ML20129G554

Forwards Presentation Matl from B&W Owners Group Steering Committee 961017 Meeting

ML20129G554
Person / Time
Issue date:	10/28/1996
From:	Rossi C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	Jordan E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
References
NUDOCS 9610300137
Download: ML20129G554 (31)

Text

October 28, 1996 MEMORANDUM TO:

Edward L. Jordan, Director Office for Analysis and Evaluation of Operational Data FROM:

Charles E. Rossi, Director Safety Programs Division Office for Analysis and Evaluation of Operational Data

SUBJECT:

PRESENTATION TO B&W OWNERS GROUP STEERING COMMITTEE The purpose of this memorandum is to summarize my participation in the B&W Owners Group Steering Committee meeting on October 17, 1996.

I attended this meeting to discuss recently completed and currently ongoing studies in the Safety Programs Division.

I have attached a copy of the handout used as the basis for my discussion which lasted about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Attachment:

As stated Distribution:

PDR/ Central File SPD R/F l

D. Ross J. Rosenthal P. Baranowsky D. Hickman QfD?3I DOCUMENT NAME:P: Westing.cer To receive a sepy of this deswnent, indioste in the box *C" copy wie attachtenol "E" sepy wIntteshlenol "N" ne copy 0FFICE SPD G.

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OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA l'

CHARLES E. (ERNIE) ROSSI 1

i SAFETY PROGRAMS DIVISION AEOD i

WASHINGTON, D.C. 20555 (301) 415-7499 i

l 2

1 t

INDEPENDENT REVIEW e

ALL LICENSEE EVENT REPORTS ARE REVIEWED IN A l

SYSTEMATIZED WAY BY AEOD REVIEWERS TO CLASSIFY FOR

[

SIGNIFICANCE i

i e

THE DATA IS ENCODED IN THE SEQUENCE CODING AND SEARCH SYSTEM TO ASSIST IN IDENTIFYING RELATED EVENTS CAUSAL AND TIME ASPECTS OF OCCURRENCES WITHIN THE EVENT SEQUENCE i

e AEOD ALSO REVIEWS FOREIGN EVENT DATA, IRS AND BILATERAL, TO ADD TO THE DOMESTIC EXPERIENCE i

[

-. [

INDEPENDENT LONG TERM STUDIES AND ANALYSES i

L e

OPERATIONAL DATA SYSTEMS AND RELIABILITY DATA COLLECTION l

t I

e BROAD SCOPE SYSTEMATIC REVIEW OF OPERATING EXPERIENCE e

IDENTIFICATION OF SAFETY ISSUES i

e INDEPTH COMPONENT, SYSTEMS, HUMAN PERFORMANCE STUDIES PERFORMANCE INDICATOR PROGRAM l

e ACCIDENT SEQUENCE PRECURSOR PROGRAM e

SYSTEM RELIABILITY STUDIES 4

l

RAB REPORTS ISSUED IN 1994 SPECIAL STUDY REPORTS DATE TITLE NO.

AUTIIOR i

02/94 OPERATING EXPERIENCE FEEDBACK REPORT-S94-01 J. BOARDMAN i

RELIABILITY OF SAFETY-RELATED STEAM TURBINE-DRIVEN STANDBY PUMPS i

09/94 TURBINE-GENERATOR OVERSPEED PROTECTION S94-02 II. ORNSTEIN SYSTEMS AT U.S. LIGIIT-WATER REACTORS l

TECIINICAL REVIEW REPORTS DATE TITLE NO.

AUTIIOR 03/94 TIIE ELECTRICAL TRANSIENT WHICH FOLLOWED TIIE T94-01 M.WEGNER LOS ANGELES EARTIIQUAKE-JANUARY 17, 1994 05/94 REVIEW OF MISPOSITIONED EQUIPMENT EVENTS T94-02 S. ISRAEL i

07/94 COMPUTER-BASED DIGITAL SYSTEM FAILURES T94-03 E. LEE l

12/94 POTENTIAL FOR BOILING WATER REACTOR EMERGENCY T94-04 J. BOARDMAN CORE COOLING SYSTEM STRAINER BLOCKAGE DUE TO l

LOSS-OF-COOLANT ACCIDENT-GENERATED DEBRIS

.5

RAB REFORTS ISSUED IN 1995 SPECIAL STUDY REPORTS DATE TITLE NO.

AUTHOR 03/95 REACTOR COOLANT SYSTEM BLOWDOWN AT S95-01 J. KAUFFMAN WOLF CREEK ON SEPTEMBER 17,1994 S. ISRAEL ENGINEERING EVALUATIONS DATE TITLE NO.

AUTHOR 07/95 OPERATING EVENTS WITH INAPPROPRIATE E95-01 J. KAUFFMAN BYPASS OR DEFEAT OF ENGINEERED SAFETY FEATURES l

TECHNICAL REVIEW REPORTS DATE TITLE NO.

AUTIIOR 03/95 MAJOR DISTURBANCES ON THE WESTERN GRID AND T95-01 M.WEGNER RELATED EVENTS 10/95 POTENTIAL DAMAGE TO LOW-PRESSURE INJECTION T95-02 E. BROWN VALVES DURING SURVEILLANCE TESTING 6

RAB REPORTS ISSUED IN 1996 3

SPECIAL STUDY REPORTS i

DATE TITLE NO.

AUTHOR 04/96 STEAM GENERATOR TUBE FAILURES NUREG/CR-6365 INEL 06/96 OCONEE ELECTRICAL SYSTEM DESIGN AND OPERATION G. LANIK 09/96 ASSESSMENT OF SPENT FUEL COOLING J. IBARRA i

ENGINEERING EVALUATIONS DATE TITLE NO.

AUTHOR 03/96 MOTOR-OPERATED VALVE KEY FAILURES E96-01 C.HSU 04/%

ANALYSIS OF ALLEGATION DATA E96-02 S. ISRAEL 09/96 INSIGHTS FROM UNDETECTED FAILURES OF E96-XX S. PULLANI SAFETY SYSTEMS I

TECHNICAL REVIEW REPORTS DATE TITLE NO.

AUTHOR 03/96 AEOD TECHNICAL REPORTS BY CATEGORY REVISION 1 T%-01 S. ISRAEL 04/96 TARGET ROCK TWO-STAGE SRV PERFORMANCE UPDATE T96-02 M.WEGNER i

09/96 RESPONSE OF B&W PLANTS ON LOSS OF T96-03 W.RAUGHLEY NONEMERGENCY AC POWER j

7

[

i STEAM GENERATOR TUBE FAILURE STUDY NUREG/CR-6365 e

PEER REVIEWED BY NRR e

INCLUDES BOTH A HISTORICAL REVIEW AND AN OVERVIEW OF EVENTS AND ACTIONS WORLDWIDE RELATED TO STEAM GENERATOR TUBE FAILURES, INCLUDING NRC GENERIC LETTER 95-05 e

COVERS THE FOLLOWING:

STEAM GENERATOR DESIGN STEAM GENERATOR DEGRADATION MECHANISMS, SITES, AND FAILURE MODES STEAM GENERATOR TUBE RUPTURES j

THERMAL-HYDRAULIC RESPONSE OF A TYPICAL PWR PLANT WITH A DEFECTIVE STEAM GENERATOR I

B 8

I STEAM GENERATOR TUBE FAILURE STUDY NUREG/CR-6365 (CONT.)

e COVERS THE FOLLOWING (CONT.):

TRANSIENT STEAM GENERATOR ACCIDENT RESPONSE RISK SIGNIFICANCE OF STEAM GENERATOR TUBE RUPTURE ACCIDENTS WORLDWIDE REGULATORY PRACTICES AND FITNESS-FOR-SERVICE GUIDELINES STEAM GENERATOR TUBE DEFECT DETECTION AND SIZING i

L i

9 i

MOTOR-OPERATED. VALVE KEY FAILURES i

e STUDY INITIATED BY EVENT AT PALO VERDE ON APRIL 18, 1995 i

e A TOTAL OF 73 REPORTS WERE IDENTIFIED FROM 46 OPERATING f

PLANTS IN THE PERIOD FROM 1990 TO SEPTEMBER 1995 j

ANTI-ROTATION KEY FAILURES 11 REPORTS VALVE OPERATOR-TO-VALVE STEM KEY FAILURES 22 i

REPORTS MOTOR PINION KEY FAILURES 40 REPORTS t

i e

SAFETY SIGNIFICANCE j

MANY KEY FAILURES WERE NOT DETECTED DURING SURVEILLANCE AND HAD EXISTED FOR SOME TIME BEFORE DISCOVERY POTENTIAL COMMON-MODE FAILURE COULD RENDER CERTAIN SAFETY-RELATED SYSTEMS INOPERABLE l

l 10

1 i

MOTOR-OPERATED VALVE KEY FAILURES (CONT.)

{

i e

ANTI-ROTATION KEY FAILURE 11 EVENTS FROM 11 PLANTS TWO CAUSES: INSTALLATION AND DESIGN DEFICIENCIES I

e VALVE OPERATOR-TO-VALVE STEM KEYS 22 EVENTS FROM 19 PLANTS. AMONG THESE 16 KEY FAILURES ON DEMAND OR DETECTED DURING SYSTEM l

OPERATIONS, THE OTHERS FAILED DURING TESTING l

DOMINANT CAUSE: LOOSE OR SLIPPING OUT (ASSOCIATED WITH SETSCREW LOOSENING OR IMPROPER STAKING) e LIMITORQUE MOTOR PINION GEAR KEYS 40 EVENTS FROM 27 PLANTS

(

THREE MAIN CAUSES: HIGH IMPACT LOAD (15 EVENTS),

IMPROPER MATERIAL (8 EVENTS), AND IMPROPER INSTALLATION SLIPPING OUT OR MISSING KEYS j

(13 EVENTS) l 11

[

CATEGORIZATION OF AEOD TECHNICAL REPORTS OVER 500 REPORTS ISSUED SINCE 1980 i

CROSS-REFERENCED IN FOUR TABLES t

PWR SYSTEMS BWR SYSTEMS ACTIVITY / DEFICIENCY TOPIC COMPILED IN AEOD/T96-01 RESOURCE FOR CHECKING OPERATING EXPERIENCE RELATED I

TO EMERGING TECHNICAL ISSUES f

CATEGORIZATION TABLES AND REPORT SUMMARIES WILL BE PUT ON INTERNET 12

i i

HUMAN PERFORMANCE HUMAN PERFORMANCE EVENT DATABASE (HPED)

A SYSTEM FOR STORING, MANIPULATING, AND RETRIEVING DATA ON HUMAN e

PERFORMANCE DURING OPERATING EVENTS e

CONTAINS RECORDS FROM AITs (40) AND HUMAN PERFORMANCE STUDIES (20) e RECORDS WILL BE ADDED FOR OTHER AITs, EVENT-DRIVEN SPECIAL l'

INSPECTIONS, AND SELECTED LICENSEE EVENT REPORTS I

e WILL PROVIDE DATA FOR STUDIES OF IIUMAN PERFORMANCE AND RELIABILITY AND WILL BE USED TO EVALUATE REQUIREMENTS FOR HUMAN l

PERFORMANCE DATA HUMAN PERFORMANCE PROGRAM PLAN (HPPP) e A SYSTEM FOR DESCRIBING AND MONITORING THE STATUS OF ALL AGENCY PROGRAMS RELATED TO HUMAN PERFORMANCE e

INTENDED TO PROMOTE INTEROFFICE COOPERATION AND COORDINATION AND AVOID DUPLICATION e

ESTABLISHED AND MAINTAINED BY THE HUMAN FACTORS COORDINATION i

COMMITTEE (WITH REPRESENTATIVES FROM AEOD, NRR, RES, AND NMSS) t o

TO BE REVISED AND PUBLISHED SEMIANNUALLY l

l 13 I

i REVIEW OF OCONEE EMERGENCY ELECTRICAL SYSTEM e

AEOD TASK PROVIDE AN INDEPENDENT EVALUATION OF SYSTEM DESIGN AND OPERATION i

REVIEW SAFETY CONCERNS AND POTENTIAL RISKS I

e METHODOLOGY REVIEW OF THE FOLLOWING:

OPERATIONAL EVENTS VULNERABILITIES l

TESTING l

HUMAN PERFORMANCE MANAGEMENT ISSUES i

KEOWEE RELIABILITY ASSESSMENT OTHER AVAILABLE PROBABILISTIC RISK ASSESSMENTS SITE VISITS 14 i

REVIEW OF OCONEE EMERGENCY ELECTRICAL SYSTEM l

(CONTINUED) 1 CONCLUSION: A LEVEL OF SAFETY COMPARABLE TO THAT OF A PLANT WITH DIESEL GENERATORS MAY BE ACHIEVED ASSUMING INSTALLATION AND TESTING OF DESIGN CHANGES PROPOSED BY OCONEE OR IDENTIFIED BY NRC - EXAMPLES ARE CHANGES TO l

ENSURE ADEQUATE VOLTAGE AND FREQUENCY DURING LOADING OF EMERGENCY POWER SOURCES TESTING TO DEMONSTRATE CAPABILITY OF SYSTEMS TO PROGRESS THROUGH A START AND LOAD CYCLE OF EMERGENCY EQUIPMENT PERIODIC TESTING OF EMERGENC" ELECTRICAL SYSTEM UPGRADE AND TEST OPERATOR PROCEDURES AND TRAINING FOR EMERGENCY POWER SYSTEM l

TESTING OF THE SAFE SHUTDOWN FACILITY TO VERIFY DESIGN AND PERIODIC TESTING TO MAINTAIN SYSTEM RELIABILITY 15

TECHNICAL REVIEW REPORT, " RESPONSE OF B&W PLANTS l

FOLLOWING A LOSS OF NONEMERGENCY AC POWER" t

i FOLLOWUP TO OCONEE EMERGENCY POWER STUDY j

CONCERN COMPLETE LOSS OF AC AT OCONEE COULD LEAD TO:

TRIPS OF ALL 3 REACTORS POTENTIAL OVERCOOLING TRANSIENTS MULTIPLE UNIT EMERGENCY CORE COOLING SYSTEM ACTUATIONS EMERGENCY AC SYSTEM LOADS BEYOND THOSE CONSIDERED i

VOLTAGE AND FREQUENCY REDUCTIONS BEYOND THOSE CONSIDERED REVIEWED:

B&W PLANT OPERATING EXPERIENCE (1987 TO 1996) 18 COMPLETE OR PARTIAL LOSSES OF NONEMERGENCY AC POWER EVENTS f

13 WHILE PLANTS SHUTDOWN 5 DURING POWER OPERATION l

FINDINGS:

NO OVERCOOLING TRANSIENTS OCCURRED NO FULL ECCS ACTUATIONS OCCURRED 4 OF THE 5 EVENTS AT POWER HAD AUTO INITIATION OF MOTOR DRIVEN EMERGENCY FEEDWATER PUMPS FROM EMERGENCY POWER 16 i

i I

l i

i INDEPENDENT SPENT FUEL POOL STUDY i

e STUDY REQUESTED BY EXECUTIVE DIRECTOR FOR OPERATIONS e

DEVELOPED GENERIC CONFIGURATIONS TO ASSESS LOSS OF SPENT FUEL POOL COOLING AND INVENTORY e

ASSESSED 12 YEARS OF OPERATIONAL EXPERIENCE e

PERFORMED SITE VISITS TO GATHER INFORMATION ON PHYSICAL CONFIGURATION, PRACTICES, AND PROCEDURES

[

e REVIEWED REGULATIONS, STANDARD REVIEW PLAN AND f

REGULATORY GUIDES l

e PERFORMED ASSESSMENTS OF ELECTRICAL SYSTEMS, INSTRUMENTATION, HEAT LOADS, AND RADIATION e

CONTRACTOR EVALUATED RISK OF LOSING SPENT FUEL COOLING 17 f

SFP STUDY FINDINGS AND CONCLUSIONS j

l LIKELIHOOD AND CONSEQUENCES l

o CONSEQUENCES OF ACTUAL EVENTS HAVE NOT BEEN SEVERE e

PRIMARY CAUSE OF EVENTS HAS BEEN HUMAN ERROR e

RELATIVE RISK IS LOW COMPARED WITH OTHER REACTOR RISKS e

HIGHLY DEPENDENT ON HUMAN PERFORMANCE AND PLANT DESIGN t

PREVENTION e

CONFIGURATION CONTROL IMPROVEMENTS CAN PREVENT AND/OR MITIGATE SFP EVENTS l

e EVALUATIONS MAY BE NEEDED AT SOME MULTI-UNIT SITES FOR

[

POTENTIAL SFP BOILING EFFECTS ON SAFE SHUTDOWN

RESPONSE

e IMPROVED PROCEDURES & TRAINING MAY BE NEEDED l

IMPROVEMENTS TO INSTRUMENTATION AND POWER SUPPLIES MAY BE e

NEEDED 18

LOSS OF INVENTORY DU RATION NUMBER OF OCCURRENCES 3

7 i

7 6

5 4

3 3

2 2

2 i

1 1@ 72 HRS 1

1@ 24 HRS 0

<1 1TO4 4TO8 8 TO 24

> 24 i

DURATION (HRS?

19 r

LEVEL DECREASE DUE TO LOSS OF INVENTORY NUMBER OF OCCURRENCES 10 8

8 6

i 4

f 2

2 2

1.. @ 6 51 N 1@

60-12OIN O

<3 3 TO 12 12 TO 60'

> 60 LEVEL DECREASE (INCHES) 20

t DURATIOl\\ OF LOSS OF COOLING EVENTS 1

NUMBER OF OCCURRENCES 25 22 20 15 1 @ 32 HRS 10 1 @ 30 HRS i

10

~

~~

'i~@ 24 HRS ~

5 5

5 3

v 0

<1 1 -- 4 4 -- 8 8--24

> 24 i

DURATION (HRS) i

TEMPERATURE INCREASE DUE TO LOSS OF COOLING EVENTS NUMBER OF OCCURRENCES 20 16 i

15 j

10 5

a 1

I"

"""I O

O O TO 20 20 TO 40 40 TO 60 TEMPERATURE INCREASE (DEG F) n

REDUCED TIME TO BOIL AT NINE MILE Ul\\IT 2 HOURS TO BOIL DAYS TO OFFLOAD 120 0

108 40 80 30 0 "

""S' B o "-

60 23 U D AYS TO OFl~LO AD 39 20 40 13 29 7

20 5

8 0

-0 1

2 3

4 REFUELING OUTAGE 23

INSIGHT FROM UNDETECTED FAILURES OF SAFETY SYSTEMS f

BACKGROUND e

33 EVENTS HAVING CONDITIONAL CORE DAMAGE PROBABILITY ABOVE 10 4

E OCCURRED DURING 1991-93 l

e THE EVENTS IDENTIFIED FROM ACCIDENT SEQUENCE PRECURSOR DATA BASE I

e THE STUDY IS AN EVALUATION OF US EVENTS, THE INPUT ORIGINALLY GIVEN TO NUCLEAR ENERGY AGENCY'S WORLD-WIDE STUDY l

l OBJECTIVE TO GAIN INSIGHTS USEFUL IN PREVENTION OR REDUCTION OF LIKELIHOOD OF SUCH FAILURES l

METHODOLOGY l

THE FAILURES WERE ANALYZED WITH RESPECT TO:

DISCOVERY METHOD TIME TO DISCOVER j

FAILURE CAUSES l

CORRECTIVE / PREVENTIVE ACTIONS BY LICENSEES t

REGULATORY ACTIONS TAKEN

[

24

i I

INSIGHT FROM UNDETECTED FAILURES OF SAFETY SYSTEMS (CONTINUED)

FINDINGS AND CONCLUSIONS Y

e FAILURES REMAINED UNDISCOVERED FOR A LONG TIME: 4 (1-10 YRS),

4 (10-18 YRS),4 POSSIBLY SINCE PLANT STARTUP e

11 EVENTS /YR OR.1 PER REACTOR YEAR e

29 AT PWRs AND 4 AT BWRs (NO RATIONALE FOR TIIE SKEW) e

> 75% DISCOVERED BY TESTING OR ANALYSIS OF OPERATIONAL PROBLEMS l

~ 70% CAUSED BY COMPONENT FAILURE, DESIGN DEFICIENCY, OR INADEQUATE e

TESTING / MAINTENANCE PROCEDURE i

l e

MOST FREQUENT LICENSEE ACTIONS: PLANT MODIFICATION, ADDITIONAL

[

TRAINING, MODIFIED OPERATING / MAINTENANCE PROCEDURE I

e 7 EVENTS WERE THE SUBJECT OF NRC GENERIC COMMUNICATIONS 4 OF THE ABOVE 7 WERE THE SUBJECT OF AUGMENTED INSPECTION TEAMS (AIT) e i

e OUT OF 12 RELATIVELY MORE SIGNIFICANT EVENTS, I WAS THE SUBJECT OF NRC

[

GENERIC COMMUNICATION AND NONE WAS TIIE SUBJECT OF AN AIT l

[

25 f

NRC ALLEGATION DATA i

e COMMISSION DIRECTED TREND ANALYSIS TO ALERT AGENCY TO CHANGES e

ALLEGATION DATA INCLUDES:

TECHNICAL CONCERNS HARASSMENT AND INTIMIDATION (H & I) l SOME WITH BOTH e

GROUPINGS:

MULTIPLE-UNIT SITES SINGLE-UNIT SITES VENDORS

[

t 5

26 m..

m

NRC ALLEGATION DATA (CONTINUED) e INEL PERFORMED ANALYSIS TO IDENTIFY OUTLIERS 1994 COUNT CHANGE FROM 1993 TO 1994 TREND OF COUNT (SLOPE OVER 5 YEARS) e RESULTS OBTAINED FOR EACH MEASURE FOR BOTH TOTAL ALLEGATIONS AND H & I ALLEGATIONS e

WATTS BAR, MILLSTONE, BURNS SECURITY HAVE A DISPROPORTIONATE NUMBER OF ALLEGATIONS 27

PERFORMANCE INDICATORS l

l e

AUTOMATIC SCRAMS WHILE CRITICAL e

SAFETY SYSTEM ACTUATIONS e

SIGNIFICANT EVENTS e

SAFETY SYSTEM FAILURES j

e FORCED OUTAGE RATE l

i i

e EQUIPMENT FORCED OUTAGES PER 1000 CRITICAL HOURS i

e COLLECTIVE RADIATION EXPOSURE i

e CAUSE CODES l

I 28

CAUSE CODES ADMINISTRATIVE CONTROL PROBLEMS LICENSED OPERATOR ERRORS t

OTHER PERSONNEL ERRORS i

i e

MAINTENANCE PROBLEMS e

DESIGN, CONSTRUCTION, INSTALLATION OR i

FABRICATION PROBLEMS i

e MISCELLANEOUS FAILURES l

1 29 I

PERFORMANCE INDICA ORS ANNUAL INDUSTRY AVERAGES

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30

t RISK BASED ANALYSIS OF l

i REACTOR OPERATING EXPERIENCE I

USE REACTOR OPERATING EXPERIENCE TO:

e ASSESS AND TREND RISK INDICATORS l

e COMPARE WITH PROBABILISTIC RISK ASSESSMENTS (PRAs) AND INDIVIDUAL PLANT EXAMINATIONS (IPEs) e IDENTIFY TECHNICAL INSIGHTS RELATING TO RISK CONTRIBUTORS i

e PROVIDE INSIGHTS TO INDUSTRY AND REGULATORY ACTIVITIES RELATED TO RISK t

I 31

i PROGRAM ELEMENTS

?

i e

ACCIDENT SEQUENCE PRECURSOR (ASP) ANNUAL REPORT l

ASP METHODS ASP DATABASE l

i f

o INITIATING EVENTS PERIODIC REPORT LOSS OF OFFSITE POWER (LOOP) DATABASE SPECIAL INITIATORS (e.g., POWER-OPERATED RELIEF l

VALVES, STEAM GENERATOR TUBE RUPTURES, HUMAN I

ERRORS) i e

SYSTEM RELIABILITY STUDIES RELIABILITY INDICATORS i

COMPONENT ANALYSES l

l i

t i

32

I L

PROGRAM ELEMENTS (CONTINUED) e COMMON-CAUSE FAILURES (CCFs)

DATABASE AND ANALYSIS SOFTWARE PERIODIC ANALYSIS i

INTERNATIONAL COMMON-CAUSE DATA l

EXCHANGE EFFORT t

I e

PERFORMANCE INDICATORS (RISK / RELIABILITY) i e

DATA SYSTEMS SEQUENCE CODING AND SEARCH SYSTEM (SCSS)

AND NUCLEAR PLANT RELIABILITY DATA SYSTEM i

(NPRDS) i RELIABILITY DATA i

s

=

l 33

ACCIDENT SEQUENCE PRECURSOR PROGRAM l

DETERMINE CONDITIONAL PROBABILITY OF SUBSEQUENT e

SEVERE CORE DAMAGE CONDITIONAL CORE DAMAGE PROBABILITY (CCDP)

GIVEN THE FAILURES DURING AN

' OPERATIONAL EVENT i

i i

l 34

OBJECTIVES OF THE ASP PROGRAM I

e IDENTIFY AND RANK RISK SIGNIFICANCE OF OPERATIONAL EVENTS i

e DETERMINE GENERIC IMPLICATIONS OF AN OPERATIONAL EVENT / CHARACTERIZE RISK INSIGHTS e

PROVIDE SUPPLEMENTAL INFORMATION ON PLANT-f SPECIFIC PERFORMANCE

[

e PROVIDE A CHECK WITH PRAS e

PROVIDE AN EMPIRICAL INDICATION OF INDUSTRY RISK AND ASSOCIATED TRENDS 35 1

DEFINITIONS e

ACCIDENT SEQUENCE PRECURSORS - EVENTS OR CONDITIONS f

I THAT ARE IMPORTANT ELEMENTS IN THE SEVERE CORE DAMAGE SEQUENCES (e.g., AN UNUSUAL INITIATING EVENT OR FAILURES OF MULTIPLE COMPONENTS THAT, WHEN COUPLED WITH ONE OR MORE POSTULATED EVENTS, COULD RESULT IN A PLANT CONDITION LEADING TO SEVERE CORE j

DAMAGE) i e

CONTAINMENT-RELATED EVENTS - FAILURES THAT COULD RESULT IN REDUCED CONTAINMENT PERFORMANCE (e.g.,

UNAVAILABILITY OF A CONTAINMENT FUNCTION SUCH AS t

CONTAINMENT ISOLATION, CONTAINMENT COOLING, CONTAINMENT SPRAY, OR POST-ACCIDENT HYDROGEN CONTROL).

l l

f 36

1994 AT-POWER PRECURSORS INVOLVING UNAVAILABILITIES j

I PLANT CCDP DESCRIPTION

[

4 HADDAM NECK 1.4 x 10 POWER-OPERATED RELIEF VALVES AND VITAL i

480-V AC BUS DEGRADED ZION 2 2.3 x 10-5 UNAVAILABILITY OF TURBINE-DRIVEN l

AUXILIARY FEEDWATER PUMP AND EMERGENCY DIESEL-GENERATOR POINT BEACH 1 & 2 1.2 x 10-5 BOTH DIESEL-GENERATORS UNAVAILABLE DRESDEN 2 6.1 x 104 MOTOR CONTROL CENTER TRIPS DUE TO IMPROPER BREAKER SETTINGS DRESDEN 2 3.1 x 104 LONG-TERM UNAVAILABILITY OF HIGH PRESSURE COOLANT INJECTION TURKEY POINT 3 & 4 1.8 x 104 LOAD SEQUENCERS PERIODICALLY UNAVAILABLE i

l

[

37 i

1994 AT-POWER PRECURSORS INVOLVING AN INITIATOR i

PLANT CCDP DESCRIPTION RIVER BEND 1.8 X 10 SCRAM, MAIN TURBINE-GENERATOR FAILS TO 5

TRIP, REACTOR CORE ISOLATION COOLING AND L

CONTROL ROD DRIVE SYSTEMS UNAVAILABLE I

CALVERT CLIFFS 2 1.3 X 10-5 TRIP, LOSS OF 13.8 KV BUS, AND SHORT-TERM i

UNAVAILABILITY OF SALTWATER COOLING SYSTEM i

1994 SHUTDOWN PRECURSORS INVOLVING AN INITIATOR PLANT CCDP DESCRIPTION i

WOLF CREEK 3.0 X 10-3 REACTOR COOLANT SYSTEM BLOWS DOWN TO REFUELING WATER STORAGE TANK DURING IIOT SHUTDOWN l

8 38 I

COMMON-CAUSE FAILURE DATABASE INSIGHTS e

ESTIMATED CCF PARAMETERS FOR OVER 40 COMBINATIONS OF RISK-IMPORTANT SYSTEMS AND COMPONENTS USING DATA FROM 1980 THROUGH 1993.

e BETA FACTORS DO NOT GENERALLY EQUAL 0.1.

j e

BETA FACTORS FOR SIMILAR COMPONENTS VARY AMONG SYSTEMS AND FAILURE MODES.

e CCF

SUMMARY

REPORTS CONTAINING CCF PARAMETER l

ESTIMATES FOR USE IN PRAS AND RISK STUDIES e

MAIN CAUSES OF FAILURE ARE INTERNAL TO THE COMPONENT (50%), DESIGN AND CONSTRUCTION (17%), AND HUMAN-RELATED (11%)

39

BETA FACTORS FOR MOTOR-OPERATED VALVES t

SYSTEM FAIL TO OPEN FAIL TO CLOSE BWR ISOLATION CONDENSER 1.1E-3 1.3E-3 BWR RESIDUAL HEAT REMOVAL 1.9E-2 2.8E-2 PWR HIGH PRESSURE INJECTION 5.0E-2 1.4E-2 PWR LOW PRESSURE INJECTION 6.1E-3 2.3E-3 PWR AUXILIARY FEEDWATER 2.7E-2 9.1E-3 l

40

i SYSTEM RELIABILITY STUDIES I

[

PURPOSE:

TO EVALUATE RELIABILITY AND PROVIDE INSIGHTS OF RISK

[

IMPORTANT SYSTEMS BASED ON OPERATING EXPERIENCE l

l OBJECTIVE:

e USE ACTUAL DEMANDS, FAILURES AND i

UNAVAILABILITIES TO ESTIMATE RELIABILITY e

ANALYZE TRENDS IN RELIABILITY o

QUANTIFY UNCERTAINTIES l

e COMPARE WITH PRA/IPE VALUES e

IDENTIFY PLANT SPECIFIC DIFFERENCES i

e PROVIDE ENGINEERING INSIGHTS i

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STUDIES BEING CONDUCTED l

i e

BOILING WATER REACTOR (BWR) SYSTEMS i

HIGH PRESSURE COOLANT INJECTION (HPCI)

REACTOR CORE ISOLATION COOLING (RCIC)

HIGH PRESSURE CORE SPRAY (HPCS)

(

ISOLATION CONDENSER (IC) l PRESSURIZED WATER REACTOR (PWR) SYSTEMS AUXILIARY FEEDWATER (AFW)

HIGH PRESSURE SAFETY INJECTION (HPSI) t e

BWR AND PWR LOW PRESSURE INJECTION SYSTEMS e

BWR AND PWR REACTOR TRIP SYSTEMS (RPS)

I e

EMERGENCY DIESEL GENERATORS (EDGs) 42 i

METHODOLOGY OVERVIEW STANDARDIZED AND SYSTEMATIC STUDY PROCEDURE l

e DETAILED EVALUATION OF EVENTS USING RISK ANALYSIS METHODS AND MODELS e

RIGOROUS MATHEMATICAL TREATMENT OF RELIABILITY AND AVAILABILITY DATA, INCLUDING UNCERTAINTIES i

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e DETAILED ANALYSIS OF RESULTS, INCLUDING l

INDEPENDENT PEER REVIEW l

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EDG INSIGHTS e

NO DISCERNIBLE TREND IN RELIABILITY FAILURE RATE AND UNPLANNED DEMAND RATE BOTH DECREASING i

i e

HIGHER FAILURE RATES FOR PLANTS LICENSED FROM 1980 TO 1990 e

THREE DISTINCT FAILURE-TO-RUN RATES l

e GENERAL AGREEMENT WITH PRAs AND IPEs i

s e

FAILURE DIFFERENCES BETWEEN ACTUAL DEMANDS AND l

ROUTINE SURVEILLANCE OR INSPECTIONS e

ACTUAL DEMAND FAILURES-TO-START NOT EASILY l

RECOVERABLE 47

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i EDG INSIGHTS (CONTINUED) l f

l e

NO COMMON-CAUSE FAILURES OF MULTIPLE DIESELS OBSERVED DURING ACTUAL UNPLANNED DEMANDS e

DEMAND RELIABILITY CONSISTENT WITH STATION l

BLACKOUT RULE e

MAINTENANCE OUT OF SERVICE (MOOS) MUCH HIGHER THAN EARLIER DATA USED IN STATION BLACKOUT RULE e

MOOS SIGNIFICANTLY HIGHER DURING SHUTDOWN THAN DURING OPERATION l

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