LR-N24-0072, Response to Request for Additional Information – Proposed Amendment to Implement Optimized Zirlotm Fuel Rod Cladding

Response to Request for Additional Information – Proposed Amendment to Implement Optimized Zirlotm Fuel Rod Cladding

ML24352A080
Person / Time
Site:	Salem
Issue date:	12/15/2024
From:	Sharbaugh D Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
LR-N24-0072, LAR S24-01, EPID L-2024-LLA-0100
Download: ML24352A080 (1)

Text

David Sharbaugh Site Vice President - Salem Generating Station - PSEG Nuclear PO Box 236 Hancocks Bridge, New Jersey 08038-0221 david.sharbaugh@pseg.com

10 CFR 50.90 LR-N24-0072 LAR S24-01 December 15, 2024 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Response to Request for Additional Information - Proposed Amendment to Implement Optimized ZIRLOTM Fuel Rod Cladding - PSEG Nuclear LLC -

Salem Generating Station Unit Nos. 1 and 2

References:

1.

PSEG letter to NRC, Application to Revise Salem Generating Station Units 1 and 2 Technical Specifications and 10 CFR 50.12 Exemption Request to Implement Optimized ZIRLOTM Fuel Rod Cladding dated July 24, 2024 (ML24206A100)

2.

NRC e-mail to PSEG, Salem Nuclear Generation Station Unit Nos. 1 and 2 - Request for Additional Information Regarding License Amendment Request to Use Optimized ZIRLO Fuel Rod Cladding dated November 21, 2024 (EPID L-2024-LLA-0100) (ADAMS Accession No. ML24326A148)

In the Reference 1 letter, PSEG Nuclear LLC (PSEG) submitted a license amendment (LAR) for Salem Generating Station (Salem) Unit 1 and Unit 2. The proposed amendment would revise Salem Technical Specifications (TS) 5.3.1 to allow the use of Optimized ZIRLOTM1 as an approved fuel rod cladding material.

In Reference 2, The U.S. Nuclear Regulatory Commission staff provided PSEG a Request for Additional Information (RAI) to support the NRC staffs detailed technical review of Reference 1. to this letter contains the response to the RAI question contained in Reference 2.

PSEG has determined that the information provided in this submittal does not alter the conclusions reached in the 10 CFR 50.92 no significant hazards determination previously submitted. In addition, the information provided does not affect the bases for concluding that 1Optimized ZIRLO is a trademark of Westinghouse Electric Company LLC o PSEG I NUCLEAR

LR-N24-0072 Page 2 10 CFR 50.90 neither an environmental impact statement nor an environmental assessment need to be prepared in connection with the proposed amendment.

There are no regulatory commitments contained in this letter.

If there are any questions or if additional information is needed, please contact Mr. Michael Wiwel at Michael.Wiwel@pseg.com.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully, David Sharbaugh Site Vice President Salem Generating Station Attachments:

1 Response t~ Request for Additional Information for Proposed Amendment to lmpleme t Optimized ZIRLO' Fuel Rod Cladding (EPID L-2024-LLA-0100) 2 Technical S ecification Page Markups cc:

Administrator, Region 1,1 NRC NRC Project Manager, Salem NRC Senior Resident Inspector, Salem Manager, NJBNE PSEG Commitment Tracking Coordinator

LR-N24-0072 LAR S24-01

Response to Request for Additional Information for Proposed Amendment to Implement Optimized ZIRLOTM Fuel Rod Cladding (EPID L-2024-LLA-0100)

LR-N24-0072 LAR S24-01

BACKGROUND Optimized ZIRLO' is described in Westinghouse Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO' (Addendum 1-A). Addendum 1-A includes the NRC staff conditions on the approval of Addendum 1. The LAR addresses the NRC staff conditions in Addendum 1-A, and states in several places that specific NRC staff conditions will be confirmed as part of the normal reload design process. The LAR implies that Addendum 1-A would need to be used when establishing core operating limits if Optimized ZIRLO' will be used in the reload core.

Technical Specification 6.9.1.9, Core Operating Limits Report (COLR), for Salem, Unit Nos. 1 and 2, specifies administrative controls for establishing core operating limits prior to each reload cycle, or prior to any remaining portion of a reload cycle. Technical Specification 6.9.1.9 requires the licensee to use the specific analytical methods listed in TS 6.9.1.9.b to determine the core operating limits. However, the LAR for Salem does not propose to add Addendum 1-A to the list of analytical methods for determining core operating limits in TS 6.9.1.9.

The LAR cites license amendments the NRC staff issued for three facilities as precedents to support the use of Optimized ZIRLO' at Salem. These facilities have TSs for the COLR that are similar to Salem TS 6.9.1.9. Among other changes, the cited license amendments for these facilities amended the TSs for the COLR to added Addendum 1-A to the list of analytical methods used to determine the core operating limits.

Request for Additional Information Identify whether a supplement to this LAR or a separate LAR will be submitted to revise TS 6.9.1.9 to add Addendum 1-A to the list of analytical methods used for determining core operating limits prior to the initial loading of fuel with Optimized ZIRLO' at the Salem units.

Alternatively, explain how the existing analytic methods in TS 6.9.1.9 can be used without modification to determine core operating limits for a reload that includes fuel with Optimized ZIRLO'. This explanation should include a discussion of why the NRC conditions in Addendum 1-A would not impact the establishment of the core operating limits for the Salem units.

Response to Request for Additional Information PSEG Nuclear LLC (PSEG) proposes changes to TS 6.9.1.9.b to include the references to WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM," to the list of analytical references used to determine the core operating limits for Salem Generating Station, Units 1 and 2 (Salem). Markups of the proposed TS changes for both Salem Unit 1 and Unit 2 are shown in Attachment 2. The proposed TS changes provided in the initial application dated July 24, 2024 (ADAMS Accession No. ML24206A100) are not impacted by this supplement and are augmented by those provided in Attachment 2 to this letter.

LR-N24-0072 LAR S24-01

Technical Specification Page Markups The following Technical Specifications page for Renewed Facility Operating License DPR-70 is affected by this RAI response:

Technical Specification Page 6.9.1.9.b 6-24a The following Technical Specifications page for Renewed Facility Operating License DPR-75 is affected by this change request:

Technical Specification Page 6.9.1.9.b 6-24a

ADMINISTRATIVE CONTROLS 6.9.1.9

a.

CORE OPERATING LIMITS REPORT (COLR)

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.4,

2.

Control Bank Insertion Limits for Specification 3/4.1.3.5,

3.

Axial Flux Difference Limits and target band for Specification 3/4.2.1,

4.

Heat Flux Hot Channel Factor, FQ, its variation with core height, K(z), and Power Factor Multiplier PFxy, Specification 3/4.2.2, and 6

Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PF4tt for Specification 3/4.2.3.

6.

Refueling boron concentration per Specification 3.9.1

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (~ Proprietary), Methodology for Specifications listed in 6.9.1.9.a.

SALEM -

UNIT 1 6-24 Amendment No. 284 Page Included for Information Only. No change on this page

ADMINISTRATIVE CONTROLS

2.

WCAP-8385, Power Distribution Control and Load Following Procedures -

Topical Report, (W Proprietary) Methodology for Specification 3/4.2.1 Axial Flux Difference.

3.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code (W Proprietary), Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

4.

WCAP-10266-P-A, The 1981 Version of Westinghouse Evaluation Model Using BASH Code, (W Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

5.

WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, (W Proprietary).

6.

CENPD-397-P-A, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology.

c.

The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.9.1.10 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.8.4.i, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Then nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment;

4. The number of tubes plugged during the inspection outage; and SALEM - UNIT 1 6-24a Amendment No. 338

7. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary).

XXX

ADMINISTRATIVE CONTROLS 6.9.1.9 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

2.

Control Bank Insertion Limits for Specification 3/4.1.3.5,

3.

Axial Flux Difference Limits and target band for Specification 3/4.2.1,

4.

Heat Flux Hot Channel Factor, Fa, its variation with core height, K(z), and Power Factor Multiplier PFxy, Specification 3/4.2.2, and

5.

Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PFt.H for Specification 3/4.2.3.

6.

Refueling boron concentration per Specification 3.9.1

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, 0!:f.. Proprietary), Methodology for Specifications listed in 6.9.1.9.a.

SALEM - UNIT 2 6-24 Amendment No. 278 Page included for information only. No change on this page

ADMINISTRATIVE CONTROLS

2.

WCAP-8385, Power Distribution Control and Load Following Procedures -

Topical Report, (W Proprietary) Methodology for Specification 3/4.2.1 Axial Flux Difference

3.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code, (W Proprietary), Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

4.

WCAP-10266-P-A, The 1981 Version of Westinghouse Evaluation Model Using BASH Code, (W Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

5.

WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, (W Proprietary).

6.

CENPD-397-P-A, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology

7.

WCAP-10054-P-A, Addendum 2, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model.

c.

The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements shall be provided upon issuance for each reload cycle to the NRC.

6.9.1.10 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.8.4.i, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG;

b.

Then nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported; SALEM - UNIT 2 6-24a Amendment No. 320

8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary)

XXX