ML20207T454

Forwards Slide Presentation & Attendee List for 870304 Meeting W/Util,Bnwl & Inel Re PRA Application Program. Comments on Program Requested

ML20207T454
Person / Time
Site:	Oconee, Mcguire, McGuire, 05000000
Issue date:	03/12/1987
From:	Brownlee V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Tucker H DUKE POWER CO.
References
NUDOCS 8703240034
Download: ML20207T454 (47)

Text

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MAR 121967 Duke Power Company

_yTh: .Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street Charlotte, NC 28242 Gentlemen:

SUBJECT:

PRA APPLICATION PROGRAM FOR OCONEE AND MCGUIRE On March 4,1987, members of the Region IL staff and consultants from Battelle and Idaho National Engineering Laboratory 1riefed members of your staff at your corporate office on the NRC's plan to develop and conduct a PRA inspection. A copy of the slide presentation and a list of attendees is enclosed.

I . trust this meeting was worthwhile and that the end result of our mutual effor*.; .1111 be beneficial in contributing to risk reduction. We would appreciate any comments you may wish to share with us regarding this program.

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Sincerely, Or197'nal i

signd b3 feebles Virgil L. Brownlee, Chief Reactor Projects Branch 2 Division of Reactor Projects

Enclosures:

1. Meeting Slides

2. List of Attendees c w/encls:

, . Tuckman, Station Manager L. McConnell, Station Manager (bcc w/encls - See page 2)

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l 8703240034 870312 PDR ADOCK 05000269 P PDR 8

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Du e Power Company 2 MAR 12137 bec w/encls:

' @ Hood,-NRR k . Pastis, NRR F. Jape,~.RII t,[ERC Resident Inspectors for Oconee and McGuire

~ State of South Carolina State of North Carolina bcc w/o enc 1 1:

ore, PNL Wright, INEL RII RIL p RII FJap AH r t 03/[D /87 03/it/87 3////87

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, . 1 ENCLOSURE 2 LIST OF ATTENDEES MARCH 4, 1987 Frank Jape NRC, Region II Jack Bryant NRC, SRI-Oconee Pierce H. Skinner NRC, SRI-Catawba J. S. Warren Duke - Licensing George Ridgeway Duke - ONS - OPS Craig Harlin Duke - Oconee - Compliance M. A. Haght Duke - NPD - Licensing i

R.'L. Gill Duke - NPD - Licensing Paul Guill Duke - NPD - Licensing Bryan Dolan Duke PRA Norma Atherton Duke NPD Compliance (MNS)

Ken Cavody Duke Nuclear Engineering W. H.'Rasin Duke Power Company L. R. Davison Duke - QA M. S. Lesser NRC, Catawba S. Guenther NRC, McGuire T. A. Peebles NRC, Region II W. T. Orders NRC, SRI - McGuire A. R. Herdt NRC, Region II, Engir.eering Branch B. Hillman NRC, Region I Bryan Gore Battelle - PNL Trung V. Vo Battelle - PNL Ron Wright INEL

ENCLOSURE 1 6

./

PRA BASED APPLICATIONS TO INSPECTIONS PRESENTED TO:

DUKE POWER COMPANY OCONEE MCGUIRE MARCH 1987

1 AGENDA DUKE POWER COMPANY MEETING 1

MARCH 4, 1987 INTRODUCTION ALAN HERDT PRA APPLICATIONS PROGRAM FOR RON WRIGHT INSPECTION OF NUCLEAR POWER PLANTS (MCGUIRE)

RANCHO SECO INSPECTION GUIDANCE BRYAN G0RE PRA BASED INSPECTION GUIDANCE FOR TRUONG V0 OCONEE UNIT 3 BRYAN GORE FUTURE INSPECTION PLANS FRANK JAPE FINAL DISCUSSIONS AND QUESTIONS ALL 4

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t l PRA APPLICATIONS PROGRAM i

l FOR INSPECTION OF l

i l NUCLEAR POWER PLANTS

) /

l RON WRIGHT l Idaho Na tional Engineering l Labora tory

[MBg5 Idaho, Inc.

INTRODUCTION o PROGRAM XISTORY AND CURRENT STATUS a

n o PROGRAM REQUIREMENTS AND PRODUCTS  :

o REGION II PLANTS l

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I PROGRAM HISTORY o PURPOSE - THIS PROGRAM INTEGRATES PRA -

INSIGHTS INTO THE NUCLEAR f POWER PLANT INSPECTION PROCESS o PROGRAM ATTRIBUTES *'

IDENTIFIES IMPORTANT PLANT SYSTEMS IDENTIFIES IMPORTANT COMPONENTS FOR RISK SIGNIFICANT SYSTEMS IDENTIFIES COMPONENTS FAILURE MODES IDENTIFIES COMPONENTS TO SPECIFIC INSPECTION MODULES PROVIDES IMPORTANT COMPONENT PLANT SPECIFIC CHECK 0FF LIST j t

i f

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! ADVANTAGES TO NRC l

1. THE PROGRAM IS A DIRECT RESPONSE TO THE COnnISSION'S POLICY AND PLANNING GUIDANCE WHICH CALLS FOR THE USE OF PRA IN SETTING INSPECTION  !

PRIORITIES.  :

2. INSPECTIONS WILL BE BETTER FOCUSED ON EQUIPMENT WHOSE FAILURE HAS THE GREATEST IMPACT ON PUBLIC .

RISK.

3. INSPECTORS CAN MANAGE THEIR INSPECTION TIME BASED l ON THE IMPORTANCE OF SYSTEMS AND THEIR COMPONENTS. THIS SHOULD RESULT IN THE MORE EFFICIENT USE OF INSPECTION TIME.

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IMPORTANT FEATURES OF THE PROGRAM

1. IT IS CONSISTENT WITH AND CONSTRUCTED AROUND THE I CURRENT IE MODULES. ,

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2. IT IS STRUCTURED FOR USE BY RESIDENT AS WELL AS REGION BASED INSPECTORS.

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3. IT COVERS ALL OF THE COMPONENTS AND ACTIVITIES THAT CONTRIBUTE SIGNIFICANTLY TO PUBLIC RISK. '

4. THE PROGRAM CAN BE APPLIED TO ANY FACILITY FOR WHICH A PRA HAS BEEN DEVELOPED.

5. THE PROGRAM CAN BE USED BY NRC INSPECTORS WITHOUT THE NEED TO CONDUCT A DETAILED REVIEW OF THE PRA. j i

6. A CLEAR DEFINITION OF WHAT CONTRIBUTES TO RISK IS PROVIDED WITHOUT REQUIRING PRA EXPERTISE.

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PRA APPLICATIONS PROGRAN FUNCTIONAL AREAS o PRA BASED INSPECTION GUIDARCE ..

o GENERIC BASED INSPECTION GUIDANCE i

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I PROGRAM STATUS l o INEL INDIAN POINT 2 SEABROOK ZION -

HADDAM NECK (CY), GENERIC o BNL LIMERICK INDIAN POINT 3 l

SHOREHAM MILLSTONE GRAND GULF o PNL OCONEE i

t REQUIRED INPUTS o PROBABILISTIC RISK ASSESSMENT -

o SYSTEM DESCRIPTIONS o TECHNICAL SPECIFICATIONS o TESTING PROGRAMS (LIST OF TITLES) o MAINTENANCE PROGRAMS (LIST OF TITLES) o EMERGENCY OPERATING PROCEDURES i o CHECK OFF LISTS (LIST OF TITLES) j k

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I BASIC PRA PLANT PROGRAM DESCRIPTION o INPUT PRA SEQUENCES o DETERMINE EVENT IMPORTANCES .,

o SYNTHESIZE SYSTEM IMPORTANCE o RANK SYSTEMS o FOR TOP SYSTEMS, PROVIDE COMPONENT AND FUNCTION ANALYSIS

_ . . - _ _ _ _ _ . . , , - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - = ' ' ~ ' ' - ' - ' ^ ^ ^ ^ - ~ ~ ' ~ ' ^ ~

TABLE 16. MOST If5NNtTANT SYSTEMS Importance for Importance for:

Codea Name Public Health Plant Damage RHR Residual heat removal .83 .59 EP Electric power .11 .07 CS Containment spray .08 .00 AFW Auxiliary feedwater .07 .16 RC Reactor coolant .01 .07 .

RP Reactor protection .01 .13 SI Safety injection .00 .08 CC Component cooling .00 .06 MS Main steam .00 .03 -

a. These codes agree with those in Tables 14 and 15.

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TA8LE 1A. RESIOUAL HEAT REMOVAL SYSTEM FAILURE MODE IDENTIFICATION l

1 i The residual heat removal system is important for long ters recirculation cooling of the reactor following successful safety -

injection. The most important system failure is the V sequence, which  !

consists of a loss of coalant accident (LOCA) via an interfacing system. This LOCA thus b; passes reactor containment. Other system  !

failures rasisit in a loss of reactor long term recircul,jation cooling due to multiple RHR system failures.  !

Conditions That Lead to Failure l

1. Reactor Coolant System to RHR Pumps Isolation Valves IMOV-RH8701 and IMOV-RH8702 Fai1 Open i

These valves line up RHR suction from the reactor coolant system.

Failure of these valves would expose the RHR piping to RCS pressure,

, thus creating a leak path which bypasses containment. This accident scenario is several times more important to public health risk than j all the other failures in Tables 1 through 9 combined. Maintenance and surveillance of these valves should be observed or reviewed to I

minimize these failures. -

2. Operator Fails to Initiate Switchover from In.iection to Recirculation Mode or Fails to Stop Pumo at RWST Low Level This is the dominant failure for the low head recirculation mode and is significant for the high head recirculation mode. It involves the i proper interpretation of plant status and proper initiation and completion of full switchover to recirculation. Operator awareness of the criteria for switchover and adherence to emergency procedures are important.

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TABLE 18. IE SWOULES FOR RESIDUAL HEAT REMOVAL SYSTEM INSPECTION Failurea Module Title Components Mode 61701 Surveillance (Complex) RHR Pumps A, B 6 61726 Monthly Surveillance MOV-RH8701, 8702 1 Observation MOV-RH8700A, 8 3 RHR Pumps A, 8 6 62700 Maintenance RHR Pumps A, 8 4 62703 Monthly Maintenance RHR Pumps A, B 4 Observation 71707 Operational Safety MOV-RH8701, 8702 1 Verification MOV-RH8700A, B 2,3 Containment Sump 5 RHR Pumps A, 8 6 71710 ESF System Walkdown MOV-RH8701, 8702 1 MOV-RHS700A, 8 3 Containment Sump 5 RHR Pumps A, 8 6

a. See Table 1A for failure identification.

TASLE IC. ISIDIFIED RESIDUAL HEAT REMOVAL SYSTEM um rnnha Component Required Actual Number Noun Name Location Position Position Electrical .

Pump 1A NIR Pump 1A Bus 149 G34 Racked In -

RHR Pump 1A DC and Ckt 8kr On Spring Charging Motor Switch Pump 18 _

RHR Pump 18 Bus 148 G33 Racked In ,

RHR Pump 18 DC and Ckt Skr On .

Spring Charging Motor Switch IMOV-RH8701 ACS to RHR Pumps Isolation MCC1391-83 On -

IMOV-RH8702 RCS to RHR Pumps Isolation MCC1381-86 On IMOV-RH8700A RHR Pump 1A Suction Isolation MCC1393C-T5 On -

IMOV-RH87008 RHR Pump 18 Suction Isolation MCC1383A-A4 On .

Valve Lineup IMOV-RH8700A RHR Pump 1A Suction Isolation 542' L22 Open IMOV-RH87008 RHR Pune 18 Suction Isolation 542* M22 Open 1MnU-RHR701 909 tn RHR Puens Isolation MR' 73G Open

TABLE 5. INDIAN POINT 2 MOST IMPORTANT SEQUENCES et Initiator Faulted Svsta== Involved Public Health.

Events insertance V A RNR .701 ETI H EP .120 ET2 H EP .120 ETilB41 F EP .003 ET3 A R2 RHR .003 ET2 E EP .002

)

ETI A R1 EP, RHR .002 ET2 A R1 EP, RHR .002 ETI E EP .002

. ET4 A OP41 SL2 PZR, PCS .001 ETilA H L1 EP, AFW .001 ET12A H L1 EP, AFW .001 ET7 H L1 EP. AFW .001 ETI A LPI ACC, RHR .001 FIRE A FZIA FP .001 FIRE A FZ14 FP .001 FIRE A FZ32A FP .001 ET4 A SLI PCS .001 Notes:

Initiator Descriotion V Interfacing System Loss of Coolant Accident ETI Large loss of Coolant Accident ET2 Medium Loss of Coolant Accident ET3 Small Loss of Coolant Accident ET4 Steam Generator Tube Rupture ET7 Soss of Main Feedwater -

l ETilA Turbine Trip ET11841 Loss of Offsite Power Eil2A Spurious Safety injection l

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e PRA Applications to inspection Carrent PftA F2, IP3, Westingineesse Generic Plam Appticaties Plaa .

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Seabreelt

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non-PRA Plant AppNesties Plom

) AppNcation Trial Plaat U

\ Generic Plea

/ Evatesation Trial Plant cvs sees

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IWORTANT SYSTDtS FOR EACH PUWT IP2 IP3 SEASA00K ZION M SW M ~

RHR SW EP SSPS EP - -

CCW CCW RWST CS CF RPS PCCW AFW EP HPI ESFAS RCS G E W US SI RECIRC EFW SI ACC RCS EP CCW FR AFW ~

pts .

AFW LPI SAS ACC CS CF -

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GENERIC METHODOLOGY l

o SYSTEM IDENTIFICATION FROM GENERIC

! DATA AND PLANT FUNCTIONAL REQUIREMENTS o IMPORTANT COMPONENTS GENERIC DATA GENERIC FAULT TREE RESULTS (ASEP)

SYSTEM FAULT TREE ANALYSIS l

ENGINEERING JUDGEMENT

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X Rancho Seco Powerplant inspection Guidance Derived from PRAs for Surrogate Plants B.F. Gore J.C. Huenefeld sanene Pacific Northwest Laboratory

_ .-.-, - -. -.. . ,, __...___._-n. _ , _,_ m

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. -x Rancho Seco Applications .

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Objective:

• Develop information for inspection planning

• Develop insight from PRA for other B&W plants *

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- ANO-1 -

Oconee

• Duke /EPRI Study

• RSS Map Study .

• No PRA information exists for Rancho Seco g a

T Rationale

• B&W Plants share important similarities

• RCS arrangement

• OTSG design; MFW & EFW inlets

• ECCS systems

- Separate / redundant trains O

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Develop Rancho Seco inspection Information 8 For each failure mode identify RS component ID numbers Operating mode (manual vs MOV)

Normal position (LO, LC)

Power supplies

• Motor power j

• Cohtrol power

• System operating procedures '

Valve line-ups

• Additional information obtained from Operator training materials System diagrams Electrical prints Operating procedures t

x Study PRAs far Surrogoto PI:nto .

• Identify dominant accident sequences

- Study PRA cut sets

- Focus on core melt probability

- >85% of total e Develop failure mode categories

- Assign cut set elements to systems

- Define categories of failures for each system i - Calculate Fussel-Vesely importances

e Relate failure mode categories to system designs I

- Study systems for ANO-1 and Oconee

- Determine similarities / differences

- Correlate failure mode importance and system design -

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- - . , , - , _ _ -__3.-,. - . .---.e- , . --

.-w-- - --- +- , - - .m -- - - e- --.- w -. y------ ----.---- -i - -- -

Conclusions

• Surrogate plant PRAs can provide many insights useful in inspection planning

• Compared with plant-specific PRA, this method is very efficient .,

e Strongest point: inference of important system failure modes from PRA results and system similarities / differences e Weakest point: quantification of results is extremely uncertain

' e The use of 2 or 3 PRAs helps to highlight how plant i

differences relate to PRA outcomes ,

• Documentation ,0f plant specific information for inspector reference can b'e efficiently combined with studies of systems to evaluate failure mode plausibility

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PRA-Based Inspection Guidande  :

i for'OCONEE Unit 3  : J<.

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Pacific Northwest Laboratory ,y

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Objectives

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1 l e identify important accident initiators and sequences l e Identify and prioritize systems associated with 98%

of the inspectable plant risk l , e Identify failure modes for important components in "

identified systems l ,

e Develop checkoff lists for identified systems l

specifying proper configuration of important _

components during normal operation o Correlate identified components with I&E inspection procedures most related to component reliability L - -- _ - _ _ - - - _ _

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Analysis of the OCONEE PRA ,

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l e Calculate risk-based Fussel-Vesely importance of systems l '

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e Prioritize and list systems associated with 98% of l ~

l inspectable risk l

e Identify components contributing to 95% of each system failure probability

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Project Status -

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j e System importances calculated -

e Ready to start component identification i

i - Seeking most efficient method .

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! Component Identification Options l

l For each system  :

e Acquire lists of importance-ranked components f e Acquire system fault tree cut sets '

! - Analyzh for component importance i e i

i e Acquire system fault tree analysis tapes

- Rerun necessary analyses e input fault trees to IRRAS code

- Run on IBM-PC .

e Generate needed fault trees

- Input to IRRAS and analyze i ,i j -

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i l Example Component importances .

Component Failure Rate F-V importance B-7 1.00E-001 5.17E-001 B-4 1.00E-001 4.94E-001 B-2 2.40E-002 4.87E-001  :

B-5 1.00E-001 3.91 E-001 i

B-6 1.00E-001 2.74E-001 '

l 2.73E-001 j B-9 1.00E-001 I B 'i O 2.51 E-001 1.92E-001 B-8 1.OOE-001 1.80E-001 i B-11 1.30E-003 1.58E-003 B-1 2.95E-006 8.09E-005 .

B-3 . . . . , . .

2.40E-006 1.19E-005 i

B-12 1.34E-005 7.22E-010

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1 Example System Level Minimum Cut Sets .

! Cut Set Percent j Elements Unavailability Contribution  ;

i-26.95%

B-2 B-4 2.40E-003 B-4 B-5 B-7 1.00E-003 11.23%

6.02E-004 6.76%

B-10 B-2 B-6 B-10 B-6 B-7 B-8 2.51 E-004 2.82%

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B-2 B-5 B-6 2.40E-004 2.69% '

1 B-6 B-7 B-8 B-9 1.00E-004 1.12%

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B-10 B-11 B-2 7.82E-006 0.09%

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Residual Heat Removal System Failure ,.

,! Mode Identification - Eum,le .

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1. RCS to RHR Pumps isolation Valves IMOV-RH8701 and IMOV-RH8702 Fail Open j- .

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2. Operator Fails to initiate Switchover from injection to ,,

Recirculation Mode or Fails to Stop Pump at BWST' Low Level

3. RHR Pump Suction isolation Valves IMOV-RH8700A, B Fail to Close i

4. RHR Pumps 1 A and 1B Unavailable Due to Maintenance

5. Blockage of Containment Sump

6. RHR Pumps 1 A and 1B Failure

lE Modules for Residual Heat Removal System inspection fuewy le Failurea Module Title Components Mode 61701 Surveillance (Complex) RHR Pumps A, B 6 61726 Monthly Surveillance MOV-RH8701, 8702 . 1 Observation MOV-RH8700A, B 3 RHR Pumps A, B 6 62700 . Maintenance RHR Pumps A, B 4 62703 Monthly Maintenance RHR Pumps A, B 4 Observation

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71707 Operational Safety MOV-RH8701, 8702 1 Verification MOV-RH8700A, B 2,3 .

Containment Sump 5 RHR Pumps A, B 6 71710 ESF System Walkdown MOV-RH8701, 8702 1 MOV-RH8700A, B 3 Containment Sump 5 RHR Pumps A, B 6

Modified Residual Heat Removal System Walkdown - gw,9/ e Component Required Actual Number Noun Name location Position Position- -

Electrical Pump 1A RHR Pump 1A Busi49 G34 Racked In

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RHR Pump 1A OC and Ckt Skr On Spring Charging Motor Switch Pump 1B RHR Pump 1B Bus 148 G33 Racked In RHR Pump 1B DC and Ckt Skr On Spring Charging Motor Switch IMOV-RH8701 RCS to RHR Pumps Isolation MCC1391-83 On IMOV-RH8702 RCS to RHR Pumps Isolation MCC1381-86 On IMOV-RH8700A RHR Pump 1A Suction Isolation MCC1393C-T5 On IMOV-RH87008 RHR Pump 18 Suction Isolation MCC1383A-A4 On Valve Lineup 1MOV-RH8700A RHR Pump 1A Suction Isolation 542' L22 Open IMOV-RH87008 RHR Pump 1B Suction Isolation 542' M22 Open RCS to RHR Pumps Isolation 568' Z30 Open IMOV-RH8701 1MOV-RH8702 RCS to RHR Pumps Isolation 568' Z27 Open

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Calculation of F-V importance of System .

1 I' e include only failures which system inspection can affect: l

- Internal events >

- Turbine building flood ,

e Exclude:

- External flood ,

- Earthquake i - Tornado ,

s l - Fire -

i l, - Mitigating systems are failed by these events.  ;

l importance calculations for those systems would be biased by including the non-inspectable failures

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! Calculation of System importance .

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e Calculate risk for each cut set i

- Associate cut set with one or more plant damage bin

- Associate each bin with appropriate release categories using containment event tree transfer functions ,

include containment bypass events as Category 2 releases

- Risk is sum of:

Release category frequency X' man rem / category Summed over associated bins and release ,

categories e . Associate cut set elements with systems .

- Exceptions: human error, recovery factors l.

e System F-V importance is sum of risk from cut sets j with system elements divided by total risk 1

f * .

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i Risk Transfer Function (RTF?

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(

4 e Facilitates calculations of system importance l

i e Single multiplication converts plant damage bin frequency to plant risk t

) '

i RTF X PDB frequency = man rem per year .

e Combines the CET transfer function with the risk consequences (man rem) for each release category RTF = (CET 1 A)X(3.0+8)+(CET 1 B)X(4.0+7)

+ (CET 2) X (1.0+8) + (CET 3) X (0.0)

+(CET 4)X(1.0+6) +(CET 5) X (0.0)

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Risk Based System importances - Preliminary Internal Event Sequences Only

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System or Event {V importance (%)

Operator Error 76 RBS 74 LPI 6C Recovery Factors 53 SRV 49 RBC 6 4 kV Vital AC 6 EFW 4 LPSW 3 LOP O.7 HPSW O.5

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RPS 0.4 IA O.3 HPI O.2 Total Internal Event Risk 81 man rem /yr -

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l Risk Based System importances - Preliminary l Turbine Building Flood Only System or Event F-V importance (%)

ccW 100 1

RBC 99

! RBS 91 i j Recovery Factors 61 l l SRV 37 -

SSF 9 '

(HPI 7)

(ASW 1)

LPSW 3

{

Operator Error 1 I

Total Turbine Building Flood Risk 106 man rem /yr o

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i Risk Based System importances - Preliminary i

I internal Events and Turbine Building Flood System or Event F-V Importance (%)

RBS 83 RBC 59 Recovery Factors 58 CCW 57 i

S RV's 42 l 34 l

Operator Error LPI 28 '

l I SSF HPl 4 LPSW 3  ;

! 4 kV AC 2 Emergency FW 2 l O.7 i SSF ASW LOP O.3 HPSW O.2 ~

i RPS 0.2 HPI O.1  !

IA O.1 l

'l Total Risk - Internal Events and TB Flood i~ 187 man rem /yr ,

FUTURE INSPECTION PLANS

- Announced Inspection

- Team Of Four or Five Inspectors, Plus Team Leader

- Inspection Requires About Two Weeks On-Site

- Requires Cooperation From Utility:

Operators Supervisors Craftsmen

.