Text

Suggests That Advisory Opinion Favoring Reduction of Radius of EPZ from 10 to 2 Miles Flies in Face of Lessons Learned. Opinion Would Violate Technical Bases of NUREG-0654 & FEMA-REP-1.Related Info Encl
	ML20206J395
Person / Time
Site:	Seabrook
Issue date:	10/10/1986
From:	Mavroules N; HOUSE OF REP.
To:	Ward D; Advisory Committee on Reactor Safeguards
Shared Package
ML20205K769	List: ML20205K763; ML20206J224; ML20206J246; ML20206J276; ML20206J291; ML20206J329; ML20206J395; ML20206J464; ML20206J485;
References
	FOIA-87-7, RTR-NUREG-0654, RTR-NUREG-654 NUDOCS 8704160085
	Download: ML20206J395 (165)
	v • d • e

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October 10, 1986 David Ward, Chairman Advisory Committee on Reactor Safeguards 1717 H Street, N.W.

Washington, D.C. 20555

Dear Chairman Ward:

During your deliberations on the status of Seabrook Station probabilistic safety analysis, I understand you will be reviewing the matter of reducing in radius the Emergency Planning Zone' from ten (10) miles to just two(2) miles.

I would forcefully suggest that an advisory opinion in favor of such a concept flies in the face of the lessons learned from the Three Mile Island accident and recently confirmed by the

( Chernobyl catastrophe. Such an opinion ~would violate the technical basis for the N..R. C. 's and F.E.M. A. 's -joint declaration of -

Criteria for Preparation and Evaluation'of Radio' logical Emergency

'hesponse Plans and Preparedness in Support of Nuclear Power Plants (NUREG-0654/ FEMA-REP-1). And such an opinion, if respected by the Commissioners, would work to destroy the credibility of the N.R.C. as an independent, regulatory body.

For the record, I am unequivocally opposed to any reduction in the radius of the' Emergency Planning Zone.

Thank you for your anticipated courtesy and consideration.

Sincerely, N cholas Mavroules Member of Congress ,

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' JOINT PEETING OF THE ACRS SUBC(ft1ITTEES ON OCCUPATIONAL AND ENVIROWEFAL NTECTION SYSTEMS AND SEVERE (CLASS 9) ACCIDENT SEAPR00K STATION SEPT 9 BER 26, 1986 t

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VINcFNT S NOONAN, DIRECTOR FWR PROJECT DIPECTORATE No. 5 DIVISION OF PWR LICENSING-A t

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REVIEW OF UPDATED PPfRABILISTIC SAFETY ASSES 9ENT FOR SEABROOK STATION ORJECTIVES

, TO REVIEW TECHNICAL ADEQJACY OF SEABROOK STATION RISK MANAGEN NT AND EN RGENCY PLANNING SIUDY AND EN RGENCY PLAMING SENSITIVIT/ STUDY ,

E0gg

. CONTAIWENT STRUCTURAL INTEGRITY

, 00NTAIENT BYPASS

. INTERFACING SYSTEMS LOCA RESULTS

( , IDENTIFICATION OF SIGNIFICANT TECHNICAL ISSIES

, EVALUATION OF MAJOR RISK CONTRIBilTORS

, IDENTIFICATION OF ADEAS FOR FURTHER RISK PEDUCTION

ee 9

I.

SLIDE 2

4 .

4 CURRENT STATUS OF SEABROOK REVIEW .

. SAFETY EVIEW FIE PROTECTION INCOMPLEIE

, EERGENCY PLANNING EVIBi ONSITE COMPLETE APPRAISALS EXERCISE OFFSITE INCOMPLETE N. H. EXERCISE PERFORW D PASS. EXERCISE TO BE PERF0 PED COORDINATION WITH FEPA

~

EGIONAL ASSISTANCE C0mITTEE

, HEARINGS EERGENCY PLANNING ONSITE AUGUST, 1983 TO BE SUPPLEE NTED BY HEARING DURING THE WEK OF SEPTEMBER 29,1986 l

_ EERGENCY ACTION LEVELS OFFSITE NEW HAWSHIRE - POSTFONED FROM AUGUST,19PE AND NOT YET ESCHEDULED PASSAGUSETTS - PLANS NOT FDPFALLY StBMITTED TO FEPA FOR EVIEW SAFEiY SEPTEEER 29,1986

_ ENVIRONENTAL OJALIFICATION OF EQUIPEhT

.. CONTPOL ROOM DESIGN EVIEW SLIDE 3

. . . . ~ . .

4 PESULTS OF LATEST SALP ,

( l PERIOD ENDING PERIOD ENDING RECENT TPEND R NCTIONAL APEA (12/ 31 /84 ) (3 /31 /86 )

A. CONSTRUCTION 2 1 CONSISTENT B. PREOPERATIONAL 1 1 CONSISTENT TESTING C. FIRE PROTECTION AND HOUSEKEEPING N/A 1 CONSISTBir D .' OPERATIONAL READINESS N/A 1 CONSIST 9rr E. EERGENCY PREPAREDNESS N/A 2 IPPROVING

('

F. ASSURANCE OF GJALITY I I CONSISTENT G. LI NSING 2 1 C0tSISTENT SLIDE 4

1 NRC STAFF PRESENTATION TO THE ACRS

-. ON THE REVIEW PLAN FOR THE SEABROOK EPERGENCY PLAlflING SENSITIVITY STUDY SCOPE AND FOCUS OF STAFF REVIEW S. LONG (NRR)

CCPPARISON OF PLGOS6 WIm NJREG-0396 D. PkTmEWS (IE)

Pkm0DOLOGY AND STATUS OF REVIEW:

SOURCE TERMS T. PRATT (BNL)

RISKANALYSIS " "

CONTAINPENT STRUCTURAL INTEGRITY C. HOFMAYER (BNL)

CONTAINMENT BYPASS R. YOUNGBLOOD (BNL) i INTERFACING SYSTEM LOCA l

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PURPOSE OF REVIEW TO IDENTIFY AND REVIEW THE PORTIONS OF THE STUDY THAT ARE M)ST SENSITIVE WITH RESPECT TO THE STUDY'S PRINCIPAL CONCLUSIONS Il0lVIDUAL RISK OF EARLY FATALITY AT SEABROOK IS WITHIN SAFETY GOAL 1 MILE EVACUATION AT SEABROOK PROVIDES SIMILAR RISK OF EARLY FATALITIES TO THE WASH-14)0 RESULTS WITH 25 MILE EVACUATION 2 PR2 ABILITIES OF SPECIFIC RADIOLOGICAL EXPOSURE LEVELS AT 1 MILE FROM SEABROOK ARE LESS THAN THE CORRESPONDING PROBABILITIES SHOWN AT 10 MILES IN NUREG-03%

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MAJOR CONTRIBUTORS TO RISK .

i DIFFERENT CONTRIBUTORS FOR DIFFEPBIT RISK COWARISONS:

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- PROSABILITY OF EARLY FATALITIES, GIVEN EVACUATION, APPEARS TO HAVE CONTRIBUTIONS FR&l SEVERAL RELEASE CATAGORIES AND EVENT INITIATORS, EVENT V WAS DOMINANT IN ORIGINAL PSA,

- DOSE VS DISTANCE OJRVES (N0 EVACUATION) ARE DOMINATED BY SINGLE RELEASE CATAGORY AND SEISMIC EVENT INITIATORS.

THESE CURVES WERE NOT PREJCED IN ORIGINAL PSA,

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e F00JS OF STAFF REVIEW EFFORT .

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FACTORS SHOWN BY CURRENT STUDY TO BE IlPORTANT FOR RISK MITIGATION O

AREAS WHERE SIGNIFICANT RISK REDUCTION OCCURRED BETWEEN UPDATE AND ORIGINAL PSA O

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CONTAlffENT RESPONSE AT HIGH PRESSURE

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c. Source: Lipinski et al,1985

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i Estimated Release Fractions from Containment for a Surry TML8' Secuence. Accountino for ;

Phenomenological Uncertainties G .

Containment Failure WASH-1400 Element Early ,. Late PWR 2 Value l .

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! Cs 0.004 to 0.8 0 to 0.004 0.5

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!, Sr 3E-6 to 0.8 0 to 0.02 0.06 Ru 3E-7 to 0.2 0 to 1E-5 0.02 .

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TABLE 2-4.* COMPARISON OF CORE MELT FREQUENCIES AND DISTRIBUTIONS OF RELEASE TYPES Rist Parameter WASH-1400 33p3A Upcatec PWR Results e Mean Core Melt Frequency (events 9.9-5* 2.3-4 2.7-4 per reactor-year) e Percent Contribution of Release Types

- Gross, Early Contairenant 34 1 0.1 Failure Gradual Containment 66 73 60 Overpressurization or -

Melt-Through

. Containment Intact 0 26 40

'Sased on WASH-1400 uncertainty ranges.

MOTE: Exponential notation is indicated in ambreviated form; i.e., 9.9-5 = 9.9 x 10-5, e

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SEABROOK STATION ,

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PROBABILISTIC SAFETY ASSESSMENT AND '

EMERGENCY PLANNING STUDY l

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Presentation to l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS .

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i October 10,1986 l

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1 Agenda e Project Overview

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O Low PowerTesting 9 Emergency Planning

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Construction 1 PreoperationalTesting 1 Fire Protection And Housekeeping 1 Operational Readiness 1 Emergency Preparedness 2 -

Assurance Of Quality 1 Licensing 1 9

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Low Power Testing e No Fuel Load Open items With NRR Or Region 1 -

4 ASLB (Low Power License)

- Open items O Equipment Qualification Time Duration i O Detailed Control Room Design Review- SPDS O Emergency Action Levels

! - Hearing Held 9/29-10/3 -

In Portsmouth, N.H.

! - Petition 50.57 (c) To Load Fuel And Conduct Precriticality Testing Submitted 8/22/86 l

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Emergency Planning e Graded Exercise Held 3/86 (N.H. Only)

( e Risk Management And Emergency Planning Study Submitted To NRC

-7/21/86 O Revision 2 Of N.H. Plan Submitted 9/86 e Gov.Of Mass. Announced RefusalTo Submit Plans To FEMA- 9/20/86 4

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l PRESENTATION OUTLINE

SSPSA (1983) OVERVIEW j e RMEPS (1985) OVERVIEW i

e WASH-1400 METHODOLOGY SENSITIVITY STUDY

TECHNICAL RESULTS

TECHNICAL BASES FOR RESULTS i

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e CONCLUSIONS !

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l SSPSA RISK MODEL DEVELOPMENT e SSPSA COMPLETED DECEMBER 1983 (PLG-0300) ,

e TECHNICAL SPECIFICATION AUGUST 1985 (PLG-0431)

UPDATE e RMEPS UPDATE DECEMBER 1985 (PLG-0432) e SENSITIVITY STUDY UPDATE APRIL 1986 (PLG-0465) e FRAGILITY UPDATE JUNE 1986 (SMA-12911.01) e SSPSA UPDATE IN PROGRESS

NRC REVIEW PLANT MODEL REVIEW - LAWRENCE LIVERMORE (1985)

, - PSNH RESPONSE TO PLANT MODEL REVIEW (MAY 1986)

- REVIEW OF CONTAINMENT RESPONSE, SOURCE TERMS, AND CONSEQUENCES -- BROOKHAVEN (BNL), NUREG/CR-4540 (FEBRUARY 1986) 1

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i SSPSA SCOPE AND COVERAGE OF .

ACCIDENT SEQUENCES e COMPREHENSIVE COVERAGE OF ACCIDENT SEQUENCES

- 58 DISTINCT INITIATING EVENT CATEGORIES

- 39 PLANT DAMAGE STATES (" BINS")

- 14 RELEASE CATEGORIES

- 16 MODULARIZED EVENT TREES e FULL TREATMENT OF DEPENDENT EVENTS

- COMMON CAUSE FAILURES (SYSTEM LEVEL)

- EXTERNAL EVENTS i

- INTERNAL PLANT HAZARDS .

- EXPLICIT MODELING OF FUNCTIONAL DEPENDENCIES e PLANT-SPECIFIC AND ENHANCED CONTAINMENT MODEL

- ASSESSMENT OF CONTAINMENT FAILURE MODES !

- QUANTIFICATION OF SOURCE TERM UNCERTAINTIES

- ENHANCED METHODOLOGY

SITE-SPECIFIC CONSEQUENCE MODEL

- MULTIPUFF RELEASE TREATMENT i - ACTUAL SITE CHARACTERISTICS

- QUANTIFICATION OF UNCERTAINTY I

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PRINCIPAL RISK CONTRIBUTORS IN THE SSPSA (1983) .

Accident Containment Response - Group Group Fraction of Sequence Group Contributing Contribution Frequency Total Core Damage 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 Effects - Seismic 24 or Once in l . lRRI 410.000 Reactor l Years I

Group II Delayed Containment Failure 1.7 x 10-4 per .73 l Latent Health -

Loss of Offsite Power 40 Reactor Year Effects -

Transients 19 'or Once in

- Fires -

15 6.000 Reactor

- Seismic 15 Years Others 11 TUU Group III Containment Intact No Health -

Transients 57 6.0 x 10-5 per .26 E f fects -

SLOCA 29 Reactor Year l -

Others 14 or Once in l "lRRT 17.000 Reactor Years Total 2.3 x 10-4 per 1.00 Reactor Year j or Once in 4.300 Reactor Years 9

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, SSPSA CONCLUSIONS (1983) .

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e EARLY HEALTH RISK -

- NRC SAFETY GOAL MET WITH LARGE MARGINS l

- INTERFACING LOCA DOMINATES e LATENT HEALTH RISK -

- NRC SAFETY GOAL MET WITH VERY LARGE MARGINS

- SUPPORT SYSTEM FAILURES DOMINATE , :

l e CORE MELT FREQUENCY - 2 x 10-*/ REACTOR YEAR l,

CONTAINMENT EFFECTIVENESS i

- PRIMARY CONTAINMENT VERY STRONG 1 - EARL FAILURE UNLIKELY .

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,LONG TIME FOR OVERPRESSURE l

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Unique Features .

e Ultimate Containment Strength e Secondary Containment Enclosure Building e Two Service Water Systems -

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Steam Generator Secondary Water Inventory

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RMEPS OBJECTIVES e REEXAMINE TECHNICAL BASIS OF THE 10-MILE EPZ (NUREG-0396) ON A PLANT-SPECIFIC BASIS

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e DEVELOP AN ENHANCED PRA METHODOLOGY FOR ESTABLISHING A PLANT AND SITE-SPECIFIC EPZ e APPLY THIS METHODOLOGY TO SEABROOK STATION

- UPDATE SSPSA RISK MODEL (1983 - 1985)

- DETERMINE RISK IMPACT OF EMERGENCY PLAN OPTIONS

~

ADDRESS UNCERTAINTIES AND SENSITIVITIES ;

e PROVIDE DOCUMENTATION AND PEER REVIEW 9

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ENHANCED METHODOLOGY FOR EPZ DETERMINATION e DEVELOP NUREG-0396 RISK OF DOSE VERSUS DISTANCE CURVES BASED ON PLANT / SITE-SPECIFIC RISK MODEL e CHARACTERIZE TOTAL POTENTIAL FOR RISK REDUCTION '

BY PROTECTIVE ACTION AS RISK WITH NO EVACUATION e QUANTIFY SPATIAL DISTRIBUTION OF NONEVACUATION RISK .

1 e CALCULATE ACTUAL RISK REDUCTION FOR PROTECTIVE ACTION STRATEGIES '

MILE EVACUATION i i

MILE EVACUATION l MILE EVACUATION

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MILE EVACUATION AND SHELTERING OUT TO 10 MILES ,l e' EVALUATE UNCERTAINTIES AND SENSITIVITIES l e COMPARE RESULTS WITH ALL AVAILABLE RISK l ACCEPTANCE CRITERIA

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RISK ACCEPTANCE CRITERIA UTILIZED e NUREG-0396 DOSE VERSUS DISTANCE CURVES FOR 1,5, 50, AND 200-REM WHOLE-BODY DOSES e WASH-1400 RISK CURVES FOR EARLY FATALITIES AND LATENT CANCER FATALITIES (MEAN AND MEDIAN RESULTS)

NRC INDIVIDUAL AND SOCIETAL RISK SAFETY GOALS e SPATIAL DISTRIBUTION OF RESIDUAL RISK i 4

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RMEPS UPDATE OF SSPSA RISK MODEL i

e UPDATED.SSPSA PLANT MODEL

- ENHANCED V-SEQUENCE MODEL

- ENHANCED SEIStilC ANALYSIS

- CONTAINMENT RECOVERY MODEL t

I - ENHANCED TREATMENT OF COMMON CAUSE FAILURES I

e UPDATED SSPSA SOURCE TERMS .

- EXISTING SSPSA SOURCE TERMS

- INCORPORATED SOME ZION IDCOR RESULTS

- PERFORMED SEABROOK/ ZION DESIGN COMPARISON

- DEVELOPED SOME SEABROOK RESULTS WITH MAAP

- REASSESSED UNCERTAINTIES

- EXAMINED SENSITIVITIES

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RMEPS RESULTS (1-MILE EVACUATION)

Accident Containment Response - Group Group Fraction of Sequence Group Contributing . Contribution Frequency Total Core Damage ,

Initiating Events Percent (mean values) Frequency l

Group I Early Contal:nment Failure 3.5-7 per Reactor .001 Early Health - Interfacing LOCA 13 Year or Once in Effects - Seisaic 87 410,000 Reactor

- APC and TMLL >1 Years

! 100 l Group II Delayed Containment Failure 1.6-4 per Reactor .58

Latent Health - Loss of Offsite Power 10 Year or Once in Effects - Transients 43 6,000 Reactor

- Fires 16 Years

- Seismic 17 .

- Others 14 105 i .

! Group III Containment Intact -

l No Health - LOSP 46 1.2-4 per Reactor 42 i Effects -

Transients 27 Year or Once in

- SLOCA 15 17,000 Reactor ,

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- Others 12 Years W

i NOTE: Expo'nential notation is indicated Total 2.8-4 per Reactor 1.00

- in abbreviated form; -

Year or Once in i 1.e., 4.4-8 = 4.4 x 10-8 *

. 4,300 Reactor Years l

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RMEPS KEY RESULTS j

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e EARLY HEALTH RISK WITH NO EVACUATION IS:

- LESS THAN WASH-1400 WITH 25-MILE EVACUATION

- MEETS NRC SAFETY GOAL WITH WIDE MARGIN

- CONFINED TO AREA CLOSE TO THE SITE -

l e VERY SMALL RISK REDUCTION BY ANY EVACUATION t

! e ALL NUREG-0396 DOSE VERSUS DISTANCE CRITERIA SATISFIED AT 1 MILE OR LESS e LATENT HEALTH RISK INSENSITIVE TO ASSUMPTIONS REGARDING EVACUATION -

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l EMERGENCY PLANNING SENSITIVITY STUDY i

J METHODOLOGY l- e PURPOSE: DETERMINE IMPORTANCE OF SOURCE TERMS !

I VERSUS PLANT-SPECIFIC FEATURES AND i ENHANCED PRA TECHNOLOGY !

APPROACH: RMEPS CALCULATIONS REDONE USING:

- W SH-1400 SOURCE TERM METHODOLOGY i

- BEET ESTIMATE ASSUMPTIONS ON ALL OTHER UNCERTAIN

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n e m PEER REVIEW GROUP e ROBERT BUDNITZ, CHAIRMAN, FUTURE RESOURCES ASSOCIATES, INC.

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DAVID ALDRICH, SCIENCE APPLICATIONS INCORPORATED e JOSEPH HENDRIE, CONSUt TANT e NORMAN RASMUSSEN, MASSACHUSETTS INSTITUTE OF TECHNOLOGY e ROBERT RITZMAN, ELECTRIC POWER RESEARCH .

INSTITUTE o WILLIAM STRATTON, CONSULTANT e RICHARD WILSON, HARVARD UNIVERSITY

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p PEER REVIEW FINDINGS i

e CONCURRED WITH PRINCIPAL STUDY FINDINGS

- OVERALL OFFSITE RISKS VERY SMALL

- EARLY HEALTH RISK LOWER THAN THOUGHT TO EXIST WHEN GENERAL EPZ ESTADLISHED

- EARLY HEALTH RISK CONFINED TO AREAS VERY CLOSE TO REACTOR e CONCLUSION ROBUST EVEN IN LIGHT OF UNCERTAINTIES e BELIEVE THE "BEST ESTIMATE" PROBABLY

! OVER-ESTIMATES ACTUAL CONSEQUENCES e SEABROOK CONTAINMENT MAJOR FACTOR I

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! FAVORABLE RESULTS DUE TO

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CONTAINMENT EFFECTIVENESS

ENHANCED V-SEQUENCE MODEL e SOURCE TERMS S

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Safety Study Results: '!

, . CONTAINMENT EFFECTIVENESS -

j I', Percent of Core Damage Frequency) 66 % 99 % 99.9 %

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.. Lj Il lE lj [ [ 34 % 1% .1 %

WASH-1400 (1975) SEABROOK STATION (1983) SEA T CONTAINMENT MODEL(1985) .

ILURI N NTACT W

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CONTAINMENT FAILURE TYPES i

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l A. SMALL LEAK (0.02 SQUARE INCHES TO 6 SQUARE INCHES)

PRESSURE RISE CONTINUES B. LOCAL FAILURE (6 SQUARE INCHES TO 60 SQUARE INCHES)

PRESSURE RISE STOPS EXTENDED RELEASE (>1 HOUR)

C. GROSS FAILURE (>60 SQUARE INCHES)

RAPID CONTAINMENT BLOWDOWN (<1 HOUR)

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I LOCAL CONTAINMENT FAILURE MODES CONSIDERED i ;

i e FLUID SYSTEMS PENETRATION e HIGH ENERGY PENETRATION e FUEL TRANSFER TUBE i '

t e ELECTRICAL PENETRATION e PURGE LINE PENETRATION l e PURGE VALVE SEALS l e EQUIPMENT HATCH e PERSONNEL LOCK e OTHER PENETRATIONS i -

e LINER TEARING

!I i

e WELD IMPERFECTIONS 1

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CONTAINMENT FAILURE MODES AND TYPE -

.i Medfan y,ggan Lyomal Fatlure Fatture l"I Failure Standerd Mode ; Pressure Type Devfation (psfal Ares g Strwetural Fallure Modes Cylinder Wall floos 231 Larges C .12 Dome Hoos or Mertdional 238 Larges C .12 I

Wall Merfdfonal 296 Larges C .12 Oase slab Shear 338 Larges C .23 Base Stab Flexure 415 Larges C .25 Wall Shear at Base 423 Larges C .30 Local Faf1ere Modes .

Feetheater Penetratfon 194 Self-Regulatingh 8 0.5 Flue Head Feeabeater Pf pe Crushing 231 Self-Regulatingb 5 .12 Fuel Transfer Tube > 260C 5el f-Regulatingb g d tellows -

Penetrations X-25, X-26, 181 0.5 Square Inch A 0.16 X-27 Each All Otherse > 231C 5el f-Regulatingh a d i

aMuch larger than 0.5 square foot.

b leak area Is self-adjusting to stop pressure rf te.

cProbabflIty of fallure Is less than 50% at ultf* ate wall hoop capacity, dFallure pressure model not legnormal. *

' Composite estf* ate of Ifner adhesfon, interocracks, weld faults, equfoment hatch, other mechanical penetrattons, and electrical penetrations.

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COMPOSITE CONTAINMENT FAILURE PROBABILITY DISTRIBUTIONS FOR TYPE B (LEAK) FAILURE, TYPE C (GROSS)

FAILURE, AND TOTAL FAILURE

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ENHANCED TREATMENT OF INTERFACING

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i SYSTEMS LOCA e MORE COMPLETE MODELING OF VALVE FAILURE MODES

!

NEW DATA ON CHECK VALVE FAILURES VERSUS LEAK SIZE

! e MORE REALISTIC TREATMENT OF DYNAMIC PRESSURE PULSE l e EXPLICIT MODELING OF RHR RELIEF VALVES e QUANTIFICATION OF RHR PIPING FRAGILITIES TO j OVERPRESSURE l e MODELING OF RHR PUMP SEAL LEAKAGE o OPERATOR ACTIONS TO PREVENT MELT CONSIDERED .

e THERMAL HYDRAULIC AND SOURCE TERM FACTORS MODELED USING MAAP e ' UNCERTAINTIES QUANTIFIED O

n - m '..

INTERFACING SYSTEMS LOCA KEY RESULTS FREQUENCY (PER REACTOR YEAR) l UPDATED EVENT SSPSA ANALYSIS VALVE RUPTURES, LOCA 1.8 x 10-s 7.8 x 10-8 VALVE RUPTURES, LOCA, 1.8 x 10-s 3.1 x 10-7 l CONTAINMENT BYPASS l

VALVE RUPTURES, LOCA, 1.8 x 10-s 4.1 x 10-s CONTAINMENT BYPASS, MELT I

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a

f SOURCE TERMS FOR SEABROOK STATION ALL ACCIDENT SEQUENCES IDENTIFIED IN SEABROOK PRA v ,

13 RELEASE CATEGORIES TO COVER ALL SEQUENCES IN SSPSA v

6 RELEASE CATEGORIES WITH SIGNIFICANT RISK CONTRIBUTION IN SSPSA

( -

v 12 SOURCE TERMS FOR RMEPS

- 6 BEST ESTIMATE

- 6 CONSERVATIVE ESTIMATE 2 SOURCE TERMS SOURCE TERMS DEVELOPED FROM FROM IDCOR (MAAP) SSPSA USING WASH-1400 (CORRAL) l METHODOLOGY AND CORRECTIONS FOR l IDCOR AND NRC SOURCE TERM v RESEARCH

)

2 SOURCE TERMS '

I CALCULATED FOR SEABROOK

( USING IDCOR (MAAP) l METHODOLOGY

n .

SEABROOK STATION SOURCE TERMS SOURCE TERM DEVELOPMENT

{ CONTAINMENT l ACCIDENT RELEASE CATEGORY FAILURE TYPE RMEPS SENSITIVITY A B C SSPSA BEST ESTIMATE CONSERVATIVE STUDY i S1 EARLY GROSS FAILURE VMT RSS MODIFIED RSS RSS METHODOLOGY SSPSA METHODOLOGY METHODOLGY T

S2 EARLY INCREASED LEAKAGE VMT (LOP) LOP RS$ SSPSA SSPSA RSS l WITH LATE OVERPRESSURE METHODOLOGY METHODOLOGY

+ UNCERTAINTY S3 LATE OVERPRESSURE OR LATE (LOP) LOP RSS IDCOR SSPSA RSS BASEMAT MELT-THROUGH METHODOLOGY (ZION-MAAP) '

METHODOLOGY

+ UNCERTAINTY SS CONTAINMENT INTEGRITY - - - RSS RSS RSS RSS MAINTAINED METHODOLOGY METHODOLOGY METHODOLOGY METHODOLOGY

+ ENCLOSURE NO ENCLOSURE s

I S6 CONTAINMENT PURGE - At - RS$ IDCOR SSPSA RSS ISOLATION FAILURE t=0 METHODOLOGY (ZION-MAAP) METHODOLOGY

+ UNCERTAINTY S7 CONTAINMENT BYPASS BYP - -

ASSIGNED SEABROOK-MAAP SEABROOK-MAAP RSS VIA RHR PUMP SEALS RNEPS TO 86 + POOL NO POOL METHODOLOGY

^

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A i

SUMMARY

l I

e EARLY HEALTH RISK VERY LOW EVEN WITHOUT IMMEDIATE PROTECTIVE ACTIONS ,

! e BENEFITS OF EVACUATION VERY SMALL AND CONFINED i l TO AREA CLOSE IN TO SITE I

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) e PEER REVIEW GROUP CONCURS WITH RMEPS AND SENSITIVITY STUDY CONCLUSIONS 1

l

e SEABROOK RISK MANAGEMENT ACTIVITIES ARE l CONTINUING I

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SEAna.PW. STul1';

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October 1, 1986 Public Service of New Hampshire CE- 23 NEW HX1PSilIRE *iANKEE DIVISION No Response Required Brookhaven National Laboratories Mr. Charles lloffmayer 32 Lewis Street Building 129 , ,

Upton, NY 11973

Dear Sir:

Seabrook PSA Studv - Piping Isometrics In.accordance with your request at our September 23rd meeting, we have en' closed isometric piping drawing for lines 4 inch and larger which penetrate primary containment. These isometrics show piping runs on both sides of the containment penetrations. If additional information is required to support your review, please feel free to call.

Very truly yours

~

R. ;. Whi _

Mechanical / Structural Supervisor REW/ daw ,

I Attachments -

l' cc: lG. Bachi w/o attachment J. E. ft1rtin w/o attachment R. Basu w/o att achment G. F. Mcdonald w/o attachment B. B. Beckley w/o attachnent D. G. McLai n w/o at tachment [

W. J. Daley, Jr. w/o attachment D. C. Mercer w/o attachment R. J. DeLoach w/o attachnent D. E. *!ondy w/o attachment ,,

E. W. Desmarais w/o attachment J. H. Moody, Jr. w/o attachment ;

DeVincent is w/o attachment w/o at tach .ent J. J. W. Stacey R. A. Garramore w/o attachment G. S. Th,r.u w/o attachment B. J. Iluselton w/o attachment J. M. V. argas w/o attachnent D. A. !! aid rand w/o attachment 1 --

bcc: B. Dec urti ( rMc) fcJ4-87-7 ,

5. L o J q ( M a c.) g fyf

\l. tJBL5 5(au)

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1. SEISMIC FRAGILITY UPDATE Seismic sequences dominate release categories S2 and S6 in the Risk Management and Emergency Planning Study (RMEPS). A separate submittal on the principal contribution to early healtn risk explains how these release categories contribute to early health risk. This write-up explains how the seismic fragility update is expected to change tne frequency of S2 and S6. A complete requantification will be included in the probabilistic safety assessment (PSA) update now in progress and i planned for completion in 1987.

In a complete seismic risk analysis, there is first performed a point estimate analysis using the plant event trees that are quantified for several discrete values of ground acceleration. From tne point estimate ,

results, dominant sequences initiated by seismic events are identified; then, these sequences are reanalyzed using a special computer code called SEIS4 In SEIS4, the seismicity curves and fragility curves are convoluted and uncertainties in these curves are propagated to obtain uncertainty distributions on the final result, whicn is eitner a core

-melt or plant damage state frequency contribution. In the following approximate analysis, the point estimate step is bypassed, so some assumptions are made atout dominant sequences. Hence, these results are only rough ape oximations and should only be used for order-of-magnitude estimates. A complete reanalysis of seismic events is currently in progress and is planned for completion in 1987.

1.1 RELEASE CATEGORY S2 This release category is dominated by earthquake and transient initiating events. These sequences can be simply represented as OG*(OT + DG + SSPS)

(1) where OG = Offsite Power Fragility

. UT = Diesel Generator Day Tank Fragility OG = Diesel Generator Fragility SSPS = Solid State Protection System (SSPS) Fragility (actually 120V AC power panel required for SSPS success) and only seismic unavailabilities are included.

Also, earthquake and large loss of coolant accident (LOCA) initiating events provide a small contribution and can be represented as LL*0G*(UT + DG + SSPS) (2) where LL = Large LOCA Fragility 1

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______J

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Equation (1) was quantified with the SEIS4 computer code and resulted in the following annual core melt frequency:

Mean = 2.84 x 10-5 Variance = 2.24 x 10-9 Based on the fragility update, SSPS and DT can be dropped from the model, based on significantly higher capacities. However, a relay chatter fragility at a relatively lower capacity has been identified in the 4,160V switchgear. This chatter could have a negative effect; e.g., trip out the diesels. Until the consequences of this chatter are evaluated, it is assumed that the chatter fails both diesels. Therefore ,

Equation (1) can be changed as follows:

OG*(chatter + UG) (3) where Chatter = Relay Chatter Fragility (4,160V switcngear)

Quantifying equations for annual core melt frequency with SEIS4 results in '

Mean = 1.8 x 10-6 Variance = 9.58 x 10-10 Comparing the quantification of Equations (1) and (3) shows a slight reduction (less than a factor of 2) in frequency. However, this assumes the chatter fails the diesels without recovery. An ongoing relay chatter

- review will determine whether this particular chatter is a real concern.

In addition. this review will determine whether there are any other relay chatters that should be considered in the model.

1.2 < RELEASE CATEGORY S6 Ths release category is dominated by earthquake and transient initiating e nts. These sequences can be simply represented as j N0G*SSPS (4) l where N0G = Offsite Power Available (negation of UG - fragility)

As described above under release category S2, the solid state protection system can be dropped from tne model. Therefore, the simple model in Equation (4) would go to zero. To actually determine the new

$6 frequency, the whole plant model needs to be requantified and unraveled to obtain new dominant sequences and frequencies. However, the trend is a reduced frequency unless the ongoing relay chatter review identifies new sequences.

2

PRINCIPAL CONTRIBUTORS TO EARLY HEALTH HISK AT SEA 8R00K STATION The purpose of this note is to describe the principal contributors to early health risk at Seabrook Station, as determined in the original probabilistic safety assessment (PSA) in 1983 (PLG-0300), in the PSA updates of 1985 (PLG-0432), and in tne sensitivity study (PLG-0465). Tne risk measures of interest here are the early fatality risk curves and the frequency of exceedance of dose and distance curves for the whole body doses of 200 rem, 50 rem, 6 rem, and I rem.

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1. SSPSA RESULTS (1983) 1.1 EARLY FATALITY RISK CURVES The only results available for-early health risk in the Seabrook Station Probabilistic Safety Assessment (SSPSA) (PLG-0300) assume a lu-mile

, evacuation zone. The contributors to risk can be expressed in a number 1

of different ways; e.g., by an accident sequence or by accident sequences grouped in different ways. Alternative ways to group accident sequences are to group the individual sequences by initiating event, by. plant damage state, and by release category. A graphic display of how sequences grouped by release category contribute to the mean risk of early fatalities in the original PSA is shown in Figure 1, whicn is taken from Figure 13.2-la in PLG-0300. As seen in this figure, release ,

, category S6 (large isolation failure or bypass) dominates tne entire curve, S2 (small isolation failure) makes a small contribution, and all other categories make negligible contributions that are at frequency levels below 10-9 per reactor year. Note that the mean risk curve,

. whose contributions are being discussed here, is the mean of a family of i surves that characterize uncertainty in the risk estimate. This family is shown in Figure 2, which is reproduced from Figure 13.1-ba in i.

PLG-0300. The fact that the mean curve falls well outside the median

(.50) risk curve indicates large uncertainties. These uncertainties are due to uncertainties in estimating the accident frequencies, source terms, and site model parameters.

, A tabular representation of the information in Figure 1 is provided in l- Table 1, which is adopted from Table 13.2-7a in PLG-0300. This table shows that more than 99% of the mean risk curve comes from release category S6 Most of the remaining contribution comes from S2. Only in

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the extreme right hand tail, at frequencies below 10-9 per reactor year, does another category appear, S1 (early containment rupture due to

, steam explosion, early overpressure, or external missile).

l j The next step in breaking down the SSPSA risk contributors is to examine the contribution of sequences grouped by initiating event. Because nearly all of the early health risk comes from S6 and a small coptribution factor, S2, it is more efficient to confine our search to -

these release categories. The initiating events that make significant contributions to S6 and S2 are provided in the table below, which was adapted from Table 13.2-Sa in PLG-0300. As can be seen, release category 56, which indicates early fatality risk, is, in turn, dominated by the interfacing system loss of coolant accident (LOCA) (V-sequence).

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Percent Contribution Initiating Event 56 S2 V, Interfacing LOCA 76. 0 ET, Seismic Transient 22. 95 EL, Seismic LOCA 2. 5 Others < 1. <1 Total 100. 100.

In a similar fashion, accident sequences can be grouped with respect to plant damage states (sometimes referred to as a bin). The following is the plant damage state breakdown of the S6 and S2 release categories.

Plant Percent Contribution -

, Damage State

S6 S2 IF 78 0 IFP 0 5 4

3F 21 'O ,

3FP 0 35 7F 1 0 7FP 0 60 Others <1 <1

Total 100 100 i

In 9mparing the previous two tables, note that 100% of the interfacing

' systems LOCAs were modeled as IF states. Hence, of the 24% of the seismic contribution to S6, 21% terminated in plant damage state 3F, 2%

! in IF, and 1% in 7F. Hence, most of the overall risk contribution comes from the interfacing system LOCA initiator and plant damage state IF.

Essentially all the remainder are seismically initiated sequences ending in plant damage state 3F.

See Table 1-2 in the Seabrook Station Risk Management and Emergency Planning Study (RMEPS) _(PLG-0432) for definitions. Numbers denote containment and reactor coolant system conditions at time of reactor vessel melt-through; letters denote status of containment systems and leak paths. F states are isolation failures or bypasses more than 3 inches in diameter; FP states are isolation failures less than 3 inches in diameter.

e

The final step in breaking down the early health risk is to examine specific accident sequences. In the SSPSA, an accident sequence is a single path that can be traced through the plant event trees from the point of entry (tne initiating event) to the point of termination (tne plant damage state). As with other PRAs, the interfacing system LOCA was analyzed as a single sequence. That is, the event was analyzed as an initiating event and assigned directly to the most severe plant damage state considered in the study and denoted as IF. This reflects the conservative assumption that multiple failures of the interfacing valves automatically result in a core melt and early large containment bypass.

All other initiating events were modeled tnrough the plant event trees, which include more than 4.5 billion sequences counting all the initiating events, tne plant damage states, and paths connecting them througn the plant event trees. Therefore, the above contributions from the .

V-sequence are from a sin come from many sequences.gle sequence, whereas the seismic contributions The nature of the specific sequences initiated by seismic events is next described. Of the seismically induced transients that make up 22% of the frequency of release category S6, the single sequence having the greatest frequency makes up only about one-fourth of this contribution and was analyzed as follows.

Event Frequency Earthquake Occurs (.3g) 1.1 x lod / yea r Offsite Power Does Not Fail .3S

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Solid State Protection Fails .041 Charging Pumps fail .88 Containment is Initially in Purge Mode .10

' Emergency Core Cooling System, Containment Isolation, and Containment Sprays Fail [ dependent failures resulting from loss of solid state protection system (SSPS)] 1.0 Other Equipment Does Not Fail .86 Total 1.2 x 10-7/ yea r The remaining three-fourths of the seismic sequences in S6 are made up from a large number of sequences, some involving station blackout and others involving failures of other equipment.

In a similar fashion, the seismic contributors to release category S2 are also spread over many sequences. The single most frequent sequence in this category is a seismically induced loss of offsite power and failure of both diesels due to either seismic causes or independent causes. Inis sequence appears several times in the scenario identification tables (in Section 13.2 of PLG-0300) once for each discrete range of ground acceleration. The total frequency of this sequence summed over all values of ground acceleration is 6.9 x 10-6 per reactor year, or about 40% of the total release category frequency.

In summary, the early health risk curves in the SSPSA, which were only performed for a 10-mile evacuation zone, were dominated by the interfacing LOCA sequence (about 76% contribution to the mean exceedance frequency in the risk curves). Most of the remaining contributions come from seismically induced transient events with failure to isolate tne ,

containment either through the purge lines (the Sb sequences) or through the reactor coolant pump seal return line (the 52 sequences).

1.2 DOSE VERSUS DISTANCE CURVES At the time this note was prepared, the work to reconstruct the dose versus distance curves that correspond with the SSPSA results for source terms and accident frequencies had not been completed. it is anticipated that, when these curves are completed, they will show dominance of release category S6, a significant contribution from S2, and very little, if any, appreciable contribution from any other release category. As noted earlier, S6 is dominated by the interfacing systems LOCA (761) and, to a lesser extent, by seismically induced transients with a failed open purge line. When the SSPSA dose versus distance curves are completed, they will be provided under separate cover.

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2. PSA UPDATE RESULTS (RISK MANAGEMENT AND EMERGENCY PLANNING STUDY, 1985) 2.1 EARLY FATALITY RISK CURVES In the RMEPS update of the Seabrook PSA, the following changes were made that had an impact on the risk levels and the ordering of the risk contributors.

e Plant Model Changes Item 1. The single sequence interfacing LOCA model was replaced by a two-event tree model, one for suction side and one for injection side residual heat removal / reactor coolant system ,

(RHR/RCS) interfacing valve ruptures. This led to a reduction in the frequency of plant damage state IF and the addition of three new plant damage states (IFV, IFPV, and 7FPV).

Two plant damage states (1FPV and 7FPV) were added to model new

,, scenarios with a submerged RHR pump seal bypass. This, in turn, led to the introduction of a new release category, 57, which takes credit for decontamination and scrubbing in the source term determination. Plant damage state IFV contains interfacing LOCAs resulting from unsubmerged piping failures.

Item 2. A conservatism in the treatment of certain seismically initiated seque1ces in release category S6 (plant damage states IF, 3F, and 7F) was eliminated. In the updated results, credit was taken for loss of instrument air to the air-operated valves (A0Vs) in the purga lines on loss of offsite power; hence,

_ a high probability of purge isolation valve closures in these instances. This resulted in a shift in some of the frequency of release category S6 to S2 because, when the large purge valves are assumed to close, there remain small open. lines with motor-operated valves that fail in these same sequences. There i still remain some seismic sequences in S6 with the purge isolation failure. Those that remain either involve a no loss of f offsite power condition or mechanical failure of the purge valves.

Item 3. In support of the effort to optimize plant technical i

specifications (PLG-0431), the PSA systems modeled were revised i

to incorporate revisions to the technical specifications and a more complete treatment of common cause failures. This led to many minor changes to individual sequence frequencies, with the most significant change being an increase to the unavailability of the primary component cooling system. This led to a slight increase {ncoremeltfrequencyfrom2.3x10-4 to j 2.7 x 10 , most of which occurred in plant damage state 80.

2 4 -

Item 4 The updated results take credit for recovery of certain j containment systems (principally the containment building spray) i during core melt scenarios initiated by loss of offsite power and 6

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involving a station blackout. Both the original and updated results took credit for recovery of electric power prior to and.

in prevention of core melt. This new recovery action results in a small shif t in frequency from release category S3 (gradual containment overpressure) to Sh (containment intact). This change does not appreciably affect the results of the RMEPS or sensitivity studies since neither S3 nor Sb contributes to early health risk with at least I mile of evacuation.

e Containment Model Changes --

Item 5 Uncertainties in source terms were reassessed for all release categories with the net effect of a reduction in the mean source terms for all categories.

- Item 6 A new release category and three new plant states for interfacing system LOCA scenarios were added.

Item 7. Interfacing system LOCAs resulting in unsubmerged RHR piping failures (plant state IFV) were reassigned from release category S6 to S1 (small conservative effect).

e Site Model Changes Item 8. Site model uncertainties were reassessed (minor effect).

Item 9. The evacuation distance and sheltering assumptions were varted.

Item 10. The Unit 2 construction workers were eliminated from the population distribution.

Although all the above changes contribute in some way to differences in the updated results, the ones that had the most significant impact on early health risk are items 1, 5, 6, and 9. A more qdantitative picture of the significance of eacn change is provided below.

The,results in the RMEPS update for early fatality risk are presented in Tables 2, 3, and 4 for evacuation cases of no evacuation,1-mile evacuation, and 2-mile evacuation, respectively. These are the

! comparison tables for Table 1 and the original SSPSA results. There are two kinds of differences exhibited in the new tables. One is that the risk levels (exceedance frequency values) are lower although less l evacuation is assumed, and, as expected, the levels decrease as the e

evacuation zone is increased from 0 to 2 miles. The other difference is that several new release categories, in addition to $6 and $2, appear as making significant contributions: 53, $7, and 51. Release category S3 contributes only under the assumption of no evacuation. This result is l viewed as purely academic because the time of release for 53 is some j 89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br /> after the initiating event, during which even ad hoc protective

actions would be effective.

i i For 1 or 2-mile evacuation, release categories 52, 56, S7, and 51 are l significant. The contribution of 52 only appears in the low consequence, i

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relatively high frequency portions of the risk curve. In comparison of these results with Table 1, the shift in the ranking of contributors is due to the following.

1. The frequency of 56 in the SSPSA is lower because of the deletion of the interfacing LOCA and some seismically initiated sequences with station blackout.

2. The frequency of S2 increased slightly from the same seismic sequence noted in 1.

3. Some of the old V-sequence frequency formerly categorized in Sb is now in S7. Wnile source terms in S7 are lower than S6, they are still great enough for potentially fatal doses.

4 The frequency of S1 increased due to the addition of the pipe break type V-sequences formerly categorized in S6 and to a smaller extent by a reassessment of some turbine missile scenarios that was done since the RMEPS.

The contributions of plant damage states and initiating events to all updated release categories are shown in Tables 5 and b, respectively.

Tables 7 and 8 define the codes used for initiating events. About 78% of the scenarios in category 51 are pipe break type interfacing LOCAs that are assigned to plant damage state IFV. The remaining scenarios in 51 include aircraft and turbine missile scenarios tnat fail the containment in plant damage states IFA, 2FA, and 6FA and a wide spectrum of transient and LOCA scenarios with containment failure due to reactor vessel steam explosions. The contributors to category 52 are the same as those in tne SSPSA; namely, seimically induced station blackout with a failed open small penetration. Release category S6 is now dominated by seismically

. Induced accident sequences with no loss'of offsite power and failure of the SSPS system with an assumed containment purge in progress. No credit ,

for operation recovery of any system or component is taken for any seismic sequence, including those that now dominate SZ and 56.

Release category S7 is composed wholly of new interfacing LOCA scenarios in which the RHR piping remains intact and the bypass occurs via a defadedandsubmergedRHRpumpseal. In assessing the uncertainty on the source term for 57, a 101 probability was assigned to tne possibility that the leak path would not be submerged. From the information provided i in RMEPS, it. is clear that there would be no contribution to early health l risk from S7 if only best estimate (submerged) source terms had been used. Similarly, had the conservative source terms not been used for the remaining release categories, the risk levels calculated in RMEPS would l .

have been much lower than they were. In fact, on the basis of using the l best estimate source terms only, release category S1 is the only category i

that produced any potential for 200-rem doses and, hence, any potential j for early fatalities.

2.2 00SE VERSUS DISTANCE CURVES l In the RMEPS results, there was found to be very little potential for i 200-rem doses, even close to the site. As seen in Figure 2-9 of RMEPS i

l (3

(PLG-0432), the frequency of exceedance scale had to be extended from .001 to .0001 to pick up the mean risk of exceeding the 200-rem dose shown on the curve. The median curve for the 200-rem dose was off-scale. The contributions to the mean risk curve at various distances are indicated in the table below.

Percent Contribution to Release 200-Wem Exceedance Frequency Category (Figure 2-9 in PLG-0432) - - ---

1 mile 1.5 miles 2 miles 51 1 3 6 .

52 13 34 0 S3 67 3 0 55 0 0 0 S6 17 50 73 S7 2 9 20 By comparing these results with those in Table 2, it is seen that the 200-rem risk curve has the same set of release category contributors as the early fatality risk curve for no evacuation. Category S3 dominates at 1 mile. Again, this result is largely academic. It is difficult to envision, even if no emergency plans existed, that any individual would be in a position to receive a large dese more than 3 days after tne initiating event. At 2 miles, the 200-rem curve is dominated by S6, with smaller contributions by 57 and S1. Hence, the overall picture of the

, risk contributors is the same for the early fatality risk curves and the 200-rem dose versus distance curves.

e

/

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3. SENSIT!vlIY STUOY UPDATE in the Seabrook Station Emergency Planning Sensitivity Study (PLG-0465, 1986), there were no changes made to the plant model . Source terms were revised to reflect the Reactor Safety Study (WASH-1400) source term methodology; i.e., were calculated using the CORRAL computer program for the Seabrook plant configuration. These CORRAL source terms had been developed during the original SSPSA. CRACIT computer program runs were made using best estimate (medium) modeling assumptions, and median

. accident frequencies-were used for consistency with-NUREG-0396 and WASH-1400. However, unlike NUREG-0396 and WASH-1400, the full treatment of dependent and external events in the Seabrook results was left unchanged.

3.1 EARLY FATALITY RISK CURVES

The early fatality risk curves for 0, 1 and 2-mile evacuations are plotted in Figure 2-1 of PLG-0465. The contributors by release category are shown in the tables below for no evacuation,1-mile evacuation, and

'f-mile evacuation.

RESULTS FOR No EVACUATION Percent Contribution to Release Early Fatality Risk Curve Category 1 Fatality 100 Fatalities 1,000 Fatalities S1 <1 <1 <1 S2 100 99 99 56 <1 1 1 Others <1 <1 <1 Total 100 100 100

?

RESULTS FOR 1-MILE EVACUATION Percent Contribution to -

Release Early Fatality Risk Curve -

Category 1 Fatality 100 Fatalities 1,000 Fatalities S1 <1 <1 <4 S2 9 95 0 S6 1 5 96

! Others <1 <1 <1 -

Total 100 100 100 10 N

RESULTS FOR 2-MILE EVACUATIUN Percent Contribution to Release Early Fatality Risk Curve Category 1 Fatality 100 Fatalities 1,000 Fatalities S1 2 2

S2 0 0 ~*

56 98 96 8 Uthers <1 <1

Total 100 100 100 .

Results below 10-9 per reactor year not shown.

By comparing these results with the RMEPS results, it can be seen that one chief difference is that S2 now has a more dominating impact than it did in RMEPS, especially for the 0 and 1-mile evacuation cases.

Category S6 dominates the low frequency tail of the 1-mile curve and completely dominates the 2-mile results. The other chief difference is that categories S3 and 57 no longer make a significant contribution to early health risk and the percent of contribution of S1 is reduced somewhat. These differences stem from the fact that application of the WASH-1400 source term methodology did not have uniform impact on all tne source terms. The application of this methodology appears to have increased the S2 source term more than the others. In addition, the

~

RMEPS results for early health risk are heavily influenced by the conservative source terms used in that study. For category S7, the conservative RMEPS source term assumed no credit for a flooded RHR vault, while such credit was taken in the sensitivity study to make the analysis consistent with WASH-1400. In WASH-1400, credit was taken for suppression pool scrubbing in some boiling water reactor scenarios.

3.2/DOSEVERSUSDISTANCECURVES The 200-rem dose versus distance curve is fully dominated by release category S2, with very small contributions from S6 and 51.

(

i 1

I 11 w re m w n

s TABLE 1. CONTRIBUTIONS OF RELEASE CATEGORIES TO RISK 0F EARLY FATALITIES AS CALCULATED IN SSPSA Number of Early Fatalities (percent contribution of release category) 1 10 100 1,000 10,000 56 (98.98) S6 (98.8) S6 (99.4) S6 (99.4) S6 (99.5)

S2 (0.92) S2 (1.10) S2 (0.52) S2 (0.49) S1 (0.5)

Others (< .1) Others (< .2) Others (< .1) Others (< .1) Others (0)

Exceedan e 4.60-7 3.87-7 3.14-7 1.78-7 6.26-10 NOTE: Exponential notation is indicated in abbreviated form; i 1.e., 4.60-7 = 4.60 x 10-7, .

t 1A?OD1nn104

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, s._

TABLE 2. CONTRIBUTIONS OF RELEASE CATEGORIES TO RISK OF EARLY FATALITIES BASED ON RMEPS UPDATE - NO EVACUATION CASE (EO)

Number of Early Fatalities (percent contribution of release category) 1 10 100 1,000 10,000 S2 (40.1) S3 (62.7) S3 (50.0) S6 (67.6) S7 (98.5) s S3 (35.4) S6 (27.6) 56 (32.7) S7-(24.9) S1 (1.5)

S6 (18.3) S7 (6.7) S7 (13.3) 51 (7.5)

Others (< 7) Others (< 3) Others (< 4) Others (= 0.0) Others (= 0.0)

Updated Frequency of 1.40-7 7.91-8 2.98-8 4.41-9 2.53-11 Exceedance SP5A 6.26-10 4.60-7 3.87-7 3.14-7 1.78-7 Res9]

NOTE: Exponential notation is indicated in abbreviated form; i.e., 1.40-7 = 1.40 x 10-7, 1429P092986

s s

TABLE 3.

CONTRIBUTIONS OF RELEASE CATEGORIES TO RISK OF EARLY FATALITIES BASED ON RMEPS UPDATE MILE EVACUATION ZONE (E Number of Early Fatalities (percent contribution of release category) 1 10 100 1,000 10,000 S2 (71.7) S6 (65.5) S6 (50.7)

N S7 (62.4) S7 (98.5)

S6 (18.6) S7 (26.4) S7 (41.2) S6 (28.3) S1 (1.5)

S7 (7.4) S1 (8.1) S1 (8.1) S1 (9.3)

Others (< 3) Others (= 0.0) Others (= 0.0) Others (= 0) Others (= 0)

Updated Frequency of 7.44-8 1.64-8 Exceedance 8.25-9 1.59-9 2.53-11 Resu ts 4.60-7 3.87-7 3.14-7 1.78-7 6.26h10 NOTE:

Exponential notation is indicated in abbreviated form; i.e., 7.44-8 = 7.44 x 10-8 i 1429P092986 i

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TABLE 4. CONTRIBUTIONS OF RELEASE CATEGORIES TO RISK 0F EARLY FATALITIES BASED ON RMEPS UPDATE MILE EVACUATION ZONE (E2)

Number of Early Fatalities i

(percent contribution of release category) 1

1 10 100 1,000 10,000 S2 (46.0) S7 (55.2) S7 (65.3) S6 (72.9) lU S7 (43.3) 56 (33.3) S6 (23.9) S7 (19.1)

S1(10.7) 51 (11.5) S1 (10.8) 51 (8.0)

! Others (= 0) Others (= 0) Others (= 0) 'Others (= 0)

Updated Frequency of 4.96-9 3.07-9 1.17-9 2.28-10 0.0 Exceedance j Resu ts 4.60-7 3.87-7 3.14-7 1.78-7 6.26-10 NOTE: Exponential notation is indicated in abbreviated form; i.e., 4.96-9 = 4.96 x 10-9 1429P100186

s TABLE 5. CONTRIBUTIONS OF PLANT DAMAGE STATES TO RELEASE CATEGORIES MAKING MAJOR CONTRIBUTIONS TO RISK OF CORE MELT FREQUENCY -

RMEPS UPDATE RESULTS Major Risk and Core Melt Frequency Contributing Release Categories (percent contribution of plant damage states to release categories) 51 S2 S3 SS 56 S7

, IFV (77.5) 7FP (49.5) 80 (72.0) 8A (82.1) 3F (92.5) IFPV (69.5)

.A IFA (7.3) 3FP (43.8) FD (14.6) 4A (15.0) 7F (7.3) 7FPV (31.5) 8A (5.0) IFP (6.6) 3D (11.3) 2A (1.6) 2FA (3.7) 6FA (1.3)

Others (5.2) Others (<'1) Others (< 3) Others (< 2) Others (< 1) Others (0.0)

Release Category 6.00-9 2.02-5 1.43-4 1.17-4 3.00-7 3.93-8 Frequency NOTE: Exponential notation is indicated in abbreviated form; i.e., 6.00-9 = 6.00 x 10-9 0

l 1 , ,

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.r TABLE 6. CONTRIBUTIONS OF ACCIDENT SEQUENCES GROUPED BY INITIATING EVENT TO '

PLANT DAMAGE STATES WITH MAJOR RISK OR CORE MELT FREQUENCY CONTRIBUTIONS - RMEPS UPDATE RESULTS Plant Oama9e States (percent contributton by Inf tfating event)

IFV IFPV IFA 2FA 7FPV 30 3F NP 44 70 FFP 84 80 vt (58.11 V5 (1001 init (86.0) TMLt (86.0) v5 (76.31 AtonF (49.11 f.34 (56.31 E.7T (55.3) Insiv (23.7) 105P (39.2) t.77 (31.4) 105P (53.01 SI (21.0)

V5 (41.9) APC (14.01 APC (14.0) VI (23.7) 1057 (14.0) E.7A (16.0) 1.5T (14.5) 110C (23.1) EITLP (79.5) E.4I (10.3) SLOCA (18.41 TLPCC (17.0)

E.3A (8.9) E.4A (0.9) El.01 (10.3) LOSP (12.1) (ITaC (11.5) t.57 (17.61 St (7.11 PtMFw (16.8)

E.77 (6.0) E.54 (6.1) E.47 (6.9) ATT (11.3) 5GTR (4.1) E.37 (14.3) St0i (6.8) Ti (13.2).

f.7A (3.Il fl.04 (4.58 f.FA (5.2) AttMF (7.1) itnFw (4.1) E.27 (12.9) TT (4.2) tyFW (9.25 El.0A (3.5) E.47 (3.7) EXTCt (2.3) ENFW (2.9) TL5u (6.0) t.57 (3.0) PLMFW (2.0) LOSP (4.0) 6 E.n (2.Si (OPr o.n t.3T (2.6) SLOC A (3.71 StoCA (2.0)

Others < 20 Others < 9 Others < $ Others e 8 Others < 10 Others < 6 Others < 5 others < 5 Plaxt t 4.65-9 2.73-0 4.36-10 2.21-10 1.20-0 1.70-5 7.77-7 8.84-6 Frequency 1.77-5 2.20-5 1.00-5 9.66-5 ?.03-4 (events per ywr) norts:

1. littlatlag event codes are def fned in Tables 2 and 8.

2. f rpoaential metation is in4fcated in abbreviated form; f.e., 4.65-9

4.65 a 10*'.

e 14N09M46

i l

l TABLE 7. BINNING OF INITIATING EVENTS THAT HAVE IDENTICAL IMPACTS New Initiating Events Binned SSPSA Initiating Events Title Frequency (events / year) Title Frequency (events / year) -

EXTAC 2.70-6 FSRAC 5.19-7 FCRAC 2.10-6 FL2SG 8.50-8 i EXTLP 1.20-3 FTBLP 6.00-4 FLLP 3.20-4 TCTL 2.76-4 EXTCR 5.43-7 TMCR 3.98-7 NCR 5.80-9 ACR 1.39-7 TLPCC 1.82-5 LPCC 1.39-6 FSRCC 3.60-6 FCRCC 9.00-6 i

FPCC 4.20-6 TMPCC 1.27-8 MPCC 5.46-9 TLSW 6.22-6 LOSW 2.52-6 FCRSW 2.10-6 FLSW 1.60-6 TLbV 4.18-1 LCV 4.18-1 TMLCV 8.30-5 NOTE: Exponential notation is indicated in abbreviated form; i.e., 2.70-6 = 2.70 x 10-0 m

,. TABLE 8. INITIATING EVENT CATEGORIES SELECTED FOR QUANTIFICATION OF THE SEABROOK STATION RISK MODEL FOR THE SSPSA Sheet 1 of 2 Group Initiating Event Categories Selected Code for Separate Quantification Designator e Loss of Coolant 1. Excessive LOCA ELOCA Inventory 2. Large LOCA LLOCA

3. Medium LOCA

4. Small LOCA MLOCA(a)

SLOCA

5. Interfacing Systems LUCA V j 6. Steam Generator Tube Rupture SGTR(a ) ,

o General 7. Reactor Trip RT Transients 8. Turbine Trip TT(b)

9. Total loss of Main Feedwater TLMFW(c)

10. Partial Loss of Main Feedwater PLMFy(c)

11. Excessive Feedwater Flow EXFW(b)

12. Loss of Condenser Vacuum LCV(b)

13. Closure of One Main Steam IMSIV(b)

Isolation Valve (MSIV)

14. Closure of All MSIVs AMSIV

15. Core Power Excursion CPEXQ

! 16. Loss of Primary Flow LOPF(b )

17. Steam Line Break Inside Containment SLBI

18. Steam Line Break Outside Containment SLB0

19. Main Steam Relief Valve Opening MSRV

20. Inadvertent Safety Injection SI l e Common Cause Initiating Events t

- Support 21. Loss of Offsite Power LOSP(d) l System Faults 22. Loss of One DC Bus L1DC e

23. Total Loss of Service Water LOSW

24. Total Loss of Component Cooling LPCC Water

- Seismic 25. 0.79 Seismic LOCA E.7L l Events 26. 1.0g Seismic LOCA i

27. 0.29 Seismic Loss of Offsite Power E1.0L E.2Tle )

' 28. 0.3g Seismic Loss of Offsite Power E.3T(8)

29. 0.4g Seismic Loss of Offsite Power E.4T(e)

a. Transient without scram scenarios are represented by a separate code, ASLOC.

b. Transient without scram scenarios are represented by a separate code, ATT.

c. Transient without scram scenarios are represented by a separate code, ALOMF.

d. Transient without scram scenarios are represented by a separate code, ALOSP.

e. Transient without scram scenarios are represented by a separate code, ExA, x = .2, .3, .4, .5, .7, 1.0.

w

TABLE 8 (continued)

Sheet 2 of 2 Group Initiating Event Categories Selected Code for Separate Quantification Designator 30 0.59 Seismic Loss of Offsite Power E.5T(e)

31. 0.79 Seismic Loss of Offsite Power E.7T(e)

32. 1.0g Seismic Loss of Offsite Power El.0T(e)

Fires 33 Cable Spreading Room - PCC Loss FSRCC 34 Cable Spreading Room - AC Power Loss FSRAC (

35 Control Room - PCC Loss FCRCC

36. Control Room - Service Water Loss FCRSW 37 Control Room - AC Power Loss FCRAC

38. Electrical Tunnel 1 FET1

39. Electrical Tunnel 3 FET3 407 PCC Area FPCC

41. Turbine Building - Loss of Offsite ~

Power FTBLP

- Turbine 42. Steam Line Break TMSLB Missile 43. Large LOCA TMLL 44 Loss of Condenser Vacuum TMLCV

45. Control Room Impact TMCR

46. Condensate Storage Tank Impact TMCST

47. Loss of PCC TMPCC

- Tornado 48. Loss of Offsite Power and Une MELF Missile Diesel Generator

49. Loss of PCC MPCC

50. Control Room Impact MCR

- Aircraft 51. Containment Impact APC Crash 52. Control Room Impact ACR

53. Primary Auxiliary Building impact APAB

- F)ooding 54 Loss of Offsite Power FLLP

/ 55. Loss of Offsite Power and One Switchgear Room FL1SG

56. Loss of Offsite Power and Two Switchgear Rooms FL2SG

57. Loss of Offsite Power and Service Water Pumps FLSW

- Others 58. Truck Crash into Transmission Lines TCTL

a. Transient without scram scenarios are represented by a separate code, ASLUC.

b. Transient without scram scenarios are represented by a separate code, ATT.

c. Transient without scram scenarios are represented by a separate code, ALOMF,

d. Transient without scram, scenarios are represented by a separate code, ALOSP.

e. Transient without scr'am scenarios are represented by a separate code, ExA, x = .2, .3, .4, .5, .7, 1.0.

1429P092986 __ _ ____ - - - - - - - - - - - - - - - - - - - - - - - - - - -

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' FIGURE 1.

CONTRIBUTION OF RELEASE CATEGORIES TO RISK OF EARLY FATALITIES (MEAN VALUES) AS CALCULATED IN SSPSA (1983)

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l FIGURE 2. RISK OF EARLY FATALITIES WITH UNCERTAINTIES FROMSSPSA(PLG-0300,1983) l l

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PicKARDe LOWE AND GARRICK, INC.

2260 UNIVERSITY ORIVE

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October 9, 1986 PSN H-1016-PLG- 53 Mr. David Maidrand New Hampshire Yankee Division Public Service of New Hampshire P.O. Box 300 Seabrook, NH 03874

Dear Dave:

KEY FACTORS INFLUENCING UPDATED ANALYSIS OF INTERFACING LOCA As promised in my letter of October 1, enclosed is the package that addresses the changes made to the V-sequence analysis, " Key Factors Influencing Updated Analysis of Interfacing LOCA."

f Very truly yours dLI $ 70/

Karl N. Fleming "N IY l Enclosure J

feza - 8'7- 7 Elu ENGINilR$ '

APPLl(D $CllNIl$f $ M AN AO(M(NT CONSULT ANT $

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KEY FACTORS INFLUENCING UPDATED ANALYSIS Of INTERFACING LOCA The purpose of this note is to highlight the key factors that resulted in a major reduction in risk levels in the updated Seabrook probabilistic safety assessment (PSA) results [per the Risk Management and Emergency Planning Study (RMEPS), PLG-0432, 19853 in comparison with the Seabrook Station Probabilistic Safety Assessment (SSPSA) (PLG-0300,1983).

Qualitatively, the key differences fall into three main areas of the analysis: initiating event frequency, plant response to various types of interfacing loss of coolant accident (LOCA) scenarios, and operator actions to prevent core melt and isolate the bypass. Each of these areas is described in the following pages.

d

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1436P100986 _ _ _ . . - _ . _ _ _ . _ , . _ _ _ _ _ _ _ .-. - _ - _ - - - - - - .

1. INITIATING EVENT FRE0VENCY The initiating event frequency model in the SSPSA considered four residual heat removal (RHR) cold leg injection paths, each having two series check valves, and two RHR hot leg suction paths, each having two series motor-operated valves (M0V). The check valve model considered successive, independent ruptures of, first, the inboard and, second, the outboard check valve. The second failure was assumed to occur at the same rate as the first at any random time between the first failure and the next test (refueling). The MOV model included a similar sequence of ruptures (inboard, then outboard), as well as the possibility that the outboard valve is already open when the inboard valve fails. The updated model included all the above failure modes plus several more. For the check valve failures, it was conservatively assumed that the ,

outboard + inboard sequence could also occur and that it would occur at the same rate as the inboard + outboard sequence. In addition, the possibility of instantaneous failure of the second valve in a sequence at the time the first valve failed was also considered. Hence, the model used in the update is more complete. The net effect of those model enhancements is worth about a factor of 2 increase in the frequency of a given leak size.

A second difference between the two studies was the definition of the initiating event. In the SSPSA, the V-sequence initiator was defined as a major rupture leading to RHR overpressurization. In the RMEPS, any rutures with a leak flow exceeding 150 gpm (capacity of one charging pump) was considered an initiator. Such flows are not capable of overpressurizing the RHR system when the RHR relief valves operate properly.

A third difference in the initiating event frequency was in the treatment of check valve data. The SSPSA used check valve rupture data--actually zero failures in a large sample of component hours per population--taken from the Indian Point 2 and 3 PSAs. In RMEPS, a different approach was based on a frequency-magnitude correlation of nuclear grade RHR and reactor coolant system (RCS) check valve experience in U.S. pressurized water reactors (PWR). These data are documented in RMEPS and in a Aeparate submittal.

i l t To put the corresponding analyses on a common footing, the frequencies of l RHR overpressurization events can be compared as follows:

SSPSA: 1.8 x 10-6/ reactor year.

RMEPS Update: 7.1 x 10- / reactor year (leak > 1,800 gpm).

Thus, the net effect of the model differences (which have an increasing

' effect in the update) and the data treatment (which has a decreasing effect in the update) is a reduction in the frequency of valve ruptures leading to RHR pressurization by a factor of 2 to 3. Hence, if no other i changes would have been made to this analysis, the V-sequence risk contribution (and its early release frequency contribution) would have decreased by this same factor.

l 2 u n a ure m

2. PLANT RESPONSE The plant response to RHR interfacing valve ruptures in the SSPSA and in most previously published probabilistic risk assessments (PRA) on PWRs has been treated rather simply, according to the following assumptions, without consideration of their incremental probability.

e Valve ruptures produce a shock wave with peak dynamic pressures significantly greater than the RCS pressure that travels down the low pressure RHR piping.

e RHR piping ruptures outside the containment.

e RCS and refueling water storage tank (RWST) inventories leak outside ,

the containment via a piping break.

e Core melt occurs with unsubmerged bypass.

e No credit is taken for any operator actions.

'Therefore, the plant response to the V-sequence in the SSPSA was treated as a single sequence. It was assigned to plant damage state 1F, which, in turn, was assigned to release category S6. In the RMEPS update, a large number of alternative scenarios were identified to provide a more complete picture of plant response. The most important variables introduced in the update to consider alternative plant responses are the size of the valve ruptures that initiate the event, the response of the RHR relief valves inside the containment, the pressure capacity of RHR low pressure piping, the response of RHR pump seals to overpressure, and the configuration of the RHR pump vaults with regard to source term implications.

The major differences in plant response to RHR interfacing valve ruptures, as modeled in the SSPSA and the RMEPS update, are illustrated in Figure 1. This figure is a highly simplified version of the event sequence model that was developed and quantified in the RMEPS update.

The chief differences in the update in this regard are a lower frequency of ynsubmerged pipe rupture-type bypasses because of the high capacity of the RHR piping. A more likely outcome is a submerged bypass via the RHR

! pump seals, l 3 l

1435P100936 _

3. OPERATOR RESPONSE Because of a different treatment of hardware and plant response, the potential for operator actions to mitigate the effects of the interfacing valve ruptures was appropriately considered in the update. The two key k actions, which are illustrated in Figure 1, are those to prevent melt and I to isolate the bypass. If the RHR piping remains intact, there is a high chance, as assesed in RMEPS, that the operators would prevent core melt whether or not the bypass was isolated. The key is to diagnose the bypass at the RHR pump seal and to provide long-term makeup of coolant to the RWST. If this action is not successful, there is some chance that the operator can isolate the bypass, but only for the discharge check valve rupture case (VI).

The net effects of the major update factors are summarized in Table 1. -

4 1435P100986 _ _ _ _

~

TABLE 1. IMPACT OF KEY FACTORS IN UPDATED V-SEQUENCE ANALYSIS Factor SSPSA RMEPS Frequency of RHR 1.8 x 10-6 7.1 x 10-7

/ System Overpressurization Reac r Year Reac or Year Percent of Overpressurization Events that End with:

e No Core Melt 0 93 e Melt with No Bypass 0 ~1 e Melt with RHR Seal Submerged 0 ~5 Bypass e Melt with Unsubmerged RHR Pipe 100 ~1

' Rupture Bypass l

5

_ _ . . . 1436P100986

7 FACmG VALVE LEAK HIN 8MME ATE I '

[ '

LEAKAGE O CHARG G O SAFETY

.,, 5150 GPM CAPACITV 5 F8-0108e07 ASSESS 3 CSSPSA:

RMEPS: 7.7 X 10-4 e ve j

/

VALVE LEAKAGE YES m f RHR SYSTEM )

PRESSURE RISE ,

LOCA VI A PR T.

v "

51.800 GPM LIMtTE D IF SOTH SY SS NO 18 X 10-4/yr 3 CSSPSA RMEPS. 7.i s 10-7/yr j qg SSPSA: DeO NOT ASSESS T RMEPS: 7 0 X IO-She j RHR SYSTEM PRESSURIZATO4 SSPSA: DIO NOT ASSESS T (PE AKS AT 2.250 psee) RMEPS: e S X 10-78, j -

1F OPE R ATOR RHR PtP NG YES m ACTIONS VES m SUCCESSFUL REMAlfeS -

TO PREVENT TERMINATION O INTACT MELT peo NO SSPSA: 1.3 X 164r T

( RMEPS: 4 6 X isP9/yr j 1P CORE OPE RATOR CORE ME LT wtTH ACTIONS TO YES esELT wlTH

" SS OR $3 U885UeMERGEO ISOLATE No SyPASS SYPASS SYPASS NO S1 (SS IN SSPSA)

DIO NOT ASSESS T PS S 10-(SSPSA:

RMEPS: 3 7 X 10-8/yr j \

r CORE ME LT wtTH SUBMERGED RHR SEAL 8vPASS 57 FIGURE 1. SIMPLIFIED EVENT SEQUENCE DIAGRAM FOR V-SEQUENCE

1

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I I

STATUS OF BNL REVIEW 0F SEABROOK STATION EMERGENCY PLANNING SENSITIVITY STUDY DEPARTMENT OF NUCLEAR ENERGY BROOKHAVEN NATIONAL LABORATORY UPTON, NY 11973

(-

PRESENTED TO A JOINT MEETING OF THE ACRS SUBCOMMITTEES ON OCCUPATIONAL AND

, ENVIRONMENTAL PROTECTION SYSTEMS AND SEVERE (CLASS 9) ACCIDENTS SEPTEMBER 26, 1986 91g-87-7 f/3b'

( BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(Illl

i PRESENTATION OF MATERIAL RISK PERSPECTIVE W. T. PRATT CONTAINMENT STRUCTURAL INTEGRITY C. HOFMAYER REVIEW OF ACCIDENT SEQUENCE FREQUENCIES R. YOUNGBLOOD

(

t BROOKHAVEN Nail 0NAL LABORATORY l} g)l AS500ATED UNIVERSITIES, INC.(Illl

~

(

BACKGROUND

- SEABROOK STATION PROBABLISTIC SAFETY ASSESSMENT (SSPSA), 1983

" FRONT END" REVIEW: LLNL

'BACK END" REVIEW: BNL (NUREG/CR-4540)

- SEABROOK STATION RISK MANAGEMENT AND EMERGENCY PLANNING STUDY, PLG-0432, DECEMBER 1985

{

SEABROOK STATION EMERGENCY PLANNING SENSITIVITY STUDY PLG-0465, APRIL 1986 C

BROOKHAVEN NATIONAL LABORATORY l} l)l

('

A5500ATED UNIVERSITIES, INC.(llll -

_ , ~ . . . _ . - .,--,n,..,. __.-,_.,.__.,_n..-

t OBJECTIVES REVIEW 0F THE EMER6ENCY PLANNING SENSITIVITY STUDY FOR SEABROOK FOCUS IS COMPARISON OF SEABROOK SPECIFIC DOSE VERSUS DISTANCE CURVES WITH GENERIC CURVES FROM NUREG-0396 4

SEABROOK DOSE CURVES BASED ON:

SSPSA WITH UPDATES IN PLG-0432 WASH-1400 SOURCE TERMS 1,

BROOKHAVEN NATIONAL LABORATORY l}

A5500ATED UNIVERSITIES, INC.(1lll

(

BASIS OF DOSE VS DISTANCE CURVES DOSE VS DISTANCE CURVES GENERATED FOR EACH RELEASE CATEGORY PROBABILITY OF EXCEEDING GIVEN DOSE IS MULTIPLIED BY APPROPRIATE RELEASE CATEGORY FREQUENCY ALL PRODUCTS ARE SUMMED TO GIVE TOTAL

( PRORABILITY OF EXCEEDING GIVEN DOSE TOTAL PROBABILITY OF EXCEEDING GIVEN DOSE NORMALIZED TO TOTAL CORE MELT FREQUENCY

( BROOKHAVEN Nail 0NAL LABORATORY l} g)l l

A5500ATED UNIVERSITIES, INC.(1lll

(

TABLE 4-1. SOURCE TERM CATEGORIES Source Analyzed Analyzed Term Containment Failure Mode in the in This Category 55PSA Study 51 Early Containment Failure Yes Yes 52 Early Increased Containment Leakage Yes Yes 53 Late Overpressure Failure Yes Yes 54 8asemat Melt-through Yes No

55 Containment Intact Yes Yes 56 Containment Not Isolated Yes Yes

( $7 Containment 8ypassed (V-sequence) No Yes

8ased on the 55PSA results, basemat melt-through sequences were assigned' to category 53 in this ' study.

(Taken frcm PLG-0465, April 1986)

, BROOKHAVEN Nail 0NAL LABORATORY l3 l)l ASSOCIATED UNIVER$lilES, INC.(II11

i j ..

4 1

l 1 TAst.E 4-3. RELEASE CATEGORIES FOR vanannr STATICII BASED 011 MRSH-1400 50lmCE TEIBl fETII000 LOGY n .

t 1 selease selease Morela9 Emery selease Fractfess ,

him itse seratten itse selease "

" faaers) fesurs) inta/s) at e.g. 3-2 cs it sa se ta feemrs) i 2.5 e.5 1.s 11.9 e.9 7-3 .7 .5 .. 3 .es .e2 4-3 siv t

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e.5 e .e3 2.1-4 4.3-3 .ses 4.2-3 2.e-3 a.4-4 a.4-5

! s2v-1 4.s 2.s 5.s-3 3.4-3 5.2-4 saw-2 6.e 4.s 2.5 e .er 5.e-4 1.w .see .ess I

52v-3 19.e le.e 15.5 e .e23 1.6-3 2.3-3 .las .147 . ele .eII 1.9-3 24.0 e.5 e .123 2.3-3 7.9-3 .2e .19 . eft .81 5 2.5-3 t 18tR 4.4 1

24.s 2.e e 4.7-4 3.3-6 3.2-5 1.7-4 1.5-4 1.9-5 1.2-5 2.4-6 53u 6.e los 1.5 e 15 1.1-3 .18 .11 82 .304 4.1-3 4.1-4 I i SW-1 1.75 .19 063 .422 .est .sel 56u-2 2.75 4.s 2.5 e 42 2.9-3 .e7 18.5 15.5 0 .32 2.2-3 .el .13 .32 . ell .42e 3.e-3

$m-3 15.75 t l 48 .887 833 5.2-3 l

TetR I.75 23.5 1.5 e .9 6.2-3 .le .43 l

! 6-5 2-5 4-6 I s7W e.5 7.4 2.8 e .9 7-6 7-4 5-4 3-4 ,

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i FOR NO IMEDIATE PROTECTIVE ACTION (RUN NUMBER 453)

(Taken from PLG-0465, April 1986).

, BROOKHAVIN NAll0NAL LABORATORY lg g g l .

j A5500ATID UNIVERSITIES, INC.IlIII f

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i TA8LE 4-2. REVISED C-MAi?.!X FOR NEW SOURCE TERM CATEGORIES

. (-

4 Plant Source Term Category Damage State S1 52 53 SS S6 S7 (frequency) 1F 1.0 (2.0-8) ,

(2.0-8)

IFV 1.0 (4.6-9) ( 4.6-9 ) .

1FP 1.0

? (1.4-6) (1.4-6) .

1FPV 1.0 (2.7-8) ( 2.7-8 )

! 2A 3.4-5 1.4-4 1.0-2 0.99 (1.9-6) (6.5-11) (2.7-10) (1.9-8) (1.9-6) j 30/70 2.0-6 8.0-5 0.95 0.05 ,

i (3.8-5) (7.6-11) (3.0-9) (3.6-5) (1.9-6)

( 3F/7F 1.0

( 3.0-7 ) ( 3.0-7 )

3FP/7FP 1.0

( 1.9-5,) (1.9-5) 4A/8A 3.1-6 1.3-4 6.2-3 U.995 (1.1-4) (3.3-10) (1.4-8) (5.5-7) (1.1-4) l 7FPV 1.0 l (1.2-8) (1.2-8) 80 1.1-6 3.1-5 0.9999 -

! (1.0-4) ( 1.1-10 ), (3.2-9) (1.0-4) ,

i Total 5.2-9 2.0-5 1.4-4 1.1-4 3.2-7 3.9-8 Frequency l NOTES: :

1. Exponential notation is indicated in abbreviated form; i.e., 2.0-8 = 2.0 x 10-B. I

2. Numbers inside parentheses are unconditional frequencies (events per reactor ;

year) based on mean values. Numbers not inside parentheses are conditional

! I frequencies of source term categories, given the indicated plant damage state,

also based on mean values. Median values of source term categories are presented in Section 3.

l (Taken from PLG-0465, April 1986).

1

-....... m - - .__,____._--.___.-.----_m.-_. _ _ , _ . . . . .--,-_,-_.___m.

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NUREG-0396 [

E2 . . ----- THIS STUDY FOR .

SEABROOK STATION

$E - . _

y8 *

...... .......... RMEPS R ESU LTS FOR

@4 - SEABROOK STATION .

mD (200 REM CURVE OFF 0g SCALE) fw g 0.1 -- -

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' ' ' ' ' 'I ' ' ' ' ' 'I ' ' ' ' ' '

. I 10 100 1,000 DISTANCE (MILES) i, FIGURE 2-3. COMPARISON OF SEABROOK STATION RESULTS IN THIS STUDY AND RMEPS WITH NUREG-0396 - 200-REM AND 50-REM WHOLE BODY DOSE -

PLOTS FOR NO IMMEDIATE PROTECTIVE ACTIONS ,

(Taken from PLG-0465. April 1986)

I BROOKHAVEN Nail 0NAL LABORATORY l} g)l

ASSOCIATED UNIVERSITIES, INC.(Illl F e

. - - , , , - - - - , , , , , - - - - , , - . - , - . , - , - - - - , ----.,---..-.---,-.,,--..--.--.,---,,.-.------------n. .- , - - , . . . . . - - - - - -

( -

COMMENTS ON THE SEABROOK DOSE VS DISTANCE CURVES RELEASE CATEGORY S2 IS ONLY CONTRIBUTOR TO CONDITIONAL FREQUENCY OF EXCEEDING 200 REM DOSE RELEASE CATEGORY S2 DOMINATED BY SEISMICALY INITIATED EVENTS WHICH RESULT IN EARLY INCREASED CONTAINMENT LEAKAGE ALL ACCIDENTS THAT MIGHT RESULT IN RELEASE CATEGORIES -

(S1 AND S6) ARE LOW FREQUENCY (< 3x10-7)

(

THIS IMPLIES LOW FREQUENCY FOR .

INDUCED EARLY CONTAINMENT FAILURE BYPASS OF CONTAINMENT FUNCTION LOSS OF CONTAINMENT ISOLATION .

INTERFACING SYSTEM LOCA l

i BROOKHAVEN NATIONAL LABORATORY [3 g} l A5500ATED UNIVERSITIES, INC.(1 ElI

. 1 ,

t FOCUS OF REVIEW CONTAINMENT STRUCTURAL INTEGRITY l

MAINTAINING CONTAINMENT FUNCTION:

CONTAINMENT ISOLATION FAILURE INTERFACING SYSTEM LOCA

(

COMPLETENESS:

OTHER' ACCIDENT SEQUENCES THAT MIGHT LEAD TO EARLY LARGE FP RELEASE EXAMPLES:

INDUCED STEAM GENERATOR TUBE RUPTURE ACCIDENTS FROM SHUTDOWN WITH CONTAINMENT l DPEN l

BROOKHAVEN NATIONAL LABORATORY l)l)l l A5500ATED UNIVERSITIES, INC.(1 ElI

(

CONTAINMENT STRUCTURAL INTEGRITY C. HOFMAYER

(.

d

BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVER5mES, INC.(llll

STATUS BNL REVIEW 0F CONTAINMENT STRUCTURAL INTEGRITY t MEETING AT BNL ON AUGUST lt4,1986 0 OBTAINED:

- DESIGN DRAWINGS AND SPECIFICATIONS ,

- CONSTRUCTION AND DESIGN REPORTS

- UE&C DESIGN CALCULATIONS

- SMA OVERPRESSURIZATION CALCULATIONS

- UE&C CONTAINMENT MODEL DATA

- PENETRATION SLEEVE DRAWINGS

( - OTHER RELEVANT INFORMATION l

0 REVIEWED RELEVANT CONTAINMENT STRUCTURAL ANALYSES PERFORMED BY APPLICANT'S CONSULTANTS.

8 PARTICIPATED IN SITE TOUR ON SEPTEMBER 8 AND 9, 1986 8 PERFORMING AXISYMMETRIC FINITE ELEMENT ANALYSIS OF CONTAINMENT. -

0 MEETING WITH APPLICANT AT NRC ON SEPTEMBER 23, 1986 9 FOLLOWUP MEETING TO DISCUSS DETAILS OF STRUCTURAL ANALYSES SCHEDULED FOR MID OCTOBER.

I BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(I ElI

ISSUES RELATED TO CONTAINMENT STRUCTURAL INTEGRITY I BASIS FOR UNCERTAINTY FACTORS ASSIGNED TO VARIOUS PRESSURE CAPACITIES.

O MECHANISM FOR TRANSFERRING LOAD FROM PENETRATION SLEEVES TO WALL UNDER HIGH STRAIN CONDITIONS.

8 NONUNIFORM CRACK GROWTH MAY LEAD TO STRAINS ON LINER HIGHER .

THAN PREDICTED.

8 COMPATIBILITY OF STRAINS IN REBARS AND LINER PLATE.

4 CONSIDERATION OF COMBINED TENSION, SHEAR AND BENDING EFFECTS AT DISCONTINUITIES.

I NONAXISYMMETRIC CONDITIONS MAY LEAD TO HIGHER STRAINS.

(. 0 EVALUATION OF PUNCHING SHEAR AT FUEL TRANSFER BUILDING.

8 MATERIAL STRENGTHS USED IN ANALYSIS.

8 EXTENT TO WHICH PIPING PENETRATIONS HAVE BEEN ANALYZED.

O BASIS FOR LEAK AREA CALCULATIONS. 1 BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES,INC.(IElI '

~ '

~

e

(

REVIEW 0F ACCIDENT SEQUENCE FREQUENCIES R. YOUNGBLOOD I

1

( BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llll

l REVIEW OF ACCIDENT SEQUENCE FREQUENCIES A FOCUSED LOOK AT: .

INTERFACING SYSTEM LOCA

- CONTAINMENT BYPASS l

BROOKHAVEN NATIONAL LABORATORY l} g} l AS500ATED UNIVERSITIES, INC.(llll i

INTERFACING SYSTEMS LOCA ' l l

l PSNH SUBMITTAL IS ESPECIALLY THOROUGH IN KEY AREAS:

PROBABLE BREAK LOCATION PLANT RESPONSE OPERATOR RESPONSE

- WE WILL BE FOCUSING ON THESE AREAS, AND SPOT-CHECKING SYSTEMS TO TRY TO COME TO GRIPS WITH COMPLETENESS

- BNL IS CONCURRENTLY WORKING ON GENERIC ISSUE 105 ON INTERFACING. SYSTEMS LOCA l

l .

BROOKHAVEN NATIONAL LABORATORY l} l]l A5500ATED UNIVERSITIES, INC.(Illl l

l e =

I i CONTAINMENT BYPASS ISSUES ISOLATION CAPABILITY OF ALL VALVES COMMUNICATING WITH CONTAINMENT ATMOSPHERE, ESP. PURGE AND VENT VALVES:

SEAL DEGRADATION DUE TO ENVIRONMENT OPERABILITY SURVIVABILITY l

(

l BROOKHAVEN NATIONAL LABORATORY l} l A5500ATED UNIVERSITIES, INC.(1 ElI

I'

'S v

SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT i

Presentation to

(

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Washington, D.C.

September 26,1986 t

e l

pair- e 7-7 -

2/37

, , . . , , - - . .c-- - - - , - - - -

5 Agenda ,

e Project Overview e SALP e Low PowerTesting ,

'O Emergency Planning -

l e

s l l

t Project Overview e Unit 1 Complete e Unit 2 24.1% Complste -

i i

I

.=

(

SALP .

FUNCTIONAL AREA RATING (1/1/85-3/31/86)

Construction 1 Preoperational Testing 1 Fire Protection And Housekeeping 1 Operational Readiness 1 Emergency Preparedness 2 Assurance Of Quality 1

, Licensing 1 9

6 ee l

t

4 1

Low Power Testing

e No Fuel Load Open items With NRR Or Region 1 e ASLB (Low Power License)

- Open items O Equipment Qualification Time Duration O Detailed Control Room

( Design Review- SPDS O Emergency Action Levels

- Hearing Scheduled For 9/29 - 10/3 in Portsmouth, N.H.

i

- Petition 50.57 (c) To Load Fuel And Conduct Precriticality Testing

. Submitted 8/22/86 .

l e m

. l

l

. \

l Emergency Planning e Graded Exercise Held 3/86 (N.H. Only) e Risk Management And Emergency ,

Planning Study Submitted To NRC

- 7/21/86 ~

e Revision 2 Of N.H. Plan Submitted 9/86

, e Gov.Of Mass. Announced RefusalTo Submit Plans To FEMA- 9/20/86 e

e e us G

de g -~www+ -- ' - - - - - " - - w-- - - -,, y -- ,, .,, ,- --- , _

es l

=

i i

! SEABROOK STATION PROBABILISTIC SAFETY ASSESSMENT l

i i

by James H. Moody, Jr. - New Hampshire Yankee Karl N. Fleming - Pickard, Lowe and Garrick, Inc.

Presentation to ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Washington, D.C.

September 26,1986 i

l 1 '

l 9

- - --.- w

i

1 l

l MEETING AGENDA -

i APPROXIMATE TOPIC DURATION SPEAKER

(min)

I i

SSPSA(1983) OVERVIEW i

i - HISTORY, SCOPE & CONCLUSIONS 10 JIM MOODY i

- METHODOLOGY & RESULTS 10 KARL FLEMING PSA UPDATE

) - OVERVIEW & CONCLUSIONS 30 JIM MOODY i

i j d METHODOLOGY & RESULTS -

30 KARL FLEMING i

l i -

1 i

l . .. .

l 4

i .

i SSPSA (1983) OVERVIEW i

HISTORY, SCOPE & CONCLUSIONS i- -

l l

l o

PROBABILISTIC SAFETY ASSESSMENT

}

e FULL-SCOPE LEVEL 3 PSA i o PUBLISHED DECEMBER 1983 e PRINCIPAL CONTRIBUTORS t

- YAEC - UTILITY PROJECT MANAGEMENT AND REVIEW l - PICKARD, LOWE AND GARRICK, INC. - PRA CONSULTANT 1

e NRC REVIEW

- LAWRENCE LIVERMORE - PLANT MODEL - (RESPONDED

, MAY 1986)

- BROOKHAVEN - CONTAINMENT CAPABILITY NUREG/CR-4540 (FEBRUARY 1986) l

! SSPSA SCOPE AND COVERAGE OF s ACCIDENT SEQUENCES .

i

! e COMPREHENSIVE COVERAGE OF ACCIDENT SEQUENCES

- 58 DISTINCT INITIATING EVENT CATEGORIES 1

- 39 PLANT DAMAGE STATES (" BINS")

- 14 RELEASE CATEGORIES

- 16 MODULARIZED EVENT TREES e FULL TREATMENT OF DEPENDENT EVENTS l

- COMMON CAUSE FAILURES (SYSTEM LEVEL) l - EXTERNAL EVENTS l - INTERNAL PLANT HAZARDS

- EXPLICIT MODELING OF FUNCTIONAL DEPENDENCIES i e PLANT-SPECIFIC AND ENHANCED CONTAINMENT MODEL

- ASSESSMENT OF CONTAINMENT FAILURE MODES

! - QUANTIFICATION OF SOURCE TERM UNCERTAINTIES

- ENHANCED METHODOLOGY '

i e SITE-SPECIFIC CONSEQUENCE MODEL

- MULTIPUFF RELEASE TREATMENT

- ACTUAL SITE CHARACTERISTICS

- QUANTIFICATION OF UNCERTAINTY

I I

i COVERAGE OF INITIATING EVENTS IN SSPSA BY MAJOR GROUP i

15 l

i l COMMON CAUSE EVENTS (38)

, A i z 10 -

I i

E e 3 O = l-5 E '

I [j ; j Z '

l

$)

J$

'~

3 Si Nl 0 '

l dl E_.

LOSS TRAN- SUPPORT EARTH FIRES FLOODS AIRCRAFT TURBINE OF SIENTS SYSTEM OUAKES (9) (4) & TRUCK & TORNADO I

, COOLANT (14) FAULTS (8) CRASHES MISSILES te) (4) (4) (s)

O 9

l l .l 1

) CONCLUSIONS (1983) i EARLY HEALTH RISK.

j - NRC SAFETY GOAL MET WITH LARGE MARGINS

! - INTERFACING LOCA DOMINATES l LATENT HEALTH RISK

- N RC SAFETY GOAL MET WITH VERY LARGE MARGINS

- SUPPORT SYSTEM FAILURES DOMINATE i

l CORE MELT FREQUENCY LOW I

i CONTAINMENT EFFECTIVENESS i - PRIMARY CONTAINMENT VERY STRONG

- EARLY FAILURE UNLIKELY

- LONG TIME FOR OVERPRESSURE

, - - - -_ _ , , , . , , . --- - -._ , _a.______ pm _ ,. _, _ __

h e

j

SSPSA (1983) OVERVIEW i

l METHODOLOGY & RESULTS -

l l

e p

^

l SSPSA METHODOLOGY ENHANCEMENTS / DIFFERENCES O PLANT ANALYSIS -

MODULARIZED EVENT TREE SCENARIO MODEL ENHANCED TREATMENT OF DEPENDENT EVENTS INTEGRATED ASSESSMENT OF ALL TYPES OF INITIATORS ENHANCED TREATMENT OF DATA l

8 CONTAINMENT ANALYSIS -

CORRECT PROBABILISTIC INTERPRETATION OF EVENT TREES

- REALISTIC ASSESSMENT OF CONTAINMENT FAILURE MODES ASSESSMENT OF SOURCE TERM UNCERTAINTIES I O SITE ANALYSIS (CRACIT) -

MULTI - PUFF / VARIABLE PLUME TRAJECTORY ACTUAL SITE CHARACTERISITICS ASSESSMENT OF MODELING UNCERTAINTIES l

e RISK ASSEMBLY & -

MATRIX SYNTHESIS OF RISK RESULTS DECOMPOSITION -

MATRIX DECOMPOSITION OF RISK CONTRIBUTORS i

INDIVIDUAL SCENARIO RANKINGS CAUSE TABLE BREAKDOWN OF SYSTEM LEVEL CONTRIBUTORS I

4 9

9

.m, _/

SSPSA TREATMENT OF DEPENDENT EVENTS e FULL SCOPE COVERAGE OF INITIATORS

" INTERNAL" EVENTS (TRANSIENTS & LOCAS)

- INTERNAL PLANT HAZARDS (FIRES, FLOODS, MISSLES, etc.)

l " EXTERNAL" EVENTS (SEIMIC, AIRCRAFT CRASH, etc.)

e EXPLICIT MODELING OF FUNCTIONAL DEPENDENCIES j - DETAILED DEPENDENCY MATRICES

- EVENT TREE STRUCTURES & IMPACT VECTORS

- CAREFUL TREATMENT OF PLANT / CONTAINMENT DEPENDENCIES S ADDITION 'OF " SPATIAL INTERACTION" TASK

- LOCATION DEPENDENT FT MODEL FOR SCREENING

- Pl. ANT WALKDOWN / INSPECTION

- SYSTEMATIC EVALUATION OF ALL POTENTIAL FIRE & FLOOD LOCATIONS 1

e ' SYSTEM LEVEL COMMON CAUSE FAILURE ANALYSIS

- SYSTEMATIC REVIEW OF OPERATING EXPERIENCE

- DESIGN SPECIFIC BETA FACTORS

- PROBABILITIES ASSIGNED TO ALL ACTIVE REDUNDANT COMPONENTS ,

~ - ---- ~ ~

^

OVERVIEW OF SSPSA EVENT SEQUENCE MODEL STRUCTURE INITIATING EVENTS AUXILIARY SYSTEM EARLY SYSTEM RESPONSE LONG TERM SYSTEM RESPONSE TR ANSIENT EVENTS OTHER EVENTS EVENT TREE EVENT TREES EVENT TREES 2 RT FSRAC

^

LOCA TT FCRCC E.7L > I 0 LLt > LL2 O TLMFW FCRSW E l.0L I PLMFW FCRAC TMLL EXFW FETI LCV FET3 l APC l l 0 APC y > APCy ---->

1 MSiv FPCC A MSIV FTBLP I CPEXC TMLCV l MLOCA ' O > ML --->

LOPF TMCA g TOPLANT St TMCST STATES LOSP TMPCC l SLOCA l 0 0 SL LIDC MELF LOSW MPCC AUX LPCC MCR O O TRAN ---, LTI E.2 T ACR e E.3 T APAB ,

E.4 T FLLP l SLBI l 0 SLI W % LT2 I E.5T FLISG I E.7 T FL2SG StaO E l.0T F LSW TMSLB i i SLO - - -

FSRCC TCTL MSRV i

,, O ATWS W

---> SGTR y b l SGTR l >

- SGTR g SGTR 3 >

l l V l p l .

n i

SUMlUARY OF PRINCIPAL CONTRIBUTORS TO RISK IN TERMS OF ACCibENT SEQUENCE GROUPS AND l

INITIATING EVENTS FROM THE SSPSA Accident Containment Response - Group Group Fraction of 3equence Group Contributing Contribution Frequency Total Release

Initiating Events Percent (meanvalues) Frequency Group I Early Containment Failure 2.4 x 10-6 p,7 ,og Early llealth -

Interfacing LOCA 76 Reactor Year or Effects -

Seismic 24 Once in 410,000

TUlf Reactor Years Group II Delayed Containment Failure 1.7 x 10-4 per .73 I Latent Health -

Loss of Offsite Power 40 Reactor Year or j Effects, -

Transients 19 Once in 6,000 i

Fires 15 Reactor Years Seismic 15 -

l Others 11 l 100 l

, Group Ill Containment Intact l Ho llealth -

Transients 57 6.0 x 10-5 per .26 l Effects -

SLOCA 29 Reactor Year or i -

Others 14 Once in 17,000

TUiT Reactor Years Total 2.3 x 10-4 per 1.00 Reactor Year or Once in 4,300 Reactor Years

~ ~

I .

i 1

9 I

l l

1

! PSA UPDATE (1986)

OVERVIEW & CONCLUSIONS l ,

l t l

I i

I .

! PSA UPDATE ACTIVITIES e RISK BASIS FOR TECHNICAL SPECIFICATIONS

e CONTINUAL REASSESSMENT OF PUBLIC HEALTH RISK

- RMEPS i

- SENSITIV!TY STUDY

- SEISMIC CAPACITY UPDATE

) e ESTABLISHED RELIABILITY AND SAFETY ENGINEERING GROUP

- MAINTAIN CURRENT RISK MODEL

- PART OF CHANGE REVIEW PROCESS

- EVALUATE IMPACT OF REGULATORY CHANGES

- PLANT RELIABILITY RESPONSIBILITIES

+'

m-m

h-l l .

RMEPS OBJECTIVES l

e REEXAMINE TECHNICAL BASIS OF THE 10-MILE EPZ (NUREG-0396) ON A PLANT-SPECIFIC BASIS e DEVELOP AN ENHANCED PRA METHODOLOGY FOR ESTABLISHING A PLANT AND SITE-SPECIFIC EPZ i

e APPLY THIS METHODOLOGY TO SEABROOK STATION

- UPDATE SSPSA RISK MODEL (1983 - 1985)

- DETERMINE RISK IMPACT OF EMERGENCY PLAN OPTIONS e ADDRESS UNCERTAINTIES AND SENSITIVITIES e PROVIDE DOCUMENTATION AND PEER REVIEW t

e 9

,-( . + A

PRINCIPAL INVESTIGATORS KARL N. FLEMING PLG ALFRED TORRI PLG 4

l ROBERT J. LUTZ WESTINGHOUSE ROBERT E. HENRY FAI l

i R. KENNETH DEREMER PLG KEITH WOODARD PLG j

[' .

KEY RESULTS 1

e EARLY HEALTH RISK WITH NO IMMEDIATE PROTECTIVE ACTIONS

- LESS THAN WASH-1400 WITH 25-MILE EVACUATION

- MEETS NRC SAFETY GOAL WITH WIDE MARGIN

- CONFINED TO AREA CLOSE TO THE SITE o VERY SMALL RISK REDUCTION BY ANY EVACUATION e ALL NUREG-0396 DOSE VERSUS DISTANCE CRITERIA SATISFIED AT 1 MILE OR LESS

e LATENT HEALTH RISK INSENSITIVE TO ASSUMPTIONS REGARDING EVACUATION t

FAVORA.BLE RESULTS DUE TO 1

e CONTAINMENT EFFECTIVENESS i

e ENHANCED V-SEQUENCE MODEL e SOURCE TERMS 1

e l #

=- . e

!i .

! RESULTS AND CONCLUSIONS OF SENSITMTY l STUDY l

l e WASH-1400 EARLY HEALTH RISK APPROXIMATELY' MET i WITH 1-MILE EVACUATION i

l e "RCs PROPOSED INDIVIDUAL RISK SAFETY GOAL MET WITH NO IMMEDIATE PROTECTIVE ACTIONS e CONDITIONAL FREQUENCY OF EXCEEDING WHOLE-BODY DOSE VERSUS DISTANCE LOWER FOR ALL CASES SEABROOK NUREG-0396 STATION j 10 MILES 1 MILE 200 REM .03 .02 50 REM .12 .03 1 REM .30 .06 l e 1-MILE EPZ JUSTlFIED EVEN ASSUMING WASH-1400 i SOURCE TERM METHODOLOGY -

i 1

i j .- .

1 ..

i

) PEER REVIEW GROUP l .

e ROBERT BUDNITZ, CHAIRMAN, FUTURE RESOURCES

~

ASSOCIATES, INC.

e DAVID ALDRICH, SCIENCE APPLICATIONS INCORPORATED e JOSEPH HENDRIE, CONSULTANT e NORMAN RASMUSSEN, MASSACHUSETTS INSTITUTE OF t TECHNOLOGY ,

l .

e ROBERT RITZMAN, ELECTRIC POWER RESEARCH INSTITUTE l

i e WILLIAM STRATTON, CONSULTANT e RICHARD WILSON, HARVARD UNIVERSITY

+_ _ _ _ _ _

4 l .

PEER REVIEW FINDINGS t

.e CONCURRED WITH PRINCIPAL STUDY FINDINGS

- OVERALL OFFSITE RISKS VERY SMALL

- EARLY HEALTH RISK LOWER THAN THOUGHT TO EXIST WHEN GENERAL EPZ ESTABLISHED

- EARLY HEALTH RISK CONFINED TO AREAS VERY CLOSE TO REACTOR e CONCLUSION ROBUST EVEN IN LIGHT OF UNCERTAINTIES e BELIEVE THE "BEST ESTIMATE" PROBABLY OVER-ESTIMATES ACTUAL CONSEQUENCES e SEABROOK CONTAINMENT MAJOR FACTOR l

\ -

i i

i SSPSA UPDATE (1986)

METHODOLOGY & RESULTS i

l l

I I S

t

,- . - - - ,., . . . -- - . - n.

l BLOCK DIAGRAM STRUCTURE OF i

SEABROOK RISK MODEL

^3 f 3 (

3 1

1 .. . -

.i am i W Ia M NI

] vent saousacs/ +----- ~ susar unusncs .--- % sete coussousaces i' Ym)YS)mb ***

j . ,

l l sysisus immaan annamas i

moosts *- saac8anse ms'moosts

-* swears W contamassas ramuna <

accsoam sannanon [

, f

% sc.o.,mo ,e aaomiocacai oows a

amosas at asert gy ,c,y ", g ac ooet

.. ,.i.,oc..,,,,,.

sino os, .oc .,,,,,

nasconsa

"*"8

..,4 easg etaar ammout iIAfonts

(YM#

, eaaluna,s aaso 1 ; so e mass l

i I

i 1

k PLANT MODEL j ( CONTAINMENT MODEL i J

\

( SITE MODEL J

t LEGEND ORIGINAL SSPSA MODEL l

RMEPS UPDATED MODEL I

t *

.I e

UPDATE OF SSPSA RISK MODEL e UPDATED SSPSA PLANT MODEL -

- ENHANCED V-SEQUENCE MODEL

- ENHANCED SEISMIC ANALYSIS

- CONTAINMENT RECOVERY MODEL

- ENHANCED TREATMENT OF COMMON CAUSE FAILURES e UPDATED SSPSA SOURCE TERMS

- EXISTING SSPSA SOURCE TERMS

- INCORPORATED SOME ZION IDCOR RESULTS

- PERFORMED SEABROOK/ ZION DESIGN COMPARISON

- DEVELOPED SOME SEABROOK RESULTS WITH MAAP

- REASSESSED UNCERTAINTIES

- EXAMINED SENSITIVITIES .

ENHANCED METHODOLOGY FOR EPZ DETERMINATION e DEVELOP NUREG-0396 RISK OF DOSE VERSUS DISTANCE CURVES BASED ON PLANT / SITE-SPECIFIC RISK MODEL e CHARACTERIZE TOTAL POTENTIAL FOR RISK ~ REDUCTION BY PROTECTIVE ACTION AS RISK WITH NO EVACUATION e QUANTIFY SPATIAL DISTRIBUTION OF NONEVACUATION :

RISK

/

e CALCULATE ACTUAL RISK REDUCTION FOR PROTECTIVE ACTION STRATEGlES '

MILE EVACUATION MILE EVACUATION MILE EVACUATION j

i

, MILE EVACUATION AND SHELTERING OUT TO 10 MILES i i

e EVALUATE UNCERTAINTIES AND SENSITIVITIES e COMPARE RESULTS WITH ALL AVAILABLE RISK ACCEPTANCE CRITERIA -

1 - ,

EMERGENCY PLANNING SENSITIVITY STUDY l METHODOLOGY o PURPOSE:

DETERMINE IMPORTANCE OF SOURCE TERMS VERSUS PLANT-SPECIFIC FEATURES AND ENHANCED PRA TECHNOLOGY -

e APPROACH: RMEPS CALCULATIONS REDONE USING:

- WASH-1400 SOURCE TERM METHODOLOGY l'

j

- BEST ESTIMATE ASSUMPTIONS ON ALL OTHER UNCERTAIN l RISK PARAMETERS l i i

i e *

.~. .

RISK ACCEPTANCE CRITERIA UTILIZED e NUREG-0396 DOSE VERSUS DISTANCE CURVES FOR 1,5, 50, AND 200-REM WHOLE-BODY DOSES e WASH-1400 RISK CURVES FOR EARLY AND LATENT HEALTH EFFECTS (MEAN & MEDIAN l RESULTS) e NRC INDIVIDUAL AND SOCIETAL RISK SAFETY GOALS e SPATIAL DISTRIBUTION OF RESIDUAL RISK g g ( . _

~

MEDIAN RISK OF EARLY HEALTH EFFECTS FOR DIFFERENT EVACUATION DISTANCES 10~3 d D 10

53'RMe4F#' WASH-1400 SOURCE

?

g

< TERM METH000 LOGY y (MEDIAN RESULTS) e WASH.1400 (MEDIAN RESULTS) gto"5 5

m E ~~

NO MM DIATE PROTECTIVE 4 %

$ 10 -

3 1. MILE N C N EVACUATION \

W N N I \

E \

g20-7 -. N g a \ g E \ g

\ ,

\

20 4 -

2 - uitE . \ \

~ -=== -- EVACUATION {

\

RMEPS RESULTS % k N ~~

/OFF SCALE i i Ns i\

l i iO-S 10 I 3 3 4 10 5 100 10 10 10 EARLY FATALITIES

j - , -

I l THE BENEFITS OF RISK REDUCTION BY l EVACUATION OR SHELTERING ARE:

i I e VERY SMALL DUE TO VERY LOW INHERENT PLANT RISK e FULLY REALIZED BY CLOSE-IN EVACUATION I

I e NOT NEEDED TO MEET NRC SAFETY GOALS M"I e g a y j - S a. .

e O,r seit rorutAlsose wetsmee 3 aests souessany g.3 .

1 I M

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f =4 - -

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2 2 a i F

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g4 Ew Cua aosemess nueves SeattIiesess 30 se eastES E

p i I I f g

4 3 4 0 0 #

Ovatap4f90se gestAeste lbestES

i I

i l COMPARISON OF INDIVIDUAL RISK WITH BACKGROUND AND SAFETY GOAL VALUES

! 10-2

\

d

-) .

y

BACKGROUND ACCIDENTAL FATALITY RISK 10-3 -

(5 FATALITIES PER 10.000 POPULATION PER YEAR) 10 4 -

4

! 5 l* $

10-5 -

f h l 9 SAFETY GOAL (.001 TIMES

! O );$i::: BACKGROUND RISK)

} [ SENS W S N O 10-g -

WITH NOIMMEDIATE

! g PROTECTIVE ACTIONS

'l =8 m

10~7 -

e WITH 1 MILE I ,

WmTM

< ::2:::

S 10-g RMEPS RESULTS z -

WITH NOIMMEDIATE

PROTECTIVE ACTIONS

\

10-8

/ ~

O 4

i

'+

- _ _ _ . _ _ __ 4

i

COMPARISON WITH NUREG-0396 - 200-REM & 50-REM WHOLE-

! BODY DOSE PLOTS FOR NO IMMEDIATE PROTECTIVE ACTIONS I.

3

) - a s aasssg : s a a a s iag s a : s a e i s.

l, - _

~

NUREG-0398 ~

en >-

as 2 - ----- SENSITMTY STUDY _

.i oO i Og - -

, >u """""*""* RMEPS RESULTS

) O[

ap_s

- (200 REM CURVE OFF SCALE) 1' i .a=2

! Om oi --

18u

. o --- _

', z4 - -

l 5z ensnas Uh x 7 ~ ' *% .

1

= =i su u

_ \ _

' Ok \ k 50 REM

>p 200 _

zM \ REM sas O 30 \

O ssa

nii; o.01 es. u

\

i I --

+

! gr

\ I' -

z 1 o =< l to -

l \ -

l OE zP .

I g

l *

\

l

i o3 .\

s uo -

i -\ ll *

/L D 50 REM 200 REM l

i l ,_,,, V.a i e e a e a :I e n i e e a n al i 8 8 8 8 ase 10 100 1.000 l 3 ,

DISTANCE (MILES) 1 .

I i

N

e COMPARISON WITH NUREG-0396 REM & 5-REM WHOLE- '

l BODY DOSE PLOTS FOR NO IMMEDIATE PROTECTIVE ACTION 1

1 1

8

_ . . . .. ...i . . . .... . . . . . . . .

! NUREG-0308 .

i .

Ez -

SENSITMTY STUDY -

i 8E D

! u

>- o .... . . ..... AMEPS RESul.TS o< . .

OU 1 REM 0s "2 om 0,1 SREM -

Ie -

! 3o -

no - -

t

! 24 ~

z

.o - _7 1

yh C n**. g _ . ~.--~-~N b

x

=J . 5., \ \ _

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\ i

\

oh

~

i \

! 55 \ I \

! 88 '

s \ \

-- o.gi . .

\ g :

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I \

. sh \ :

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1 8

pg .

g

\1 REM -

! UE ze on I..

5 REM () \

A g

. g UO s.

1 REM -

s . 1 I

.5 REM i., g I i

, , ,.'..H .i . . . l..I . . . ' ' ' ' '

10 100 ,

IM 1

DISTANCE (MILES)

l t

Safety Study Results: '

i CONTAINMENT EFFECTIVENESS i (Percent of Accident Scenarios) l . !

l I

66 % 99 % 99.9 %

\ i i

\

I 34 % 1% .1 %

i WASH-1400 (1975) '

SEABROOK STATION (1983) SEABROOK STATION

\

IMPROVED LOCA OUTSIDE i

CONIAINMENTMODEL L1985) 1 i -"1 EARLY l

! DEGRADED 4' CONTAINMENT CONTAINMENT OR FAILURE i CONTAINMENTINTACT i

t 8 - -

! PRINCIPAL CONTRIBUTORS TO :

l EARLY RELEASE FREQUENCY i

i INITIATING EVENTS '

WITH CONTAllGAENT BYPASS

- INTERFACING LOCAs #

- STEAM GENERATOR i TUBE RUPTURE l

i EXTERNAL EVENTS WITH POTENTIAL CONTAINMENT DAMAGE FOR

- MCRAFT CRASH > EARLY

- TURBINE MISM RELEASE ,

m LOSS OF CONTAINMENT m i

STRUCTURAL INTEGRITY ALL OTHER NTIATING EVENTS m

CONTAINMENT ISOLATION .

FAILURE i .

l _

CONTRIBUTIONS TO EARLY RELEASE 9 FREQUENCY (S1, S6, S7)

t SCENARIO TYPE PERCENT

~

1. EARTHQUAKE WITH OPEN PURGE 87

2. INTERFACING LOCA 13

3. TURBINE MISSILE IMPACTS CONTAINMENT BUILDING <1

4. REACTOR VESSEL STEAM EXPLOSION <1 l S. AIRCRAFT IMPACTS CONTAINMENT BUILDING <1 i

TOTAL 100 I

t l .

1

t I ENHANCED TREATMENT OF INTERFACING

' SYSTEMS LOCA o MORE COMPLETE MODELING OF VALVE FAILURE MODES e NEW DATA ON CHECK VALVE FAILURES VERSUS LEAK SIZE t e MORE REALISTIC TREATMENT OF DYNAMIC PRESSURE PULSE '

e EXPLICIT MODELING OF RHR RELIEF VALVES e QUANTIFICATION OF RHR PIPING FRAGILITIES TO OVERPRESSURE ,

e MODELING OF RHR PUMP SEAL LEAKAGE o OPERATOR ACTIONS TO PREVENT MELT CONSIDERED

.e THERMAL HYDRAULIC AND SOURCE TERM FACTORS MODELED USING MAAP '

j e UNCERTAINTIES QUANTIFIED t

INTERFACING SYSTEMS

~

LOCA KEY RESULTS .

FREQUENCY (PER REACTOR-YEAR)

UPDATED EVENT SSPSA ANALYSIS VALVE RUPTURES, LOCA 1.8 x 10 -6 7.8 x 10-6 VALVE RUPTURES, LOCA, 1.8 x 10-6 3.1 x 10-7 CONTAINMENT BYPASS

-6 l VALVE RUPTURES, LOCA, 1.8 x 10 4.1 x 10-8 ,

! CONTAINMENT BYPASS, '

i i MELT l .

j 1

);

4 .

-- - - w _.-.. _ _ u_. m. . .. _ a a. .u.. . ,. ._,. _. . _. _.r.- . . .

. o i .

CONTAINMENT FAILURE TYPES A. SMALL LEAK (0.02 SQ. INCHES TO 6 SQ. INCHES)

PRESSURE RISE CONTINUES l

B. LOCAL FAILURE (6 SQ. INCHES TO 60 SQ. INCHES)

PRESSURE RISE CONTINUES '

LEAK RATE INCREASES UNTIL PRESSURE RISE STOPS C. GROSS FAILURE ( > 60 SQ. INCHES) l l

RAPID CONTAINMENT BLOWDOWN ( < 1 HOUR) ,

l -

l

\ _ _ -. _ .

i l

i L

i LOCAL CONTAINMENT FAILURE MODES CONSIDERED '

e FLUID SYSTEMS PENETRATION e HIGH ENERGY PENETRATION e FUELTRANSFERTUBE e ELECTICAL PENETRATION e PURGE LINE PENETRATION e PURGE VALVE SEALS S EQUIPMENT HATCH e PERSONNEL LOCK e OTHER PENETRATIONS l , e LINERTEARING

! O WELDIMPERFECTIONS l

j e -

4

COMPOSITE CONTAINMENT FAILURE PROBABILITY' -

DISTRIBUTIONS FOR TYPE B (LEAK) FAILURE, TYPE C (GROSS) FAILURE, AND TOTAL FAILURE

\

5

yg-4 -

l.

[ ,/

p 's

= ausescas w' = :

/m'u=~~'

\ -

5 j -

f / me r esausmcas

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: l l I w' = : i

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-: 1

.! ,I .

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=: I a l :l I -

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- = 1 j

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> n.- -e -

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+=

4

(

l . .

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t i !

SUMMARY

O EARLY HEALTH RISK VERY LOW EVEN WITHOUT IMMEDIATE PROTECTIVE ACTIONS l -

I i

) e BENEFITS OF EVACUATION VERY SMALL & CONFINED !

j TO AREA CLOSE IN TO SITE O PEER REVIEW GROUP CONCURS WITH RMEPS &

SENSITIVITY STUDY CONCLUSIONS i i e SEABROOK RISK MANAGEMENT ACTIVITIES ARE -

CONTINUING l

' U SEABROOK STATION TWIN PRESSURIZED WATER REACTOR UNITS N EACH GENERATING 1,150 MEGAWATTS OF ELECTRICITY -

\. SEA 8 ROOK, NEW HAMPSHIRE N

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54. Refuelmg water storage tank

55. Spray chemical ecoten tank Ib* / ' $6. Recovery test tares (2)

. t ;

./ 26. Feedwater controlvalves 57.Weste test tants (2)

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27. Main control Doord $8. Reactor water make.up tans

. g,# y<' 28 Reactorprotectionandothe' logic coomete

$0. Boron weste storage tares (2)

60. Pnmary dram tan = (2)

K 2g. Computer room 61. Aereted drain tana (2)

30. Technical support cente' y . 1. Reactor vesses 62. Rose wesie tana

2. Steam generator 31.Caew spreedmg room 63. Sohd weste processeg I .

3. Prosaunzer 32. Station bettenes 64. Weste houid degassifier b . 4. Reactor coolant pump 33. Switengeer 65. Steam generator 04owoown

5. Emipment hatch 34. Diesei generator recovery unit ..

6. Refuehng cavity 35. Fuel storage tar

06. Cochng tower (service weter)

i. Po6er gentry crane 36. Startmg air compressor currge

/, 8.Contamment sprey pong 37. Desee exhaust sdencer 67. Coohng tower ceramic fill y f) Weste oleposalDutung

a. Misade sneed

10. Man steam knes

38. New fuel storage 3g. Fueltranster canai

68. Fans 60 Cochngtowerswitchgear J) Tann farm 11. Reector cooient pipe 40. Cask loading eroe 70. Circuletog weter pumps Ir) Coonngtower 12. Wgn pressure turome 41. Spent fuel storage 71. Traveling screens l

L) Emergencyfeedwaterpump 13. Low prosaure turemos(3) 42. Spent fuel pool heet 72. Intake fiume and forecay

!, budding 14.Generetor exchanger 73. Decnerge velves for back.

l

' 'MS W) Secunty Duieng N) Servicewaterpumsshouse

15. Esciter 16.Monsture separator renester

43. Spent fuelcook encomo eree

44. Bonc acid storage tar *a flush controt

74. Discharge nser snaft

0) Circulategwaterpurrohouse 17. Tureme hait crane 45. Bonc ac*$ mizeg 75. Demmerenzed weter

", do. Let4own dogassifier . _ storage tenas P) Dischargetunnel(transition 18. Steam generator feed pumpe Key to drewWig neer) ig. Condenser shell 47. Volume control ter* 76. See won A) Reectorcontainment(or Q) inisme mannel(trerwtion neer) 20. HigrH> secure feedwater 48. Pnmary plant ventHation 77. Gas ineuteted transmisse reactor Duieng) R) Cheonnetebudding heaters (6m stage) eir supply kne.

8) Turt>ne building $) SteamgeneretorDiow4own 21. Secondary component cool. 4g. Auxihary buildmg exhaust 78. Ausihary transformers (4)

C) Hesterbey(orfeedwater treatment ing heet emenengers air filters 7g. Step up transformers (3)-

noeter buieng) 1) Mainsteam&feedwaterune 22. Loweressure feedwater 50. Let4own domineralizers 80. Auxer ary boiler D) Aomen6etratlentuieng encicoure toteam kne hoeters and hnere 81. Water treatment ptont E) ControsDudding enciosure) 23. Nornogregated phase staten $1. Pnmary cG. ,~., .; cochng 82. Boder stack F) DeoosigeneretorDuddmg U) Emergencycorecoonng service aus pumpe 83.Conoenseto storage tank i O Pnmaryauxderybunding etoment veurt 24. Heeter boy crone 52. Pnmary component coonng 84. Pnmary vent stack H) Fued storage bundmg V) 345-KV Switchyard 25. Mem steem controi velves expensen tarts 85. Sennce water pong

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i

- 10 MI'LE EMERGENCY PLAN,NING ZONE"-

UPON 4 PRINCIPAL CONSIDERAT10 ifs ,

1) DBA DOSES LESS THAN PAGs EEYOND 10 MILES

2) MOSTCDP.EMELTACCID$NTDOSESLESSTHAN '

PAGs EEYOND 10 MILES

3) FOR WORST ACCIDENTS -- PROMPT DEATHS -

GENERALLY WITHIN 10 MILES

4) 10 MILES - BASE AREA EXPAND AD-HOC IF NECESSARY -

ee e

0 e

9 e e

E6 O

i CORE MELT ACCIDENTS CALCULATIONS - EMERGENCY PLANNING ZONE ABOUT 10 MILE FOR PLUME EXPOSURE

1_ -

~

o _

o o -

1 REM op . *

, mz mm 0.1 - 5 REM o'9 n -

Io i S4 -

~

I 0F ~

l Zd og -

i Em UE 50 REM l Xo mu -

O z

>- m '

t-> 0.01 --

O - ,

, m -

4 -

m -

O E -

200 REM a.

e e

1 10 100 1000 DISTANCE (MILES)

T

- ~

l ..

i

,u J

CORE MELT ACCIDENTS CALCULATIONS .

l -

EMERGENCY PLANNING ZONE l.

OF ABOUT 10 MILES FOR PLUME EXPO About 30% chance of exceeding i

PAG doses at 10 miles Less than one chance in 100 of exceeding life threatening do'ses l beyond about 10 miles t

9 g 9

ML20206J395

Contents

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Dear Sir:

Dear Dave:

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ML20206J395

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Dear Chairman Ward:

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Dear Sir:

Dear Dave:

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