ML23103A468

NRR E-mail Capture - Comanche Peak Request for Additional Information - License Amendment Request to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model

ML23103A468
Person / Time
Site:	Comanche Peak
Issue date:	04/11/2023
From:	Dennis Galvin NRC/NRR/DORL/LPL4
To:	Hicks J Vistra Operations Company
References
L-2022-LLA-0171
Download: ML23103A468 (6)

Text

From: Dennis Galvin Sent: Tuesday, April 11, 2023 5:04 PM To: Jack Hicks (Jack.Hicks@luminant.com)

Cc: Nic Boehmisch (Nicholas.Boehmisch@luminant.com)

Subject:

Comanche Peak - Request for Additional Information - License Amendment Request to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model. (EPID: L-2022-LLA-0171)

Attachments: Comanche Peak Adopt FSLOCA LAR - RAI Issued 2023-04-11.pdf

Dear Mr. Hicks,

By letter dated November 21, 2022 (ML22325A324), Vistra Operations Company LLC ("Vistra OpCo", the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for Comanche Peak, Units 1 and 2 (Comanche Peak). The proposed amendment would revise the Comanche Peak Technical Specification (TS) to reflect the adoption of topical report (TR) WCAP-16996-P-A, Revision 1, "Realistic LOCA [loss-of-coolant accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (Full Spectrum LOCA Methodology), (FSLOCA)" (ML16343A238). The proposed amendments revise TS 2.1.1.2 reactor core safety limit (SL), to reflect the peak fuel centerline melt temperature specified in TR WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (ML17335A334), and revise the TS 4.2.1 reactor core fuel assemblies design feature by removing the discussion of Zircalloy fuel rods and ZIRLO lead test assemblies for Comanche Peak Unit 1 and Unit 2. The amendment also proposes to delete references 11 through 14 in the TS 5.6.5.b list of analytical methods and to add a new approved analytical method, i.e., FSLOCA.

The NRC staff has determined that additional information is needed to complete its review. The requests for additional information (RAIs) were transmitted to the licensee in draft form on March 27, 2023. A clarification call was held with your staff on April 6, 2023, and the licensee agreed to provide responses to the RAls by June 5, 2023. The NRC staff agrees with this date.

Respectfully, Dennis Galvin Project Manager U.S Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 4 301-415-6256 Docket Nos. 50-445 and 50-446

Hearing Identifier: NRR_DRMA Email Number: 2037 Mail Envelope Properties (SA1PR09MB8111118152832D54B631CE9CFB9A9)

Subject:

Comanche Peak - Request for Additional Information - License Amendment Request to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model. (EPID L-2022-LLA-0171)

Sent Date: 4/11/2023 5:04:24 PM Received Date: 4/11/2023 5:04:00 PM From: Dennis Galvin Created By: Dennis.Galvin@nrc.gov Recipients:

"Nic Boehmisch (Nicholas.Boehmisch@luminant.com)" <Nicholas.Boehmisch@luminant.com>

Tracking Status: None "Jack Hicks (Jack.Hicks@luminant.com)" <Jack.Hicks@luminant.com>

Tracking Status: None Post Office: SA1PR09MB8111.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1831 4/11/2023 5:04:00 PM Comanche Peak Adopt FSLOCA LAR - RAI Issued 2023-04-11.pdf 388051 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK UNIT 1 AND 2 DOCKET NOS. 50445 AND 50446 INTRODUCTION By letter dated November 21, 2022 (Reference 1), Vistra Operations Company LLC ("Vistra OpCo," the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for Comanche Peak, Units 1 and 2 (Comanche Peak). The proposed amendment would revise the Comanche Peak Technical Specification (TS) to reflect the adoption of topical report (TR) WCAP16996P-A, Revision 1, "Realistic LOCA

[loss-ofcoolant-accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (Full Spectrum LOCA Methodology), (FSLOCA)" (Reference 2). The proposed amendments revise TS 2.1.1.2 reactor core safety limit (SL), to reflect the peak fuel centerline melt temperature specified in TR WCAP17642P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (Reference 3), and revise the TS 4.2.1 reactor core fuel assemblies design feature by removing the discussion of Zircalloy fuel rods and ZIRLO lead test assemblies for Comanche Peak Unit 1 and Unit 2. The amendment also proposes to delete references 11 through 14 in the TS 5.6.5.b list of analytical methods and to add a new approved analytical method (Reference 2).

After reviewing the LAR (Reference 1), the staff requests responses to the requests for additional information (RAIs) given below.

RAI 1

Regulatory Basis:

The regulations in 10 CFR 50.46(b) require the following criteria to be met during LOCA events:

(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200° F.

(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

Regarding errors in evaluation models, 10 CFR 50.46(a)(3)(i) states:

Each applicant for or holder of an operating license or construction permit issued under this part, applicant for a standard design certification under Part 52 of this chapter (including an applicant after the Commission has adopted a final design certification regulation), or an applicant for or holder of a standard design approval, a combined license or a manufacturing license issued under Part 52 of this chapter, shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature different by more than 50 °F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 °F.

10 CFR 50.46(a)(3)(ii) states in part For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or holder of a construction permit, operating license, combined license, or manufacturing license shall report the nature of the change or error and its estimated effect on the limiting ECCS

[emergency core cooling system] analysis to the Commission at least annually as specified in § 50.4 or § 52.3 of this chapter, as applicable. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with § 50.46 requirements.

RAI:

LAR Attachment 4 indicates that it was prepared in 2019. Further, in LAR, Attachment 4, Limitation and Condition Number 2, under Compliance, the licensee indicates that the analyses reflects changes to the FSLOCA methodology as described in LTR-NRC1830 (Reference 4) but does not incorporate changes to address errors described in LTR-NRC196 (Reference 5), which reports two errors were discovered in the WCOBRA/TRAC-TF2 code that can occur under certain conditions in the ECCS evaluation models. The letter LTR-NRC196 report that these errors were found to have negligible impact on analysis results with the FSLOCA EM. In addition, the LTR-NRC196 report (and similar Westinghouse annual submittals) states:

It is noted that plant-specific peak cladding temperature (PCT) variations are not addressed in this letter.

i. Confirm that the applicable changes and errors described in LTR-NRC1830 are reflected in the Comanche Peak analyses and the results are quantitively obtained using the revised code.

ii. For each change to or error discovered in the FSLOCA model not reflected in the analyses supporting LAR Attachment 4, including those described in Westinghouse LTR-NRC235, dated March 10, 2023 (Reference 6), which reports two errors discovered in the

WCOBRA/TRAC-TF2 code, quantitatively describe its plant-specific effect on the results, errors and uncertainties.

RAI 2

Regulatory Basis:

Same as in RAI 1 RAI:

In LAR, Attachment 4, Section 1.3, the description of the small break OCA (Region I) analysis does not provide the range of break areas/sizes that were analyzed. LAR, Attachment 4, Section 1.3.2 identifies the break sizes and LAR, Attachment 4, Tables 8A and 8B, represent the time sequence of events for the PCTs results given in LAR, Attachment 4, Tables 7A and 7B. However, corresponding break sizes and sequence of events are not provided for maximum local oxidations (MLOs), core-wide oxidations (CWOs) results given in LAR, Attachment 4, Tables 7A and 7B., Identify the range of range of break areas/sizes that were analyzed and the break areas/sizes for which MLOs and CWOs, and the time sequence of events for the Region I results given in LAR, Attachment 1, Tables 7A and 7B.

RAI3 Regulatory Basis:

Same as in RAI 1 RAI:

In LAR, Attachment 4, Section 1.4, the description of the large break LOCA (Region II) analysis does not provide the break spectrum scenarios that were analyzed. LAR, Attachment 4, Tables 9A and 9B, represent the time sequence of events for the PCTs results given in LAR, , Tables 7A and 7B. However, the corresponding break sizes are not provided for the PCTs results, and the corresponding break sizes and sequence of events are not provided MLOs,and CWOs results given in LAR, Attachment 4, Tables 7A and 7B. Provide the scatter pots of the PCT vs break size, transient equivalent cladding reacted vs PCT, core wide oxidation vs PCT, and the time sequence of events for the Region II results given in LAR, attachment 1, Tables 7A and 7B.

RAI 4

Regulatory Basis:

10 CFR 50.46(b)(4) on Coolable Geometry: Calculated changes in core geometry shall be such that the core remains amenable to cooling.

RAI:

LAR, Attachment 4, Section 1.5, addresses compliance with 10 CFR 50.46. Regarding compliance with 10 CFR 50.46(b)(4), the last sentence of third paragraph, states:

Inboard grid deformation due to combined LOCA and seismic loads is not calculated to occur for Comanche Peak.

The LAR does not provide any summary of the calculations or the results or other technical basis to support this statement. Provide a summary of the calculations and the results or other technical basis to support the statement.

REFERENCES

1. Vistra Operations Company LLC ("Vistra OpCo") letter to NRC, Application to Revise Technical Specifications to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model. LAR 22-002 License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt FULL SPECTRUMTM LOCA Methodology, dated November 21, 2022 (ML22325A324).

2. Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), WCAP16996P-A, Revision 1, dated November 2016 (ML16343A238).

3. Westinghouse Performance Analysis and Design Model (PAD5), WCAP17642P-A, Revision 1, dated November 2017 (ML17335A334).

4. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, LTR-NRC1830, dated July 18, 2018 (ML19288A174).

5. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, LTR-NRC-19-6, dated February 7, 2019 (ML19042A378).

6. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2022, LTR-NRC-23-5, dated March 10, 2023 (ML23072A071).