ML20092L044

Forwards Comments on Low Power Shutdown Accident Frequencies Program/Senior Consulting Group 900314-15 Meeting

ML20092L044
Person / Time
Site:	Surry, Grand Gulf
Issue date:	04/05/1990
From:	Marie Pohida Office of Nuclear Reactor Regulation
To:	Rasmey Robinson NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML20092K912	List: ML20092K908 ML20092L019 ML20092L044
References
FOIA-91-267 NUDOCS 9202260131
Download: ML20092L044 (80)

Text

{{#Wiki_filter:e /q)ril 5,1990 lcII Ir.it RichuT1 C. Rchisrxx1 httubilistic RirJ: Awamt Rraldi Division of Systrec Peccarrh offico of liaclear Raplatory Piccattti T104: 14ario A. Ithida RirJ: Iq:plications Brardt Division of Radiation Pmtretion ard Dantjency heputdress offico of liuclear Ikoctor Ikvalation SUlUlrT: Inf.TWDt NID SIUIUCM!1 AOCIDIRT ITdTNDiCID3 IYCGWi SDi10R CttiSULTDM GOJP (SCU) PJIrDD,- WJOf 14 N!D 15,1990 L i theloccd are iny ccennts regan111r3 the alxne referetrxd rmtira tlnt a2n to be 'ifrctporatcd in the SCG group cuments. If you rhould lave any qJostions picano contact to at X21063. i .1 CC. RhuTett AEl-Dusioni l '? r I L 9202260131 910726 .PDR.FOIA z. REDDA91-267 PDR V)UlfA

i-i a:sments 1. 7e ccqute caTonent mintemrm umvailability, the tire botvten - placirg the safety tags on an1 off tho equilmmt rust in ured. %is tiro inte.tval my to inach greater than the actini (guitrent mroir tire, no bware of using the plant mintenvm Icg trxsks.

2. 'Using liPRD3 to ocqmte ruintenuco umvailabilitics should tu terrottni with caution duo to the inxnsistency of utility rerotting aid the q>xial inture of the chutdwn operaticns.

3. Daily plant events wrortable un3er 10C3RSO.72 ("50.72 cvents") rhould lxt Irviwcd to evaluate recent abrcIm1 occurreuxs at nhutdcun. Short tern: losrcs of thutdcun ecolin) aru frcquently mported as 50.72 events. We i AIDD Core Damgo Pmcursor Sttdies chould also in consulted as a "nanity check". A rS[52i 'alpituch to this shutdcun im is preferred ircause of the 4. campickity,of th9 prubicm. - mio fist rharo thould concentrate on fislDg/aocideht seemrlos ard identifyirg significant humn Sensitivity / studico uhould be uxd to identify thcco humn erro m. actions that nhenld my itq.liro refincd rrdels. 5. To coquta or verify the tire duration of each shutdcm rhase within each rncde, the data into frtra which the AIDD " Grey reck" is ocq 11cd can to usad. Wo tire dumtion of these rhutdcun rhases is present<d in this data inco. We uugo t Out this informtlan my in plant crocific. ' mis mfemnao can to uncd to assccs tho variability of f.hase duration betwoca Surry and Grand Gulf as caqaltd to other plants.

6. - Ibcause humn ermts will ecostituta a lattje roztlon of the coru ctwage soernrios, Event Scquerce Diagram (130c) kould le a useful tcol to present the individual accident ocquerces. ISD1 w.nid be cqwially useful-for the Itactor crerators sho do rot have im c>qnrierce to be able to;prcr,'ide infut durirg the !!RA ptroccs.

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SUMMARY

OF GRAND GULF FULL POWER PRA i ~ DONNIE W. WHITEHEAD SANDIA NATIONAL LABORATORIES LOW POWER AND SHUTDOWN ACCIDENT FREQUENCIES PROGRAM-SENIOR CONSULTING GROUP MEETING MARCH 14 AND 15, 1990 ALBUQUERQUE, NEW MEXICO R l

.b ') INITIATING EVENTS [ T1: LOSS OF 0FFSITE POWER TRANSIENT 0.11 T2: TRANSIENTS WITH LOSS OF POWER CONVERSION 1.62 SYSTEM (PCS) T3A: TRANSIENTS WITH PCS INITIALLY AVAILABLE 4.51 l T3B: TRANSIENTS INVOLVING LOSS OF FEEDWATER (LOFW) 0.76 BUT WITH THE STEAM SIDE OF THE PCS INITIALLY AVAILABLE T3C: TRANSIENTS CAUSED BY AN INADVERTENT OPEN 0.14 l t RELIEF VALVE (IORV) ON THE REACTOR VESSEL i TIAS: TRANSIENT CAUSED BY LOSS OF INSTRUMENT AIR 8.1E-4

9 ~ re 4 . SYSTEMS MODELED HPCS HIGH PRESSURE CORE SPRAY RCIC-REACTOR' CORE ISOLATION COOLING CRD CONTROL ROD DRIVE SLC STANDBY LIQUID CONTROL ADS AUTOMATIC DEPRESSURIZATION CDS CONDENSATE LPCS LOW PRESSURE CORE SPRAY LPCI LOW PRESSURE COOLANT INJECTION SSW X-TIE STANDBY SERVICE WATER CROSSTIE u

L 7 -e SYSTEMS' MODELED (CONCLUDED) EVS EMERGENCY VENTILATING SYSTEM IAS INSTRUMENT AIR SYSTEM RPS REACTOR PROTECTION SYSTEM SPMU SUPPRESSION POOL MAKEUP i SBGT STANDBY GAS TREATMENT i CI CONTAINMENT ISOLATION HYDROGEN IGNITERS HzI

r. ~ l '- q EVENT TREES LARGE LOSS OF COOLANT ACCIDENT (LOCA) INTERMEDIATE LOCA SMALL LOCA SMALL-SHALL LOCA LOSS OF 0FFSITE POWER (LOSP) STATION BLACK 0UT LOSS OF POWER CONVERSION SYSTEM

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4 N ?O b?o b cry ro 4% y 907. N ) .+ : Nf STAfiON BLACKOUT (590) - rAl(UR( QF Cl[$CL C(N(PATCR$ (QGs) AND REACTOR CORE iSOLAT10N C00DNG (RCIC) TO START AND RUN S80 - FA: LURE OF CCs AND RCIC TO START AND RUN AND ONE STUCK-CAEN REUEF VALNE (50Rv)- ANiiciRATED TRANSIENT WITHOUT SCRAM (ATWS) 2 580 - FAILURE OF OCs AND BATTERY DERLETION - FAILURE OF cgs AND TWO SORVs - F:, LURE OF OCs AND FIREWATER OTHER TRANSlENT Figure 1-1. Grand Gulf Core Damage Frequency Types. 4 i 14- ~ ',P

x. m s v s u y t nr 7 7 4 4-1 iMOST:SIGNIFICANT RISKLREDUCTION EVENTS s i r* - ' FAILURE'TOTREC0VER OFFSITE POWER e i FAILURE 0F THETRCICLSYSTEM TURBINE DRIVEN PUMP TO RUN 4 j: 7 I FAILURETTO RECOVER THE DIESEL GENEEATORS (DGs)

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[ o l j1 FAILURE OF THE DGs TO START 1 l -.i COMMON'CAUSE: FAILURE'0F_THE BATTERIES i .i i i COMMON-CAUSE FAILURE TO START'0F TWO DGs i l 1 1 i. + s .,+,,c. %m

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THE UNCERTAINTY IN THE:

~ CORELDAMAGE FREQUENCY l .l 4 . FAILURE:0F THE DGsLTOLRUN N t -FAILURE 0F THE DGs'TO START 1 l l COMMON'.CAUSE' FAILURE.0F THE.DGs i' i f I t 1 j' k i ~.. - .... - n

A METHODOLOGY FOR BWRLLOW-POWERL

AND SHUTDOWN: PROGRAM DONNIE:W. WHITEHEAD.

1 -SANDIA NATIONAL-LABORATORIES LOW POWER AND SHUTDOWN ACCIDENT FREQUENCIES l PROGRAM. SENIOR CONSULTING GROUP MEETING t 4 j MARCH 14 AND 15, 1990 i 4 l ALBUQUERQUE, NEW MEXICO l i: l -[ l

3 1.. LOBJECTIVES N 9 t* ?T0; ASSESS' THE FREQUENCIES)l0F SEVERE ACCIDENTS INITIATED DURINGEPLANT:'0PERATIONAL MODES OTHER THAN FULLt POWER = OPERATION ' FOR. A COMMERCIAL BWR- . ' GRAND GULF UNIT 1 (STUDIED IN-NUREG-1150) TO. COMPARE'RESULTS.OF THIS STUDY: i-L - ESTIMATED CORE DAMAGE FREQUENCIES - IMPORTANT ACCIDENT SEQUENCES i - OTHER. QUALITATIVE AND. QUANTITATIVE RESULTS WITH.THOSE ACCIDENTS INITIATED DURING FULL POWER .0PERATION -(AS ASSESSED IN NUREG-1150) T0' DEMONSTRATE METiiODOLOGIES FOR EVALUATING FREQUENCIES OF ' ACCIDENTS INITIATED DURING PLANT OPERATING MODES 10THER THAN FULL POWER

~ 3. TECHNICAL APPR0ACH CONSISTS OF 13 STEPS " i

1. IDENTIFICATION OF PLANT OPERATIONAL MODES (P0Ms)

2. DETERMINE APPLICABLE. INITIATING EVENTS (IEs) FOR EACH OM

3. DETERMINE APPLICABLE SYSTEMS AND SUCCESS CRITERIA FOR EACH P0M AND IE

4. DEVELOP EVENT TREES FOR EACH INITIATING EVENT GROUP

5. CONSTRUCT SYSTEM-FAULT TREES

6. DEVELOP NON-FULL-POWER DATA BASE ( % *<)

7. CONDUCT DEPENDENT FAILURE AND HUMAN RELIABILITY ANALYSES-

_i,

9 .N 1. IDENTIFICATION 0F PLANT OPERATIONAL MODES (P0Msl e MODE 1 - POWER OPERATION (LOW POWER < 15%) MODE 2 - STARTUP r MODE 3 - HOT SHUTDOWN MODE 4 - COLD SHUTDOWN MODE 5 - REFUELING

- o 3. DETERMINE APPLICABLE SYSTEMS AND SUCCESS CRITERIA FOR EACH P0M SYSTEM OPERABILITY DEPENDENT ON INITIAL CONDITIONS OF PLANT (E.G., RCS PRESSURE / TEMPERATURE, AVAILABILITY OF STEAM, ETC.) CONSIDER SAFETY SYSTEMS AS WELL AS NONSAFETY SYSTEMS THERMAL HYDRAULIC CONSIDERATIONS Dj ; g g- ]r f g ,~ " "" -9 1

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E 5. CONSTRUC,T, SY,STEMLF, AULTF, TREES ~ o i G ib MODIFY: EXISTING -FAULT TREES (FTs) BASED ON': LOW = ' POWER / SHUTDOWN SUCCESS CRITERIA' CONSTRUCTTNEW FTs: FOR SYSTEMS NOT MODELED1 IN GRAND t GU LF ! ANALYSIS 4-h -YIjf' -"""'""-"'" '" * "" 1 FTs SHOULDLINCLUDELALL MAJOR COMPONENTS 4 i .FAILURELMODES AND. EVENTS SHOULD NOT BE. GROUPED i tiXTERNAL: EVENTS !: INTERFACE - -i IMPORTANCE CALCULATIONS DEPENDENCIES AND CORRELAT7.0NS l i g-ga9 MORE DETAIL.SHOULD INCREASE DEFENSIBILITY AND j " SCRUTABILITY: l 4 t

7

.

CONDUCT DEPENDENT FAILURE AND HUMAN

RELIABILITY ANALYSES DEPENDENT FAILURE ' ANALYSIS EXAMINES 3 CATEGORIES OF

' DEPENDENT FAILURES DIRECT FUNCTIONAL DEPENDENCIES INCORPORATED IN THE FAULT TREE MODELS AND INCLUDE SUCH THINGS AS EFFECTS OF INITIATING EVENTS,- -SUPPORT SYSTEM DEPENDENCIES, AND -SHARED-EQUIPMENT DEPENDENCIES COMMON CAUSE FAILURES LIMITED TO WITHIN. SYSTEM AND "LIKE" COMPONENTS BETA FACTOR-APPROACH I

8; QUANTIFY ACCIDENT SEQUENCES SOLVE SYSTEM FAULT TREES USING IRRAS (OR OTHER SOFTWARE) COMBINE SYSTEM FAULT TREES AS PERSCRIBED BY THE ACCIDENT SEQUENCE EVENT TREES S0LVE ACCIDENT SEQUENCES BY GENERATING CUT SETS AND POINT ES(IMATES :(SCREENING QUANTIFICATION) TURNCATE SEQUENCES BELOW 10-8 (I IDENTIFY.AND ADD REC 0VERY ACTIONS TO CUT SETS REEVALUATE PROBABILITIES FOR DOMINANT FAILURE EVENTS o RECALCULATE POINT ESTIMATES FOR THE SURVIVING ACCIDENT SEQUENCES (FINAL QUAM 1IFICATION)

s i.: i-10. PERFORM A FIRE ANALYSIS i IDENTIFY FIRE ZONESI DETERMINE INITIATING' EVENT FREQUENCIES FOR FIRES 4 SCREEN FIRE ZONES VERIFYLIMPORTANTLFIRElZONES ' DETERMINE RESPG.'SE OF EQUIPMENT IN REMAINING FIRE l ZONES FINALIZE FIRE SEQUENCES .i 7

_4 'es*..' j 'h ..N' n. 12:. CONDUCTLAN UNCERTAINTYEAN.ALYSIS i.

LHS/TEMAC USED'T0 CALCULATE UNCERTAINTIES

.j y FLEXIBLE: UNCERTAINTY DISTRIBUTIONS' .C0RRELATION OF EVENTS POSSIBLE GENERATESLIMPORTANCE CALCULATIONS: 4 UNCERTAINTY CALCULATIONS'FOR: .EACH' ACCIDENT SEQUENCE: SURVIVING SCREENING l 'EACH PLANT DAMAGE' STATE J l f .-TOTAL CORE MELT MODEL I h I i .... ~

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SUMMARY

L a r -METHODOLOGIES USED.WILL BE' CONSISTENT WITH CURRENT a -LEVEL.1 TECH 0NOLOGY -i PROVIDE ESITIMATE OF.C0RE DAMAGE FREQUENCY AS A ' RESULT 10FL INTERNAL EVENTS FOR' MODES :0F OPERATION l OTHER THAN FULL'POWERT 'CANiBE COMPARED'.WITH1NUREG-4550 e' .j { F L 1

y c PLANT OPERATIONAL MODES CHARACTERIZATION BWR LOW POWER AND SHUTDOWN iACCIDENT FREQUENCIES PROJECT 1 i TANIA M. HAKE .SANDIA NATIONAL LABORATORIES 2 SENIOR CONSULTING GROUP MEETING MARCH 14-15, 1990

! j Y e s tc L e E l u C e f e I E t R n R 't f E I M o 3. G R 0 P e n O 4 s n t 1 o i N i m a W 5 l t e y O A ut t ma E T f p f D f U e p N R o S 5 4 s n kco lr L'- N e W t f = O it n D h I M 0 0 R u D t g 1 n I n 4 s U N 4 i l "D e y A T L ut p O f C e 5 y* R a 4I 4 s I-w C ' I'I io' m'o C/5 NW f O N E O D O W M 0 C t M u O T m G ' S U s' h N$ I i T fa om 8 tP O 3 F IL C ) ) D t s N R S in t A T n C D o' .i e N p t s A E R ge' n TS U i s a T s gr 1 A I pp it 5 T s 1 O R ' ir pt

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M E P P / s U P s 4( 5 e E 6 4 c p I U T 0 0 m 4 T A 1 1 e R YI ( 1 A 5 N T A S p p 2 n) 1 ) ) ) t s e er ee n t r n u ww i t o o e s ep N E p i s b O R t s e t U ge-n r I n I T is a p u5 A A s gr R R p pi t gL R 1 ( E i s iS ' 7. t o-P n P r pt sM O u M t pp( 0 R t R 4( 5 e i 6 4 c0 h1 c E 0 0 x0 W f t t 1 1 e8 iu O u ( wo P A Sb ( a p p p 1 f i n E ME R t a l l U ( i'W A oL f T k i E u0 E A 0 C t A R I t mR M Sl C R M5 5 E 1 aR E A i A ( l R g W H 8 El R P i f E aO A D$ E M i g a C 1t P G b P O O V [ Sp M P A 1 s 1 l l 4 A! i I ' + I ?J.

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~ ~ uT ~: so ~ .TECHNICAESPECIFICATIONLOPERABILITYJREQUIREMENTS 7: i a L CHARACTERIZE-POMs INLTERMS OFLSYSTEM OPERABILITY' L CONSIDERATIONiOF SAFETY SYSTEMSLAS WELL'AS "NON-- t i. SAFETY" SYSTEMS:FOR.WHICHLSAFETY CREDIT MAY BE'GIVEN l I l c* DEVELOPMENT 0F TECHNICAL 4 SPECIFICATION' SYSTEM-1 OPERABILITY REQUIREMENTSLTABLE, TO-INCLUDE: 1 i. l l SYSTEMS 1 INCLUDED:IN: FULL-POWER STUDY. F gm mu l SYSTEMS-NOT GONSIDERED :IN FULL-POWER STUDY BUT l i WHICH MAY.:BELRELEVANT FOR THE SHUTDOWN MODES l 1 4 i l ) . DETERMINE TIMEiDURING WHICH PLANT IS OUTSIDE a b TECHNICAL SPECIFICATIONS S03THAT THIS' TABLE CAN BE. USED TO CHARACTERIZE! ACTUAL SYSTEM AVAILABILITY I i 5. t a ? d i I

1 .t I GRAND. GULF TECHNICAL SPECIFICATION SYSTEM OPERABILITY REQUIR TABLE OF CONTENTS, cont'd PAGE /4 mu SYSTEMS NOT INCLUDED IN TULL-POWER STUDY 9-l Saf ety/Relie f Valves (SRVs) 9 HPCS Service Water System 9 Ultimate' Heat sink 10 Main Steam Line Isolation valves-(MSIVs) 10 Containment and;Dryvell Icolation valves 10 MSIV. Leakage Control System 10 reedvater Leakage Control System 31 - Primary Containment Integrity 11 Drywell Integrity 11 Containment: Air Locks (and.Drywell Air Locks) 11 - Secondary-containment Integrity 12 Suppr e s s i on -- Pool 12 containment Hydrogen Recombiner. Systems 12 - Combustible Gas Control-System 13 Control Room Emergency Tiltration System Recirculation Loops ' ~ 13 13 Jet' Pumps I 14 Idle Recirculation. Loop Startup RCS Leakage Detection: System ~ 14 15

Onsite-Power Distribution Systems 15 RPS Electric Power-Monitoring 16 Isolation-Actuation-Instrumentation 16 ECCS Ac'tuation Instrupentation 16 RCIC System' Actuation Instrumentation 16-Control' Rod Block' Instrumentation 16 Remote Shutdown Instrumentation and Controls 16-Accident' Monitoring Instrumentation 17 Plant SystemsLActuation Instrumentation 17 Control Rods-18

-Control Rod-Scram Accumulators 18 Control Rod. Drive Coupling 19 Control-Rod Position Indication 19 Control. Rod Drive Housing Support 19 Red Withdrawal 19 Rod Pattern Control Sys'.en m 4-v -w ,w =

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' PRELIMINARY SYSTEMlAND SUCCESSECRITERIA LIST p i: DONNIE W. WHITEHEAD L SANDIA NATIONAL LABORATORIES ? a LOW POWER AND SHUTDOWN ACCIDENT FREQUENCIES I PROGRAM SENIOR CONSULTING GROUP' MEETING [ .. I t

?

MARCH 14 AND 15, 1990 i I ALBUQUERQUE, NEW MEXICO j 4 t i L 4. f

' ~ PRELIMINARY SYSTEM LIST

(CONCLUDED 1 MODES r

SYSTEM 1 2-3 4 5 SSW X X X -X X EVS/EHV 'X X -X X .X IAS': X ?- ? -? ? RPS X X ? ? ? SPMU X X-X' ? ? SBGT X. 'X X X X' CI X X X X X HzI X X ? ? ? ADHRS ? ? ? X X RWCU ? X X X X SFPCCU ? ? ? X X CCW. NA .? ? ? ?- TBCW NA ? ? ? ? 4 PSW NA ? ? ? ? CInc W NA ? ? ? ? l 1

=; x py i ~ f I SUCCESS CRITERIA MODE:2? " FULL ' POWER-INITIATORS" WILL HAVE: THE SAME SUCCESS [ -CRITERIA LOCA INITIATORSL (" MAINTENANCE INDUCED", " RECOVERABLE-j. l DIVERSION",'AND " PIPING OR COMPONENT FAILURE IN '0PERATING' SYSTEM") .WILL HAVE' '.THE SAME SUCCESS

CRITERIA AS THE APPROPRIATE SIZED " FULL = POWER LOCA" i

l. EXCEPT:THE'(....)1 SYSTEMS:MAY BE FAILED AND.NO LONGER i CONSIDERED-IN THE: SUCCESS CRITERIA LIST I DECAY HEAT' REMOVAL -(DHR) INITIATORS WILL HAVE THE [ SAME LTYPE. OF SUCCESS CRITERIA AS " FULL POWER p TRANSIENT INITIATORS" EXCEPT THE SYSTEM INVOLVED IN l THE :DHR : INITIATOR MAYS BE FAILED AND NO LONGER CONSIDERED IN THE~ SUCCESS CRITERIA'. LIST h

INITIATING EVENT IDENTIFICATION AND FREQUENCIESL BWR LOW : POWER-AND SHUTDOWN -ACCIDENT FREQUENCIES PROJECT TANIA M.. HAKE SANDIA NATIONAL LABORATORIES SENIOR CONSULTING GROUP MEETING-MARCH 14-15, 1990 5 i

+. FIVE GROUPS 0F INITIATING EVENTS CONSIDERED: i i' 1. TRANSIENT EVENTS 2. LOSS OF COOLANT ACCIDENT EVENTS j 3. DECAY HEAT REMOVAL CHALLENGE INITIATORS j i 4. SPECIAL EVENTS: 4 i 5. HAZARDS EVENTS J N . ~

+_ INITIATING EYENTS E 2. LOCAs M 0 D'E LOW HOT. COLD LOCA CATEGORY POWER STARTUP SHUTDOWN-SHUTDOWN REFUELING-A LARGE LOCA X X X-X X S1 INTERMEDIATE LOCA X X X X X S2 SMALL LOCA' l X X X X X - S3 SMALL-SMALL LOCA X X X

X X

4 V INTERFACING SYSTEM LOCA (FAILURE OF HIGH-TO-LOW PRESSURE INTERFACE) N/A TBD X* X* X* R VESSEL RUPTURE N/A N/A N/A N/A N/A H DIVERSION OF VESSEL INVENTORY'THROUGH CONNECTED SYSTEM: -RHRS X X X X HPCS X X X X LPCS X X X X J LOCA IN OPERATING CONNECTED SYSTEM: RHRS TBD X X X RWCU X X X X K MAINTENANCE-INDUCED LOCA X X X X 4 X* = EVENTS CONSIDERED UNDER ANOTHER LOCA EVENT CATEGORY ("H", "J", OR "K") r-

a i. POSSIBLE SCREENING CRITERIA FOR MAINTENANCE-INDUCED LOCAs: " DOUBLE FAILURE" REQUIRED FOR CONSIDERATION <. MUST REVIEW MAINTENANCE PROCEDURES IN DETAIL TO IDENTIFY PROBLEMS As:3 SCREEN SEGMENTS OF SYSTEMS SCREEN ON MULTIPLE DETECTION CAPABILITY SCREEN ON NUMBER OF POSSIBILITIES TO MITIGATE SEQUENCE 1. l m

s, ~ INITIATING EYENTS 4. SPECIAL EVENTS M0DE LOW HOT COLD 'l SPECIAL EVENTS POWER STARTUF SHUTDOWN SHUTDOWN REFUELING T4 CRITICALITY EVENTS: ROD WITHDRAWAL ERROR X X X X X REFUELING ACCIDENT X T5' LOSS OF ANY SERVICE WATER SYSTEM: COMPONENT COOLING WATER X X X X STANDBY-SERVICE WATER X X X X TBCW X X X X PLANT SERVICE WATER X X X X CIRCULATING WATER SYS-X X X TIAS LOSS OF INSTRUMENT AIR SYSTEM "T2 X X X X TORV INADVERTANT OPEN RELIEF VALVE (SHUTDOWN) (T3C) (T3C) X TIOP INADVERTANT OVERPRESSURIZATION EVENT TBD TBD TBD i c

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AREAS-REQUIRING FURTHER INVESTIGATION

• LSTEPS'IN STARTUP OR SHUTDDWN PROCEDURES'WHICH COULD' SERVE AS INITIATING EVENTS 3
• WHETHER VESSEL RUPTURE FREQUENCY IS SIGNIFICANTLY INCREASED FOR THE STARTUP. MODE
• COLD OVERPRESSURIZATION OF VESSEL IN SHUTDOWN
• EXTENT OF MAINTENANCE ACTIVITIES IN STARTUP DOSSIBILITY FOR MAINTENANCE-INDUCED TRANSIENTS

.VARIOUS LOCAs INVOLVING INTERFACING SYSTEMS' IS FRACTION OF TIME RHRS-OPERATES IN STARTUP SIGNIFICANT? l - LOCA IN OPERATING RHRS DURING STARTUP - LOSS OF RHR-SDC FOR DHR IN STARTUP

• HOW OFTEN RCIC IS USED FOR NORMAL SHUTDOWN EVOLUTION

- LOSS OF RCIC AS DHR SYSTEM (STARTUP/ HOT SHUTDOWN) 4

• WHETHER TRANSIENT EVENTS CAUSING SCRAM ON HIGH NEUTRON FLUX ARE APPLICABLE OR HAVE-SAME EFFECT AT LOW POWER ("T3A")

(CONTINUED) x

c. LOSP FREQUENCY CALCULATION ~ BACKGROUND + THREE CAUSES FOR LOSF EVENTS:

1. PLANT-CENTERED (SWITCHYARD)

2. GRID-RELATED

3. SEVERE WEATHER EVENT CATEGORIES:

IA: NO 0FF-SITE POWER AVAILABLE FOR LONGER THAN 30 MIN. (WITH UNIT TRIP) IB: NO OFF-SITE POWER AVAILABLE FOR LESS THAN 30 MIN. (WITH UNIT TRIP) II: LOSS OF BACKUP OFF-SITE POWER, BUT UNIT REMAINS CONNECTED TO NORMAL 0FF-SITE POWER (N0 UNIT TRIP) III: LOSS OF NORMAL OFF-SITE POWER, BUT BACKUP 0FF-SITE POWER AVAILABLE THROUGH SWITCHING (N0 UNIT TRIP) IV: LOSP DURING A COLD SHUTDOWN BECAUSE OF SPECIAL MAINTENANCE CONDITIONS UNIQUE TO SHUTDOWN

m i l LOSP APPROACH FOR LOW POWER / SHUTDOWN l l LOW POWER /STARTUP: I l

• USE ALL EVENTS INCLUDED FOR 1150 I

(PLUS ADDITIONAL EVENTS OCCURRING IN LATTER HALF OF 1987)

• APPLY ALL EVENTS OVER PLANT CALENDAR YEARS ~

~ ~

• USE SAME COMPUTER PROGRAM AS IN 1150 TO GENERATE LHS SAMPLES

- THE FREQUENCY FOR EACH MODE IS FOUND BY MULTIPLYING THE OVERALL FREQUENCY BY THE FRACTION OF TIME SPENT IN A PARTICULAR MODE. L f

I. LOSP FREQUENCY EQUATIONS, MODES 1 THROUGH 5: LOSP1 = F1 * {(PC + G&W)/(SITE YRS)} I = 1,2 LOSP1 = F1 * {(PC + G&W)/ (SITE YRS) + (CAT IV)/(Tso)} I = 3,4,5 WHERE: = FREQUENCY OF LOSP DURING MODE I UYR] LOSP1 = FRACTION OF TIME SPENT IN MODE I F1 PC = NUMBER OF PLAHT-CENTERED EVENTS OCCURRING THRU 1987 G&W = NUMBER OF GRID & WEATHER EVENTS OCCURRING THRU 1987 SITE YRS = CUMULATIVE PLANT CALENDAR YEARS CAT IV = NUMBER OF CATEGORY IV EVENTS OCCURRING THRU 1987 Tso = PLANT SHUTDOWN YEARS CALCULATED BY CORRECTING THE SITE YEARS BY THE COMPLEMENT OF THE PLANT AVAILABILITY FACTOR

165P EVE 2TS vet 1US MSC (All Yeer3'Through 1987) wofE: 1st cote centaine einste-telt alte evente; 2nd cottsm containe suttfpts telt site eventa (counted se 0.5 in totets) - formet: plent;dete (see etteched plant stereviettens) e.g. grunewick Unit 1 in 04/83 = 981;0483 3 >15% POWER

• 15% POWER-ST Atita.

'sf[ WCT sNUTDOWN' COLD SMUTDetAs )' atrUELimG LantaB0hAl MODE '-

1. CATEGORY IV sv;0TT8 Aut;0978 FA;0977 cs;1177 AN2;wd8 Ps;0186 mlt;0985 FA;1083 582;0375 40;0481 501:1180 Ps1;0474 f f;0379 541;0683 08:1079..CC1:078T w;0884 Pv1;1085

.MI2:0985 fv;0$a3 IP2:0TTT TP3:0585 Pe2:0474 ff;10TS :IP3:1184-Da;1165 CC2;0787 Ct;0371 OC1;01T4 FC;0375.. MC1;0987 FC;0276-OR1:0885 m;CJ76 PS1:0271 us;0884.:MC2:0967 FC;08T7 DR2:0885 m;1173 P11;0780 no;0684. Ps2:1084 CI;1073 lP2:0TT2 PA;0971~SL1:0578 - 0V;1183 OC2;0585 Mm;0468 IP2;0680 PL;1186 TP4:0577 PA;0184 SA1:0684 m ;0769 MC1:0884 PL;1187 50;0673. TP3;0485 W;0772 Ml2;0776 Ts;1165 $0:0480 r NN;0174.M11:0876 YR;0584 I DY;0973 MI2:0876 ? PA;0977 0C2:0174 .PA;078T Pv1;1085 PL;0577 PI2:0780 PL;0278.cC1;11TT R0;0186. 0C2:1177 QC2:0682 $01;1185 SL1:0577 $ul;0784 SN2:0784 TP3;0473 TP3:0473 TP3:0474 'iP4;o474 TP3;0674 1P4:0674 TP3:0577 iP3;0284 TP4:0284 TP3;02S4 j TP4;0284 [ TP4:0585 y j 10TAL.16 TTL=16.5 101AL.1 TofAL.0 10fAL.2 : TOTAL.1 101AL.1 TOTAL-1 TOTAL.9 10TL.3.5 10TAL.1 TOTAL.1 TOTAL.0 total.1 TOTAL. IT CAT IV (10f/2). (10t/2) (tof/2) (10T/2) (TOT /2) (10T/2) l (107/2) i TOTAL NON CAT.'IV EvtuTS CoustDERED: $3 (+ 1 tmh 0hAf seCDE EVENT) I" Pf acinf ACE (*15% + <15%

• STARTUP/ MOT sisT). 36.5/53 68.8%

i PERCENTAGE (tsof 5/D + COLD S/D + REFUEL) = 16.5/53 = 30.5% - I i I

4 L FURTHER LOSP WORK REQUIRED: t. k CALCULATION OF REC 0VERY TIME--CATEGORY IV EVENTS-ADD.0NE CATEGORY IV EVENT MISSED (NOT INCLUDED IN NSAC-118)- EXTEND. SCOPE THROUGH 1988??'r -(1987 USED IN ORDER TO BE COMPATIBLE WITH NUREG-1150 RESULTS) 1 1

TRANSIENT EVEbTS BASED ON FREQUENCIES AND DISTRIBUTIONS USED FOR 1150. ALL FREQUENCIES ARE FOUND BY SUMMING THE FREQUENCIES FOR THE INDIVIDUAL EPRI EVENTS COMPRISING A TRANSIENT CATEGORY. ALL FREQUENCIES ARE TO BE CORRECTED BY THE FRACTION OF TIME SPENT IN A PARTICULAR MODE.

ne LLOSSLOFLC00LANT / ACCIDENTS l e L i MAINTENANCE-INDUCEDHLOCAsLAND DIVERSION 0F VESSEL' INVENTORY l THROUGH HPCSLAND LPCS REQUIRE FURTHER4 INVESTIGATION BEFORE: f SCREENING FREQUENCIES CANLBE: ASSIGNED. t SEVERAL FREQUENCIES'HAVE BEEN CALCULATED USING A BINOMIAL l c. > COMPUTER TOLESTIMATE THE. MEDIAN AND:ERROP. FACTOR, GIVEN'X NUMBER-0F ' EVENTS - OVER. A L CERTAIN AMOUNT OF TIME. A LOGNORMAL-F DISTRIBUTION HAS BEEN ASSIGNED-FOR ALL. EVENTS IN.0RDER TO-CALCULATE"AlMEAN. i. 0 t t i x.

~. DECAY HEAT REMOVAL CHALLENGE INI"IATORS SOME OF THE DHR CHALLENGE EVENTS HAVE BORROWED EPRI TRANSIENT EVENTS FROM THE 1150 TRANSIENT CATEGORIES. IN THESE CASES, THE UPPER B0UND ESTIMATE REPORTED IN NUREG/CR-3862 FOR THE EPRI EVENT HAS BEEN USED, AND AN ERROR FACTOR OF FIVE ASSIGNED. FURTHER INFORMATION IS REQUIRED BEFORE SCREENING FREQUENCIES FOR EVENTS Elc, ISOLATION FROM OPERATING RWCU, E2C, LOSS OF OPERATING RWCU, AND E2E, LOSS OF OPERATING RCIC SYSTEM CAN BE ASSIGNED.

I L SPECIAL INITIATORS. THE SCREENING FREQUENCY FOR EVENT T4$, REFUELING ACCIDENT, IS NOT MEANT TO BE CORRECTED BY ANY. FACTORS, BECAUSE THIS EVENT HAS BEEN CALCULATED USING DATA UNIQUE TO REFUELING. SCREENING FREQUENCIES FOR ALL THE T5 EVENTS, LOSS OF ANY SERVICE WATER SYSTEi4,. ARE TO BE DETERMINED VIA FAULT TREES FOR THE VARIOUS SYSTEMS. THE INADVERTANT OVERPRESSURIZ TION EVENT, TIOP, REQUIRES FURTHER INVESTIGATION BEFORE SCREENING FREQUENCIES CAN BE ASSIGNED TO THIS EVENT.

PRELIMINARY EVENT TREES DONNIE W. WHITEHEAD. I

SANDIA.. NATIONAL. LABORATORIES LOW POWER AND SHUTDOWN ACCIDENT FREQUENCIES PROGRAM SENIOR CONSULTING GROUP MEETING MARCH 14.AND 15, 1990 b

ALBUQUERQUE, NEW MEXICO

s. EVENT TREES FOR MODE 2. EVENT TREE DESCRIPTION LOSP SAME AS MODE 1 LOCAs PRIMARY SYSTEM SAME AS MODE 1 MAINTENANCE INDUCED DEPENDING ON BREAK

SIZE, SIMILAR TO PRIMARY SYSTEM LOCAs

" MAINTAINED" SYSTEM MAY BE FAILED REC 0VERABLE DIVERSION DEPENDING ON BREAK

SIZE, SIMILAR TO PRIMARY SYSTEM LOCAs

" REC 0VERABLE" SYSTEM MAY BE FAILED PIPING OR COMPONENT DEPENDING ON BREAK SIZE, SIMILAR FAILURE IN OPERATING TO PRIMARY SYSTEM LOCAs CONNECTED SYSTEM " FAILURE" SYSTEM MAY BE FAILED

EVENT TREES F0R. MODE 2 (CONTINUED) EVENT TREE DESCRIPTION FAILURE 0F OPERATING RCIC REACTOR SUBCRITICAL, REC 0VER

FAILURE, RCS 0VERPRESSURE PROTECTION, CORE

COOLING, CONTAINMENT OVERPRESSURE PROTECTION

.RWCU. REACTOR SUBCRITICAL, RECOVER

FAILURE, RCS OVERPRESSURE PROTECTION, CORE

COOLING, CONTAINMENT OVERPRESSURE.

i PROTECTION INABILITY TO ESTABLISH SDC REACTOR SUBCRITICAL, ESTABLISH.SDC, RCS OVERPRESSURE PROTFrTION, CORE

COOLING, CONTAINMENT OVERPRESSURE PROTECTION

f f I } HUMAN RELIABILITY ANALYSIS METHODOLOGY F0R GRAND GULF. SHUTDOWN STUDY i e MARCH 15, 1990 TERESA-T. SYPE DIVISION 6412 i r

o

1. *, /;.

' l' 4 i 1 1

E o PROCEDURES T0. ENSURE EFFICIENT USE OF CHOSEND HRA: METHODOLOGY-TALENT'- BEINGNEGOTIATEDWITHLIV5RMOREANDNRd SHARP-<.< TEAM

NUCLARR DATA BASE -INEL USEFUL TOOL FOR REFERENCE PURPOSES 1 I

HUMAN FACTORS SYSTEMS ANALYST IDENTIFIES HUMAN ACTION AND HRA BEGINS TWO CATEGORIES OF HUMAN ACTIONS 1. PRE-ACCIDENT RESTORATION OF COMPONENTS AFTER TEST OR MAINTENANCE ACTIVITIES 2. POST-ACCIDENT RESTORATION OF A FAILED FUNCTION

k-HRA METHODOLOGY ASEP THERP CHAPTER 7 0F NUREG/CR-4550, METHODOLOGY DOCUMENT ALLOWS DIRECT COMPARIS0N TO FULL-POWER STUDY ACCEPTABLE TO HRA COMMUNITY COST-EFFECTIVE EASY TO USE PRODUCE NEEDED/ REQUIRED HERs DOCUMENTED / TRACEABLE SIMULATOR UTILITY PARTICIPATION SUPPLY REACTOR OPERATORS USE OF SIMULATOR E

J. [ PRIOR TD PLANT VISIT 1. PROCEDURES ADMINISTRATIVE SURVEILLANCE CALIBRATION TEST AND MAINTENANCE E0Ps SYSTEM DESCRIPN ONS PLANT LAYOUT DRAWINGS TECHNICAL SPECIFICATIONS

tt-P PRIOR T0 PLANT VISIT :(CONT.) 2. ARRANGE ORIENTATION TOUR 0F PLANT (CONTROL: ROOM) P H O T O G R A P H S. e.;,- ~ s.:...s. ~; - i

• TALK-THROUGHS WALK-THROUGHS OPERATORS SIMULATOR 4

i

DURING PLANT VISIT TASK ANALYSIS DATA SHEET DEVELOPED AS DETAILED RECORD KEEPING TOOL' STUDY HOW TASK INTENDED TO BE CARRIED OUT AND ADMINISTRATIVE CONTROL EMPLOYED INTERVIEW AND OBSERVE SUBJECT MATTER EXPERTS SAMPLE OF REAL TASKS TALK-THROUGHS WALK-THROUGHS TIMING-STOPWATCH CHECKLISTS PROCEDURES ADMINISTRATIVE CONTROL HANDLING AND TYPE OF TAGS - RESTORATION SIMULATOR CLOSE UP PHOTOGRAPHS TO READ DISPLAYS AND LABELS mE

f _;ot. .m_- u.-m -- - u,.,._m m e. m m.-u u m m m - m ,w w _,w.,. - - - m m, t SURRY SilUTDOWN STUDY APPROACil AND PROGRESS REPORT PRESENTED llY: T.L. CllU DEPARTMENT OF NUCLEAR ENERGY IIROOKIIAVEN NATIONAL LAllORATORY PRESENTED TO: QA COMMI'ITEE MARCII 14-15, 1990 G --e hh Q \\ 5l

__.+-mx-wma.a-_wam-mmmmmme . -. - ~ A PWR AT SilUTDOWN 235 LOSS OF RilR EVENTS AT U.S. PWRs (Up to 1986) lY SCENARIO h1AY LEAD TO CORE DAh1 AGE IN 30 A11NUTES FEW TECilNICAL SPECIFICATION REQUIREh1ENTS - SAFETY INJECTION SYSTEhl, DIESEL GENERATORS, CONTAINh1ENT INTEGRrlY IIIGIIER h1AINTENANCE UNAVAILAllILITY - 4ky ESSENTIAL llUS - REACTOR COOLANT SYSTEh! A1AY llE PARTIALLY DRAINED - STEAh! GENERATOR,.TURillNE. DRIVEN AUXILIARY FEED PUhlP we.-. 4

LOSS OF 11ESIDUAL i1EN1' ItEh10 VAL SYSTEh! AT DIAllLO CANYON, UNIT 2 APitiL 10,1987 ONE WEEK AFTElt SilUTDOWN h11D LOOP OPERATION 4 EQUIPh1ENT IIATCll OPEN .. INADVEllTENT DRAINING OF RCS IlOTil Rllit PUA1PS IlECAh1E VAPOR llOUND RESTORATION-OF VESSEL LEVEL WAS DELAYED RCS WAS IlOILING RilR RESTORED IN 89 A11NUTES [ ,e d((,2.*2,7$,,$sSU'"- bO ! a

'w'- d-l =i n

SUMMARY

0F EXISTIriG PRA STUDIES j STUDY-SCOPE: CORE DAMAGE FREQUENCY ~ NSAC-841(Zion)-_ LOSS OF RHR. 1.8x10* LOCA LOW-TEMPERATURE OVERPRES-SURIZATION-NUREG/CR-5015' (Generic / Zion) LOSS OF RHR 5.22x10~5- .l-LOSS OF-0FFSITE POWER. - l_ s. LOCA f SEABROOK MODE 4, 5, 6 4.5x10*- a i TRANSIENTS LOCAs EXTERNAL EVENTS N1RGELBREAKLOCA-LARGE BREAK:LOCA IN MODES SHALLER THAN THAT FOR - 3 3 AND 4 NODE 1 7 1: FRENCH TECH. SPEC. STUDY LOSS:0F.RHR 4x105 LOCA r-i- l SURRY SHUTDOWN STUDY N0ff-FULL POWER 1' FIRE, FLOOD } 3 k m __l_ -~ g

SCOPE OF SURitY SilUTDOWN STUlW LOW POWEll AND SilUTDOWN FUEL IN Tile COllE CIIARACTElllZATION OF PLANT Ol'EllATIONAL MODES LEVEL 1 INTERNAL EVENT Pila WITil Filles AND FLOODS LEVELS 2 AND 3 WILL llE PEltFORMED LATEll N' h og

g

=./rtr 201-_.J

i j -= ) IDENTIFICATION AND CllAllACTEltiZATION OF PLANT OPEllATIONAL MODES . - DEFINITION OF PLANT OPEllATIONAL MODES (l'OM) -1 i i - DETEltMINATION OF Al'PLICAlli,E INITIATING EVENTS (IEs) ESTAllLISil APPLICAllLis SYSTEMS AND SUCCESS CitlTEltlA FOR EACll POM AND IE DEVELOPMENT OF NON FULL POWEll DATA IIASE I t ..g . %T_m " W-- _. m;_mmm, V.-~-.~gne-bb 4 Ys e ne v{N 6stt stua41 1 & tres f.'Wrf . wit.nT7,s esis.i t.. Dut Li --==,m m._- su

) I ) DURATIONS Atl0 CHARACTERIZATION OF NIASES 0F THREE TYPES OF OUTAGES (Used in NUREG/CR-5015) j START

• END*

DURATION PLANT OUTAGES PHASE (HR) (HR) (HR) CONDITIONS REFUEllNG 1 54 167 113 RCS COOLING DOWN, RCS FILLED l REFUELING 2 167 587 420 RCS DRAINED. SG EDDY CURRENT TEST l l REFUEllHG 3 587 1087 500 REFUELING CAVITY FILLED, FUEL Sl10FFLING VESSEL HEAD OFF REFUEllHG 4 1087 1996 909 RCS FILLED, MAINTENANCE 7 RAINED 1 21 83 62 RCS COOLING DOWN, RCS MAINTENANCE FILLED DRAINED 2 83 179 96 RCS DRAINED, HAINTENANCE MAINTENANCE 4 i DRAlHED 3 179 982 803 RCS FILLED, HAINTENANCE MAINTENANCE i NONDRA:NED 1 21 146 125 RCS FILLED, HAlHTENAllCE MAINTENANCE

• TIME AFTER SHUTDOWN i

h m- "r,'wMN_ Ni as.m.,g4=mese.es ,m, ,g ete _.mm_____._ ":::::::::::::~ RC\\ F ,,e-,- ww.w,.e--.-,.n~- .w... ,.,, ~. ,.,,-,m--,---.----v,, wn, ~, - -,,, -+---.a,r,,--,-

f DEFINITION AND CilAllACTEltlZATION OF PO51s t POhls INCLUDE OUTAGE 'lTI'ES AND PilASES WITillN EACll OUTAGE TYPE i itEFUEl.ING, DitAINED hlAINTENANCE. NON.DitAINED hlAINTENANCE, hilD.I 001' Ol'EllATIONS, I'lANT OPEllATIONAl, h10 DES AS DEFINED IN TECil. Sl'EC. PAllAMETEllS USED IN DEFINITION llisACflVITY, COOLANT TEhlPEllATUltE, itCS PitESSUllE, VESSEI, I.EVEl, Tih1E AlTElt SilUTDOWN, DUltATION j r. CilAllACTEltlZATION OF POMs FitEQUENCY, PLANT CONFIGUllATION, SYSTEh! AVAllAllilEIT, SilOTDOWN ACflVilY, TihlE TO COllE UNCOVEltY, hlAINTENANCE i UNAVAllAllllJIY, itCS INTEGit!TY, CONTAINh1ENT INTEGill'lY 4 APPitOACil (INFOI(MATION NEEDED) PIANT VISIT, itEVIEW OF EVOLUTION OF PLANT PAST OUTAGES (LOG BOOKS), itEVIEW OF PROCEDURES, TilElth1Al, llYDilAULIC ANAIXSIS, DISCUSSION WITil PIANT PCitSONNEl, USE OF EXPEltlENCE FitOh! EXISTING SIIUrDOWN STUDIES ..,.- - - m,... . m.. .--, m.sr m w a m a m u,. w w w a m m. m m,.___,. t g g~em g,yp-n -. n -n -m nm_m ar 4 ( t ..=r====m

==c n.r.=m

iI' j1 {l1jllfjjl!j'

1l g

= d ra d = n.e 6, 6 - 6 S. d 6 6, - 2, 2 = a o t STM 1 3 4 - 5 5 5 6 = 6 6 5 5 5 3 1 i t .e S. d 5, 3, 2, 2, _ 2, 2, 4, 1 t o uSTM 6 4 3 2 2 1 1 7 1 1 1 1 3 6 -) g 5 y i 3 0 - r 5 s 2 5 r 0 3 E p 2 4 u 5 2 S 4 2 3 0 0 00 4 5' 0 45 0 3 T 7 5 3 2 11 5 4 ( 3 5 0 0. 0 0 = 1 d 1 - 1' K' 1 e ~ =' 7 1 = 1 d p 1 p e o f o l l o c l e l l y i i s e l d t d F = s v l i i i e e u = M v M S VL f a C C R = 3 6 r y e at 2 4 r w ca 2 8 o 5 ee /) / P 1 dh . ) 7y -) 8 n n n

6. rr 8o 4

0 o s e o e m O 4i 1u 1i 7 6 i 5Z 1S 1Z d i 8 t o T -( ( ( 2 a n r e l p a n 7 I O n - w I = r I v v - op e o v r - v - rn e i rn - do n g-g tg-n n n n o .o rn n fn-o S eo w t eo ' w w w dl - pi oi i_ a i pi C .- p a wi wi. o r ot o o o ed_ ot fl - l ) - ot=_R p-ot P ni_l ttu. Pa e Pa d - d - d _ oa e e_ ne oa u-p r l .l - oR a n_de lu u_ iu-lr= l r rG-r 1 i O wc o G. : o R - l Saa.op lf f af =de =l - we 1 oH-oHro_ip iee_reip =i Cet u oy o t LC C S. CR CR Dt _ MO FR R_DR MO=F RH SS LO. F na l r P o d t t e y cp r p r ai Ra Rp r er Ht Ho u RT RS Rt S S ( I

.. _ ~.--= = P SUltitY TECilNICAl, SI'ECIFICATIONS Ol'EllATIONAl,510DE DEFINITIONS DO NOT Fol LOW STANDARD TECll. SPECS. IlOUNDARIES OF DIFFERENT REQUlllEMENTS . CRITICAL.lTY . Tavy = 350 F and P = 450 isig l . Tavg = 200 F (Cold Shutdown) . REFUEL. LNG OPERATIONS ItEQUIREMENT ON OTilElt UNIT . AUNILIAltY FEEDWATEll . CllARGING PUMP a-~....--==--=======-.=-=====-===-==, ((((.((C[C[, na] $[r[ _ rr :r r r e _ (.

5j ~ h*'f L E iX. 3 4 i,; jI ,l9jijg! !J j {!t d ae h e h F t 0 h 0 t i T 2 w NAE r LR o m OU d OT e CA T n R o EE F F F F F i GP s AM 0 0 9 F 0 0 ne RE 5 5 5 0 4 t ET 3 3 3 0 2 1 V 5 y A 1 1 1 3 1 l lu f n 3 a h N t O .R s I . E s S W e I l DO V EP E s T t R AL l RA o b M C %R e E E 5 5 r P H u S T 1 0 0 0 0 so l P. c C d E a T e h D l R YK e i A T s D s I e N VN v A IO T TI e S CT 9 9 9 9 9 5 h .t AI 9 9 9 9 9 9 W ED h RN 0 0 0 0 0 0 t O i C 1 1 w lesse .v i t N ar O eo I N ht T N W c A Y. W O ya ae R B O D cr E D D T G e P N T U N de O P A U H I h U T H S L gtn R T S S E i n E R 'D U di W A T T L F u d O T O O O E ll e E P S H H C R cev xuo D EF m O

• e M

1 2 3 4 5 6

• r L

li 4 )' 2, I I s1}1 , :i ; I

l

!; i;,i4li! 4l:

77%.ab'X. .K.'J.*.FM N Wal.4 - w"J. 4.4 J :.a. .J y '.y"* ,La...;,- tf.... T s y,_ . g;, ;, -.: SUltitY TECll. SI'EC.

1. REMELINJ1EllUII)OWN1f1EI)lILON (6)

WilEN Tile ltEACTOlt IS SUllCitlTICAL l'Y AT LEAST SC'c ak/k AND Tyc, IS $140 F AND FUEL IS SCllEDUI.ED TO llE AlOVED TO Olt FitOM Tile IREACTOlt COitE.

2. COLILS111IUlOWNXilNiji'00E (5)

WilEN Tile ltEACTOlt IS SUllCillTICAL llY AT LEAST IG ak/k AND Tyc; IS $200 F.

3. INIliltMIsllIKLE SI[UTDOWN111NIID10N WilEN Tile ItEACTOlt IS SUllCillTICAL llY AN AMOUNT GitEATEll TilAN Olt EQUAL TO 1.77"o e k/k AND 200' F < Tg (;

< 547 F.

4. IlOT_StilRD.OWN CONI)D10E (3)

WilEN Tile llEACTolt IS SUllCillTICAL llY AN AMOUNT GilEATElt TilAN Olt EQUAL TO 1.77C' ak/k AND Tyt; IS 2547 F. / c 't

5. REMlfllLCRl'DCAL WilEN TIIE NEUTitON CilAIN ItEACTION IS SELF-SUSTAINING AND kg= 1.0.

t..,

~:=.r==m.u z =:=~- - " m -

,==a+.:st.==.==.====a=~=a=.= .. =a

==:. :=:=;;._; 2 x=_

e :;u=x==a:vu.ez su;m = g,

'.i E.73T.7 - - i.7.4 =7G. I'A-z'CET '~% ' 7'Ja' i.i.~ - -. s g gA prw f p -== =. = = --

j l SURRY TECII. SPEC. (Continued) j i i

6..POWEll OPEIMTION (1)

WilEN Tile REACTOR IS CRITICAL, AND Tile NEUTRON FLUX POWER RANGE INSTRUMENTATION INDICATES GREATER TilAN 27o OF-RATED POWER. t 4

7..REFUEldNG OPERAILON r

ANY OPERATION INVOINING MOVEMENT OF CORE COMPONENTS WilEN Tile VESSEL llEAD IS UNilOlXED OR t

REMOVED,

( t. L -t 8 1-i I l.t r. i Y -n- .n ~.,... _ _ _. dU; pn ..........u..... asks ist eo (myg.ogit ig g, e.g _ m.w- --w,m-ww y r-- ar,,--c-m-,-,-wn--,,,,,n,_,,,.,n,, _m,.,_,,,,_en,.,,,,m.,, n r. ,c~.,,,,-,.,e,- , c v, c o n y..--. c n

---

-___ ______,_.m._

• /.,,

SURitY TECll. SI'EC. FOR Tavg > 350 F AT LEAST 'lWO REACTOlt COOLANT LOOL'S SilAl,L llE Ol'EllAllLE WITil AT 1. EAST ONE IN Ol'ERATION AT LEAST ONE OF Tile 'lWO SGs IN NON ISOLATED 1.001' SIIALL llE Ol'EllAllLE CONTAINMENT INTEGRITY (Iteiluired Unless Tavg < 200 F) TWO DIESEL GENERATOltS (Seven Days AOT liefore Cold Shutdown) TWO MOTOR DRIVEN AUX 1LIARY ! EEDWATER l'UMI'S (Three Day AOT liefore flot Sluitdown) 'IWO CCW l'UMI'S (One Unit Ol>eration) TilllEE CCW l'UMI'S (Two Unit Operation) ONE CilARGING l'UMI' COOL.ING SYSTEM SIIALL llE Ol'EllATING. Tile Sl'ARE SUllSYSTEM SilALL llE OPEllAllLE - WITil 24110U11 AOT. '6,000 GALLON IN CONDENSATE SToltAGE TANK ) TWO DIESEL GENEllATORS, 'lWO EMERGENCY llUSES (4160 AND 480) ._ _ m _ __ _ _ _. n = _-. _ _ _

"3tJ-h*1XMM'" '--+ "'Wf6.*iMM MJRdddes2.ib--(~XM'M',J4: 4-' MVe% Aki!L'# b MW444.N L%LMMTUnhEJJ%%f4&ff E;.m?47 SUltitY TEC11. SI'EC.17011 Tavg < 350 F A MINIMUM OF TWO NON ISOLATED LOOL'S, CONSISTING OF ANY COMillNATION OF ltEACTOlt COOLANT LOOL'S Olt itillt LOOL'S SilALL llE Ol'EllAlli,E, EXCEL'T IN MODE 7, ONE LOOL' OF Rillt MUST llE Ol'EllAllLE AND CAN llE ItEMOVED ONE !!OUR PEll EIGilT 110U11 l'EltlOD. ONE l'ItESSUltlZElt SAFETY YAINE SilALL llE Ol'EllAllLE WilEN Tile llEAD IS ON Tile VESSEL. TWO l'OltVs WITil SETI'ING OF < = 435 psig Olt A llUllllLE IN Tile l'ItESSUltlZEll FOlt LESS TilAN 72 llOURS Olt Tile ItCS VENTED WITil ONE OPEN l'OltV. A MAXIMUM OF ONE CilAltGING l'UMI' SilALL llE Ol'EllAllLE. ONE Fl.OW PATil FOR 110111C ACID INJECTION. O ACCUMULATOlt IlOllON CONCENTRATION IN ISOLATED LOOL'S. TWO TitAINS OF Tile MAIN CONTRO'. ROOM AND EMERGENCY SWITCll itOOM VENTILATION AND Allt CONDITIONING SYSTEM (ItEQUlitED IF AllOVE COLD SI1UTDOWN). ,,,.,....o CONTAINMENT INTEGRITY '(Tavg > 200 F) -==_____=-_.m_m.__, mm..m m__.m tv4Z4: regncmg- ---v -~v;;;We,rf ur.2-tis 4& art %Tmm I g& 4.* L __==_

{

---

_ _ _ _m-mm--

m___m_.____ i i SURRY TECll. Sl'EC. FOR REFUEL. LNG Ol'ERATION 1 i I t CONTAINMENT INTEGRITY, e, l , - c. t i ONE R11R l' UMP AND llEAT EXCilANGElt SilALL llE i Ol'ERAlli E. IT CAN llE REMOVED ONE IlOUR l'ER EIGilT IlOUR l'ERIOD. r .- - NO MOVEMENT OF IRRADIATED FUEL, IN Tile REACTOR CORE ilEFORE 100 llOURS ALTER REACTOR TRil' r t-ese4W.t MNW d MM gig h my. y_mmhqMd q y ' wmmwatsah%"Cr _ - WJN%JtM Fiaprim e YYYb.is p . assi a n. ,.b,4 _ aur - b OW!bSN* .ssa _mm im-~-,-.,-,, v...s,

l n un.. -muvaan-~~ .mma-acmanu-w - ~- ~ ~~- INITIATING EVENT ANAINSIS IDENTIFICATION OF INITIATING EVENTS SEAltCil OF COhlPUTERIZED DATA IIASES SCSS, NI E;' NPRDS REYlEW OF SilUTDOWN DATA REPoltTS NSAC 52, AEOD/C503 ItEVIEW OF EXISTING SilUTDOWN STUDY NSAC 84, SEAllROOK, FitENCil STUDY (NEED llELP TO GET llEPOllT), NUREG/CR 5015 4 ItEVIEW OF PitOCEDUltES USED AT SilUTDOWN - OPERATING, TEST, h1AINTENANCE, Eh1ERGENCY, OFF. NORh1AL, AllNOlth!AL REVIEW OF IEs FOR POWER OPERATIONS FOP. APPLICAlllLITY REVIEW OF GENERIC LETTERS, INFOlth1ATION NOTICES, ilULLETINS (TO ADDRESS NitC CONCERNS) REVIEW OTilCR STUDIES TilAT IDENTIFIED SCENARIOS AT SilUTDOWN: NUREG/CR 4999 LTOP NUREG/CR 4982 SPENT FUEL POOL ACCIDENT, FUEL TRANSFElt CASK DROP 4 4 ~ NUREG/CR 5368 REACTIVITY ACCIDENTS WESTINGilOUSE TilERSIAL llVDitAULIC ANALYSIS ON hilD LOOP OPEllATIONS * ^' ' WESTINGilOUSE PRA OF LOCAs IN A10 DES 3 AND 4 (NEED itEPORT) .,;gT;;ggggt Mw =mye wy ggm_r + + ~y-Mw, +=- =gyqy 3~ c -,-~~_3xg mm m z.m yp -- - - -, m m-... ..e yggg m yga s v .e 4 % l 8 ?

h l l IDENTIFIED SCENARIOS AT SHUTDOWN t s 1 l' I I SCENARIO' STUDY-CORE DAMAGE FREQUENCY j i p 3.. LTOP " + - HUREG/CR-4999 2.6x10' (INDIAN POINT). NUREG/CR-4550,'VOL.3 <10* (SURRY) ~ NSAC-84 NOT ASSESSED (VERY LOW) d .PEFUELING CAVITY i., NUREG/CR-4982-10* (GIriNA) ll l'SEALFAILURE

f i

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LOSS OF ltillt DATA IIASE (PWils) A TOTAL OF 235 EVENTS (UP TO DECEhlllElt 1986) SOUllCES: NSAC 52: IRESIDUAI, llEAT llEh10 VAL EXPEltlENCE ItEVIEW AND SAFETY ANALYSIS, PitESSUltlZED WATElt ItEACTOllS, lANUAltY 1983.- AEOD C503: DECAY llEAT ltEh10 VAL PitOllLEh18 AT U.S. PitESSUltlZED WATEll itEACTollS, DECEh111Elt 19h5. NUllEG/CR 5015: Ih1PitOVED itELIAlllLITY OF llESIDUAL llEAT REh10 VAL IN PWits AS RELATED TO itESOLUTION OF GENEltlC ISSUE 99. SEAllROOK STATION PitollAlllLISTIC SAFETY STUDY, SilUTDOWN (510 DES 1,5 AND 6), h1AY 1988. 1 _ _ __ _._ _ y wa~... --w smu- --- -nm ..---. _ _._' ~ 4% 5 iti O W t. t

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1 4M'fEJaeeW $7.sWX4TalFM4 JTrJEMP.4WP3Cig tiMJM34 MM ENEAME-61TrJ rJ Mh2T+? A" wMt14rJWM S tNCatt;J77ti-@m" '-"41W hu.A1Mr= 1 LOSS OF OFFSITE POWER CATEGORIES (NUREG/CR 3992) fKuiG1111LI: FAILURE OF Tile PREFERRED, Ai!!'ERNATE 4 AND UNIT POWER SOURCES EITilElt WilEN Tile REACTOR WAS AT l'OWER-OR WilEN SilUT DOWN, ilUT IF TIIE PLANT WAS SilUT DOWN Tile I.OSS WOULD llAVE OCCURitED IIAD Tile PLANT llEEN OPERATING. l'OWElt WAS NOT f(ESTol(EI) WITillN 30 AllNlTl'ES. e l'OWEll WAS 1(ESTOl(EI) WITillN 30 511NUTES. ENDiGDJ1Lll: Fall.URE OF Tile PREFERRED AND A!!!'ERNATE POWER SOURCES. Tile UNIT DID NOT TRIP OR, IF NOT OPERATING, WOULD NOT llAVE TitlPPED llAD IT llEEN OPERATING. CN1'EGORY-ll]: FAILURE OF Tile PREFERRED AND UNIT OFFSITE POWER SOURCES. AN Alli'ERNATE OFFSITE POWER SOURCE WAS AVAILAllLE IlY MANUAL SWITClllNG. fKl'liGQl(Y IV: FAILURE OF ALL OFFSITE POWElt DURING A COLD SilUTDOWN. TilESE SPECIAL MAINTENANCE CONDITIONS DO NOT OCCUR DURING OR IMMEDIATELY FOLLOWING OPERATION. fKl'EGORY V: LOW YOl!!' AGE CONDITIONS ON Tile TRANSMISSION SYSTEM. Vol!!' AGE WAS NEAR OR llELOW 90% OF NORMAL. .3y.- --._+_-r e ms;myw I i. . 0 +

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b I 4 t 1 D cluatre GIUDNfRMJd 'JMIL1E NMLYS11UT31SrJUMMuR1M pun *ML0 TWIT 1WS 1. h@Jgi Arulysis of Rink at Itu hver an! shutdwn O.ntitioru. I 2. tmh _leMeri Ricturv) Ihbimon, Pithibillutic Rir.k Amlyain tunn:h, Division of Dyutern feroatth, Of floo of thelear Ergulatory Remarth. 3. ID9ri Traditiorully, pitinbilistic rick arnlyres of revere occidents in nuclear twar plants (Ireltdito t1wa of the imC utaf f's rcant IU10n-1150 nmlysis) invo considerrd the ret of initiatin) evea.ts ' notentially orurring durin) full puer q mmtion. Scro rettening amlyrm of accident initiators darlig lw pur, rhutdx1, and s +1er trdan of plant geration other tJun full }wer luvo tren parforred. 1hese ran)cstol tJnt rirls darlig there mica were tral rnlativo to tJvne oxurrin) darity full pwr gr: ration. Ikuner, other stadies (dicatrmi later) ani the Q)enx27/1 accident, which cmurrtd during Im pwr testing exextiros, cuy3estcd tlut Iw Fuer an3 chutdwn accident rirls oxid bo significant. 'Ibe treent event at Vcytic Unit 1, khile at cold rhutdwn, further eginsitco the twd to systextatically arrl cuqprrhensivnly evaltato plant rafety when cynratirn at other than full guer. As such, the arnlynin of the fnvparcico, consequences, ani rinha of thono accidents was identif1(d as om tank (Tari 1.4C) in the IRC staf f a ctuly of the inplications of this accident to U. j k\\\\

1. ?, S. comotuial suc1 cur pwr plants. 4. 1M)LB1mrol 1he cbjcetivce of the tank atv Ub acre.s the invpen:les of mvere accidents initiattd during plant cpiratican1 axles, other tinn full gu2r geration, for a ocutercial pzrnmarind water tractor (ikH) an3 a inflity water inv: tor (thB); 7b cxrbim accident firquemics with accident prvJtumion, rattro terrn, ard offsito con %queJKU amlyral to yloid OStimtal of EcVGTO accident rinks frun those plant cption raies in the stulicd 1%R and thHf ard 7b cnqure the estimtid core dungo frcquercies, irgortant accident repenoes ard other cpulitativo aid cpantitativo trallts of this stuly with thom of accidents initiatni during full powr creration-(as arzeral in !URDJ-1150). - 5. Sqt;xyl . As dicalsred alxwe, the work perfornd urder this task involvea ' invectirjation of -two. cpreratiin ocrwarial reactors, a l'hH ani a ibH, at plant oporrtioral nr. des (104) other tlan full power geration. -1ho curIrnt - l L plan canists of a tso-11ntal ajpimch in ottier to pluvido an early t-

t I r 3 arulyuin cue.tvlw arti to highlight any lotential pnblan alwtu. Ihre 1 in &dicatal to pruicity prul(mimty 11A noults, liclulinJ intenal fitu f ard fico31n] amlyres, for oiler rulated tAulics unhavay in tho IUc. Ihva 2 la to pnxiucu a timi lie aralysin, guida3 Ly the linro 1 amulta to ptiqurticvntely ocmentanto the ef foit anorg tho varicxis gerutiny rcdon, the denirnnt rcquarms, ani lertimnt dita ituto aamniiny to their irgurtan;o to curv dringo flujoen;y aid rirJ;. '1ho txqu of thin tank does rot inclu.1a any coltaic arnlysis, tot oms influjo the f ollwirn (for Loth 1 parcs): (i) Identification of Plant Oletatiomi Wden (104a) ard Ikranetenas f 'lhere are rm oral plant q vraticrnl 1:oies Lewicies figl1 Irwr qmution duracterizcd by ruramicra uudt as nuctor criticality, reactor crolant systin (IC3) pluntre, IC3 ttqcrainte, ani }etumt thermi pNor. .9.3d1 Inica of quration atui Icv Jo.er, startup, hot stardby, hot rhuttlwn, cold rhutdwn, and refunlisq. '1hus, thin tank is to defino the plant ruics of cguration of intenst as a founbtlan f or }erfomin) rink amlysis. (ii) Detemim the Iqplicablo Initiatirn Dmnts for Fach itti: '1ho coore of thin tank is to detemim a ret of initiatin3 cvcots for eacn Ici that lotentially Instit in oore damge, incitdity thcre initiatily events are.cciated with ani resultif y f rtn minternJoe activitiec ani plant rolif ic.iticm, as well as tJan ar.r.cciated with

j 3. -- fires ani fIoxls-inteam1 to the plant. (iii) Establish the Jmlicable Systens ard Stxxvss critoria for Fadi Im and Initiatiny Event: 1he initial corditions of the plant, eqxclally IK:S prxmute ard trzp?rature, ani the availability of - stea:n will af ftet tJw cparability of en3 nxutd rafety systtma -(ud can systena int i defirgxi as " safety" int for which caf ety citdit my be given). 1his task ' is to identify the alplicable systenu for cadt im ani initiating event vith the cortigon31ng success criteria so tJut system radels can be omstructsd. (iv) Develop the Non-Full-Iber Ibta tuse: In order to develop a data tuso for run-pwer cparaticuni rodes, plant testin3 ard minternnce practicts, proxdures, ard.loys will be crunined urder this task to duracterize equi [wnt an1 systma umvailabilitics for the various Ims. Mean tino daration (per year) will also be establichxl for cadt of the Ims. qnrating pruxduma vill be Irviewoi to detetuim if ani what systrm my be byIncrai durin3_a given I m. Utchnical qwcifications will in reviewcd to detemine stut zulaxatiorw will be in ernet dur.ing the given Im. (V) Arulysis of Accident Praguoncics: I:

.S. Ihred on identifi 1 initiatin] events for each -Ici, accident frequercy amlysis till be ' carried out, enoaupTssity) (1) data amlysis, (2) systs amlysis, (3) event troo amlysis, (4) Interml fitu and flcodiin amlysis, (5) deperdent failum amlysis, (6) humn tullability amlysis, (7) accident ocyaenae c sntification, (8) plant dwage state arnlysis, ard (9) urcertajnty amlysis. (vi) Accident Progtussion anl cantainmnt Amlysis: 'nie scope of this task inclu3cs exantimtion of appi bble ttchnical qxcifications for each Kti to identify containmnt status aid systces availability ard develep accident pluJtrcsion ard containwnt event trces, and to cuTy out quantification ard develop accident plc.p.mion bins fur each plant drage state. -(vii) Soutre Tem Amlysis: 'Ihe rext step in the risk calculation is the scxum tem analysis. The results of the scaum tem amlysis aru release fractiors for - groups of chemically sirailar radionuclides with associatcd enen3y content, tire, and duration of rulease for each accident pluJncsion bin. Source tonn for shutdcun aid lo.ur guer events will' te corrccttd for reducrd fission pruiuct ard dccay heat levels fmn the full pcuer soutre tenra. ._viii) Consequence Amlysis: (

6- 'Ihe fimi stop in the risk rpantification vill b3 tha offsite ~ corngaence amlysis for soutre tents defimd in the provicus stcp. h specific consequerce masuits my inaltx3e corly fatalities, latent cancer fatalities, }xpulation dcce, etc. After the performnce of risk quantification for two plants, generic insights and spccific trecrmerristions, if nocxmary, will be developed to nduce estirates of frequercies and conscquences for accidents which my cmir durhg the law power or shutdown nodes of creration. 6. liork_INEriptdonsi In this &ction, in addition tc discussing sna of the work parfontud dirrctly for the task describxl in Scction 5, ran tulatal sttdies perfomx! by others are also identified. Findings frun them studies are described in the rext scetion. Under this task, a sttdy of c thR plant is in prognw_.s at the Sardia fiatiom1 Iaboratory (SNL) aid a study of a IkB plant is in pitgruss at the Brookhaven Natioml laboratory (EUL). For the selected thR plant, scrc of the cngoing sork incitrics review of the Fiml Safety Amlysis Peport (FEM) ard technical specificaticrs to develop a mtrix which describes each cperatiry taxle in tenns of the relectal parancters such as treperature, pressure, etc. Also, tabulated informtion on technical specification rcqailtrents-versus operating txdes is also bairn developxi. Various docanents, including safety amlyses contaimd l

in the plant ITJJ1 are licenoce event reports, tuvo bcen tuviewod to _ define initiatirq events relevant to Icu pcuer/chutckwn modes. -It is antic 3}utal tJut the pinliminary (nnse 1) accident frequency cpsntification for 1xath plants will be complettd by the mickile of 1991 ard a roIm oxprshensive arnlysis (Huso 2) by the miclile of.1992, to be follcued by the fiml risk comptstations approximtely six renths later. Scmc of the other complettd stulics relevant

to this tank ircitde the follcuing:

(1) Seabrook Pninbilistic Safety Study - Mxles 4, 5, and 6 (Paf.1); (2) Brurruick Dx:ay I! eat Rem /al Prdubilistic Safety Sttrly (Ref. 2); .(3) . Zion Nuclear Plant Residtul lleat Pc: eval IRA (Ref 3); .(4) Inprovcd Reliability of Residual licat Rccoval (bpability in IWu as Related to Pesolution of Generic Issue 99 (Refs. 4 ard 5); (5) Residual llcat Rext/al Experience Revic=# and Safety Analysis - Ptmincd Water Reactors (Ref 6); (6) ' Residual float Re::cval Experience Reviev and Safety Arnlysis - Ibilire Water Reactors (Ref. 7); (7) Reactivity Accidents - Reassescamt of the Drign-Ibsis Events (Ref. 8) ; (8) Prd>abilistic Atulysis of 900 ne Prurch IER Shutdcun Technical Sgcifications (Ref.15); and (9) PRAs of the French 900 M4e ard 1300 K4e IWRs (Ref.16). 7. . fln. dirnst.

~8-Befom describin; finiings fran sttdies listed in the pmvlous section, scraa general discussion of the irmes is useful. In most of tJw nus, it ins boon assuwd thst the level of risk av.ociattd with accidents initiated durinJ full pwer cforation, while rmall, is substantially greater tJun that associatal with accidents during Icv power or chutdcun. 'Ihis assurption is sunotted by the fact that hraune of the lower decay heat levels ani cmller radionuclide inventory during Icu Tcuer/rhutdcun mxies there is generally nam tiro available to recover frun adverse sittntions during these naix of creration. licuever, there atu other factors khich might exacettete the situation durirq accidents at Icv pwx/shutdcun, Scraa of t]nw factors are: (1) the fact tlat mny of the autcmtic safety system my have toen disabkd d" ring these nxies requiring greater cperator intervention; (2) high cquipmnt b..tvailability due to planned mintenance; (3) potential minterance configurations rtquiring minimum RCS coolant inventory; (4) open containmnt penetrctic = :n3 hatdies; ard (5) imdegaacy of full rcuer emryemy prtoxtures to ackdress emiyencies at 1cv pcuer/rhutdakn nodes. In addition to the above factors, mrtain expariences and events at c$erating reactors provide further irretus to study ris daring Icu power /rhutdcun axles of cperation (e.g., Refs. 4, 5, 6, 9, 10). Om type of event is the Gerrrbyl ' type of event, that is, rapid insettion of reactivity causing accidents. Other-types of events represent loss of dccay heat rumval functions, locs of coolant inventory, ard inadvertent prrcsuriration. To systemtically exantim thcse cormms, two NUFM-11SO (Ref.11) plants are chocen under this task. In Fef. 8, reporting on k'ork perforred in support of Task 2.1A, a study of

accidents khich result frun latye ruactivity insertions was carricd out for a Iwt plant an$ a Isa plant. 'nw potential reactivity accidents have bcen -catcgorized in that study as follcus: W E Dients 1. Miing diluttd acrunulator water during refueling 2. M31rq diluted 1661' water durity chuttkAn 3. IOCA with diluted EOCS water 4. IDCA vith sunp water dilutcd 5. LOCA/SGIR with seconiary diluting prinu,f 6. Inadvertent toron-dilution at shutdown 7. Startup of reactor coolant }utp after inprop1r di.lution 8. Beyord-design-basis rod ejection accidents 9. 'IheIml-hydraulic transients with positive MIC 10. Other tcyondMesign-basis even'.s Y.m Events-1. .Beyond-design-basis red drcp accident 2.- Rod ejection accickmt 3., Beyord-design-basis overpressurization events 4. Plushirg of boron during an A'IkS S. Operaticn in region of instability 6. Refueling accidents 7. Other beyord-design-basis events

N 4 a 1. Few of the above listed weru identificd tryuiring futther arnlysis harai on the. estimttd frrquercy of worst accident coprres. Ibring the current task, there events will be exantincd for their applicability to cpuntioral rxdes of interust and further arnlysis. Iboently, a Iovel 3 IBA for the scabruck station has bcen coqplettd to evaluate the likelihocd of revere corn danv3e with various Inths for offsite tulcare for the plant in Fb3e 4 (hot shutzkun), Ftde 5 (cold rhutdown), or Fb3e 6 (tufuelin3) (Ref. 1). Radiological soutuo tam and tusultant public health correguerres weru also evaluatcd. Findin3s ard cunclusions frun this stuly att rerxxttd as follcws (Ref.1): (i)- With the benefit of rulatively Icu ort.rMificaticus and adatint.strative conttuls, the frequency of core damge durirg chutdown is sanll, but nan-negligible, in orparison to power opctration. 'Ihe inptuverents include: Instntaantation and alarm to iqxuve operator action ard to foretoll incipient 1cca of (mm) system during the tim khen the Rt3 is-draincd to the hot Icq midplane; L Pro:cdurm ani training to cover the poc:sible abnorm 1 plant l corditions ard alterrative cooling schems; aid i l'

4- = - - Administrative controls to minimize the tim in the drairal-dcun configuration, -to assum that altcITutive coolin) mthods are available, ard to aratre conttul of contahn2nt integrity. (ii) 'Iho follcuirg surmarizes scro quantitative mnclusions of tJ'o sttdy: 'Ihc nmn wre damge frc.quercy when nhutdcun is loca tJun during full puer operation by about a factor of 6; and 'Iho early fatality risk frm diutdcun in atout an ottler of mgnittdo less tJan full pcuer creration. Entlts. of two EIRI-sponsomd stixlies to evaluato experiences with the IUR systers for both the WR and SG plants arv summarizcd in Pofs. 6 arri 7, respectively. In Paf. 6, scn3 251 shutdown events frun 1977 thtu)3h 1981 for W'R plants worn evalustad. Nxut 100 of there events involved an actual locs or significant degradation in the Pims while it was cycrating in a dccay heat reroval mxle. 'Ihe mjor safety imlications of these events fall into -three catrgories: (1) Loss of reactor coolant inventory via the IUR; (2) Imdvertent cold oveIpressurization of the IG; and (3) loco of long-tenn decay heat recrual capability via the lum, i Similarly, Ref. 7 surveycd 480 BG events involving the lum, ard 90 of those events involvcd an actual loss or significant degradation in the IUm. 'Ibe safety inplicat. ions are the same as thoco discusred for It'Rs. l l..

k I As a follow-on staty to Ref. 6 findings, DEI also quisonxl a txsiutul hmt nnwal prxinbLlistic stn3y on the Zion (a ihm) plant. We stniy crn:Inid, by conptrison with the mmits of Zion Pralnbilistic Safety Sttdy, tJut the annati frequency of fuel damge frun events initiating durirg diutdcun is less, by a factor of 5 to 20, tiun the firquency of corn damge from t_rvsients initiatcd at power. licuuver,- the shutdown rink is highly depen 3ent on cperator erttr, and wider uncertainty exist a for the uhutdown ordel mmit thui for the full pcar results. A sintilar prdnbilistic study was also parforwM for 11runswick Unit 1 (a lhR). In styrort of the troolution of Generic Issue (GI) 99, which deals with locs of residtn1 heat renwal events in NBs, INL mamlytod the Zion sttdy (Ref. 4) by applying som mMificati:= in the definition of outage pruses ard their duration ard rodeling of humn cognitive errors. '1he estimted core damgo firqueny. mpresented a non-triviar contribution to overall colu damge fragtharcy (at full pcuer ard shutdown). Ref. 14 describes an inspcction team repart of a rco2nt IIcident at Vogtle. Le plant was creratire in a mic-Icrp cordition (ttduccd inventory) when a lacs of offsite pcser occurrvxd due to an accident in the switchyard. One of the onsl% diccal generators was dcun for mintervuce, an3 the other diesel failcd to operate. (boliry for dccay heat renwal was Icst for 36 minutes. Ref. '15 is a conference ptper that describes mxarth by Electricite de Franx (EDF) that used IFA raathodology to arnlyze the iguct of twhnical sprifications

on cure relt frapency of a IwJd) 900 Mb IKR durity cold diutdcun. By -irplcm nti19 dutyos in rhutdcun todinical g ecifications in the arma of sdicdultd urnvallabilities of certain system (rainly affecting lam), the cott telt firquency was rtdaatd by a factor of fcur. Faf.16 sunnarizes the results of two lims: one done by IDF for a 900 Kb IW, ani the other by Institut de Prtttetion et de Surete llucleaire (IIGN) for a 1300 Eh IMR. Ibth Pl&s invcstigatcd the inportitrce of risk khen the reactor is not at c5 erating [wcr (i.e., rhut dcun with the ICR opunting, or when refuelirg). '1hcse states accountrd for over ralf (551) of the total colm Irelt frcqueary in the 13004Me plants, ard alrost one-thirti in the 9004Me plants. Firdings of the st1 dies discusscd abcne will to Irvicmi in this task, ard insights gailyxi will be uscd to develcp trdels ani carry out arulyses. 8. Comlusiorn A urder this task is still in prtyress. licuever, as discusmo in the previous scction, note work ins already teca dare to evaltate risk during nan-full power rrdes. As a resolution of Cereric Iccue 99 ard events occurring at opuntirq plants, several inforvation reticrs luve teen issuecl to licensees (e.g., Eefs. 12 aid 13). Ahinistrative ard pruxdural charucs tave twn) evaltnted and inpleunital. Several hcudoure duny:c lave also occurrcd. 9. DrurM _____o

. At the acrnpletion of this task, two IURIE/01 ITicrts adirtcsiry risks at a IMR plant ard at a IMR plant during the Icw }xucr/chutdown redes are expectrd. 'Ibese reports, in Inrt, will deal with the ricks which ray result frun the accidents initiatcd by the reactivity insertion (01e2ntyl type accident). In addition they will also acktress risks initiated by systcn mlfun:tions, (OmrnWyl could arguably have been initiated by hunnn ertur) ard other events. 'Ihese stulles my lecd to rc<xmendations rtganiing pocsible hald.nru, procedural, training, va staffing durges. Follcuirg the establishemnt of imC's researid1 program in this arva, a shutdcun event (describcd in Section 7) cocurrtd at the Vcytle plant. In responxa to this ircident ard grcuiry corocIn in this atua, the imC Executive Dirrctor for Operations issued a June 21, 1990 mencranium rtganling follcu-up actions _ to the IGC inoprtion team's rufort on Vogtle (Ref.14). Subsequently, a task plan was fonallated to evahnte plant cafety durin3 shutdcun operations to ensuru that risk during all ades of operaticn is ac ptably Icu. '11eco evaluations will form the insis for: (1) any pttpcncd durges to current todinical specifications which govern shutdcun cperations, (2) dunges in direction royanling the new-standard tcchnical specifications that are urder developemnt by the staff, (3) reccmmerdations to industry regan11ng cmr13ency Iesponse pruccdures ard cutage, mmgement ard contrul, and (4) rodifications to the lac insprtion pingram. It is further planned to develcp a korking agrt m 2nt with irdustry tupresentatives to ensure coop 2rative efforts in ailrossirq chutdown risk. 'Ibpics which will clearly involve significant interaction with industry groups l-include technical sprifications, ernrgency operatirg procedures, ard risk mrngemnt applied to shutdcun activities. 'Ihe mjor tedinical acpects aru l

~. h - expxttd to be ccarpletc4 by aiproximtely mid 1991. 1 10. Rnfercng;n t 1. Fbody,- J. H., Jr., et al., "Seabrrok Station Pinbc bilistic Safety Stuly - Shutdcun tbdes 4, 5 & 6, PrtxMngs of Conferurre on Prolabilistic Safety Assessnent," Pittsburgh,1989. 2. Iloldemess, J. _ II., et al., "Blumick Decay Heat Ikmaral Probabilistic t Safety Study," DRI/lEAC IWport, !GAC-83, pre {nred by Impell-Corp., CL h, 1985. ~3. . Bley, D. C., et al., " Zion Nuclear Plant Residual lleat Remwal TEA," DRT/NSAC Report, IEAC-84, prepared by Pickard, Ime, aid Garrick, Inc., July, 1985. 4. mu, T-L, et al., " Improved Reliability of Residual licat Removal Cainbility in IMRs as Related to Resolution of Generic Issue 99," Brookluven National laboratory, NURD3/CR-5015, April,1988. r 5.- Spano, A. H., "Rcqulatory Amlysis for the Ecsolution of C<meric Issue 99: Imm of RHR Capability in IKRs," U.S. Nuclear Regulatory Comission, h'JREG-3340, February, 1989. 6. Vine, - G., et al., " Residual 'Ileat IVroval - Dperience ' Review and Safety Arnlysis - Pressurizcd Wat'r Reactors," FJRT/NSAC, Report IEAC-52, Jantury, 1983. 7.

Vine, G., et al,, "Raidual Heat Emoval Dporience RevicW ard Safety-Arnlysis - Doiling Water Reactors," FJRI/NSAC, Report NSAC-88, Rarch, 1986.

8. Diarord, D. J., et al., " Reactivity Accidents - A Reascessment of the

m. ____.-._.-_m__.---.___ 4 4 6 0 Dwign-Basis Events," thuoktuven Natioml laboratory, NURD3/m-5368,.Dmft ~ Povision 2,' June,1989. b ~

- 9.

Omstein, H., " Decay lleat Pemoval Prublems at U.S. Pressurized - Water Peactors," AIDD Case St1dy Frport, AfDD C-503, U.S. Nuclear Pagulatory Ceregnission, Decefier,1985.

10. -

NRC Aug~ented Insprtion Team, "Ircs of Decay lleat Rawal System," Diablo Canyon Unit 2, April 10,1987, U.S.11uclear Regulatory Conunission, NURai-1269,. June, 1987, 11.' HRC, " Severe Accident Rinks: An Accerement for Fivo U.S. Nuclear Ihcr Plants," U.S. Nuclear Regulato)y Cantnission, NURD2-1150, SecoM _ Draft, June, 1989.

12.

NRC Infonration Notice No. 88-36, "Possible Sui 3en Incs of ICS Inventory ' During Izu Coolant IcVel Op2 ration," June 8,1988. - 13.' . NRC Generic Intter 87-12, "Icas of Residual IIcat Renaval hhile the Reactor Ctrlant System is fertially Filled," June 9,1987. 9 114. NRC, "Ioss of Vital AC 1%er ard the Residual !! cat Renwal' System During - Mid-Inop operations at Vogtle Unit 1 on Mardi 20, 1990,_" U.S. Nuclear Regulatory Cbmmission, NURIn-1410, June 1990. 15. Mattei, J. M., & InIT1, G., "PIdxtbilistic Amlysis of 900 MWo IkTt Shutdown Technical - Specifications," PSA 87, Vol. 3, pp 801-806, IMS & SNS In*mtioml - 1bpical' Conferenoo, AtrJ '_30 - Sep 4, 1987, Zdrich, K61n: Verlag 1UV Rhein1ard 1939. -16. EDF & IPS!1, "PIdaabilistic Safety Sttdies in Fratre: Mut is the Core Melt Protability for the lhTbs?," SFIN semimr, May 16, 1990, Paris. I l= ,,}}