ML20205F596

Forwards Comments on Preliminary Accident Sequence Percursor (ASP) Analysis of 980624 Operational Event at Dbnps,Unit 1, as Transmitted by NRC Ltr

ML20205F596
Person / Time
Site:	Davis Besse
Issue date:	03/27/1999
From:	Campbell G CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2593, NUDOCS 9904060325
Download: ML20205F596 (2)

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0 ta e R Oak Harbor, Ohio 43449-9760 Guy G. Campbell . 419-3214S88 '

%:e President- Nuclear Fax:419-321 8337 Docket Number 50-346 License NumberNPF-3 Serial Number 2593 March 27, 1999 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Comments on Preliminary Accident Sequence Precursor Analysis of June 24,1998 Operational Event at Davis-Besse Nuclear Power Station, Unit Number i Ladies and Gentlemen:

l l

Attached please find the FirstEnergy Nuclear Operating Company's (FENOC) comments on the  !

preliminary accident sequence precursor (ASP) analysis of the June 24,1998, operational event at Davis-Besse Nuclear Power Station, Unit Number i as was transmitted by Ncclear Regulatory Commission (NRC) letter (Log Number 5422) dated February 18,1999. On March 24,1999, the FENOC staff briefed the NRC Davis-Besse Project Manager by telephone on the status ofits review and its planned submittal of comments by March 31,1999. The comments provided herein should be reflected in the analysis discussion and cut sets to ensure that the correct characterization of the plant design and the event are portrayed.

'l If you have any questions, or require further information, please contact Mr. James L. Freels, Manager - Regulatory Affairs at (419) 321-8466.

Very truly yours, L

RMC/dle Attachment cc: J. E. Dyer, Regional Administrator, NRC Region Ill S. J. Campbell, DB-1 NRC Senior Resident Inspector (Acting)

A. G. Hansen, DB-1 NRC/NRR Project Manager Utility Radiological Safety Board 9904060325 990327 7 PDR ADOCK 05000346 S

PDR3

o Docket Number 50-346 r s . ,' _. License Number NFP-3 f

Serial Number 2593 Attachment Page1 af1 i

COMMENTS ON PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF JUNE 24,1998 EVENT AT DAVIS-BESSE NUCLEAR POWER STATION, UNIT N' UMBER 1 The following comments are provided to ensure the accident sequence precursor (ASP) analysis is consistent with the Davis-Besse Nuclear Power Station (DBNPS), Unit I design and Individual Plant Evaluation (IPE).

1. Page 2 of the preliminary ASP analysis, Event Description, states that the emergency diesel generators (EDGs) are qualified to 120 F. This is not correct in that the only design characteristic that relates to this teriperature value is the EDG Room ventilation system that is sized to maintain each " operating" EDG room at 120 F assuming at 95 F outside air (Reference US AR 9.4.2.1.2.3).- The elevated temperatures will only affect the 40-year life of the EDG in terms of days (based on continuous operation during this period). There fore, although the EDG was declared inoperable per plant procedures, :t was in fact available and continued to provide essential electric power during the event. l

2. Page 6 of the preliminary ASP analysis, third bullet of Failure of EDG 1 room ventilation l recirculation damner, states the maximum room temperature reached was 122 F. The maximum  ;

temperature recorded for the EDG 1 room, as cited in Potential Condition Adverse to Quality Report (PCAQR) 98-1294, was 125 F at 1135 hours0.0131 days <br />0.315 hours <br />0.00188 weeks <br />4.318675e-4 months <br /> on June 25,1998.

3. Page 6 of the preliminary ASP analysis, third bullet of Failure of EDG 1 room ventilation recirculation damoer. also reflects that a 120 F room temperature is an " operability limit," which is incorrect as it relates to qualification of the EDGs. Per comment 1 above, this is a procedural operability limit and a design parameter for the ventilation system. l

4. Page 6 of the preliminary ASP analysis, third bullet of Failure of EDG 1 room ventilation recirculation damner, states that the most limiting temperature of components in the room was 132 F. Evaluation of control cabinet components for PCAQR 98-1294 indicated that the EDG differential relays were the most limiting components, being certified for continuous operation at 55 C (131"F).

5. The Analysis Results on page 7 of the ASP analysis assumed that a reactor coolant pump (RCP) real Loss of Coolant Accident (LOCA) could occur despite the loss of the RCPs due to failure of offsite power. This is not consistent with the seal failure model used in the DBNPS Probabilistic Safety Assessment (PSA). The seal failure model used by DBNPS is described in detail in the DBNPS IPE, Part 3, Section 4.4.3 submitted by Serial Number 2119, dated February 26,1993.

It was concluded, based on testing and the design of the Byron Jackson RCP seal, that the seals will not experience gross leakage due to loss of support systems as long as operators take appropriate actions to trip the RCP (Reference DB-OP-02523, Component Cooling Water System Malfunctions). The DBNPS PSA model incorporates this conclusion by assuming a RCP seal LOCA only if plant operators fail to trip the affected RCP. This approach is consistent with other plants using Byron Jackson pumps and the same model/ design seals. For the June 24, 1998, event, the RCPs tripped upon loss of offsite power; therefore, a RCP seal LOCA should not be assumed anytime offsite power is not avaLble.