ML20059M714

Intervenor Exhibit I-MFP-13,consisting of Re LER 1-92-001-00

ML20059M714
Person / Time
Site:	Diablo Canyon
Issue date:	08/17/1993
From:	Rueger G PACIFIC GAS & ELECTRIC CO.
To:
References
OLA-2-I-MFP-013, OLA-2-I-MFP-13, NUDOCS 9311190261
Download: ML20059M714 (11)

Text

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April 30, 1992

'93 CCT 23 P5 :15 PG&E Letter No. DCL-92-Ill U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.

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Re:

Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Licensee Event Report 1-92-001-00 Violation of Technical Specification 4.0.5 due to a Previously Unidentified Check Valve Safety Function Gentlemen:

Pursuant to 10 CFR 50.73(a)(2)(1)(B), PG&E is submitting the enclosed Licensee Event Report concerning a violation of Technical Specification 4.0.5 due to a previously unidentified check valve safety function.

This event has in no way affected the health and safety of the public.

Sipcerely, f

sm, y Gregcry M. Rueger cc:

Ann P. Hodgdon John B.-Martin Philip J. Morrill Harry Rood CPUC Necan ucutucu couruson Diablo Distribution 1NPO

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UCENSEE EVENT REPORT (LER):

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On April 1,1992, it was detemined that Units I and 2 were in' violation of Technical Specffication 4.0.5, since-volume control. tank.' outlet-chack valve :

CVCS-8440 was not being tested in-accordance with ASHE'Section'XI. 'This condition was identified during a review of an INPO Nuclear Network Operating. Experience entry regarding a different potential recirculation flow leakage path..

1 The roo6cause for"fne exclusion of CVCT"8440 from-the ASMETection'XI Inservice-fisting' program was that its safety function.had not previously been identified. 'Because LCV-Il2B and LCV-112C upstream of CVCS-8440 were considered to' bet the flowpath. boundary -

valves, the function of CVCS-8440 was thought only to betisolation:from another closed piping system.

~

Corrective actions to prevent recurrence will include:

(1)l development of a valve; test:-

procedure for CVCS-8440 to periodically test for potential > post-LOCA recirculation gross J

leakage, (2). revision of DCPP Design Criteria Memorandum S-8 to incorporate ir. formation deterstned as a result of the current investigations.-and ?(3) a review lof all valves.in.

the post-LOCA recirculation flow path to determine if'any other valves perfom a-previously unidentified safety function, due to*a similar configuration.'

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Plant Conditions Units 1 and 2' h' ave bon in various modes at various power levels with'out-i periodic, reverse-flos inservice testing of check valve CVCS-8440.

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II.

Descriotion of Event 1

}

A.

Sussiary:

On April 1, 1992, it was determined that Units-l-and 2 were-in-violation.of Technical. Specification (TS) 4.0.5.

Volume control' tank-(VCT). (CB)(TK) outlet check valve CVCS-8440. (CB)(V) perfoms a-safety.

j function as a boundary valve during post-LOCA recirculation.-

1 Therefore, it should have been included:in the Diablo Canyon Power Plant (DCPP) Inservice Testing;(IST) program and tested periodically' 1

in accordance with ASME Section.XI. This condition was identified during a review of,an INPO Nuclear h twork Operating Experience lentiyL

-j regarding a different potential recirculation' flow leakage. path-identified at another nuclear power plant'

(

l TS 4.0.5 requires ~ that inservice testing of ASME Code Class ! J2,~

i and 3 components shall be perfomed in accordance with-Section XI ~of' l

the ASME Soiler and Pressure Vessel-Code. -Section XI specifies.

~

required testing intervalsifor Class'1, 2, or.3 valves:that perform a r

}

function in mitigating the consequences'of. an ' accident 'or in bringing -

l the plant to a cold shutdown condition.:

B.

Background:

j i

f On February 7-9, 1983, a working meeting was held between the.NRC-(NRR, Region V resident inspectors, contractor EG&G,IdahtInc.):and-PG&E.4DCPP staff) regarding the proposed DCPP'IST orogram. :CVCS-8440 i

l was specifically discussed and it was agreed that ihe valve was

~J j

nonsafety-related. Motor-operated VCT discharge isolation valves 1

[

LCV-ll2B and LCV-Il2C.(CB)(ISV) upstream'of.CVCS-8440 were considered j

to perform the safety function of; isolating the VCT. ' Potential ~

]

j backleakage past CVCS-8440 to the : seal water heat exchanger was not identified.

d i

On May 2,1985, Supplement 31. to the Diablo Canyon 1 Safety Evaluationt i

Report was-issued providing the.NRC's evaluation of the DCPP IST-j program. CVCS-8440 was not discussed in the safety evaluation.

4 l

On August 29, 1988, NRC Information Notice-(IN) 88-70,~* Check Valve.

i Inservice Testing Program Deficiencies," was issued'to provide j.

notification of potential problems identified:during NRC' inspections j

of check valve IST programs at several. plants.~ PG&E initiated a i

review of safety-related check valves at' DCPP.

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On April 3, 1989, NRC Generic Letter (GL) 89-04, " Guidance-on-4 Developing Acceptable Inservice Testing Programs," was -issued noting; similar generic concerns and requiring that implementing test procedures be~ reviewed and revised as necessary. No response.to GL 89-04 was required from DCPP because the NRC had already completed their review of the DCPP IST. program (in 1985)..Howeverp the testing methods for check valves in the DCPP IST program were: reviewed for consistency with GL 89-04 guidance in 1991,' as a result of _ check valve testing deficiencies discussed in LER l-84-044-01.

On July 11, 1989, the review of safety-related' chebk valves initiated -

in response to IN 88-70 was completed. On' December 7.-1989,-the DCPP:

Plant Staff Review Committee reviewed.and: approved its response to-IN 88-70.. Two check valves.(8998A and'89988) were found not.to have q

been leak' tested in accordance with'ASME Section XI.

On July 16, 1990, LER l-84-044-00 was. submitted to the NRC to report-the violation of TS 4.0.5 resulting from'the failure to fully test check valves 8998A and 89988.- The valves were-being. tested in_ the open position to verify flow, but were not being tested in the closed position.

Design Criteria Memorandum.(DCM) S-8, "Chemicalland Volume Control-System(CVCS), dated December-10, 1990, does-not specify any safety functions performed by-VCT outlet check valve CVCS-8440. 'However, CVCS-8440 does in fact perform:the safety-related function of preventing: loss of reactor coolant inventory during post-LOCA recirculation.

Failure of this valve to'close could potentially allow post-LOCA recirculation fluid to' pressurize the piping fros' the seal waterheatexchangerl(CB)(HX). The' increased pressure could lift the seal water heat exchnger relief valve,'which discharges 'to' the VCT (outside of containment).

On April 2, 1991, LER l-84-044-01 was submitted to'the.NRC to. update-LER l-84-044-00. Additional deficiencies (other check valves in the IST program that were not being fully tested) had been discovered as a result of a corrective action in LER 1-84-044-00 to review the IST program plan for consistency with GL 89-04. 'The revised corrective actions to prevent recurrence included a review of testing and; design-basis requirements for check valves in the IST program plan to' ensure consistency with the guidance in GL 89-04. The LER corrective ~ actions addressed proper testing of valves that were ~already within'the" scope-of the IST program. 'The corrective actions did not. identify any problems with.CYCS-8440_(CVCS-8440 was"not in the:IST program because c

it did not have an identified safety function)(

A 1991 INPO Nuclear Network Operating Experience entry identified a potential recirculation flow leakage path outside of cantainment.-.

That scenario identified certain conditions where the' centrifugal 10215/85K Y

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un on charging pump ~(CCP) (8Q)(P) minimum-flow isolation' valves (8Q)(ISV) would be open during the post-LOCA recirculation. phase. This would allow the CCPs to pressurize the ~ seal water heat exchanger piping and-lift its relief. valve to the VCT,.stallar to thel situation ~ postulated

]

1 in the current scenario.

i C.

Event

Description:

On October 21, 1991, System Engineering initiated an ' evaluation to I

detenmine whether CVCS-8440 performs a safety function in the l

reverse-flow direction. An INPO-Nuclear Network Operating Experience-

)

entry had described a potential recirculation flow leakage path identified at another nuclear power plant. Review of this scenario.

indicated that it was not applicable to DCPP because-of a difference in valve control configuration. - However, System Engineering. realized.

l i

that CVCS-8440 appeared to protect against a'similar scenario,-

although no such safety function was specified in the DCM for the 3

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system.

t On November 7,1991, initial results indicated that' CVCS-8440 does i

perform a safety function in the reverse-flow direction; i.e., it' i

prevents loss of reactor coolant inventory 'during post-LOCA

~

recirculation. However, System' Engineering did not' consider the i

current scenario.to'be an immediate concern, because:

(1) the

~

scenario was thought to be' applicable only to a very small range of l

LOCA break sizes, (2) there were no indications of an actual-equipment problem, and (3) usually, it-is direct leakpaths: to ats~ahere that -

are of immediate concern. Although the evaluation indicated that the leakpath was possible, System Engineering did.not' consider it.to; be a' l

credible accident scenario.

i

" On February 20, 19 % a Technical Revie 7 Group (TRG) was convened to i

review the potential leakpath scenario. -Backleakage through CVCS-8440 1

could cause a pathway for release of post-LOCA recirculation fiuid in 1

this scenario (see Figure 1). During post-LOCA recirculation, wotor-i operated isolation valves LCV-1128 and LCV-112C upstream of CVCS-8440 l

would be closed to isolate the VCT from recirculation backflow.

j However, piping from the seal water heat exchanger, while upstream of.

CVCS-8440, is downstream of LCV-1128 and LCV-112C. Therefore, backleakage past CVCS-8440 could theoretically pressurize this piping j

until its relief valve, RV-8123, lifted and allowed flow to-the VCT..

i If the VCT were to fill and overpressurize, its relief valve, RV-8120, j

could lift and allow flow to the liquid bid-up tanks (LHUTs)

(CA)(TK).

Both the VCT and the LHUTs are located outside of J.

containment.

l The LHUTs have a large capacity; however, if they were also to fill and overpressurize, they would relieve to the unfiltered auxiliary-1 j

building atmosphere.

In addition, the piping fros'RV-8120 to the t

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UCENSEE EVENT REPORT (LER) TEXT CONTINUATION 18330'G 1

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tta, ( m LHUTs is not seismically supported. Therefore, assuming no operator intervention, this pathway was initially thought.to represent the.

potential for:~ (1) diversion of a large quantity of recirculation' inventory (initial estimate was 93,000 gal, conservatively.-' assuming full reverse-flow through CVCS-8440), and-(2) an unfiltered post-LOCA release outside of containment.

On April _1, 1992, the TAG determined th'e. event:was reportable to the NRC. The TRG received information fros Westinghouse and detemined thatCVCS-8440performsasafetyfunction'inmitigatingthe'findicated' consequences of an accider.t.. Discussions with Westinghouse c

that the potential leakpath scenario had not been previously identified by. Westinghouse or other utilities.= Therefore,-the valve 3

should have been included in the DCPP:Section XI-IST program, and its exclusion was a violation of TS 4.0.5.. Because there.is no one j

specific data at which the safety function should have been realized, the event d.te'is the discovery date.

4 D.

Inoperable Structures, Components, or Systems that Contributed to the Event:

None.

E.

Dates and' Approximate' Times ~for Major Occurrences:.

1.

October 21, 1991:

An evaluation of. CVCS-8440 was initiated.

2.

April 1,1992:

Event / Discovery :hte.. The TRG determined that j

periodic. inservice: testing of CVCS-8440 should

'have been performed,-asfrequired by TS 4.0.5.:

F.

Other Systems or Secondary-Functions Affected: +

_i None.

G.

Method of Discovery:

The TS violation was discovered by utility-(System Engineering).

personnel. During a review of an'INPO Nuclear Network: Operating-Experience entry. System Engineering initiated an investigation to determine any-safety functions performed'by CVCS-8440 in the:

reverse-flow direction.

H.

Operator Actions:

None.

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Safety System Responses:

fi None..

III.

Cause of the Event A.

Isumediate Cause:

I I

ASME Section XI inservice testing was not.being' performed.on check valve CVCS-8440 to verify that' it would perform its' safety function..

I i

B.

Root Cause:

i The root cause for the exclusion of CVCS-8440 from the IST program was t

that neither the industry, through operating' experience, nor.

Westinghouse, in its. original _desion,-had previously:iden_tified the safety function'of preventing: post-LOCA leakage.via this scenario.

Because LCV-112B'and LCV-112C~ upstream of-CVCS-8440 were considered to be the flowpath boundary valves,. theLfunction of CVCS-8440 was thought-only to be isolation fros' another closed piping' system. :The' potential leakpath discus?ed'in this LER is,a subtle one,' discovered.as a-result

~

-)

of increasing industry expertise regarding potential-post-LOCA recirculation leakpaths IV.

Analysis of the Event Review of CVCS-8440. indicates that'it would perform a safety-related

~-

i function during the post-LOCA recirculation' phase. The valve. is 'necessar/

to provide a barrier to isolate the VCT from potentially radioactive post-LOCA recirculation flow.. Westinghouse has verified ~the valve's design basis.

Thefollowingoperatingandmaintenancehistorysummahindicatesthat CVCS-8440 (on both Units 1 and 2) is operable and. capable of performing.

its intended safety function.

1.

The Unit 1 CVCS-8440 valve was inspected duringLthe'UnitT1 fourth refueling outage in' March 1991. Thwre was no' valve internal:: damage l

indicated and:a seating surface blue check was(satisfactory without lapping the' seating. surfaces. The Unit 2 CVCS-G440 valve'was-similarly inspected during the Unit 2 second refueling ^ outage in'.

October 1988, with similar results.

There are 16 valves of this same make-and;model installed: at DCPP. :

2. -

Four of these'are in' safety-related: applications, including theLtwo-CVCS-8440 locations. Review of the. maintenance: history of these 16 1

valves shows one case of maintena.ce required;to restore seat' leaks

-)

n tightness. This maintenance was required when valve RCS-1-8028?inLtheL miscellaneous relief valve header was-found:to have rust on. its!

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s seating surfaces. This valve application is in a stagnant system with only periodic flow when a relief valve discharges.

~

3.

Results of a Nuclear Plant Reliability Data System'(NPRDS) search indicate a total of 10 valves of this model.in industryssafety-related.

applications. There were no industry reports of leakage problems on-j this valve model other than the one experienced:at DCPP.-

4.

Tha CVCS-8440 valve application is in high purity, controlled

}

chemistry, 110*F filtered water.- This is not a stagnant line as water-is flowing in this line for normal operation'at all times when the charging pumps are operating.

t Although.PG&E considers the valves to be' operable,.the potential. effects of valve impalment are discussed below. Total valve disc failure-or failure in the fully open. position-is'not considered to'be a credible failure mode, based on the maintenance history of the valve, inspection results, and the service environment; and-in accordance:with'the single failure criterion defined by the'NRC in SECY 77-439. Therefore, any ;

j potential valve impaiment'is considered to befin'the fom of backleakage. -

As a typical illustration of this point, Final-Safety Analysis Report Update Table 6.3-1 for Emergency Core Cooling System check' valves-indicates a backleakage design parameter of onlyf3 cc per hour per inch of nominal pipe size.

(CVCS-8440 is a 4-inch valvef and its required factory

~

seat leakage test was 10 cc per hour per inch.)

During-the post-LOCA recirculation phase,;backleakageipast CVCS-8440 would.

allow a leak path for potentially radioactive. containment s_ ump water-from the residual heat removal (RHR). pumps-(BP)(P)Lthroughithe seal water heat exchanger to relief valve RV-8123- (see Figure 1).: ; RV-8123 is set to.

relieve pressure at a nominal 150 psig.- Preliminary worst. case' estimates indi rte that the pressurt at RV-8123, without subtracting'line losses,-

could be as high as 180 psig plus containment pressure for a small break i

LOCA and 146 psig plus containment ~ pressure for a large break: LOCA..- This piping would be pressurized only during the recirculation phase'(i.e.,

when the RHR pumps are supplying containment ~ sump water to:the CCP suction), and the pressure would depend on the~ RHR pump discharge pressure. Therefore, in these worst case scenarios,.RV-8123 could lift and discharge sump water to the:VCT.

However, qualitative evaluations' of the expected cases indicateLthatlthe-containment should be sufficiently cooled and depressurized by the" time-recirculation is initiated,-such that RV-8123 would not be challenged.

In addition, fuel damage would not necessarily occur in-_the-'small_ break LOCA -

scenario that would be required to. maintain RHR pump-discharge pressure high enough to lift RV-8123 for an extended period. : Dose concerns.would not: approach limiting values unless fuel damage occurred and the'resulting increased radioactivity. leaked out'to the atmosphere.

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These qualitative evaluations; indicate'it;is unlikely;that2the combination of containment pressure,.RHR pump ^ discharge pressure,:CVCS-8440:

backleakage rate, and duration of pressure / leakage _ would be such thati backleakage past CVCS-8440_ would fill the free volume of the VCT.1;The amount of postulated backleakage ;is very small'(on the' order of.: tens' of cc/hr), as discussed'above. However,_even; assuming a backleakage rate of up to 0.3 gpa (which Westinghouse considers to be the maximum credible!

leakage rate for-a functioning check valve of this' size),;it would take:

appiv/.imately 25_ hours to fill the free volume of the VCT after reaching; its high alarm setpoint. PG&E's qualitativ'e evaluation.of aflimited range; of LOCA break sizes indicates that the plant.would be cooled down and?

depressurized to cold shutdown within 25~ hours. following a _LOCA. -

Therefore, RHR pump discharge pressure would.not. remain;high.enough to lift RV-8123 long enough to fill.the VCT, even assuming no operator actions to stop the leakage.-

Although potentially radioactive sump water has? not previously been anticipated in the V M or vicinity, no radioactive releases would occur to a

the auxiliary building atmosphere-from the VCT -since'it would haveL.-..

sufficient capacitylo contain the postulated leakage' flow 1until the lift pressure of RV-8123 was;no longer exceeded. -In addition, the potential for 1oss of sump recirculation inventory is~ bounded by.~a previous evaluation.

Thus, the failure to perform periodic: inservice-testing.on'CVCS-8440 did not adversely affect the health and safety of the public.

V.

Corrective Actions A.

Immediate Corrective Actions:

• ~ 1.

Operability =EvaluationOE92-02was: performed,whichdavidesthe bases for considering CVCS-8440 operable and capable of performing its safety function in the event of a"LOCA.

B.

Corrective Actions to Prevent Recurrence:-

1.

A valve test procedure will be developed for CVCS-8440f and ~added to the IST program, to. test for potential post-LOCA recirculation:

gross leakage (i.e., to test that the valve will closeiand prevent reverse-flow), with initial' perfonsance scheduled for -

IR5/2R5.

2.

The' check valve safety function discussed in this LER had not~-

previously been identified by Westinghouse,;PG&E,;or'the -

industry. DCM S-8'will be revised to incorporate the information.

resulting from the current investigations..

1021S/85K

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All valves in the post-LOCA recirculation flow path will be-reviewed to determine if any other valves perform a previously-unidentified safety function, due to a.similar configuration (i.e., a configuration where potential leakage is thought to be contained by a closed piping system).

]

VI.

Additional Information A.

Failed Components:

None.

B.

Previous Similar. Events:

)

1.

LER l-84-044-01, " Check' Valve Back flow Inservice Testing i

Deficiencies" This LER reported incomplete testing of-certain valves in the IST program. The root cause was personnel error during the original preparation of the IST program; however, the LER did not concern valves excluded from the IST' program. Because the corrective actions addressed proper testing for valves already in the IST-program, these corrective actions did not prevent the current LER.

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