ML18270A094

Request for Additional Information License Amendment Request for Transition to Westinghouse Methodology for Selected Accident and Transient Analyses

ML18270A094
Person / Time
Site:	Wolf Creek
Issue date:	10/04/2018
From:	Balwant Singal Plant Licensing Branch IV
To:	Heflin A Wolf Creek
Wengert T, NRR, 301-415-4037
References
CAC MF9307, EPID L-2017-LLA-0211
Download: ML18270A094 (6)

Text

Mr. Adam C. Heflin UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 4, 2018 President and Chief Executive Officer, Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1-REQUEST FOR ADDITIONAL INFORMATION RE: LICENSE AMENDMENT REQUEST FOR TRANSITION TO WESTINGHOUSE METHODOLOGY FOR SELECTED ACCIDENT AND TRANSIENT ANALYSES (CAC NO. MF9307; EPID L-2017-LLA-0211)

Dear Mr. Heflin:

By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), as supplemented by letters dated March 22, May 4, July 13, October 18, and November 14, 2017; and January 15, January 29, April 19, June 19, and August 9, 2018 (ADAMS Accession Nos. ML17088A635, ML17130A915, ML17200C939, ML17297A478, ML17325A982, ML18024A477, ML180338024, ML18114A115, ML18177A198, and ML18232A058, respectively), Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a license amendment request to transition to Westinghouse methodology for selected accident and transient analyses for Wolf Creek Generating Station, Unit 1 (WCGS).

On September 25, 2018, the staff from WCNOC/Westinghouse and the U.S. Nuclear Regulatory Commission (NRC) participated in a conference call to clarify a draft request for additional information (RAI) that was sent to the licensee by e-mail on September 5, 2018. At the conclusion of the call, it was agreed that the NRC staff would revise the draft RAI to further clarify the request. The final RAI is enclosed with this letter. You are requested to respond to this RAI request within 30 days from the date of the letter.

If you have any questions, please contact me at 301-415-3016 or via e-mail at Balwant.Singal@nrc.gov.

Docket No. 50-482 Enclosure RAI cc: Listserv Sincerely,

'£:, ~:+ t-&, ~--~

Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST TO TRANSITION TO WESTINGHOUSE METHODOLOGY FOR SELECTED ACCIDENT AND TRANSIENT ANALYSES WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), as supplemented by letters dated March 22, May 4, July 13, October 18, and November 14, 2017; and January 15, January 29, April 19, June 19, and August 9, 2018 (ADAMS Accession Nos. ML17088A635, ML17130A915, ML17200C939, ML17297A478, ML17325A982, ML18024A477, ML18033B024, ML18114A115, ML18177A198, and ML18232A058, respectively), Wolf Creek Nuclear Operating Corporation submitted a license amendment request to transition to Westinghouse methodology for selected accident and transient analyses for Wolf Creek Generating Station, Unit 1 (WCGS). The U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is needed to complete its review, as described below.

Background

Attachment I, "Evaluation," and Enclosure I, WCAP-17658-NP, Revision 1, "Wolf Creek Generating Station Transition of Methods for Core Design and Safety Analyses - Licensing Report," to the letter dated January 17, 2017, refer to Westinghouse Letter L TR-NRC-12-18, dated February 17, 2012, "Westinghouse Response to December 16, 2011 NRC Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary)" (ADAMS Accession No. ML12053A105).

In Attachment I of the letter dated January 17, 2017, the licensee stated in part:

Regarding the impact of the issue of fuel thermal conductivity degradation (TCD) on the Westinghouse codes and methods, Westinghouse... justified continued operation of the plants analyzed with Westinghouse codes and methods. The Westinghouse codes and methods applied in the non-LOCA analyses discussed in Enclosure I are consistent with those evaluated for TCD in [LTR-NRC-12-18],

and therefore the conclusions presented in [L TR-NRC-12-18] are applicable to theWCGS.

Section 2, "Accident and Transient Analysis," of Enclosure I of the letter dated January 17, 2017, contains a similar passage.

Section 2 of Enclosure I also presents results for two specific design-basis accidents, among others, that are the subject of this request. The first is the Rod Cluster Control Assembly (RCCA) ejection accident (CREA), discussed in Section 2.5.6 of Enclosure 1. The second is the main steamline break (MSLB) accident, discussed in Section 2.2.5 of Enclosure I.

Enclosure Request for Additional Information Rod Cluster Control Assembly Ejection Accident Please demonstrate that the CREA analysis results discussed in Section 2.5.6 of Enclosure I are consistent with General Design Criterion (GDC) 28, "Reactivity limits," of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10 of the Code of Federal Regulations (10 CFR) Part 50 for long term cooling, and with the assumptions employed with the radiological analyses supporting the implementation of the alternative source term in accordance with 10 CFR 50.67, "Accident source term." GDC 28 requires that the effects of postulated reactivity accidents neither damage the reactor coolant pressure boundary greater than limited local yielding, nor impair the ability to cool the core. Meanwhile, the radiological consequence analysis of the CREA assumes no more than 10-percent of fuel centerline melt at the hot spot.

Since the NRC position stated in Regulatory Guide 1.77, "Assumptions used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," May 1974 (ADAMS Accession No. ML003740279), indicating that acceptance criteria of 280 calories per gram (cal/gm) should be applied to the event, and Westinghouse's position that 200 cal/gm and 10-percent fuel centerline melt criteria contained in WCAP-7588, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," were established, the NRC staff determined that more restrictive acceptance criteria were needed to assure the core would remain in a coolable geometry, given the potential for high temperature cladding failure and pellet cladding interaction. This position is documented in an NRC staff internal memo dated April 3, 2015, titled, "Results of Periodic Review of Regulatory Guide 1. 77" (ADAMS Accession No. ML15075A311 ). The position is supported by the NRC staff internal memo dated January 19, 2007, titled, "Technical and Regulatory Basis for the Reactivity Initiated Accident Interim Acceptance Criteria and Guidance" (ADAMS Accession No. ML070220400).

a) Please justify why 200 cal/gm and 10-percent fuel melt are appropriate acceptance criteria given that much more is known now about fuel damage behavior than when the method was approved in 1975.

b) Please discuss how the calculations appropriately include the effects of nuclear fuel TCD in the evaluation against the acceptance criteria. The hot spot fuel melt limit of 10-percent relates to the prevention of fuel dispersal into the coolant, and is reflected in the assumptions employed in the radiological analyses.

Main Steamline Break Accident Please demonstrate how the MSLB analyses contained in Section 2.2.5 of Enclosure I are consistent with the requirements of 10 CFR Part 50, Apendix A, GDC 27, "Combined reactivity control systems capability," and GDC 28. GDC 27 requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes so that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. GDC 28 requires that the effects of postulated reactivity accidents neither damage the reactor coolant pressure boundary greater than limited local yielding, nor impair the ability to cool the core. Per the Westinghouse analytic methods proposed for implementation, the MSLB is analyzed to demonstrate that fuel damage criteria are satisfied, including departure from nucleate boiling limits and fuel melt limits.

In order for the NRC staff to determine whether the MSLB analyses assure compliance with the requirements of GDC 27 and 28, please discuss how the MSLB results and acceptance criteria appropriately include the effects of TCD, including:

a) The hot zero power and hot full power departure from nucleate boiling ratio, and b) The hot full power peak linear heat rate.

ML18270A094 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/SRXB/BC NAME BSingal PBlechman w/comment JWhitman DATE 10/2/18 9/28/18 10/2/18 OFFICE NRR/DSS/SNPB/BC NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME RLukes RPascarelli BSingal DATE 10/3/18 10/4/18 10/4/18