ML25073A111
| ML25073A111 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/24/2025 |
| From: | Blake Purnell NRC/NRR/DORL/LPL2-2 |
| To: | Mcfeaters C Public Service Enterprise Group |
| Kim J | |
| References | |
| EPID L-2024-LLA-0100 | |
| Download: ML25073A111 (1) | |
Text
April 24, 2025 Charles V. McFeaters President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 350 AND 332 TO PERMIT USE OF OPTIMIZED ZIRLO' FUEL ROD CLADDING (EPID L-2024-LLA-0100)
Dear Mr. McFeaters:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 350 and 332 to Renewed Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, respectively. The amendments are in response to the PSEG Nuclear LLC (PSEG) application submitted by letter dated July 24, 2024 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML24206A100), as supplemented by letter dated December 15, 2024 (ML24352A080). The amendments revise the Salem technical specifications to allow for the use of Optimized ZIRLO' fuel rod cladding material and make other editorial changes.
The July 24, 2024, letter from PSEG also requested an exemption from certain emergency core cooling system requirements to support the issuance of the amendments. By letter dated April 24, 2025 (ML25078A006), the NRC staff granted PSEGs exemption request.
C. McFeaters A copy of the NRC staffs safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Blake Purnell, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311
Enclosures:
- 1. Amendment No. 350 to DPR-70
- 2. Amendment No. 332 to DPR-75
- 3. Safety Evaluation cc: Listserv PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 350 Renewed License No. DPR-70 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by PSEG Nuclear LLC dated July 24, 2024, as supplemented by letter dated December 15, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-70 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 350, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: April 24, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.04.24 14:50:55 -04'00'
Renewed License No. DPR-70 Amendment No. 350 ATTACHMENT TO LICENSE AMENDMENT NO. 350 RENEWED FACILITY OPERATING LICENSE NO. DPR-70 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NO. 50-272 Replace the following page of Renewed Facility Operating License No. DPR-70 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 5-4 5-4 6-24 6-24 6-24a 6-24a Renewed License No. DPR-70 Amendment No. 350 (4)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 350, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.
(3)
Deleted Per Amendment 22, 11-20-79 (4)
Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this renewed license) with less than four (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this renewed license.
(5)
PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety
DESIGN FEATURES
============================================================
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 47 psig. Containment air temperatures up to 351.3°F are acceptable providing the containment pressure is in accordance with that described in the UFSAR.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO or Optimized ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies.
The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
SALEM - UNIT 1 5-4 Amendment No. 350
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 3/4.1.1.4, 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 5.
Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PFH 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, (W Proprietary), Methodology for Specifications listed in 6.9.1.9.a.
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-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary).
SALEM - UNIT 1 6-24 Amendment No. 350
ADMINISTRATIVE CONTROLS 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. 350 PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 332 Renewed License No. DPR-75 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by PSEG Nuclear LLC dated July 24, 2024, as supplemented by letter dated December 15, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-75 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 332, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: April 24, 2025 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2025.04.24 14:52:02 -04'00'
Renewed License No. DPR-75 Amendment No. 332 ATTACHMENT TO LICENSE AMENDMENT NO. 332 RENEWED FACILITY OPERATING LICENSE NO. DPR-75 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 DOCKET NO. 50-311 Replace the following page of Renewed Facility Operating License No. DPR-75 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 5-4 5-4 6-24 6-24 6-24a 6-24a Renewed License No. DPR-75 Amendment No. 332 (3)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended (4)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source or special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; (5)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 3459 megawatts (thermal).
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 332, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
DESIGN FEATURES
============================================================
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 47 psig. Containment air temperatures up to 351.3ºF are acceptable providing the containment pressure is in accordance with that described in the UFSAR.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO or Optimized ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies.
The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirement specified in Section 4.1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 650ºF, except for the pressurizer which is 680ºF.
SALEM - UNIT 2 5-4 Amendment No. 332
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, FQ, 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, PFH 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, (W Proprietary), Methodology for Specifications listed in 6.9.1.9.a.
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.
SALEM - UNIT 2 6-24 Amendment No. 332
ADMINISTRATIVE CONTROLS 8.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary) 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. 332 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 350 AND 332 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By application dated July 24, 2024 (Agencywide Documents Access and Management System Accession No. ML24206A100), as supplemented by letter dated December 15, 2024 (ML24352A080), PSEG Nuclear LLC (the licensee) submitted a license amendment request (LAR) for Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2. The amendments would revise the Salem technical specifications (TSs) to allow for the use of Optimized ZIRLO' fuel rod cladding material and make other editorial changes.
The licensees July 24, 2024, letter also requested an exemption from certain emergency core cooling system requirements to support the issuance of the amendments. By letter dated April 24, 2025 (ML25078A006), the U.S. Nuclear Regulatory Commission (NRC, or Commission) staff granted the licensees exemption request.
The licensees December 15, 2024, supplemental letter was in response to an NRC request for additional information dated November 21, 2024 (ML24326A148). The supplemental letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 1, 2024 (89 FR 79971).
2.0 REGULATORY EVALUATION
2.1 Background
The Westinghouse Electric Company (Westinghouse) manufactures both ZIRLO and Optimized ZIRLO' fuel cladding. According to the application, Optimized ZIRLO' was developed to meet the needs of longer operating cycles with increased fuel discharge burnup and fuel duty. Optimized ZIRLO' has a lower tin content and different microstructure than ZIRLO, which reduces the corrosion rate. Optimized ZIRLO' is described in the Westinghouse Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO' (Westinghouse Addendum 1-A).1 By letter dated June 10, 2005, the NRC staff found that this topical report is acceptable for referencing in licensing applications to the extent specified and under the conditions and limitations (henceforth referred to as conditions) delineated in the topical report and associated staff safety evaluation. The NRC staffs letter and safety evaluation with ten conditions on the approval of the topical report are included in Westinghouse Addendum 1-A.
2.2 Description of Proposed Changes TS 5.3.1, Fuel Assemblies, for Salem, Unit Nos. 1 and 2, currently states, in part, that each fuel assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods. The proposed amendment would revise this to state that each fuel assembly shall consist of a matrix of Zircaloy, ZIRLO or Optimized ZIRLO' clad fuel rods.
TS 6.9.1.9, Core Operating Limits Report (COLR), for Salem, Unit Nos. 1 and 2, provides a list of analytical methods that the licensee must use to calculate core operating limits for each reload cycle or portion of a reload cycle. The proposed amendment would add Westinghouse Addendum 1-A to the list of acceptable analytical methods in TS 6.9.1.9, which will enable the licensee to calculate core operating limits for fuel with Optimized ZIRLO' cladding material.
2.3 Regulatory Requirements and Guidance The regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.46, Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors, require, in part, that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents (LOCAs) conforms to the criteria set forth in 10 CFR 50.46(b). In addition, paragraph I.A.5, MetalWater Reaction Rate, of 10 CFR 50, Appendix K, requires the Baker-Just equation be used to calculate the rate of energy release, hydrogen generation, and cladding oxidation from the metal-water reaction in the core.
The regulations in 10 CFR 50.46 and the Baker-Just equation in 10 CFR Part 50, Appendix K, presume the use of either Zircaloy or ZIRLO fuel rod cladding. By letter dated April 24, 2025, the NRC staff granted the licensees request for an exemption from 10 CFR 50.46 and 10 CFR Part 50, Appendix K, to support the use of Optimized ZIRLO' fuel rod cladding in Salem, Unit Nos. 1 and 2. The exemption was limited to the cladding material such that all other requirements in 10 CFR 50.46 and 10 CFR Part 50, Appendix K, remain applicable to Salem, Unit Nos. 1 and 2.
On July 11, 1967, the Atomic Energy Commission published a revised and expanded set of 70 draft general design criteria (GDC) for public comment in the Federal Register (32 FR 10213). On February 20, 1971, the Atomic Energy Commission published in the Federal Register (36 FR 3255) a final rule that added Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. Differences between the 1967 draft GDC and the final GDC included a consolidation from 70 to 64 criteria. The LAR states that the Salem units were designed in accordance with the draft GDC, and the LAR discusses the relationship between 1 Transmittal letter and non-proprietary version are available under ADAMS Accession Nos. ML062080563 and ML062080569, respectively.
the relevant draft and final GDC. Although the licensee is not required to demonstrate compliance with the final GDC, the final GDC are an acceptable standard for evaluating fuel cladding design changes for the Salem units. The NRC staff considered the following GDC in 10 CFR Part 50, Appendix A, to be relevant to the LAR:
GDC 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any conditions of normal operation, including the effects of anticipated operational occurrences.
GDC 27, Combined reactivity control systems capability, requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
GDC 35, Emergency core cooling, insofar as it requires that a system to provide abundant emergency core cooling following a LOCA to prevent fuel and cladding damage that could interfere with effective core cooling and limit the metal-water reaction on the fuel cladding to negligible amount.
Section 4.2, Revision 3, Fuel System Design (ML070740002), of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR
[Light-Water Reactor] Edition, provides regulatory guidance to the NRC staff for the review of fuel system design, which includes fuel rod cladding. Section 4.2 states, in part, that the fuel system safety review provides assurance that (1) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs), (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained.
The regulations in 10 CFR 50.36, Technical specifications, establish the regulatory requirements related to the content of TSs. This regulation requires that TSs include: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls. In accordance with 10 CFR 50.36(c)(4), the design features to be included in the TSs are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories (1), (2), or (3). In accordance with 10 CFR 50.36(c)(5), the TSs must include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
3.0 TECHNICAL EVALUATION
By letter dated June 10, 2005, the NRC staff found that Westinghouse Addendum 1-A is acceptable for referencing in licensing applications to the extent specified and under the conditions delineated in the topical report and associated staff safety evaluation. The NRC staffs safety evaluation for Westinghouse Addendum 1-A concluded, in part, that the ECCS performance criteria in 10 CFR 50.46 and the Baker-Just correlation in 10 CFR Part 50, Appendix K, are applicable to Optimized ZIRLO'. The NRC staffs safety evaluation considered GDC 10, 27, and 35, and states:
The NRC staff reviewed the effects of Optimized ZIRLO' using the appropriate fuel design requirements of [NUREG-0800 Section 4.2] and 10 CFR Part 50, Appendix A, General Design Criteria and found that the [Topical Report] provided reasonable assurance that under both normal and accident conditions, Westinghouse fuel assembly designs implementing Optimized ZIRLO' fuel cladding would be able to safely operate and comply with NRC regulations.
This generic conclusion is applicable to Salem, Unit Nos. 1 and 2, as the licensee is requesting to use Westinghouse fuel assembly designs with Optimized ZIRLO' fuel rod cladding.
Therefore, the NRC staffs review of this LAR focused on the proposed TS changes and how the licensee addressed each of these conditions.
3.1 Condition 1: Exemption from Regulation Condition 1 states that implementation of Optimized ZIRLO' fuel clad requires an exemption from 10 CFR 50.46 and 10 CFR Part 50, Appendix K. As previously noted, the NRC staff granted the licensees request for an exemption from 10 CFR 50.46 and 10 CFR Part 50, Appendix K, for Salem, Unit Nos. 1 and 2. Therefore, the NRC staff determined that Condition 1 was met.
3.2 Condition 2: Burnup Limit Condition 2 in relevant part limits the NRC staffs approval of Westinghouse Addendum 1-A to fuel rod burnup limits of 62 gigawatt days per metric ton of uranium (GWd/MTU) for Westinghouse fuel designs.
The licensee stated:
Salem Units 1 and 2 use Westinghouse fuel designs and the current design basis rod burnup limit is 62 GWd/MTU. For any fuel using Optimized ZIRLO fuel rod cladding, the maximum fuel rod burnup limit for Westinghouse fuel designs will continue to be 62 GWd/MTU until such time that a new fuel rod burnup limit is approved for use. The fuel burnup limit will be confirmed as part of the normal reload design process.
The NRC staff determined that Condition 2 was met because the licensee uses Westinghouse fuel and the current design basis for Salem, Unit Nos. 1 and 2, limits fuel rod burnup to 62 GWd/MTU.
3.3 Condition 3: Corrosion Limits Condition 3 states that the maximum fuel rod waterside corrosion, as predicted by the best-estimate model, will meet specified proprietary limits of hydrides for all locations of the fuel rod.
The licensee stated:
The maximum fuel rod waterside corrosion for fuel using Optimized ZIRLO fuel rod cladding will be confirmed to be less than the specified proprietary limits for all locations of the fuel rod. Evaluations will be performed to confirm that the appropriate corrosion limits are satisfied as part of the normal reload design process.
The NRC staff determined that Condition 3 was met because the licensee will confirm that the maximum fuel rod waterside corrosion limits are met as part of the normal reload design process. This is assured because TS 6.9.1.9 requires the licensee to calculate core operating limits for each reload cycle using NRC-approved methods, which will include Westinghouse Addendum 1-A with the proposed amendments.
3.4 Condition 4: Use of Approved Methodologies Condition 4 states that all the conditions listed in previous NRC safety evaluation approvals for methodologies used for standard ZIRLO' and Zircaloy-4 fuel analysis will continue to be met, except that the use of Optimized ZIRLO' cladding in addition to standard ZIRLO' and Zircaloy-4 cladding is now approved.
The NRC staff determined that Condition 4 was met because the licensee stated that the Optimized ZIRLO fuel rod analysis will continue to meet all conditions associated with approved methods. This is assured because TS 6.9.1.9 requires the licensee to calculate core operating limits for each reload cycle using NRC-approved methods, which will include Westinghouse Addendum 1-A with the proposed amendments.
3.5 Condition 5: Use in the Applicable Data Range Condition 5 states that all methodologies will be used only within the range for which ZIRLO and Optimized ZIRLO' data were acceptable and for which the verifications discussed in Westinghouse Addendum 1-A and responses to requests for additional information were performed.
The NRC staff determined that Condition 5 was met because the licensee stated that (1) the application of ZIRLO and Optimized ZIRLO' in approved methodologies will be made consistent with the approach in Westinghouse Addendum 1-A and (2) confirmation of these conditions is required as part of the normal reload design process. This is assured because TS 6.9.1.9 requires the licensee to calculate core operating limits for each reload cycle using NRC-approved methods, which will include Westinghouse Addendum 1-A with the proposed amendments.
3.6 Condition 6 - Lead Test Assembly Data Condition 6 required Westinghouse to provide information to the NRC regarding Optimized ZIRLO' lead test assembly data and confirm applicability of the approved fuel performance models. By letter dated August 3, 2016 (ML16173A354), the NRC staff informed Westinghouse that condition 6 had been satisfied and licensees no longer needed to provide additional data to adopt Westinghouse Addendum 1-A.
3.7 Condition 7: Cycle Data Condition 7 required Westinghouse to provide information to the NRC regarding Optimized ZIRLO' growth and creep data and confirm applicability of the approved fuel performance models. By letter dated August 3, 2016, the NRC staff informed Westinghouse that condition 7 had been satisfied and licensees no longer needed to provide additional data to adopt Westinghouse Addendum 1-A.
3.8 Condition 8: Yield Strength Condition 8 states that the licensee shall account for the relative differences in unirradiated strength between Optimized ZIRLO' and standard ZIRLO in cladding and structural analyses until irradiated data for Optimized ZIRLO' has been collected and provided to the NRC staff.
The licensee stated that the Optimized ZIRLO' fuel rod analysis for Salem will use the yield strength and ultimate tensile strength as modified by the specifications in Condition 8 for Westinghouse fuel designs. The NRC staff determined that Condition 8 was met because (1) Salem uses Westinghouse fuel and (2) the licensee stated that confirmation of this condition is required as part of the reload design process. This is assured because TS 6.9.1.9 requires the licensee to calculate core operating limits for each reload cycle using NRC-approved methods, which will include Westinghouse Addendum 1-A with the proposed amendments.
3.9 Condition 9: LOCBART or STRIKIN-II Early Peak Cladding Temperature Condition 9 states that for plants licensed with the LOCBART or STRIKIN-II computer codes and a limiting peak cladding temperature that occurs during blowdown or early reflood, the limiting LOCBART or STRIKIN-II calculation will be rerun using the specified Optimized ZIRLO' material properties.
The licensee stated that the Salem units are licensed with the LOCBART computer code, but not the STRIKIN-II computer code. However, the licensee also stated that the peak cladding temperature does not occur during the blowdown or early reflood portion of the transient. Based on this information, the NRC staff determined that Condition 9 is not applicable to the Salem units.
3.10 Condition 10: Peak Cladding Temperature Limit for Locked Rotor Event Condition 10 states that due to the absence of high temperature oxidation data for Optimized ZIRLO', the Westinghouse coolability limit on peak cladding temperature during the locked rotor event shall be a specified proprietary value.
The NRC staff determined that Condition 10 was met because the licensee stated that confirmation of this condition is required as part of the normal reload design process. This is assured because TS 6.9.1.9 requires the licensee to calculate core operating limits for each reload cycle using NRC-approved methods, which will include Westinghouse Addendum 1-A with the proposed amendments.
3.11 Technical Conclusion By letter dated June 10, 2005, the NRC staff found that Westinghouse Addendum 1-A is acceptable for referencing in licensing applications to the extent specified and under the conditions delineated in the topical report and associated staff safety evaluation. As discussed in Sections 3.1 through 3.10, the NRC staff determined that each of the conditions was either met or not applicable to Salem, Unit Nos. 1 and 2. Therefore, the NRC staff concludes that the use of Westinghouse Addendum 1-A and Optimized ZIRLOTM fuel rod cladding at Salem, Unit Nos. 1 and 2, is acceptable.
3.12 TS Changes The licensee proposed to add Optimized ZIRLO' clad fuel rods to the list of allowed cladding materials in TS 5.3.1 and make other editorial changes. The NRC staff evaluated the proposed changes to TS 5.3.1 and determined that 10 CFR 50.36(c)(4) will continue to be met because TS 5.3.1 will specify the allowed fuel rod cladding materials, which, if altered or modified, would have a significant effect on safety.
The licensee proposed to add Westinghouse Addendum 1-A to the list of acceptable analytical methods in TS 6.9.1.9, which will enable the licensee to calculate core operating limits for fuel with Optimized ZIRLO' cladding material. The NRC staff evaluated the proposed changes to TS 6.9.1.9 and determined that 10 CFR 50.36(c)(5) will continue to be met because TS 6.9.1.9 will continue to include administrative controls necessary to assure operation of the facility in a safe manner.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendments on March 14, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (89 FR 79971; October 1, 2024). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Summer Sun, NRR Blake Purnell, NRR Date: April 24, 2025
- via e-mail NRR-058 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL1/LA NRR/DSS/SNSB/BC NAME BPurnell KEntz DMurdock DATE 3/21/2025 3/17/2025 2/24/2025 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL1/BC NAME SMehta KBernstein HGonzález DATE 3/21/2025 4/4/2025 4/11/2025 OFFICE NRR/DORL/LPL2-2/PM NAME BPurnell DATE 4/24/2025