ML20210C954
| ML20210C954 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/08/1986 |
| From: | Doolittle E Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8609190020 | |
| Download: ML20210C954 (60) | |
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Docket l(o. 50-443 APPLICANT:
Public Service Company of New Hampshire FACILITY: Seabrook Station, Unit 1
SUBJECT:
SUMMARY
OF MEETING WITH PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE TO DISCUSS SEABROOK RISK MANAGEMENT AND EMERGENCY PLANNING STUDY A meeting was held with Public Service Company of New Hampshire on August 27, 1966 at NRC Headquarters in Bethesda, Maryland. The NRC staff was represented by itembers of the Office of Nuclear Reactor Regulation (NRR) and Brookhaven National Laboratory (Bill). The applicant was represented by members of Public
(
Service Company of New Hampshire, New Hampshire Yankee Division (PSNH),
Westinghouse Electric Corporation (W) and Pickard Lowe & Garrick (PL&G). A list of attendees is included as Enclosure 1.
l The purpose of this meeting was to discuss the Seabrook plant model and specifically treatment of Event V and containment bypass in the study.
l Mr. Jirn Moody began the applicant's presentation with an introduction and overview of results. He said that the two most important contributors to early release frequency are containnent isolation and interfacing systems LOCA.
Mr. Moody reviewed the containment response for principal contributors to risk in terms of accident sequence groups and initiating events from the SSPSA, categorizing them as early containment failure, delayed containment failure, l
and intact containment. He described four updated areas of the plant risk model: enhanced Y sequence, enhanced seismic analysis, containment recovery j
nodel and enhanced treatment of common cause failures. The applicant showed i
l that earthquake with cpen purge contributes 87% to early release frequency.
Mr. Karl Fleming (PL&G) continued the applicant's presentation with a discus-sion of the Seabrook plant model. He noted that the model is the same as the PRA model with updates.
It achieves a more complete coverage of accident sequences and full treatment of dependent events.
Ernie Rossi of the NRC staff asked how treatment of event V at Seabrook dif-fers from treatment at other plants. The applicant stated that valve l
configuration is different and there are differences in frequency of occurrence and failure rates of valve ruptures. These have been treated more realis-tically in the current study. Also, design differences include relief valves inside containment.
l Mr. Rich Barrett of the NRC staff asked if there had been changes in surveil-lance requirements as part of the new work.
It was stated that testing l
frequency did influence the analysis.
G609190020 060900 f
PDR ADOCK 05000443
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2-Two V sequence possitilities associated with the RCS boundary were identified.
These were cold leg injection path arrangement and RHR section path arrange-ment. The applicant stated that the most important aspect of the enhanced treatment of interfacing systems LOCA is the explicit consideration of RHR relief valves in containment.
Mr. Jim Moody continued the applicant's presentation with a discussion of treatment of containment isolation in the updated study. He stated that a total of nine penetrations were included in the model for quantification.
In the small penetration isolation failure model, the earthquake coninon cause initiating event dominates its contribution to risk. For larger containment purge valves the earthquake with loss of offsite power is the dominant con-tributor to risk.
Other aspects of the study discussed were SG Tube Rupture and the consequence analysts. The applicant showed that SG Tube Rupture is not important because of the considerable amount of time available for operator action.
The meeting was adjourned at 3:00 p.m'.
A list of slides used by the applicant during his presentation are included as Enclosure 2.
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A Elizabet Doolittle, Project Manger PWR Pr ect Directorate #5 Division of PWR Licensing-A Enclosures i
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.I Meeting Summary Distribution
< Docket or. Ceritral: File NRC Participants NRC PDR local POR E. Doolittle PD#5 Reading File V. Noonan J. Partlow T. Novak V. Noonan E. Rossi Project Manager G. Bagchi OGC-Bethesda David H. Jaffee E. Jordan Kate Silburgh R. Grimes S. l.ong ACRS (10)
V. Benaroya M. Rushbrook Don Hickman J. Shapaker cc: 1.icensee and Plant Service 1.i s t
m-9 Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrooi Nuclear Power Station cc:
Thomas Dignan, Esq.
E. Tupper Kinder, Esq.
John A. Ritscher, Esq.
G. Dana Bisbee, Esq.
Ropes and Gray Assisti.nt Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 708 Stite Hosue Annex Concord, New Hampshir; 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Restrient Inspector Post Office. Box 330 Seabrook Nuclear Powtr Station Manchester, New Hampshire 03105 c/o US Nuclear, Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire.03874 Sun Valley Association 209 Sumer Street Fr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and I.icensing Yankee Atomic Electric Company Robert A. Backus, Esq.
1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 1.owell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 Harmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C.
20009 Mr. Philip Ahrens, Esq.
Assistant Attorney General State House, Station #6 Augusta, Maine 04333 Jo Ann Shotwell, Esq.
Office of the Assistant Attorney General Environmental Protection Division Mr. Warrer. Hall One Ashburton Place Public Service Company of Boston, Massachusetts 02108 New Hampshire Post Office Box 330 D. Pierre G. Cameron, Jr., Esq.
Seabrook, New Hampshire 03874 General Counsel Public Service Company of New Hampshire Seacoast Anti-Pollution League Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region !
U.S. Nuclear Regulatory Comission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Hampe, Esq.
New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301 i
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J Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:
Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. I.etty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C.
20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.
Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman Committee Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Commission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C.
20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874
J ATTENDANCE 8/27/86 Seabrook EP Sensitivity Study Meeting 1.
E. Doolittle NRC Project Manager 2.
Vince Noonan NRC/NRR/PWR-A 3.
T. M. Novak NRC/NRC/PWR-A 4.
Ernie Rossi NRC/NRR/PWR-A 5.
Goutam Bagchi NRC/NRR/PWR-A 6.
Moon-Hyun Chun Brookhaven National Laboratory 7.
Arthur G. Tingle Brookhaven National Laboratory 8.
Bruce Miler Brookhaven National Laboratory 9.
Robert G. Fitzpatrick Brookhaven National Laboratory
- 10. George P. Semienko YankeeAtcmicElectricCo.(Syst.)
- 11. John DeVincentis New Hampshire Yankee
- 12. Robert E. Sweeney NHY Bethesda Office
- 13. Lawrence A. Walsh OPSMGR NHY Seabrook Station
- 14. David A. Maidrand Asst. Proj. Manager-YNSD
- 15. Robert J. Lutz, Jr.
Westinghouse Electric Corp.
- 16. Daniel W. Stillwell Pickard, Lowe and Garrick Inc.
- 17. David H. Jaffe NRR/NRR/PWR-B
- 16. Jim Moody New Hampshire Yankee
- 19. Karl fleming Pickard Lowe & Garrick, Inc.
- 20. Kate Silburgh Harmon & Weiss
- 21. Steve Long NRC/iikR/PWR-A
- 22. Victor Benaroya NRC/NRR/PWR-A
- 23. Ocn Hickman NRC/NRR/PAF0
- 24. J. W. Shappaker NRC/NRR/PSB-A
- 25. Trevor Pratt Brookhaven National Laboratory
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SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT PLANT MODEL REVIEW by James H. Moody, Jr. - New Hampshire Yankee Karl N. Fleming - Pickard, Lowe and Garrick, Inc.
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Presentation tp U.S. NUCLEAR REGULATORY COMMISSION AND j
BROOKHAVEN NATIONAL LABORATORY j
Betitesda, Maryland August 27,1986 i
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l Pickard,Lowe and Garrick,Inc.
Engineers Applied Scientists Management Considiants e
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CONTAINMENT PENETRATION i
STEAM GENERATOR DAMAGED LARGE LOCA ECCS NOT DAMAGED CBS NOT DAMAGED j
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i AIRCRAFT CRASH MODEL POINY OF LOSS IMPACT OF CONTROL
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N = ANNUAL NUMBER OF AIRCRAFT OPERATIONS j
A = CRASH RATE (PER MILE FLOWN) u = VELOCITY g = GLIDE RATIO i
h =: ALTITUDE I
Pickard, Lowe and Garrick, Inc.
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MEETING OBJECTIVES 1
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l e REVIEW SSPSA PLANT MODEL AND UPDATES i
I e IDENTIFY PRINCIPAL CONTRIBUTORS TO EARLY RELEASE l
FREQUENCY l
e FULLY SUPPORT NRC/BNL REVJEW EFFORT i
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l MEETING AGENDA l
j APPROXIMATE l
TOPIC DURATION SPEAKER i!
INTRODUCTION 5
JIM MOODY OVERVIEW OF RESULTS 15 JIM MOODY l
SEABROOK PLANT MODEL 15 KARL FLEMING
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PRINCIPAL CONTRIBUTORS TO 5
JIM MOODY I
EARLY RELEASE FREQUENCY o CONTAINMENT ISOLATION 15 JIM MOODY e INTERFACING LOCA 30 KARL FLEMING
\\l e STEAM GENERATOR TUBE RUPTURE 10 KARL FLEMING e EXTERNAL EVENTS 15 KARL FLEMING l
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PROBABILISTIC SAFETY ASSESSMENT i
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- FULL-SCOPE LEVEL 3 PSA
- PUBLISHED DECEMBER 1983 e PRINCIPAL CONTRIBUTORS i
l YAEC - UTILITY PROJECT MANAGEMENT AND REVIEW k
PICKARD, LOWE AND GARRICK, INC. - PRA CONSULTANT e NRC REVIEW LAWRENCE LIVERMORE - PLANT MODEL - (RESPONDED l
MAY 1986)
BROOKHAVEN - CONTAINMENT CAPABILITY NUREG/CR-4540
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(FEBRUARY 1986)
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l SSPSA PLANT MODEL DEVELOPMENT I
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e SSPSA COMPLETED DECEMBER 1983 (PLG-0300) o TECHNICAL SPECIFICATION AUGUST 1985 (PLG-0431) i UPDATE o RMEPS UPDATE DECEMBER 1985 (PLG-0432) e SENSITIVITY STUDY UPDATE APRIL 1986 (PLG-0465)
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e FRAGILITY UPDATE JUNE 1986 (SMA-12911.01) i j
e CONTINUATION OF UPDATE'
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SUMMARY
OF PRINCIPAL CONTRIBUTORS TO RISK IN TERMS OF ACCIDENT SEQUENCE GROUPS AND INITIATING EVENTS FROM THE SSPSA Containment Response -
Group Group Fraction of Accident Contributing Contribution Frequency Total Release Sequence Group Initiating Events Percent (mean values)
Frequency Group I Early Containment Failure 2.4 x 10-6 per
.01 Early Health
- Interfacing LOCA 76 Reactor Year or Effects
- Seismic 24 Once in 410,000 100 Reactor Years Group II Delayed Containment Failure 1.7 x 10-4 per
.73 Latent Health
- Loss of Offsite Power 40 Reactor Year or
- Effects,
- Transients 19.
Once in 6,000
- Fires 15 Reactor Years
- Seismic 15
- Others 11 100 i
Group III Containment Intact No llealth
- Transients 57 6.0 x 10-5 per
.26 Effects
- SLOCA 29 Reactor Year or
- Others 14 Once.in 17,000 100 Reactor Years Total 2.3 x 10-4 per 1.00 Reactor Year or Once in 4,300 Reactor Years Pickard, Lowe and Garrick, Inc.
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l UPDATE OF PLANT RISK MODEL ENHANCED V-SEQUENCE MODEL ENHANCED SEISMIC ANALYSIS CONTAINMENT RECOVERY MODEL t
ENHANCED TREATMENT OF COMMON CAUSE FAILURES t
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'6 RMEPS RESULTS (1-MILE EVACUATION)
Containment Response -
Group Group Fraction of Accident Sequence Group Contributing Contribution Frequency Total Release Initiating Events Percent (mean' values)
Frequency Group I -
Early Containment Failure 3.5-7 per Reactor
.001 Early Health Interfacing LOCA (4.4-8) 13 Year or Once in Effects
- Seismic (3.0-7) 87 410,000 Reactor
- APC and TMLL (7.43-10) 100 Years Group II Delayed Containment Failure 1.6-4 per Reactor
.58 Latent Health Loss of Offsite Power 10 Year or Once in
- Jffects,
- Transients (7.06-5) 43 6,000 Reactor (S2r
- Fires 16 Years
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.Others 14 TOU Group III Containment Intact No Health
- LOSP 46 1.2-4 per Reactor
.42 Effects Transients 27 Year or Once in
- SLOCA 15 17,000 Reactor h5)
- Others 12 Years
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Exponential notation is indicated Total 2.8-4 per Reactor 1.00 in abbreviated form; Year or Once in i.e., 4.4-8 = 4.4 x 10-8 4,300 Reactor Years Pickard, Lowe and Garrick, Inc.
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4 UPDATED RESULTS FOR NO IMMEDIATE PROTECTIVE ACTIONS EARLY HEALTH RISK AT SEABROOK STATION IS:
- ABOUTTWO ORDERS OF MAGNITUDE LESS THAN NRC SAFETY GOAL 10-2 Y
S BACKGROUND ACCIDENTAL FATALITY RISK 3 10-3 (5 FATALITIES PER 10,000 POPULATION PER YEAR) 5 H
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O Safety Study Results:
CONTAINMENT EFFECTIVENESS (Percent of Accident Scenarios) r 66 %
99 %
99.9 %
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- COMPREHENSIVE COVERAGE OF ACCIDENT SEQUENCES 58 DISTINCT INITIATING EVENT CATEGORIES l
39 PLANT DAMAGE STATES (" BINS")
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- 14 RELEASE CATEGORIES
- 16 MODULARIZED EVENT TREES i,
l e FULL TREATMENT OF DEPENDENT EVENTS
- COMMON CAUSE FAILURES (SYSTEM LEVEL)
- EXTERNAL EVENTS
- INTERNAL PLANT HAZARDS
- EXPLICIT MODELING OF FUNCTIONAL DEPENDENCIES l
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FAULTS (8)
CRASHES MISSILES (6)
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OVERVIEW OF SSPSA EVENT SEQUENCE MODEL STRUCTURE INITI ATING EVENTS AUXILI ARY SYSTEM EARLY SYSTEM RESTONSE LONG TERM SYSTEM RESPONSE TR ANSIENT EVENTS OTHER EVENTS EVENT TREE EVENT TREES EVENT TREES RT FSRAC LOC TT FCRCC E.7 L LLI LL2 TLMFW FCRSW E 1.0L PLMFW FCRAC TMLL EXFW FET1 LCV FET 3 l APC l
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1 MODELED CONTRIBUTORS TO l
COMPONENT FAILURE i
t INITIAL UNAVAILABILITY t
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e HUMAN ERRORS IN TEST OR MAINTENANCE i
MISSION FAILURES t
- HARDWARE FAILURES ON DEMAND j
- INDEPENDENT
- COMMON CAUSE l
- HARDWARE FAILURES IN OPERATION I
-- INDEPENDENT l
- COMMON CAUSE e' HUMAN ERRORS IN OPERATION Pickard, Lowe and Garrick, Inc.
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DEFINITION OF 39 PLANT DAMAGE STATES USED IN SSPSA RISK MODEL
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CONDITIONS AT TIME OF RE ACTOR CONTAINMENT INT ACT AT TIME OF CORE MELT START VE15EL MELT-TitROUGH YES NO CORE MELT TIME PR E SSUR E RwST
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VE EL NITIAT D FISSIO 4 IIEAT PRODUCT (g 3cor$3A3 FRODUCT REMOVAL REMOVAL NONE flLTERED OPENING OPENING AIRCRAFT PEMOVAL ONLY ONLY
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CORE MELT WITH NotstSOL ATED SIE AtA GENERATOR TUDE RUPEURE 191 PRECluDEO SY SE ABROOK HOT USED SECAUSE OF UNCERTAINTIES PHYSICALLY POssistE SUT NOT U$ED s
STATIOtt DESIGr4 F E ATURES IN FAN COOLE R CAPAtiLITY IN 5Sr5A SISK MODEL Pickafd, Lowe and Garrick, Inc.
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TABLE 4-4 SOURCE TERM CATEGORIES Source Identified Analyzed Term Containment Failure Mode in the in This Category SSPSA St'dy u
S1 Early Containment Failure Yes Yes 52 Early. Increased Containment Leakage Yes Yes S3 Late Overpressure Failure Yes Yes 54 Basemat Melt-through Yes No*
SS Containment Intact Yes Yes 15 6 Containment Not Isolated Yes Yes S7 Containment Bypassed (V-sequence)
No Yes
- Sased on the SSPSA results, basemat melt-through sequences were assigned to category S3 in this study.
4 4
1300P112685 4-59
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l PRINCIPAL CONTRIBUTORS TO
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EARLY RELEASE FREQUENCY l
l INITIATING EVENTS WITH CONTAINMENT BYPASS
- INTERFACING LOCAs
- STEAM GENERATOR TUBE RUPTURE EXTERNAL EVENTS WITH POTENTIAL I
FOR CONTAINMENT DAMAGE EARLY l
- AIRCRAFT CRASH
- TURBINE MISSLE RELEASE 1
LOSS OF CONTAINMENT m
m STRUCTURAL INTEGRITY ALL OTHER INITIATING
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EVENTS l
CONTAINMENT ISOLATION m
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l CONTRIBUTIONS TO EARLY RELEASE FREQUENC (S1, S6, S7)
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SCENARIO TYPE PERCENT l
- 1. EARTHQUAKE WITH OPEN PURGE 87 i
- 2. INTERFACING LOCA
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- 3. TURBINE MISSILE IMPACTS CONTAINMENT BUILDING 41 i
- 4. REACTOR VESSEL STEAM EXPLOSION 41
- 5. AIRCRAFT IMPACTS CONTAINMENT BUILDING
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TOTAL 100
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ANALYSIS OF CHECK VALVE DATA 0
610 CHECK VALVE EVENTS FOUND IN LER SUMMARIES (NPE)
.(PWRs AND BWRs) 0 163 EVENTS INVOLVED LEAKAGE O
21 EVENTS INVOLVED LEAKAGE NORMALLY SEATED VALVES IN PWR PLANIS AT THE RHR/RCS INTERFACE AND ACCUMULATORS 6
0
'8 BWR " PRECURSORS" REVIEWED AND FOUND TO BE INAPPLICABLE 0'
ESTIMATE OF EXPOSURE TIME TAKEN FROM NUREG/CR-1363 1.0 x 10 8 HOURS
t RHR PIPING AND HEAT EXCHANGER STRENGTH
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e DYNAMIC EFFECTS NOT IMPORTANT BASED ON RECENT EVALUATIONS FOR THE,SEABROOK STATION
- AT NORMAL RCS PRESSURE 2,250 psia, HOOP STRESS IN LARGER RHR PIPING BELOW YlELD STRENGTH
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e PRESSURE AT ULTIMATE STRESS EQUALS 5,400 psia e PROBABILITY OF PIPE FAILURE ESTIMATED I
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o PROBABILITY OF PIPE FAILURE 1.0 g
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8 7
6 WHERE P ' CORRESPONDS TO A p
5 LOGNORMAL DISTRf 8UTION = y = 8.1673
~
~
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P ' =.01 AT YlELD p
P ' =.99 AT ULTIMATE 3
F p (2,250) = 10'3 + 5 x 10'3 = 6 a 10'3 P
2 10-3 _Po_
9 -
8 -
7 -
6 5 -
4 -
3 2 -
MATERIAL YlELD MATERIAL ULTIMATE (e - 35,000) to = 80.000) 10'd I
I I
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I 1.000 2.000 3.000 4.000 5.000 6.000 7,000 PR ESSURE (PSI A)
Pickard, Lowe and Garrick, Inc.
?
i RHR PUMP SEAL LEAKAGE i
i I
i e SEALS REMAIN INTACT BUT EXCESSIVE LEAKAGE
{
DEVELOPS
.09 inch 2 = 50 gpm e SUMP PUMPS CAPACITY
- 1.05 inches 2 ~ 550 gpm AT 425 psig (<RHR RELIEF VALVE SETTING) o SEALS ARE BLOWN OUT BUT MECHANICAL SEAL ASSEMBLY REMAINS i
- 1.3 inches 2/ PUMP
- MECHANICAL SEAL ASSEMBLY IS BLOWN OUT i
- 6.5 inches 2/ PUMP i
Pickard, Lowe and Garrick, Inc.
1 OPERATOR ACTIONS l
\\
1.
OPERATORS DIAGNOSE THE RHR SYSTEM LOCA
- PRESSURIZER RELIEF TANK LEVEL
- RHR SYSTEM RRESSURE 1
- SUMP PUMP OPERATING 2.
OPERATORS ISOLATE THE RHR SYSTEM LOCA
- COLD LEG INJECTION PATHS
- CROSS-CONNECT 3.
' OPERATORS PROVIDE MAKEUP TO THE RWST
- BORIC ACID TRANSFER PUMP
- REACTOR MAKEUP WATER PUMP j
- DEMINERALIZED WATER TRANSFER PUMP Pickard, Lowe and Garrick, Inc.
INTERFACING SYSTEMS LOCA KEY' RESULTS FREQUENCY (PER REACTOR-YEAR)
EVENT UPDATED SSPSA ANALYSIS
-6
-6 VALVE RUPTURES, LOCA 1.8 x 10 7.8 x 10
-6
-7 VALVE RUPTURES. LOCA, 1.8 x 10 7.3 x 10 CONTAINMENT BYPASS
-6
-8 VALVE RUPTURES. LOCA, 1.8 x 10 4.4 x 10 CoHTAINHENT BYPASS.
MELT I
Pickard, Lowe and Garrick, Inc.
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1 LOCA 2.9665E-06 i
2 1FV 1.424CE-10 I
3 LOOA 2.9163E 09 y
4 DILO 7.252CE-09 I
3 7FPV 3.5710E-11 8
6 DILOC 8.0577E-10 1
7 1FPV 3.9677E-12 8
DILOC 8.9631E-10 I
9 7FPV 4.4135E-12 1C DILO
- 9. 9790E-J 1 I
11 IFPV 4.9039E-13 t
F 12 DIwCC 1.03710-C3 I
13 7FPV 5.106LE-11 14 D: LOC
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15 1FPV
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- 9. 00'e 2E-10 l
I 17 1FPV 1.00CTE-10 18 DILCO
(.1131E-10 0
19 1FPV 1.236f2-11 F
20 DILOO 1.2C17h'-09 I
21 7FPV 6.3115E-12 22 IFPV 3.4312E-10 a
F-23 7FPV 4.7680E-11 I
F-24 1FPV 5.2978E-12 a-25 7FPV 5.Is930E-12 26 afPV 6.5476E-13
-=
r F-27 7FPV 6.8184E-11 a-28 1FPV 7.5759E-12 F---
29 DILDO 5.8930E-12 f
I p-30 1FPV 6.5478C-13 a-31 DILCO 7.2835E-13 I
a-32 IFPV 8.09282-14
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w-34 1FPV 9.3635E-13 r~a-a-a 35 DILOC 1.0387E-07 I
36 7FPV 5.1149E-10 t
37 1FPV 1.159BE-06 7.--o -- -w--a
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38 7FPV 6.8294E-10 b
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40 DILOC 1.2956E-07 1
4g 7ppy g,37,6E-10 42 1FPV 1.4466E-09 F-Cr-F-r F-43 7FPV 8.5162E-10 F-44 1FPV 9.4646E-11 u - U - F---O 8-45 7FPV 2.6571E-09 W-46 1FPV 2.9523E-10 7-r-F-2 F-47 7FPV 1.7384E-11 g
48 IFPV 1.9316E-12 2
49 1FV 1.76C3E-09 50 1FV 1.4CB3E-11 Fickard, Lowe and Garrick,, Inc.
i i
l
SGTR ANALYSIS J
- D INITIATING EVENT FREQUENCY BASED ON LEAKS > 75 spm - 1.38 X 10-2 1
~
ASSUME SINGLE TUBE OFFSET RUPTURE - 7.24 X 10 0-SEPARATE EVENT TREES O
ANALYSIS BOUNDS RISK' CONTRIBUTION OF MULTIPLE TUBE RUPTURES 0
GEPARATE EVENT TREE MODULES FOR SGTR INCLUDE:
4 INCLUDED PRESSURIZED THERMAL SH0dK SCENARIOS i
RCP SEAL LOCA SCENARIOS INCLUDED FRONT lit?E SYSTEMS FOR SHUTDOWN, INVENTORY CONTROL, RCS HE AT R EMOV AL, CONTAINMENT SPRAY (SUPPORT SYSTEMS MODELED IN SEPARATE TREE) 4 07E RATOR A CTIONS TO CONTRO L BREAK FLOW, ISOLATE SG,
DEPRESSURIZE RCS, PROVIDE LONG TERM RCS MAKEUP AND HEAT CHECKED ON OPERATIONS TR AINING SIMULATOR
-~ REMOVAL INCLUDED;
' UNIQUE RELEASE PATHS TRACKED WITH SEP ARATE PLANT DAM STATES MAIN STEnMLINE DAMAGE DUE TO SG OVERFILLING MODELED
' MALL CONTRIEUTOR TO RISK O
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e RELATIVE RISK CONTRIBUTION OF SGTR ACCIDENT SEQUENCES.
ALL MODELED SGTR SEQUENCES
. SEQUENCES CORE MELT l
. FREQUENCY 2.8 X 10-4 1.7 X 10-6 (events / year l
<'PERCEhi OF 60/% MR,T I
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< 0.1 FAILURE EARLY GROSS 01
<< 01 FAILURE l
STEPS IN TURBINE MISSILE AWYSIS 1.
IDENTIFY MISSILE SCENARIOS O
FREQUENCY OF MISSILE GENERATION f g NORMAL SPEED OVERSPEED 0
MISSILE TRAJECTORY
]
]
HIGH
]
]
f2 f3 LOW
]
]
]
O TARGET IMPACT AND DAMAGE
]
~
0 IMPACT ON PLANT SYSTEMS 2.
QUANTIFY MISSILE SCENARIO FREQUENCIES f i f 2 f 3 AS PLANT MODEL INITIATING EVENTS 3.
ASSIGN IMPACT VECTOR (EVENT TREE TOPS) [1001110.. 0]
4.
QUANTIFY PLANT EVENT SEQUENCE MODEL 5.
D5TERMINE RISK IMPACT 6
j l
1
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l
4 TURMINE MISSILE FREQUENCY i
l SOURCE FREQUENCY FRACTION OF FAILURES AT
+
(EVENT PER YEAR)
NORMAL SPEED OVERSPEED l
I BUSlf AND IIEASLER 1.6 X 10-4
.72
.28 GE STATISTICS 1.4 X 10-4 I
CE ANALYSIS 1.4 X 10-8
.64
.36 SSPSA (MEAN)
- 8. 3. X 10-5
.76
.24' s
1 Wan 5th Percentile Median 95th Percentile I
I
~8
-7
-6
-5
-3 1,0 10 10 3p 10,-4 1p o
TURBINE MISSILE FREQUENCY (EVENTS PER REACTOR - YEAR)
I
TURDINE MISSil_E SCEN'ARIOS lill_TN CONTAINUENT DAMAGE
.i Turbine Missile Turbine Speed Missile Conditional Scenario Trajectory Frequency.of Frequency Generation Containment (per reactor (per reactor (fraction)
Damage
. year) year)
O o
Low
.76 Normal
-9 4.5 x 10 '
2.9 x 10 8.3 x 10 High 0
-5 0
Unit 1 Low
.24 Overspeed 8
5.8 x 10 1.2 x 10
-5 High
-5 2.6 x 10 8
4.1 x 10 Low
.76 Normal 8
8.7 x 10 5.5 x 10
-5 8.3 x 10 High
-5 8
7.1 x 10 1.4 x 10 d
Unit 2 Low
.24 Overspeed
-o 5.4 x 10 I.! x 10 d
-o High 6.1 x 10
\\
OVERVIEW OF AIRCRAFT CRASH ANALYSIS l
e CONSIDERED ALL SOURCES OF OVERHEAD TRAFFIC l
MILITARY (PEASE AFB)
COMMERCIAL (INCLUDING LOGAN AIRPORT TRAFFIC)
OTHER SMALL AIRPOJ1TS e MODELED SEVEN SITE STRUCTURES AND DIFFERENT PLANT IMPACT l
e MODELED CRASHES DUE TO MIDAIR COLLISIONS l
~
e UTILIZED 28 MONTHS OF AIR TRAFFIC DATA FROM PEASE AFB CONTROL TOWER 4
l e SIMPLIFIED / CONSERVATIVE STRUCTURAL FRAGILITY ANALYSIS e AIRCRAFT CRASH INTO PRIMARY CONTAINMENT ONLY I
NOTICEABLE RISK CONTRIBUTOR l
j Pickard, Lowe and Garrick, Inc.
i CONTAINMENT ISOLATION l
PENETRATION REVIEW I
\\
PENETRATION CONTAINMENT TWO PUR'GE PENETRATION DIAMETER REVIEW
> 3 INCHES PENETRATlONS l
NO l
I 1r i
j SEVEN l
PENETRATIONS 1r 1r SEQUENCES SEQUENCES PDS: 1 FP,3FP,7FP PDS: 1F, 3F, 7F l
RC: S2 RC: S6 Pickard, Lowe and Garrick, Inc.
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l i
l I
SMALL PENETRATION ISOLATION FAILURE MODEL l
l CORE MELT AC POWER YES ACTUATION YES m
ONAL SEQUENCE AVAILABLE EE NO NO l
V BOTH
/SUCCESSFUL OPERATOR YES U
CONTAINMENT NO
/
j m
E CONTAINMENT i
RECOVERY ISOLATION I OMTION VALVES FAIL NO YES V
[SMALL m/ PENETRATION
"\\ ISOLATION
\\ FAILURE i
9 l
Pickard, Lowe and Garrick, Inc.
CONTAINMENT PURGE ISOLATION FAILURE MODEL l
BOUNDARY OF CONTAINMENT l
[ ISOLATION SYSTEM MODEL
______________q l
I 1
CORE MELT l
CONTAINMENT NO l
m I
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Pickard, Lowe and Garrick, Inc.
i l
1 j
i SYSTEMS ANALYSIS RESULTS DOMINATING RISK e EARLY HEALTH. RISK (RC: S6)
)
- EARTHQUAKE INITIATING EVENTS
- SOLID STATE PROTECTION SYSTEM FAILURE
- CONTAINMENT ISOLATION FAILURE 1.0
=
- PURGE VALVES OPEN 0.1
=
I I
e LATENT HEALTH RISK (RC: S2)
. STATION BLACKOUT OR SOLID STATE PROTECTION SYSTEM FAILURE
- CONTAINMENT ISOLATION FAILURE 1.0
=
- PURGE VALVES CLOSED 1.0
=
l l
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- RHR FAILURE AND CONTAINMENT BYPASS l
- NO CREDIT FOR THE OPERATOR e LOW FREQUENCY BUT DOMINANT CONTRIBUTOR TO l
EARLY HEALTH RISK l
i e ANALYSIS BELIEVED TO BE CONSERVATIVE
]
RHR PIPING FAILURE NOT GUARANTEED
~
RHR RELIEF VALVES MAY MITIGATE SOME SEQUENCES RHR PUMP SEALS MOST PROBABLE POINT OF FAILURE RHR VAULT FLOODING LIKELY
. OPERATOR CAN ISOLATE AND RECOVER MANY SEQUENCES e POTENTIAL NONCONSERVATISMS
- NEED TO REDEFINE INITIATING EVENT AS ANY VALVE I
FAILURES THAT LEAD TO RHR SYSTEM PRESSURIZATION
- LEADS TO INCREASE IN INITIATING EVENT FREQUENCY Pickard, Lowe and Garrick, Inc.
}
COLD LEG INJECTION PATH ARRANGEMENT ACC ff!Gil LOW PRESSURE PRESSUR E 3,,yg M
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TRAIN 8 E
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RH V2G SIV35 RH V63 Ril 29 I
INSIDE CONTAINMENT PIPE TUNNEL RHR VAULT 4---
Pickard, Lowe and Garrick, Inc.
1 RHR SECTION PATH ARRANGEMENT I
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PRESSURE '
l
^ PRESSURE l
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Pickard, l. owe and Garrick, Inc.
r ENHANCED TREATMENT OF INTERFACING SYSTEMS LOCA l
i
- MORE COMPLETE MODELING OF VALVE FAILURE MODES e NEW DATA ON CHECK VALVE FAILURES VERSUS LEAK SIZE e MORE REALISTIC TREATMENT OF DYNAMIC PRESSURE PULSE 4
l
)
e EXPLICIT MODELING OF RHR RELIEF VALVES e QUANTIFICATION OF RHR PIPING FRAGILITIES TO OVERPRESSURE e MODELING OF RHR PUMP SEAL LEAKAGE j
e OPERATOR ACTIONS TO PREVENT MELT CONSIDERED i
e THERMAL HYDRAULIC AND SOURCE TERM FACTORS MODELED USING MAAP e UNCERTAINTIES QUANTIFIED a
l Pickard, Lowe and Garrick, Inc.
SIMPLIFIED MODEL i
l CbcctYul6 b
l LEAKAGE LEAKAGE RHR RHR PLANT
> 150 GPM
< RV CAP PIPING SEALS OPERATOR IMPACT gT I
LOCA i
>QR l"
LOCA i
i i
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i LOCA/
3 BYPASS i
MELT l
1 MELT Pickard, Lowe and Gyrick, Inc.
1
i
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VALVE FAILURE MODES i
M@ f W o
l e DISC RUPTURE OF SERIES VALVES (CHECK VALVES AND 3
j MOVs) i W
l OL Y e DISC RUPTURE (STEM MOUNTED LIMIT SWITCH) AND DISC FAILING OPEN WHEN INDICATING CLOSED i
- NEW MODEL IS MORE COMPLETE
\\
~
i l
OLD MODEL NEW MODEL 1
<APATH > = V212T
<APATH > = 1 T + 211 + M d
g i
i Pickard, Lowe and Garrick, Inc.
w
VALVE FAILURE DATA i
e NUCLEAR POWER EXPERIENCE REVIEW (1972-1984)
- NO DISC RUPTURE EVENTS e 8 BWR EVENTS CONSIDERED
- 21 EVENTS WITH LEAK RATES FROM 5 TO 200 gpm e TOTAL EXPOSURE TIME ESTIMATED 108 CHECK VALVE HOURS e DATA POINTS PLOTTED AGAINST LEAK RATE Pickard, Lowe and Garrick, Inc.
c 94 SEP 19o6 Two V sequence possibilities asscciated with the RCS boundary were identified.
These were cold leg injection path arrangement and RHR section path arrange-ment. The applicant stated that the most important aspect of the enhanced treatment of interfacing systems LOCA is the explicit consideration of RHR relief valves in containment.
Mr. Jim Moody continued the applicant's presentation with a discussion of treatment of containment isolation in the updated study. He stated that a total of nine penetrations were included in the model for quantification.
In the small penetration isolation failure model, the earthquake common cause initiating event dominates its contribution to risk.
For larger containment purge valves the earthquake with loss of offsite power is the dominant con-tributor to risk.
Other aspects of the study discussed were SG Tube Rupture and the consequence analysis. The applicant showed that SG Tube Rupture is not important because of the considerable amount of time available for operator action.
The meeting was adjourned at 3:00 p.m.
A list of slides used by the applicant during his presentation are included as Enclosure 2.
l Elizabeth Doolittle, Project Manger PWR Project Directorate #5 Division of PWR Licensing-A Enclosures UTC :P PLA NAME :E ttle :
__...:...g.......
DATE :9/ A/86 v
0FFICIAL RECORD COPY