ML102810285

Request for Additional Information Regarding Davis-Besse Notification of Significant Change to the small-break loss-of-coolant Accident Emergency Core Cooling Model in Accordance with 10 CFR 50.46 (a)(3)

ML102810285
Person / Time
Site:	Davis Besse
Issue date:	10/19/2010
From:	Michael Mahoney Plant Licensing Branch III
To:	Allen B FirstEnergy Nuclear Operating Co
Mahoney, M NRR/DORL/LPLIII- 2 415-3867
References
TAC ME4780
Download: ML102810285 (3)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 October 19, 2010 Mr. Barry S. Allen Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING DAVIS-BESSE NOTIFICATION OF SIGNIFICANT CHANGE TO THE SMALL-BREAK LOSS-OF-COOLANT ACCIDENT EMERGENCY CORE COOLING MODEL IN ACCORDANCE WITH 10 CFR 50.46 (a)(3) (TAC NO. ME4780)

Dear Mr. Allen:

By letter dated September 2, 2010 (Agencywide Documents Access and Management System Accession No. ML102530281), FirstEnergy Nuclear Operating Company (FENOC), submitted a notification of a significant change to the small-break loss-of-coolant accident emergency core cooling model in accordance with Title 10 of the Code of Federal Regulations, Section 50.45(a)(3) for Davis Besse Nuclear Power Station, Unit 1.

The Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on October 7, 2010, it was agreed that FENOC would provide a response within 60 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3867.

S7!TJJI)

Michael Mahoney, Project Plant licensing Branch 111-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosure:

Request for Additional Information cc w/encl: Distribution via listserv

REQUEST FOR ADDITIONAL INFORMATION DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-346 By letter dated September 2, 2010 (Agencywide Documents Access and Management System Accession Number ML102530281), FirstEnergy Nuclear Operating Company (FENOC), the licensee for Davis-Besse Nuclear Power Station, Unit 1, sent a notice reporting a change or error discovered in an evaluation model or in the application of such a model that affects the peak cladding temperature (PCT) calculation. This report was submitted pursuant to the requirements of Title 10 of the Code of Federal Regulations, Section 50.45(a)(3), which requires, in part, that licensees report a change in the evaluation model used resulting in a significant change in PCT (greater than 50 degrees Fahrenheit (OF)). The intent of this requirement is to enable the staff to establish the safety significance of this change (Federal Register, Volume 53, No. 180, pp. 35996-36005).

Because of the magnitude of the reported change, 225°F, and because the means by which this value was determined is not clear to the staff, the U.S. Nuclear Regulatory Staff does not have sufficient information to establish the safety significance of this change to the evaluation model.

Requested Information Please provide additional information regarding FENOC's evaluation of the impact of this peak cladding temperature modeling error. This evaluation should include a discussion of the causes of the error and evidence to support a conclusion that the model as a whole remains adequate to predict PCT. Please include a discussion of the impact of this model error on the full spectrum of postulated break sizes as well as FENOC's planned corrective actions and actions to prevent reoccurrence. If a plant-specific assessment regarding the modeling errors was not performed, justify the use of any generic evaluation.

MI-102810285 I\IRR-088 *Via E-mail OFFICE LPL3-2/PM LPL3-2/LA LPL3-2/BC LPL3-2/PM NAME MMahoney THarris MDavid for RCarlson MMahoney DATE 10/08/10 10/14/10* 10/19/10 10/19/10