ML17100A266

Supplemental Information Needed for Acceptance of Requested Licensing Action Transition to Westinghouse Core Design and Safety Analyses

ML17100A266
Person / Time
Site:	Wolf Creek
Issue date:	04/18/2017
From:	Balwant Singal Plant Licensing Branch IV
To:	Helfin A Wolf Creek
Singal B, DORL/LPLIV, 415-3016
References
CAC MF9307
Download: ML17100A266 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 18, 2017 Mr. Adam C. Heflin President, Chief Executive Officer, and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION-SUPPLEMENTAL INFORMATION NEEDED FOR ACCEPTANCE OF REQUESTED LICENSING ACTION RE:

TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSES (CAC NO. MF9307)

Dear Mr. Heflin:

By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a license amendment request (LAR) for Wolf Creek Generating Station (WCGS). The proposed LAR would replace the existing WCNOC methodology for performing core design, non-loss-of-coolant-accident (non-LOCA) and LOCA safety analyses to the standard Westinghouse U.S. Nuclear Regulatory Commission (NRC)-approved methodologies for performing these analyses and associated technical specification (TS) changes at WCGS. In addition, the LAR proposes to revise the WCGS licensing basis for adopting the alternative source term radiological analysis methodology in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.67, "Accident source term."

The purpose of this letter is to provide the results of the NRC staff's acceptance review of this LAR. The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review.

The acceptance review is also intended to identify whether the application has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant.

Consistent with 10 CFR 50.90, an amendment to the license (including the TSs) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required. This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.

The NRC staff has reviewed your application and concluded that the information delineated in the enclosure to this letter is necessary to enable the staff to make an independent assessment regarding the acceptability of the proposed LAR in terms of regulatory requirements and the protection of public health and safety and the environment.

A. Heflin In order to make the application complete, the NRC staff requests that WCNOC supplement the application to address the information requested in the enclosure by May 5, 2017. This will enable the NRC staff to begin its detailed technical review. If the information responsive to the NRC staff's request is not received by the above date, the application will not be accepted for review pursuant to 10 CFR 2.101, and the NRC will cease its review activities associated with the application. If the application is subsequently accepted for review, you will be advised of any further information needed to support the staff's detailed technical review by separate correspondence.

The information requested and associated timeframe in this letter were discussed with Mr. William Muilenburg of your staff on April 10, 2017.

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

Sincerely,

~-+~k,~

Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure Supplemental Information Needed cc w/encl: Distribution via Listserv

SUPPLEMENTAL INFORMATION NEEDED LICENSE AMENDMENT REQUEST TO TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANL YSES WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 By letter dated January 17, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17054C103), Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a license amendment request (LAR) for Wolf Creek Generating Station (WCGS). The proposed LAR would replace the existing WCNOC methodology for performing core design, non-loss-of-coolant-accident (non-LOCA) and LOCA safety analyses to the standard Westinghouse U.S. Nuclear Regulatory Commission (NRC)-approved methodologies for performing these analyses and associated technical specification (TS) changes at WCGS.

The proposed amendment also revises the Updated Safety Analysis Report, Chapter 15, radiological consequence analyses and associated TSs using an updated accident source term consistent with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, "Accident source term."

The NRC staff has reviewed your application and concluded that the following supplemental information is necessary to enable the NRC staff to make an independent assessment regarding the acceptability of the proposed LAR in terms of regulatory requirements and the protection of public health and safety and the environment.

The NRC staff recognizes that the information being requested is likely to be part of the radiological accident analysis calculations performed by the licensee, but not submitted as part of the LAR documents. The licensee has the option of providing the NRC staff a detailed summary of the analyses including inputs, assumptions, methodology technical basis, and results or submit the analyses clearly identifying the requested information, as necessary.

The supplemental information needed to complete the acceptance review is as follows:

1. Page 87 of Enclosure VI I to the letter dated January 17, 2017 (ADAMS Accession No. ML17054C229), states, in part:

For the fuel handling accident, two possible accident types are considered: (1) in containment and (2) in fuel building. For a fuel handling accident occurring in containment, the transport path is through the open equipment hatch. For a fuel handling accident occurring in the fuel building, the radionuclides released are Enclosure

drawn by the building ventilation system to the unit vent stack and eventually exhausted to the environment through the unit vent stack [emphasis added].

Section 50.36, 'Technical specifications," of 10 CFR requires the TSs to be derived from the analyses and evaluation included in the safety analysis report. Per WCGS TS Bases B 3.7.13, the emergency exhaust system's design basis is established by the consequences of the limiting design-basis accidents including the fuel handling accident (FHA) analysis.

A note in WCGS TS 3.7.13, "Emergency Exhaust System (EES)," allows the fuel building boundary to be opened intermittently under administrative controls, and Condition E allows two EES trains to be inoperable (due to an inoperable fuel building boundary) for 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> during the movement of irradiated fuel (when an FHA could occur). Enclosure IV to the letter dated January 17, 2017 (ADAMS Accession No. ML17054C227), Table 15B-1, "Conformance with Regulatory Guide [RG] 1.183 Main Section," the licensee states that the WCGS analysis conforms to RG 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession No. ML003716792),

Regulatory Position 5.1.2, "Credit for Engineered Safeguard Features," which states, in part, that credit may be taken for accident mitigation features that are required to be operable by TS. Therefore, when the fuel building boundary is inoperable (during movement of irradiated fuel) it should not be credited in the FHA analysis.

However, the technical analysis section of the LAR does not include consideration for a scenario where the FHA occurs while the fuel handling building boundary is open or not operable as allowed by the WCGS TS 3.7.13. For example, Enclosure VII, Table 4.1.2-3(a), "Calculated xtQ (sec/m 3 ) for the Emergency Control Room Intake Vent," which includes the atmospheric dispersion factor (x/Q) values determined for the control room and technical support center (TSC) intakes does not contain an xtQ values for a release from the fuel building boundary location to the control room and TSC.

To be consistent with the allowances of TS 3. 7 .13:

a. Please submit for the NRG staff's review, a detailed summary of the results of the radiological consequences of an FHA that supports: 1) the fuel building boundary being open intermittently under administrative control and closed during an FHA and 2) an inoperable fuel building boundary (as allowed by Condition E) for the duration of the event. Please show that the dose results for these scenarios meet the limits in General Design Criterion (GDC) 19, "Control room," of 10 CFR 50 Appendix A and 10 CFR 50.67. In addition, please provide the inputs, assumptions, methodology and technical basis for the analysis.

or

b. Please provide a proposed change to TS 3. 7.13 that is consistent with the analysis proposed in the LAR for the FHA analysis.

2. Page 51 of Enclosure VII to the letter dated January 17, 2017, states:

The fuel handling accident (FHA) analysis conservatively assumed that 100% of the fuel will not [emphasis added] meet the Footnote 11 from Regulatory Position 3.2 of Regulatory Guide 1.183. Footnote 11 contains burnup and linear heat generation rate limits for the gap fractions contained in Table 3 of Regulatory Guide 1.183 as well as an alternative method for determining gap fractions provided they bound the limiting projected plant-specific power history for the specific load. The analysis conforms with the position such that alternative gap fractions were used which were appropriate for the assumption of not meeting the Footnote 11 limits. These gap fractions are obtained from Regulatory Guide 1.25, as modified by NUREG/CR-5009, which provides higher, i.e.,

more conservative gap fractions than Regulatory Guide 1.183, which are not constrained by the Footnote 11 burnup limits as they can be applied to higher burnup to bound power history [emphasis added].

RG 1.183, Footnote 11 for Table 3, "Non-LOCA Fraction of Fission Product Inventory in Gap," states, in part:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR [light-water reactor] fuel with a peak burnup up to 62,000 MWD/MTU [megawatt-days per metric ton of uranium] provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft [kilowatt per foot] peak rod average power for burnups exceeding 54 GWD/MTU [gigawatt-days per metric ton of uranium]. As an alternative, fission gas release calculations performed using NRG-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load.

The LAR also proposes to use Table 2, "PWR [pressurized-water reactor] Core Inventory Fraction Released into Containment" of RG 1.183, which is contained in Regulatory Position 3.2, "Release Fractions." Regulatory Position 3.2 has a footnote (Footnote 10) that states, in part:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU.

Because of the cited text above from the LAR, it is unclear whether the proposed LAR meets the burnup limits in Footnote 10. Please clearly state the peak proposed burnup limits assumed in support of the analyses. If it is not less than or equal to 62,000 MWD/MTU (peak rod burnup), please provide a technical justification why Table 2 in RG 1.183 is applicable to burnups higher than 62,000 MWD/MTU. Note that RG 1.183, Regulatory Position 2 provides attributes of an acceptable alternative source term (AST) that the NRC staff will use for its review of an AST different than the one provided in RG 1.183.

3. Enclosure IV, Table C, "Conformance with Regulatory Guide 1.183, Appendix B (Fuel Handling Accident)," of letter dated January 17, 2017, states that the analysis in the LAR conforms with RG 1.183 Appendix B, Regulatory Position 5.3, which states:

If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

The Table entitled "Regulatory Guide 1.194 Comparison" in the LAR states that the analysis in the LAR conforms with RG 1.194, [Revision 0, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants" (ADAMS Accession No. ML031530505), Regulatory Position 3.2.4.2, which states:

Since leakage is more likely to occur at a penetration, analysts must consider the potential impact of building penetrations exposed to the environment within this modeled area. If the penetration release would be more limiting, the diffuse area source model should not be used.

Releases from personnel air locks and equipment hatches exposed to the environment, or containment purge releases prior to containment isolation, may need to be treated differently. It may be necessary to consider several cases to ensure that the xtQ value for the most limiting location is identified.

Section 50.36 of 10 CFR requires the TSs to be derived from the analyses and evaluation included in the safety analysis report. Per WCGS TS Bases B 3.9.4, the applicable safety analysis for the containment penetrations is the most severe radiological consequences from an FHA.

WCGS TS 3.9.4, "Containment Penetrations," allows many different containment configurations during Core Alterations and during the movement of irradiated fuel assemblies within containment, including allowances for the equipment hatch, emergency air lock and other containment penetrations to be unisolated under administrative controls. However, the NRC staff could not locate a complete evaluation for the FHA occurring in containment with the individual penetrations (allowed to be unisolated under administrative controls by the Note in TS 3.9.4(c) open.

To be consistent with the allowances of TS 3.9.4:

a. Please submit for the NRC staff's review a detailed summary of the radiological consequences of an FHA in containment that supports the containment penetrations being open under administrative control and closed during an accident to justify the most severe radiological consequences from an FHA.

Please show that the dose results for these scenarios meet the limits in GDC 19 of 10 CFR Part 50, Appendix A and 10 CFR 50.67. In addition, please provide the inputs, assumptions, methodology and technical basis for the analysis.

or

b. Please provide a proposed change to TS 3.9.4 that is consistent with the analysis proposed in the LAR for the FHA in containment.

4. Consistent with 10 CFR 50.90, an amendment to the license (including the TSs) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Issue 1, "Level of Detail Contained in LARs" of the NRC's Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," dated March 7, 2006 (ADAMS Accession No. ML053460347) states, in part, that an "AST amendment request should describe the licensee's analyses of the radiological and non-radiological impacts and provide a justification [emphasis added] for the proposed modification in sufficient detail to support review by the NRC staff."

In Tables 4.3-5 through 4.3-16 of Enclosure VII to letter dated January 17, 2017, the licensee provided a brief generic "Reason for [the] Change." In some cases the reasons for the change provided can also be used as a justification for the change. For example, the NRC staff understands that some values are justified because they "allow for additional surveillance testing margin" or are a result of conforming to a "Regulatory Guide update."

In other cases the "Reasons for [the] Change" do not provide justifications for the changes. For example the "RCS [Reactor Coolant System] mass, maximum" in Table 4.3-6 for the AST is 8.42E+5 lbm [pounds mass] and the CLB [current licensing basis] is 4.94E+5 lbm. The "Reason for [the] Change" is specified as a "Modeling update (not AST specific)." While this is the reason for the change, the NRC staff needs to understand the technical basis behind the change and why the new value or assumption is acceptable. In Table 4.3-8, the "Time of Control room isolation (including delays) (sec)" for the AST is 120 seconds, but this parameter is marked N/A for the CLB.

The "Reasons for [the] Change" provided is "No Change" although this appears to be a change. In these cases and for several others in Tables 4.3-5 through 4.3-16, the information provided does not justify the proposed changes.

Please review the "Reasons for [the] Change" and update Tables 4.3-5 through 4.3-16 to include a justification (a technical basis that addresses why these changes are acceptable) for the changes made to the CLB.

ML17100A266 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DRA/ARCB/BC NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME BSingal PBlechman US hoop RPascarelli BSingal DATE 4/13/17 4/12/17 4/17/17 4/18/17 4/18/14