ML14027A162

Redacted, Request for Additional Information, Round 2, License Amendment Request to Approve Transition to Westinghouse Core Design and Safety Analysis and Adoption of Full Scope Alternate Source Term

ML14027A162
Person / Time
Site:	Wolf Creek
Issue date:	01/28/2014
From:	Lyon C Plant Licensing Branch IV
To:	Matthew Sunseri Wolf Creek
Lyon C
References
TAC MF2574
Download: ML14027A162 (8)

Text

OFFICIAL USE ONLY- PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 28, 2014 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) 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. Portions of the letter dated August 13, 2013, contain proprietary information and, therefore, those portions have been withheld from public disclosure.

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 January 17, 2014. Please provide a response to the questions by February 28, 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 NOTICE: Enclosure 1 to this letter contains Proprietary Information. Upon separation from , this letter is DECONTROLLED.

OFFICIAL USE ONLY- PROPRIETARY INFORMATION

OFFICIAL USE ONLY- PROPRIETARY INFORMATION M. Sunseri 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

Enclosures:

1. Request for Additional Information (proprietary)

2. Request for Additional Information (non-proprietary) cc w/Enclosure 2: Distribution via Listserv OFFICIAL USE ONLY- PROPRIETARY INFORMATION

ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION (NON-PROPRIETARY)

TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSIS WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482

OFFICIAL USE ONLY- PROPRIETARY INFORMATION 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

1.0 INTRODUCTION

By letter dated August 13, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13247A075), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) submitted a license amendment request (LAR) to revise Safety Limits (SLs) 2.1.1, "Reactor Core SLs," Technical Specification (TS) 3.3.1, "Reactor Trip System (RTS)

Instrumentation," TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation," TS 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation,"

TS 3.4.1, "RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits," TS 3.7.1, "Main Steam Safety Valves (MSSVs)," and Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to replace the existing WCNOC methodology for performing core design, non-loss-of-coolant-accident (non-LOCA) and LOCA safety analyses (for Post-LOCA Subcriticality and Cooling only) with standard Westinghouse developed and U.S. Nuclear Regulatory Commission (NRC)-approved analysis methodologies.

As part of the transition to the generic Westinghouse NRC-approved methodologies, the licensee performed instrumentation setpoint and control uncertainty calculations based on the current Westinghouse Setpoint Methodology. This amendment request also includes the adoption of Option A of Technical Specification Task Force (TSTF) TSTF-493-A, Revision 4, "Clarify Application of Setpoint Methodology for LSSS [Limiting Safety System Settings]

Functions."

2.0 REGULATORY REQUIREMENTS The NRC staff evaluated the LAR against the regulatory requirements and guidance listed below to ascertain whether there is reasonable assurance that the systems and components affected by the LAR will perform their required safety functions when called upon to do so.

2.1 Regulatory Requirements The NRC staff considered the following regulatory requirements:

Title 10 of the Code of Federal Regulations ( 10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," establishes the fundamental regulatory requirements.

Specifically, Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 provides, in part, that

... an application for a design certification, combined license, design approval, or manufacturing license, respectively, must include the principal design criteria for a proposed facility. The principal design criteria establish the minimum necessary design, fabrication, construction, testing, and performance OFFICIAL USE ONLY- PROPRIETARY INFORMATION

OFFICIAL USE ONLY- PROPRIETARY INFORMATION requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

General Design Criterion (GDC) 13, "Instrumentation and Control," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 requires that instrumentation be provided to monitor variables and systems and that controls be provided to maintain these variables and systems within prescribed operating ranges.

GDC 20, "Protection System Functions," of Appendix A to 10 CFR Part 50 requires that the protection system be designed to initiate the operation of appropriate systems to ensure that specified acceptable fuel design limits are not exceeded.

In 10 CFR 50.36, "Technical Specifications," the Commission established its regulatory requirements related to the contents of the TS. Specifically, 10 CFR 50.36 states that "each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section." Specifically, the regulations in 10 CFR 50.36(c)(1 )(ii)(A) state, in part, that Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

Additionally, 10 CFR 50.36(c)(3) states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The NRC staff reviewed the proposed LAR against these requirements to ensure that there is reasonable assurance that the systems affected by the proposed LAR will perform their required safety functions.

2.2 Regulatory Guidance Regulatory Guide (RG) 1.1 05, Revision 3, "Setpoints for Safety-Related Instrumentation,"

December 1999 (ADAMS Accession No. ML993560062), describes a method that the NRC staff finds acceptable for use in complying with the NRC's regulations for ensuring that setpoints for safety-related instrumentation are initially within, and will remain within, the TS limits. RG 1.105 endorses Part I of Instrument Society of America (ISA)-S67.04-1994, "Setpoints for Nuclear Safety Instrumentation," which is subject to NRC staff clarifications.

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OFFICIAL USE ONLY- PROPRIETARY INFORMATION In Regulatory Issue Summary (RIS) 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006 (ADAMS Accession No. ML051810077), the NRC addresses requirements on limiting safety system settings that are assessed during the periodic testing and calibration of instrumentation.

By letter dated September 7, 2005, from Patrick L. Hiland (NRC) to the Nuclear Energy Institute's Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values" (ADAMS Accession No. ML052500004), footnotes are described that should be added to surveillance requirements related to setpoint verification for instrument functions on which a safety limit has been placed. This letter also addresses the information that should be included within TSs to ensure operability of the instruments following surveillance tests related to instrument setpoints.

2.3 Supplemental Guidance The PWR [Pressurized-Water Reactor] and BWR [Boiling-Water Reactor] Owner's Groups' TSTF-493, Revision 4, dated January 5, 2011, and an errata sheet dated April 23, 2010 (ADAMS Accession No. ML100060064), addresses the NRC staff's concerns stated in RIS 2006-17, and the May 11, 2010, "Notice of Availability of the Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-493, Revision 4, 'Clarify Application of Setpoint Methodology for LSSS Functions"' (75 FR 26294), documents NRC's position on adoption of TSTF-493, Revision 4.

3.0 REQUEST FOR ADDITIONAL INFORMATION The NRC staff has reviewed the instrumentation and controls aspects of the licensee's LAR and concludes that additional information is needed to complete the review:

EICB-RAI-1 On page 56 of WCAP-17602-P, Revision 0, "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control, Protection, and Indication Systems," August 2013 (non-proprietary version available in ADAMS at Accession No. ML13247A079), for ((

)) Please provide the necessary information to demonstrate how these numbers have been established to comply with the 95/95 confidence level specified in RG 1.1 05. If they are based on plant drift analysis, then please provide sample calculations to demonstrate the validity of the data.

EICB-RAI-2 Please provide similar information for ((

)) Explain how THE same value for AL T and AFT will provide adequate detection of instrument system degradation during surveillance testing. Provide the criteria beyond which AL T and AFT are assigned different values.

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OFFICIAL USE ONLY- PROPRIETARY INFORMATION EICB-RAI-3 Please provide samples of plant procedures to demonstrate that adequate measures are being implemented to ensure compliance to TSTF-493-A, Revision 4, Option A, requirements.

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ML14027A053 (proprietary); ML14027A164 (non-proprietary); *memo dated 1/14/14 OFFICE NRRIDORULPL4-1/PM N RR/DORULPL4-1/LA NRR/DE/EICB/BC* NRRIDORULPL4-1/BC NRR/DORULPL4-1/PM NAME Flyon JBurkhardt JThorp MMarkley Flyon DATE 1/27/14 1/27/14 1/14/14 1/28/14 1/28/14