ML19341C467

Slide Presentation Entitled, Probabilistic Risk Assessment of Big Rock Point Plant

ML19341C467
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/19/1981
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML19341C463	List: ML19341C462 ML19341C467
References
NUDOCS 8103030556
Download: ML19341C467 (30)

Text

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t ENCLOSURE 2 I

PROBABILISTIC RISK ASSESSMENT

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OF THE BIG ROCK POINT PLANT JANUARY 19, 1981 i

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i CONSUMERS POWER COMPANY AND SCIENCE APPLICATI.0NS, INC.

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PRESENTATION CONTENT 1.

MANAGEMENT OVERVIEW 1.1 PURPOSE OF THE STUDY SCOPE AND APPROACH 1.2 1.3 METHODOLOGY 1.3.1 INITIATOR SELECTION 1.3.2 ACCIDENT SEQUENCE DEFINITION 1.3.3 RELIABILITY DATA 1.3.4 IN-PLANT CONSEQUENCE ANALYSIS 1.3.5 EX-PLANT CONSEQUENCE ANALYSIS 1.4 FORMAT OF RESULTS AND RECOMMENDATIONS 1.5 COMPARABILITY T0 " DECISION RULES" 0F NUREG-0739 1.G USE OF RESULTS 2.

TECHNICAL PRESENTATION ON SPECIAL ISSUES 2.1 ATWS ISSUE 2.1.1 CURRENT RISK CONTRIBUTION 2.1.2 EFFECT OF RECIRC. PUMP TRIP 2.1.3 POTENTIAL MODIFICATIONS UNDER CONSIDERATION 2.2 APPROACH TO SHIELDING EVALUATION

D.

STUDY ~ PURPOSE i

EMPLOY THE TECHN10VES OF PROBABILISTIC RISK ASSESSMENT

'(PRA) TO SUPPORT THE CONTINUED SAFE OPERATION OF THE BIG ROCK POINT NUCLEAR PLANT 4

J l

e t'

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A SCOPE OF STUDY COMPLETE BASELINE PRA'.

o SEQUENCE DEVELOPMENT AND PROBABILISTIC o

-QUANTIFICATION IN-PLANT AND EK-PLANT CONSEQUENCES o

ANALYZED THOROUGH CONSIDERATION OF POTENTIAL o

PLANT MODIFICATIONS ON-GOING DEFINITION OF RISK MINIMlZATION o

PROGRAM

rw I

APPR'0ACH EMPLOYED o

COMPLETE BASELINE PRA

+

INITIATOR SPECIFIC TO PLANT

+

ACCIDENT SEQUENCES (EVENT TREES AND FAULT TREES)

+

PLANT SPECIFIC DATA

+

IN-PLANT AND EX-PLANT CONSEQUENCES o

DIFFICULT ISSUES TREATED DIRECTLY

+

COMMON CAUSE FAILURES

+

INTERNAL EVENTS (E.G., FIRES AND HIGH ENERGY LINE BREAKS)

+

EXTERNAL EVENTS (E.G., SEISMIC AND WIND LOADINGS)

+

EQUIPMENT ENVIRONMENTAL QUALIFICATION o

INCLUDED IN SCOPE

+

UNIQUE APPROACHES TO ASSURING COMPLETENESS

+

FORMULATION AND INVESTIGATION OF EFFECT OF VARIOUS PLANT MODIFICATIONS

+

SIGNIFICANT CPCo PARTICIPATION

'o EXCLUDED FROM SCOPE

+

SABOTAGE

+

DETAll.ED QUANilFICATION OF PROBABILITY OF Fall.URE TO SCRAM b

T r - -

l REVIEW PLANT AND INDUSTRY g

EXPERIENCE FOR-PRECURSORS I TO SIGNIFICANT ACCIDENT i-SEQUENCES I

L _ _ _ _.. _.. _ _ _ _ _J i

y o

CAREFULLY EXAMINE DEVELOP EVENT TREE DEVELOP FAULT TREE EACH BRANCH IN FAULT IDENTIFY ACCIDENT MODELS OF PLANT MODELS OF IMPORTANT REE T

M h8 hNGEVENTS

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EVENT TREE HEADINGS INITIATING EVENTS AN ACCIDENT SEQUENCE g

a e '

FLOW DIAGRAM 0F ITERATIVE PROCESS TO ASSURE COMPLETENESS IN PRA ACCIDENT SEQUENCE DEFINITION 4

l 6

INITIATING EVENTS FOR BRP PRA FOR WHICH EVENT TREES WERE DEVELOPED

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INITIATING EVENT FREQUENCY (YR~)

TURBINE TRIP 1.4 LOSS OF MAIN CONDENSER

.06 SPURIOUS CLOSURE OF MSIV

.06 LOSS OF FEEDWATER

.16 LOSS OF 0FFSITE POWER

.13 LOSS OF INSTRUMENT AIR

.06 SPURIOUS OPENING OF TURBINE

.1 BYPASS VALVE SPURIOUS CPENING OF RDS 1.2x10-3 ISOLATION VALVE SPURIOUS CLOSURE OF BOTH 2.1x10-3 RECIRCULATION LINE VALVES STUCK OPEN SAFETY VALVE 2.6x10-4 INTERFACING LOCA 2.6x10-3 HIGH ENERGY LINE BREAK IN 3.9x10-7

~

RECIRCULATION PUMP ROOM HIGH ENERGY LINE BREAK 3.8y10-6 IN PIPE TUNNEL a

1.0x10-4 SMALL LOCA 1.0x10-5 MEDIUM LOCA 1.0x10-6 LARGE LOCA SMALL STEAM LINE BREAK 1.0x10-4 INSIDE CONTAINMENT O

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5 INITIATING EVENTS FOR BRP PRA FOR WHICH EVENT TREES WERE DEVELOPED (CCSTIGUED)

J NI.TIATlNG_EY.ENI EREQUENCY_LYR__1 1.0x10-5 MEDIUM STEAM LINE BREAK INSIDE CONTAINMENT LARGE STEAM LINE BREAK 1.0x10-6 INSIDE CONTAINMENT SMALL STEAM LINE BREAK 1.0x10-4 OUTSIDE CONTAINMENT MEDIUM STEAM LINE BREAK 1.0x10-5 0UTSIDE CONTAINMENT LARGE STEAM LINE BREAK 1.0x10-6 0UTSIDE CONTAINMENT FIRE IN CABLE PENEIRATION ROOM 1.8x10-3 INSIDE CONTAINMENT WHICH AFFECTS ALL CORE COOLING SYSTEMS FIRE IN CABLE SPREADING ROOM 9.0x10-4 OUTSIDE CONTAINMENT WHICH AFFECTS ALL CORE C0OLING SYSTEMS FIRE IN STATION POWER ROOM WHICH 3.3x10-3 AFFECTS ALL CORE COOLING SYSTEMS FIRE IN CONTROL ROOM WHICH AFFECTS 1.0x10-4 l

ALL CORE COOLING SYSTEMS LNAGE EARTHOUAKE (0.16 PEAK 4 1x10-5 GROUND ACCELERATION 40.459)

MEDIAN =.239 MEDIUM EARTHOUAKE (.0539< PEAK 1x10-4 GROUND ACCELERATION 50.16g)

MEDIAN =.0849 SMALL EARTHQUAKE (.016 <; PEAK 1x10-3 GROUND ACCELERATION s,0539)~

MEDIAN =.03 -

9 LOSS OF CONTROL ROOM 0.14

(

HABITABILITY (9)

J6 METHODOLOGY FOR DEFINING COMPONENT FAILURE RATES FOR THE BIG ROCK POINT PRA THE COMP 0HENT FAILURE RATE DATA WAS USED IN EVENT TREE AND o

FAULT TREE QUANTIFICATION DATAWASTAKENFROMBOTHPLANTSPECIFICANDGENERICDATASOURCES o

PLANT SPECIFIC DATA WAS PREFERRED WHERE IT WAS AVAILABLE AND o

CONSIDERED APPROPRIATE DATA WAS INAPPROPRIATE WHEN THE NUMBER OF DEMANDS OR OPERATING o

HISTORY, WHICH WAS DEDUCED FROM THE PLANT RECORDS, WAS CONSIDERED TO BE NONREPRESENTATIVE (E.G., CONTROL VALVE DEKANDS)

GENERIC DATA WAS USED WHERE PLANT SPECIFIC DATA WAS NOT AVAILABLE o

OR NOT APPROPRIATE k

0 e

45 PLANT SPECIFIC DATA INFORKATIONUSEDTOCOMPILEPLhNTSPECIFICCOMPONENTF o

. RATES WAS DERIVED FROM PLANT RECORDS SOURCES OF INF0hrtAT10N INCLUDED:

o PLANT MAINTENANCE RECORDS; WHICH PROVIDED A DESCRIPTION OF MAINTENANCE ACTIVITIES CONTROL ROOM LOG BOOKS; THESE PROVIDE THE DAY-TO-DAY OPERATING HISTORY SURVEILLANCE TESTS; PROCEDURES BY WHICH SAFETY RELATED COMPONENTS AND INSTRUMENTATION CAN BE TESTED AGAINST STANDARD OF NORMAL OPL?ATION DOCUMENTS WHICH DESCRIBE UNUSUAL OR ABNORMAL EVENTS; E.G., LERs, ERs, DRs, ETC, i

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I i

GENERIC DATA o

SOURCES OF GENERIC DATA INCLUDED:

(1)

WASH-1400, REACTOR SAFETY'. STUDY, AUGUST 1974 (2)

GE-22A2589, RECOMMENDED COMPONENT FAILURE RATES, MAY 1974 (3) lEFE-90, COMPONENT RFIIAPillTY DATA,.1977 (4)

CRNL-704, COMPONENT RrLI ABIL11Y DATA, DECEMBER 1971 (5)

Al-67-TRD-15, RELIABILITY DATA C0'iPILAT10NS, FEBRUARY 1968 (6)

NUREG/CR-1363, DATA SUM" ARIES OF LERs 0F VALVES, JUNE 1980 (7)

NUREG/CR-1205, DATA SUMMARIES OF LERs 0F PUMPS, JANUARY 1980 o

THE RECOMMENDED GENERIC VALUE, USED FOR A COMPONENT FAILURE RATE, UAS TAKEN FROM THE SOURCE MOST COMPATIBLE Wilil THE TYPE AND MODE OF OPERATION OF THAT COMPONENT AT BIG ROCK POINT, t

0 e

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f EXAMPLES OF COMPONENT FAILURE DATA USEDINBIGROCKPOINTPRA o

EMERGENCY DIESEL GENERATOR (PLANT SPECIFIC)

FAILURE TO START - 12/669 0 = 1,79 x 10-2/D FAILURE TO RUN - 7/355 A = 1.97 x 10-2/HR

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o MOTOR OPERATED VALVES (PLANT SPECIFIC)

FAILURE TO OPEN - 7/989 0 = 7'.07 x 10-3/D

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FAILURE TO CLOSE - 10/639 0 = 1,56 x 10-2/D FAILURE TO REMAIN CLOSED -1/1254970 A = 8.81 x 10-7/HR o

GENERIC VALUES FOR MOTOR OPERATED VALVES NOT USED IN ANALYSIS BUT SHOWN FOR COMPARISON FAILURE TO OPEN 0 = 1 x 10-3/D FAILURE TO CLOSE Q = 1 x 10-3/D FAILURE TO REMAIN CLOSED A = 1. 6 x 10-7/HR

ESTIMATES OF HUMAN ERROR PROBABILITIES FOR BIG ROCK POINT o

MANY OF THE BACKUP SYSTEMS FOR BRP SAFETY FUNCTIONS DEPEND ON OPERATOR ACTION o

DETERMINING PROBABILITY OPERATOR WOULD PERFORM ACTIONS REQUIRED TO ALIGN BACKUP SYSTEMS

, USED SWAIN AND GUTTMANN'S " HANDBOOK 0F HUMAN o

RELI ABILITY WITH EMPHASIS ON NUCLEAR POWER PLANT APPLICATIONS" o

FACTORS WHICH DEIERMINE HUMAN ERROR PROBABILITIES

~

EXPERIENCE TRAINING PROCEDURES STRESS i

G

4 IN-PLANT CONSEQUENCE ANALYSIS o

ASSESS POTENTIAL FOR COR'E MELT o

DEFINE RANGE OF SEQUENCE CHARACTERISTICS (E.G., TIMING, CONTAINMENT CHALLENGE) o EMPLOY RACAP TO CALCULATE RANGE OF RELEASES FOR VARIOUS CONTAINMENT STATES o

CATEGORIZE, RELEASES BY SIMILARITY OF TIMING AND QUANTITY RELEASED i

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POTENTIAL CONTAINMENT FAILURE MODES SIGNIFICANT

+

ENCLOSURE ISOLATION FAILURE

+

SHORT-TERM OVERPRESSURE FAILURE (ATWS)

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PRIMARY SYSTEM ISOLATION FAILURE UNIMPORTANT

+

LONG-TERM OVERPRESSURE FAILURE j

+

HYDROGEN COMBUSTION

+

IN-VESSEL STEAM EXPLOSION

+

EX-VESSEL STEAM EXPLOCION

+

BASEMAT PENETRATION

+

NORMAL CONTAINMENT LEAKAGE 7

0 I

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RISK MINIMillNG 1ALIURS l

EXPERIENCEDOPERATINGSiAFF o

)

o LOW RATIO 0F POWER TO CONTAINMENT VOLUME (<0.2 SURRY) o LOW RADIONUCLIDE INVENTORY (~0,1 SURRY) o LOW POPULATION SITE d

1 4

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OUTPUTS OF STUDY o

DESCRIPTION OF RISK-CONTRIBUTING SEQUENCES o

SUMMARY

OF PLANT OPERAflNG EXPERIENCE o

RISK EVALUATION OF RECOMMENDED DESIGN CHANGES o

QUANTITATIVE DESCRIPTION OF ACCIDENT PROCESS AND SOURCE TERMS o

COMPARIS0N OF HEALTH EFFECTS DISTRIBUTIONS CONSIDERING SITE POPULATION f.ND METEOROLOGY o

PLAN FOR PROGRAM TO DEPICT QUANTITATIVELY THE AGING PROCESS h

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SUMMARY

OF DOMINANT SEQUENCES PERCENTAGE CONTRIBUTION TO CORE DAM &SE SEQRENCE CLASS (NO. OF SEQUENCES)

TURBINE TRIP (3) 0,08 LOSS.0F FEEDWATER (1) 0,04 LOSS OF MAIN CONDENSER (6)'

O,38 LOSS OF'0FFSITE POWER (15) 4,57 4,37 LOCA (5)

STEAM LINE BREAK INSIDE 11,18

~ CONTAINMENT (3)

LOSS OF INSTRUMENT AIR (6) 3.35 SPURIOUS CLOSURE OF MSIV (4) 0.33 SPURIOUS OPENING 0F TURBINE EYPASS VALVE (5) 7,22 ATWS (18) 4,78 SPURIOUS OPENING 0F RDS 1,73 ISOLATION VALVE (2)

HIGH ENERGY LINE BREAK (2) 0,15 INTERFACING LOCA (2) 8,84 FIRE (6) 23,37 STUCK OPEN SAFETY (8) 29.47 TOTAL (86 SEQUENCES)

+100, I

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e TYPES OF MODIFICATIONS BEING CONSIDERED 4

o PROCEDURAL CHANGES t

4 EXPANDED USE OF. EXISTING FEATURES

.o o

MODIFICATIONS TO REDUCE HUMAN ERROR PROBABILITY o

EXPANDED EQUIPMENT QUALIFICATION o

PHYSICAL DESIGN MODIFICATIONS l

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. LIST OF RISK OUTLIERS AND SE00ENCE CLASSES AFFECTED E

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- sS EMERGENCY CONDENSER X

X X X X X

X MAKEUP ENVIRONMENTAL QUALI-X X X X X X

X X X

X X

X X

FICATI ON LIMITED FW DURING X

ATWS MSIV BACKUP VALVE X

FAILURE POST INCIDENT SYS-X X XX X X

X X X

X X

X TEM RELIABILITY RDS/CS RELIABILITY X

XX X X

X X X

X X

X STANDBY DIESEL REll-X ABILITY INSTRUMENT AIR SYS-x TEM REPAIR LEAKING RDS VALVES X

SINGLE VALVE ISOLA-X TION OF PRIMARY SYSTEM PROXIMITY OF SAFETY X

SYSTEM PIPING TO HIGH EhtRGY 1.INES CONCENTRATION OF X

SAFETY SYSTEM ELEC-TRICAL CABLES IN SINGLE LOCATIONS LATE AUTOMATIC X

ISOLATION OF MAIN STEAM LINE ON LOSS OF COOLANT SECONDARY SYSTEM X

X X

X INSTABILlTIES

TOTAL CORE DAMAGE FREQUENCY (yr-I) l' l

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.ie BASELINE FREQUENCY l

MODIFIED BASELINE MODIFICATION 1 g

MODIFICATION 1 AND 2 8,

g M00. 1, 2, AND 3 33 M00. 1, 2, 3 AND 4 2

n MOD. 1 THRU 5 ob M0D. 1 THRU 6 MOD. 1 THRU 7 MOD. 1 THRU 12

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1 FRAGILITY CURVE FOR EQUIPMENT ENVIRONMENTAL QUALIFICATION

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FAILURE QUALIFICATION PROBABILITY TEMPERATURE

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TEMPERATURE t

TOTAL CORE DAMAGE FREQUENCY (yr',I)

ENVIRONMENTAL QUALIFICATION BASELINE ENVIRONMENTAL QUALIFICATION MODIFICATIONS MODIFIED ENVIRONMENTAL QUALIFICATION BASELINE MODIFICATION 1 MODIFICATIONS 1 AND 2 y,

8 C

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MODIFICATIONS 1, 2, AND 3 5

ll M00. 1 THRU 4

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MOD 1, 2, 3, AND 5 i

9 M00. 1, 2, AND 6

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MOD.1, 2, 6. AND 7 a

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FIGURE QUALITATIVE COMPARISON OF BIG ROCK POINT RISK WITH DECISION RULES PROPOSED IN NUREG-0739

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LIMITS ON OCCURRENCE OF HAZARD STATE BIG ROCK POINT DECISION RULE ON MEAN FREQUENCY BIG ROCK POINT P0TENTIAL POST M0D.

HAZARD STATE GOAL LEVEL UPPER LIMIT PRE-MOD. STATUS STATUS 3

SIGNIFICANT CORE

<3x10-4/RY

<1x10/RY)

BELOW G0AL BELOW GOAL DAMAGE LARGE SCALE FUEL

<1x10-4/RY

<5X10-4/RY ABOVE LIMIT BELOW G0AL MELT (LSFM)

LARGE SCALE UNCON-

<0,01

<0.1 ABOVE LIMIT BETWEEN G0AL AND TROLLED RELEASE FROM LIMIT FOR MOST CONTAINMENT [GIVEN SEQUENCES (1)

LSFM) (1)

(1)

THIS DECISION SEEMS TO BE ARBITRARY, OPEN TO INTERPRETATION, AND LIKELY UNACCEPTABLE BECAUSE IT IS SO STRONGLY RELATED TO THE SEQUENCE CHARACTERISTICS AND INDEPENDENT OF THE SEQUENCE PROBABILITY.

FIGURE QUALITATIVE COMPARISON CF BIG ROCK POINT RISK WITH DECISION RULES PROPOSED IN NUREG-0739

__LJMITS ON RISK TO YOST EXPOSED INDIVIDUAL (1)

DECISION RULE ON MEAN BIG ROCK POINT FREQUERCY PER SITE-YEAR _

BIG ROCK POINT POTENTIAL POST-MOD.

ERQBABILITY GOAL GOAL LEVEL UPPER LIMIT PRE-MOD. STATUS STATUS INDIVIDUAL PROBABILITY <5x10-6/ SITE- <2.5x10-5/SI- -

YEAR YEA 3 0F DELAYED CANCER DEATH (MOST EXPOSE 3 PERSON)

PROBABILITY OF EARLY

<1x10-6/ SITE- <5X10-6/ SITE-vEAR BELOW G0AL BELOW G0AL YEAR DEATH (MOST EXPOSED PERSON) 1 (1)

DECISION RULES ON MEAN FREQUENCY PER LARGE SCA'_E FUEL MELT HAVE NOT YET BEEN ESTIMATED,

I

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FIGURE QUALITATIVE COMPARISON OF BIG ROCK P0 INT

~

RISK WITH DECISION RULES PROPOSED IN NUREG-0739 SOCIEIAL HEALTH RISK LIMITS BIG ROCK POINT DECISION RULES BIG ROCK ?0 INT POTENTIAL POST-MEASURE OF RISK G0AL LEVEL DPPER LIMIT PRE-MOD. STATUS MOD. STATUS

~

EXPECTED VALUE OF

<2

<10 10 10 DELAYED CANCER PER 10 KWh PER 10 KWh BELOW G0AL BELOW G0AL DEATHS EXPECTED FREQUENCY

<0.4

<2 10 10 0F EARLY DEATHS PER 10 KWh PER 10 KWh BELOW GOAL BELOW G0AL (RAISED TO THE 1,2 POWER)

Time AvM i nble te, Opernter Lo leile et, I.letulil l'olnen Prevent.l tig Ilir.

Transier:

Time Low Level Transier:s Ani: FCM I

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Loss of feedvster and "r.uti FC M @ 60s.>

RDS cannot be transients irv ving prevented opening of :.e turbine

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bypass valve Hir,h Pressure Trt..sie-:: without Feedwnter Aui-C'7T T

Loss of cffsite cver "s..f_ 'OM @ 60s.

RDS cannot be I

prevented transients

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High Pressure Tra.sients with Feedwater An: FCM 180 s.

from Hotwell Mr.usi Pept @ 60s.

120 s.

Loss of enin : nianner and turbir.e tri; transients 5: ?~PT 0 s.

without by;$_ss

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VP - 302 v

VP - 301 e 4 CV 4050 A

,s REACTOR CIRC. PUMP #1 REACTOR CIRC. PUMP #2 A

/s LIQUID POISON SYSILM FLOW DIAGRAM

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Loss of Loss of Hisc.

Total Core Damage Modification Condenser Offsite Power Screma Frequrne:r for ATVS

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1.

Restrict Reject 1.2x10 NC 3.5x10 NC 1.7x10 NC 4.6x10 2.5x10 Line-(Prevent FW trip on TBPV i

opening)

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Load Rejection NC 3.5x10 NC 2.0x10 Capability

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Evironmentally 6.1x10 NC NC 3.9x10 qualify LPS

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Automatic LPS 6.1x10 NC NC 3.8x10"

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Auto RCPT &

3.3x10 NC NC 3.5x10 i

Env Qual LPS

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Restrict Reject 6.1x10 NC NC 1.8x10 LDie Environ.

t il LPS

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-5 r :t rict Reject 1.0x10 NC NC 1.1x10 i

! ine Env Qual LPS Makeup from CDST l

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-5 Rcstrict Reject 1.1x10 NC NCr 1.0x10 Line Env Qual l

LPS Makeup from CDST Auto RCPA

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Restrict Rej 1.9x10-NC NC 9.8x10 Line Auto LPS Makeup from CDST 1

Auto RCPT 4

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e Al'I'RUACll 10 I VAlllAIING UTILITY OF CONTAINMENT SHIELD WALL o

FIRST DEFINE IMPORTANT ACCIDENT SEQUENCES o

ASSESS SPECTRUM 0F ACCIDENTS LEADING T0. SOURCE TERMS IN CONTAINMENT o

DEFINE MAGNITUDE OF POTENTIAL RADIONUCLIDE SOURCES TO CONTAINMENT FOR VARIOUS SE0llENCES (MEl.T AND NON-MI.IT) o CONSIDLR CURRI.Li!VL ALiIUN ROI.L 01-UPIRAIUR IN SEQUENCES i

o ASSESS LOCATT '1 REQUIRING OPERATOR PRESENCE (FOR INF0E.n110N GATHERING OR LOCAL ACTIONS) o ASSESS THE ADEQUACY OF ASSUMPTIONS ON OPERATOR ACTION DURING SEQUENCES o

ASSESS POTENTIAL CONSERVATISMS IN OPERATOR ACTION ASSUMPTIONS o

ASSLSS ACIl0NS PRI:VI.NIED BY PRESENCE OF SOURCE IERM i

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. _. _ - - -.,,