ML20056G417

Topical Rept Evaluation of Rev 4 to OPPD-NA-8303, Transient & Accident Methods & Verification. Proposed Changes in Rev 4 Acceptable Except for Use of Cents Computer Code for Transient Analyses

ML20056G417
Person / Time
Site:	Fort Calhoun
Issue date:	08/18/1993
From:	Office of Nuclear Reactor Regulation
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ML20056G411	List: ML20056G410 ML20056G417
References
NUDOCS 9309030099
Download: ML20056G417 (2)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO l

OPPD-NA-8303 REVISION 4 TRANSIENT AND ACCIDENT METHODS AND VERIFICATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT 1  :

DOCKET NO. 50-285 1.0 liiTRODUCTION By letter dated February 5,1993, Omaha Public Power District (0 PPD) submitted proposed changes to Topical Report OPPD-NA-8303, " Reload Core Analysis Methodology, Transient and Accident Methods and Verification". The report i desc.ribes OPPD's reload core transient and accident methods for application to Fort Calhoun Station Unit 1. The proposed changes would be incorporated as ,

Revision 4. l 2.0 EVALUATION The large and small break loss-of-coolant accident (LOCA) are analyzed for OPPD by Westinghouse (Ref.1). OPPD confirms the assumptions used in these .

analyses are valid for each reload core and, if reanalysis is required, it is 1 to be performed by Westinghouse. The applicability of the Westinghouse LOCA evaluation models for Fort Calhoun has been approved by the NRC (Ref. 2). i However, use of the CENTS code for transient analysis has not been approved at this time. The reference to the CETOP Code for calculating the minimum departure from nucleate boiling ratio (DNBR) in the control element assembly (CEA) withdrawal event has been changed to TORC since the TORC Code will now be used for calculating the required overpower margins. This is acceptable since both codes have been approved by the NRC for calculating fuel rod DNBR.

Revision 4 incorporates the OPPD steam generator tube rupture methodology. ,

This methodology was approved by the NRC in a safety evaluation dated May 22, ,

1991 (Ref. 3). l Various other minor revisions have been made to reflect current approved methodology, to remove cycle specific results, to make editorial corrections,  !

and to update references to the most recent revisions. These various changes have been found to be acceptable.

9309030099 930818 i PDR ADOCK 05000285 P PDR_

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j1 3.0 CONCly1193 The staff has reviewed the proposed changes in Revision 4 to OPPD-NA-8303 and finds them acceptable. Use of the CENTS computer code by OPPD for transient l analyses has not been approved at this time. The applicability of the Westinghouse LOCA evaluation models for Fort Calhoun and the steam generator tube rupture analysis methodology have been previously approved.

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4.0 REFERENCES

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1. " Westinghouse ECCS Evaluation Model for Analysis of CE-NSSS,"

WCAP-13027- P, July 1991.

l 2. Letter from D. L. Wigginton (NRC) to W. G. Gates (OPPD), Loss of Coolant Accident Analyses for Fort Calhoun Station, Unit 1, (TAC No. M81831),

March 26, 1992.

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3. Letter from W. C. Walker (NRC) to W. G. Gates (0 PPD), Steam Generator Tube Rupture Methodology, (TAC No. 66801), May 22,1991.

Principal Contributor: L. Kopp, DSSA Date: August 18, 1993