full Power Moderator Temperature Coefficient vs Cycle Burnup (Typical) 4.3-24 Total Power Coefficient-SOL, EOL (Typical) 4.3-25 Total Power Defect-SOL, EOL (Typical) 4.3-26A Rod Cluster Control Assembly Pattern - Unit 1 4.3-26B Rod Cluster Control Assembly Pattern - Unit 2 4.3-27 Accidental Simultaneous Withdrawal of 2 Control Banks EOL, HZP, Banks D&B Moving in the Same Plane 4.3-28 Design Trip Curve 4.3-29 Normalized Rod Worth vs Percent Insertion All Rods But One 4.3-30 Axial Offset vs Time-PWR Core With a 12-Foot Height and 121 Assemblies 4.3-31 XY Xenon Test Thermocouple Response Quadrant Tilt Difference vs Time 4.3-32 Calculated and Measured Doppler Defect and Coefficients at BOL, Two-Loop Plant, 121 Assemblies, 12-Foot Core 4.3-33 Comparison of Calculated and Measured Boron Concentration for 2-Loop Plant, 121 Assemblies, 12-foot Core 4.3-34 Comparison of Calculated and Measured CB 2-Loop with 121 Assemblies, 12-Foot Core 4-xiii SGS-UFSAR Revision 17 October 16, 1998
LIST OF FIGURES (Cant) Figure Title 4.3-35 Comparison of Calculated and Measured ce 3-Loop Plant, 157 Assemblies, 12-Foot core 4.4-1 Peak Fuel Average and Surface Temperatures During Fuel Rod Lifetime vs Linear Power 4.4-lA Peak Fuel Average and Surface Temperatures During Fuel Rod Lifetime vs Linear Power for Vantage-5H fuel 4.4-2 Peak Fuel Centerline Temperature During Fuel Rod Lifetime vs Linear Power 4 ., 4-2A Peak Fuel Centerline Temperatures During Fuel Rod Lifetime vs Linear Power for Vantage-SH fuel 4.4-3 Thermal Conductivity of uo (Data Corrected to 95% Theoretical 2 Density)
4. 4-4 Axial Variation of Average Clad Temperature for Rod Operating at
'5.43 kW/ft 4.4-SA Comparison of Measured to Predicted 17 x 17 DNB Data 4.4-58 Measured vs Predicted Critical Heat Flux WRB-1 Correlation 4.4-SC Measured Critical Heat Flux - WRB-2 Correlation 4.4-6 TDC vs Reynold's Number for 26-Inch Grid Spacing 4.4-7 Normalized Radial Flow and Enthalpy Distribution at 4-Foot Elevation 4.4-8 Normalized Radial Flow and Enthalpy Distribution at 8-Foot Elevation SGS-UFSAR 4-xiv Revision 18 April 26, 2000
LIST OF FIGURES {Cont) Figure 4.4-9 Normalized Radial Flow and Enthalpy Distribution at 12-Foot Elevation (Unit 2) 4.4-10 Void Fraction vs Thermodynamic Quality H-HSAT/Hg-HSAT 4.4-11 PWR Natural Circulation Test 4.4-12 Comparison of a Representative Westinghouse Two-Loop Reactor Incore Thermocouple Measurements With THINC-IV Predictions 4.4-13 Comparison of a Representative Westinghouse Three-Loop Reactor Incore Thermocouple Measurements With THINC-IV Predictions 4.4-14 Hanford Subchannel Temperature Data Comparison With THINC-IV 4.4-15 Hanford Subcritical Temperature Data Comparison With THINC-IV 4.4-16 Distribution of Incore Instrumentation - Unit 1 4.4-17 Distribution of Incore Instrumentation - Unit 2 4.5-1 Typical Salem Unit 1 Loading Pattern 4.5-2 Typical Salem Unit 1 Burnable Absorber Configuration 4.5-3 Typical Salem Unit 2 Loading Pattern 4.5-4 Typical Salem Unit 2 Burnable Absorber Configuration 4-xv SGS-UFSAR Revision 23 October 17, 2007
SECTION 4 REACTOR 4.1
SUMMARY
DESCRIPTION This chapter describes the following: 1) the mechanical components of the reactor and reactor core including the fuel rods and fuel assemblies, reactor internals, and the control rod drive mechanisms, 2) the nuclear design, and 3) the thermal-hydraulic design. The reactor core is comprised of an array of fuel assemblies which are similar in mechanical design and fuel enrichment. The Salem Unit 1 and 2 cores may consist of any combination of fuel designs including Vantage 5H, Vantage+, and standard Robust Fuel Assembly (RFA and RFA-2, which further enhances the anti-fretting characteristics with improved mid grids) as described in Section 4. 2. 1. 2. The most significant difference between the Vantage+ and RFA fuel and the others is the
. TM application of Zlrlo cladding, guide thimble and instrument tubes. The Vantage+
and RFA are modifications of the NRC-approved Vantage 5H fuel assembly design (Reference 1) A detailed description and evaluation of the Vantage+ and RFA features is provided in References 2, 4, 5 and 6. The core is cooled and moderated by light water at a pressure of 2250 psia in the Reactor Coolant System. The moderator coolant contains boron as a neutron absorber. The concentration of boron in the coolant is varied as required to control relatively slow reactivity changes including the effects of fuel burnup. Additional boron, in the form of burnable absorber rods and/ or IFBAs, may be employed in the core to establish the desired initial reactivity. Two hundred and sixty-four fuel rods are mechanically joined in a square array to form a fuel assembly. The fuel rods are supported in intervals along their length by grid assemblies which maintain the lateral spacing between the rods throughout the design life of the assembly. The grid assembly consists of an "egg-crate" arrangement of interlocked straps. The straps contain spring fingers and dimples for fuel rod support as well as coolant mixing vanes. The fuel rods consist of slightly enriched uranium dioxide ceramic cylindrical pellets contained in slightly cold worked Zircaloy-4 or Zirlo TM tubing which is plugged and seal welded at the ends to 4.1-1 SGS-UFSAR Revision 28 May 22, 2015
encapsulate the fuel. All fuel rods are pressurized with helium during fabrication to reduce stresses and strains and to increase fatigue life. In addition, the Zirlorn fuel rods may be oxide coated at the lower end for additional protection against fretting. RFA fuel rods will utilize annular pellets at the top and bottom 6" to provide lower rod internal pressures. The center position in the assembly is reserved for the in-core instrumentation, while the remaining 24 positions in the array are equipped with guide thimbles joined to the grids and the top and bottom nozzles. Depending upon the position of the assembly in the core, the guide thimbles are used as core locations for rod cluster control assemblies, neutron source assemblies, and burnable absorber rods. The remaining guide thimbles may be fitted with plugging devices to limit bypass flow. The use of plugging devices is optional. The bottom nozzle is a box-like structure which serves as a bottom structural element of the fuel assembly and directs the coolant flow distribution to the assembly. The top nozzle assembly functions as the upper structural element of the fuel assembly in addition to providing a partial protective housing for the rod cluster control assembly or other core components. The rod cluster control assemblies each consist of a group of individual absorber rods fastened at the top end to a common hub or spider assembly. These assemblies have rods containing absorber material to control the reactivity of the core under operating conditions. The control rod drive mechanisms are of the magnetic latch type. The latches are controlled by three magnetic coils. They are so designed that upon a loss of power to the coils, the rod cluster control assembly is released and falls by gravity to shut down the reactor. The components of the reactor internals are divided into three parts consisting of the lower core support structure (including 4.1-2 SGS-UFSAR Revision 18 April 26, 2000
the entire core barrel and thermal shield), the upper core support structure and the in-core instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between the fuel assemblies and control rod drive mechanisms, direct coolant flow past the fuel elements and to the pressure vessel head, provide gamma and neutron shielding, and provide guides for the in-core instrumentation. The nuclear design analyses and evaluation establish physical locations for control rods and burnable absorbers, and physical parameters such as fuel enrichments and boron concentration in the coolant such that the reactor core has inherent characteristics which, together with corrective actions of the Reactor Control, Protection and Emergency Cooling Systems provide adequate reactivity control even if the highest reactivity worth rod cluster control assembly is stuck in the fully withdrawn position. The design also provides for inherent stability against diametral and azimuthal power oscillations. The thermal-hydraulic design analyses and evaluation establish coolant flow parameters which assure that adequate heat transfer is assured between the fuel cladding and the reactor coolant. The thermal design takes into account local variations in fuel rod dimensions, power generation, flow distribution, and mixing. The mixing vanes incorporated in the fuel assembly spacer grid design induces additional flow mixing between the various flow channels within a fuel assembly as well as between adjacent assemblies. Instrumentation is provided in and out of the core to monitor the nuclear, thermal-hydraulic, and mechanical performance of the reactor and to provide inputs to automatic control functions. The reactor core design together with corrective actions of the Reactor Control, Protection and Emergency Cooling Systems can meet the reactor performance and safety criteria specified in Section 4.2. 4.1-3 SGS-UFSAR Revision 17 October 16, 1998
To illustrate the effects of the change in fuel design, Table 4.1-1 presents principal nuclear, thermal-hydraulic, and mechanical design parameters for the Salem 17 x 17 Vantage 5H, Vantage+, and RFA fuel assemblies. The effects of fuel densification were evaluated (Reference 3). The analytical techniques employed in the core design are tabulated in Table 4.1-2. The loading conditions considered in general for the core internals and components are tabulated in Table 4.1-3. Specific or limiting loads considered for design purposes of the various components are listed as follows: fuel assemblies in Section 4.2.1.1.2; reactor internals in Section 4.2.2.3 and Table 5.1-10; neutron absorber rods, burnable absorber rods, neutron source rods, and thimble plug assemblies (if used) in Section 4.2.3.1.3; control rod drive mechanisms in Section 4.2.3.1.4. 4.1.1 Reference for Section 4.1
1. Davidson, S.L. (Ed.), et al., "Vantage 5H Fuel Assembly Reference Core Report,"
WCAP-10444-P-A and Appendix A, September 1985; Addendum 2-A, March 1986; Addendum 2-A, April 1988.
2. Davidson, S.L., Nuhfer, D.L. (Eds.), "Vantage+ Fuel Assembly Reference Core Report," WCAP-12610-P-A, April 1995.

3. Hellman, J. M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP-8219-A (Nonproprietary), March 1975.

4. Letter from W. J. Rinkacs (Westinghouse) to M.M. Mannion (PSE&G), "Westinghouse Fuel Features Recommendation for Cycle 11", July 22, 1998.

5. Letter from W. J. Rinkacs (Westinghouse) to M.M. Mannion (PSE&G), "Westinghouse Generic Safety Evaluation for the 17x17 Standard Robust Fuel Assembly", October 1, 1998.

6. Garde, A., et al., "Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO," WCAP-12610-P-A & CENPD-404-P-A, Addendum 2-A, October 2013.
4.1-4 SGS-UFSAR Revision 28 May 22, 2015
TABLE 4.1-1 THERMAL AND HYDRAULIC DESIGN Reactor Core Heat Output, MWt 3459 6 Reactor Core Heat Output, 10 Btu/hr 11,806 (Unit 1) 11,844 (Unit 2) Heat Generated in the Fuel, % 97.4 Nominal System Pressure, psia 2250 Assumed Initial System Pressure for DNB (1) Transients, psia 2218 (STDP ) (2) 2250 (RTDP ) (3) Minimum DNBR for Design Transients V-5H 1.24 (RTDP) RFA (7) 1.24 (RTDP, Typ) RFA 1.22 (RTDP, Thm) (3) DNB Correlation V-5H WRB-1 RFA(7) WRB-2 Coolant Flow 6 (5) Total Thermal Design Flow Rate, 10 lb/hr 125.3 (Unit 1 & 2) Effective Flow Rate for Heat 6 (5) Transfer, 10 lb/hr 116.3 (Unit 1) 115.7(5) (Unit 2) Effective Flow Area for Heat 2 (3) Transfer, ft V-5H 51.3 RFA 51.1 (3) Average Velocity Along Fuel Rods, ft/sec V-5H 14.1 RFA 14.2 6 2 (3) (5) Average Mass Velocity, 10 lb/hr-ft V-5H 2.27 (5) RFA 2.28 Page 1 of 5 SGS-UFSAR Revision 33 October 24, 2022
TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Coolant Temperature (5) Nominal Inlet, °F 542.7 (5) Average Rise in Vessel, °F 70.4 (5) Average Rise in Core, °F 75.2 (5) Average in Core, °F 582.4 (Unit 1) 582.6(5) (Unit 2) (5) Average in Vessel, °F 577.9 Heat Transfer 2 Active Heat Transfer Surface Area, ft 59,700 Heat Flux Hot Channel Factor, F 2.40 Q 2 Average Heat Flux, Btu/hr-ft 192,470 Maximum Heat Flux for Normal Operation, 461,930 2 Btu/hr-ft Average Thermal Output, kW/ft 5.52 Maximum Thermal Output for Normal 13.3 Operation, kW/ft Peak Linear Power for Determination <22.4 of Protection Setpoints, kW/ft Peak Fuel Center Temperature at Maximum Thermal <4700 Output for Maximum Overpower Trip Point, °F Page 2 of 5 SGS-UFSAR Revision 33 October 24, 2022
TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Fuel Assemblies Design RCC Canless Number of Fuel Assemblies 193 uo Rods per Assembly 264 2 Rod Pitch, in 0. 4 96 Overall Dimension, in 8.426 X 8.426 Weight of Fuel (as uo ) in Core, lbs STD, VSH, V+ 222,739 2 RFA (9 ) 217 I 565 Weight of Zircaloy in Core, lbs All STD 50913 All VSH, V+ 52541 All RFA 53847 Number of Grids per Assembly STD 8 Inconel VSH 2 Inconel (Top & Bottom) 6 4 (Mid Grids) V+ 2 Inconel (Top & Bottom) 6 Zirlo' (Mid Grids) RFA 2 Inconel (Top & Bottom) 1 Inconel (Protective Grid) 6 Zirlo' (Mid Grids) 3 Zirlo' (Intermediate Flow Mixing Grids) Loading Technique 3 Region Non-uniform rue.:. Rods Number in Core 50,952 Outside Diameter, in 0.374 Ciametral Gap, in 0.0065 Clad Thickness, in 0.0225
lad Material STD, VSH Zircaloy-4 V+,RFA Zirlo' Page 3 of 5 SGS-UFSAR Revision 18 April 26, 2000
TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Fuel Pellets Material uo2 Sintered Density, % of Theoretical Diameter, in 95.5 0.3225(lO) I RFA Annular Pellet I.D., in 0.155{ll) Length, in STD 0.530 V-5H( 3 ) 0.387 RFA Solid 0.387 RFA Annular 0.462 or 0.500( 12 } Rod Cluster Control Assemblies Neutron Absorber Ag-In-Cd Cladding Material Type 316L Ionnitride Surface Clad Thickness, in 0.0185 Number of Clusters 53 Number of Absorbers per Cluster 24 Core Structure Core Barrel, ID I OD, in 148.0 I 152.5 Thermal Shield, ID I OD, in 158.5 I 164.0 Nuclear Design Parameters: Structure Characteristics Core Diameter, in (Equivalent) 132.7 Core Average Active Fuel Height, in 143.7 Page 4 of 5 SGS-UFSAR Revision 20 May 6, 2003
TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Reflector Thickness and Composition Top - Water Plus Steel, in ~10 Bottom - Water Plus Steel, in -10 Side - Water Plus Steel, in -15 H20/U, Molecular Ratio, Lattice (cold) 2.41 (1) Standard Thermal Design Procedure. (2) Revised Thermal Design Procedure. (3) Also valid for V+ assemblies without Intermediate Flow Mixing Grids. (4) Deleted (5) For analyses where high average core temperature is bounding. (6) Deleted (7) With Intermediate Flow Mixing Grids. (8) Deleted (9) With annular axial blankets. (10) Applicable to solid or annular pellets. (11) Top and bottom 6n of RFA fuel stack height. (12) Starting with Unit 1 Region 17 and Unit 2 Region 15. Page 5 of 5 SGS-UFSAR Revision 20 May 6, 2003
TABLE 4.1-2 ANALYTIC TECHNIQUES IN CORE DESIGN Section Analysis Technique Computer Code Referenced Mechanical Design of Core Internals Loads, Deflections, and Static and Dynamic Blowdown code, FORCE, Stress Analysis Modeling Finite element structural analysis code, and others Fuel Rod Design Fuel Performance Characteristics Semi-empirical thermal Westinghouse fuel rod 4.2.1.3.1 (temperature, internal pressure, model of fuel rod with design model 4.3.3.1 clad stress, etc.) consideration of fuel 4.4.2.2 density changes, heat 4.4.3.4.2 transfer, fission gas release, etc. Nuclear Design
1) Cross Sections and Group Microscopic data Modified ENDF/B library 4.3.3.2 Constants Macroscopic constants LEOPARD/CINDER type or 4.3.3.2 for homogenized core PHOENIX-P regions Group constants for HAMMER-AIM or PHOENIX-P 4.3.3.2 control rods with self-shielding PARAGON or NEXUS 4.3.3.2

2) X-Y and X-Y-Z Power 2-Group Diffusion TURTLE (2-D) or 4.3.3.3 Distributions, Fuel Depletion, Theory ANC(2-D or 3-D)
Critical Boron Concentrations, x-y and X-Y-Z Xenon Distributions, Reactivity Coefficients
3) Axial Power Distributions 1-D, 2-Group Diffusion PANDA or APOLLO 4.3.3.3 Control Rod Worths, and Theory Axial Xenon Distribution 1 of 2 SGS-UFSAR Revision 31 December 5, 2019
Analysis
Technique TABLE 4.1-2 (Cont)
Computer Code Section Referenced
4) Fuel Rod Power Integral Transport Theory LASER 4.3.3.1 Effective Resonance Monte Carlo Weighting REPAD Temperature Function Thermal-Hydraulic Design

1) Steady-state Subchannel analysis of THINC-IV 4.4.3.4.1 local fluid conditions in rod bundles, including inertial and crossflow resistance terms, solu-tion progresses from core-wide to hot assembly to hot channel

2) Transient DNB Analysis Subchannel analysis of THINC-I (THINC-III) 4.4.3.4.1 local fluid conditions in rod bundles during transients by including accumulation terms in conservation equations; solution progresses from core-wide to hot assembly to hot channel SGS-UFSAR 2 of 2 Revision 6 February 15, 1987
TABLE 4.1-3 ~ DESIGN LOADING CONDITIONS FOR REACTOR CORE COMPONENTS
1. Fuel Assembly Weight

2. Fuel Assembly Spring Forces

3. Internals Weight

4. Control Rod Scram (equivalent static load)

5. Differential Pressure

6. Spring Preloads

7. Coolant Flow Forces (static)

8. Temperature Gradients

9. Differences in thermal expansion

a. Due to temperature differences

b. Due to expansion of different materials

10. Interference between components
~ 11. Vibration (mechanically or hydraulically induced)
12. One or more loops out of service

13. All operational transients listed in Table 5.1-10.

14. Pump overspeed

15. Seismic loads (operation basis earthquake and design basis earthquake)

16. Blowdown forces (due to cold and hot leg break)
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
4.2 MECHANICAL DESIGN The plant conditions for design are divided into four categories in accordance with their anticipated frequency of occurrence and risk to the public: Condition I - Normal Operation; Condition II - Incidents of Moderate Frequency; Condition III - Infrequent Incidents; Condition IV- Limiting Faults. The reactor is designed so that its components meet the following performance and safety criteria:
1. The mechanical design of the reactor core components and their physical arrangement, together with corrective actions of the Reactor Control, Protection, and Emergency Cooling Systems (when applicable) assure that:

a. Fuel damage* is not expected during Condition I and Condition II events. It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the Plant Cleanup System and are consistent with the plant design bases.

b. The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged*
although sufficient fuel damage might occur to preclude resumption of operation without considerable outage time.
c. The reactor can be brought to a safe state and the core can be kept subcritical with acceptable heat transfer geometry following transients arising from Condition IV events.
Fuel damage as used here is defined as penetration of the fission product barrier (i.e., the fuel rod clad).
4.2-1 SGS-UFSAR Revision 6 February 15, 1987
2. The fuel assemblies are designed to accommodate expected conditions for design for handling during assembly inspection and refueling operations and shipping loads.

3. The fuel assemblies are designed to accept control rod insertions in order to provide the required reactivity control for power operations and reactivity shutdown conditions.

4. All fuel assemblies have provisions for the insertion of in-core instrumentation necessary for plant operation.

5. The reactor internals, in conjunction with the fuel assemblies, direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements can be met for all modes of operation. In addition, the internals provide core support and distribute coolant flow to the pressure vessel head so that the temperature differences between the vessel flange and head do not result in leakage from the flange during the Condition I and II modes of operation. Required inservice inspection can be carried out as the internals are removable and provide access to the inside of the pressure vessel.
4.2.1 Fuel 4.2.1.1 Design Bases The fuel rod and fuel assembly design bases are established to satisfy the general performance and safety criteria presented in Section 4.2 and specific criteria noted below. The same design bases apply to the 17 x 17 standard (STD), 17 x 17 Vantage 5H, 17 x 17 Vantage+ and 17 x 17 Standard Robust Fuel Assembly (RFA) designs. 4.2-2 SGS-UFSAR Revision 18 April 26, 2000
4.2.1.1.1 Fuel Rods The integrity of the fuel rods is ensured by designing to prevent excessive fuel temperatures, excessive internal rod gas pressures due to fission gas releases, and excessive cladding stresses and strains. This is achieved by designing the fuel rods so that the following conservative design bases are satisfied during Condition I and Condition II events over the fuel lifetime:
1. Fuel Pellet Temperatures The center temperature of the hottest pellet is to be below the melting temperature of the uo2 (melting point of 5080°F(l) unirradiated and reducing by 58°F per 10,000 MWD/MTU). While a limited amount of center melting can be tolerated, the design conservatively precludes center melting. A calculated centerline fuel temperature of 4700°F has been selected as an overpower limit to assure no fuel melting. This provides sufficient margin for uncertainties, as described in Sections 4.4.1.2 and 4.4.2.10.1.

2. Internal Gas Pressure - The internal gas pressure of the lead rod in the reactor will be limited to a value below that which would cause (a) the diametral gap to increase due to outward clad creep during steady-state operation, and (b) extensive departure from nucleate boiling (DNB) propagation to occur.

3. Clad Stress - The effective clad stresses are less than that which would cause general yield of the clad. While the clad has some capability for accommodating plastic strain, the yield strength has been accepted as a conservative design basis.

4. Clad Tensile Strain - The clad tangential strain range is less than one percent. The clad strain design basis addresses slow transient strain rate mechanisms where the clad effective stress never reaches the yield 4.2-3 SGS-UFSAR Revision 6 February 15, 1987
strength due to stress relaxation. The 1 percent strain limit has been established based upon tensile and burst test data from irradiated clad. Irradiated clad properties are appropriate due to irradiation effects on clad ductility occurring before strain-limiting fuel clad interaction during a transient event can occur.
5. Strain Fatigue - The cumulative strain fatigue cycles are less than the design strain fatigue life. This basis is consistent with proven practice.
Radial, tangential, and axial stress components due to pressure differential and fuel clad contact pressure are combined into an effective stress using the maximum-distortion-energy theory. The von Mises' criterion is used to evaluate if the yield strength has been exceeded. The von Mises' criterion states that an isotropic material under multiaxial stress will begin to yield plastically when the effective stress (i.e., combined stress using maximum-distortion-energy theory) becomes equal to the material yield stress in simple tension as determined by an uniaxial tensile test. Since general yielding is to be prohibited, the volume average effective stress determined by integrating across the clad thickness increased by an allowance for local nonuniformi ty effects before it is compared to the yield strength. The yield strength correlation is that appropriate for irradiated clad since the irradiated properties are attained at low exposure whereas the fuel/clad interaction conditions which can lead to minimum margin to the design basis limit always occurs at much higher exposure. The detailed fuel rod design established such parameters as pellet size and density, clad-pellet diametral gap, gas plenum size, 4.2-4 SGS-UFSAR Revision 17 October 16, 1998
and helium pressure. The design also considers effects such as fuel density changes, fission gas release, clad creep, and other physical properties which vary with burnup. Irradiation testing and fuel operational experience has verified the adequacy of the fuel performance and design bases. This experience and testing are discussed in References 2 and 3. Fuel experience and testing results, as they become available, are used to improve fuel rod design and manufacturing processes and assure that the design bases and safety criteria are satisfied. The safety evaluation of the fuel rod internal pressure design basis is presented in Reference 4. 4.2.1.1.2 Fuel Assembly Structure Structural integrity of the fuel assemblies is assured by setting limits on stresses and deformations due to various loads and by determining that the assemblies do not interfere with the functioning of other components. Three types of loads are considered.
1. Nonoperational loads such as those due to shipping and handling

2. Normal and abnormal loads which are defined for Conditions I and II

3. Abnormal loads which are defined for Conditions III and IV.
These criteria are applied to the design and evaluation of the top and bottom nozzles, the guide thimbles, the grids, and the thimble joints. The design bases for evaluating the structural integrity of the fuel assemblies are: 4.2-5 SGS-UFSAR Revision 25 October 26, 2010
1. Nonoperational 4g axial and 6g lateral loading with dimensional stability.

2. Normal Operation (Condition I) and Incidents Moderate Frequency (Condition II).
For the normal operating (Condition I) and upset conditions (Condition I I) , the fuel assembly component structural design criteria are classified into two material categories, namely, austenitic steels and Zircaloy. The stress categories and strength theory presented in the ASME Boiler and Pressure Vessel Code, Section III, are used as a general guide. The maximum shear-theory (Tresca criterion) for combined stresses is used to determine the stress intensities for the austenitic steel components. The stress intensity is defined as the numerically largest difference between the various principal stresses in a three-dimensional field. The allowable stress intensity value for austenitic steels, such as nickel-chromium-iron alloys, is given by the lowest of the following:
a. 1/3 of the specific minimum tensile strength or 2/3 of the specified minimum yield strength at room temperature

b. 1/3 of the tensile strength or 90 percent of the yield strength at temperature but not to exceed 2/3 of the specified minimum yield strength at room temperature.
4.2-6 SGS-UFSAR Revision 11 July 22, 1991
The stress limits for the austenitic steel components follow: Stress Intensity Limits Categories Limit General Primary Membrane Stress Intensity Sm Local Primary Membrane Stress Intensity 1.5 Sm Primary Membrane plus Bending Stress Intensity 1.5 Sm Total Primary plus Secondary Stress Intensity 3.0 Sm TM The Zircaloy and ZIRLO structural components which consist of guide thimble and fuel tubes are in turn subdivided into two categories because of material differences and functional requirements. The fuel tube design criteria are covered separately in Section 4.2.1.1.1. The maximum stress theory is used to evaluate the guide thimble design. The maximum stress theory assumes that yielding due to combined stresses occur where one of the principal stresses are equal to the simple tensile or compressive yield stress. The Zircaloy and TM ZIRLO unirradiated properties are used to define the stress limits. Abnormal loads during Conditions III and IV worst case represented by combined seismic and blowdown loads.
1. Deflections of components cannot interfere with the reactor shutdown or emergency cooling of the fuel rods.

2. The fuel assembly component stresses under faulted conditions are evaluated using primarily the methods outlined in Appendix F of the ASME Pressure Vessel Code Section 3. Since the current analytical methods utilize elastic analysis, the stress allowables are defined as the smaller value of 2.4 Sm or 0.70 Su for primary membrane and 3.6 Sm or 1.05 Su for primary membrane plus primary bending. For the austenitic steel fuel assembly components, the stress intensity is defined in accordance with the rules described in the previous section TM for normal operating conditions. For the Zircaloy and ZIRLO components the stress limits are set at two-thirds of the material yield strength, Sy, at reactor operating temperature. This results in Zircaloy stress intensity limits being the smaller of 1.6 Sy or 0.70 Su for primary membrane and 2. 4 Sy or 1. 05 Su for primary membrane plus TM bending. For conservative purposes, the Zircaloy and ZIRLO unirradiated properties are used to define the stress limits. The grid component strength criteria are based on experimental tests. The grid component strength criterion is based on the lower 95 percent confidence level on the true mean from distribution of grid crush strength data at temperature.
4.2-7 SGS-UFSAR Revision 21 December 6, 2004
4.2.1.2 Design Description Fuel assembly and fuel rod design data are given in Tables 4.1-1 and 4.3-1. Two hundred sixty-four fuel rods, twenty-four guide thimble tubes, and one instrumentation thimble tube are arranged within a supporting structure to form a fuel assembly. The instrumentation thimble is located in the center position and provides a channel for insertion of an in-core neutron detector if the fuel assembly is located in an instrument core position. The guide thimbles provide channels for insertion of either a rod cluster control assembly, a neutron source assembly, a burnable absorber assembly or a plugging device (if used), depending on the position of the particular fuel assembly in the core. Figure 4.2-1 shows a cross section of a fuel assembly array, and Figure 4.2-2 shows a standard fuel assembly full length view. The fuel rods are loaded into the fuel assembly structure so that there is clearance between the fuel rod ends and the top and bottom nozzles. The design changes from the 17 x 17 STD design to the Vantage 5H design include reduced guide thimble and instrumentation tube diameters, and replacement of the six intermediate (mixing vane) Inconel grids with Zircaloy grids. The debris filter bottom nozzle (DFBN) design has been incorporated into the Vantage 5H fuel assembly. The DFBN is similar to the standard bottom nozzle design except that it is thinner and has a new pattern of smaller flow holes. The DFBN helps to minimize passage of debris particles that could cause fretting damage to fuel rod cladding. The Vantage+ assembly skeleton is identical to that previously described for Vantage 5H except for those modifications necessary to accommodate intended fuel operation to higher burnup levels. The Vantage+ assembly skeleton is made of low cobalt material and the spring height is slightly increased for the reduction in fuel assembly height. The modifications consist of the use of TM ZIRLO guide thimbles as necessary and small skeleton dimensional alterations to provide additional fuel assembly and rod growth space at the extended burnup levels. The Vantage+ fuel assembly is 0.200 inch shorter than the Vantage 5H assembly. The grid centerline elevations of the Vantage+ are identical to those of the Vantage 5H assembly, except for the top grid. The Vantage+ top grid has been moved down by the same 0.200 inch. The RFA assembly provides further design enhancements from the Vantage+ design including modified Low Pressure Drop mid-grids and IFMs, thicker guide tubes, and a protective bottom TM grid. The RFA continues to utilize the ZIRLO fuel rods and skeletons. 4.2-8 SGS-UFSAR Revision 21 December 6, 2004
However, since the Vantage+ and RFA fuel are intended to replace either the Westinghouse LOPAR or Vantage 5H, their assembly exterior envelope is equivalent in design dimensions, and the functional interface with the reactor internals is also equivalent to those of previous Westinghouse fuel assembly designs. Also, the Vantage+ and RFA are designed to be mechanically and hydraulically compatible with the LOPAR and Vantage 5H, and the same functional requirements and design criteria as previously established for the Westinghouse Vantage 5H fuel assembly remain valid for the Vantage+ and RFA (References 18, 21). The fuel rod clad oxidation and hydriding design criteria for ZIRLO cladding has been updated in Reference 27. The Vantage+ and RFA design parameters are provided in Table 4.1-1. Figure 4.2-2A compares the Vantage+ and the Vantage 5H fuel assembly designs. The RFA design is shown in Figure 4. 2-2B. The feed fuel loaded into Salem Unit 1 (starting with Regions 17A & 17B) and Salem Unit 2 (starting with Regions 14A & 14B) contains further enhancements to the RFA fuel assembly design in order to reduce performance limitations associated with rod internal pressure. The design changes from the 17 x 17 RFA design to the 17 x 17 RFA Z IRL0'+2 design include: an increased fuel rod length, increased fuel assembly length and elimination of the external grip top end plug. The RFA ZIRL0'+2 fuel assembly is 0. 2 inches longer than the RFA assembly due to an increase in thimble and guide tube lengths. The small dimensional alterations to the skeleton provide rod growth space for the longer fuel rods. The grid centerline elevations of the RFA ZIRL0'+2 design are identical to those of the RFA fuel assembly, except for the top grid. The RFA ZIRL0'+2 top grid has been moved upward by 0. 2 inches. The top nozzle spring height is slightly decreased due to the increase in overall assembly height. The RFA ZIRL0'+2 assembly design is shown in Figure 4.2-2C. The feed fuel loaded into Salem Unit 1 (starting with Regions 19A and 19B) and Salem Unit 2 (starting with Regions 17A and 17B) contains further enhancements to the RFA mid grid design in order to further reduce grid to rod fretting. The design changes from the RFA mid grid to the RFA-2 mid grid are changes to the spring window cut-outs and the spring and dimple contact areas. Compared to RFA mid grid, RFA-2 mid grid offers improved resistance to fuel rod fretting without significantly affecting any other thermal-hydraulic or mechanical TM performance. Both RFA and RFA-2 mid grids are made from ZIRLO material and both mid grids have the same mass and volume. Each fuel assembly is installed vertically in the reactor vessel and stands upright on the lower core plate, which is fitted with alignment pins to locate and orient the assembly. After all fuel assemblies are set in place, the upper support structure is installed. Alignment pins, built into the upper core plate, engage and locate the upper ends of the fuel assemblies. The upper core plate then bears downward against the fuel assembly top nozzle via the holddown springs to hold the fuel assemblies in place. 4.2.1.2.1 Fuel Rods The Vantage+ and RFA fuel rod designs represent a modification to the Vantage
. TM TM 5H fuel rod in claddlng of ZIRLO as compared to Zircaloy-4. ZIRLO is a zirconium alloy similar to Zircaloy-4, which has been specifically developed to enhance corrosion resistance. The Vantage 5H, Vantage+, and RFA fuel rods contain enriched uranium dioxide fuel pellets, and Integral Fuel Burnable Absorber ( IFBA) coating on some of the fuel pellets. Additionally, the RFA fuel rods contain axial blankets (top and bottom 6 inches) of annular pellets.
The annulus represents a 25% void by volume of the pellet. Schematics of the fuel rods are shown in Figures 4.2-3, 4.2-3A, and 4.2-3B. 4.2-9 SGS-UFSAR Revision 28 May 22, 2015
The Vantage+ and RFA fuel rods have the same wall thickness as the Vantage 5H. The Vantage+ fuel rod length is shorter to provide the required fuel rod growth room. To offset the reduction in plenum length the Vantage+ fuel rod has a variable pitch plenum spring. The variable pitch plenum spring provides the same support as the Vantage 5H plenum spring, but with less spring turns which means less spring volume. The RFA fuel rod is longer than the Vantage+ and Vantage 5H designs but maintains sufficient room for fuel rod growth. This provides additional plenum length to accommodate fission gas release associated with high burnup. In addition, the RFA utilizes the variable pitch plenum spring and axial blankets of annular pellets to provide additional plenum margin. The 6 inches (top and bottom) of annular pellets effectively provide 3 additional inches of plenum volume beyond the current VANTAGE+ design. The bottom end plug has an internal grip feature to facilitate rod loading on the Vantage+, RFA, and Vantage 5H designs and provides appropriate lead-in for the removable top nozzle reconstitution feature. The RFA design also incorporates the external grip top end plug into its fuel rod design. This feature provides additional capability to reposition the fuel during manufacturing and will simplify reconstitution of the fuel rod from the top of the assembly. The Salem Vantage+ and RFA fuel rods also may posses a zirconium dioxide (Zr0 ) coating on the 2 bottom outside surface of the fuel rod. This fuel feature may be incorporated as an option to be determined on a reload basis. The protective oxide coating covers the bottom end plug, the bottom end plug weldment and a portion of the cladding. A metallurgical-bonded layer of zro uniformly covers a minimum of 4.5 inches of the bottom end of 2 the fuel rod. The minimum coating length was chosen to ensure that the coating would extend through the top of the current bottom Inconel structural grid, independent of the fuel rod loading position or fuel assembly design. The coating, which is 2 to 6 microns thick, provides a hard, wear resistant, surface layer of ZrO for additional debris damage resistance, thereby improving fuel 2 reliability. This extra layer of oxide coating provides additional rod fretting wear protection. The RFA design incorporates the protective bottom grid and modified fuel rod bottom end plug. These enhanced debris mitigating features, described in Section 4.2.1.2.2, diminish the need for the oxide coating. Therefore, for core designs utilizing the RFA, the oxide coating may be incorporated as an option to be determined on a reload basis. The RFA ZIRL0'+2 fuel rod design represents a modification to the RFA fuel rod in terms of increased fuel rod length and increased fuel assembly length. The RFA ZIRL0'+2 fuel rod length is 0. 2 inches longer than the RFA fuel rod length. Sufficient fuel rod growth margin is accomplished by increasing the length of the instrument thimble and guide thimbles. The RFA ZIRL0'+2 design incorporates a shorter (0.12 inches) external top end plug without a gripper feature. There is no longer any functional purpose for this top end plug gripper feature as fuel rods can still be handled through the top of the assembly for reconstitution. Both the longer fuel rod and shorter external top end plug provide additional plenum length to accommodate fission gas release. The plenum spring free length was increased to accommodate the increase in fuel rod plenum length, while the spring rate was decreased to maintain the same fuel stack hold-down force as in the RFA fuel rod design. A schematic of the RFA ZIRL0'+2 fuel rod is shown in Figure 4.2-3C. The Salem Vantage 5H, Vantage+, and RFA fuel uses a standardized fuel pellet design which is a refinement to the chamfered pellet design. The standard design helps to improve manufacturability while maintaining or improving performance (e.g., improved pellet chip resistance during manufacturing and handling) . The Vantage 5H, Vantage+, and RFA IFBA coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin boride 4.2-10 SGS-UFSAR Revision 21 December 6, 2004
coating on the pellet cylindrical surface. Coated pellets occupy the central portion of the fuel column. The number and pattern of IFBA rods within an assembly vary depending on specific application. The ends of the coated and uncoated pellets are dished to allow for greater axial expansion at the pellet centerline and void volume for fission gas release. To accommodate a six inch axial blanket length at the top and bottom of the Robust Fuel Assembly (RFA) fuel rods, the annular pellets have a longer length than the solid pellets to obtain an integer number of fuel pellets which will equal a six inch length. To avoid overstressing of the cladding or seal welds, void volume and clearances are provided within the rods to accommodate fission gases released from the fuel, differential thermal expansion between the cladding and the fuel, and fuel density changes during burnup. Shifting the fuel within the cladding during handling or shipping prior to core loading is prevented by a stainless steel helical spring which bears on top of the fuel. At assembly the pellets are stacked in the cladding to the required fuel height, the spring is then inserted into the top end of the fuel tube and the end plugs pressed into the ends of the tube and welded. All fuel rods are internally pressurized with helium during the welding process in order to minimize compressive clad stresses and creep due to coolant operating pressures. Fuel rod pressurization is dependent on the planned fuel burnup as well as other fuel design parameters and fuel characteristics (particularly densification potential). 4.2.1.2.2 Fuel Assembly Structure The fuel assembly structure consists of a bottom nozzle, top nozzle, guide thimbles, and grids, as shown on Figures 4.2-2 and 4.2-2A. Bottom Nozzle The bottom nozzle is a box-like structure which serves as a bottom structural element of the fuel assembly and directs the coolant flow distribution to the assembly. The square nozzle is fabricated from Type 304 stainless steel and consists of a perforated plate and four angle legs with bearing plates as shown on Figure 4.2-2. The legs form a plenum for the inlet coolant flow to the fuel assembly. The plate itself acts to prevent a downward ejection of the fuel rods from the fuel assembly. The bottom nozzle is fastened to the fuel assembly guide tubes by locked screws which penetrate through the nozzle and mate with an inside fitting in each guide tube. The debris filter bottom nozzle (DFBN) design was introduced into the Salem fuel assemblies to help reduce the possibility of fuel rod damage due to debris-induced fretting. The Vantage+ and RFA assemblies have a low cobalt 4.2-11 SGS-UFSAR Revision 28 May 22, 2015
stainless steel DFBN. The DFBN design incorporates a modified flow hole size and pattern, as described below, and a decreased nozzle height and thinner top plate to accommodate the high burnup fuel rods. The DFBN retains the design reconstitution feature which facilitates easy removal of the nozzle from the fuel assembly. The relatively large flow holes in a conventional bottom nozzle were replaced with a new pattern of smaller flow holes in the DFBN. The holes are sized to minimize passage of debris particles large enough to cause damage. The holes were also sized to provide sufficient flow area, comparable pressure drops, and continued structural integrity of the nozzle. Tests to measure pressure drop and demonstrate structural integrity have been performed to verify that the DFBN is totally compatible with the current design. Salem Unit 2 Region 23 and Salem Unit 1 Region 26 are the first regions to implement the Standardized Debris Filter Bottom Nozzle (SDFBN) design into Salem RFA fuel assemblies. The SDFBN is designed to have a loss coefficient that is the same, independent of supplier. The SDFBN eliminates the side skirt communication flow holes as a means of improving the debris mitigation performance of the bottom nozzle. This nozzle meets all of the applicable mechanical and thermal-hydraulic design criteria. The RFA fuel assembly with the SDFBN is illustrated in Figure 4.2-2C. Coolant flow through the fuel assembly is directed from the plenum in the bottom nozzle upward through the penetrations in the plate to the channels between the fuel rods . The penetrations in the plate are positioned between the rows of the fuel rods. Axial loads (holddown) imposed on the fuel assembly and the weight of the fuel assembly is transmitted through the bottom nozzle to the lower core plate. Indexing and positioning of the fuel assembly is controlled by alignment holes in two diagonally opposite bearing plates which mate with locating pins in the lower core plate. Any lateral loads on the fuel assembly are transmitted to the lower core plate through the locating pins. Top Nozzle The top nozzle assembly functions as the upper structural element of the fuel assembly in addition to providing a partial protective housing for the rod cluster control assembly or other components. It consists of an adapter plate, enclosure, top plate, and pads. The integral welded assembly has holddown springs mounted on the assembly as shown on Figure 4.2-2. The reconstitutable top nozzle (RTN) design contains hold-down springs and screws made of Inconel 718 and Inconel 600, respectively, whereas other components are made of Type 304 stainless steel. The feed fuel loaded into Salem Unit 2 (starting with Regions 14A and 14B) and Salem Unit 1 (starting with Regions 17A and 17B) contains recons ti tutable top nozzles with hold-down screws made of Inconel 718, as opposed to Inconel 600, in order to increase resistance to stress corrosion cracking. Vantage+, RFA, and Vantage 5H fuel assemblies use the reconstitutable top nozzle (RTN). The RTN design for the Vantage 5H fuel assembly differs from the conventional design in two ways: 1) a groove is provided in each thimble thru-4.2-12 SGS-UFSAR Revision 28 May 22, 2015
hole in the nozzle plate to facilitate attachment and removal, and; 2) the nozzle plate thickness is reduced to provide additional axial space for fuel rod growth. Additional details of this design feature, the design bases and evaluation of the reconstitutable top nozzle are given in Section 2.3.2 in Reference 15. The square adapter plate is provided with round and obround penetrations to permit the flow of coolant upward through the top nozzle. Other round holes are provided to accept sleeves which are welded to the adapter plate and mechanically attached to the thimble tubes. The ligaments in the plate cover the tops of the fuel rods and prevent their upward ejection from the fuel assembly. The enclosure is a sheet metal shroud which sets the distance between the adapter plate and the top plate. The top plate has a large square hole in the center to permit access for the control rods and the control rod spiders. Holddown springs are mounted on the top plate and are fastened in place by screws and clamps located at two diagonally opposite corners. The clamps are attached to the nozzle by a specific arrangement of tack welds or tack weld(s) in combination with a stainless steel clamp screw, depending on the manufacturing process in place at the time a given fuel region was built. The spring screws apply a load directly to the base of the hold-down springs. The clamps do not have any bearing surfaces that load the spring to the nozzle, but primarily provide a stationary location for attachment of lock wires that prevent rotation of the spring screws. On the other two corners, integral pads are positioned which contain alignment holes for locating the upper end of the fuel assembly. Salem Units 1 and 2 later implemented the Westinghouse Integral Nozzle (WIN) design in RFA-2 fuel assemblies. The WIN design, while similar to the RTN, incorporates design and manufacturing improvements to eliminate the Inconel 718 spring screw for attachment of the holddown springs. In the WIN nozzle, the springs are assembled into the nozzle pad and pinned in place. The WIN design provides a wedged rather than a clamped (bolted) joint to transfer the fuel assembly holddown forces into the top nozzle structure. A replacement reconstitutable top nozzle (RRTN) design may be used in a reload cycle to replace the original reconstitutable top nozzle (RTN) or the WIN on an irradiated fuel assembly. The mechanical features of the RRTN are the same as those for the RTN (see Figure 4.2-2) or the WIN with some minor dimensional differences in the top nozzle adapter plate thimble hole to facilitate attachment to an irradiated fuel assembly. The RRTN design contains hold-down springs and screws made of Inconel 718, whereas, other components are made of Type 304 stainless steel. Guide and Instrument Thimbles The guide thimbles are structural members which also provide channels for the neutron absorber rods, burnable poison rods, or neutron source assemblies. Each TM one is fabricated from Zircaloy-4 or ZIRLO tubing having two different diameters. The larger diameter at the top provides a relatively large annular area to permit rapid insertion of the control rods during a reactor trip as well as to accommodate the flow of coolant during normal operation. Four holes are provided on the thimble tube above the dashpot to reduce the rod drop time. The lower portion of the guide thimbles has a reduced diameter to produce a dashpot action near the end of the control rod travel during normal operation and to accommodate the outflow of water from the dashpot during a reactor trip. The dashpot is closed at the bottom by means of an end plug which is provided with a small flow port to avoid fluid stagnation in the dashpot volume during normal operation. The top end of the guide thimble is fastened to a tubular insert by three expansion swages. The insert engages into the top nozzle and is secured into position by the lock tube. The lower end of the guide thimble is fitted with an end plug which is then fastened into the bottom nozzle by a locked screw. 4.2-13 SGS-UFSAR Revision 31 December 5, 2019
Fuel rod support grids are fastened to the guide thimble assemblies to create an integrated structure. Since welding of the Inconel grid and Zircaloy thimble is not possible, the fastening technique depicted on Figures 4.2-5 and 4.2-9 is used for all but the top and bottom grids in a fuel assembly. An expanding tool is inserted into the inner diameter of the Zircaloy or Zirlo TM thimble tube to the elevation of the zircaloy sleeves that have been welded to the Zircaloy middle grid assemblies. The four-lobed tool forces the thimble and sleeve outward to a predetermined diameter, thus joining the two components. The top grid-to-thimble attachment for the Vantage 5H, Vantage+, and RFA design is shown on Figure 4.2-7. The Zircaloy or ZIRLOTM thimbles are fastened to the top nozzle inserts by expanding the members as shown on Figure 4. 2-7. The inserts then engage the top nozzle and are secured into position by the insertion of lock tubes. The bottom grid assembly is joined to the fuel assembly as shown on Figure 4.2-
11. The stainless steel insert is spot welded to the bottom grid and later captured between the guide thimble end plug and the bottom nozzle by means of a stainless steel thimble screw.
The described methods of grid fastening are standard and have been used successfully since the introduction of Zircaloy guide thimbles in 1969. The central instrumentation thimble of each fuel assembly is constrained by seating in counterbores in each nozzle. This tube is a constant diameter and guides the incore neutron detectors. This thimble is expanded at the top and mid grids in the same manner as the previously discussed expansion of the guide thimbles to the grids. 4.2-13a SGS-UFSAR Revision 18 April 26, 2000
With the exception of an increased length above the dashpot, the Vantage+ guide thimbles are identical to those in the Vantage 5H design. For the RFA, the thimble tube thickness has been increased by 25% ( 4 Mils) relative to the V5H and Vantage+ designs. The thimble tube outer diameter of both the upper spans (major diameter section above the dashpot) and the dashpot spans have been increased. The inner diameter of the thimble tube in both the upper spans and in the dashpot spans are unchanged from the current designs. With the thicker guide thimble tube, the cross-sectional area is increased by 26% in the major diameter section and 29% in the dashpot section. The Vantage+, RFA and Vantage 5H guide thimble ID provides adequate clearance for the control rods and sufficient diametral clearance for burnable absorber rods and source rods. The Vantage+, RFA and Vantage 5H instrumentation tube diameter has sufficient diametral clearance for the flux thimble to traverse the tube without binding. Grid Assemblies The fuel rods, as shown on Figures 4. 2-2, 4. 2-2A and 4. 2-2B, are supported laterally at intervals along their length by grid assemblies which maintain the lateral spacing between the rods throughout the design life of the assembly. Each fuel rod is afforded lateral support at six contact points within each grid by the combination of support dimples and springs. The grid assembly consists of individual slotted straps interlocked and welded in an "egg-crate" arrangement to join the straps permanently at their points of intersection. The straps contain spring fingers, support dimples, and mixing vanes. The magnitude of the grid restraining force on the fuel rod is set high enough to minimize possible fretting, without overstressing the cladding at the points of contact between the grids and fuel rods. The grid assemblies also allow axial thermal expansion of the fuel rods without imposing restraint sufficient to develop buckling or distortion of the fuel rods. Up to four types of grid types are used in each fuel assembly: Mid-grids (structural grids with flow mixing vanes), Intermediate Flow Mixing (IFM) grids (non-structural grids with flow mixing vanes), top and bottom structural grids without mixing vanes, and the protective bottom grid ( P-grid) . Table 4.1-1 provides the breakdown of the grid types, number and grid material used in each of the fuel designs. Flow mixing vanes project from the edge of the inner grid strap into the coolant stream to promote mixing of the coolant in the high heat flux regions of the fuel assembly. 4.2-13b SGS-UFSAR Revision 18 April 26, 2000
The IFMs are positioned at the mid-spans of the four uppermost mid-grids to further increase the flow turbulence in the axial zone where departure from nucleate boiling (DNB) is limiting. Each IFM grid cell contains four dimples, which are designed to prevent mid-span channel closure in the spans containing IFMs and fuel contact with the mixing vanes. For the RFA, the modified low pressure drop mid-grids and IFM grids are embossed to accept the larger diameter guide thimble tube. The P-grid is a partial height grid similar in configuration to the mid-grid, but without mixing vanes. It is located between the bottom Inconel grid and the bottom nozzle, nearly on the surface of the bottom nozzle. The intersections of the inner straps of the P-grid align with the flow holes of the DFBN, effectively bisecting the flow path through the flow hole into four quarters. This provides an effective barrier against small debris. In conjunction with the P-grid, the fuel rod bottom end plug is changed to a longer design such that the portion of the fuel rod engaged in the P-grid and extending up past the top of the P-grid is solid end plug material. This provides a protective zone where trapped debris cannot fret through the fuel rod and cause a failure. The hydraulic effects of the P-grid are minimized by positioning the fuel rods 0.085 inches above the bottom nozzle. The combination of the lowered fuel rod position and longer fuel rod end plug results in no change to the axial fuel stack height from the previous Vantage+ fuel region. For the application of the P-grid, the bottom Inconel grid was welded to 20 of the 24 inserts. The remaining four inserts are spot-welded to the P-grid at the four outermost corners on the grid diagonal. Salem Unit 2 Region 23 and Salem Unit 1 Region 26 are the first regions to implement the Westinghouse Robust Protective Grid (RPG) which was developed as a result of observed failures of the P-grid in the field during Post Irradiation Exams (PIE) performed at several different plants. It was determined by Westinghouse that observed failures were the result of two primary issues; 1) fatigue failure within the protective grid itself at the top of the end strap and 2) stress corrosion cracking ( SCC) primarily within the rod support dimples. The RPG implements design changes such as increasing the maximum nominal height of the grid, increasing the amount of material at the ends of the dimple window cutouts, increasing the radii of the dimple window cutouts, and the welding of four additional inserts for a total of 8 welded inserts out of the 24 total inserts in order to help better support the grid. 4.2-13c SGS-UFSAR Revision 28 May 22, 2015
The nominal height of the grid was increased to allow "V-notch" window cutouts to be added to help minimize flow-induced vibration caused by vortex shedding at the trailing edge of the inner grid straps. Figure 4.2-2C shows RFA fuel with the increased nominal height due to implementation of the RPG. These design changes incorporated into the RPG design help address the issues of fatigue failures and failures due to stress corrosion cracking. These changes do not impact the thermal hydraulic performance of the RPG as there is no change to the pressure loss coefficient. In addition, the RPG retains the original protective grid function as a debris mitigation feature. The outside straps on all the grids contain mixing vanes which, in addition to their mixing function, aid in guiding the grids and fuel assemblies past projecting surfaces during handling or during loading and unloading of the core. During 1989, snag-resistant grids were introduced. These grids contain outer grid straps which are modified to help prevent assembly hangup from grid strap interference during fuel assembly removal. This was accomplished by changing the grid strap corner geometry and the addition of guide tabs on the outer grid strap. 4.2-13d SGS-UFSAR Revision 28 May 22, 2015
4.2.1.3 Design Evaluation 4.2.1.3.1 Fuel Rods The fuel rods are designed to assure the design bases are satisfied for Condition I and II events. This assures that the fuel performed, and safety criteria (Section 4.2.1.1) are satisfied. Materials - Fuel Cladding The desired fuel rod cladding is a material which has a superior combination of neutron economy (low absorption cross section), high strength (to resist deformation due to differential pressures and mechanical interaction between fuel and clad), high corrosion resistance (to coolant, fuel, and fission products) , and high reliability. Zircaloy-4 and ZIRLOTM have this desired combination of cladding properties. There is considerable pressurized water reactor (PWR) operating experience on the capability of Zircaloy as a cladding material (2). Clad hydriding has not been a significant cause of clad perforation since current controls on fuel-contained moisture levels were instituted. Metallographic examination of irradiated commercial fuel rods have shown occurrences of fuel/ clad chemical interaction. Reaction layers of <1 mil in thickness have been observed between fuel and clad at limited points around the circumference. Westinghouse metallographic data indicates that this interface layer remains very thin even at high burnup. Thus, there is no indication of propagation of the layer and eventual clad penetration. Stress corrosion cracking is another postulated phenomenon related to fuel/clad chemical interaction. Out-of-reactor tests have shown that in the presence of high clad tensile stress, relatively large concentrations of iodine, or cadmium in solution in liquid cesium can stress corrode zirconium alloy tubing and lead to eventual clad cracking. Extensive post irradiation examination has 4.2-14 SGS-UFSAR Revision 17 October 16, 1998
produced no conclusive evidence that this mechanism is operative in commercial fuel. Creep collapse and creep-down, along with the associated irradiation stability of cladding, have been evaluated using the models described in references 7 and 28. It has been established that the design basis of no clad collapse during planned core life can be satisfied by limiting fuel densification and by having a sufficiently high initial internal rod pressure. Materials - Fuel Pellets Sintered, high density uranium dioxide fuel is chemically inert, with respect to the cladding, at core operating temperatures and pressures. In the event of cladding defects, the high resistance of uranium dioxide to attack by water protects against fuel deterioration although limited fuel erosion can occur. As has been shown by operating experience and extensive experimental work, the thermal design parameters conservatively account for changes in the thermal performance of the fuel elements due to pellet fracture which may occur during power operation. The consequences of defects in the cladding are greatly reduced by the ability of uranium dioxide to retain fission products including those which are gaseous or highly volatile. Observations from several operating Westinghouse PWRs (2) have shown that fuel pellets can densify under irradiation to a density higher than the manufactured values. Fuel densification and subsequent incomplete settling of the fuel pellets results in local and distributed gaps in the fuel rods. Fuel densification has been minimized by improvements in the fuel manufacturing process and by specifying a nominal 95.5 percent initial fuel density. The effects of fuel densification have been taken into account in the nuclear and thermal-hydraulic design of the reactor described in Sections 4.3 and 4.4, respectively. Materials - Strength Considerations One of the most important limiting factors in fuel element duty is the mechanical interaction of fuel and cladding. This fuel- cladding interaction produces cyclic stresses and strains in the cladding, and these in turn consume cladding fatigue life. The 4.2-15 SGS-UFSAR Revision 30 May 11, 2018
reduction of fuel-cladding interaction is therefore a principal goal of design. In order to achieve this goal and to enhance the cyclic operational capability of the fuel rod, the technology for using prepressurized fuel rods in Westinghouse PWRs has been developed. Initially the gap between the fuel and cladding is sufficient to prevent hard contact between the two. However, during power operation a gradual compressive creep of the cladding onto the fuel pellet occurs due to the external pressure exerted on the rod by the coolant. Cladding compressive creep eventually results in hard fuel-cladding contact. During this period of fuel-cladding contact, changes in power level could result in significant changes in cladding stresses and strains. By using prepressurized fuel rods to partially offset the effect of the coolant external pressure, the rate of cladding-creep toward the surface of the fuel is reduced. Fuel rod prepressurization delays the time at which substantial fuel-cladding interaction and hard contact occur and hence significantly reduces the number and extent of cyclic stresses and strains experienced by the cladding both before and after fuel-cladding contact. These factors result in an increase in the fatigue life margin of the cladding and lead to greater cladding reliability. If gaps should form in the fuel stacks, clad flattening will be prevented by the rod prepressurization so that the flattening time will be greater than the fuel core life. Steady-State Performance Evaluation In the calculation of the steady-state performance of a nuclear fuel rod, the following interacting factors must be considered:
1. Clad creep and elastic deflection

2. Pellet density changes, thermal expansion, gas release, and thermal properties as a function of temperature and fuel burnup 4.2-16 SGS-UFSAR Revision 6 February 15, 1987

3. Internal pressure as a function of fission gas release, rod geometry, and temperature distribution These effects are evaluated using an overall fuel rod design model (Reference

17) which include appropriate modifications for time dependent fuel densification. With these interacting factors considered, the model determines the fuel rod performance characteristics for a given rod geometry, power history, and axial power shape. In particular, internal gas pressure, fuel and cladding temperatures, and cladding deflections are calculated. The fuel rod is divided lengthwise into several sections and radially into a number of annular zones. Fuel density changes, cladding stresses, strains and deformations, and fission gas releases are calculated separately for each segment. The effects are integrated to obtain the internal rod pressure. The initial rod internal pressure is selected to delay fuel/clad mechanical interaction and to avoid the potential for flattened rod formation. Clad flattening for Salem Nuclear Generating Station (SNGS) fuel is evaluated using the models described in Reference 7.
The gap conductance between the pellet surface and the cladding inner diameter is calculated as a function of the composition, temperature, and pressure of the gas mixture, and the gap size or contact pressure between clad and pellet. After computing the fuel temperature of each pellet annular zone, the fractional fission gas release is assessed using an empirical model derived from experimental data (17). The total amount of gas released is based on the average fractional release within each axial and radial zone and the gas generation rate which, in turn, is a function of burnup. Finally, the gas released is summed over all zones and the pressure is calculated. The model shows good agreement in fit for a variety of published and proprietary data on fission gas release, fuel temperatures, and clad deflections ( 17) . Included in this spectrum are variations in power, time, fuel density, and geometry. The in-pile fuel 4.2-17 SGS-UFSAR Revision 22 May 5, 2006
Temperature measurements' comparisons used are shown in Reference 17. Typical fuel clad inner diameter and the fuel pellet outer diameter as a function of exposure are presented on Figure 4.2-4. The cycle-to-cycle changes in the pellet outer diameter represent the effects of power changes as the fuel is moved into different positions as a result of refueling. The gap size at any time is merely the difference between clad inner diameter and pellet outer diameter. Total clad-pellet surface contact occurs near the end of Cycle 2. The figure represents hot fuel dimensions for a fuel rod operating at the power level shown on Figure 4.2-6. Figure 4.2-6 illustrates representative fuel rod internal gas pressure and linear power for the lead burnup rod vs. irradiation time. In addition, it outlines the typical operating range of internal gas pressures which is applicable to the total fuel rod population within a region. The "best estimate" fission gas release model was used in determining the internal gas pressures as a function of irradiation time. The clad stresses at a constant local fuel rod power are low. Compressive stresses are created by the pressure differential between the coolant pressure and the rod internal gas pressure. Because of the prepressurization with helium, the volume average effective stresses are always less than -10,000 psi at the pressurization level used in this fuel rod design. Stresses due to the temperature gradient are not included in this average effective stress because thermal stresses are, in general, negative at the clad inner diameter and positive at the clad outer diameter and their contribution to the clad volume average stress is small. Furthermore, the thermal stress decreases with time during steady-state operation due to stress relaxation. The stress due to pressure differential is highest in the minimum power rod at the beginning of life (BOL) (due to low internal gas pressure) and the thermal stress is highest in the maximum power rod (due to steep temperature gradient). 4.2-18 SGS-UFSAR Revision 22 May 5, 2006
Tensile stresses could be created once the clad has come in contact with the pellet. These stresses would be induced by the fuel pellet swelling during irradiation. As shown on Figure 4. 2-4, there is very limited clad pushout after pellet-clad contact. Fuel swelling can result in small clad strains (< 1 percent) for expected discharge burnups, but the associated clad stresses are very low because of clad creep (thermal and irradiation-induced creep). Furthermore, the 1 percent strain criterion is extremely conservative for fuel-swelling driven clad strain because the strain rate associated with solid 7 1 fission products swelling is very slow (-5 x 10- hr- ) In-pile experiments ( 8) have shown that Zircaloy tubing exhibits "superplastici ty" at slow strain rates during neutron irradiation. Uniform clad strains of greater than 10 percent have been achieved under these conditions with no sign of plastic instability. Transient Evaluation Method Pellet thermal expansion due to power increases is considered the only mechanism by which significant stresses and strains can be imposed on the clad. Power increases in commercial reactors can result from fuel shuffling (e.g., Region 3 positioned near the center of the core for Cycle 2 operation after operating near the periphery during Cycle 1), reactor power escalation following extended reduced power operation, and control rod movement. In 4.2-19 SGS-UFSAR Revision 11 July 22, 1991
the mechanical design model, lead rods are depleted using best estimate power histories as determined by core physics calculations. During the depletion the amount of diametral gap closure is evaluated based upon the pellet expansion-cracking model, clad creep model, and fuel swelling model. At various times during the depletion, the power is increased locally on the rod to the burnup dependent attainable power density as determined by core physics calculations. The radial, tangential, and axial clad stresses resulting from the power increase are combined into a volume average effective clad stress. The von Mises' criterion is used to evaluate if the clad yield stress has been exceeded. This criterion states that an isotropic material in multi-axial stress will begin to yield plastically when the effective stress exceeds the yield stress as determined by a uniaxial tensile test. The yield stress correlation is that for irradiated cladding since fuel/clad interaction occurs at high burnup. Furthermore, the effective stress is increased by an allowance, which accounts for stress concentrations in the clad adjacent to radial cracks in the pellet, prior to the comparison with the yield stress. This allowance was evaluated using a two-dimensional (r, e) finite element model. Slow transient power increases can result in large clad strains without exceeding the clad yield stress because of clad creep and stress relaxation. Therefore, in addition to the yield stress criterion, a criterion on allowable clad positive strain is necessary. Based upon high strain rate burst and tensile test data on irradiated tubing, 1 percent strain was determined to be the lower limit on irradiated clad ductility and thus adopted as a design criterion. In addition to the mechanical design models and design criteria, Westinghouse relies on performance data accumulated through transient power test programs in experimental and commercial reactors, and through normal operation in commercial reactors. 4.2-20 SGS-UFSAR Revision 6 February 15, 1987
It is recognized that a possible limitation to the satisfactory behavior of the fuel rods in a reactor which is subjected to daily load follow is the failure of the cladding by low cycle strain fatigue. During their normal residence time in reactor, the fuel rods may be subjected to -1000 cycles with typical changes in power level from 50 to 100 percent of their steady-state values. The assessment of the fatigue life of the fuel rod cladding is subjected to a considerable uncertainty due to the difficulty of evaluating the strain range which results from the cyclic interaction of the fuel pellets and claddings. This difficulty arises, for example, from such highly unpredictable phenomena as pellet cracking, fragmentation, and relocation. Nevertheless, since early 1968, Westinghouse has been investigating this particular phenomenon both analytically and experimentally. Strain fatigue tests on irradiated and nonirradiated hydrided Zircaloy-4 claddings were performed which permitted a definition of a conservative fatigue life limit and recommendation of a methodology to treat the strain fatigue evaluation of the Westinghouse reference fuel rod designs. However, Westinghouse is convinced that the final proof of the adequacy of a given fuel rod design to meet the load follow requirements can only come from in-pile experiments performed on actual reactors. The Westinghouse experience in load follow operation dates back to early 1970 with the load follow operation of the Saxton reactor. Successful load follow operation has been performed on reactor A (300 load follow cycles) and reactor B (150 load follow cycles) . In both cases, there was no significant coolant activity increase that could be associated with the load follow mode of operation. The following paragraphs present briefly the Westinghouse analytical approach to strain fatigue. A comprehensive review of the available strain-fatigue models was conducted by Westinghouse as early as 1968. This included the 4.2-21 SGS-UFSAR Revision 6 February 15, 1987
Langer-O'Donnell model (9), the Yao-Munse model, and the Manson-Halford Model. Upon completion of this review and using the results of the Westinghouse experimental programs discussed below, it was concluded that the approach defined by Langer-0' Donnell would be retained and the empirical factors of their correlation modified in order to conservatively bound the results of the Westinghouse testing program. The Langer-O'Donnell empirical correlation has the following form: E 100 1n + 4.jN; 100 RA where: s 1/2 E ~ s pseudo - stress amplitude which causes a t
.2 failure in Nf cycles (lb/ln ) ~s total strain range (in/in) t .2 E Young's Modulus (lb/ln )
Nf number of cycles to failure RA reduction in area at fracture in a uniaxial tensile test (percent) 2 S endurance limit (lb/in ) e Both RA and S are empirical constants which depend on the type of material, e the temperature, and the irradiation. The Westinghouse testing program was subdivided in the following subprograms:
1. A rotating bend fatigue experiment on unirradiated Zircaloy-4 specimens at room temperature and at 725°F.
4.2-22 SGS-UFSAR Revision 6 February 15, 1987
Both hydrided and nonhydrided Zircaloy-4 cladding were tested.
2. A biaxial fatigue experiment in gas autoclave on unirradiated Zircaloy-4 cladding both hydrided and nonhydrided.

3. A fatigue test program on irradiated cladding from the CVTR and Yankee Core V conducted at Battelle Memorial Institute.
The results of these test programs provided information of different cladding conditions including the effect of irradiation, hydrogen level, and temperature. The Westinghouse design equations followed the concept for the fatigue design criterion according to Section 3 of the ASME Boiler and Pressure Vessel code; namely:
1. The calculated pseudo-stress amplitude (S ) has to be multiplied by a
a factor of 2 in order to obtain the allowable number of cycles (Nf
) .
2. The allowable cycles for a given Sa is 5 percent of Nf, or a safety factor of 20 on cycles.
The lesser of the two allowable number of cycles is selected. The cumulative fatigue life fraction is then computed as: k
<1 1
where: nk number of diurnal cycles of mode k. 4.2-23 SGS-UFSAR Revision 11 July 22, 1991
The potential effects of operation with waterlogged fuel are discussed in Section 4.4.3.6. Waterlogging is not considered to be a concern during operational transients. 4.2.1.3.2 Fuel Assembly Structure Stresses and Deflections The potential sources of high stresses in the assembly are avoided by the design. For example, stresses in the fuel rod due to thermal expansion and TM Zircaloy or ZIRLO irradiation growth are limited by the relative motion of the rod as it slips over the grid spring and dimple surfaces. Clearances TM between the fuel rod ends and nozzles are provided so that Zircaloy or ZIRLO irradiation growth will not result in end interferences. As another example, stresses due to holddown springs in opposition to the hydraulic lift force are limited by the deflection characteristic of the springs. Stresses in the fuel assembly caused by tripping of the rod cluster control assembly have little influence on fatigue because of the small number of events during the life of an assembly. Welded joints in the fuel assembly structure are considered in the structural analysis of the assembly. Appropriate material properties of welds are used to ensure the design bases are met. Assembly components and prototype fuel assemblies made from production parts have been subjected to structural tests to verify that the design bases requirements are met. The fuel assembly design loads for shipping have been established at 4 g axial and 6g lateral. Probes are permanently placed into the shipping cask to monitor and detect fuel assembly displacements that would result from loads in excess of the criteria. Past history and experience have indicated that loads which exceeded the allowable limits rarely occur. Exceeding the limits requires reinspection of the fuel assembly for damage. Tests on various fuel assembly components such as the grid assembly, sleeves, inserts, and structure joints have been performed to assure that the shipping 4.2-24 SGS-UFSAR Revision 17 October 16, 1998
design limits do not result in impairment of fuel assembly function. Dimensional Stability The Vantage 5H Mechanical Test Program description and results are given in Reference 16 and are considered to be applicable to Vantage+ as the two assemblies are structurally essentially identical. The development of the RFA design included a comprehensive set of mechanical and hydraulic tests: pressure drop, assembly vibration, fuel rod vibration, bulge joint strength, grid crush. A description of these tests and results is provided in References 19, 20 and 21. The coolant flow channels are established and maintained by the structure composed of grids and guide thimbles. The lateral spacing between fuel rods is provided and controlled by the support dimples of adjacent grid cells. Contact of the fuel rods on the dimples is assured by the clamping force provided by the grid springs. Lateral motion of the fuel rods is opposed by the spring force and the internal moments generated between the spring and the support dimples. Grid testing is discussed in Reference 10. No interference with control rod insertion into thimble tubes will occur during a postulated loss-of-coolant accident (LOCA) transient due to fuel rod swelling, thermal expansion, or bowing. In the early phase of the transient following the coolant break, the high axial loads which potentially could be generated by the difference in thermal expansion between fuel clad and thimbles are relieved by slippage of the fuel rods through the grids. The relatively low drag force restraint on the fuel rods will only induce minor thermal bowing not sufficient to close the fuel rod-to-thimble tube gap. This rod-to-grid slip mechanism occurs simultaneously with control rod drop. 4.2-25 SGS-UFSAR Revision 18 April 26, 2000
Vibration and Wear The effect of the flow induced vibration on the V5H and Vantage+ fuel assembly and individual fuel rods is minimal. The cyclic stress range associated with deflections of such small magnitude is insignificant and has no effect on the structural integrity of the fuel rod. The conclusion that the effect of flow induced vibrations on the fuel assembly and fuel rod is minimal is based on test results and analysis documented in Reference 11. Full flow vibration tests have been performed on the RFA covering both assembly and fuel rod vibration. These tests have shown the Robust Fuel Design is not susceptible to flow induced assembly vibration and provides improved fuel rod vibration performance over the prior designs. In addition, the RFA-2 mid grid design provides further enhancements that improve fuel rod vibration performance over the RFA design. The reaction on the grid support due to vibration motions is also correspondingly small and much less than the spring preload. Firm contact is therefore maintained. No significant wear of the cladding or grid supports is expected during the life of the fuel assembly, based on out-of-pile flow tests, performance of similarly designed fuel in operating reactors (2), and design analyses. Clad fretting and fuel rod vibration have been experimentally investigated as shown in Reference 11. No significant guide thimble tube wear due to flow-induced vibration of the control rods is predicted. Based on a conservative wear analyses, Westinghouse concluded that the integrity of the guide tube is maintained during normal operation, accident conditions, and nonoperational loading condition for at least 250 weeks (> 3 cycles) of fuel assembly operation. The Nuclear Regulatory Commission (NRC) has concluded ( 12) that the Westinghouse analyses probably accounts for all the major variables in the wear process. However, the NRC requested additional confirmatory information supporting the absence of significant thimble wear (no wear hole formation) for the 17 x 17 fuel assembly design. Examination of 1434 guide thimble tubes in six fuel assemblies examined at Salem Unit 1 shows no wear hole formation. Four of the assemblies had control rods in the parked 4.2-26 SGS-UFSAR Revision 25 October 26, 2010
position (7 1/2 inches into the guide thimble) for Cycle 1, and two assemblies had control rods parked for Cycles 1 and 2. The parked position of the control rods has the greatest potential for causing guide thimble wear due to flow induced vibrations. The results of this surveillance program satisfy the NRC request to verify the wear analysis conclusion of no wear holes. Evaluation of the Reactor Core for Limited Displacement RPV Inlet and Outlet Nozzle Breaks The STD fuel assembly response resulting from the most limiting main coolant pipe break was analyzed using time history numerical techniques. Since the resulting vessel motion induces primarily lateral loads on the reactor core, a finite element model similar to the seismic model described in Reference 10 was used to assess the fuel assembly deflections and impact forces. The reactor core finite element model which simulates the fuel assembly interaction during lateral excitation consists of fuel assemblies arranged in a planer array with inter-assembly gaps. For the Salem Station, 15 fuel assemblies which correspond to the maximum number of assemblies across the core diameter were used in the mode. The fuel assemblies and the reactor baffle support are represented by single beam elements as shown on Figure 4.2-25. The time history motion for the upper and lower core plates and the barrel at the upper core plate elevation are simultaneously applied to the simulated reactor core model as illustrated on Figure 4. 2-25. The three time history motions were obtained from the time history analysis of the reactor vessel and internals finite element model. 4.2-27 SGS-UFSAR Revision 17 October 16, 1998
The fuel assembly response, namely the displacements and grid impact forces, were obtained from the reactor core model using the core plate and barrel motions resulting from a reactor coolant pump outlet double ended break. The maximum fuel assembly deflection was determined to occur in a peripheral fuel assembly. The fuel assembly stresses resulting from this deflection indicated significant safety margins compared to the allowable values. The grid maximum impact force for both the seismic and lateral blowdown accident conditions occurred at the peripheral fuel assembly locations adjacent to the baffle wall. The grid impact forces were appreciably lower for fuel assembly locations inward from the peripheral fuel. For the lateral blowdown case, only a small (outer) portion of the core experienced significant grid impact forces. The maximum grid impact force obtained from the limiting rupture break was found to be less than the minimum grid strength (using the 95 x 95 value as determined by tests at reactor operating safe shutdown temperatures) . The maximum square-root-of-the-sum-of-the-squares combination of the pipe rupture and safe shutdown earthquake loads for the limiting grid location was found to be less than the minimum grid strength. The major components that determine the structural integrity of the fuel assembly are the grids. Mechanical testing and analysis of the Vantage 5H Zircaloy grid and fuel assembly have demonstrated that the Vantage 5H structural integrity under seismic/LOCA loads will provide margins comparable to the STD 17 x 17 fuel assembly design and will meet all design bases. Since the Vantage+ assembly is structurally similar to that of Vantage 5H, the seismic and LOCA analysis for the Vantage 5H assembly are applicable to Vantage+ assembly. The use of ZIRLOTM guide thimbles will not affect the seismic and LOCA loads. A Salem plant specific seismic and LOCA analysis was performed for the RFA (Reference 22). The results of this analysis demonstrate that the RFA and V5H fuel assembly designs are capable of maintaining a coolable core geometry and control rod insertability under the combined seismic and LOCA loading for both a homogeneous and mixed core of fuel assembly designs. 4.2.1.3.3 Operational Experience Westinghouse has had considerable experience with Zircaloy-clad fuel since its introduction in the Jose Cabrera plant in June 1968. This experience is extensively described in Reference 2. 4.2-28 SGS-UFSAR Revision 22 May 5, 2006
4.2.1.3.4 Test Rod and Test Assembly Experience. This experience is presented in Sections 8 and 23 of Reference 3. 4.2.1.4 Testing and Inspection Plan 4.2.1.4.1 Quality Assurance Program The quality assurance program plan of the Westinghouse Nuclear Fuel Division for Salem is summarized in Reference 13. 4.2-28a SGS-UFSAR Revision 11 July 22, 1991
THIS PAGE INTENTIONALLY BLANK 4.2-28b SGS-UFSAR Revision 11 July 22, 1991
The program provides for control over all activities affecting product quality, commencing with design and development, and continuing through procurement, materials handling, fabrication, testing and inspection, storage, and transportation. The program also provides for the indoctrination and training of personnel and for the auditing of activities affecting product quality through a formal auditing program. Westinghouse drawings and product, process, and material specifications identify the inspections to be performed. 4.2.1.4.2 Quality Control Quality control (QC) philosophy is generally based on the following inspections being performed to a 95 percent confidence that at least 95 percent of the product meets specification, unless otherwise noted. Fuel System Components and Parts The characteristics inspected depend upon the component parts; the QC program includes dimensional and visual examinations, check audits of test reports, material certification, and nondestructive examination, such as X-ray and ultrasonic. Pellets Inspection is performed for dimensional characteristics such as diameter, density, length, and squareness of ends. Additional visual inspections are performed for cracks, chips, and surface conditions according to approved standards. Density is determined in terms of weight per unit length and is plotted on zone charts used in controlling the process. Chemical analyses are taken on a specified sample basis throughout pellet production. 4.2-29 SGS-UFSAR Revision 6 February 15, 1987
Rod Inspection The fuel rod inspection consists of the following nondestructive examination techniques and methods, as applicable:
1. Each rod is leak tested using a calibrated mass spectrometer, with helium being the detectable gas.

2. Rod welds are inspected by ultrasonic test or X-ray in accordance with a qualified technique and Westinghouse specifications.

3. All rods are dimensionally inspected prior to final release. The requirements include such items as length, camber, and visual appearance.

4. All fuel rods are inspected by gamma scanning or other approved methods to ensure proper plenum dimensions.

5. All fuel rods are inspected by gamma scanning or other approved methods to ensure that no significant gaps exist between pellets.

6. All fuel rods are active gamma scanned to verify enrichment control prior to acceptance for assembly loading.

7. Traceability of rods and associated rod components is established by QC.
Assemblies Each fuel assembly is inspected for compliance with drawing and/or specification requirements. Other incore control component inspection and specification requirements are given in Section 4.2.3.4. 4.2-30 SGS-UFSAR Revision 8 July 22, 1988
Other Inspections The following inspections are performed as part of the routine inspection operation:
1. Tool and gage inspection and control, including standardization to primary and/ or secondary working standards. Tool inspection is performed at prescribed intervals on all serialized tools. Complete records are kept of calibration and conditions of tools.

2. Audits are performed of inspection activities and records to ensure that prescribed methods are followed and that records are correct and properly maintained.

3. Surveillance inspection, where appropriate, and audits of outside contractors are performed to ensure conformance with specified requirements.
Process Control To prevent the possibility of mixing enrichments during fuel manufacture and assembly, strict enrichment segregation and other process controls are exercised. The uranium dioxide powder is kept in sealed containers. The contents are fully identified both by descriptive tagging and preselected color coding. A Westinghouse identification tag completely describing the contents is affixed to the containers before transfer to powder storage. Isotopic content is confirmed by analysis. Powder withdrawal from storage can be made by only one authorized group, which directs the powder to the correct pellet production line. All pellet production lines are physically separated from each other and pellets of only a single nominal enrichment are produced in a given production line at any given time. 4.2-31 SGS-UFSAR Revision 6 February 15, 1987
Finished pellets are placed on trays identified with the same color code as the powder containers and transferred to segregated storage racks within the confines of the pelleting area. Samples from each pellet lot are tested for isotopic content and impurity levels prior to acceptance by QC. Physical barriers prevent mixing of pellets of different nominal densities and enrichments in this storage area. Unused powder and substandard pellets are returned to storage in the original color-coded containers. Loading of pellets into the clad is performed in isolated production lines, and again only one enrichment is loaded on a line at a time. A serialized traceability code is placed on each fuel tube to provide unique identification. The end plugs are inserted and then inert-welded to seal the tube. The fuel tube remains coded and traceability identified until just prior to installation in the fuel assembly. At the time of installation into an assembly, the traceability codes are removed and a matrix is generated to identify each rod in its positions within a given assembly. The top nozzle is inscribed with a permanent identification number providing traceability to the fuel contained in the assembly. Similar traceability is provided for burnable poison, source rods, and control rods, as required. 4.2.1.4.3 Onsite Inspection Surveillance of fuel and reactor performance is routinely conducted on Westinghouse reactors. Power distribution is monitored using the ex-core fixed and in-core movable detectors, and the BEACON (Best Estimate Analyzer for Core Operations Nuclear) on-line core monitoring system. BEACON is also known as the Power Distribution Monitoring System (PDMS) in the Technical Specifications. Coolant activity and chemistry is followed which permits early detection of any fuel clad defects. 4.2-32 SGS-UFSAR Revision 19 November 19, 2001
Visual fuel inspection is routinely conducted during refueling. Additional fuel inspections are dependent on the results of the operational monitoring and the visual inspections. 4.2.2 Reactor Vessel Internals 4.2.2.1 Design Bases The design bases for the mechanical design of the reactor vessel internals components are as follows:
1. The reactor internals, in conjunction with the fuel assemblies, shall direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation. In addition, required cooling for the pressure vessel head shall be provided so that the temperature differences between the vessel flange and head do not result in leakage from the flange during reactor operation.

2. In addition to neutron shielding provided by the reactor coolant, a separate thermal shield is provided to limit the exposure of the pressure vessel in order to maintain the required ductility of the material for all modes of operation.

3. Provisions shall be made for installing in-core instrumentation useful for the plant operation and vessel material test specimens required for a pressure vessel irradiation surveillance program.

4. The core internals are designed to withstand mechanical loads arising from operating basis earthquake (OBE), design basis earthquake (DBE), and pipe ruptures and meet the requirement of Item 5 below.
4.2-33 SGS-UFSAR Revision 6 February 15, 1987
5. The reactor shall have mechanical provisions which are sufficient to adequately support the core and internals and to assure that the core is intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.

6. Following the design basis accident (DBA), the plant shall be capable of being shut down and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling.
The functional limitations for the core structures during the DBA are shown in Table 4.2-1. To ensure column loading of rod cluster control guide tubes, the upper core plate deflection is limited to not exceed the value shown in Table 4.2-1. 4.2.2.2 Description and Drawings The reactor vessel internals are described as follows: The components of the reactor internals consist of the lower core support structure (including the entire core barrel and thermal shield), the upper core support structure, and the in-core instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and control rod drive mechanisms, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding, and guides for the in-core instrumentation. The coolant flows from the vessel inlet nozzles down the annulus between the core barrel and the vessel wall and then into a plenum at the bottom of the vessel. It then reverses and flows up through the core support and through the lower core plate. Flow passages in the lower core plate are sized to provide the desired inlet flow distribution to the core. After passing through the 4.2-34 SGS-UFSAR Revision 6 February 15, 1987
core, the coolant enters the region of the upper support structure and then flows radially to the core barrel outlet nozzles and directly through the vessel outlet nozzles. A small portion of the coolant flows between the baffle plates and the core barrel to provide additional cooling of the barrel. Similarly, a small amount of the entering flow is directed into the vessel head plenum and exits through the vessel outlet nozzles. All the major material for the reactor internals is Type 304 stainless steel. Parts not fabricated from Type 304 stainless steel include bolts and dowel pins which are fabricated from Type 316 stainless steel and the radial support clevis inserts and bolts which are fabricated of Inconel. The only stainless steel materials used in the reactor core support structures which have yield strengths greater than 90,000 pounds are the 403 series used for holddown springs. The use of these materials is compatible with the reactor coolant and is acceptable based on the 1971 ASME Boiler and Pressure Vessel Code, Case Number 1337. All reactor internals are removable from the vessel for the purpose of their inspection as well as the inspection of the vessel internal surface. Lower Core Support Structure The major containment and support member of the reactor internals is the lower core support structure, shown on Figure 4.2-8. This support structure assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the thermal shield, and the core support which is welded to the core barrel. All the major material for this structure is Type 304 stainless steel. The lower core support structure is supported at its upper flange from a ledge in the reactor vessel and its lower end is restrained from transverse motion by a radial support system attached to the vessel wall. Within the core barrel are an axial baffle and a lower core plate, both of which are attached to the core barrel wall and form the enclosure periphery of the core. 4.2-35 SGS-UFSAR Revision 6 February 15, 1987
The lower core support structure and core barrel serve to provide passageways and direct the coolant flow. The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies. The lower core plate is a member through which the necessary flow distribution holes for each fuel assembly are machined. Fuel assembly locating pins (two for each assembly) are also inserted into this plate. Columns are placed between this plate and the core support of the core barrel in order to provide stiffness and to transmit the core load to the core support. Adequate coolant distribution is obtained through the use of the lower core plate and core support. The one-piece thermal shield is fixed to the core barrel at the top with rigid bolted connections. The bottom of the thermal shield is connected to the core barrel by means of axial flexures. This bottom support allows for differential axial growth of the shield/core barrel but restricts radial or horizontal movement of the bottom of the shield. Rectangular specimen guides in which material samples can be inserted and irradiated during reactor operation are welded to the thermal shield and extended to the top of the thermal shield. These samples are held in the rectangular specimen guides by a preloaded spring device at the top and bottom. Vertically downward loads from weight, fuel assembly preload, control rod dynamic loading, hydraulic loads, and earthquake acceleration are carried by the lower core plate into the lower core plate support flange on the core barrel shell and through the lower support columns to the core support and thence through the core barrel shell to the core barrel flange supported by the vessel flange. Transverse loads from earthquake acceleration, coolant cross flow, and vibration are carried by the core barrel shell and distributed between the lower radial support to the vessel wall, and to the vessel flange. Transverse loads of the fuel assemblies are transmitted to the core barrel shell by direct 4.2-36 SGS-UFSAR Revision 6 February 15, 1987
connection of the lower core plate to the barrel wall and by upper core plate alignment pins which are welded into the core barrel. The radial support system of the core barrel is accomplished by "key" and "keyway" joints to the reactor vessel wall. At six equally spaced points around the circumference, an Inconel clevis block is welded to the vessel inner diameter. An Inconel insert block is bolted to each of these clevis blocks, and has a "keyway" geometry. Opposite each of these is a "key" which is welded to the lower core support. At assembly, as the internals are lowered into the vessel, the keys engage the keyways in the axial direction. With this design, the internals are provided with a support at the furthest extremity, and may be viewed as a beam fixed at the top and simply supported at the bottom. Radial and axial expansion of the core barrel are accommodated, but transverse movement of the core barrel is restricted by this design. With this system, cyclic stresses in the internal structures are within the ASME Section III limits. In the event of an abnormal downward vertical displacement of the internals following a hypothetical failure, energy absorbing devices limit the displacement of the core after contacting the vessel bottom head. The load is then transferred through the energy absorbing devices of the lower internals to the vessel. The energy absorbers are mounted on a base plate which is contoured on its bottom surface to the reactor vessel bottom internal geometry. Their number and design are determined so as to limit the stresses imposed on all components except the energy absorber to less than yield (ASME Code Section III valves). Assuming a downward vertical displacement, potential energy of the system is absorbed mostly by the strain energy of the energy absorbing devices. 4.2-37 SGS-UFSAR Revision 6 February 15, 1987
Upper Core Support Assembly The upper core support assembly, shown on Figures 4.2-10 and 4.2-12, consists of the upper support assembly and the upper core plate between which are contained support columns and guide tube assemblies. The support columns establish the spacing between the upper support assembly and the upper core plate and are fastened at the top and bottom to these plates. The support columns transmit the mechanical loadings between the upper support and upper core plate. The guide tube assemblies shield and guide the control rod drive shafts and control rods. They are fastened to the upper support and are guided by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the upper guide tube which is attached to the upper support. The upper core support assembly, which is removed as a unit during refueling operation, is positioned in its proper orientation with respect to the lower support structure by slots in the upper core plate which engage flat-sided upper core plate alignment pins which are welded into the core barrel. At an elevation in the core barrel where the upper core plate is positioned, the flat-sided pins are located at angular positions of 90 degrees from each other. As the upper support structure is lowered into the lower internals, the slots in the plate engage the flat-sided pins in the axial direction. Lateral displacement of the plate and of the upper support assembly is restricted by this design. Fuel assembly locating pins protrude from the bottom of the upper core plate and engage the fuel assemblies as the upper assembly is lowered into place. Proper alignment of the lower core support structure, the upper core support assembly, the fuel assemblies and control rods are thereby assured by this system of locating pins and guidance arrangement. The upper core support assembly is restrained from any axial movements by a large circumferential spring which rests between the upper barrel flange and the upper 4.2-38 SGS-UFSAR Revision 6 February 15, 1987
core support assembly. The spring is compressed when the reactor vessel head is installed on the pressure vessel. Vertical loads from weight, earthquake acceleration, hydraulic loads, and fuel assembly preload are transmitted through the upper core plate via the support columns to the upper support assembly and then into the reactor vessel head. Transverse loads from coolant cross flow, earthquake acceleration, and possible vibrations are distributed by the support columns to the upper support and upper core plate. The upper support plate is particularly stiff to minimize deflection. In-Core Instrumentation Support Structures All bottom-mounted in-core instrumentation support structures consist of a system to convey and support flux thimbles penetrating the vessel through the bottom (Figure 7. 7-6 shows the Basic Flux-Mapping System) . Specifically, the flux thimbles enter the reactor vessel through the bottom penetration nozzles. Conduits extend from the bottom of the reactor vessel down through the concrete shield area and up to a thimble seal line. The minimum bend radii are about 144 inches and the trailing ends of the thimbles (at the seal line) are extracted approximately 15 feet 4.2-39 SGS-UFSAR Revision 11 July 22, 1991
during refueling of the reactor in order to avoid interference within the core. The thimbles are closed at the leading ends and serve as the pressure barrier between the reactor pressurized water and the containment atmosphere. Mechanical seals between the retractable thimbles and conduits are provided at the seal line. During normal operation, the retractable thimbles are stationary and move only during refueling or for maintenance, at which time a space of approximately 15 feet above the seal line is cleared for the retraction operation. The in-core instrumentation support structure is designed for adequate support of instrumentation during reactor operation and is rugged enough to resist damage or distortion under the conditions imposed by handling during the refueling sequence. These are the only conditions which affect the in-core instrumentation support structure. Reactor vessel surveillance specimen capsules are covered in Section 4.5.1. As part of the conversion to an all bottom mounted in-core instrumentation, the support structures for the original top entry core exit thermocouple system have been removed. Specifically, the five thermocouple columns on the upper support plate have been removed and their corresponding reactor vessel head penetrations have been cut and capped. However, the support bases for the columns were left on the upper support plate so as not to create any flow openings across the upper support plate. The core exit thermocouple system went from a top entry system to a bottom entry system for the following reasons as stated in letter NPE-85-1035:
1. Resistance readings were taken on the thermocouples and many were found to have shorts or were declared inoperable. These would provide erroneous indication at normal operating conditions.
4.2-40 SGS-UFSAR Revision 11 July 22, 1991
2. Thermocouple columns were bent/damaged several times during refueling operations and not all thermocouples were recoverable. The installation of bottom-mounted thermocouples provided an easier method of reactor disassembly and lower probability of damage.

3. Installation of a bottom entry system would also allow the elimination of the five instrument ports and resolve the numerous problems encountered with the qualification of the reference junction boxes and thermocouples on the top-mounted system.
4.2.2.3 Design Loading Conditions The design loading conditions that provide the basis for the design of the reactor internals are:
1. Fuel Assembly Weight

2. Fuel Assembly Spring Forces

3. Internals Weight

4. Control Rod Scram (equivalent static load)

5. Differential Pressure

6. Spring Preloads 4.2-40a SGS-UFSAR Revision 11 July 22, 1991
THIS PAGE INTENTIONALLY BLANK 4.2-40b SGS-UFSAR Revision 11 July 22, 1991
7. Coolant Flow Forces (static)

8. Temperature Gradients

9. Differences in Thermal Expansion

a. Due to temperature differences

b. Due to expansion of different materials

10. Interference Between Components

11. Vibration (mechanically or hydraulically induced)

12. One or More Loops Out of Service

13. All Operational Transients Listed in Table 4.1-10

14. Pump Overspeed

15. Seismic Loads (OBE and DBE)

16. Blowdown Forces (due to cold and hot leg break)
Combined seismic and blowdown forces are included in the stress analysis as a design loading condition by assuming the maximum amplitude of each force to act concurrently. The main objectives of the design analysis are to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Both low and high cycle fatigue 4.2-41 SGS-UFSAR Revision 6 February 15, 1987
stresses are considered when the allowable amplitude of oscillation is established. As part of the evaluation of design loading conditions, extensive testing and inspection are performed from the initial selection of raw materials up to and including component installation and plant operation. Among these tests and inspections are those performed during component fabrication, plant construction, startup and checkout, and during plant operation. 4.2.2.4 Design Loading Categories The combination of design loadings fits into either the normal, upset, or faulted conditions as defined in the ASME Section III Code. Loads and deflections imposed on components due to shock and vibration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces, and thermal gradients for the determination of the total stresses of the internals. The reactor internals are designed to withstand stresses originating from various operating conditions summarized in Table 5.1-10. The scope of the stress analysis problem is very large requiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of operation under consideration. Allowable Deflections For normal operating conditions, downward vertical deflection of the lower core support plate is negligible. 4.2-42 SGS-UFSAR Revision 6 February 15, 1987
For the LOCA plus the DBE condition, the deflection criteria of critical internal structures are the limiting values given in Table 4.2-1. The corresponding no loss of function limits are included in Table 4.2-1 for comparison purposes with the allowed criteria. The criteria for the core drop accident are based upon analyses which have been performed to determine the total downward displacement of the internal structures following a hypothesized core drop resulting from loss of the normal core barrel supports. The initial clearance between the secondary core support structures and the reactor vessel lower head in the hot condition is approximately 1/2 inch. An additional displacement of approximately 3/4 inch would occur due to strain of the energy absorbing devices of the secondary core support; thus the total drop distance is about 1 1/4 inches which is insufficient to permit the grips of the rod cluster control assembly to come out of the guide thimble in the fuel assemblies. Specifically, the secondary core support is a device which will never be used, except during a hypothetical accident of the core support (core barrel, barrel flange, etc.). There are four supports in each reactor. This device limits the fall of the core and absorbs the energy of the fall which otherwise would be imparted to the vessel. The energy of the fall is calculated assuming a complete and instantaneous failure of the primary core support and is absorbed during the plastic deformation of the controlled volume of stainless steel, loaded in tension. The maximum deformation of this austenitic stainless piece is limited to approximately 15 percent, after which a positive step is provided to ensure support. 4.2.2.5 Design Criteria Basis The basis for the design stress and deflection criteria is identified below. 4.2-43 SGS-UFSAR Revision 6 February 15, 1987
Allowable Stress For normal operating conditions, Section III of the ASME Nuclear Power Plant Components Code is used as a basis for evaluating acceptability of calculated stresses. Both static and alternating stress intensities are considered. Under Code Case 1618, bolt material Type 316 stainless steel is now covered in ASME Section III and is so treated. It should be noted that the allowable stresses in Section III of the ASME Code are based on unirradiated material properties. In view of the fact that irradiation increases the strength of the Type 304 stainless steel used for the internals, although decreasing its elongation, it is considered that use of the allowable stresses in Section III is appropriate and conservative for irradiated internal structures. The allowable stress limits during the DBA used for the core support structures are based on the January 1971 draft of the ASME Code for Core Support Structures, Subsection NG, and the Criteria for Faulted Conditions. 4.2.3 Reactivity Control System 4.2.3.1 Design Bases Bases for temperature, stress on structural members, and material compatibility are imposed on the design of the reactivity control components. 4.2.3.1.1 Design Stresses The Reactivity Control System is designed to withstand stresses originating from various operating conditions as summarized in Table 5.2-10. Allowable Stresses: For normal operating conditions, Section III of the ASME Boiler and Pressure Code is used as a general guide. 4.2-44 SGS-UFSAR Revision 18 April 26, 2000
Dynamic Analysis: The cyclic stresses due to dynamic loads and deflections are combined with the stresses imposed by loads from component weights, hydraulic forces, and thermal gradients for the determination of the total stresses of the Reactivity Control System. 4.2.3.1.2 Material Compatibility Materials are selected for compatibility in a PWR environment, for adequate mechanical properties at room and operating temperature, for resistance to adverse property changes in a radioactive environment, and for compatibility with interfacing components. 4.2.3.1.3 Reactivity Control Components The reactivity control components are subdivided into two categories:
1. Permanent devices used to control or monitor the core

2. Temporary devices used to control or monitor the core.
The permanent type components are the rod cluster control assemblies, control rod drive assemblies, neutron source assemblies, and thimble plug assemblies. Although the thimble plug assembly does not directly contribute to the reactivity control of the reactor, it is presented as a Reactivity Control System component in this document because it can be used to restrict bypass flow through those thimbles not occupied by absorber, source, or burnable absorber rods. The temporary component is the burnable absorber assembly. The design bases for each of the mentioned components are in the following paragraphs. 4.2-45 SGS-UFSAR Revision 11 July 22, 1991
Absorber Rods The following are considered design conditions under Subsections NG and NB of the ASME Boiler and Pressure Vessel Code Section III.
1. The external pressure equal to the Reactor Coolant System operating pressure

2. The wear allowance equivalent to 1,000 reactor trips

3. Bending of the rod due to a misalignment in the guide tube

4. Forces imposed on the rods during rod drop

5. Loads caused by accelerations imposed by the control rod drive mechanism

6. Radiation exposure for maximum core life

7. Temperature effects at operating conditions The absorber material temperature shall not exceed its melting temperature (1470°F for Ag-In-Cd absorber material) (14)
Burnable Absorber Rods Two kinds of discrete burnable absorber rods may be used at Salem. The first is the borosilicate glass (PYREX) burnable absorber. The second is the Wet Annular Burnable Absorber (WABA) design. See References 23 and 24 of this section for a more detailed discussion of WABA. Reference 24 extends the allowable lifetime of the WABA rods from 18,000 to 40,000 EFPH, and allows utilization of the WABA rods for up to two eighteen month cycles without requiring additional inspections of the WABA rods. Discrete burnable absorbers are utilized to meet nuclear design requirements of UFSAR Section 4. 3. Some comparative data is also shown in UFSAR Table 4.3-1. The burnable absorber rod clad is designed using Subsections NG and NB of the ASME Boiler and Pressure Vessel Code, Section III, 1973 as a general guide for Conditions I and I I . For abnormal loads during Conditions I I I and IV, Code stresses are not considered limiting. Failures of the burnable absorber rods during these conditions must 4.2-46 SGS-UFSAR Revision 23 October 17, 2007
not interfere with reactor shutdown or emergency cooling of the fuel rods. The burnable absorber material is nonstructural. The structural elements of the burnable absorber rods are designed to maintain the absorber geometry even if the absorber material is fractured. The PYREX rods are designed so that the borosilicate absorber material is below its softening temperature ( 1492°F* for reference 12.5 weight percent boron rods). The WABA rods utilize an aluminum oxide/boron carbide (Al 20 3 -B 4C) absorber material which is a sintered ceramic and has a very high melting temperature. In addition, the structural elements are designed to prevent excessive slumping. Neutron Source Rods The neutron source rods are designed to withstand the following:
1. The external pressure equal to the Reactor Coolant System operating pressure

2. An internal pressure equal to the pressure generated by released gases over the source rod life Thimble Plug Assembly If used in the core (optional), the thimble plug assemblies satisfy the following:

1. Accommodate the differential thermal expansion between the fuel assembly and the core internals
Borosilicate glass is accepted for use in burnable absorber rods if the softening temperature is 1510 + 18°F. The softening temperature is defined in ASTM C 338.
4.2-47 SGS-UFSAR Revision 18 April 26, 2000
2. Maintain positive contact with the fuel assembly and the core internals

3. Limit the flow through each occupied thimble to acceptable design value 4.2.3.1.4 Control Rod Drive Mechanisms The mechanisms are Class I components designed to meet the stress requirements for normal operating conditions of Section III of the ASME Boiler and Pressure Vessel Code. Both static and alternating stress intensities are considered.
The stresses originating from the required design transients are included in the analysis. A dynamic seismic analysis is required on the control rod drive mechanism (CRDM) when a seismic disturbance has been postulated to confirm the ability of the mechanism to meet ASME Code, Section III allowable stresses and to confirm its ability to trip when subjected to the seismic disturbance. The CRDM design used for the 17 x 17 fuel assembly control rod is identical to the 15 x 15 CRDM. The seismic analysis and response of 17 x 17 CRDM will be identical to those of the 15 x 15 mechanisms. Control Rod Drive Mechanism Operational Requirements The basic operational requirements for the CRDMs are as follows:
1. 5/8-inch step

2. 150-inch travel

3. 360-pound maximum load

4. Step in or out at 45 inches/minute (72 steps/minute) 4.2-48 SGS-UFSAR Revision 6 February 15, 1987

5. Power interruption shall initiate release of drive rod assembly

6. Trip delay of less than 150 ms - Free fall of drive rod assembly shall begin less than 150 ms after power interruption no matter what holding or stepping action is being executed with any load and coolant temperatures of 100°F to 550°F

7. 40-year design life with normal refurbishment 6

8. 28,00 complete travel excursions which is 13 x 10 steps with normal refurbishment 4.2.3.2 Design Description Reactivity control is provided by neutron absorbing rods and a soluble chemical neutron absorber (boric acid) The boric acid concentration is varied to control long-term reactivity changes such as:

1. Fuel depletion and fission product buildup

2. Cold to hot, zero power reactivity change

3. Reactivity change produced by intermediate-term fission products such as xenon and samarium

4. Burnable absorber depletion The rod cluster control assemblies provide reactivity control for:

1. Shutdown

2. Reactivity changes due to coolant temperature changes in the power range 4.2-49 SGS-UFSAR Revision 11 July 22, 1991

3. Reactivity changes associated with the power coefficient of reactivity

4. Reactivity changes due to void formation If soluble boron were the sole means of control, the moderator temperature coefficient could be positive. It is desirable to have a negative moderator temperature coefficient throughout the entire cycle in order to reduce possible deleterious effects caused by a positive coefficient during loss-of-coolant or loss-of-flow accidents. This is accomplished by installation of burnable absorber assemblies.
The neutron source assemblies and spontaneous fission neutron sources associated with the irradiated fuel assemblies provide a means of monitoring the core during periods of low neutron activity. The most effective reactivity control components are the rod cluster control assemblies and their corresponding drive rod assemblies which are the only kinetic parts in the reactor. Figure 4.2-13 identified the rod cluster control and drive rod assembly, in addition to the arrangement of these components in the reactor relative to the interfacing fuel assembly, guide tubes, and CRDM. The guidance system for the control rod cluster is provided by the guide tube as shown on Figure 4.2-13. The guide tube provides two regimes of guidance. In the lower section a continuous guidance system provides support immediately above the core. This system protects the rod against excessive deformation and wear due to hydraulic loading. The region above the continuous section provides support and guidance at uniformly spaced intervals. The envelope of support is determined by the pattern of the control rod cluster as shown on Figure 4.2-13. The guide tube assures alignment and support of the control rods, spider body, 4.2-50 SGS-UFSAR Revision 25 October 26, 2010
and drive rod while maintaining trip times at or below required limits. In the following paragraphs, each reactivity control component is described in detail. 4.2.3.2.1 Reactivity Control Components Rod Cluster Control Assembly The rod cluster control assemblies are divided into two categories: control and shutdown. The control groups compensate for reactivity changes due to variations in operating conditions of the reactor, i.e., power and temperature variations. Two criteria have been employed for selection of the control groups. First, the total reactivity worth must be adequate to meet the nuclear requirements of the reactor. Second, in view of the fact that some of these rods may be partially inserted at power operation, the total power peaking factor should be low enough to ensure that the power capability is met. The control and shutdown groups provide adequate shutdown margin which is defined as the amount by which the core would be subcritical at hot shutdown if all rod cluster control assemblies are tripped assuming that the highest worth assembly remains fully withdrawn and assuming no changes in xenon or boron concentration. A rod cluster control assembly comprises a group of individual neutron absorber rods fastened at the top end to a common spider assembly, as illustrated on Figure 4.2-14. The absorber material used in the control rods is silver-indium-cadmium single piece absorber rod which is essentially "black" to thermal neutrons and has sufficient additional resonance absorption to significantly increase its worth. The alloy is in the form of extruded rods which are sealed in stainless steel tubes to prevent the rods from coming in direct contact with the coolant. In construction, the silver-indium-cadmium rods are inserted into cold-worked stainless steel tubing which is then sealed at the bottom and the top by welded Type 308L stainless steel end plugs as shown on Figure 4. 2-15. Sufficient diametral and end clearance is provided to accommodate relative thermal expansions. The cladding surface has been ion-nitrided for hardening and corrosion resistance. The bottom plugs are made bullet-nosed to reduce the hydraulic drag during reactor trip and to guide smoothly into the dashpot section of the fuel assembly guide thimbles. The upper end plug is threaded for assembly to the spider and is machined with a reduced diameter shank to provide flexibility to the joint for any misalignment condition. 4.2-51 SGS-UFSAR Revision 17 October 16, 1998
The spider assembly is a one-piece machined casting in the form of a central hub with radial vanes containing cylindrical fingers from which the absorber rods are suspended. Handling detents and detents for connection to the drive rod assembly are machined into the upper end of the hub. A coil spring inside the spider body absorbs the impact energy at the end of a trip insertion. A centerpost which holds the spring and its retainer is threaded into the hub within the skirt and welded to prevent loosening in service. The spider casting material is CF3M cast 316 stainless steel. The absorber rods are fastened securely to the spider assembly as shown in Figure 4.2-15 to assure trouble free service. The threaded end of the upper end plug is inserted into the bottom of the spider boss hole. A nut is tightened on and welded to the spider boss to prevent loosening. A lock pin is inserted into the aligned holes of the spider base and upper end plug and welded to prevent the end plug and rod from backing off. The overall length is such that when the assembly is withdrawn through its full travel the tips of the absorber rods remain engaged in the guide thimbles so that alignment between rods and thimbles is always maintained. Since the rods are long and slender, they are relatively free to conform to any small misalignments with the guide thimble. 4.2-52 SGS-UFSAR Revision 17 October 16, 1998
Burnable Absorber Assembly Each burnable absorber assembly consists of burnable absorber rods attached to a holddown assembly. Burnable absorber assemblies are shown on Figure 4.2-16. The PYREX absorber rods consist of borosilicate glass tubes contained within Type 304 stainless steel tubular cladding which is plugged and seal welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin wall tubular inner liner of Type 304 stainless steel. The top end of the liner is open to permit the diffused helium to pass into the void volume and the liner overhangs the glass. The liner has an outward flange at the bottom end to maintain the position of the liner with the glass. A PYREX burnable absorber rod is shown in longitudinal and transverse cross sections on Figure 4.2-17. The WABA consist of aluminum oxide/boron carbide pellets contained within Zircaloy tubular cladding which is plugged and seal welded at the ends to encapsulate the pellets. The pellets are also supported along the length of their insider diameter by a thin wall tubular liner of Type 304 stainless steel. The top and bottom of the inner tube, or liner, is open to allow for coolant flow. There is a void above the Al 2 0 3 -B 4 C pellets between the inner and outer tube to contain diffused helium. A cross section drawing of the WABA rod is shown in Figure 4.2-17A. The rods are statically suspended and positioned in selected guide thimbles within specified fuel assemblies. The absorber rods in each fuel assembly are grouped and attached together at the top end of the rods to a holddown assembly by a flat, perforated retaining plate which fits within the fuel assembly top nozzle and rests on the adaptor plate. The retaining plate (and the absorber rods) is held down and restrained against vertical motion through a spring pack which is attached to the plate and is compressed by the upper core plate when the reactor upper internals assembly is lowered into the reactor. This arrangement assures that the absorber rods cannot be ejected from the core by flow forces. Each rod is permanently attached to the base plate by a nut which is lock welded into place. The clad in the rod assemblies is Zircaloy or slightly cold worked Type 304 stainless steel. All other structural materials are Type 304 or 308 stainless steel except for the springs which are Inconel 718. The borosilicate glass tube and I or aluminum oxide/boron carbide pellets provide sufficient boron content to meet the criteria discussed in Section 4.3.1. 4.2-53 SGS-UFSAR Revision 19 November 19, 2001
Neutron Source Assembly A neutron source assembly can be used to provide a base neutron level to monitor core multiplication to changes in core reactivity. Since there is very little neutron activity when the core is subcri tical, such as refueling and approach to criticality, neutron source assemblies are placed in the reactor if required to provide a count rate greater than 1 cps in Mode 3 prior to starting the approach to critical on the source range monitors. The source range monitors, which receive their signal from the source range detectors, are used primarily when the core is subcritical and during special subcritical modes of operation. In addition to having a greater than 1 cps count rate, the source channels should maintain a signal to noise ratio of at least two in Mode 3 prior to beginning the approach to critical. The source assembly also supplements detection of changes in the core multiplication factor during core loading, refueling and approach to criticality. This can be done since the multiplication factor is related to an inverse function of the detector count rate. Therefore a change in the multiplication factor can be detected during addition of fuel assemblies while loading the core, a change in control rod positions, and changes in boron concentration. Both primary and secondary neutron source rods are used. The primary source rod, containing a radioactive material, spontaneously emits neutrons during the initial core loading and reactor startup. After the primary source rod decays beyond the desired neutron flux level, neutrons are then supplied by the secondary source rod. The secondary source rod contains a stable material which must be activated by neutron bombardment during reactor operation. The activation results in the subsequent release of neutrons. This becomes a source of neutrons during periods of low neutron flux, such as during refueling and subsequent startups. The initial reactor core employs four source assemblies, two primary source assemblies, and two secondary source assemblies. Each primary source assembly contains one primary source rod and between 0 and 23 burnable absorber rods. Each secondary source assembly contains a symmetrical grouping of 4 or 6 secondary source rods and between 0 and 2 0 burnable absorber rods. The 4 rodlet secondary source utilizes a single encapsulated design. The 6 rodlet secondary source utilizes a double encapsulated design which provides additional margin against source material leakage. Locations not filled with a source or burnable absorber rod may contain a thimble plug (optional). Source assemblies are shown on Figures 4.2-18 and 4.2-19. A comparison of the single and double encapsulated secondary source design is provided in Table 4.2-2. 4.2-54 SGS-UFSAR Revision 25 October 26, 2010
Following the initial operating cycle, the actual source range count rate will depend on the core loading and the outage duration. Normally for subsequent reloads, the primary sources are removed and the secondary sources continue functioning. The core loading determines the placement of the secondary sources and other inherent neutron sources (i.e., spontaneous fission from the irradiated fuel) relative to the location of the source range detectors. For secondary source 124 assemblies made of antimony-beryllium (Sb-Be, gamma-neutron reaction), sb has a 60.2 day half-life. Thus, extended outages may impact the effectiveness of the secondary sources. For such extended outages, inherent neutron sources are sufficient to provide the required greater than 1 cps count rate in Mode 3 prior to beginning the approach to critical. Likewise, if a reload design has sufficient spontaneous fission neutrons to ensure the minimum required count rate response on the source range channels, then there is no need to install the secondary source assemblies. The primary and secondary source rods both utilize the same cladding material as the absorber rods. The single encapsulated secondary source rods contain approximately 500 grams of Sb-Be pellets in one rod. The double encapsulated secondary source rods contain approximately 338 grams of Sb-Be pellets in one rod. The primary source rods contain capsules of Californium source material and alumina spacer rods to position the source material within the cladding. The rods in each assembly are permanently fastened at the top end to a holddown assembly, which is identical to that of the burnable absorber assemblies. The other structural members are constructed of Type 304 stainless steel except for the springs. The springs exposed to the reactor coolant are wound from an age hardened nickel base alloy for corrosion resistance and high strength. The springs, when contained within the rods where corrosion resistance is not necessary, are oil tempered carbon steel. Thimble Plug Assembly Thimble plug assemblies may be used in order to limit bypass flow through the rod cluster control guide thimbles in fuel assemblies which do not contain either control rods, source rods, or burnable absorber rods. The thimble plug assemblies as shown on Figure 4. 2-20 consist of a flat base plate with short rods suspended from the bottom surface and a spring pack assembly. 4.2-55 SGS-UFSAR Revision 25 October 26, 2010
The 24 short rods, called thimble plugs, project into the upper ends of the guide thimbles to reduce the bypass flow area. Similar short rods are also used on the source assemblies and burnable absorber assemblies to plug the ends of all vacant fuel assembly guide thimbles. At installation in the core, the thimble plug assemblies interface with both the upper core plate and with the fuel assembly top nozzles by resting on the adaptor plate. The spring pack is compressed by the upper core plate when the upper internals assembly is lowered into place. Each thimble plug is permanently attached to the base plate by a nut which is locked to the threaded end of the plug by a small lock-bar welded to the nut. All components in the thimble plug assembly, except for the springs, are constructed from Type 304 stainless steel. The springs are wound from an age hardened nickel base alloy for corrosion resistance and high strength. 4.2.3.2.2 Control Rod Drive Mechanism All parts exposed to reactor coolant are made of metals which resist the corrosive action of the water. Three types of metals are used exclusively: stainless steels, Inconel and cobalt based alloys. Wherever magnetic flux is carried by parts exposed to the main coolant, 400 series stainless steel is used. Cobalt based alloys are used for the pins and latch tips. Inconel is used for the springs of both latch assemblies and Type 304 stainless steel is used for all pressure containing parts. Hard chrome plating provides wear surfaces on the sliding parts and prevents galling between mating parts. A position indicator assembly slides over the CRDM rod travel housing. It detects the drive rod assembly position by means of 42 discrete coils that magnetically sense the entry and presence of the rod drive line through its center line over the normal length of the drive rod travel. 4.2-56 SGS-UFSAR Revision 25 October 26, 2010
Control Rod Drive Mechanism Control rod drive mechanisms are located on the dome of the reactor vessel. They are coupled to rod control clusters which have absorber material over the entire length of the control rods and derive their name from this feature. The CRDM is shown on Figure 4.2-21 and schematically on Figure 4.2-22. The primary function of the CRDM is to insert or withdraw rod control clusters within the core to control average core temperature and to shut down the reactor. The CRDM is a magnetically operated jack. A magnetic jack is an arrangement of three electro-magnets which are energized in a controlled sequence by a power cycler to insert or withdraw rod control clusters in the reactor core in discrete steps. The CRDM consists of four separate subassemblies. They are the pressure vessel, coil stack assembly, the latch assembly, and the drive rod assembly.
1. The pressure vessel includes a latch housing and a rod travel housing which are connected by a threaded, seal welded, maintenance joint which facilitates replacement of the latch assembly. The closure at the top of the rod travel housing is a threaded plug with a canopy seal weld for pressure integrity.
4.2-57 SGS-UFSAR Revision 22 May 5, 2006
The latch housing is the lower portion of the vessel and contains the latch assembly. The rod travel housing is the upper portion of the vessel and provides space for the drive rod during its upward movement as the control rods are withdrawn from the core.
2. The coil stack assembly includes the coil housings, an electrical conduit and connector, and three operating coils: 1) the stationary gripper coil, 2) the moveable gripper coil, and 3) the lift coil.
The coil stack assembly is a separate unit which is installed on the drive mechanism by sliding it over the outside of the latch housing. It rests on the base of the latch housing without mechanical attachment. Energizing of the operation coils causes movement of the pole pieces and latches in the latch assembly.
3. The latch assembly includes the guide tube, stationary pole pieces, moveable pole pieces, and two sets of latches: 1) the moveable gripper latch, and 2) the stationary gripper latch.
The latches engage grooves in the drive rod assembly. The moveable gripper latches are moved up or down in 5/8 inch steps by the lift pole to raise or lower the drive rod assembly while the moveable gripper latches are repositioned for the next 5/8 inch step. 4.2-58 SGS-UFSAR Revision 22 May 5, 2006
4. The drive rod assembly includes a flexible coupling, a drive rod, a disconnect button, a disconnect rod, and a locking button.
The drive rod has 5/8 inch grooves which receive the latches during holding or moving of the drive rod. The flexible coupling is attached to the drive rod and produces the means for coupling to the rod control cluster assembly. The disconnect button, disconnect rod, and locking button provide positive locking of the coupling to the rod control cluster assembly and permits remote disconnection of the drive rod. 4.2-58a SGS-UFSAR Revision 13 June 12, 1994
THIS PAGE INTENTIONALLY LEFT BLANK 4.2-58b SGS-UFSAR Revision 13 June 12, 1994
The CRDM is a trip design. Tripping can occur during any part of the power cycler sequencing if power to the coils is interrupted. The CRDM is threaded and seal welded on an adapter on top of the reactor vessel and is coupled to the rod control cluster assembly directly below. The mechanism is capable of handling a 360 pound load, including the drive rod weight, at a rate of 45 inches per minute. Withdrawal of the rod control cluster is accomplished by magnetic forces while insertion is by gravity. The mechanism internals are designed to operate in 650°F reactor coolant. The pressure vessel is designed to contain reactor coolant at 650°F and 2500 psia. The three operating coils are designed to operate at 392°F with forced air cooling required to maintain that temperature. The CRDM shown schematically on Figure 4.2-22 withdraws and inserts its control rod as electrical pulses are received by the operator coils. An ON or OFF sequence, repeated by silicon controlled rectifiers in the power programmer, causes either withdrawal or insertion of the control rod. Position of the control rod is measured by 42 discrete coils mounted on the position indicator assembly surrounding the rod travel housing. Each coil magnetically senses the entry and presence of the top of the ferro-magnetic drive rod assembly as it moves through the coil center line. During plant operation the stationary gripper coil of the drive mechanism holds the control rod withdrawn from the core in a static position until the movable gripper coil is energized. Rod Cluster Control Assembly Withdrawal The control rod is withdrawn by repetition of the following sequence of events: 4.2-59 SGS-UFSAR Revision 6 February 15, 1987
1. Movable Gripper Coil (B) - ON The latch locking plunger raises and swings the movable gripper latches into the drive rod assembly groove. A 1/16 inch axial clearance exists between the latch teeth and the drive rod.

2. Stationary Gripper Coil (A) - OFF The force of gravity, acting upon the drive rod assembly and attached control rod, causes the stationary gripper latches and plunger to move downward 1/16 inch until the load of the drive rod assembly and attached control rod is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.

3. Lift Coil (C)- ON The 5/8 inch gap between the movable gripper pole and the lift pole closes and the drive rod assembly raises one step length (5/8 inch).

4. Stationary Gripper Coil (A) - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 1/16 inch. The 1/16 inch vertical drive rod assembly movement transfers the drive rod assembly load from the moveable gripper latches to the stationary gripper latches.

5. Movable Gripper Coil (B) - OFF 4.2-60 SGS-UFSAR Revision 6 February 15, 1987
The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove.
6. Lift Coil (C) - OFF The gap between the movable gripper pole and lift pole opens. The movable gripper latches drop 5/8 inch to a position adjacent to a drive rod assembly groove.

7. Repeat Step 1 The sequence described above (1 through 6) is termed as one step or one cycle. The control rod moves 5/8 inch for each step or cycle.
The sequence is repeated at a rate of up to 72 steps per minute and the drive rod assembly (which has a 5/8 inch groove pitch) is raised 72 grooves per minute. The control rod is thus withdrawn at a rate up to 45 inches per minute. Rod Cluster Control Assembly Insertion The sequence for control rod insertion is similar to that for control rod withdrawal, except the timing of lift coil (C) ON and OFF is changed to permit lowering the control rod.
1. Lift Coil (C) - ON The 5/8 inch gap between the movable gripper and lift pole closes.
The movable gripper latches are raised to a position adjacent to a drive rod assembly groove. 4.2-61 SGS-UFSAR Revision 6 February 15, 1987
2. Movable Gripper Coil (B) - ON The latch locking plunger raises and swings the movable gripper latches into a drive rod assembly groove. A 1/16 inch axial clearance exists between the latch teeth and the drive rod assembly.

3. Stationary Gripper Coil (A) - OFF The force of gravity, acting upon the drive rod assembly and attached control rod, causes the stationary gripper latches and plunger to move downward 1/16 inch until the load of the drive rod assembly and attached control rod is transferred to the movable gripper latches. The plunger continues to move downward and swings the stationary gripper latches out of the drive rod assembly groove.

4. Lift Coil (C) -OFF The force of gravity separates the movable gripper pole from the lift pole and the drive rod assembly and attached control rod drop down 5/8 inch.

5. Stationary Gripper (A) - ON The plunger raises and closes the gap below the stationary gripper pole. The three links, pinned to the plunger, swing the three stationary gripper latches into a drive rod assembly groove. The latches contact the drive rod assembly and lift it (and the attached control rod) 1/16 inch. The 1/16 inch vertical drive rod assembly movement transfers the drive rod assembly load from the movable gripper latches to the stationary gripper latches.

6. Movable Gripper Coil (B) - OFF 4.2-62 SGS-UFSAR Revision 6 February 15, 1987
The latch locking plunger separates from the movable gripper pole under the force of a spring and gravity. Three links, pinned to the plunger, swing the three movable gripper latches out of the drive rod assembly groove.
7. Repeat Step 1 The sequences are repeated, as for control rod withdrawal, up to 72 times per minute which give a control rod insertion rate of 45 inches per minute.
Holding and Tripping of the Control Rods During most of the plant operating time, the CRDMs hold the control rods withdrawn from the core in a static position. In the holding mode, only one coil, the stationary gripper coil (A), is energized on each mechanism. The drive rod assembly and attached control rod hang suspended from the three latches. If power to the stationary gripper coil is cut off, the combined weight of the drive rod assembly and the rod cluster control assembly is sufficient to move latches out of the drive rod assembly groove. The control rod falls by gravity into the core. The trip occurs as the magnetic field, holding the stationary gripper plunger half against the stationary gripper pole, collapses and the stationary gripper plunger half is forced down by the weight acting upon the latches. After the drive rod assembly is released by the mechanism, it falls freely until the control rods enter the buffer section of their thimble tubes. 4.2.3.3 Design Evaluation 4.2.3.3.1 Reactivity Control Components The components are analyzed for loads corresponding to normal, upset, emergency, and faulted conditions. The analysis performed 4.2-63 SGS-UFSAR Revision 6 February 15, 1987
depends on the mode of operation under consideration. The scope of the analysis requires many different techniques and methods, both static and dynamic. Some of the loads that are considered on each component where applicable are as follows:
1. Control Rod Trip (equivalent static load)

2. Differential Pressure

3. Spring Preloads

4. Coolant Flow Forces (static)

5. Temperature Gradients

6. Differences in thermal expansion

a. Due to temperature differences

b. Due to expansion of different materials

7. Interference Between Components

8. Vibration (mechanically or hydraulically induced)

9. All Operational Transients Listed in Table 4.1-10

10. Pump Overspeed

11. Seismic Loads (OBE and DBE)
The main objective of the analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the 4.2-64 SGS-UFSAR Revision 6 February 15, 1987
functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Standard methods of strength of materials are used to establish the stresses and deflections of these components. The dynamic behavior of the reactivity control components has been studied using experimental test data (11) and experience from operating reactors. The design of reactivity component rods provides a sufficient cold void volume within the burnable absorber and source rods to limit the internal pressures to a value which satisfies the criteria in Section 4.2.3.1. The void volume for the helium in the burnable absorber rods is obtained through the use of glass in tubular form which provides a central void along the length of the rods. Helium gas is not released by the neutron absorber rod material; thus the absorber rod only sustains an external pressure during operating conditions. The internal pressure of source rods continues to increase from ambient until end-of-life at which time the internal pressure never exceeds that allowed by the criteria in Section 4.2.3.1. The stress analysis of reactivity component rods assumes 100 percent gas release to the rod void volume, considers the initial pressure within the rod, and assumes the pressure external to the component rod is zero. Based on available data for properties of the borosilicate glass and on nuclear and thermal calculations for the burnable absorber rods, gross swelling or cracking of the glass tubing is not expected during operation. Some minor creep of the glass at the hot spot on the inner surface of the tube could occur but would continuously until the glass came in contact with the inner liner. The wall thickness of the inner liner is sized to provide adequate support in the event of slumping and to collapse locally before rupture of the exterior cladding if unexpected large volume changes due to swelling or cracking should occur. The top of the inner liner is open to 4.2-65 SGS-UFSAR Revision 11 July 22, 1991
allow communication to the central void by the helium which diffuses out of the glass. Sufficient diametral and end clearances have been provided in the neutron absorber, burnable absorber, and source rods to accommodate the relative thermal expansions between the enclosed material and the surrounding clad and end plugs. There is no bending or warping induced in the rods although the clearance offered by the guide thimble would permit a postulated warpage to occur without restraint on the rods. Bending, therefore, is not considered in the analysis of the rods. The radial and axial temperature profiles have been determined by considering gap conductance, thermal expansion, and neutron and/or gamma heating of the contained material as well as gamma heating of the clad. The maximum neutron absorber material temperature was found to be less than 850°F which occurs axially at only the highest flux region. The maximum borosilicate glass temperature was calculated to be about 1200°F and takes place following the initial rise to power. The glass temperature then decreases rapidly for the following reasons: (1) reduction in power generation due to depletion; (2) better gap conductance as the helium produced diffuses to the gap; and (3) external gap reduction due to borosilicate glass creep. Rod, guide thimble, and dashpot flow analysis performed indicates that the flow is sufficient to prevent coolant boiling and maintain clad temperatures at which the clad material has adequate strength to resist coolant operating pressures and rod internal pressures. Analysis on the rod cluster control spider indicates the spider is structurally adequate to withstand the various operating loads including the higher loads which occur during the drive mechanism stepping action and rod drop. The reactivity control component materials selected are considered to be the best available from the standpoint of resistance to irradiation damage and compatibility to the reactor environment. The materials selected partially dictate the reactor environment 4.2-66 SGS-UFSAR Revision 25 October 26, 2010
(e.g., Cl control in the coolant). The current design type reactivity controls have been in service for more than 10 years with no apparent degradation of construction materials. With regard to the materials of construction exhibiting satisfactory resistance to adverse property changes in a radioactive environment, it should be noted that on work on breeder reactors in current design, similar materials are being applied. At high fluences the austenitic materials increase in strength with a corresponding decreased ductility (as measured by tensile tests) but energy absorption (as measured by impact tests) remains quite high. Corrosion of the materials exposed to the coolant is quite low and proper control of Cl and 0 2 in the coolant will prevent the occurrence of stress corrosion. All of the austenitic stainless steel base materials used are processed and fabricated to preclude sensitization. Analysis of the rod cluster control assemblies shows that if the drive mechanism housing ruptures, the rod cluster control assembly will be ejected from the core by the pressure differential of the operating pressure and ambient pressure across the drive rod assembly. The ejection is also predicted on the failure of the drive mechanism to retain the drive rod/ rod cluster control assembly position. It should be pointed out that a drive mechanism housing rupture will cause the ejection of only one rod cluster control assembly with the other assemblies remaining in the core. 4.2-67 SGS-UFSAR Revision 17 October 16, 1998
Ejection of a burnable absorber or thimble plug assembly (if used) is conceivable based on the postulation that the holddown bar fails and that the base plate and burnable absorber rods are severely deformed. In the unlikely event that failure of the holddown bar occurs, the upward displacement of the burnable absorber assembly only permits the base plate to contact the upper core plate. Since this displacement is small, the major portion of the burnable absorber material remains positioned within the core. In the case of the thimble plug assembly, the thimble plugs will partially remain in the fuel assembly guide thimbles thus maintaining a majority of the desired flow impedance. Further displacement or complete ejection would necessitate the square base plate and burnable absorber rods be forced, thus plastically deformed, to fit up through a smaller diameter hole. It is expected that this condition requires a substantially higher force or pressure drop than that of the holddown bar failure. Experience with control rods, burnable absorber rods, and source rods is discussed in Reference 2. The mechanical design of the reactivity control components provides for the protection of the active elements to prevent the loss of control capability and functional failure of critical components. The components have been reviewed for potential failure and consequences of a functional failure of critical parts. The results of the review are summarized below. Rod Cluster Control Assembly
1. The basic absorbing material is sealed from contact with the primary coolant and the fuel assembly and guidance surfaces by a high quality stainless steel clad.
4.2-68 SGS-UFSAR Revision 18 April 26, 2000
Potential loss of absorber mass or reduction in reactivity control material due to mechanical or chemical erosion or wear is therefore reliably prevented.
2. A breach of the cladding for any postulated reason does not result in serious consequences. The absorber material, silver-indium-cadmium, is relatively inert and would still remain remote from high coolant velocity regions. Rapid loss of material resulting in significant loss of reactivity control material would not occur.

3. The individually clad absorber rods are doubly secured to the retaining spider finger by a threaded top end plug secured by a nut welded to the finger and a welded lock pin.
It should also be noted that in several instances of control rod jamming caused by foreign particles, the individual rods at the site of the jam have borne the full capacity of the CRDM and higher impact loads to dislodge the jam without failure. The guide tube card/guide thimble arrangement is such that large loads are required to buckle individual control rods. The conclusion to be drawn from this experience is that this joint is extremely insensitive to potential mechanical damage. A failure of the joint would result in the insertion of the individual rod into the core. This results in reduced reactivity which is a fail safe condition. Further information is given in Reference 2. 4.2-69 SGS-UFSAR Revision 17 October 16, 1998
4. The spider is a one-piece machined casting and includes the radial vanes and fingers. Reliability is increased by not using brazed joints. Casting allows the rod holes to be drilled to the positional tolerances prior to assembly to ensure the rods will align with the guide cards.
4.2-70 SGS-UFSAR Revision 17 October 16, 1998
5. The spider hub being of a one-piece machined casting is very rugged and of extremely low potential for damage. It is difficult to postulate any condition to cause failure. Should some unforeseen event cause fracture of the hub above the vanes, the lower portion with the vanes and rods attached would insert by gravity into the core causing reactivity decrease. The rod could then not be removed by the drive line, again a fail safe condition. Fracture below the vanes cannot be postulated since all loads, including scram impact, are taken above the vane elevation.

6. The rod cluster control rods are provided a clear channel for insertion by the guide thimbles of the fuel assemblies. All fuel rod failures are protected against by providing this physical barrier between the fuel rod and the intended insertion channel.
Distortion of the fuel rods by bending cannot apply sufficient force to damage or significantly distort the guide thimble. Fuel rod distortion by swelling, though precluded by design, would be terminated by fracture before contact with the guide thimble occurs. If such were not the case, it would be expected that a force reaction at the point of contact would cause a slight deflection of the guide thimble. The radius of curvature of the deflected shape of the guide thimbles would be sufficiently large to have a negligible influence on rod cluster control insertion. Burnable Absorber Assemblies The burnable absorber assemblies are static temporary reactivity control elements. The axial position is assured by the holddown assembly which bears against the upper core plate. Their lateral position is maintained by the guide thimbles of the fuel assemblies.
4. 2-71 SGS-UFSAR Revision 17 October 16, 1998
The individual rods are shouldered against the underside of the retainer plate and securely fastened at the top by a threaded nut which is then locked in place. The square dimension of the retainer plate is larger than the diameter of the flow holes through the core plate. Failure of the holddown bar or spring pack therefore does not result in ejection of the burnable absorber rods from the core. The only incident that could potentially result in ejection of the burnable absorber rods is a multiple fracture of the retainer plate. This is not considered credible because of the light loads borne by this component. During normal operation the loads borne by the plate are approximately 5 lbs per rod, or a total of 100 lbs. distributed at the points of attachment. Even a multiple fracture of the retainer plate would result in jamming of the plate segments against the upper core plate, again preventing ejection. Excessive reactivity increase due to burnable poison ejection is therefore prevented. The same type of stainless steel clad used on rod cluster control rods is also used on the burnable absorber rods. In this application there is even less susceptibility to mechanical damage since these are static assemblies. The guide thimbles of the fuel assembly afford the same protection from damage due to fuel rod failures as that described for the rod cluster control rods. The consequences of clad breach are also similarly small. The absorber material is borosilicate glass which is maintained in position by a central hollow tube. In the event of a hole developing in the clad for any postulated reason the expected consequence is only the loss of the helium produced by the absorption process into the primary coolant. The glass is chemically inert and remains remote from high coolant velocities; therefore significant loss of absorber material resulting in reactivity increase is not expected. 4.2-72 SGS-UFSAR Revision 11 July 22, 1991
Drive Rod Assemblies All postulated failures of the drive rod assemblies either by fracture or uncoupling, lead to the fail safe condition. If the drive rod assembly fractures at any elevation, that portion remaining coupled falls with and is guided by the rod cluster control assembly. This always results in reactivity decrease for the control rods. 4.2.3.3.2 Control Rod Drive Mechanism Material Selection All pressure-containing materials comply with Section III of the ASME Pressure Vessel Code, and, with the exception of the needle vent valve, will be fabricated from austenitic (Type 304) stainless steel or CF-8 stainless steel. The vent valve is a modified austenitic stainless steel cap screw. Magnetic pole pieces are fabricated from Type 410 stainless steel. All nonmagnetic parts, except pins and springs, are fabricated from Type 304 stainless steel. Haynes 25 is used to fabricate link pins. Springs are made from Inconel-X. Latch arm tips are clad with Stelli te to provide improved wearability. Hard chrome plate and Stellite are used selectively for bearing and wear surfaces. At the start of the development program, a survey was made to determine whether a material better than Type 410 stainless steel was available for the magnetic pole pieces. Ideal material requirements are as follows:
1. High magnetic saturation value 4.2-73 SGS-UFSAR Revision 22 May 5, 2006

2. High permeability

3. Low coercive force

4. High resistivity

5. High Curie temperature

6. Corrosion resistant

7. High impact strength

8. Nonoriented

9. High machinability

10. Radiation damage After a comprehensive material trade-off study was made it was decided that the Type 410 stainless steel was satisfactory for this application.
The cast coil housings require a magnetic material. Both low-carbon cast steel and ductile iron have been successfully tested for this application. The choice, made on the basis of cost, indicated that ductile iron will be specified on the CRDM. The finished housings are zinc plated to provide corrosion resistance. Coils are wound on bobbins of molded Dow Corning 302 material, with double glass-insulated copper wire. Coils are then vacuum impregnated with silicon varnish. A wrapping of mica sheet is secured to the coil outer surface. The result is a well-insulated coil capable of sustained operation at 200°C. The drive shaft assembly utilized a Type 410 stainless steel drive rod. The coupling is machined from Type 403 stainless steel. Other 4.2-74 SGS-UFSAR Revision 6 February 15, 1987
parts are Type 304 stainless steel with the exception of the springs which are Inconel-X and the locking button which is Haynes 25. Radiation Damage As required by the equipment specification, the CRDMs are designed to meet a radiation requirement of 10 Rads/Hr. Materials have been selected to meet this 6 requirement. The above radiation level which amount to 1. 7 53 x 10 Rads in 20 years will not degrade control rod drive mechanism life. Control rod drive mechanisms at Yankee Rowe which have been in operation since 1960 have not experienced problems due to radiation. Positioning Requirements The mechanism has a step length of 5/8 inches which determines the positioning capabilities of the control rod drive mechanism. (Note: Positioning requirements are determined by reactor physics.) Elevation of Materials Adequacy The ability of the pressure housing components to perform throughout the design lifetime as defined in the equipment specification is confirmed by the stress analysis report required by the ASME Boiler and Pressure Vessel Code, Section III. Internals components subjected to wear will withstand a minimum of 3,000,000 steps without refurbishment as confirmed by life tests. Results of Dimensional and Tolerance Analysis With respect to the CRDM systems as a whole, critical clearances are present in the following areas:
1. Latch assembly (Diametral clearances) 4.2-75 SGS-UFSAR Revision 20 May 6, 2003

2. Latch arm-drive rod clearances

3. Coil stack assembly-thermal clearances

4. Coil fit in coil housing The following write-up defines clearances that are designed to provide reliable operation in the CRDM in these four critical areas. These clearances have been proven by life tests and actual field performance at operating plants.
Latch Assembly - Thermal Clearances The magnetic jack has several clearances where parts made of Type 410 stainless steel fit over parts made from Type 304 stainless steel. Differential thermal expansion is therefore important. Minimum clearances of these parts at 650°F minimum clearance is 0.0045 inch, and at the maximum expected operating temperatures of 550°F is 0.0057 inch. Latch Arm - Drive Rod Clearances The CRDM incorporates a load transfer action. The movable or stationary gripper latch is not under load during engagement, as previously explained, due to load transfer action. Figure 4.2-23 shows latch clearance variation with the drive rod as a result of minimum and maximum temperatures. Figure 4.2-24 shows clearance variations over the design temperature range. Coil Stack Assembly - Thermal Clearances The assembly clearance of the coil stack assembly over the latch housing was selected so that the assembly could be removed under all anticipated conditions of thermal expansion. 4.2-76 SGS-UFSAR Revision 6 February 15, 1987
At 70°F the inside diameter of the coil stack is 7.308/7.298 inches. The outside diameter of the latch housing is 7.260/7.270 inches. Thermal expansion of the mechanism due to operating temperature of the CRDM results in minimum inside diameter of the coil stack being 7. 310 inches at 222°F and the maximum latch housing diameter being 7.302 inches at 532°F. Under the extreme tolerance conditions listed above, it is necessary to allow time for a 70°F coil housing to heat during a replacement operation. Four coil stack assemblies were removed from four hot CRDMs mounted on 11.035-inch centers on a 550°F test loop, allowed to cool, and then replaced without incident as a test to prove the proceeding. Coil Fit in Coil Housing Control rod drive mechanism and coil housing clearances are selected so that coil heatup results in a close or tight fit. This is done to facilitate thermal transfer and coil cooling in a hot CRDM. 4.2.3.4 Tests, Verification, and Inspections 4.2.3.4.1 Reactivity Control Components Tests and inspections are performed on each reactivity control component to verify the mechanical characteristics. In the case of the rod cluster control assembly, prototype testing has been conducted, and both manufacturing test/inspections and functional testing at the plant site are performed. 4.2-77 SGS-UFSAR Revision 6 February 15, 1987
During the component manufacturing phase, the following requirements apply to the reactivity control components to assure the proper functioning during reactor operation:
1. All materials are procured to specifications to attain the desired standard of quality.

2. All clad/end plug welds are checked for integrity by visual inspection and X-ray, and are helium leak checked. All the seal welds in the neutron absorber rods, burnable poison rods, and source rods are checked in this manner.

3. To assure proper fitup with the fuel assembly, the rod cluster control, burnable poison, and source assemblies are installed in the fuel assembly and checked for binding in the dry condition.
The rod cluster control assemblies (RCCA) are functionally tested following core loading, but prior to criticality to demonstrate reliable operation of the assemblies. Each assembly is operated (and tripped) one time at no flow/cold conditions and one time at full flow/hot conditions. In addition, selected assemblies, amounting to about 15 to 20 percent of the total assemblies are operated at no-flow/operating temperature conditions and full flow/ambient conditions. Also the slowest rod and the fastest rod are tripped 10 times at no-flow/ambient conditions and at full flow/operating temperature conditions. Thus each assembly is tested a 4.2-78 SGS-UFSAR Revision 17 October 16, 1998
minimum of 2 times or up to 14 times maximum to ensure that the assemblies are properly functioning. In order to demonstrate continuous free movement of the RCCAs and to ensure acceptable core power distributions during operations, partial movement checks are performed on every RCCA, as required by the Technical Specifications. In addition, periodic drop tests of the RCCAs are performed at each refueling shutdown to demonstrate continued ability to meet trip time requirements. If an RCCA cannot be moved by its mechanism, adjustments in the boron concentration of the coolant ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move one RCCA can be tolerated. More than one inoperable RCCA could be tolerated but would impose additional demands on the plant operator. Therefore, the number of inoperable RCCAs has been limited to one. 4.2.3.4.2 Control Rod Drive Mechanism Quality assurance procedures during production of CRDMs include material selection, process control, mechanism component tests during production, and hydrotests. After all manufacturing procedures had been developed, several prototype CRDMs and drive rod assemblies were life tested with the entire drive line under environmental conditions of temperature, pressure, and flow. All acceptance tests were of duration equal to or greater than service required for the plant operation. All drive rod assemblies tested in this manner have shown minimal wear damage. These tests include verification that the trip time achieved by the CRDMs met the design requirement of 2. 7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry. Trip time requirement will be confirmed for each CRDM at periodic intervals after initial reactor operation. In
addition, 4.2-79 SGS-UFSAR Revision 16 January 31, 1998
a Technical Specification has been set to ensure that the trip time requirement is met. It is expected that all CRDMs will meet specified operating requirements for the duration of plant life with normal refurbishment. However, a Technical Specification pertaining to an inoperable RCCA has been set. If an RCCA cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move one RCCA can be tolerated. More than one inoperable RCCA could be tolerated, but would impose additional demands on the plant operator. Therefore, the number of inoperable RCCAs has been limited to one. In order to demonstrate continuous free movement of the RCCAs and to ensure acceptable core power distributions during operation, partial-movement checks are performed on every RCCA at least every 31 days during reactor critical operation. In addition, periodic drop tests of the full length RCCAs are performed at each refueling shutdown to demonstrate continued ability to meet trip time requirements, to ensure core subcri ticali ty after reactor trip, and to limit potential reactivity insertions from a hypothetical RCCA ejection. During these tests the acceptable drop time of each assembly is not greater than 2.7 seconds, at full flow and operating temperature, from the beginning of decay of stationary gripper coil voltage to dashpot entry. To confirm the mechanical adequacy of the fuel assembly and RCCA, functional test programs have been conducted on a full scale control rod. The prototype assembly was tested under simulated conditions of reactor temperature, pressure, and flow for approximately 1000 hours. The prototype mechanism accumulated about 3,000,000 steps and 600 trips. At the end of the test the CRDM was still operating satisfactorily. A correlation was developed to predict the amplitude of flow excited vibration of individual fuel rods and fuel assemblies. Inspection of the drive 4.2-80 SGS-UFSAR Revision 16 January 31, 1998
line components did not reveal significant fretting. The control rod free fall time against 125 percent of nominal flow was less than 1. 5 seconds to the dashpot; about 10 feet of travel. Actual experience on the Ginna, Mihama No. 1, Point Beach No. 1, and H. B. Robinson plants indicates excellent performance of CRDMs. All units are production tested prior to shipment to confirm ability of CRDMs to meet design specification-operational requirements. Periodic tests are also conducted during plant operation to confirm brake core operation. During refueling, tests are also conducted to confirm condition to stator windings. 4.2.4 References for Section 4.2
1. Christensen, J. A.; Allio, R. J.; and Biancheria, A. "Melting Point of Irradiated uo ," WCAP-6065, February 1965.
2
2. Skari tka, J., "Operational Experience with Westinghouse Cores," WCAP-8183 (latest revision - updated annually) .

3. Eggleston, F. T., "Safety Related Research and Development for Westinghouse Pressurized Water Reactor - Program Summaries, Winter 1976 -
Summer 1978," WCAP-8768, Revision 2, October 1977.
4. Risher, D. H. (Ed.), "Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963-P-A, (Proprietary) and WCAP-8964-A, (Nonproprietary) August 1977.

5. Not used.
4.2-81 SGS-UFSAR Revision 22 May 5, 2006
6. Not used.

7. George, R. A., Lee, Y. C.; and Eng, G. H., "Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Nonproprietary), July 1974.

8. Watkins, B. and Wood, D. S. "The Significance of Irradiation - Induced Creep on Reactor Performance of a Zirca1oy-2 Pressure Tube",
Applications - Related Phenomena for Zirconium and its Alloys, ASTM STP 458, American Society for Testing and Materials, pp. 226-240, 1969.
9. O'Donnell, W. J. and Langer, B. F., "Fatigue Design Basis for Zircaloy Components," Nuclear Science and Engineering, 20, 1-12, 1964.

10. Gesinski, L.; Chiang, D.; and Nakazato, S., "Safety Analysis of the 17 x 17 Fuel Assembly For Combined Seismic and Loss-of-Coolant Accident,"
WCAP-8236 (Proprietary) and WCAP-8288, (Westinghouse Nonproprietary), December 1973 and Addendum 1.
11. Demario, E. E. and Nakazato, S., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Nonproprietary),
February 1974.
12. NUREG-0641, "Control Rod Guide Tube Wear in Operating Reactors," U.S.
NRC, Division of Operating Reactors, April 1980.
13. Moore, J., "Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 5A, November 1979.

14. Cohen, J., "Development and Properties of Silver Base Alloys as Control Rod Materials for Pressurized Water Reactors," WAPD-214, December 1959.

15. Davidson, S. L. (Ed.), et al, "Vantage 5 Fuel Assembly Reference Core Report," WCAP-10444-P-A, September 1985.
4.2-82 SGS-UFSAR Revision 25 October 26, 2010
16. Davidson, S. L. (Ed.), et al, "Vantage 5H Fuel Assembly," WCAP-10444-P-A, Addendum 2-A, February 1989.

17. Foster, J. P., Sidener, S., Westinghouse Improved Performance Analysis and Design Model (PAD 4.0), WCAP-15063-P-A, Revision 1, with Errata, July 2000.

18. Davidson, S. L. and Ryan, T. L., VANTAGE+ Fuel Assembly Reference Core Report, WCAP-12610-P-A, April 1995.

19. Letter from W. J. Rinkacs (Westinghouse) to M. M. Mannion (PSE&G),
Westinghouse Fuel Features Recommendation for Cycle 11, July 22, 1998
20. Letter from W. J. Rinkacs (Westinghouse) to T. K. Ross (PSE&G), Design Reviews for Fuel Feature Changes Proposed for Cycle 11, July 22, 1998

21. Letter from W. J. Rinkacs (Westinghouse) to M. M. Mannion (PSE&G),
Westinghouse Generic Safety Evaluation for the 17x17 Standard Robust Fuel Assembly, October 1, 1998
22. Letter from B. W. Gergos (Westinghouse) to T. K. Ross (PSE&G), Seismic/
LOCA Analysis of the Robust Fuel Assemblies, February 4, 1999
23. Iorii, J. A. and Petrarca, D. J., Westinghouse Wet Annular Burnable Absorber Evaluation Report, WCAP-10021-P-A, Revision 1, October 1983

24. Letter from I. R. Williamson (Westinghouse) to F. D. Rankowski (PSEG),
Extended Lifetime Wet Annular Burnable Absorber, NF-PSE-06-5, February 16, 2006.
25. Davidson, S.L. (Ed.), Westinghouse Fuel Criteria Evaluation Process, WCAP-12488-A, October 1994.

26. Sepp, H.A., Addendum 1 to WCAP-12488-A Revision to Design Criteria, WCAP-12488-A, Addendum 1-A, Revision 1, January 2002.

27. Garde, A., et al., Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO, WCAP-12610-P-A & CENPD-404-P-A, Addendum 2-A, October 2013.

28. Kersting, P. J, et al., Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel, WCAP-13589-A, March 1995.
4.2-83 SGS-UFSAR Revision 30 May 11, 2018
TABLE 4.2-1 MAXIMUM DEFLECTIONS ALLOWED FOR REACTOR INTERNAL SUPPORT STRUCTURES No-Loss-of-Allowable Function Deflections Deflections Component (in.) (in.) Upper Barrel radial inward 4.1 8.2 radial outward 0.5 1.0 Upper Package 0.10 0.15 Rod Cluster Guide Tubes 1.00 1.75
SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
TABLE 4.2-2 COMPARISON OF SINGLE AND DOUBLE ENCAPSULATED SECONDARY SOURCE DESIGNS PA.RAMETE:R SINGLE ENCAPSULA'J:ED DOUBLE ENCAPSt.T.LATED Number of rodlets 4 6 Outer Clad OD, in. 0.361 +I- 0.001 0.381 +I- 0.001 Outer Clad ID, in. 0.344 +!- 0.0005 0.344 +I- 0.0005 Inner Clad OD, in. N/A b.344 +I- 0.001 Inner Clad ID, in. N/A 0.297 +I- 0.0005 Pellet OD, in. 0.338 +0.002/-0.001 0.292 +/- 0.001 Pellet Stack Length, in. 88.00 86.00 Pellet Stack Weight, grams 500/535 338 +/- 10 Sp=in; Clip Material Carbon steel - plated 410 Stainless steel Outer Pressuriza~ion, psig 625 +/- 50 625 +/- 50 Inner ~ressurization, psig N/A 250 +/- 20 1 of 1 SGS-UFSAR Revision 18 April 26, 2000
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LR-N22-0086, Unit 2, Updated Final Safety Analysis Report, Revision 33

Contents

Text

SUMMARY

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7.1 INTRODUCTION

8.1 INTRODUCTION

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14.1 DESCRIPTION

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REFERENCE:

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LR-N22-0086, Unit 2, Updated Final Safety Analysis Report, Revision 33

Text

SUMMARY

SUMMARY

SUMMARY

SUMMARY

7.1 INTRODUCTION

8.1 INTRODUCTION

SUMMARY

14.1 DESCRIPTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

REFERENCE:

SUMMARY

SUMMARY

SUMMARY

-==

Response

RESPONSE

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