ML121000281

Request for Additional Information, Round 2, License Amendment Request, Revise Technical Specification 5.6.5, Core Operating Limits Report (Colr), to Replace Large-break Loss-of-Coolant Accident Analysis Methodology

ML121000281
Person / Time
Site:	Wolf Creek
Issue date:	04/05/2012
From:	Hall J Plant Licensing Branch IV
To:	Wideman S Wolf Creek
Hall, J R, NRR/DORL/LPL4, 301-415-4032
Shared Package
ML121000271	List: ML121000281
References
TAC ME4996
Download: ML121000281 (2)

Text

REQUEST FOR ADDITIONAL INFORMATION WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 LICENSE AMENDMENT REQUEST RELATING TO THE ADDITION OF A NEW ANALYTICAL METHODOLOGY FOR THE BEST-ESTIMATE LARGE BREAK LOSS-OF-COOLANT ACCIDENT TAC NUMBER ME4996 By letter dated November 4, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML1032002090), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee), submitted a License Amendment Request (LAR) for a change to the Technical Specifications (TSs) for the Wolf Creek Generating Station (WCGS). Specifically, WCNOC requested a revision to TS 5.6.5, Core Operating Limits Report (COLR), to replace the existing best estimate large break loss-of-coolant accident (LOCA) analysis methodology with a methodology based on the NRC-approved topical report WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM). WCNOC provided supplemental information in letters dated October 19, 2011, and January 31, 2012.

The NRC staff has reviewed the subject LAR and the supplemental information and has determined that the additional information requested below is needed to complete its review.

This information is necessary to enable the staff to determine whether the approved methods of WCAP-16009-P-A can be acceptably applied to the plant-specific LOCA analyses for WCGS, and whether the proposed changes comply with the requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

NRC Information Notice (IN) 2011-21, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation, describes an error in the Westinghouse ASTRUM emergency core cooling system evaluation model due to the discrepancy between PAD 4.0-predicted fuel initial conditions and actual conditions based on Halden Research Reactor data. The error has been demonstrated to affect the predicted peak cladding temperature significantly, as described in IN 2011-21 and in additional evaluations performed subsequent to the issuance of IN 2011-21.

Please discuss how the evaluation proposed for implementation at Wolf Creek addresses this error. If it does not, please estimate the effect of the error and its correction on the Wolf Creek-predicted peak cladding temperature. If the effects of the error and its correction are significant, as described in 10 CFR 50.46(a)(3), discuss what steps will be taken to re-analyze the Wolf Creek emergency core cooling performance. If the effects of the error and its associated

2 correction cause the results of the Wolf Creek evaluation to exceed 10 CFR 50.46(b) acceptance criteria, describe what actions will be taken to ensure that emergency core cooling performance remains acceptable as set forth in 10 CFR 50.46.