ML13345B335

Request for Additional Information, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternative Source Term

ML13345B335
Person / Time
Site:	Wolf Creek
Issue date:	12/13/2013
From:	Lyon C Plant Licensing Branch IV
To:	Matthew Sunseri Wolf Creek
Lyon C
References
TAC MF2574
Download: ML13345B335 (4)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 13, 2013 Mr. Matthew W. Sunseri President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION- REQUEST FOR ADDITIONAL INFORMATION RE: TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSIS (TAC NO. MF2574)

Dear Mr. Sunseri:

By application dated August 13, 2013, to the U.S. Nuclear Regulatory Commission (NRC)

(Agencywide Documents Access and Management System (ADAMS) package Accession No. ML13247A075), Wolf Creek Nuclear Operating Corporation (the licensee) requested a license amendment for Wolf Creek Generating Station, to revise the Technical Specifications to support transition to the Westinghouse core design and safety analysis.

The NRC staff has reviewed the information provided in your application and determined that additional information is required in order to complete its review. The enclosed questions were provided to Mr. S. Wideman of your staff on December 11, 2013. Please provide a response to the questions by January 31, 2014.

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 NRC staff resources. If circumstances result in the need to revise the requested response date, please contact me at 301-415-2296 or via e-mail at Fred. Lyon@nrc.gov.

Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSIS WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 By application dated August 13, 2013, to the U.S. Nuclear Regulatory Commission (NRC)

(Agencywide Documents Access and Management System (ADAMS) package Accession No. ML13247A075), Wolf Creek Nuclear Operating Corporation (the licensee) requested a license amendment for Wolf Creek Generating Station, to revise the Technical Specifications to support transition to the Westinghouse core design and safety analysis.

The NRC staff has reviewed the information provided in the application and determined that additional information, as listed below, is required in order to complete its review.

EMCB-RAI-1 Implementation of an Alternate Source Term (AST) in accordance with Section 50.67, "Accident source term," of Title 10 of the Code of Federal Regulations (1 0 CFR) could affect structures, systems, and components (SSCs) which were not previously evaluated for the consequences of a design-basis accident (DBA) and, as such, may not be seismically qualified. Appendix A to 10 CFR Part 100, "Reactor Site Criteria," requires that SSCs necessary to assure the capability of the plant to mitigate the consequences of accidents, which could result in exposures comparable to the guideline exposures provided in 10 CFR Part 100, be designed to remain functional during and after a safe shutdown earthquake. In accordance with the 10 CFR Part 100 requirements, please identify SSCs which may be affected by the implementation of the proposed AST and address the following:

a) Indicate whether any non-safety-related SSCs are being credited in the proposed AST license amendment.

b) For any nonsafety-related SSCs credited in the AST, confirm that the SSCs have been seismically qualified in accordance with the plant licensing basis.

c) Indicate whether the SSCs are new or existing.

d) Describe the location of the SSCs and the seismic qualification method employed to demonstrate the seismic ruggedness of these SSCs, such as the plant licensing basis or an NRC-endorsed industry standard.

e) Summarize the results of the seismic qualification of the equipment, indicating whether any modifications or re-design will be necessary in support of the AST.

Enclosure

ESGB-RAI-1 Please identify all sources of post-loss-of-coolant accident (LOCA) strong acid generation in containment and time dependent values of strong acid concentrations in the sump for a period of 30 days post-LOCA.

ESGB-RAI-2 Please describe the ORIGEN-S computer code used to determine the pH in the sump water post-LOCA and the differences between it and ORIGEN-2/ARP. Please provide the input and output data of the program.

• ML133458335 *email dated **memo dated OFFICE NRRIDORULPL4-1/PM NRRIDORULPL4-1/LA NRRIDE/EMCB/BC NRR/DE/ESGB/BC NRRIDORULPL4-1/BC NRRIDORULPL4-1/PM NAME Flyon JBurkhardt AMcMurtray* GKulesa** MMarkley Flyon DATE 12/13/13 12/12/13 12/06/13 10/18/13 12/13/13 12/13/13