ML20003E727

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Slide Presentation from 810317-18 Meeting in Bethesda,Md Re Licensee Probabilistic Risk Assessment
ML20003E727
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 03/17/1981
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20003E724 List:
References
NUDOCS 8104100315
Download: ML20003E727 (84)
v • d • e


Text

{{#Wiki_filter:' ~ i 4 9 BIG ROCK POINT PRA EXECUTIVE

SUMMARY

l MOTIVATION FOR THE STUDY STUDY OBJECTIVES RESULTS CPCo CONCLUSIONS AND PLAN OF ACTION 1 I-i s 1810.410 0 3 f S FWBucKMAN 7 '17 /n1 . ~.. _. - _ _,

A MOTIVATION FOR STUDY t CPCO Is CONCERNED As0uT PROJECTED' CAPITAL 8XPENDITURES BIG ROCK POINT HAS PERFORMED WELL IN TERMS 0F SAFETY AND AVAILABILITY CVER THE YEARS A BASIS FOR A CORPORATE DECISION ON FUTURE OPERATION OF BIG ROCK POINT Is NEEDED 4 FWBUCKMAN 7 /1 7 /01

t 7 6 ACTUAL .m m es M O 19 5 E v di -w ' E4

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1 J 1980 82 84 86 03 90 92 94 96 98 2000 YEAR OF PIM OPEPATION FWBUCKMAN 3/17/81

I STUDY OBJECTIVES PERFORM A COMPREHENSIVE AND SYSTEMATIC EVALUATION OF PLANT SAFETY t MAINTAIN TECHNICAL OBJECTIVITY IN IHIS EFFORT EXPLORE MEANS OF REDUCING PLANT RISK ~ l EVALUATE EFFECTIVENESS OF REGULATORY ISSUES USING PRA DEVELOP A Cost EFFECTIVE Risk MANAGEMENT PROGRAM FOR BIG ROCK POINT i p FWBUCKMAN 7 /17/91

I PROBABILISTIC RISK ASSESSMENT OF BIG ROCK POINT I OBJECTIVE: i To ACHIEVE COST EFFECTIVE RISK REDUCTION v DEVELOP IDENTIFY SELECT SHORT TERM PLAN LONG IERM PLAN h:5 ~ %kte.t DespWic BRP RISK PLANT >.-IfiPLEtiCNT[10DS y. IMPLEMENT CPMP l PRA OUTLIERS MODIFICATIONS ADDRESS REG ISSUES '. Lpl8*ed No 3 MODEL , DEFINE CRMP 6 O FEUCKMAN 7/17/01

6 10~3 { ~ Big Rock Point. Base Case 10~4 x N U ~ z k10-b 'g ~. u ~ 2 b10'0 a i. 2 b 10-7 f Average Curve from MASH-1400 g. 2 k -0 10 ~ 1 g' f. 2 _f 1 f f f f f f f f f fl I f f f8fff 8 9 t ffffs a f ff$ggl 0 1 2 3 4 5 10 10 10 10 10 10 Latent Cancer Fatalities Per Year. X C'O.40AR150N OF CCDF FOR LATENT FATALITIES PER YEAR BET 11EEtt BIG ROCr. POINT (BASE CASE) AND THE AVERAGE CURVE FROM WASH-1400 FWBUCKMAN RES97E09

f t SLM MRY.OF DOMINANT SEQUENCES 3 i Core Damage t Sequence Class (?:o. of Seouences)_ Frequency (yr-l) TurbineTrip(3) 8'.5 x 10-7 f> F ' Loss of Feedwater (1)

4 x 10-7 Loss of Main Condenser (6) 3.7 x 10-6 C

Loss of Offsite^ power (15) 4.5 x 10-5 L'OCA(Sj 4.3 x 10-5 M Steam Line Break Inside 1.1 x 10-4 / Containment (3) 5 LossofInstrumentAir(6) 3.3 x 10-5 Spurious Closure of MSIV (4) 3.2 x 10-6 Spurious Opening of Turbine 7.1 x 10-5 Bypass Valve (5) j r ~' A1WS (18) 4 7 x 10-5

2. 7 V'a Spurious Opening of RDS 1.7 x 10-5 Isolation Valve (2)

High Energy Line Break (2) 1.5 x 10-6 Interfacing LOCA (2) 8.7 x 10-5 9,/ ~ Fire (6) 2.3 x 10-4 Stuck Open Safety Valve (8) 2.9 x 10-4 TOTAL (86 sequences) 9.8R x 10-4 / 1 FWBUCKMAN 3/17/P1

,.s 1.,---, b ns.A .~._s a n.s s ...1..na s. x u._.+- h a 4 g 8 a e g RISK' REDUCTION ACTION PLAN 6 L 4 fi e e PLANT MODIFICATIONS J t CONTINUING RISK MANAGEMENT' PROGRAM .t i 't 1 4 4 1 ~ 5 i i i d FWBUCKMAN 3/17/81

SHORT TERM PLAN 1. CPC0 WILL SUBMIT PRA DOCUMENT 3/31 2. CPC0 WILL SUBMIT PRA-BASED DOCUMENTATION FOR DEFERRAL ITEMS ON 3/31 3. CPC0 WILL INITIATE DETAILED DESIGN ACTIVITIES ON PRA - RECOMMENDED MODIFICATIONS AS SOON AS THE NRC ISSUES A PROVISIONAL ACCEPTANCE OF THE CPC0 POSITION' STATED IN ITEM 2 ABOVE. THIS PROVISIONAL ACCEPTANCE INVOLVES DEFERRAL OF IMPLEMENTATION DATES FOR THESE ITEMS PENDING COMPLETf0N OF DETAILED REVIEW 4. CPC0 WILL COMPLETE IMPLEMENTATON OF THE PRA-RECOMMENDED MODIFICATIONS SUBSEQUENT TO COMPLE-i TION OF NRC REVIEW GIVEN NRC ACCEPTANCE OF PRA AS AN EVALUATION BASIS FOR FUTURE REGULATORY ISSUES e +- ,wi -,e.,, e4, y - = vv - r r

11 I CONCLUSIONS BRP CAN BE OPERATED AT AN ACCEPTABLY LOW LEVEL i 0F RISK TODAY. PRA~ INDICATES THAT THERE ARE SOME HI'GHLY COST-EFFECTIVE ACTIONS THAT CAN BE TAKEN TO REDUCE RISK. P f BASED UPON PROJECTED CAPITAL EXPENDITURES CPCo HEEDS SOME ASSURANCES THAT PRA WILL SERVE AS AN EFFECTIVE TOOL IN DECIDING UPON PLANT MODIFICATIONS. s i 9 4 e FWBUCKMAN 3/17/81

D-7 SCOPE OF STUDY - t o ' COMPLETE ~ BASELINE PRA ~ o SEQUENCE DEVELOPMENT AND PROBABILISTIC QUANTIFICATION P o IN-PLANT AND EX-PLANT CONSEQUENCES. ANALYZED o THOROUGH CONSIDERATION OF POTENTIAL-PLANT MODIFICATIONS o ON-G0ING DEFINITION OF. RISK MINIMIZATION PROGRAM s 9 / 4 o T 3p- --a--- ~4, ggy ,y .---wgmm-yr y,p,-ge,. ,n y 4-w 4,,-,sy, 4yp-y y 4 y y, m 4,y,,

.~ i 1. -f ..( .) STUDY PURPOSE-. T ' EMPLOY THE TECHN100ES.0F'PROBABILISTIC RISK ASSESSMENT' (PRIJ T0. SUPPORT THE CONTINUED SAFE OPERATION 0F THE . BIG. ROCK POINT NUCLEAR PLANT' J I s e f i - 1 1 t k .4 i 5 h. 1 i i s

APPROACH 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 + FORM'JLATION AND INVESTIGATION OF EFFECT OF VARIOUS PLANT MODIFICATIONS + SIGNIFICANT CPCo PARTICIPATION o EXCLUDED FROM SCOPE + SA130TAGE +. DETAILED QUANTIFICATION OF PROBABILITY OF FAILURE TO SCRAM

~. CORE DAMAGE AND RADI0il0CLIDE nELEASE EVALUATION ACCIDENT DOES THE CORE DOES THE CONTAINMENT DOES THE CONTAINMENT SIGilIFICAtlT YES SEQUENCE SUFFER SEVERE ISOLATE FAIL RADIONUCLIDE = z = = DAMAGE + PRESSURE SOURCES RELEASES + STREf!GTH + HEAT REJECTION CAPABILITY fl0 NO y g SIGNIFICANT CONTAINED RADIONUCLIDE RADI0fiUCLIDE ~ I RELEASE INVEtlTORY f 4 e

COMPARISONS OF CONTAINMENT PHYSICAL PARAMETERS BIG ROCK POINT SURRY POWER LEVEL (MW ) 240 2,400-t 5 6 CONTAINMENT VOLUME (CUBIC FEET) 9.4x10 1.8x10 CONTAINMENT DESIGN PRESSURE 27.0 45.0 '(PSIG) ENERGY RELEASES IN LARGE LOCA INFORMATION REQUIRED o PRIMARY VOLUME (CUBIC FEET) 3,639 (2,718 CU. 8,387 FT. OF WATER) o ' PRIMARY TEMPERATURE (*F) 566*F 572*F o PRIMARY PRESSURE (PSIA) 1,350 2,295 PRESSURE IN CONTAINMENT PEAK PRESS. 39.3 PSIG FOLLOWING A LARGE LOCA 20 PSIG ASSUMING NO CONTAINMENT SPRAY (PSIG) 3 4 VOLUME OF STEAM PRODUCED AT 3.82x10 3.88x10 ATMOSPHERIC PRESSURE BY 1% (3 100*C) (a 100*C)- DECAY POWER (CUBIC FEET PER MINUTE) DIVIDED BY CONT, VOLUME 4.1x10-3 2.2x.10-2 MASS OF U02 (POUNDS) 27,500 LBS. 175,600 LBS. MASS OF Z(RCONIUM IN CORE 11,270 LBS. 36,300 LBS. (POUNDS) PERCENT OF H IN CONTAINMENT 7 To 8% H 12 To 14% H 2 2 2 (ASSUMING ALL ZlR, CON!UM REACTS) REMOVAL RATE CONSTANT'(liR-1) I o IODINE (NATURAL) 1.5 1.4 o PARTICULATE (NATURAL) 0.9 0.6 o IODINE (SPRAY) 0.1 3.0 (BORIC ACID), (IF HYDROXIDE IN SPRAY, 30) o PARTICULATE (SPRAY) 0.6 20.0

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SUMMARY

OF DOMINANT SEQUENCES SEQUENCE CLASS (NO OF SEQUENCES) FREQUENCY (YR~) UNCERTAINTY

  • TURBINE TRIP (3) 8.7 x 10-7 20 LOSS OF FEEDWATER (1) 4 x 10-7 98 LOSS OF MAIN CONDENSER (6) 4.0 x 10-6 35 LOSS OF 0FFSITE POWER (15) 4.7 x 10-5 29 LOCA (5) 4.5 x 10-5 19 STEAM LINE BREAK INSIDE 1.1 v. 10-4 49 CONTAINMENT (3) i LOSS OF INSTRUMENT AIR (6) 3.5 x 10-5 18 SPURIOUS CLOSURE OF MSIV (4) 3.2 x 10-6 40 SPURIOUS OPENING OF IURBINE 7.1 x 10-5 22 BYPASS VALVE (5)

ATWS (13) 2.7 x 10-5 11 SPURIOUS OPENING OF RDS ISOLATION VALVE (2) 1.7 x 10-5 g HIGH ENERGY LINE BREAK (2) 1.5 x 10-6 10 INTERFACING LOCA (2) 9.1 x 10-5 y FIRE (6) 2.3 x 10-4 13 STUCK OPEN SAFETY VALVE (8) 2.9 x 10-4 45 TOTAL (81 SEQUENCES) 9.75 x 10-4 15 UNCERTAINTY TAKEN TO BE 95% CONFIDENCE LIMIT +MEAN t

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STEPS OF PROCESS FOR DEFINING PLANT N0DIFICATIONS RESULTING FROM PRA SENSITIVITY + STUDY ON = ALL MODS ^ LIST OF MODS LIST OF LIST OF TO ADDRESS REDilCED LIST DOMINANT RISK RISK OF MODS SEQUENCES OUTLIERS OUTLIERS ENGINEERING JUDGEMENT ON FEASIBILITY OF ~ ALL MODS CONCEPTUAL LIST OF DETAILED LIST OF MODS DESIGN AND POTENTIALLY ENGINEERING TO BE COST ESTIMATE COST-EFFECTIVE DESIGN AND IMPLEMENTED FOR MODS ON MODS FOR RISK COST-ESTIMATES.: REDilCED LIST REDUCTION UPDATFD PRA I /

SAFE SHUTDOWN AREA RECOMMENDATIONS FOR FIRE PROTECTION 10CFR 50 APPENDIX R BIG ROCK PRA EQUIPMENT ON PANEL FOR SNORT TERM COOLING MSIV CONTROL AND POWER MSIV CONTROL AND POWER EMERGENCY CONDENSER OUTLET EMERGENCY CONDENSER OUTLET VALVE CONTROL AND POWER (1 LOOP) VALVE CONTROL AND POWER (2 LOOPS) EMERGENCY CONDENSER MAKEUP EMERGENCY CONDENSER MAKEUP VALVE FROM FIRE SYSTEM CONTROL VALVE FROM FIRE SYSTEM AND POWER CONTROL AND POWER MANUAL REACTOR IRIP CAPABILITY INSTRUMENTATION l DRUM LEVEL. DRUM LEVEL EMERGENCY CONDENSER LEVEL EMERGENCY CONDENSER LEVEL (ANNUNCIATOR) (ANNUNCIATOR) EMERGENCY CONDENSER VALVE POSITION EMERGENCY CONDENSER VALVE INDICATION MAKEUP E OUTLET VALVE POSITION MAKEUP & OUTLET VALVE l MSIV POSITION INDICATION MSIV POSITION i

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EMERGENCY CONDENSER MAKEUP FAILURE MODIFICATIONS a WATER FROM FIRE SYSTEM 4 bl [ ) { k J t a WATER FROM DEMIN. SYSTEM v j LIQUID TO STEAM FROM PCS PCS MOD 1 REPLACE VEC-1 WITH REMOTE MANUAL D.C. POWERED VALVE MOV FAILS TO OPEN 7.1E-3 OPERATOR FAILS TO ACTUATE 1.0E-3 8,1E-3 MOD 2 REPLACE VEC-1 WITH AUTOMATIC D.C. POWERED VALVE MOV FAILS TO OPEN 7.lE-3 MOD 3 PLACE DEMIN. PUMP AND AN AIR COMPRESSER ON THE EMERGENCY BUS 1.0E-2 CHECK VALVES FAIL TO OPEN 2.0E-3 DEMIN. PUMP OUT FOR MAINT. 7.3E-3 DEMIN. PUMP FAIL TO RUN 2.7E I4 LOSS OF AIR 2.7E-5 1.0E-2 l ,,. - - ~

RESULTS OF EMERGENCY CONDENSER MAKEUP SENSITIVITY STUDY LOSS OF 0FFSITE STilCK OPEN SAFETY Al.L AFFECTED TOTAL CORE POWER SEQUENCES VALVE SEQUENCES SFOUENCES DAMAGE. FREQUENCY MODIFICATION W/0 MOD W. MOD W/0 MOD W. MOD W/0 MOD W. MOD W/0 MOD W. MOD I 1. REPLACE VEC-1 WITH i REMOTE MANUAL D.C. POWERED VALVE 4.7E-5 1.11E-5 2,9E-4 6,8E-5 4[E-4 1.6E-4 9.75F-4 6.76E-4 2. REPl ACE VEC-1 WITH AUTOMATIC D.C. POWERED VALVE 4.7E-5 1.10E-5 2,9E-4 6.7E-5 4hE-4 1,59E-4 9.75E-4 6.75E-4 3. PLACE DEMIN PilMP AND AN AIR COMPRESSER ON EMERGENCY BilS 4.7E-5 1.6E-5 2,9E-4 1.9E-4 43E-4 3,2E-4 9.75E-4 8,36E-4 7 l h

1 I 1 THE NEED FOR SHIELDING AT BIG ROCK POINT NEED MAY COME FROM REQUIREMENTS IN NUREG-0737 OR COMPARISON BETWEEN-THE PLACEMENT OF SHIELDING VS AFFECT ON CORE MELT PROBABILITY 1 4 RMMARUSICH 3/17/81

NUREG-0737 ACTION GUIDANCE 1. REVIEW VITAL AREAS WHICH MAY REQUIRE OCCUPANCY. 2. REVIEW SYSTEMS WHICH MAY CONTAIN LARGE QUANTITIES OF RADIO NUCLIDES. 3. DETERMINE WHICH OF THESE SYSTEMS AFFECT THE VITAL AREAS. 4. DETERMINE THE MAGNITUDE OF THE SOURCE AT THE TIME OF THE ACTION (NO SOURCE, GAP RELEASE, MELT RELEASE, ETC). 5. DETERMINE DOSE. CRITERIA - 15 MR/HR AVERAGE OVER 30 DAYS FOR CONTINUOUS OCCUPANCY AREAS. 5 REM TOTAL EXPOSURE FOR ACCESS INTO AREAS INFREQUENTLY.- t-RMMARUSICH 3/17/81-

METHODOLOGY AND ASSUMPTIONS 1. RADIOACTIVITY WORST CASE FOR DIRECT SHINE - LARGE LOCA WORST CASE FOR AIRBORNE - LOSS OF STATION POWER WITH FAILURE OF THE PCS TO ISOLATE RELEASES NOT AT T=0 BUT OVER TIME OBTAINED FROM INCOR CODES. 2. CALCULATIONAL METHODOLOGY SHINE - USE EQUATION CONCERNING DOSE TO A POINT FROM A FINITE SPHERICAL SOURCE. INHALATION - P.ISK ASSESSMENT METHODOLOGY PROVIDED -RELEASE FRACTIONS VS TIME; X/0 CALCULATED FOR OUR CASE; VENTILATION CHARACTERISTICS USED; FACTOR OF 10-3 USED TO DETERMINE IODINE CONCENTRATIONS CHEMICALLY REDUCING ENVIRONMENT (ALWAYS WATER PRESENT AS STEAM OR WATER) NO OXIDIZING ENVIRONMENT (CON - TAINMENT P -TB P). 3. DOSES 25 REM FOR ACCIDENT CONDITIONS FROM NCRP REPORT 39 & DOE IN EP FACTOR OF 20 REDUCTION IN THYROID DOSE WHEN KI PILLS TAKEN, NCRP REPORT 55. 5 REM WHOLE BODY, 30 REM THYROID FOR CONTINUOUS OCCUPIED AREAS. 4. SCENERIO LARGEL0tA LOSS OF POWER T=0 INITIATION T=0 INITIATION T=15M MELT STARTS T=15M MELT STARTS T=lHR AS NDS T=1.5HR MELT PHASE ENDS T=3 HRS VESSEL MELT THROUGH T=4.4HR REMAINING NUCLIDES RELEASED T=5 HRS REMAINING NUCLIDES RELEASED RMMARUSICH 3/17/81

l REVIEW CONSIDERED VITAL ARFJLS. 1. ALL OPERATOR ACTIONS SPECIFIED BY THE PROCEDURE. 2. FUEL OIL DELIVERY, STANDBY EDG SET-UP, INGRESS AND EGRESS. 3. AREAS OF POSSIBLE MAINTENANCE (CORE SPRAY, HX ROOM, SCREEN HOUSE). 4. INHABITED AREAS.(CONTROL ROOM, TECHNICAL SUPPORT CENTER, ELECTRIC EQUIPMENT ROOM, SAMPLE ANALYSIS AREA). SOURCES 1. CONTAINMENT. 2. POST INCIDENT RECIRCULATION PIPING. MAGNITUDES OF RADIOACTIVITY 1. NORMAL OPERATION LEVELS. 2. GAP RELEASE (APPROXIMATELY 2% MELT RELEASE). 3. MELT RELEASE. RMMARUSICH 3/17/81

OPERATOR ACTIONS REVIEWED THE PROCEDURES AND DETERMINED THE ACTIONS AND WHERE THEY TAKE PLACE. THE VAST MAJORITY OF THE ACTIONS TAKE PLACE BEFORE RDS. AFTER MELT ACTIONS - OCCUPY CONTROL ROOM, TSC GET TO SAMPLE ANALYSIS AREA, ELECTRIC EQUIPMENT ROOM FILL FUEL OIL TANKS (HEATING BOILER, EDG, DIESEL FIRE PUMP, STANDBY EDG) INGRESS AND EGRESS SWITCH TO RECIRCULATION (MANUAL VALVES) OTHER POSSIBLE ACTIONS - REPAIR CORE SPRAY SYSTEM COMPONENTS IN SCREENHOUSE. REPAIR POST INCIDENT SYSTEM COMPONENTS IN CORE SPRAY HEAT EXCHANGER ROOM. ) OPEN MANUAL VALVE IN CORE SPRAY HEAT EXCHANGER ROOM SHOULD MOV FAIL. Sw1TCH STRAINER IN BN "r STRAINERS SWITCH LOADS SHOULD 0; ~~ POWER BE LOST AND CONNECT STANDBY EDG RMMARUSICH 3/17/81 a

l

SUMMARY

OF SHIELDING ISSUE NO ADDITIONAL SHIELDING IS REQUIRED TO SATISFY !!UREG-0737 RESULTS OF BASELINE PRA WILL NOT BE SIGNIFICANTLY AFFECTED BY ADDTION OF SHIELD WALL OR LOCAL SHIELDING POTENTI AL LOCAL SHIELDING REQUIRED FOR PLANT MODIFICATION WILL BE ADDRESSED IN THE DETAILED MODIFICATION DESIGN PACKAGE POTENTIAL FOR FURTHER REDUCTION IN THE DOSE OBTAINED FROM INGRESS OR EGRESS WILL BE CONSIDERED m RMMARUSICH 3/17/81

J RESULTS ALL AREAS CONTINUOUSLY OCCUPIED; < 5 REM. ALL AREA INFREQUENTLY OCCUPIED l (INCLUDES SHORT AND LONG-ACTION); < 25 REM. INHALATION DOSES IN CONTINUOUSLY OCCUPIED AREA; < 0.1 REM TO THE THYROID (POTASSIUM IODINE PILLS TAKEN, REALISTIC IODINE RELEASE-A FACTOR OF 1000 LESS THAN NOBLE GAS RELEASE) FILL OIL TANXS; < 25 REM INGRESS AND EGREES; <12 REM WORST CASE 7 REM TYPICAL INVOLVE ONLY LONG TERM MAINTENANCE ACTIVITIES AND ARE NOT REQUIRED TO BE PERFORMED. SCREENHOUSE - TO FIX COMPONENTS OF CORE SPRAY SYSTEM (FIRE PROTECTION SYSTEM) PRIOR TO RDS IF THE NEED ARISES; DOSE >25 REM (MORE DETAILS FURTHER BACK) i CORE SPRAY HEAT EXCHANGER ROOM - TO FIX COMPONENTS OF THE POST INCIDENT SYSTEM PRIOR TO RECIRCULATION, IF THE NEED ARISES; DOSE >25 REM (MORE DETAILS FURTHER BACK) RMMARUSICH 3/17/81

ACTION REVIEWED RISK ASSESSMENT DOMINENT SEQUENCES REVIEWED TIME TO CORE MELT VS CORE SPRAY FAILURE AND PIS FAILURE CORE SRPAY FAILURES - CANNOT SIGNIFICANTLY AFFECT SYSTEM FAILURE PROBABILITY PRIOR TO CORE MELT PIS. FAILURE - CAN SIGNIFICANTLY AFFECT SYSTEM FAILURE PROBABILITY PRIOR TO CORE MELT FOR NON-BREAK SCENERIOS (17 HOURS UNTIL CORE MELT FROM INITIAL LTC ACTION) HOWEVER DOMINENT SEQUENCES WITH PIS FAILURE MAKE UP ONLY 2.4E-4 0F THE 9.8E-4 MELT PROBABILITY SO FIXING PIS WOULD NOT SIGNIFICANTLY AFFECT TOTAL CORE MELT PROBABILITY AS A RESULT SHIELDING WILL NOT BE PLACED IN THESE AREAS UNLESS AS A RESULT OF THE REVIEW OTHER ISSUES ARE RAISED I t RMMARUSICH 3/17/81

IYPICAL ACCIDENT SCENERIO LOSS OF OFFSITE POWER SCRAM EMERGENCY POWER ACTIVATION MSIV'S CLOSE (ON LOSS OF AC) EMERGENCY CONDENSER TRIED TO REMOVE DECAY HEAT BUT IT FAILS RESULTS IN RAPID PRESSURIZATION. PRESSURE CONTINUES TO RISE UNTIL SAFETY VALVE SETPOINT IS REACHED. SAFETIES OPEN PLANT COOLING WITH FEED AND BLEED BUT SPHERE IS FILLING WITH WATER. MUST USE CORE SPRAY /RDS,IF THIS FAILS, NO COOLING AND CORE MELTS. l IF RDS/ CORE SPRAY GOES, NEED LONG TERM COOLING. IF LONG TERM COOLING FAILS, CORE WILL MELT LATER. i RDS/ CORE LIC SPRAY FAILURE FAILURE CORE GAP RELEASED (RDS) 5.9 HRS 5.9 HRS CORE MELT STARTED 6.3 HRS 40 HRS MELT 80% 11.5 HRS 46 HRS RMMARUSICH 3/17/81 ~

CALCULATIONS CONTROL ROOM, TSC ELECTRICAL EQUIPMENT ROOM (720 HRS CONTINUOUS) DAMPER CLOSED BY PROCEDURE ON HIGH AREA MONITOR ALARM AT 0.09 MR/HR (5CFM LEAK) SHINE FROM CONTAINMENT < 5 REM FOR ISOLATED CONTAINMENT CASE. SHINE FROM AIRBORNE CLOUD <1 REM FOR NONISOLATED CASE. 4 SAMPLE ANALYSIS AREA SHINE FROM T=1 HR 27 R/HR CONTAINMENT 5HR 6.3 R/HR (ISOLATED) 24 HR 0.7 R/HR SHINE FROM AIRBORNE RELEASE < 5 REM OVER 720 HOURS FILL Oil T NKS, START STANDBY EDG 4: 25 REM G RMMARUSICH 3/17/81

INGRESS & EGRESS (WORST CASE - LARGE LOCA) TIME (FR0n DOSE RATE DOSE RECEIVED INITIATION R/HR (USING 36 SECOND ENTRY TIME) 0 0 0 15 MIN 30 R/HR 0.3 REM 1 HR 1275 R/HR 12.75 REM 2 HR 1087 R/HR 10.87 REM 8 HR 435 R/HR 4.35 REM 24 HR 120 R/HR 1.20 REM TYPICAL CASE - LOSS OF 0FFSITE POWER - CS FAILS 0 0 0 1.75 HR 0 0 2.25 HR 11.25 R/HR 0.11 REM 4.24 HR 675 R/HR 6.75 REM 8 HR 405 R/HR 4.05 REM 24 HR 108 R/HR 1.08 REM' TYPICAL CASE - LOSS OF 0FFSITE POWER -PIS FAILS 0 0 0 5.9 HR 0 0 8 HR 4.05 R/HR 0.04 REM 24 HR 1.5 R/HR 0.015 REM 46 HR 67.5 R/HR 0.675 REM RMMARUSICH 3/17/81

PROBLEM AREAS SCREEN HOUSE TIME AFTER TRIP, HR MELT RELEASE T=1 2834 R/HR 5 667 24 77.5 720 16.2 CORE SPRAY NEAT EXCHANGER ROOM SHINE FROM CONTAINMENT INSIDE ROOM ROOM NEAR PUMP TIME (AFTER DOSE RATE DOSE RATE GENERAL FIELD _IRIP,HR) R/HR R/HR R/HR T* 1 426 4260 1239 5 46.8 468 771 24 4.9 48.8 161 720 1.8 18.1 6.2 I l RMMARUSICH 3/17/81

i e NUREG 0737 ITEMS AFFECTED BY PRA DIRECT AFFECT POTENTIAL AFFECT II.B.2 vfy DESIGN REVIEW OF PLANT SHIELDING AND I.A.l.3 SHIFT MANNING ENVIRONMENTAL QUALIFICATION OF ELEC-I.C.1 GulDE FOR EVALUATION AND DEVELOPMENT OF TRICAL EaulPMENT FOR SPACES / SYSTEMS VCR PROCEDURES FOR TRANSIENTS AND ACCIDENTS WHICH MAY BE USED IN POST-ACCIDENT ONTROL ROOM DESIGN REVIEWS OPERATIONS II.B.3 sfqe: POST-ACCIDENT SAMPLING CAPABILITY . I. D. 2,te SAFETY PARAMETER DISPLAY II.B.1-".=: REACTOR COOLANT SYSTEM VENTS II.F.2 yq,mINSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING I I.D.1,ym PERFORMANCE IESTING OF BWR AND PWR PELIEF ~ AND SAFETY VALVES-II.K.3.14 ISOLATION OF ISOLAITON CONDENSERS ON 'IVet HIGH RADIATION I I. F.1 3 b f,-. TT.5 CONTAINMENT WATER ~ LEVEL MONITOR !II.K.3.19 INTERLOCK ON RECIRCULATION PUMPS 'N" LOOPS .II.K.3.20 -LOSS OF SERVICE NATER FOR BIq ROCK POINT bl% i III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES II.K.3.25 EFFECT OF LOSS OF ALTERNATING CURRENT 'ec (lluREG-0696) m.'e z POWER ON PUMP SEALS III.D.3.4 CONTROL ROOn liABITABILITY II.K.3.44 EVALUATIOi! 0F ANTICIPATED TRANSIENTS WITH ' '/em ( ^ 'e i SINGLE FAILURE TO VERIFY NO FUEL FAILURE ~ III.A.2 IMPROVINe LICENSEE. EMERGENCY PREPAREDNESS-LONG IERf1 GUREG-0E54/ FEMA-REP-1) I. I

s .0THER ISSUES AFFECTED BY PRA ' I. SYSTEMATIC EVALUATION PROGRAM (SEP)- SEISMIC AND - SYSTEMS INTERACTION II. ELECTRICAL EQUIPMENT QUALIF1 CATION III. FIRE PROTECTION IV. GENERIC ISSUES 4 EXAMPLES: ONSITE AND OFFSITE POWER SOURCES CONTROL OF HEAVY LOADS (NUREG-0612). CONTAINMENT INTEGRITY ATWS 4 V. IE BULLETINS EXAMPLES: IE BULLETIN 80 BROWNS FERRY FAILURE TO SCRAM IE BULLETIN 81 MECHANICAL SNUBBER RELIABILITY m.. ,,s_..._,__ , _ _ ~

ACCIDENT CONSEQUENCE ANALYSIS RACAP_ CODE _NETjiORK B0ll CORE UNC0VERY ~ AND MELTDOWN PVMELT D 5 -PRESSURE VESSEL d gy' HEAT TRANSFER AND E Y MELTTHROUGH E~ ATMOSPHERE EXCHANGE FLO S INTER CORE-CONCRETE INTERACTION CONTAINMENT CONDITIONS u m Se' CORRAL FISSION PRODUCT JVEtiT_ d2 C RELEASE RADIONUCLIDE SCHEDULE TIMitK 25 RETENTI0tj gm, cc ATMOSPHERIC SOURCE TERM H Tf ACCIDENT S PROBABILITY S' RADIONUCLIDE DISPERSION WEATHER AND I, AND PUBLIC CONSEQUENCq POPULATION y DATA U RISK ESTIMATE

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O EfGURE_ .00ALITA IVE) COMPARISON OF BIG ROCK POINT RIS'l WITH DECISION RULES PROPOSED IN NUREG-0739 ~ LIMITS ON OCCURRENCE OF HAZARD STATE ~ M BIG ROCK POINT DECISION RULE ON MEAN FREQUENCY BIG ROCK POINT k N ENII b POST M0D. HAZARD STATE G0AL LEVEL UPPER LIMIT PRE-MOD. STATUS STATUS SIGNIFICANT CORE <3x10-4/RY '<lxlh/RY) -BELOUG0AL -B510W'GOA1: DAMAGE i x '6TYA

s. w os' 3.,- t,,.u r aa=w co *-

A6 vr5* LARGE SCALE FUEL <1x10-4/RY <5x10-4/RY AB0VE/ LIMIT BELOW GOAL MELT (LSFM) 1 x #6/rg 6 4 no* 1 'I~0VEUMIT RtfWEEhG0AL;ANI-LARGE SCALE UNCON-J0.01 <0.1 B TROLLED RELEASE FROM ~ o. 4 Igl@IT4.0RMOST CONTAINMENT [GIVEN SEQUENCsS2(14 LSFM) (1) A[g"'7 (Imt.uOs#( M O s I (1) THIS DECISION SEEMS TO BE ARBITRARY, OPEN TO INTERPRETATION, AND LIKELY UNACCEPTABLE i BECAUSE IT IS SO STRONGLY RELATED TO THE SEQUENCE CHARACTERISTICS AND INDEPENDENT OF ) s \\.,. THE SEQUENCE PROBABILITY. / esem - e aw m,w=

!(FIGURE _. QUALITATIVE)COMPARIS0N OF BIG ROCK POINT RISK WITH DECISION RULES PROPOSED IN NUREG-0739 LIMITS ON RISK TO MOST EXPOSED INDIVIDUAL (1) DECISION RULE ON MEAN BIG ROCK POINT FREQUENCY PER SITE-YEAR BIG ROCK POINT P0TENTIAL POST-MOD. PROBABILITY G0AL G0AL LEVEL UPPER LIMIT PRE-MOD. STATUS STATUS INDIVIDUAL PROBABILITY <5x10-6/ SITE- <2.5x10-5/ SITE-YEAR YEAR 0F DELAYED CANCER DEATH (MOST EXPOSED PERSON) PROBABILITY OF EARLY <lx10-6/ SITE- <5X10-6/ SITE-YEAR BELOW G0AL BELOW G0AL YEAR DEATH (MOST EXPOSED PERSON) (1) DECISION RULES ON MEAN FREQUENCY PER LARGE SCALE FUEL MELT HAVE NOT YET BEEN EST 1 e 9 ..~

-(j ~ N E GURE_ QUALITATIVE. COMPARISON OF BIG ROCK POINT RISK WITH~ DECISION RULES PROPOSED IN NUREG-0739 SOCIETAL HEALTH RISK LIMITS BIG ROCK POINT DECISION RULES BIG ROCK POINT POTENTIAL POST-MEASURE OF RISK GOAL LEVEL UPPER LIMIT PRE-MOD. STATUS MOD. STATUS EXPECTED VALUE OF <2 <10 t 10 10 DELAYED CANCER PER 10 KWh PER 10 KWh BELOW GOAL BELOW GOAL DEATHS I 10 l OF E RLY DE THS PER KWh PER 10 KWh BELOW G0AL BELOW GOAL (RAISED TO THE 1.2 POWER) I 1 i i

DATA COLLECTION FOR THE BIG ROCK POINT NUCLEAR PLANT RISK ASSESSMENT PLANT. SPECIFIC DATA COLLECTED FOR USE IN BIG ROCK POINT FAULT TREE QUANTIFICATION DATA USED TO DETERMINE COMPONENT FAILURE RATES, MAINTENANCE AND TEST UNAVAILABILITIES, REPAIR AND RESTORATION TIMES, AND OPERATOR ERROR PROBABILITIES SOURCES OF PLANT SPECIFIC DATA (1) Plant Maintenance Orders. - pl' ant M0's provided information on the cause of component failure and the time to accomplish repairs. The data was used to coapute failure rates, failure modes and repair times. (2) Control Room Log Books - CRLB's are the daily operating history of the plant. They provide a history of the operation of equipr.ent and actions taken by the operator during plant operations. From this source, data on equipment operation and outages, failure nodes and failure rates can be obtained (3) Surveillance Tests-SRVT's are procedures by which safety related components and instrumentation can be tested against standards.of normal cperation. Information on failure rates, component operating history and test and maintenance unavailabilities can be obtained. (4) Licensee Event Reports, Event Reports, Deviation Reports, Plant Review Coranittee Meeting Minutes and fiRC Correspondence - from these 4 documents additional component failure information was obtained. They also provided additional insight on failures found in other sources which did not elaborate on the causes. DATA GATHERED FOR TEN-YEAR PERIOD 1970-1979 i 4 6

DETERMINATION OF COMP 0i4ENT FAILURE RATES COMPONENT FAILURE MODES SUMMARIZED FROM DATA COMPONENT DEMANDS AND OPERATING TIMES DETERMINED FROM DATA SOURCES AB0VE INFORMATION COMBINED TO DETERMINE COMPONENT FAILURE RATES 4 e t 4 I e e 9 S n .Oe e 6 . - -, - ~.,, _,,, _, ,ry,y_ ,...,_..._-_---m-. .,o -._y,_r y.-,_.,,-,.,,r,__-,, _.y,,_.,

...-a Sur?W(Y OF BIG It0CK "Oltti COMPONENT FAILURES Arl0 OUTAGES EMERGENCY O!ESEl. GENERATOR ~ 1%,f ( g490'iENT DESCRIPT10rt SOURCE FAILURE t*CDE i 11/15/74 Emergency diesel genera tor Tagged out; I hour CRLD155 11/22/74 Emergency diesel generator Replaced fuel transfer pump; M074EP5527 outage 2 hours M074EPS32413 CRLB151 l l l !?/5/74 Emergency diesel generator Selector switch to off; CRLB156 l 15 minutes !4/10/75 Emergency diesel generator Failed to start; started af ter A0-9-75 ll] hand priming; approximately 30 minutes i l {7/17/75 Emergency diesel generator , Tagged out; I hour CRLB163 {3/24/76 Emergency diesel genera tor Tripped on high temperature; ORPI-050155-032476 FT9 inlet screen plugged; estimated outage.9 hours 5/!3/76 Emergency diesel generator Tripped af ter running I hour BRP1-050155-051676 Fip l l and 15 minutes on high tempera-M976EPS12806 ture; cleaned suction screen; [ . outage 9 hours 5/!?/76 Emergency diesel generator No indication of cooling water CRLB173 FT9 flow af ter start; diesel BRPI-050155-057877 shutdown; repacked coolant. M076EP513806 pump; cleaned suction screen; M076EPS13901 i outage 12.5 hours 1 ! 5/5/76 Emergency diesel-generator No voltage indication; outage CRLB173 FT5 6.5 hours M076EP51009 8/5/76 Emergency diesel generator Did not start within 15 seconds; ER-0-76-22 FTS p not retested to see if it would l start t 8/12/76 Emergency diesel riencra tor Failed to start; battery cable ER-B-75-22 FT5 M976EPS22508 burned.off;_ outage I hour

I OF BIG It0CK P0;Ni C0;.u'ONLill i ALLURES AND OUfAGtS S UMo'I i' ump 5 i. t I i i i SOURCE - FAILURE :COE .D/. ',E COMPONENT t .D.'._5_.C.R I PT 1.0__:1 L i 7/6/73

  1. 1 Control rod drive pump f P.eplaced relief valve; outage M073CR0408

' I hour ICRLD137 r f 9/21/73

  1. 1 Control rod drive pump iTagged out; reason unknown; iCRLB140 outage 4 hours j

I l l 9/24/73

  1. 1 Control rod drive pump

- Tagged out; 6.5 hours iCRLul40 I 9/25/73

  1. 1 Control rod drive pump Tagged out; 11.5 hours

!CRLB140 lCRLB140 [ 9/27/73

  1. 1 Control rod drive pump

' Tagged out; 16.5 hours CRLB414 FT6 ltG73CR0692 11/5/73 I

  1. 1 Control rod drive pump Relief valve lifting; valve I

replaced; outage I hour I .i/14/75

  1. 1 Control rod drive pump iLeak in discharge ell'ow; outage !M374CR0029 FTp 25 hours

!CRLIll44 1 1/26/74

  1. 1 Control rod drive pump Repacked pistons; outage 4 hours jt974CRD058

,CRLB144 i 5/31/74 -f1 Control rod drive pump l Pump packinri blown; est. outage !M'J74CR033006 FTB l l1 hour 7/31/74

  1. 1 Control rod drive pump 1 Packing 1<>.ik; outage 3 hours

!N.574CR0419 iCRLB152 l I C/15/74

  1. 1 Control rod drive pump
Brnken vaiv.
spring ieplaced;
Md74CR0447 FTD l outage?.Shours j

e I a t i

~- SOfWA*(Y OF O!G ROCK l'0!'O COI1PbbErlh f A!!.URES AND OUTAGES co mtot. vat.vt.f. .EtCE C0f*04ENT DESCRIPT10f t SOURCE ' FAILURE MODE ~/17/77 CV4094 Failed to close durin9 Lest_ CRLB190 FTC 1/27/77 CV4095 Failed te close during test CRLQ190 FTC !0/10/79 CV4094 & CV4095 Failed leak rate test Sury. test T180 01 Il part A .~/ 31/ 75 CV4096 Failed to seat properly; AD 75 IL l packing misadjusted 1 5/0/77 CV4096 Replaced limit switch MW77CIS15802 CRLB202 IL ~/5/78 i CV4096 1. CV4097 Failed pressure test g I i CV4097 ' Flange leak; bolts tightened BRP1-050155-042674 XL ?/26/7. 7/21/75 CV4097 Failed to seat durin9 test M675C1507901 IL 5/26/75 j CV4097 ^ Failed pressure test AS-14-/5 IL ./17/75 l CV4097 Failed leak rate test jBRPl-050155-II. i l04177 3/19/76 CV4097 Failed leak rate test BRP1-050155-IL l 06197 I ./20/78 I CV4097 Failed Icak rate test BRP1-050155-IL 012078 /21/73 l CV4097 Fai'ied to open on first attempt; M678CIS02303 FTD ~ i ' opened on second 0-BRP-78-09 CV4097 Spurious closure; failed to M678CIS02506 FTRA.

/25/79 i

I reopen I i

~^' 1.i__. _ .. ~. j SUMMAin UF BIG HUCK Pull:t LOMP0tilfil l'All,URES AND OUTAGES MOTOR OPERATED val.VES DATE COMPONErlT DESCRIPTION, SOURCE FAILURE ?'0DE 10/30/78 HOV7062 (Encrgency Condenser ' Failed to close by remote MA78EC530303 FTC Inlet Valve) manual controller; valve stem cleaned 12/10/79 N07062 Valve failed to open af ter hand E-BRP-79-41 FT0 tightened against backscat; estimated outage 93 hours 11/11/73 M07053 (Emeroency Condenser Emergency condenser outlet CRLB141 FTC OutletValve) valve would not close ,11/14/73 M07053 Would not close CRLB141 FTC 14/5/73 M07053 , Failed to open; adjusted A0-8-73 'FTB packing 6/5/78 M07053 Was inoperable af ter hand DRP1-050155-060578 tightened against backscat; estimated outage 2184 hours 1/25/72 M07063 (Emergency Condenser Failed to close; motor burned March 3, 1972 FTC outlet valve) out 'a f ter improperly set' letter to AEC torque switch, motor replaced CRLB117 l10/23/75 M07063 Control pulled to stop for work CRL0l65 on RE070; 15 minutes CRL0l66 '11/11/75 M07056, 7057,7058, Shutdown inlet and outlet 7059 (Shutdown cooling valves control placed in pull system primary and secondary to stop position; 3 hours valves) 0-BRP-77-104 FTC i8/27/77 MD7051 (Core Spray Valve) Failed to close O I I i

l A?,l E 111-3 f ALLURE RATES CU4PuiED f ROM E!G ROCK POINT SPEC 1F1C D4TA TOTAL TOTAL T OT Al. FAllU;E CD'.?O';E NT FAILURE LODE FAILURES DE!'. ANDS OPEP.4 TION RAT E E.arstncy Diesel failure to sta'rt 12 669 1,79x10-!/d St r.e ra to r failure to run 7 355

1. 97 x 10";h'r e

. CR0 Pump failure to run 13 67894 1.91 x 10/hr -2 . Feedwater Pump failure to start 4 297 1.34 x 10 /d failure to run 8 119520 6.69 x 10 /hr 87648 . Service Water Fump 1.43x10-!/d ). Condenser failure to start 3 209 Circulating failure to run 3 119520 2.5 x 10$hr Unter Pump Demineralized failure to run 3 44820 6.69 x 10-5/hr 2. Water Pump

7. reactor Cleanup failure to run 18 59760 3.01 x 10

,'h r Fump -5

0. Shutdown Cooling failure i.o run 1

27888 3.58 x 10 /hr Pump 2.16 x 10-3/d i). Condensate failure to start 1 462 T u,1p failure to run 2 119520 ~1.67 x 10- /hr >). Reactor Cocling failure to run 1 87648 1.14 x IC- /hr Water Pump

1. Fuel Pit Pump failure to start 1

572 1.74xloi/d failure to run 11 87648 1.25 x 10 /hr -3

2. Electric Fire failure to start 2

355 399 5.63 x 10 /d Pump

3. Diesel Fire failure to start 1

326 146 3.06 x 10 /d Pump ~

TABI E !!!-3 FAILURE RAIES CD:;duiED IRW. BIG ROCK POINI SPECIFIC OATA TOTAL TOTAL TOTAL FAILL2E CO'?C::ENT FAILURE MODE FAILURES DE!G.NDS OPEP.ATION RATE 3.84 x 10~2/d

14. I'SIV (F37050) failure to close 2

52 1.42 x 10-1/d

15. CV4C'.4 (Turbine failure to open 4

28 Bypass Valve)

16. Control Valves failure to open 16 750 2.13 x 10' /d failure to clo*.e 16 6?1

?.57 x 10~ /it failure to remain 4 66 9 59/60 1.01 x 10'*/hr open 7.6' x 10','hr failure to remain 3 66 0 59760-closed

17. Motor Operated failure to open 7

989 21 0 59760 7.07 x 10~ /d Val.es failure to close 10 639 1.96 x 10':/d failure to remain 1 1754970 8.31 x 10~ /'n r closed -2

18. Cori Spray failure to open 3

230 4 0 59760 1.3 x 10 /d Val.es (MJ7051, 7061, 7070, & 150032 7071) 2

19. Emer;ency failure to open 2

125 1.6 x 10 /d Con f e r.sc r Val.es 20. t failure to 19 J?/

8. 7 / v 'S Isi. nion *!alves isolate (CV: 25, 402/,'

failure of leak 16 160 .10/d 6031, 4091, 4092, test 4092, 40?4, 4095, 409s, 4097, 4102, 6103, 4117, 4200, M07:50, 7065. 7067, VF',19, VRW304) 4.65 x 10-3/d

21. RDS isolation failure to open 1

215 15240 Valve q-- ,--v e ,n g-r --,,y m-

GENERIC DATA USED~IN BIG ROCK POINT RISK ASSESSMENT . GENERIC DATA WAS USED WHERE PLANT SPECIFIC DATA WAS EITHER UNAVAILABLE OR CONSIDERED INAPPROPRI ATE SOURCES OF GENERIC DATA (1) WASH-1400, REACTOR SAFETY STUDY, AUGUST 1974 (2) GE RECOMMENDED FAILURE RATES, GE-22A2589, MAY 1974 (3) IEEE-500, COMPONENT RELIABILITY DATA, 1977 (4) NUREG/CR-1363, DATA SUMMARIES OF LER's, JUNE 1980 (5) NUREG/CR-1205, DATA SUMMARIES OF LER's 0F PUMPS, ' JANUARY 1980 ENGINEERING JUDGEMENT USED TO DETERMINE RECOMMENDED VALUE FOR A PARTICULAR COMPONENT a "Y--,.--- a a rm -c'w d mm 'Jm ~ m e L--- A,- A--m------ -a w ----- -r-m,---_-_.-

Table III-4a GENERIC DATA - VALVES EVElli 10ENTIFICATION SdURCES III

4. CRNL-704(2)
5. A!(3)
6. EGG Recountended
1. WA.S.il.1400
3. IEE_E-500
2. GE VALVES i

y A A 1 1 A Q A Q ~ MotorOperated(MOV) FT0 1.0 1.5 2.5 1.0 1.5 1.0 FTC 1.0 1.6 '2. 5 L=0,5

0. 0 1.6 1.0 34 FTRO 0.1 0.15 0.124 U=30 0.15 0.1 FTRC 0.1 0.16 0.124 0.16 0.1 i

i Safety FTO 0.01 1.1 L-1 ~ 8/d(6) l l 8'(7) ~ FTC 2.7 U=10 11.4 3.0 2.7 1.2 FTRC 10 0.42 3/d 0.42 Check FT0 0.1 0.16 L=0*f 0.08 0.1 0.08 0.1 FTC i 0.3

0. 5
0. 3
0. 5 U=10 2.3 Reverse Leak 0.3 1.6
0. 5 0.5 Diaphragm FT0 0.3 2.1
1. 0 2.1
0. 3 FTC 0.3 2.1
0. 8 2.1 0.3 FTRO 0.1 2.0 (0.15) 0.1 FTRC 0,1 (0.16) 0.1 Solenoid (SOV)

FT0 1.0 3.9 1.0 3.9 1.0 FTC 1.0

3. 9 L=1
0. 8
3. 9 1.0 FTRO 0.15 0.14 U=40 0.15 (0.1)

FTRC 0.15 0.14 0.15 (0.1) ~ ~ ~ 4 gwm~~U $dd (1) IEEE-500 data applies to actuation only. WO '" M 8fd-(2) CRNL-704 data includes upper (U) and lower (L) bounds; brackets indicate rate is for all failure modes >,g4 # (3) Af data rates are for all L.ilure modes (hence brackets) (6) A is hourly failure rate reher indicates failures per million hours M M# / (5) Q is demand failure rate, nn.her indicates failures per thousand demands ac.W' M M (6) Value for BWR relief valve. gg, (7) From CE data w

,8 MAINTENANCE AND TES) UNAVAILABILITIES USED IN BIG ROCK l POINT RISK ASSESSMENT MAINTENANCE. PHILOSOPHY AT BIG ROCK POINTLIS 10 REPAIR FROBLEMS AS THEY OCCUR ONLY MAINTEllANCE OUTAGES WHICH OCCURRED WHILE THE PLANT 'WAS AT FULL POWER WERE CONSIDERED NOTESTUNAVAILABILlilESOF-SAFETYSYSTEMSDURINGFULL POWER OPERATION t G 9 4 4 Oe oe-oog =g

I?BIE 111-5 !%lli1E::A';CE V'; AVAIL ABILITI ES C07.PUTE0 FAS",THE BlG ROCK F0lti SPECIFIC DATA ... - =

UY3ER Of TOTAL

%1:11E!;ANCE CC:;;0';EtiT C0:00' E! lS CUi/GE U::AVAit AElliTY

1. Er.ergency Diesel 1

180 3.01 x 10-3 Cer.erator

2. CR0 Pu;np 2

380 3.17 x 10-3

3. Feeditater Pump 2

127 1.06 x 10-3

4. Service Water Pu:.p 2

24 2.0 x 10~4

5. Cor. denser Circulating 2

285 2.4 x 10-3 Water Pump -3

6. Demineralized Water Pump 1

434 7.26 x 10

7. Reactor Cleanup Pump 1

117 1.95 x 10-3 ~4 E. Shutdown Cooling Pump 2 61 5.1 x 10

9. Ccndensate Pump 2

58 4.85 x 10-4

10. Reactor Cooling Water 2

192 1.60 x 10-3 Pump

11. Fuel Pit Pump 2

4.5 3.76 x 10-5 -4

12. Electric Fire Pump 1

48 8.03 x 10

13. Diesel Fire Frp 1

8 1.33 x 10-#

14. CVa014 (Turbine Bypass 1

151 2.52 x 10-3 Valve)

15. Motor Operated Valves 21 85 7.49 x 10-5
16. Core Spray Valves 4

10 6.2' x 10-5/ valve (M070T., 7061, 7070 & 7071) '7. Emergency Condenser 4 2277 9.5_x 10-3 Valves g4 ..e e a, 4._e = ....r.

HUMAN ERROR PROBABil.ITIES USED IN BIG ROCK POINT RISK ASSESSMENT MANY BACKUP SYSTElis REQUIRE OPERATOR ACTION TO FUNCTION USED SWAIN AND GUTTMANN'S " HANDBOOK OF HUMAN RELIABILITY WITH EMPHASIS ON NUCLEAR POWER PLANT APPLICATIONS" AS FOUNDATION FACTORS WHICH DETERMINE HUMAN ERROR PROBABILITIES (1) F.XPERIENCE ~ (2) TRAINING (3) ADEQUATE PROCEDURES (4) STRESS MANY HUMAN ERROR PROBABILITIES USED IN BRP RISK ASSESSMENT ~ HAD TO BE EVALUATED AS A FUNCTION OF TIME 4 4

4 I figure Ill. 2 Frobability of failure to enter containment to open valve VEC-1 vs. tiine prior to safety valve actuation (in hours) at which penthic need to open VEC-1 is reto,)nited l 1.0 .9 .6 .7 l .6 3 a = e-i O (o.5 c. E i o Cg .4 i .3 1 .2 4 I .1 12 + O n-5, n-4 n-3 n-2 n-1 n Time (hours) i r

s.o - 9 .1 - .6 E a .o.5 e .o O L. c .a L 3 -..3 u. t .2 .1 o a 4 a e 18 '2 Time (minutes) Figure 111-3 Estirnation of Failure to Initiate Liquid Poison Given an ATWS e 9 9 ._y ..m, ~.,.,.r.: -, s w w we -m e - ev

TAllLE 111-7 SU. WARY OF lluMAN ERROR PROBABILITIES FOR 111G ROCK POINT 1. Operator. performs an action for which there is no reason, i.e., closing a nonnally open valve necessary for flow or opening a normally closed circuit breaker necessary for power 1 x 10-4 2. Operator fails to open an M0V or start a pump when this action is necessary for successful system operation, 4 a procedure is available, and location is familiar 1 x 10-3 (low stress) 2 x 10-3 (moderate stress) -3. Operator fails to return valves or components to service following maintenance or test 1 x 10-3 ~ 4 Operator fails to open an MOV, start a pump or close a circuit breaker when location is unfamiliar or not frequently used 3 x 10-3 (low stress) 6 x 10-3 (moderate stress) S. Operator fails to close circuit breaker after LOSP when procedure is available but vague but sufficient indicators are available to alert him to action 1 x 10-2 4 6 Operator fails to shed loads and place demin.. inp in operation when procedure is vague following LOSP. 0.25 7 Operator fails to enter containnent and open fire water valve for emergency condenser makeup see Figure 111.2 8. Operator fails to place standby diesel generator in service before RDS operation takes place following a failure of the emergency diesel generator See Figure 111.5 9. Operator fails to shed loads anct place control rod drive pump on diesel generator with defined procedures following LOSP . l.

9 i ~ REPAIR Al4D RES10 RAT 10N TIMES. USED IN BIG ROCK POINT RISK ASSESSMENT-CUMULATIVE REPAIR DISTRIBUTIONS FORMULATED FROM BRP' f%INTENANCE DATA REPAIRDATAFOREMEilGENCYDIESELGENERATORUSEDIN EVENT TREE QUANTIFICATION i RESTORATION OF Of FSITE POWER USED IN EVENT TREE QUANTIFICATION 9 i 3 e9 pg p. y -,. 9 -n-- --,.y. --ge- .yg9r93.,- y .,fr-.* 9-,---%-py-.9.-p--. -.- y-4,r.,- y.y-- -wii,+we...--1 ye w-- = 9 ge w, ,-ys-+v, +e-.

w M r 1 W I h .n X o e e4 C g 4 .-e .a e. x L a x

.e t

= W c s-e- m r-sa 'C L. m O U .= = w L= b X Cs .-s t-X d w c a 4J x e4 t. r .C ?.. ". -a .s-sF Cz O-Ca e e i i f. 1 1 4 u O. ~ m. e. v. 4 !!19P40Jd i q ,y.

,,,-+y,

...,,,y-- -,q. -y,.. ..,m_,,#..y.,g ,y_m

i f r-l~ e 1 0 0 0 1 s ~ 0 0 1 a ta D r i X a pe X R p m g u P 0 1 e X h t X X o X t x X t X ) i X s F ru n X o o y h i ( tu X e b m i i r T ts i D X 3 lam r X o N g o L 7 1 1 1 eru g I i F 1 0 4 0 0 8 6 2

  • s av"Ea.

3 su 4 I. .l! il1 i i7!,. !/!a !li l il 4

i

-+ -n RAW DAT A iOR LOSS 0F OfI 511E I'0'.1ER AT lite BIG ROCK l'0INI SilE .. q DATE DURA 110N (MINUll5) l 46 KV LINE 138 XV LINE 5/2/69 <5 6/26/69 <5 8/4/69 <5 10/19/69 619 7/8/70 <5 9/11/70 <5 10/2/70 <5 11/19/70 <5 12/3/70 (1) 8/22//l 187 9/28//l <5 <5 1/25/72 119 20 6/1/12 61 11/3/72 <5 11/7/72 <5 5/21/74 <5 7/22/75 <5 12/8/75 <5 2/2/76 596 3/12/76 10 l 5/5/76 <5 10/21/76 <5 I 8/31/77 201 l 4/6/78 63 4/17/78 142 5/31/78 8 l 4/6/79 1329 [ 4/12/79 75 5/6/79 817 9/76/79 <5 10/10//9 94 (1) Nine power interruptions occurred on this date.

d I i X 1.0-l X 1 X 2 .8 X 6 X I \\ X i X i i i X .6 U 5 ~ t i $.4 i s. I G. t .2 i i 0 1 Id 10~3 10-2 10-I 1 10 TIwE TO RESTORE POWER, HRS i i ~ Garra Distribution F1L to the Restoration or Grisite Power Data. Figure !!!-8. 4

COMMON CAUSE FAILURE ANALYSIS: METHODOLOGY I. IDENTIFICATION OF COMMON MODE MECHANISM A. "LIKE" VS. "UNLIKE" COMPONENTS B. SUSCEPTIBILITY TO COMMON CAUSE c. OPPORTUNITY FOR COMMON CAUSE FAILURE 11. QUANTIFICATION A. "lIKE" COMPONENTS 1. CONSERVATIVE FILTER - 10% COUPLING FOR ALL REDUNDANT COMPONENTS, SUBSYSTEMS, OR SYSTEMS NUCLEAR IECHNOLOGY, VOL. 46, DECEMBER 1979 2. EXTERNAL EVENTS B. "UNLIKE" COMPONENTS 1. EXTERNAL EVENTS Ill. INTEGRATION INTO FAULT TREE AND EVENT TREE MODELS i O l \\ I

fw e4 i s Of COMMON CAUSE FAILURE MECHANISMS S INTERNAL - MANUFACTURER LOCATION TEST MAINTENANCE e EXTERNAL - FIRE EARTHOUAKE TORNADO FLOOD HIGH ENERGY LINE BREAK CONTROL ROOM HABITABILITY (SM0KE, RADIATION) AIRPLANE CRASH HUMAN ERROR e ,,,,r,- r = w T-"* --*W-- F-"W

DOMINANT COMON MODE FAILURE MECHANISMS 8 INTER M - MANUFACTURER LOCATION 8 EXTERNAL - FIRE HIGH ENERGY LINE BREAK

UNCERTAINTY ANALYSIS l. METHODOLOGY A. DOMINANT SEQUENCES ANALYZED 1. DETERMINE ALL COMPONENTS OR EVENTS INCLUDED'IN DOMINANT ACCIDENT SEQUENCES 2. DEVELOP A DOMINANT ACCIDENT SEQUENCE " FAULT TREE" B. ASSUMEPROBABILITYDENSITYFUNbTIONS(PDFS)FORALLEVENTS PROPAGATEEVENTUNbE,RTINTIESTOARRIVEATSEQUENCE C. (AND CDF) UNCERTAINTY D. PERFORM PDF SENSITIVITY STUDY II. PDFS A. ENGINEERING JUDGEMENT, BASED UPON DATA SOURCE (I.E. WASH-1400, EG&G, OR PLANT), AND/OR DATA POPULATION SIZE (SEVERAL TRIALS VS. FEW TRIALS) B. PDFS CHOSEN FOR: 1. COMPONENTS - GENERIC (LOG-NORMAL, POISSON) - PLANT SPECIFIC (BINOMIAL, POISSON, GAMMA, F) 2. HUMAN ERRORS OR ACTIONS - (NORMAL, LOG-NORMAL) 3. INITIATORS (LOG-NORMAL, POISSON, GAMMA) Ill. PROPAGATION A. USE METHOD OF MOMENT PROPAGATION 1. FIRST AND SECOND MOMENTS ABOUT THE ORIGIN 2. FASTER COMPUTER RUNNING IIME IHAN MONTE CARLO METHOD, l WITH SMALL ACCURACY LOSS l 3. 957. CHEBYCHEV CONFIDEMCE LIMIT IS CONSERVATIVE l 1

B. RESULTS ARE: MEAN, STANDARD DEVIATION, 95% CHEBYCHEV CONFIDENCE LIMIT (FOR NORMAL DISTRIBUTION)- l IV. SENSITIVITY STUDY A. CHOOSE DIFFERENT PDFS FOR 1. COMPONENTS OR EVENTS 2. INITIATORS B. RESULTS INDICATE SENSITIVITY TO INITIATOR PDFS t 0 'i e 4

w,----e,-

e re.v., --g-

PDF EXAMPLES INITIATORS: LOCA LOG-NORMAL, EF = 20 LOSS OF FEEDWATER LOG-NORMAL, EF = 10 FIRES GiMMA HUMAN ERRORS: FAILURE TO TRANSFER LOG-NORMAL', EF = 3 DEMIN. PUMP TO 2B BUS INADVERTANT OPENING / LOG-NORMAL, EF = 10 CLOSING OF VALVE FAILURE TO ENTER NORMAL CONTAINMENT FOLLOWING DEMIN. PUMP FAILURE COMPONENTS: GENERIC - MOVS ' LOG-NORMAL, EF = 3 SOLENOIDS LOG-NORMAL, EF = 3 CHECK VALVES POISSON

a PDF EXAMPLES.(CONTINUED) COMPONENTS: (CONTINUED) PIANTSPECIFIC-PUMPFAILSTOSTART BINOMIAL ~ PUMP FAILS TO RUN POISSON VALVE FAILS TO REMAIN GAMMA OPEN MSIV FAILS TO'CLOSE F o b l l 0 i 1 ~~ - - m, (

  • -,l - -,._,
  • * ~ ~.::;.
'- L;;:: ~~;;;:3

= _ - RESULTS OF BRP PRA UNCERTAINTY ANALYSIS bFfb!kE)}MT EF(g@) CASE

  • MEAN STANDARD DEVIATION A

9.85x10-4/vR 4.4x10-3 1.5x10-2 15 BASE CASE 3.1x10-3 1.1x10-2 11 B 1.8x10-3 6.6x10-3 6 C 2.2x10-3 7.9x10-3 8 D 1.5x10-2 4.9x10-2 49 E 1.6x10-2 5.3x10-2 53 F

  • CASE A = L0s-NORMAL INITIATING FREQUENCIES (IFS) EXCEPT FIRE (GAMMA), EFS = 10, 20.

MIXED SPECTRUM FOR COMPONENT DISTRIBUTIONS B = CASE A IFS, LOG-NORMAL COMPONENTS C = L0s-NORMAL IFS (ExCEPT FIRE), EFS = 3, 10. CASE B COMPONENTS D = LOG-NORMAL IFS (EXCEPT FIRE), LOCA EFS = 20, OTHERS ESTIMATED WITH .95 CASE B COMPONENTS 2 E = X IFS (ExCEPT FIRE), CASE B COMPONENTS F = CASE E IFS, CASE A COMPONENTS -}}

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