ML20052G217
| ML20052G217 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 03/19/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20052G213 | List: |
| References | |
| 22A7007, 22A7007-R02, 22A7007-R2, NUDOCS 8205140413 | |
| Download: ML20052G217 (20) | |
Text
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UNITED. STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
General Electric Company ) Docket No. STN 50-447 Standard Plant )
AMENDMENT NO. 2 TO APPLICATION FDA REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendinent No. 2 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No. 2 consists of two parts, a non proprietary portion and a portion considered by the General Electric Company to be proprietary.
The pages considered to be proprietary are so marked and are transmitted under separate cover.
Amendment No. 2 amends GESSAR I! oy adding Appendix 15D to Chapter 15.
Appendix 150 provides a Probabili Aic Risk Assessment (PRA) which quantifies the risk associated with operating a BWR/6 standard plant with a reference Mark III containment for the purpose of electrical power generation.
l The PRA demonstrates that the risk of operating the BWR/6 standard plant f is minimal relative to other societal risks and addresses issues that have been raised with. regard to severe accidents. This PRA is provided l for the Nuclear Regulatory Commission's review of the 238 Nuclear Island
(' GESSAR II Severe Accident Design.
Respectfully submitted, General Electric Company by: AM G. G. Sherwood, ManaVeY Nuclear Safety and Licensing Operation hb[knohho$$00N7 K PDR Subscribed and sworn to before me this 19 day of March 1982.
$ o $ 17Ls?s? h by: Y ra
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U f fy ) b b dJ f- KAREN S. VOGEL4 USER // 8 g k.-g. NOTARY PUBUC CAWORNIA SANTA CLARA COUNTY 9
g My Commission Expires Dec. 21,1984 bo3OQososooosooosoouwupo&3 V
UNITED STATES 0F AMERICA NUCLEAR REGULAT0RY C0MMISSION In the Matter of )
General Electric Company ) Docket No. STN 50-447 Standard Plant )
AMENDMENT NO. 2 TO APPLICATION FDA REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No. 2 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
Amendment No. 2 consists of two parts, a non proprietary portion and a portion considered by the General Electric Company to be proprietary.
The pages considered to be proprietary are so marked and are transmitted under separate cover.
Amendment No. 2 amends GESSAR II by adding Appendix 150 to Chapter 15.
Appendix 150 provides a Probabilistic Risk Assessment (PRA) which quantifies the risk associated with operating a BWR/6 standard plant with a reference Mark III containment for the purpose of electrical power generation.
The PRA demonstrates that the risk of operating the BWR/6 standard plant is minimal relative to other societal risks and addresses issues that have been raised with regard to severe accidents. This PRA is provided for the Nuclear Regulatory Commission's review of the 238 Nuclear Island GESSAR II Severe Accident Design.
Respectfully submitted, General Electric Company by: s/G. G. Sherwood G. G. Sherwood, Manager i
Nuclear Safety and Licensing Operation Subscribedandsworntobeforemethisjj$_dayofMarch1982.
by: s/ Karen S. Vogelhuber Karen S. Vogelhuber Notary Public - California Santa Clara County i
My Commission Expires December 21, 1984 175 Curtner Avenue San Jose, CA 95125 l
l l csc:rf/202E2 I
GEk$81LF103 -
R;v. 2
- ' 238 NUCLEAR ISLAND E
APPENDIX ISD SEVERE ACCIDENTS l
l l
l I
GESSAR II Rev. 2
, ' 238 NUCLEAR ISLAND o
APPENDIX 150 SEVERE ACCIDENTS n
CONTENTS 4
150.1 EXECUTIVE SumARY 150.2 BWR/6 PREVENTION AND NITIGATION CAPA8ILITY l
4 15D-i
_ ~'
GESSAR II ~ 22A700F
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238 NUCLEAR ISLAND R2v. 2
- CONTENTS (continued) .
15D.3 BWR/0 STANDARD PLANT PROBABILISTIC DISK ASSESSMENT 1 INTRODUCTION 1.1 PROBABILISTIC RISK ASSESSMENT (PRA) BACKGROUND 1.1.1 References 1.2 DEFINITIONS 1.2.1 Probabilistic Risk Assessment 1.2.2 Frequency of Core Damage 1.2.3 Risk 1.3 OBJECTIVE AND SCOPE 1.4 PRA BASIS 1.4.1 Plant and Site 1.4.2 Failure Probability and Fielu Experience 1.4.3 Initiating Accident Events 1.4.4 System Interactions and Common Cause Failure 1.4.5 Human Reliability 1.4.6 Reliability Model Definitions 1.4.7 Initial and End-Point Conditions 1.4.8 Source Ters and Core Melt Phenomenology 1.5 METHODOLOGY 1.5.1 Outline of Analysis 1.5.2 Fault Tree and Event Tree Analysis 1.5.3 Containment Analysis 1.5.4 External Consequence Analysis 1.5.5 Complementary Cumulative Frequency Function (CCFF)
I 1.5.6 Frequency and Probability 1 2 PLANT DEFINITION FOR BWR/6 PROBABILISTIC RISK ASSESSMENT
2.1 INTRODUCTION
AND PLANT DEFINITION 2.1.1 Reactor Core Isolation Cooling System Modifications 2.1.1.1 RCIC System Automatic Reset 2.1.1.2 RCIC System Break Detection Logic r
GESSAR II 22A7007
- 238 NUCLEAR ISLAND R2v. 2 CONTENTS (continued) 2.1.2 Automatic Depressurization System (ADS) Logic Modification 2.1.3 Alternate 3A ATWS and Scram Discharge Volume Design Modifications 2.1.4 References
, 2.2 SITE DESCRIPTION 3 PROBABILITY OF CORE DAMAGE 3.1 ACCIDENT INITIATORS 3.1.1 References 3.2 EQUIPMENT RELIABILITY AND AVAILABILITY 3.2.1 System and Component Failure Data 3.2.2 Dependent Failure Treatment 3.2.2.1 General Considerations 3.2.2.2 Multiple Component Failures 3.2.2.3 Cascading Failures 3.2.2.4 Human Error 3.2.2.5 Interdependencies 3.2.2.6 External Causes 3.2.3 Human Error Prediction 3.3 ACCIDENT SEQUENCE ANALYSIS 3.3.1 Success Criteria 3.3.1.1 Core Cooling 3.3.1.2 Containment Heat Removal 3.3.2 ATWS Success Criteria 3.3.3 Accident Sequence Event Trees
! 3.3.4 Classification of Accident Classes 3.3.5 References 3.4 FREQUENCY OF CORE DAMAGE 15D-lii l
l i
GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 2 CONTENTS (continued) 4 PROBABILITY OF'RADI0 ACTIVE RELEASE 4.1 RELEASE PATH 4.1.1 Release Quantification Model 4.1.2 Potential Release Paths 4.2 CONTAINMENT PRESSURE CAPABILITY 4.3 FISSION PRODUCT RELEASE PATHS 4.3.1 Overpressurization of Containment 4.3.2 Steam Explosions 4.3.2.1 Introduction 4.3.2.2 Steam Explosion Definition 4.3,2.3 In-Vessel Steam Explosion 4.3.2.4 Ex-Vessel Steam Explosion 4.3.2.4.1 Explosion Scenario 4.3.2.4.2 Molten Corium/ Water Interaction 4.3.2.5 Conclusions 4.3.3 Impact of Hydrogen Combustion 4.3.4 Conclusions 4.3.5 References 4.4 CONTAINMENT EVENT TREES 4.5 CONSOLIDATED RELEASE SEQUENCES 4.6 FREQUENCY OF THE RELEASES 5 MAGNITUDE OF RADI0 ACTIVE RELEASE 5.1 PHENOMENA 0F CORE DAMAGE AND CONTAINMENT OVERPRESSURIZATION 5.1.1 Core Meltdown 5.1.2 RPV Melt Through 5.1. 3 Corium Concrete Interaction 5.1.4 Suppression Pool Boiloff l 5.1.5 References ISD-iv i
awwu GESSAR II Rev. 2 238 NUCLEAR ISLAND CONTENTS (continued)
=
5.2 DESCRIPTION
OF ACCIDENT SEqL'?NCES 5.2.1 Isolation With Loss of Coolant Makeup to Reactor Failure of Coolant Inventory Makeup Following a Small/
5.2.2 ,
Intermediate LOCA Failure of Coolant Inventory Makeup Following a large LOCA 5.2.3 5.2.4 Loss of Containment Heat Removal Loss of Containment Heat Removal With a Large LOCA 5.2.5 5.2.6 Anticipated Transient Without Scram (ATWS) With Loss of Containment Integrity Prior to Core Melt 5.2.7 References 5.3 FISSION PRODUCT SPECIES AND DECONTAMINATION FACTORS 5.3.1 Fission Product Transport Modeling 5.3.2 References 6 CONSEQUENCES OF RADI0 ACTIVE RELEASE 6.1 CALCULATIONS OF 0FFSITE CONSEQUENCES 6.1.1 CRAC Code Calculations of Consequences 6.1.2 Estimated Early and Latent Fatalities 6.1.3 References
6.2 CONCLUSION
S 7
SUMMARY
AND CONCLUSIONS l
I 7.1
SUMMARY
OF RESULTS 7.1.1 Frequency of Core Damage 7.1.2 The Risk Curves 7.1.3 Risk t
7.2 COMPARISON WITH WASH-1400 f References 7.2.1 7.3 COMPARISON TO OTHER RISKS 7.3.1 References 150-v
22A700y GESSAR II 238 NUCLEAR ISLAND R2v. 0 CONTENTS (continued) 7.4 OVERVIEW OF CONDITIONS AND LIMITATI0;is 7.4.1 Plant and Data 7.4.2 Scope 7.4.3 Methodology 7.4.4 Uncertainty
7.5 CONCLUSION
S A INPUT DATA FOR PROBABILISTIC EVALUATION A.1 ACCIDENT INITIATORS A.1.1 Transient Accident Initiators A.1.2 LOCA-Initiators A.1.3 Reactor Pressure Vessel Failure Rate A.I.4 References A.2 COMPONENT FAILURE DATA A.2.1 Introduction A.2.2 Failure Data Description A.2.3 Other Sources A.2.4 References A.3 UNAVAILABILITY OF ON-SITE AC POWER A.3.1 Failure Probability for a Single Diesel Generator A.3.2 Failure Probability for Two Diesel Generators A.3.3 Failure Probability for Three Diesel Generators A.3.4 References A.4 UNAVAILABILITY DUE TO ON-LINE TESTING AND MAINTENANCE A.4.1 References A.5 HUMAN ERROR PREDICTION A.5.1 Introduction l A.5.2 General Discussion A. S. 3 Instrument Calibration Human Error Prediction A.S.4 References 15D-vi
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GESSAR II 22A7007 238 NUCLEAR ISLAND RIv. 2 CONTENTS (continued)
A.6 UNAVAILABILITY OF OFF-SITE POWER A.6.1 Overview A.6.2 NUREG/CR-1464 A.6.3 Estimated Average Data A.6.4 Pennsylvania - New Jersey - Maryland (PJM) Data A.6.5 RSS (WASH-1400) Analysis A.6.6 Conclusions A.6.7 References B REDUNDANT REACTIVITY CONTROL SYSTEM DESCRIPTION B.1 REDUNDANT REACTIVITY CONTROL SYSTEM (RRCS)
B.2 ALTERNATE ROD INSERTION FUNCTION (ARI)
B.3 SCRAM DISCHARGE VOLUME CHANGES B.4 STANDBY LIQUID CONTROL SYSTEM (SLCS) DESIGN MODIFICATIONS B.5 REACTOR RECIRCULATION SYSTEM MODIFICATIONS B.6 FEEDWATER CONTROL SYSTEM B.7 NUCLEAR BOILER SYSTEM (NBS) MODIFICATIONS C EVENT TREES C.1
SUMMARY
C.1.1 Introduction C.1. 2 Results C.1.3 Symbols C.1.4 Event Tree Example C.2 REACTOR SHUTDOWN EVENT TREE C.2.1 Planned Shutdown C.2.2 Unplanned Shutdown C.3 TURBINE-TRIP EVENT TREES C.3.1 Turbine-Trip with Scram C.3.2 Turbine-Trip without Scram ISD-vii
238 NUCLEAR ISLAND R2v. 2 t
CONTENTS (continued)
C.4 ISOLATION EVENT TREES C.4.1 Isolation with Scram C.4.2 Isolation without Scram C.5 LOSS OF OFF-SITE POWER EVENT TREES C.5.1 LOOP with Scram C.S.2 LOOP without Scram C. 6 INADVERTENT OPENING OF S/R VALVE (IORV)
C.6.1 10RV with Scram C.6.2 10RV without Scram C.7 LARGE-BREAK LOCA IN THE DRYWELL C.8 INTERMEDIATE-BREAK LOCA IN THE DRYWELL C.9 SMALL-BREAK LOCA IN THE DRYWELL C.10 LARGE-BREAK LOCA IN THE CONTAINMENT C.11 INTERMEDIATE-BREAK LOCA IN THE CONTAINMENT C.12 SMALL-BREAK LOCA IN THE CONTAINMENT C.13 LARGE-BREAK LOCA OUTSIDE THE CONTAINMENT C.14 INTERMEDIATE-BREAK LOCA OUTSIDE THE CONTAINMENT C.15 SMALL-BREAK LOCA OUTSIDE THE CONTAINMENT l C.16 CONTAINMENT EVENT TREES C.16.1 Transient LOCA and ATWS (Class III)
C.16.2 Loss of Heat Removal C.16.3 ATWS (Class IV)
C.17 INPUTS TO CORRAL AND CRAC D FAULT TREES D.1 FUNCTIONAL FAULT TREES D.1.1 References D.2 SYSTEM FAULT TREES 150-viii
GESSAR II 22A7007
,' Rev. 2 238 NUCLEAR ISLAND CONTENTS (continued)
(
E EMERGENCY PROCEDURE GUIDELINES E.1 OVERVIEW E.1.1 Raferences '
E.2 USE OF GUIDELINES IN PROBABILISTIC RISK ASSESSMENT F DESCRIPTION OF COMPUTER MODELS AND METHODS F.1 THE WAM COMPUTER CODES F.1.1 Introduction F.1.2 Application F.1.3 References F.2 CORE DAMAGE AND CONTAINMENT RESPONSE F.2.1 Method Description F.2.2 BWR Meltdown Scenario Analysis by MARCH Model F.2.3 MARCH Calculational Results F.2.4 Acceptable Delay Times for Coolant Injection and Containment Heat Removal Initiation F.3 FISSION PRODUCT RELEASE AND TRANSPORT F.4 CONSEQUENCE G BWR/6 MARK III STANDARD PLANT CONTAINMENT STRUCTURAL SYSTEM PRESSURE-CARRYING ASSESSMENT G.1 INTRODUCTION G.1.1 References G.2 DETERMINATION OF STRESS DISTRIBUTION IN THE GE MARK III STANDARD PLANT CONTAINMENT VESSEL G.2.1 References 150-ix
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rov. 2 CONTENTS (continued)
G.3 DETERMINATION OF PRESSURE-CARRYING CAPABILITY OF GE MARK III STANDARD PLANT CONTAINMENT VESSEL G.3.1 Application of Level A Stress Intensity Limit Given in NE of ASME Section III Supplemented by the Compression Allowable Stress in Accordance with GESSAR II Appendix 3F G.4 ASSESSMENT OF THE LOWER PORTION OF THE GE STANDARD PLANT MARK III CONTAINMENT SMFI,L G.4.1 References G.5 ASSESSMENT OF THE CONTAINMENT PRESSURE VESSEL ANCHORAGE SYSTEM G.5.1 References G.6 PRESSURE CAPABILITY ASSESSMENT OF THE ORYWELL WALL AND THE ROOF SLAB G.6.1 References G.7 PRESSURE-CARRYING CAPABILITY OF DRYWELL HEAD G.7.1 References G.8 INVESTIGATION OF POSSIBLE FAILURE MODES OF THE CONTAINMENT VESSEL SHELL G.8.1 Plastic Yield G.8.2 Buckling G.8.3 Fracture G.8.4 References G.9 PROBABILITY OF LOSS OF CONTAINMENT INTEGRITY G.9.1 References G.10 ASSESSMENT OF CONTAINMENT RESPONSE TO PRESSURIZATION PHENOMENA G.10.1 Assessment of Response to Noncondensible Gas Generation G.10.2 Assessment of Local or Global Hydrogen Combustion Effects G.10.3 Assessment of Hydrogen Detonation Effects G.10.4 References G.11 CONCLUSIONS l
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238 NUCLEAR ISLAND R2v. 8 :
)
i CONTENTS (continued) i H PhE;4JEENA 0F STEAM EXPLOSIONS H.1 INTRODUCTION H.1.1 References H.2 STEAM EXPLOSIONS AS MODELED IN THE REACTOR SAFETY STUDY (WASH-1400)
H.2.1 In-Vessel Explosion H.2.2 Ex-Vessel Explosion H.2.3 References H.3 RELATIONSHIP TO PREVIOUS REACTOR EXPERIENCE H.3.1 References H.4 TYPICAL BWR CONFIGURATIONS H.S STEAM EXPLOSION PHENOMENA - IN-VESSEL H.5.1 High System Pressure H.5.2 Low System Pressures (Large Break Sequence)
H.6 INTERPRLTATION OF SANDIA THERMITE EXPERIHENTS H.6.1 References H.7 APPLICATION TO BWR DEGRADED CORE CONDITIONS H.8 STEAM EXPLOSIONS - EX-VESSEL H.9 EVALUATION OF SANDIA NATIONAL LABORATORY DATA H.9.1 Iron Thermite Tests - Water Rich Environment H.9.2 Iron Thermite Tests - Equal Volume Environment H.9.3 Spontaneous Trigger H.9.4 Corium-Air Tests H.10 NOHENCLATURE 150-xi
1 GESSAR II 22A7007 1 238 NUCLEAR ISLAND R2v. 2 CONTENTS (continued)
I. HYDROCEN PHENOMENA I.1 INTRODUCTION AND
SUMMARY
I.2 DEFINITIONS I.3 HYDROGEN COMBUSTION CHARACTERISTICS I.3.1 Flammability Limits of Hydrogen I.3.2 Application of Flammability Limits in PRA I.3.3 Ignition Sources I.3.4 Application of Ignition Sources in the PRA I.3.5 General Characteristics of Hydrogen Combustion I.3.5.1 Pressure Rise from Local or Global Combustion I.3.5.2 Pressure Rise from Local or Global Detonation I.3.5.3 Application in PRA I.3.6 References I.4 ACCIDENT SEQUENCES AND COMBUSTION PROBABILITIES I.5 EFFECT OF LOCAL COMBUSTION I.6 CONCLUSIONS 150-xii
GESSAR II 22ii7007 Rev. 2 238 NUCLEAR ISLAND CONTENTS (continued) 1
50.4 CONCLUSION
S ISD-xiii
GESSAR II N
.- 238 NUCLEAR ISLAND Rav. 2 APPENDIX 150 SEVERE ACCIDENTS ,
150.1 EXECUTIVE SUte4ARY (To be submitted 5/82) 15D.1-1
usvenaa su -
l 238 NUCLEAR ISLAND Rov. 2
. APPENDIX 15D SEVERE ACCIDENTS 15D.2 BWR/6 PREVENTION AND MITIGATION CAPABILITY (To be submitted 5/82) 15D.2-1
Mn R:v. 2 238 NUCLEAR ISLAND APPENDIX 15D SEVERE ACCIDENTS 150.3 BWR/6 STANDARD PLANT PRA GE PROPRIETARY - Provided under separate cover s
150.3-1 through 15D.3-808
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GESSAR 11 Rw.2 238 NUCLEAR ISLAND APPENDIX 150 SEVERE ACCIDENTS
.F 1
50.4 CONCLUSION
S f (To be submitted 5/82) ,
150.4-1
_ _ -