ML20056F343

Provides Summary of Plant IPE Submitted on 930630,per GL 88-20, Individual Plant Exam for Severe Accident Vulnerabilities

ML20056F343
Person / Time
Site:	Point Beach
Issue date:	08/10/1993
From:	Hansen A Office of Nuclear Reactor Regulation
To:	Hannon J Office of Nuclear Reactor Regulation
References
GL-88-20, TAC-M74452, TAC-M74453, NUDOCS 9308270012
Download: ML20056F343 (2)

Text

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5 E UNITED STATES k i' ^! NUCLEAR REGULATORY COMMISSION g-***** f

/ WASHINGTON, D.C. 20555-0001 August 10, 1993 Docket Nos. 50-266  :

and 50-301 I

MEMORANDUM FOR: John N. Hannon, Director  !

Project Directorate III-3 1 Division of Reactor Projects III/IV/V  ;

Office of Nuclear Reactor Regulation l FROM: Allen G. Hansen, Project Manager i Project Directorate III-3 1 Division of Reactor Projects III/IV/V Office o. Nuclear Reactor Regulation

SUMMARY

OF POINT BEACH NUCLEAR PLANT IPE (TAC M74452 and M74453)

The Wisconsin Electric Power Company (WEPCo) submitted its Individual Plant Examination (IPE) report for the Point Beach Nuclear Plant (PBNP) on June 30, 1993, in accordance with Generic Letter 88-20, " Individual Plant Examination for Severe Accident Vulnerabilities." The Probabilistic Safety Assessment (PSA) methodology, as identified in NUREG/CR-2300, was utilized to perform the l IPE. The IPE consists of Level 1 PSA and a limited-scope Level 2 PSA of PBNP including an internal flood analysis to identify any severe accident vulnerabilities from internally-initiated events.

l The core damage frequency (CDF) including internal flooding calculated by the Point Beach Level 1 PSA is 1.15E-4/ year for each PBNP unit. Based on certain plant improvements made since the PSA model freeze date of September 5,1990, the current estimated CDF is about 1.0E-4/ year / unit. The largest initiator contributors to CDF are a large loss of coolant accident (LOCA, 22%), followed by station blackout (SB0,13%), transients without power conversion systems such as main feedwater or steam dumps (10%), internal flooding (9%) and medium LOCA (9%). The largest single contributors to CDF involve failure to manually align the emergency core cooling system (ECCS) for containment sump recircu-lation (particularly for large LOCA initiators) and failure to manually align alternate sources of water to the auxiliary feedwater (AFW) pumps upon depletion of the water in the condensate storage tanks (CSTs).

A conservative fission product release frequency (FPRF) is calculated by the Level 2 PSA to be 2.43E-5/ year for each PBNP unit. The Point Beach Level 2 PSA conservatively assumes that any core damage results in complete core melt and failure of the reactor vessel lower head, does not credit any operator actions or equipment recoveries following onset of core damage, and assumes that any core damage sequence for which containment pressure approached the best estimate failure pressure or is still trending upward at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> fails containment. The largest initiator contributors to FPRF are' transients and-SB0 (total of 74%), which result in containment failure due to overpressure at g' roughly 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. However, steam generator tube rupture core damage sequences l (26% of the FPRF) are also significant because they bypass containment.

9308270012 930810 @h l PDR ADOCK 05000266 P

l MC RE CENTER COPY

- John N. Hannon August 10, 1993-Containment isolation failures and interfacing system LOCA sequences, which also bypass the containment, are very small contributors (about 0.5%) to the PBNP FPRF.

Several procedural revisions, plant modifications, and implementation of a severe accident management-program are planned for PBNP to address insights from the IPE and to further reduce the probability of core damage and fission )

product release. '

The planned procedural revisions are:

1. improve the manual alignment of the ECCS to the containment sump recirculation mode in the event of a LOCA; and

2. improve the manual alignment of alternate sources of water to the suction of the AFW pumps upon deplei.icn of the CST inventory.

The planned PBNP Modifications include:

1. facilitating rapid alignment of fire water'to refill the CSTs;

2. reversing certain equipment access doors to address a flooding ,

concern in the AFW pump room; and l

3. installation of a 3rd and 4th emergency diesel generator (initiated for reasons other than IPE).

I Severe Accident Management (SAM). guidelines, which will be tailored to PBNP 1 based on the IPE results, are planned to be implemented as part of a PBNP SAM j program.

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Together, all of these improvements are expected to reduce the CDF to roughly )

8E-5/ year per unit and the FPRF by roughly an order of magnitude to about ,

3E-6/ year per unit by the end of 1995. l ORIGINAL SIGNED BY Allen G. Hansen, Project Manager l Project Directorate III-3 Division of Reactor Projects III/IV/V DISTRIBUTION:

Docket Files OGC NRC & Local PDRs EJordan (MNBB3701)

PDIll-3 Reading ACRS(10) s TMurley/FMiraglia (12-G-18) AGody,Jr., EDO (17-G-21) )

JPartlow (12-G-18)

WBeckner (10-E-4)

JRoe JSchiffgens (10-E-4)

JZwolinski AThadani (12-G-18)

JHannon JWiggins (12-G-18)

MRushbrook RBarrett (8-H-7) l AHansen RHernan (14-C-7) l JKing EGreenman, DRP. j

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0FFICE PD3-3:LA PD3-3: INTERN PD3-3:PM m. - PD3-3k NAMF MRhihrook JFKing/bj/sw/k AHanse M JHannon  !

DATE 1/S/93 O

~7 / %/93 [//o/93 g ///'/93 0FFICIAL RECORD i DOCUMENT NAME: g:\ptbeach\ptbchipe.mts l

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