ML20198H401

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Summary of 971208 Meeting W/Bg&E in Rockville,Md Re Plant Pra.List of Attendees & Viewgraphs Encl
ML20198H401
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 01/05/1998
From: Dromerick A
NRC (Affiliation Not Assigned)
To:
NRC (Affiliation Not Assigned)
References
NUDOCS 9801130278
Download: ML20198H401 (33)
v • d • e


Text

. .

I

  • Jannry 5,1998 LICENSEE: Bahimore Gas and Electric Company Facility: Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2

SUBJECT:

SUMMARY

OF THE DECEMBER 8,1997, MEETING REGARDING THE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 - PROBABILISTIC RISK ASSESSMENT On December 8,1997, the NRC and Baltimore Gas and Electric Company (BGE), the licensee for Calvert Cliffs Nuclear Power Plant, held a meeting in One White Flint North, Rockville, MD to discuss the Calvert Cliffs probabilistic risk assessment (PRA). Enclosure 1 is a list of attendees. Enclosure 2 is a copy of ths viewgraphs distributed at the meeting.

The purpose of the meeting was for BGE to preser' an overview of their updated PRA and to discuss unique key features of the study that may have contributed to Calvert Cliffs calculated core damage frequency. BGE stated that the PRA update !ncluded external events, the inclusion of two additional diesel generators, the removal of the high pressure safety injection pumps dependancy for cooling water, more thorough tre'iment of common cause failures, and linkage between human actions and potentiat instrumentation failures. Basci on the most recent update, BGE estimated that Calvert Cliffs' core damage frequency including extemal events to be 3E-4 for Unit 1 and 4E-4 for Unit 2.

There was considerable staff interest in i a licensee's treatment of common cause failures and the linkage between human actions and potentialinstrumentation failures since the licensee's approach appeared to be different from common industry practice. The staff is currentbj considering what follow up actions should be pursued.

Sincerely, Original Signed by:

Alexander W. Dromerick, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - l/II Office of Nuclear Reactor Regulation Docket Nos. 50-317 1 and 50 318 g

Enclosures:

1. List of Attendees P. Viewgraphs b )

cc w/encls: See next page DOCUMENT NAME: G:\CC12\MTG-12.08 To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure 'N" = No copy n CFFICE PM:Pj)%p 1 f)M lE LA:PDI 1) Q lE DIPDI 1 M l l I l NAME- ADrfgIrN M SL 1 t t L f.7 S8ajwe O # O( . .

DATE M/df/9f 12/ d /97 91 - /^'- ll $ \ 9 $

12/ /97 12/ /97

-Official-Record Copy y]OS]}

9001130278 900105 PDR ADOCK 05000317 ll**2 llllllll o . . -

P PDR

4 4 January 5,1998L LICENSEE: Baltimore Gas and Electric Company Facility: Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 $

SUBJECT:

SUMMARY

OF THE DECEMBER 8,1997, MEETING REGARDING THE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT - i NOS.1 AND 2 - PROBABILISTIC RISK ASSESSMENT r On December 8,1997, the NRC end Baltimore Gas and Electric Company (BGE), the licensee for Calvert Cliffs Nuclear Power Plant, held a meeting in One White Flint North, Rockvilie, MD to discuss the Calvert Cliffs probabilistic risk assessment (PRA). Encloture 1 is a list of -

attendees. Enclosure 2 is a copy of ths viewgraphs distributed at the meeting. .

The purpose of tho' meeting was for BGE to present an overview of their updated PRA and to _

discuss unique key features of the study that may have contributed to Calvert Cliff's calculated-core damage frequency. _ BGE stated that the PRA update included extemal events, the inclusion of two additional diesel generators, the removal of the high pressure safety injection pumps

= dependancy for cooling water, more thorough treatment of common cause failures, and linkage

. between human actions and potentialinstrumentation failures. Based on the most recent -

update, BGE estimated that Calvert Cliffs' core damage frequency includbg extemal events to be 3E-4 for Unit 1 and 4E-4 for Unit 2.

' There was considerable staff interest in the licensee's treatment of common cause failures and the linkage between human actions and potentiel instrumentation frilures since the licensee's approach appeared to be different from common industry practice. The staff is currently considering what follow up actions should be pursued.  ;

r Sincerely,

, Original Signed by: l 1

Alexander W Dromerick, Senior Project Manager i Project Directorate 1-1 Division of Reactor Projects - l/II ' l Office of Nuclear Reactor Regulation  ;

l Docket Nos. 50 317 l and 5F,-318 l l

Enclosures:

1. List of Attendees l
2. Viewgraphs j

'cc w/encis: See next page

]

DOCUMENT NAME: G:\CC1-2\MTG-12.08 i To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure L"E" = Copy with l

- att:chment/ enclosure ~ "N" = No copy n

, DFilCE PM:Ppki ffM lE LAPDlaibM IE D PDI-1 sA/1 l I I NAME- AD M M M Stittl4 7 SteJwe NW_

  • ^^*

DATE M/sT/W 12/:N #97 ll 5l U 12/ /97- 12/ /97 Official Record Copy -

a tw_ -~---l y- 7 --- -en .w- , - - -

.er- m - - . - - , - .-,,--.,,w -

N -

~ . . . ~ .. . - . - . - - - -

?.- nace y S UNITED STATES l t y NUCLEAR REGULATORY COMMISSION: 1

  • -i

,, wAswiweTow,o.c. -

January 5 1998 4 * ***.

OCENSEE: Baltimore Gas and Electric Company

~ FACILITY: Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 -.

SUBJECT:

SUMMARY

OF THE DECEMBER 8,1997, MEETING REGARDING CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 - PROBABILISTIC RISK ASSESSMENT ,

On December 8,1997, the NRC and Baltimore Gas and Electric Company (BGE), the licensee for Calvert Cliffs Nuclear Power Plant, held a meeting in One White Flint North, Rockville, MD to discuss the Calvert Cliffs probabilistic risk assessment (PRA). Enclosure 1 is a list of attendees. - Enclosure 2 is a copy of the viewgraphs distributed at the meeting.

The purpose of the meeting was for BGE to present an overview of their updated PRA and to discuss unique key features of the study that may nave contributed to Calvert Cliff's calcuinted core damage frequency. BGE stated that the PRA update included extemal events, the inclusion

. of two additional diesel generators, the removal of the high pressure safety injection pumps dependancy for cooling water, more thorough treatment of common cause failures, and linkage between human actions and potential instrumentation failures. Based on the most recent -

. update, BGE estimated that Calvert Cliffs' core damage frequency including extemal events to be 3E-4 for Unit 1 and 4E-4 for Unit 2. ,

There was considerable staff interest in the licensee's treatment of common cause failures and ,

the linkage between human actions and potentialinstrumentation failures since the licensee's -

approach appeared to be different from common industry practice. The staff is currently considering what follow up actions should be. pursued.

Sincerely, Alexan r W. Dromerick, Senior Project Manager Project Dir3ctorate 11 Division of Reactor Projects - t/11 Office of Nuclear Reactor Regulation  !

Docket Nos. 50-317 and 50 318

Enclosures:

1. List of Attendees

- 2. Viewgraphs cc w/encls: = See next page l

l

i

- Meeting Summary Distribution E-Mail (Encicsure 1) Hard Coov (all enclosgga)

8. Collins /F. Miraglia - - Docket File :

R. Zimmerman (RPZ) .

PUBLIC B.. Boger PD#11 Reading

- A. Dromerick OGC S. Little -ACRS ~

T. Martin (SLM3) _

M. Cheok cc: Licensee & Service '

C. Liang Ust B. Giardina ,

G. Parry.

M. Wohl

'J. Flack-A. El Bassioni B. Palla N. Saltos F. Bower J. .Trapp B. LeFave B. - McCabe (BCM) -

D

. ,. _ , . - - , . . _ . . .m, . .~,, ;_ . . . . . _

'.. -  : f.  : +3 3 4 2

=.

_g v.

1 Meeting Summary Distribution . + ,

y- , ,

E-Mail (Enclosure 11 dats.Q.oov (all enclosuresE -

I- S Collins /F. Miraglia .

Docket' Fili [ ' r ' -

R. Zimmerman (RPZ) '

^ PUBLIC. -

j B. Boger PD#1-1 Reading -

'.' A. Dromerick i .OGC~

S. Uttle "

ACRS T. Martin (SLM3) -

M. Chook cc: Licensee & Service '

C. Liang List

. B. Giardina'-

G. Parry .

Mc Wohl J. Flack - .

A. ' EbBassioni 3 B. Palla .

N. Saltos .

F, Bower J.,Trapp B. LeFave

= B. McCabe (BCM)'

s ,4a _:

4

. -g ,

I We

- . .-~ _ _ _ . _ _ _ - _ _ , . _ . _.. . _ _ _ . _ _ _ _ _ . _ _ , . _,

g. .,

=

. Baltimore Gas & Electric Company; - Calvert Cliffs Nuclear Power Plant Unit Nos.1 and 2:

- oc:

Precident . Mr. Joseph H. Walter, Chief Engineer '

' Calvert County Board of.'

' Public Service Commission of Commissioners - Maryland t

-175 Main Street . Engineering Division Prince Frederick, Md 20678 - 6 St. Paul Centre .  ;

Baltimore, MD 21202-6806

  • James P. Bennett, Esquire  :

Counsel Kristen A. Burger, Esquire

. Baltimore Gas and Electric Company '

Maryland People's Counsel

. P.O. Box 1475 6 St. Paul Centre Baltimore, MD 21203 Suite 2101  ;

Baltimore, MD 21202 1631

' Jay E. Silberg,- Esquire Shaw, Pittman, Potts and Trowbridge Patricia T. Sim;e, Esquire 2300 N Street, NW. Co Director.

Washlagton, DC 20037 - Maryland Safe Energy Coalition P.O. Box 33111 Mr. Thomas N. Pritchett, Director . . Baltimore, MD 21218 NRM Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatell 1650 Calvert Cliffs Parkway NRC Technical Training Center Lsby, MD 20657-4702- 5700 Brainerd Road Chattanooga, TN 37411-4017 Resident inspector clo U.S. Nuclear Regulatory '

Mr. Charles H. Cruse

Commission Vice President- Nuclear Energy P.O. Box 287 - Baltimore Gas and Electric Company St. Leonard, MD 20685 Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway i

Mr. Richard I. McLean, Manager - Lusby, MD 20657-4702 Nuclear Programs

! . Power Plant Researen Program

, - Maryland Dept. of Natural Resources Tawes State Office Building, B3 -

-- Annapolis, MD 21401

, Regional Administrator, Region i U.S. Nuclear Regulatory Commission P

475 Allendale Road

- Kmg of Prussia, PA- 19406 L

s l

.)

4

?

i i

LIST OF AIIENDEES BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 AND 2 l DECEMBER 8.1997 MAmt Ornanization AlexanderW. Dromerick NRR/DRPE John C. Lane RES Mike Cheok NRR/DSSA '

Gareth Parry NRRIDSSA Millard Wohl NRR/SPSB ,

Alex Dong BGE/PRA Group John Osbome BGE/ Nuclear Regulatory Matters Robert F. Cavedo BGE/PRA Group t Jeff Stone BGE/PRA Group Eric Schade BGE/PRA Group Biff Bradley NEl John Koelbel BGE/PRA Group John Flack NRR/SPSB Adel El Bassioni NRR/SPSB Steve Rockford BGE Bob Palla NRR/SCSB Nick Saltos NRR/DSSA/SPSB t Fred Bower - NRC,RI Jim Trapp -

NRC/Rl/SRA ED Rodrick RES/PRAB Bill LeFave NRR/SPLB I

Enclosure 1 7-y .* y--- y -

% ,,.m,

__ __ _ . .. _ _ _ _. _ = ._

r . ,

I

$,?*

ik Baltimore Gas & Electric Company

\

& W.

Calvert Chffs PRA f Li'

$id

@ Bruce Mrowca

.i December 8,1997 -

t 9

f

Objectives l~ ,

n Provide an overview of CCPRA

% a Identify key features potentially unique to CCPRA 15 ;

Nk

't t

Enclosure 2  :

. 4 l

l :PRA ModelStatistics Initiating Events: 318  ;

i Event Trees: 10 yr i

j e TopEvents:

. m...

277 i

i l n SplitFractions: 3,900 1

d: r. a FunctionalGroups: 3,100 .

l a Basis Events: 6,100

; a Components: 5,400 i

, 4 i

! j 5 a

i 1

j delStatistics (continued)

Human Action Designators
140 0

a Plant Specific Data y[~in Designators: 285 "a TotalSequences: 575,000

-3

.4 l

8 4 s:

'l 6 4.

1 3

1

-y,,.,ev.- e,7-rr-w--- ---yy- .-,-n,,,,,,,.-e 4,w,.-,,-owmeww-er,--,..wa-w-r,wg.w-e,,-,----w-,*m-v.w-,,m,mww.wwe,----...-- - - - - - - - - - - - - - - - -

a-. - - - - - ,

i Calvert Cliffs NuclearPowerPlant

,...n

c. Type:

. Tm 860MWe C E PWRs

,a Location: 40 miles south ofAnnapolis

~l:' y,a Commercist Operations: Unit 1: MV75 Unit 2: 4/1/77 Ultimate Heat Sink: Chesapeake Bay

  • n Containments: Large DryPost-tensioned

$ll n EmergencyPower: 5 ErnergencyDieselGenerators 2

i CCPRA Description 1

'n' Unit 1 model(Unit 2 differences are estimated)

$ n Large event-tree (RISKMAN 8.0) 4,Ij

  • Rule based
  • Extensive support system trh>dels
a Extensiveplant specific data y a HRA hybridmethodology n Includes Selsmic, Fire andHigh Winds PRAs 3

}

4

9 .

1

.) Quality Control

al Qualification Process fororiginatorandreviewers l

':

' Proceduralizedprocess forIndependent reviews on completeddocuments

] a Peerreviews performed on selected areas W

n Key inputs and Assumptions Database

  • StartedFall1995
  • 740 KeyInputs and 230 Assumptions
  • 2,300 References s

3 Completion of documentation is ongoing r i l

1 l mUnit 1 CDF Contribution

( ,

-d ,

' '((

40 %

N Fim

, 5 Selsmic g

EWind 6hW. .

OLOCAs j

D Flood 5 Transients 8%

M% 1%

l I

- , - - , - , , , . , w. ,, ,-a,, , .~,-a. r--, - - - , - - . ~ , , - . , , , , - - , - - , , , - - - - , , - - - - , - . - - - - - - - - - - - ~ - - - - - - - - - -

e 0 l ..lnitiating Events t -

. ' fn; Inlilating Events 318 fY

  • Flooding: 47
  • Fire: 177
  • Seismic: 30
  • High Hinds: 6 a impacts explicit andbounding a Common Cause on multiple train systems considered

\

hyRlant Model(Event Trees)

N

jn, Detailed electricalsystem models a Batterydepletion andLOOPrecoveryexplicitly modeled Entry into Functional Recovery Procedure explicitly modeled NEoP8S
=DA = S*DB " S *D on S *DD e S *(E 1 e S *E4 = S + E2e S *E3 e S)

NEoPoueXA = S *XB e S *XC o S *XD e S *(ewe S *EZ = S +EX= S

  • eye S) *(E5 eS* BUS 11+F6*S* BUS 14) plus Macros: NEPSAS, NEP8AL, NEOP8H, NEOP80 m HVAC dependencies for Control Room / Cable ,

Spreading Rooms, Switchgear Room, ECCS Pump  ;

Rooms and AFW Turbine Pump Rooms addressed l

10 l

l 1

/ '

Large Event Tree Method .

! <.- fj

.- . y The GeneralTransient (OT)

Event Tree has 216 branch l points. Each branch point

is represented by a fault i

tree top event.

This Eent Tree is uud:o ~~"""~~~"['"

directly quantify the Plant Damage States (PDSs). [,"""""l,[',~~~~

l ff

.Large Event Tree Branches M Each Large Event Tree branch represents a train or

[ plant model function. 'Ihese are some top events which h appear in the GT Event Tree:

rm

  • 51 Salt Water llender 11 Malntains Adequate Flow I
  • AV . MTC adequate to prevent an ATWS event I
  • El . 120 VAC Panel 11 remains Energlied in the Short Tenn
  • EW. 120 VAC Panel II remains Energized in the Long Term
  • OP. Off site power not lost as a result of the Unit trip aCD. The operator aligns Fire Protection to CCW to suppon head tank make up
  • IlX The operator provides adequate AFW control ft

y AFWSplit Fraction Binning

, r<The AFW pump related top events (F7, TF TO, and F9) fhave these associated split fractions:

Top System Analysis Plan:Model Event Split Tractions Split Tractions t

T7 6 6 l TT 12 8 TO 40 12 T9 48 32 These pump related top events are highly conditional due to maintenance and common cause.

13 i AFW Turbine Driven Pump 11 (TF) eiSplit Fraction Binnin.g i  ; 1 TFS rVa 866 0 TF1

.['. 3

_ Jg e=

Corrt)~ FC4TAS*(F04* F73UPP47=8)

Tfi FC*STJ=STG=S* F7SWP TF5 FC=STJ=STG=S17=s TF7 FC=STASTG=FT7s3

~fv3 TF3 FCm87 #8*F73UPP t ~Tf2 Corrd) FC47#(F0=8*h)

I a. U2 FC=STJ4TG=S* F7SUPP 1F6 FCsSTAFTG=ST7ss TF8 FC=ST#FTG=FT7sS

~fF4 TFT #D@~f75iTPP TF1 TTT PC477=PTJ4104 TA TTF FC477=FTRVDas TFs TF11 FC477dT541GW

~ TpC TFt2 h>FFo*

~iFT tva 1 14

OperatJr Aligns Portable V Ventilation to SWGR Room (HZ) e 4

x HZ$ NOTRIP

- 24 Nff *L OSOf0P= 8 *0Ce 8 *00* $*Ab s*DC* S

, Hl1 OP=8'(OC=$*0De$rA>8HbF'( M1NNTDC*SHbS*L MfT*L@W HZ2 (OC*F%D*FTAb8H5aF'( M1Nff,*DC*SHEoPeSH8=$*( WrT*LOS0W 23 HbF'(OC=F*oD*FrDCoS*NEOP48Wa8*( WrT*LOSOW H2e (OC.F 00 FrAbBHbf *( MtNfTPDCe8Hb$*( WIT *LOS0W j

HZ7 HS*F*DCs$*H>l*(. NtToLOS0W HIF 1 16

) l Human Actions --

. t p ni SLIM-MAUD Hybridmethodology lb

  • Success Likelihood coemcients are established using the Human Cognitive Reliability (HCR) Model and EPRI Operator Reliability Expen*ments (oER) data
  • Three phase operatorresponse: Identification, diagnosis and perforinance
  • Knowledge-based, rule-based or skill based cognitive pitnss selected forphase 23 performance shaping factors are evaluated through table.

topinterviews f6

p Performance Shaping Factors g.

gy)la Rush Perceived by Operators

.'.'S Operator Training and Experience ProceduralDirection Avallable to the

[ Operator 1 a Personnel Availability and Communica! Ions a Plant Indications ti

Performance Shaping Factors

+(continued) nN Consequences Associated with the Action Operator Confusion a Equipment Location ft

i p sHuman Actions (continued) a

. ,a Bounding scenados and available time are rf determinedpriortoInterviews Application constraints are developed a Entry Into functional recovery procedure (EOP-8) is monitoredin the plant model degradation factor j applied when appropriate a Instrumentationpowerdependencies determined-loss of minimum dependencies result in failure of action ff ulction Failure / Success Logic i

hk(:J Idenklication Disposis f%rfwnance Action success

' Action Fallure P, (r.p,)(1.p,yp, P,

Action Fallure (1-P,)P, P,

ActionFallure P,

20

)

j Closed System Leakage tai Historic CCWand SRWsystem leakage considered

, , ni R3 lief valve andhandvalve transferopen failure rate i modeled a Recovery times based on leakage distribution

] n Domineralized Water, Condensate, Fire Protection, Salt Water (for SRWonly) make up systems modeled i

Et

. Common Cause J.

'. . jm Approach: Muldple Greek Letter a EPRI TR-100382 (1992) used to develop the prior distributions a EPRITR 102747methodusedto customized a 19 Demand and 15 Operating equipment types considered i

~

n

d y Common Cause (continued) g Demand Failures Opersting failures e .- + na AFWPumps(Turbine 4dnis, a AFWPutnps(Turbine drivers,

%". .- motor 4 rivers, pumps) motor 4 rivers, pumps 4a EmerpencyOsselGenerators a Emerpency04selGenerators a Bettenes a Bettenes a HVACFans a HVACFans AC units a ACUnits a AirCompressors a AirCompressors a Relays a Releys(selectedcases) a ControlVsins,Ostnpers a ControlValves press &lemp i a SolenoidVelves a Instrument AirCVs felttocycle a PORVs a MainFeedwaterPumps a Msin Steam Safety Valves a BetteryCherpers a MotoroperatedVsfees a VoltageRegulators a Check Valves a inverters a Trip LogicModules a Transformers ($00KV,13KV, a ControlSwitches 4KV) a Breakers, Reactor Trip Bkts a UVCods ss t

&AFWPump Common Cause su Common Cause Groups i Q

  • Allpumps (turbine andmotor) including cross-

~

connect (excludes drivers)

  • All Unit 1 pumps (turbine and motor) (excludes ddvers)
  • Unit 1 and 2 motorpump drivers Unit 1 and2 turbine pump drivers un l

l l

+ Spurious Safety System Actuation fa SSSA:Inadvertentactuationof:

  • Engineered Safety Features Actuation System (ESFAS)

' Auxiliary Feedwater Actuation System (AFAS)

I,

  • ReactorProtection System (RPS) a Results from the failure of two 120V AC Buses a impacts MFW, PORVs, EDGs and AFW 36 4

Y  %

bLoad Faults

. i.y a Likelihood of a load fault and the probability a 4KV or gg 480VAC breaker opens on demand is modeled DC power dependency for 4KV breakers to open on

, demandisincluded

  • a load fault with bss of DC results in the loss of an entire facility a Significantissue forFire and Floods 1 "

. a-

. Flooding (Internal Pipe Breaks) y a Minimalscreening-rooms screenedwhen
O
  • No Rooding source exists L:)
  • No PRA Components and no Roodpropagation out of the

' room a a No floodinitiating events were screened a Flood scenarios are binned to liko-impact Initiating events (47inillating events) a Elecidcalimpact ofload faults on buses considered i

t

,7

( Flooding (continued) re j "di "Calvert Cliffs performed a more ngorous analysis that accumulated the combined effects of needy a dozen flood scenados (rather than using a sedes of n Individualscreening arguments) ... Itis unclear

,l whether the flooding contdbution willbe significantly higher for otherplants if they use the screening i approach employed by Calvert Cliffs." (NUREG 1560

} Draft, page 3-53) i se

, Cyclical Systems ln Systems with multiple demands are carefully l fassessed

<r u)

  • CWand SRWHead Tanks

1 n

l Fire PRA i, n Routh' and analysis of 5,500 cables

":;

  • Eva.. sted cables for allPRA components ni Impact of smoke explicitly considered:
  • Human actions failure rates accelerated by area: Aux Bldg.

Turbine Bldg, CR, Intake, Yard I

  • Inadvertent Halon actuation evaluated
  • CR evacuation evaluated a Electricalimpact ofload faults on buses considered a Cross zone fires modeled II c

. - r,. < .

. -. . - _ _ . _ . . _ _ = - -._ -. .- - .. . - . _ - _ _ - .

.e .

a l

1 i Seismic PRA

'  :.i in includes a sunogale to address screened components

' considers human actionImpacts e uses three models: Small. medium andlarge selsmic events

  • each model used dMerent human action failure probabilit4s l

d 31

iS ie smic HRA

$ O : 1

'4

. P5T Influtare Tnckr es $riamic g ClaselficeWon 1 ie< ..

ee J:: .

9' e . . me.**t ...

  • Ies es.

ei se .

, es es se is se is as s s n .m..

shen team &# tseem * * *

  • W tene Actow 32

l . Draft NUREG 1560 Issues .__

l

~ JAddressed by latest model:

n' PORVs are credited for feed and bleed

'It

  • requires both PORVs to open
  • earfy and late init!alion is considered I a Steam Generatordspressurizationis considered a ATWS recovery via boration is considered

)

]

3J l iSummary ln CCPRA is a comprehensive PRA

]ml Minimaluse of screening techniques a Modelevolutioncontinues

  • New DieselGenerators are beingincorporated

,

  • Development of KeyInputs and Assumptions continues
  • Document close-out continues a CCPRA contains conservatisms (e.g., toom H/U

, rates) 4

  • There is a carefulbalance betroen defendebuity and conservatism 34

l l Fina! Thought

,.4',l Qi Scope and qualityimpact the bottom line number.

1 .m.

l 4

I t '

4 38 1 l

I l

l l

l 1

1 l

I I

CCPRA-Key Input 142 T1tle: SRW Head Tank !I and 12 Iml ladication loop power supplies Plant Fumeties: Design Systesa: SRW PRA Analpls: Human Actions Applicetles: Either lYO9 or lY10 through IR01 A provkle power to the imi transmitters for SRW Tiend Tank 11 and 12. The level indicator for 11 SRW Head Tank (1.LIA.1$79) is powered from lYO9. The levelindicator for 12 SRW Head Tank (1.LIA.1565) fd powe:M from lY10, However. these rower supplies do not power the armurdcators which are powered from a DC bus. SRW Head Tank level 11 and 12 alarm on iCl3 mi Kl9 and K23.

Revlelost 2 Regiolom Dewription: Clari$ed indicator power supply Originaten JLS Date: 6/23/96 Reviewen BBM Dates 6/23/96 References Type Reference Revision Desedption Section Page rue-6., pt?0 spinoos 3 tm,Darwa I savia hur Head Twa 14ve:

ILTIS79 Drawing 98701 &lto006 4 1my Dayun i2 Sav;ee Weser Head Tasa ILTI$63 Applicable CCPRA Documents Docuenent Description .

Type WlETET Op povides adeg to SRWCCW using rr epsen whhis H sminima Humes Amien wiuW7 Or Fw6de udew to SRWCCW using 77 eyame whhis 120 mimam Hamas Amien Mitsh(T Or povides adeg to SRW using 8W estan whhin 1$0 minnas Hanan Amien MitsW3 op povide adeg to skW using SW'epuen wrin ill ownma Hanne Aaiao

)

Page i

4 .

. e . l

'. l CCPRA Assumption 119

= l

Title:

Time available for early FP Syssem make up of SRW/CCW is 95 mintues Plant Functies: Operadon System: SRW PRA Analysis: Human Actions Aneumpilos: It is assumed that 93 mintues is available froen the imidation of 4 plant trainsient to matteg SRW or CCW using the Fire Protection System.

Applicatlos This assumption is used to determine operator response time, Bases: RAN 94 008 Secdon 4,10 establishes several operator response dmes and their related failure probabilities. The period of 95 to 119 mintues represents the peri:4 of time with the second highest likelihood of success. This aedon (BlIETET) was inteniewed for the limidng time of 95 minutes. The operator was then asked how much time he would neeJ in order to haw more than enough time to complete the action, which was determined to be 120 minuta (This time >

was used for BliEFMT). Therefore, Bl!EFET was given the timeframe of 93119 minutes to fill in the timeline up to 120 minutes although, as stated before, it was intervi;wed for the limidng time of 95 minutes.

Attachment C of this RAN identifies the istmalad operator action sequence etevents for this medon as determined during an interview with a Senior Control Room Operator (Bill Raia=).

The sedmated sequence for the longest sucomes path is as follows:

00:00 Initiator ImofOff sitePower 00:10 Completes EOP 0 00;6f. Responits to Main Control alarms after plant stabilires 00 70 Completes Alarm Manual ICl3 Windows K 19 and K23 actions, enters AOP.7B 00;80 Complets AOP.7B, Loss of SRW, enters 0115 00.93 Completes 0115 Fire Protection alignment action Revision: 1 Redslon

Description:

Clarified the maximum time available Originator: EAS Date: 10/97 Reviewer BBM Date: 1/10/97 References

,7ype Pefereuce Revision Descripdon Section Page ILAN Moos 0 Evaluation of sRW and CC Meeup 4 10 s3 ILAN 4 o00 0 Evaluation of sRW and CC Mdeup Amadenent C l2 Applicable CCPRA Documents Dxument Description Type NitrET o rprovides ade up to sRWCCW using IT tystem within 9) adnutes Heman Action Page I

RELIABILITY ENGINEERING l REU QUALITY RECORD l

IIUAfAN ACTION DATA DESIGNATOR BHEFET Operator provides make-up to SRW or CCW using FP

! System within 95 minutes -

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ORIGINATOR:

DATE REVIEWER:

DATE APPROVAL:

DATE i

RAN: H-026 Rev 0 RSN: 132 A:\REV0\HAl\BHEFET. DOC l

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A. Description of Requested Human Action Analysis

1. Task to be performed l i

operator action is required to alien the Fire Protection System to the service Water or Component Cooling Water system to avoid depletion of the head tanks for the system in  !

quest 60n within 95 minutes of a plant trip.  !

2. Timing bases for performing action ,,

l a) Perform action within 95 minutes This response time P based the second blehest operator likelihood of suooess ,

time bin used in the % valuation of SRW and CC Makeup *. RAN 94-008. This time bin is establishing the constraint for the use of this adion. The timing of this action is considered self<ieterminin0 (AssumpC'vi 135) and is based on belancing the likelihood of success of the human action with the cost of the seloded timing to the plant.  !

SRW and/or CCW isakage rates which fail these systems in less than 95 minutes can not be recovered by this action.

i Assuming that the alarm for the service water or component cooling head tart occurs early in the event. It is estimated that the action to respond to the alarm could be initiated as late as 65 minutes into the event leaving 30 to 64 minutes to diagnose and perform the required actions. During the first 85 minutes it is possible that the operators would be responding to higher priority plant lesues, if the head tank alarm occur after 65 minutes, the operators will likely have the full 95 minutes to dia0 nose and perform the action (for leakage rates which oeuse fall these systems in 95 minutes or greater) a;nos competition with other plant issues would be minimal.

See Assumption 119 for the timing bases.

3. Failur Ateria

! Operator fails to align Fire Protection to either Service Water or Component Cooling ,

Water within 95 minutes following a plant trip.

4. Scenerlos for which the action is evaluated After a plant transient. tt.e fc' lowing events occur.

Pre trip SRW/CCWleak exists Domin Water system is providing make-up 00:00 LOOP occurs EDGs start and run successfully 00:00 to 00:65 SRW or CCW head tank alarm (worst case timing)

<01:35 < Analyzed action perfomwd>

RAN: 96 034 Rev 0 R8N: 132- A:UtEV0WAhBHEFET. DOC f

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Thors are three possible success paths for this action:

  • While veritying SRW/CCW Pump te start earty in EOP 02,
  • Loss of Offsite Power *, ,

Section IV. Step E.1, the operator recognizes an SRW or CCW head tank low level  !

alarm, i I

e The operator may also recognize the head tank low level alarm later in EOP 02, i Section IV, Step 8.I. (Key input 171) while determining Aux System availability.  ;

  • And lastly, the alarm may also occur after the plant stabilizes. Once the plant is i fotumed to amfe shutdown condition, the operators will look for any irregular alarms  :

and attempt to clear them. l The operator will attempt to respond to the Low Servios Water Head Tank Alarm in accordance with the Alarm Manual (Key input 152), which directs him to AOP.78.  :

AOP.78, in tum, directs him to 01 15. 01 15 provides the specific procedural  ;

direction to perform this sollon (Key input 143).

The operator will attempt to respond to the Component Cooling Water Head Tank j Alarm in accordance with the Alarm Manual, which directs him to AOP.70. AOP.

7C, in tum, directs him to 01 18. 01 16 provides the specific procedural direction to  !

perform this action (Key input 214).

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8. Preceding related actions:

There are no proceding related actions assoolated with this action. ,

S. Constralnts This adlon ma'y be used in all event sequereces, except:  !

e if a LOCA, SGTR, or SLS exists

  • - entry into EOP 8 The EOP 00 factor must be applied to use this action under these exceptions .

(Assumption 09). '

This eation was interviewed for a LOOP, thus implementing EOP 02, it is assumed that if the operators enter another EOP for the conditions above, the distractions will be si0nificantly worse along with a hi0her rush factor. The guidergs may also be i si0nifloantly effected when entering into the EOPs for the conditions above.-

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' RAN: 98428 Rev 0

, R8N: 132 A:\REVO\HAhBHEFET.00C u

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B. Specific actions required and equipment dependencies:

INDICATION:

ACTION SPECIFIC SUPPORT REFERENCE EQUIPMENT SYSTEM Low SRW Head Tank alarm 1 ANNK19 (1C13) 1YO9,2001 & 1R01A KEY INPUT 142,402 11 or 12 SRW Head Tank 1ANNK23 (IC13) 1Y10,2001 & 1R01A 1YO9 & 1R01 A 1LIA1579 (1C13) ,,

1Y10 & 1R01A KEYINPUT 108 ILIA 1565 (IC13) 11 CCW Head Tank low 1 ANNK17 (IC13) 1YO9,2001 & 1R01A KEY INPUT 198,402 level alarm 1LIA3820 (1C13) 2001 KEY INPUT 108 INDICATION DEPENDENCIES = SRW IF (1YO9 on 1Y10) AND 1R01 A AND 2D01 ARE AVAILAaLE THEN THE HUMAN ACTION SUCCEEDS. (1YO9 AND 1Y10)is usED slNCE THE SRW HEAD TANKS ARE NORMALLY Cross-C0"NECTED, CCW IF 1Y09,2001, At.D 1R01 A ARE AVAILABLE THEN THE HUMAN ACTION SUCCEEDS.

DIAGNOSIS:

ACTION SPECIFIC SUPPORT REFERENCE EQUlPMENT SYSTEM Determines that NONE Fire Protection KEY INPUT 143, Condensate and Water System KEY INPUT 214 Domineralized Water are not available and that Fire Protodion must be used for makeup, DI AGNOSIS DEPEND 5NCIES = Fute PRoTecTxW WATER SUPPt.Y PERFORMANCE:

ACTION SPECIFIC SUPPORT REFERENCE EQUlPMENT SYSTEM 1, ENSURE SHUT the OHVDW 248 N/A Key input 143, foliowin9 valves 1HVCD -429 Key input 214 (1HVCD 145 for SRW only)

2. INSTALL a temporary fire HS 1216 None Key input if 3, hose Key input 214 3, FLUSH FIRE Main if plant HS 1216 Fire Protection Key input 143 Conditions permit (N/A for Water Supply CCW)

RAN: 96-026 Rev 0 RSN: 132 A:\REV0\HAf\BHEFET. DOC

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B. Specific actions required and equipment dependencies (cont'd):

ACTION SPECIFIC SUPPORT REFERENCE ECUIPMENT SYSTEM

4. CONNECT hose to FIRE 1HVCD -457 None Key input 143 HOSE CONN ISOL valva Key input 214 and PERFORMIndependent verification i .-
5. OPEN valves to make up 1HVCDw57 Fire Protodion l Key input 143 water to the head lank (s) (1HVCD.144 for Water Supply SRW only)

HS 1216

6. THROI 'LE valves to 11 1HVSRW.106 Fire Protection Key input 143 and 12 SRW Head Tanks as 1HVSRW.114 Water Supply necessary (SRW onl/)
7. Open makeup to 1HVCD 145 Fire Protection Key input 214 component cooling water Water Supply system (CCW only)
8. Periodically MONITOR 1LG1579 Fire Proledion Key input 143 the SRW Head Tanks AND 1LG1580 Water Supply FILL as necessary (SRW only)

PERFORMANCE DEPENDENCIES = FIRE PMoTEcTION WATLR SUPPLY RAN: 95426 Rev 0 RSN: 132 A:\REVO\HAhBHEFET. DOC

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D. References  ;

CCPRA k:z; % 89 Entry into EOP 8 increases the failors probability o(some human  !

actions 6y a factor o(5.0 l CCPRA E: a % 119 Tune available for early FP System makup of SRW/CCW is 95

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mistues .' .

j CCPRA k r; e 13! Self Cr;; ' ' ; Human Action Basis CCPRA KeyInput 142 SRW Head Tank 11 and 12 level indiostion loop power supplies CCPRA Keyinput - 143 AOP.7B directs the reAlling SRW Hand Tanks with the FP System

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per OI 15 - .

[CCPRA Keyinput 152 Alarm Manual 1Cl3. Windows K 19 & K 23, SRW head tank low level alarm CCPRA Keyinpe- 171 EOP 02, ao specine SRW/CCW low head tank gu. dance in EON CCPRA Keyinput - 198 CCW Head Tank level indication power su/91ies '

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, CCPRA Keyloput 214 AOP 7C Directs Operator to Fill CCW Head Tank With Fire System per 0116 l

CCPRA Keyinput 402 1ROI A(2RS! A) is powered ham (l(2)Yo9 or 1(2)Y10) 2 i

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n MAN: 96426.Rev 9

-R$N:132L A:\REV 4 ULAR BHEFET . DOC

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