ML20198P564

Summary of 980114 Meeting W/B&W Owners Group Steering Committee Re Recently Completed & Currently Ongoing Studies in Safety Program Division.Copy of Handout Used at Meeting & List of Attendees Encl

ML20198P564
Person / Time
Issue date:	01/20/1998
From:	Rossi C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	Martin T NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
References
NUDOCS 9801220155
Download: ML20198P564 (65)

Text

Jcnuary 20, 1998 MEMORANDUM TO: Thomas T. Martin, Director Office for Analysis and Evaluation of Operational Data FROM:

Charles E. Rossi, Director original signed by:

- 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 January 14,1998. I attended this meeting to discuss recently completed and currently ongoing studies in the Safc y 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 />, and a list of the attendees.

Attachments: As stated Distribution w/atts.:

Public M:.A SPD R/F PBaranowsky JRosenthal DHickma 1 jYO? } I

@ 22g55 g20 gg_l3 l

PDR

~

y, y 1 p.Q H:\\SPD\\ROSSi\\B&WPUBLI.WPD OFFICE D:SPD g

g'} {

gg NAME--

CRossi:mg DATE 014f98 Fh' /3 hg gg OFFICIAL RECORD COPY

U 4

B&W Owners Group Steering Committee January 14,1998, Attendees Framatome Technologies J. J. Kelly Framatome Technologies Rick Edwards EPRI Gary Vine Duke Energy Corporation W.W. Foster Entergy Operations, Inc.

J A. Selva Florida Power Corporation Ken Wilson Toledo Edison Company F.L. Swanger H.C. Crawford GPU Nuclear, Inc.

Attachment

ao S

(,

1; OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA 1

... i i..

h

- i. _

CHARLES E. (ERNIE) ROSSI l

SAFETY PROGRAMS DIVISION AEOD WASHINGTON, D.C. 20555 (301) 415-7499 2

INDEPENDENT REVIEW-ALL LICENSEE EVENT REPORTS ARE REVEWED IN A SYSTEMATIZED WAY BY AEOD REVIEWERS TO CLASSIFY FOR SIGNIFICANCE THE DATA IS ENCODED IN TIE SEQUENCE CODING AND SEARCH SYSTEM TO ASSIST IN IDENTIFYING RELATED EVENTS CAUSAL AND TIME ASPECTS OF OCCURRENCES WITHIN THE EVENT SEQUENCE l

AEOD ALSO REVEWS FOREIGN EVENT DATA, IRS AND BILATERAL, TO ADD TO THE DOMESTIC EXPERIENCE 3

INDEPENDENT LONG TERM STUDIES AND ANALYSES i

e OPERATIONAL DATA SYSTEMS AND RELIABILITY DATA COLLECTION BROAD SCOPE SYSTEMATIC REVIEW OF OPERATING e

EXPERIENCE e

IDENTIFICATION OF SAFETY ISSUES INDEPTH COMPONENT, SYSTEMS, HUMAN e

PERFORMANCE STUDIES e

PERFORMANCE INDICATOR PROGRAM e

ACCIDENT SEQUENCE PRECURSOR PROGRAM e

SYSTEM RELIABILITY STUDIES 4

i l

t

~

ITEMS OF RECENT INTEREST i

i Senior Management Meeting Information Base l

Event Report Rules and Guidance I

i Reliability and Availability Database 9

5 l

\\

i ASSESSMENT OF PRESSURIZED-WATER

~

REACTOR PRIMARY SYSTEM LEAKS (Draft for peer reviewD A total of 240 leak events in 1985 through September 1996.

41 events considered risk significant.

Leak events that could be regarded as core damage precursors.

Six events contributed to transient-induced loss-of-coolant accident frequency.

l e

t Seven events contributed directly to small-break loss-of-coolant initiating event frequency.

j i

Five events analyzed by AEOD Accident Sequence Precursor j

Program.

e j

w Previously (prior to 1985) unidentified degradation mechanisms.

' Turbulent penetration and resulting thermal cycling causing.

thermal fatigue cracking of base metal and welds.

l Cavitation-induced vibratory fatigue failure of small-diameter piping.

Primary water stress corrosion cracking'of primary pressure boundary penetrations made of Alloy 600.

Transgranular stress corrosion cracking of a spare control element drive mechanism housing.

'7

s Leaks that have a potential for relatively rapid growth such that a small-break loss-of-coolant accident could occur.

Leaks through thermal fatigue cracks of branch lines.

Leaks through reactor coolant pump seals.

Significance of thermal fatigue cracking.

Thermal fatigue cracks in base metal were not expected and generally inspection was not required.

8' l

[

Detection and sizing of cracks in small-diameter branch lines is difficult.

i inspection was not required.

=

\\

The' leak-before break concept not applicable.

Failure mechanism not well understood.

l L

Timely detection of cracks become difficult if crack growth rates are high.

Small-break loss-of-coolant accident frequency estimated consistent with the generic small-break loss-of-coolant accident i

initiating event frequency (1E-2) used in many probabilistic risk assessments i

9

4 Frequency of reportable leak events has ' decreased since 1985' s

Improved valve packing.

Induced by vibrational fatigue.

t Frequency of pressure boundary leaks and leaks through bolted connections do not show a statistically significant trend.

t Current leak detection system.

Effective in detecting 1 gpm within an hour.

Not effective in detecting very small leakage.

Not effective in determining leak source location.

i i

i

10.

' 1.25Eo00 Point'est. and 90% confidence interval Fitted rate

-- - - 90% confidence band on the fitted rate 1.00E+00 l

s s

y 0.75E N 4"

+

O s

3 s

y 1

1, s s.

W 0.50E-00.

s.

s, 9

A

~4 f

~

~

~

~

l

~<-

s.

~~%s

~

~

0.25E-00.

~

p, s.

~ ~.

4, 1

_t 0.00E+00 85 86 87 88 89 90 91 92 93 94 95 96 Year C" * '

Statistical analysis of a trend in the frequency of reportable leak' events. Point estimates, 90 percent confidence bands, and 90 percent confidence intervals are shown.

11 1

\\

ONGOING REVIEW OF DESIGN BASIS j

ISSUES IN 10 CFR 50.72/73 REPORTS Review initiated in response to recent design basis issues identified at several plants.

}

The review is of events reported to the NRC as event i

notifications under 10 CFR 50.72 and as licensee event reports under 10 CFR 50.73.

The present scope of review includes only events occurring in CY 1997.

i The review is to identify the extent of design basis issues, where they exist (plants and systems), and their safety i

significance.

j 12 l

l i

GENERAL OBSERVATIONS a

BASED ON REVIEW OF LERs THROUGH JUNE 1997:

34 percent (296 out of 871) of licensee event reports included design basis. issues.

I 29 percent of plants submitted no licensee event reports with design basis issues.

i Older plants (operating license:

1964-1974) had significantly more design basis issues per plant per month than newer 4

plants.

l i

Point Beach 1 and 2, Millstone 1 and 3, and Cristal River 3 (all older plants) accounted for 28 percent or design basis issues.

-13 i

---..-4

l Four risk important systems (emergency core cooling, primary reactor systems, emergency ac/dc power, and containr'ent and containment isolation) were the top four contributors and accounted for 42 percent of design basis issues (see Figure 1).

Based on AEOD screening criteria, 29 percent of licensee event reports with design basis issues were,iudged important.

i Of all preliminary Accident Sequence Precursor analyses done to date, two events had conditional core damage probabilities i

greater than 1.0 E-6.

Both events involved design basis issues.

f i

l 14 l

Figure 1 YTD (through June 1997) Design Basis issues in LERs & Plant Systems involved Acc. Mon. Instr. (1) M 11 Aux 1Emerg. FW (2) N 15 '

Comb. Gas Control (3) M 7 Comp. Cool Water (4) to Cont. & Cont. Iso. (5) 27 Cont. Cool. Syst. (6) N 16, Cont.Rm. Em. Vent. (7) -

10 l

Emerg. AC/DC Power (8) 32 Emerg. Core Cooling (9) 55 ESF Instrument. (10) m 13 f

Essent. Comp. Air (11) M 2 Essent. Ser. Wtr. (12) m l 19 I

Firo DetJSuppr. (13) M g l

Isolation Conden. (14) M 4 LTIOP Protection (15) 0 MSIVs (16) W 2 Pri. Reactor Sys. (17) 36 i

Rad. Mon. Instr. (18) M 6 Rx. Core is. Cool (19) 0 Rx. Trip Instr. (20) m 6 RHR Syst. (21) M 9

{

Safe. & Rel. Viv. (22) M 4 i

Spent Fuel Syst. (23) m 11 Stdby. Liq. Cont. (24) 0 l

Ultimate Heat Sink (25) W 2 Other (26) 53 0

10 20 30 40 50 60 COUNT OF LERs WITH DBis i

u s

y

" REVIEW OF UNDETECTED FAILURES OF

^

SAFETY SYSTEMS" Issued on September 26,1997

)

Study initiated to review undetected failures' of safety systems in nuclear power plants during a selected 3-year period The study is a further analysis of data provided to Nuclear Energy Agency (NEA) in support of a worldwide study i

i The study identified a set of 33 such failures (events) by a I

search of 70 events in the Accident Sequencc Precursor l

(ASP) database for the period 1991-93 t

i Undetected failures, include significant " events" or " conditions" where equipment j

was inoperable or where it was subsequently determined that the equipment would not i

perform its safety function. Therefore, the undetected failures include more than purely

{

equipment failures.

16 l

t

[

.,c.-

Undetected failures were analyzed with? respect lto their:

Discovery methods Failure rate Failure causes Corrective and preventive actions by licensees

. 17

..i-l i

.sr-.

i-:i.m--

--s

i..- -. - - -....-..-.....-........

m.

---......, i..

- - -....--- -----.-.-i..... -. -- i.. -.. --... - - -

m s,-r-

- '

• i -

FINDINGS AND CONCLUSIONS Nearly 50 percent of the ASP events involved an undetected failure Failures existed in systems important to safety Failures remained undiscovered for a long time, some up.to 1

18 years (see Figure 1)

More.than 75 percent of the failures were discovered via testing

{

or analysis and evaluation of operational problems i

i Although testing was the most frequent discovery method, in approximately 60 percent of the LERs the deficiencies revealed I

were-not within the scope of test and were inadvertently discovered under fortuitous circumstances.

18 i

4

2 No. of Events 12

- ~ - - - - -

i 10 i

i l

g 1

i i

i i

6

~

i 4

I i

I i

2 l

0 --

y/

/

J

/

/

\\

A o

O O

A o

g s

e so 1

y+o#

+

M 4

s s

1 4

j Time Figure 1:

T!me Elapsed Before Discovery 1

19 i

L

" REVIEW OF. INDUSTRY EFFORTS TO MANAGE PRESSURIZED WATER REACTOR FEEDWATER NOZZLE, PIPING, AND FEEDRING CRACKING 1

AND WALL THINNING (NUREG/CR-6456 (E97-01) i Issued March 1997 l-Comprehensive overview of pressurized-water reactor events l

4 Initial feedwater pipe cracking in 1979 4

Increased event frequency in early 1990's Reviewed operating experience from 1979 through 1996 i

L Operating experience from foreign plants 1

20 n

Babcock and Wilcox Company plants IJnique design and no events Focused on feedwater system adjacent to nozzle Main and auxiliary feedwater piping, thermal sleeve, feedring

-and J-tubes i

i l

Understand technical issues to establish safe operating limits i

!t Concentrate on causes, mechanisms, conditions, inspections, procedures, corrective actions f

t l

t 21 l-l

\\.

I i

~

l:

i

[

TECHNICAL ASPECTS

'ft l

p l

Main and auxiliary feedwater system

,y Feedwater system design and safety significance of rupture Fatigue cracking experience 1

i Flow-accelerated corrosion-induced wall thinning experience

{

c Steam generator water hammer damage experience 4r Degradation mechanisms i

e t

Inservice inspection metheds

?

c Mitigation, monitoring and replacement 22 1

i

FINDINGS Cracking and wall thinning can be managed effectively.

Combined analysis, inspection, monitoring, mitigation (including operational procedures), and replacement techniques Tools to manage issue are available Management process requires detailed knowledge in several areas Component and system design, construction, and materials Operating procedures (especially extended operation at startup or hot standby with automatic auxiliary feedwater control) 23

indepth understanding of factors that cause thermal fatigue c

and flow accelerated corrosion l

Adequate training in the use of predictive analysis methods t

Use of advanced inspection techniques i

1 j

't.

I p

l f

t i

i s

I i

i 3

24 I

!=

i

?-

i-

AEOD ENGINEERING EVALUATION:

" NUCLEAR POWER PLANT COLD WEATHER PROBLEMS AND PROTECTIVE MEASURES" issued December 19,1997 Study initiated as a result of frazil ice event at Wolf Creek, January -30,1996 icing of cooling water intake trash racks and traveling screens.

l Subsequent loss of an essential service water system train, with complications.

Reviewed events occurring between January 1991 and April 1997 25

Events discussed include both. actual failures due to cold weather and some design vulnerabilities which-could lead to failures

-There were 37 events at 23 different sites.

Conditional core damage probability of Wolf Creek event 1.2E-4.

1 1

Other events below Accident Seouence Precursor screening

-criteria.

Report identifies previously issued NRC and industry communications includes corrective actions taken to resolve issues.

2S

3 l

I f

il l

l,

.sno 7

i 2

tac inu m

m o

s c

n y

o i

r t

t a

s r

u a

d e

p n

c e

n r

i d

n e

p i

S n

g r

r a

e e

i N

s p

h O

C e

x t

g d

R e

a n

I e

i T

N n

g w

d c

i n

A e

s e

a z

r i

V e

t d

t t

i p

s a

i l

l r

o a

t R

s s

e c

e a

E e

e t

d n

h p

r e

S k

g o

o y

h t

l r

i e

B u

e r

f a

s l

o n

u v

O c

c w

e o

f i

i t

v w

t t

c n

o r

o e

n o

e o

e n

f o

o i

i n

v t

f t

t s

g e

t a

e a

r a

n n

l e

t i

n s

u e

e e

t e

i u

s n

t n

m r

v d

e e

o n

i i

o t

e l

i n

f p

c n

g g

o f

o i

f e

n n

c o

m k

f h

s s

i i

o u

s s

c p

s s

c s

t e

a i

i n

s L

n n

t M

M n

I i

4 e

u e

v a

v E

C E

s c

n i

<l

j e

t Missing or ineffective seals for electrical conduits and

{

cabinets.

i Grease and oil viscosity concerns during coid weather.

1 Several components / systems affected by cold weather.

Icing of intake structures.

I Process line freezing.

instrument line waterproofing, heat tracing.

1 i

Reduced emergency diesel generator oil and grease viscosities.

s h

l Essential chiller operation at low heat loads.

c Electrical systems.

I l-l 4

28 j

f I

l Actions taken by the' NRC

~

Actions planned to revise inspection Procedure 71714, " Cold Weather Preparations" using inforrriation contained in the report to make the inspection procedure more explicit and operating experience focussed.

Information Notice to be issued.

l l

29

1..,...

RAB REPORTS ISSUED IN 1996 SPECIAL STUDi REPORTS DATE TITLE NO.

AUTilOR 04/96 STEAM GENERATOR TUBE FAILURES NUREGICR-6365 INEL

- 06/%

OCONEE ELECTRICAL SYSTEM DESIGN AND OPERATION G. LANIK 09/% -

ASSESSMENT OF SPENT FUEL COOLING J. IBARRA i

ENGINEERING EVALUATIONS DATE TITLE NO.

AUTIIOR 03/%

MOTOR-OPERATED VALVE KEY FAILURES E %-01 C. IISU 04/%

ANALYSIS OF ALLEGATION DATA E%02 S. ISRAEL 09/%

INSIGIITS FROM UNDETECTED FAILURES OF E%XX S.PULLANI SAFETY SYSTEMS TECIINICAL REVIEW REPORTS DATE TITLE NO.

AUTilOR 03/%

AEOD TECIINICAL REPORTS BY CATEGORY REVISION i T96-01 S. ISRAEL I

04/%

TARGET ROCK TWO-STAGE SRV PERFORMANCE UPDATE T96-02 M. WEGNER j-09/96 RESPONSE OF B&W PLANTS ON LOSS OF T96-03 W. RAUGIILEY NONEMERGENCY AC POWER i

30

RAB REPORTS ISSUED IN 1997 CASE STUDY REPORTS DATE TITLE NO.

AUTHOR 03/97 Grid Performance Factors C97-01 M. Wegner l

l I

SPECIAL STUDY REPORTS DATE TITLE NO.

AUTHOR 03/97 Oconee Electrical System Design S97-01 G. Lanik and Operation H. Ornstein W. Raughley J. Thompson 31 1

RAB REPORTS ISSUED IN 1997 (Cont.)

ENGINEERING EVALUATIONS DATE TITLE NO.

AUTHOR 04/97 Review of Industry Efforts to Manage NUREG/CR-6456 Pressurized Water Reactor Feedwater (E97-01) i Nozzle, Piping, and Feedring Cracking l

and Wall Thinning 09/97 Review of Undetected Failures of E97-02 S. Pullani Safety Systems E. Brown 12/97 Nuclear Power Plant Cold Weather E97-03 M. Padovan Problems and Protective Measures TECHNICAL REVIEW REPORTS DATE TITLE NO.

AUTHOR 01/97 Design Errors in Nuclear Power Plants T97-01 S. Pullani 32

Future Products Development of Financial Indicators for use by Senior Management Air-Operated Valve Operating Experience Safety-Related Electrical Breaker Problems l

Emergency Diesel Generator Control Systems Common Findings in Diagnostic Evaluation Team Rer its Design Basis Findings Turbine Hall Hazards Effectiveness of Post Trip Review 33 Human Performance ?n Significant Events

RISK BASED ANALYSIS OF REACTOR OPERATING EXPERIENCE USE REACTOR OPERATING EXP ~.'RIENCE TO:

e ASSESS AND TREND RISK INDICATORS COMPARE WITH PROBABILISTIC RISK ASSESSMENTS (PRAs) AND INDIVIDUAL PLANT EXAMINATIONS (IPEs)

IDENTIFY TECHNICAL INSIGHTS RELATING TO RISK e

CONTRIBUTORS i

PROVIDE INSIGHTS TO INDUSTRY AND REGULATORY e

ACTIVITIES RELATED TO RISK 34

PROGRAM ELEMENTS ACCIDENT SEQUENCE PRECURSOR (ASP) ANNUAL REPORT ASP METHODS ASP DATABASE e

INITIATING EVENTS PERIODIC REPORT LOSS OF OFFSITE POWER (LOOP) DATABASE i

j SPECIAL INITIATORS (e.g., POWER-OPERATED RELIEF VALVES, STEAM GENERATOR TUBE RUPTURES, HUMAN ERRORS)

SYSTEM RELIABILITY STUDIES RELIABILITY INDICATORS COMPONENT ANALYSES 4

35 1

[

PROGRAM ELEMENTS (CONT]NUED) b COMMON-CAUSE FAILURES (CCFs) i.

DA, ABASE AND ANALYSIS SOFTWARE PERIODIC ANALYSIS INTERNATIONAL COMMON-CAUSE DATA EXCHANGE EFFORT L

i i

PERFORMANCE INDICATORS (RISK / RELIABILITY) i DATA SYSTEMS j

SEQUENCE CODING AND SEARCH SYSTEM (SCSS) i' AND EQUIPMENT PERFORMANCE AND INFORMATION EXCHANGE (EPlX)

RELIABILITY DATA

[

36 L

l i

l i-ACCIDENT SEQUENCE PRECUBSOR PROGRAM I

i DETERMINE CONDITIONAL PROBABILITY OF SUBSEQUENT i

i

~

SEVERE CORE DAMAGE CONDITIONAL CORE DAMAGE PROBABILITY (CCDP)

GIVEN THE FAILURES DURING AN OPERATIONAL EVENT i

[

l j

i l

37 l

OBJECTIVES OF THE ASP PROGRAM IDENTIFY AND RANK RISK SIGNIFICANCE OF l

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

(

i PROVIDE SUPPLEMENTAL INFORMATION ON PLANT-l SPECIFIC PERFORMANCE l

[

t PROVIDE A CHECK WITH PRAS t

I PROVIDE AN EMPIRICAL INDICATION OF INDUSTRY RISK l

AND ASSOCIATED TRENDS i

!r 38 i

(

i i

l 1

E.--

t 1996 AT-POWER PRECURSORS INVOLVING UNAVAILABILITIES i

PLANT ACDP DESCRIITION t

i IIADDAM NECK 1.1 X 10 POTENTIALLY INADEQUATE RIIR PUMP NPSH 4

FOLLOWING A LARGE-OR MEDIUM-BREAK LOCA SEABROOK' 4.6 X 10-5 TURBINE-DRIVEN EFW PUMP UNAVAILABLE i

BECAUSE OF A MECIIANICAL SEAL FAILURE t

SALEM 1 & 2 5.8 X 104 CIIARGING PUMP SUCTION VALVES FROM TIIE RWST POTENTIALLY UNAVAILABLE BECAUSE OF PRESSURE LOCKING h

IIADDAM NECK 2.9 X 104 AFTER AN RIIR FUMP SEIZED, IT WAS

[

I DETERMINED TO IIAVE BEEN SUSCEPTIBLE TO FAILURE SINCE TIS OVERIIAUL IN 1987 e

i t

MCGUIRE 2 1.8 X 104 2B EDG INOPERABLE DUE TO SLOW i

[

INSTRUMENTATION RESPONSE i

c l

l 39 i

i l

1996 AT-POWER PRECURSORS i

INVOLVING AN INITIATOR FLANT CCDP DESCRIITION CATAWBA 2 2.I x 10-3 LOOP WITII EDG B UNAVAILABLE j

WOLF CREEK 2.1 x 10" REACTOR TRIP WirII LOSS OF TRAIN A OF ESSENTIAL SERVICE WATER AND TIIE TURBINE-DRIVEN AFW PUMP 1

PRAIRIE ISLAND 1 & 2 5.3 x 10-5 LOOP TO SAFEGUARDS BUSES ON BOTII UNITS i

LASALLE 1 & 2 7.0 x 10 CONCRETE SEALANT FOULS COOLING WATER 4

i SYSTEMS ANO, UNIT 1 5.6 x 10 REACTOR TRIP AND SUBSEQUENT SG DRYOUT 4

1996 SHUTDOWN PRECURSORS INVOLVING AN INITIATOR PLANT CCDP DESCRIITION i

BYRON 1 I.7 X 10-5 TRANSFORMER BUS FAULT CAUSES A LOOP 40 l

l i

1997 AT-POWER PRECURSORS INVOLVING UNAVAILABILITIES i

i i

PIAhl ACDP*

DESCRIPTION MAINE YANKEE 1.3 X 10-5 REACTOR COOLANT SYSTEM IIOT LEG I

RECIRCULATION VALVES SURJECT TO PRESSURE LOCKING BECAUSE OF POST-LOCA TIIERMAL EXPANSION OF TIIE TRAPPED WATER l

• DENOTES PRELIMINARY RESULTS

)

1997 AT-POWER PRECURSORS INVOLVING AN IlxITIATING EVENT PLANT CCDP*

DESCRIFFION OCONEE 3 a.9 X 175 TWO IIIGII PRESSURE INJECTION PUMPS WERE DAMAGED BECAUSE OF A LOW WATER LEVEL IN TIIE LETDOWN STORAGE TAhX i.

• DENOTES PRELIMINARY RESULTS 41 i

f f

i

~

SYSTEM RELIABILITY STUDIES l

i PURPOSE:

TO EVALUATE RELIABILITY AND PROVIDE INSIGHTS OF RISK j

IMPORTANT SYSTEMS BASED ON OPERATING EXPERIENCE L

1 OBJECTIVE:

i i

}

USE ACTUAL DEMANDS, FAILURES AND e

UNAVAILABILITIES TO ESTIMATE RELIABILITY l

e ANALYZE TRENDS IN RELIABILITY

)

e QUANTIFY UNCERTAINTIES COMPARE WITH PRA/IPE VALUES e

IDENTIFY PLANT SPECIFIC DIFFERENCES e

i e

PROVIDE ENGINEERING INSIGHTS 42 i

b

METHODOLOGY OVERVIEW 1

e STANDARDIZED AND SYSTEMATIC STUDY PROCEDURE l

1 e

DETAILED EVALUATION OF EVENTS USING RISK L

ANALYSIS METHODS AND MODELS I

e RIGOROUS MATHEMATICAL TREATMENT OF l

l RELIABILITY AND AVAILABILITY DATA, INCLUDING UNCERTAINTIES DETAILED ANALYSIS OF RESULTS, INCLUDING I

INDEPENDEl'.'T PEER REVIEW l

i r

43 l

L

-l 1

RELIABILITY STUDIES i

Recent Accomplishments

[

l RCIC System Reliability report issued i

Fire Events Study issued I

t

[-

Ongoina Studies t

Auxiliary Feedwater System (PWR)

Reactor Protection System t

High-Pressure injection (PWR) 1

{

l High-Pressure Core Spray (BWR) i i

initiating Events t

e t

44 j

/

1

y SYSTEM RELIABILITY STUDY RESULTS

SUMMARY

~

OPERATIONAL MISSION Unplanned Consistency UnrF.m aity Demand vs Mean

. Demand Failure Rate Unreliability with vs. - Age Test Failure Study Unreliability Trend Trend Trend PRA/IPEs Trend Differences General HPCI:

O.056 Decreasing Decreasing Steady agreement-Few None Yes-plants lower than op. experience General EDG (RG1.108).

0.044 Decreasing Decreasing Steady agreement-FTR None Yes higher in PRAs General IC -

O.02 Steady Steady Steady agreement-Nature None No of failures differ RCIC General Short (< 15 min) 0.04 Decreasing Steady Decreasing agreement-Restart None No Long (> 15 min) 0.18-different in PRAs injection-factor of HPCS (DRAFT /

O.075 Decreasing '

Steady Steady five greater than None N/A PRA/IPEs Average factor of AFW (DRAFT 1 3C 5 Decreasing Steady Steady six greater than None N/A PRA/IPEs 45

_.__m_

CURRENT PERFORMANCE INDICATORS

^

AUTOMATIC SCRAMS WHILE CRITICAL SAFETY SYSTEM ACTUATIONS SIGNIFICANT EVENTS SAFETY SYSTEM FAILURES FORCED OUTAGE RATE EQUIPMENT FORCED OUTAGES PER 1000 CRITICAL HOURS COLLECTIVE RADIATION EXPOSURE CAUSE CODES 46 l

l l

. MOVE TO RISK-BASED PERFORMANCE INDICATORS ~

e Support PRA Implementation Plan e

Support PRA Policy Statement L

e Characteristics of Risk-Based Performance Indicators i

Decomposition of risk into constituent parts Collection of data relating to those parts Analysis of data to produce indicators and trends 47

8 y

4 S

t o

R i

l t

i O

b se T

a sc i

en A

e ca l

C r

om I

e r

D h

pro N

t f

Mr I

n e

E Mp i

C a

St t

N n

a e a A

d hl M

g t p R

n rloa i

O s

f u F

u d d R

s nv i

E l

P P

ai d t n d

r D

i e

o E

s pd S

a en A

b r a I

B k

P yr s

t K

e s ir S

e h u t s t

I d

R n a r n eb oi T

ma f

s N

et ss a

E l

l e

pd P

M my s

d s E

e a i t L

i P

di sd l

nb a n M

aa b a l

I i

pa k r O

ov so T

i le a r t i

N vd e n e n s o A

Da Um LP e

e r

f Ill

n~,

n CANDIDATES FOR RISK-BASED PERFORMANCE INDICATORS l

e initiating Events Frequency e

Risk Important Systems Reliabilit.y e

ASP-type Analysis of Significant Events 6

e Integrated CDF Indicator 49 dei w

Y

}T

(.

RELIABILITY AND AVAILABILITY DATABASE Davelop a specification for the reliability and availability database e

Collect data from INPO's EPIX system and.SSPI e

Gather risk insights into plant-specific and industry-wide safety e

performance Use information to monitor and assess trends in performance 50

NUREG-1022, REVISION 1 EVENT REPORTING GUIDELINES,10 CFR 50.72 AND 50.73 i

BACKGROUND Regulatory impact Survey,1989 and 1990 Event Reportiieg Workshops,1990 Draft of Revision 1 published,1991 i

Public meetings,1992 and 1993

)

Second draft of Revision 1 published,1994 i

i 51-t i

i

a' l

1 PUBLIC COMMENTS t

A.

Burden B.

ESF actuation t

C.

Relief valve example i

D. Discovery date definition i

E.

Suitable redundancy example F.

Design basis guidance G.

PRA guidance H. Isolation valve example l.

Eight hour standard i

J.

Charging pump example 3'

K.

Voluntary LER l

1 52

{

L KEY CHANGES IN GUIDANCE i

i

' Telephone notification is not required for some minor reports to other e

agencies.

a i

Licensees may retract telephone notifications and cancel licensee event reports where appropriate.

l Credit may be taken for limiting condition for operation a6: awed outage time before reporting a missed surveillance test.

Actuation of specific systems used to mitigate consequences of a significant event 60uld be reported as engineered safety feature actuations (a list is i

provided).

Actuation of a single engineered safety feature component is usually not l

+

reportable, however, actuation to mitigate an event is reportable.

1 i

e Lack of redundancy should be reported as outside design basis.

New guidance is providet' Ior emergency conditions discovered after the fact.

i Additional guidance is provided for licensee event report preparation.

P

l PLANS Publish Revision 1 in final form in early 1998 Initiate rulemaking

)

Move towards risk-informed regulation Resolution of issues with current rules

\\

i Better alignment with current NRC needs 1

l 54

x-RULEMAKING Issues with current rules Not sufficiently risk-informed One-hour time limit Outside design basis

=

Other nonemergency events

=

Outside design basis of the plant Extent of design basis

=

Extent.of reporting

=

Actuation of an engineered safety feature l

Consistency I

l

=

Risk-significance of systems

=

Operation or condition prohibited by Technical Specifications

=

Lai.a or missed surveillance tests Clarity regarding administrative requirements

=

Potential new directions Significance testing (e.g., seriously degraded condition)

Shutdown events (e.g., reactor coolant system draining)

Licensee event report simplification (e.g., reduce effort 55 involved) 1

I l

.I IMPROVEMENTS TO l

THE INFORMATION BASE OF THE SENIOR MANAGEMENT MEETING t

56 i

l l

COMMISSION REQUIREMENTS Evaluate the development of indicators that can provide a bases for judging whether a plant should be placed on.or removed from the watch list 1

Identify objective, meaningful, leading performance indicators Consider the extent to which objective performance indicators can be used in the decision processes, with new indicators being phased in as appropriate 57

a L

PERFORMANCE PARAMETERS MONITORED Safety system failures Forced outage rate Collective radiation exposure Equipment forced outages Administrative cause code Maintenance cause code i

Design cause code Other personnel cause code 1

58

,.r a

PLANT 103 - TREND MODEL 8

6-m 2-A A

a

+

L-A a

a-

-c a-

• ----o--.--- c a

c a

o 0

192 292 392 492 193 293 393 493 194 294 394 494 195 295 395 495 196 296 396 496 Hits -- _- Threshold a

Industry Average 100 Forced Outage Rate (percent) 10

• 'Y EI" *" "

80 -

8 60 -

6-40 -

4

~

g 9

192 392 193 393 194 394 195 395 196 396 192 392 193 393 194 394 195 395 158 396 (Hits based on 6-quarters of data) 59

4 PLANT 103 REGRESSION MODEL 100 --

y 80 --

e

& 60 --

2 40 --

.8

$ 20==.---1~~~

~ ~ ~ L* * -

. m.- _ _ _ _ _, _ _ _ _.,........

0 i

192 392 193 393 194 394 195 395 196 396 Tir.re

[~

ProbabHity - - - - - -Regional Average- -industry Average l Capacity Factor (24 mo.)

Administrative Cause Code (12 mo.)

100 30 l

S-so.

60 -

2

~~E

=

5-40 10 -

''I~~~*----

20 -

5-w 0

0 192 392 193 393 194 394 195 395 196 396 192 392 193 393 194 394 195 395 196 396 60

j i

iq ECONOMIC TRENDS l

4 i

Evaluated correlation of economic parameters with 1

past discussion plant list i

L Site indicators showed better correlation than i

corporation indicators l

l

[

Developed a trending model with multiple indicators s

t 61 I

h e

~

SITE MODEL INDICATORS

~

t Revenue factor i

Coverage i

Production cost per megawatt hour generated l

a Nonfuel operations and maintenance costs i

i 62

t; SCHEDULE / MILESTONES

'j Trial use in recent { January 1998? Senior Management Meeting Process Public comment period and workshop: Spring 1998 l

e Recornmendation for Commission decision:

Summer 1998 i

Implementation: End of Calendar Year 1998 i

i t

i 63 i

i