ML17347A800

Forwards Evaluation of Safety Issues Re Regulatory Analysis for plant-specific Emergency Feedwater Sys Backfit

ML17347A800
Person / Time
Site:	Oconee, Turkey Point, 05000000
Issue date:	03/06/1986
From:	Congel F Office of Nuclear Reactor Regulation
To:	Parr O Office of Nuclear Reactor Regulation
Shared Package
ML17347A796	List: ML17347A795 ML17347A797 ML17347A798 ML17347A800
References
NUDOCS 8807050491
Download: ML17347A800 (13)

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MEMORANDUM FOR:

0. Parr, Chief Plant Electrical, Instrumentatfon and Control Systems

Branch, DPWRL-B FROM:

SUBJECT:

Frank J.

Congel, Chfef Reliability 5 Risk Assessment

Branch, DSRO RRAB INPUT TO OPMRL-B's REGULATORY DECISION ON EFM SYSTEM BACKFIT AT OCONEE'ITE

REFERENCE:

Memorandum from D. M. Crutchffeld, OPAL-B to B. Sheron,

DSRO, dated January 7,

1986 Regarding Regulatory Analysis for Plant Specific Backfit of Oconee The above reference identified the need for a regulatory analysis of the emergency feed-water (EFM) system at Oconee nuclear units.

Enclosed fs our evaluation of the Safety Issue discussed fn the above reference.

Our preliminary evaluation indicates that the upper bound estimate of the cost of any proposed EFM System ffx should not exceed

$600,000, based on a

$1000/person-rem criterfa.

Me suggest that any proposed ffx of the EFM system could consider the plant improvements whfch have been previously implemented by the licensee, Duke Power Company.

If you need further information regarding rfsk perspective.

E. Chellfah could be contacted at x28338.

Enclosure:

As stated

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T. Spefs F. Mfraglia F. Schrocdcr B. Sheron D. Crutchffcld K. Knfel J. Stolz H. Nicholas M. Rubin J. Nermfel Frank J. Congel, Chfcf Rclfabilfty 4 Risk Assessment Branch Division of Safety Revfew and Oversfght 8807050491 880616 PDR ADQCK 05000250 PDR l

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ENCLOSURE RRAB INPUT TO DPMRL-B'S REGULATORY DECISION ON EFM SYSTEM BACKFIT AT OCONEE SITE

REFERENCE:

Memorandum from D. M. Crutchfilcd, OPAL-B to B. Sheron,

DSRO, dated January 7, 1986 Regarding Regulatory Analysis for Plant Specific Backfit of Oconee The above reference fdentfffcd the need for a regulatory analysfs of the emergency feed-water (EFM) system at the Oconce nuclear units.

Basically, staff's review of the licensee, Duke Power Company (DPC), response to generic letter 81-14 concerning seismic qualification of the EFM system indicates that a safe shutdown earthquake (SSE) of 0.1g magnitude could cause a pfpe rupture of thc non-scfsmfc condcnscr cfrculatfon'ater (CN) system; the resulting flood in the turbine building could result fn the loss of all three trains of the EFM System.

When the above adverse systems interaction fs compounded by a postulated single active failure fn the standby shutdown facility (SSF) system, the decay heat removal capability through the stc'am generators could bc lost.

The staff believes that DPC should correct thc above postulated scenario.

Because thc original licensing basis for Oconec nuclear units dfd not require wgu3atory compliance with general design crfter fa NC-2 and GDC-34 for thc EFM system. staff considers the above postulated safety issue resolution a backfit for Oconee.

Thus. thc above reference requested DSRO staff to perform an initial cost/benefit analysis fn order to establish an upper bound on cost of chc desired backfit that could be justified as a realfstfc safety improvement

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at the pconec site.

The purpose of this memorandum is to a provide risk perspective regardfng seismic core damage sequences fncludfng the postulated accident scenarfo indicated fn the above reference and to provide an upper bound estimate of the dollar worth of safety improvement that could be realfzed from the seismic safety issue discussed in the above reference.

Our preliminary review of Oconce Probabilistic Risk Assessment

{PRA) indicates that the total mean core damage frequency estimate fs about 3E-4 per reactor year.

The core damage frequency fs dominated by various internal flooding events

{not seismically induced) at the turbfne bufldfng.

The frequency contribution of sefsmfc events of varfous magnftudes ranging from 0.01g to 0.6g to the total core damage fs about 454 per reactor year.

The above estimate has a very large uncertainty (a frequency range of 1.0E-7 to 1.0E-4 pcr reactor year).

The Oconee PRA has identified three dominant structural 4

failure modes which contrfbute to the total seismic sequence frequency of 6E-5 per reactor year.

These failure modes are:

(I) the collapse of auxfliary building (AB) internal aassonry walls causing loss of the high.pressure injection [HPI] and EFM system and system canponcnts; (2) collapse af the blockwalls fn the blockhouse through which all 4.16KV emergency buses pass; and

{3) ruoturc of the condenser and failure of CN pipes cannected to the condenser which cause a large flood fn the TB, resultfng fn the loss of three EFM trains and HPI pumps fn AB.

The median fragility estimate of the condenser and connected CCM piping fs about 0.21'fth a considerable uncertainty bound

~hich is abaut Br Bv 0.14.

Thus. a postulated SSE could cause CN pipe failure which could in turn cause a large flood resulting fn failure of three trains of the EFM System.

The postulated SSE. fn combfnatfon with a sfngle failure in

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,0 SSF System or a failure of stuck open safety relief va1ve in the reactor primary system could cause core damage.

The mean frequency of the above core damage sequence is estimated to be less than 2K-S per reactor year.

The Oconee containment is a 1arge dry containment.

The Oconee PRA predicts that the seismic capacity of the Oconee containment is about 2g (Median} with an uncertainty bound about Br 0.26 and Bv 0.29.

Because of the larger capacity of the containment, we believe that a seismic event of SSE magnitude will not cause a containment gross failure.

However, postulated core damage due to the TB flooding sequence could cause a delayed contaiment failure due to late overpressure buildup inside the containment.

The overpressure buildup could be due to the production of non-condensible gases with no containment heat t

removal at about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the onset of core damage.

Thus the release estimates from the seismically induced flooding sequence are estimated to be similar to the release estimates of a release category 4 as defined in the Oconee PRA.

?f a conditional probability of 0.9 for the containment overpressure failure mode and a conditional whole body dose with a maximum estimate of 1.0E+6 per event are assumed.

then the risk estimate of the postulated SSK induced flooding sequence could not be sore than about 1f person rem per reactor year.

There could also be about 0.1 probability that the containment may not fa)l with core damage, but release could occur through containment leakage.

The conditional release through a leakage type failure mode ts similar to a release category 5 as defined in the Oconee PRA, which results in an estimated conditional whole body dose of not more than 1.0 E+4 person rem per event.

The risk estimate due to the leakage failure mode is

hen less than one person-rem per reactor year.

Therefore, the total risk

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estimate of the SSE induced flooding sequence should not to exceed about 18 person-rem per reactor year. If the Ccemfssfon's proposed safety goal of $ 1000 per person-rem for the benefit/cost guidance fs applied and a reactor life of 30 years is assumed, then the postulated SSE induced flooding sequence could have a maximum risk worth of about

$540,000.

It is very important to note that thc above rfsk worth estimate of $540,000 is more likely an upperbound estimate.

This is primarily because of the use of the Reactor Safety Study (RSS) type source term methods in estimating Oconee releases.

The staff believes that the use of RSS-type source term methods could yield a somewhat hfgher risk estimate for sequences fnvolvfng late containment failures.

Also, the frequency estimate of an SSE induced flooding sequence fs a lfttle hfghcr than the realfstfc estfwiate.

The higher estimate of sequence e

frequency fs prfmarfly due to the fnclusfon of CN structural faflure contribution beyond the sefsmfc intensity of 0.2g and up to 0.6g.

However, the failure contributor beyond 0.2g fs not expected to be signfffcantly higher.

At this time, we do not know the nature of EFM Systeo and/or CCM System backfft and therefore, do not know the potentfal rfsk reduction of the backfft.

However, we suggest that the NaxfNNN risk worth estimate of $540,000 due to thc SSE induced flood sequence could be considered as an additional basis that the staff can reassess the nccd for any kind of backfft to thc EFM System at Oconee.

It fs also very important to note that DPC has already implemented many plant ffxes to mftfgate the TB flood and fts fapact on safety systems including EFM System pumps: therefore DPC has reduced the overall flood sequence frequency substantially.

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I ENCLOSURE 2

MEETING WITH FPSL ON 6/2/88 TO DISCUSS SEISMIC ADEQUACY (GENERIC LETTER 87-02)

NAt<E G.

E. Edison L.

8 ~ Marsh G. G~ise-Koch G.

Bagchi R.

Luke J.

Chandra P.

Pace M. Dryden R. Walpole W. Shields R. Aigstei 1 H. Berkow J. Richardson P.

T.

Kuo Pei-Ying Chen A. E. Friend J.

D. Stevenson J.

K. Hays J.

N. Burford ORGANIZATION HRC/Licensing NRC/EMEB t~RC/t(RR NRC/tIRR FPSL EBASSO FPSL - Licensing FPSL - Licensing EBASSO/Licensing NUBARG EBASSO/Envisosphere NRC/NRR NRC/NRR NRC/EMEB NRC/NRR/Et1EB FPSL/Licensing SEA/FPSL FPSL/0:Nuclear Licensing FPSL/Nuclear Licensing

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APPENDIX B