ML25023A278
| ML25023A278 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/07/2025 |
| From: | Scott Wall Plant Licensing Branch III |
| To: | Rhoades D Constellation Energy Generation |
| Wiebe J | |
| References | |
| EPID L-2023-LLA-0136 | |
| Download: ML25023A278 (39) | |
Text
March 7, 2025 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 237, 237, 236, AND 236 REGARDING SPENT FUEL POOL CRITICALITY ANALYSIS (EPID L-2023-LLA-0136)
Dear David Rhoades:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 237 to Renewed Facility Operating License No. NPF-72 and Amendment No. 237 to Renewed Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2, respectively, and Amendment No. 236 to Renewed Facility Operating License No. NPF-37 and Amendment No. 236 to Renewed Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2, respectively. The amendments are in response to your application dated September 29, 2023, as supplemented by letters dated October 23, 2024, and January 15, 2025.
The amendments make changes to spent fuel pool boron concentration and the spent fuel pool criticality analysis to include fuel from Framatome and Westinghouse.
A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Scott P. Wall, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454, and STN 50-455
Enclosures:
- 1. Amendment No. 237 to NPF-72
- 2. Amendment No. 237 to NPF-77
- 3. Amendment No. 236 to NPF-37
- 4. Amendment No. 236 to NPF-66
- 5. Safety Evaluation cc: Listserv CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 Renewed License No. NPF-72
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated September 29, 2023, as supplemented by letters dated October 23, 2024, and January 15, 2025, 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 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; 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 renewed 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. NPF-72 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.03.07 11:44:50 -05'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 Renewed License No. NPF-77
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated September 29, 2023, as supplemented by letters dated October 23, 2024, and January 15, 2025, 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 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; 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 renewed 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. NPF-77 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.03.07 11:45:24 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 237 AND 237 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating Licenses REMOVE INSERT License NPF-72 License NPF-72 License NPF-77 License NPF-77 Technical Specifications REMOVE INSERT 3.7.15 - 1 3.7.15 - 1 3.7.16 - 1 3.7.16 - 1 3.7.16 - 2 3.7.16 - 2 3.7.16 - 3 3.7.16 - 3 4.0 - 2 4.0 - 2
(2)
Constellation Energy Generation, 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; (3)
Constellation Energy Generation, 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; (4)
Constellation Energy Generation, 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 (5)
Constellation Energy Generation, 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Renewed License No. NPF-72 Amendment No. 237
(2)
Constellation Energy Generation, 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; (3)
Constellation Energy Generation, 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; (4)
Constellation Energy Generation, 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 (5)
Constellation Energy Generation, 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.
The 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 237 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Renewed License No. NPF-77 Amendment No. 237
I Spent Fuel Pool Boron Concentration 3.7.15 BRAIDWOOD UNITS 1 & 2 3.7.15 1 Amendment 237 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be 2000 ppm.
APPLICABILITY:
Whenever fuel assemblies are stored in the spent fuel pool.
ACTIONS
NOTE-------------------------------------
LCO 3.0.3 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Spent fuel pool boron concentration not within limit.
A.1 Suspend movement of fuel assemblies in the spent fuel pool.
AND A.2 Initiate action to restore spent fuel pool boron concentration to within limit.
Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit.
In accordance with the Surveillance Frequency Control Program
I I
Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 1 Amendment 237 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 Each spent fuel assembly stored in the spent fuel pool shall, as applicable:
a.
Region 1 of spent fuel pool storage racks Have an initial nominal enrichment of 5.0 weight percent U-235 to permit storage in any cell location.
b.
Region 2 of spent fuel pool storage racks Have a combination of initial enrichment and burnup within the Acceptable Burnup Domain of Figure 3.7.16-1.
APPLICABILITY:
Whenever fuel assemblies are stored in the spent fuel pool.
ACTIONS
NOTE-------------------------------------
LCO 3.0.3 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the LCO not met.
A.1 Initiate action to move the noncomplying fuel assembly into a location which restores compliance.
Immediately
Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 2 Amendment 237 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial nominal enrichment of the fuel assembly is 5.0 weight percent U-235.
Prior to storing the fuel assembly in Region 1 SR 3.7.16.2 Verify by administrative means the combination of initial enrichment and burnup, as applicable, of the fuel assembly is within the Acceptable Burnup Domain of Figure 3.7.16-1.
Prior to storing the fuel assembly in Region 2
Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment curve Loadi ng Curve 50 45 40 tao 35 30 r i 25
" 20 7() 15 10 5
0 2
ACCEPTABLE BURNUP DOMAIN UNACCEPTABLE BURNUP DOMAIN 2.5 3
3.5 4
Initial Fuel Enrichment (wt% U-235) 4.5 5
Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 3 Amendment 237 Figure 3.7.16-1 (page 1 of 1)
Region 2 Fuel Assembly Burnup Requirements (Spent Fuel Pool Storage Racks)
I Design Features 4.0 BRAIDWOOD UNITS 1 & 2 4.0 2 Amendment 237 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality The spent fuel storage racks are designed and shall be maintained, as applicable, with:
a.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; b.
keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.
c.
A nominal 10.888 inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and d.
A nominal 8.97 inch center to center distance between fuel assemblies placed in Region 2 racks.
4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.
4.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 236 Renewed License No. NPF-37
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated September 29, 2023, as supplemented by letters dated October 23, 2024, and January 15, 2025, 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 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; 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 renewed 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. NPF-37 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 236 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.03.07 11:46:18 -05'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 236 Renewed License No. NPF-66
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Constellation Energy Generation, LLC (the licensee) dated September 29, 2023, as supplemented by letters dated October 23, 2024, and January 15, 2025, 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 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; 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. NPF-66 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 236, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 7, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.03.07 11:46:44 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 236 AND 236 RENEWED FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating Licenses REMOVE INSERT License NPF-37 License NPF-37 License NPF-66 License NPF-66 Technical Specifications REMOVE INSERT 3.7.15 - 1 3.7.15 - 1 3.7.15 - 2 3.7.16 - 3 3.7.16 - 3 4.0 - 2 4.0 - 2
(2)
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 Updated Final Safety Analysis Report, as supplemented and amended; (3)
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; (4)
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 (5)
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.
The 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 236 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Deleted.
(4)
Deleted.
Renewed License No. NPF-37 Amendment No. 236
(2)
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 Updated Final Safety Analysis Report, as supplemented and amended; (3)
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; (4)
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 (5)
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.
The 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 236, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Deleted.
Renewed License No. NPF-66 Amendment No. 236
I Spent Fuel Pool Boron Concentration 3.7.15 BYRON UNITS 1 & 2 3.7.15 1 Amendment 236 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be 2000 ppm.
APPLICABILITY:
Whenever fuel assemblies are stored in the spent fuel pool.
ACTIONS
NOTE-------------------------------------
LCO 3.0.3 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Spent fuel pool boron concentration not within limit.
A.1 Suspend movement of fuel assemblies in the spent fuel pool.
AND A.2 Initiate action to restore spent fuel pool boron concentration to within limit.
Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit.
In accordance with the Surveillance Frequency Control Program
Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment curve Loadi ng Curve 50 45 40 tao 35 30 r i 25
" 20 7() 15 10 5
0 2
ACCEPTABLE BURNUP DOMAIN UNACCEPTABLE BURNUP DOMAIN 2.5 3
3.5 4
Initial Fuel Enrichment (wt% U-235) 4.5 Spent Fuel Assembly Storage 3.7.16 BYRON UNITS 1 & 2 3.7.16 3 Amendment 236 Figure 3.7.16-1 (page 1 of 1)
Region 2 Fuel Assembly Burnup Requirements
Design Features 4.0 BYRON UNITS 1 & 2 4.0 2 Amendment 236 4.0 DESIGN FEATURES 4.2 Reactor Core (continued) 4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium, hafnium, or a mixture of both types.
4.3 Fuel Storage 4.3.1 Criticality The spent fuel storage racks are designed and shall be maintained, as applicable, with:
a.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; b.
keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.
c.
A nominal 10.888 inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and d.
A nominal 8.97 inch center to center distance between fuel assemblies placed in Region 2 racks.
4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.
4.3.3 Capacity The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 236 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AND AMENDMENT NO. 236 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 CONSTELLATION ENERGY GENERATION, LLC DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455
1.0 INTRODUCTION
By letter dated September 29, 2023 (Agencywide Documents and Management System (ADAMS) Accession No. ML23272A201), as supplemented by letters dated October 23, 2024 (ML24297A657), and January 15, 2025 (ML25015A220), Constellation Energy Generation LLC (CEG, the licensee) requested amendments to the technical specifications (TS) for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Unit Nos. 1 and 2 (Byron). In its letters, CEG described its criticality safety evaluation (SE) for fuel assembly to support the requested TS change.
The criticality SE CEG performed included a near-term fuel design change to Framatomes GAIA fuel, a future transition to 24-month operating cycles concurrent with a fuel design change to Framatomes Advanced Fuel Management (AFM-GAIA) 17x17, and increased fuel enrichment to 6.5 weight percent (w/%) in Uranium-235 (U-235). However, CEGs September 29, 2023, letter, did not request the necessary exemption to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.68(b)(7) or TS changes to implement fuel enriched to 6.5 w/% U-235.
CEGs September 29, 2023, letter, requested the following TS changes:
TS 3.7.15, Spent Fuel Pool Boron Concentration to increase the required spent fuel pool (SFP) boron concentration to be > 2000 ppm.
TS 3.7.16, Spent Fuel Assembly Storage to update Figure 3.7.16-1 to include fuel from Framatome and Westinghouse.
TS 4.3.1.b, Fuel Storage, Criticality replaced with keff1 < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the 1 Keff and K-effective are synonymous.
Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.
TS 4.3.1.c and d, Fuel Storage, Criticality replace For Holtec spent fuel pool storage racks, a with A. (Braidwood only)
By letter dated October 23, 2024, CEG revised the requested amendments to the TS for Braidwood and Byron to only cover the near-term fuel design change to Framatomes GAIA fuel.
The criticality SE supporting the request was revised to only address the near-term transition to Framatomes GAIA fuel. Removed from consideration were the future transition to 24-month operating cycles, future transition to AFM-GAIA 17x17, or increased fuel enrichment to 6.5 w/%
in U-235.
CEGs October 23, 2024 letter, requested the following revised TS changes:
TS 3.7.15, Spent Fuel Pool Boron Concentration to increase the required SFP boron concentration to be > 2000 ppm. (No change)
TS 3.7.16, Spent Fuel Assembly Storage to update Figure 3.7.16-1 to include fuel from Framatome and Westinghouse. (Revised change)
TS 3.7.16, Spent Fuel Assembly Storage Limiting Condition for Operation (LCO) a.
and b. are revised to remove Holtec from the first line. (Braidwood only) (Revised change)
SR 3.7.16.2 is revised to remove decay time. (Braidwood only) (Revised change)
TS 4.3.1.b, Fuel Storage, Criticality replaced with keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the UFSAR and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR. (No change)
TS 4.3.1.c and d, Fuel Storage, Criticality replace For Holtec spent fuel pool storage racks, a with A. (Braidwood only) (No change)
CEGs January 15, 2025 letter revised its implementation schedule for the amendments. CEGs letter dated September 29, 2023, as supplemented by its letters dated October 23, 2024, and January 15, 2025, are collectively referred to as the license amendment request (LAR). The supplements did not change the scope as described in the notice of consideration in the Federal Register dated November 28, 2023 (88 FR 83167).
2.0 REGULATORY EVALUATION
In accordance with the licensees amendment request, the regulatory requirements and guidance which the U.S. Nuclear Regulatory Commission (NRC, the Commission) staff considered in assessing the proposed TS change are as follows:
Section 182a of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commissions regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, Technical specifications. The TS requirements in 10 CFR 50.36 include items in the following categories: (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. The requirements for system operability during movement of irradiated fuel are included in the TSs in accordance with 10 CFR 50.36(c)(2), Limiting conditions for operation. As required by 10 CFR 50.36(c)(4), [d]esign features to be included 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 described in paragraphs (c) (1), (2), and (3) of [10 CFR 50.36].
The applicable regulatory requirements for criticality safety analysis for SFPs are contained in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 62, Prevention of criticality in fuel storage and handling, and in 10 CFR 50.68, Criticality accident requirements.
The regulations in 10 CFR Part 50, Appendix A, Criterion 62 state that:
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
The regulations in 10 CFR 50.68(b)(2) state, in part, that:
The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
The regulations in 10 CFR 50.68(b)(3) state, in part, that:
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.
The regulations in 10 CFR 50.68(b)(4) state, in part, that:
If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.
In accordance with 10 CFR 50.92(a), when determining whether to issue a license amendment, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. In determining whether the proposed TS remedial actions should be granted, the staff typically applies the reasonable assurance standard derived from the requirements of 10 CFR 50.40(a) and 50.57(a)(3). The regulation at 10 CFR 50.40(a) states that in determining whether to grant the licensing request, the Commission will be guided by, among other things, consideration about whether the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in Part 20 of this chapter, and that the health and safety of the public will not be endangered. The regulation at 10 CFR 50.57(a)(3) states that the Commission may issue an operating license when, in part, there is reasonable assurance that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and such activities will be conducted in compliance with the regulations in this chapter.
In March of 2021, the NRC staff issued Regulatory Guide (RG) 1.240, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127). RG 1.240 describes an approach that the NRC staff considers acceptable to demonstrate that applicable regulatory requirements are met for the subcriticality of fuel assemblies stored in fresh fuel vaults and SFPs at light-water reactor (LWR) power plants. RG 1.240 also endorses, with clarifications and exceptions, the Nuclear Energy Institute (NEI) guidance document NEI 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants (ML19269E069). NEI 12-16, Revision 4, is a comprehensive guide that compiles previously issued NRC guidance and clarification letters regarding performing SFP and fresh fuel storage criticality analyses. The NRC staff primarily uses RG 1.240 in conjunction with NEI 12-16, Revision 4, in its review.
In January of 2001, the NRC staff issued NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061). NUREG/CR-6698 provides guidance on the validation and bias and uncertainty quantification of calculational tools used to perform nuclear criticality safety (NCS) analyses. Guidance is also provided for establishing upper safety limits on subcriticality. The NRC staff primarily used this document to evaluate the determined the acceptability of applying MCNP6.2 and CASMO5 codes in the criticality analysis of the Braidwood and Byron SFPs.
Additional guidance is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, particularly Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 2 (ML052350536).
Section 9.1.1 provides the existing recommendations for performing the review of the NCS analysis of SFPs.
3.0 TECHNICAL EVALUATION
3.1 Background
The licensees letter dated October 23, 2024, revised the LAR to solely reflect the fuel transition from Westinghouse Vantage 5, Vantage 5+, and Optimized Fuel Assembly (OFA) to Framatome GAIA fuel assemblies. The Framatome GAIA fuel assemblies are designed for Westinghouse-type plants with a 17x17 fuel rod array. The cladding material is M5, a Framatome advanced zirconium alloy. The design utilizes a 144-inch fuel stack length of UO2 or UO2-Gd2O3 (Gadolinia).
The licensee provided a short description of the new fuel vaults (NFVs) in section 2.2.1 of (non-proprietary) and Attachment 8 (proprietary) (hereafter, referred to as
/8) of its letter dated September 29, 2023. The NFVs are dry storage of new fuel in three rows of 44 storage cells. The licensee provided a short description of the SFP in section 2.2.2 of Attachment 1/8 of its letter dated September 29, 2023. The SFPs are divided into two regions. Region 1 is for storage of both fresh fuel and burned fuel and has a capacity of 396 cells. Region 1 is a flux-trap configuration and uses Boral as a permanently installed neutron absorbing material (NAM). Region 2 is for storage of burned fuel only and nonburned fuel is not authorized for storage in Region 2. Region 2 has a storage capacity of 2588 cells at Byron and a storage capacity of 2568 cells at Braidwood. Region 2 does not have a flux-trap configuration but also uses Boral as a permanently installed NAM.
3.2 Technical Specifications TS 3.7.15, Spent Fuel Pool Boron Concentration to increase the required SFP boron concentration to be > 2000 ppm.
This TS supports the licensees analysis in its LAR and the NRC staffs review documented in Section 3.4.5 of this SE. The SFP boron concentration is an operating restriction that is an initial condition of a criticality transient in the SFP.
Based on its review, the NRC staff concludes that the TS meets 10 CFR 50.36(c)(2)(ii)(B), and is acceptable.
TS 3.7.16, Spent Fuel Assembly Storage to update Figure 3.7.16-1 to include fuel from Framatome and Westinghouse.
This TS supports the licensees analysis in its LAR and the NRC staffs review documented in Sections 3.4.2 through 3.4.4 of this SE. The combination of enrichment and burnup depicted on TS Figure 3.7.16-1 is an initial condition of a criticality transient in the SFP. Based on its review, the NRC staff concludes that the TS meets 10 CFR 50.36(c)(1), and is acceptable.
TS 3.7.16, Spent Fuel Assembly Storage Limiting Condition for Operation (LCO) a.
and b. are revised to remove Holtec from the first line. SR 3.7.16.2 is revised to remove decay time. (Braidwood only)
These changes to the TS are editorial in nature and do not change the intent of the TS. Based on the above the NRC staff concludes that they are acceptable.
TS 4.3.1.b, Fuel Storage, Criticality replaced with keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the UFSAR and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.
This TS supports the licensees analysis in its LAR and the NRC staffs review documented in Section 3.4.1 of this SE. The requirements in this TS are an initial condition of a criticality transient in the SFP. Based on its review, the NRC staff concludes that the TS meets 10 CFR 50.36(c)(1), and is acceptable.
TS 4.3.1.c and d, Fuel Storage, Criticality replace For Holtec spent fuel pool storage racks, a with A. (Braidwood only)
This change to the TS is editorial in nature and does not change the intent of the TS.
Based on the above the NRC staff concludes that it is acceptable.
3.3 Method of Review In order for the LAR to be acceptable, the licensee must submit a plant-specific SFP criticality analysis that includes technically supported margins. The NRC staff reviewed the analysis to ensure that the assumptions made are technically substantiated. The NRC staff reviewed the application and supplemental information to determine whether the submittals provide reasonable assurance that the regulatory requirements will continue to be met. As discussed below, the NRC staff finds that the licensee provided the technical information needed for the NRC staff to complete its review of the LAR.
This SE involves a review of the licensees NCS analyses for the Byron and Braidwood NFVs and SFP. A summary of the licensees NCS analyses was provided in Attachment 1/8 to the licensees September 29, 2023 LAR and updated through the licensees letter dated October 23, 2024. The review was performed consistent with NUREG-0800, section 9.1.1, as well as RG 1.240.
3.4 SFP NCS Analysis Review 3.4.1 SFP NCS Analysis Method There is no generic or standard NRC-approved methodology for performing NCS analyses for fuel storage and handling by commercial power reactor licensees. Each analysis is unique to its analyzed system. An overview of the methods used for the licensees NCS used to support the LAR was provided in Attachment 1/8 to the licensees letter dated September 29, 2023, and updated by the licensees letter dated October 23, 2024.
3.4.1.1 Computational Methods The licensees SFP Region 1 storage racks NCS analysis does not consider the decrease in fuel reactivity typically seen in pressurized-water reactors (PWRs) as the fuel is depleted during reactor operation. Region 1 is for the storage of fresh fuel before it is irradiated or for irradiated fuel that does not meet the requirements for storage in Region 2. Validation of the depletion code does not affect this portion of the analysis, but validation of the computer code used to model the fuel in the Region 1 storage rack is necessary. The same validation provided for the computer code used to model the fuel is applicable to both the Region 1 and Region 2 storage racks in this analysis.
The licensees SFP Region 2 storage racks NCS analysis considers the decrease in fuel reactivity typically seen in PWRs as the fuel is depleted during reactor operation. This approach is frequently used in PWR NCS analyses and is sometimes referred to as burnup credit (BUC).
BUC NCS requires a two-step process. The first step relates to depletion where a computer code simulates the reactor operation to calculate the changes in the fuel composition of the fuel assembly. The second step is a modeling of the depleted fuel assembly in the SFP storage racks and the determination of the system Keff. The validation of the computer codes in each step is a significant portion of the analysis. Since the licensees NCS analysis credits fuel burnup, it is necessary for the NRC to consider validation of the computer codes and data used to calculate burned fuel compositions, and the computer code and data that utilize the burned fuel compositions to calculate Keff for systems with burned fuel. This portion of the NCS analysis establishes a burnup/enrichment loading curve that fuel assemblies must meet to be safely stored in the Bryon and Braidwood SFP Region 2 storage racks.
3.4.1.1.1 Depletion Computer Code Validation In its LAR, the licensee states that CASMO5 with ENDF/B-VII.1 cross-section library was used for fuel depletion calculations to determine the isotopic composition of spent fuel for Region 2 criticality analysis. CASMO5 is a two-dimensional multigroup transport theory code for burnup calculations of [boiling water reactor (BWR)] and PWR fuel assemblies. The NRC reviewed and approved the CASMO5 code in a letter dated September 15, 2017 (ML17236A419), and has accepted the use of CASMO5 as the depletion code for NCS analyses in the past (e.g., NRC letter dated May 10, 2023 (ML23093A095)). The licensees usage of CASMO5 is therefore acceptable.
3.4.1.1.2 SFP Keff Computer Code Validation In its LAR, the licensee states that, All racks are analyzed using the MCNP6.2 Monte Carlo neutron transport program and ENDF/BVII.1 cross-section library. The Monte Carlo N-Particle (MCNP) is developed and maintained by the Los Alamos National Laboratory. According to Los Alamos National Laboratory, The MCNP, Monte Carlo N-Particle, code can be used for general-purpose transport of many particles including neutrons, photons, electrons, ions, and many other elementary particles, up to 1 TeV/nucleon. The transport of these particles is through a three-dimensional representation of materials defined in a constructive solid geometry, bounded by first-, second-, and fourth-degree user-defined surfaces. In addition, external structured and unstructured meshes can be used to define the problem geometry in a hybrid mode by embedding a mesh within a constructive solid geometry cell, providing an alternate path to defining complex geometry.
Validation of MCNP6.2 was performed by benchmarking it against a variety of established criticality experiments and comparing the effective multiplication factor generated by MCNP6.2 to that measured from the critical experiments. In its LAR, the licensee states, The validation methodology was based on recommendations contained in NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculation Methodology. Using NUREG/CR-6698 is consistent with the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0. The information provided by the licensee in its LAR demonstrates that MCNP6.2 was properly validated, therefore, the NRC staff finds use of MCNP6.2 for this purpose to be acceptable.
3.4.2 SFP and Fuel Storage Racks 3.4.2.1 SFP Water Temperature Guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 states the NCS analysis should be done at the temperature corresponding to the highest reactivity. The licensees analysis performed sensitivity analyses on SFP water temperature and determined the highest reactivity is at 124 °C. The licensees analysis applied a bias when calculations were performed at a different temperature. The water densities used were adjusted consistent with the water temperatures being modeled. The effects of water temperature variation on the maximum Keff value were included as bias terms for all regions both with and without soluble boron. Therefore, the NRC staff finds that the licensees criticality analysis was done at the water temperature and density corresponding to the highest reactivity, and is therefore acceptable.
3.4.2.2 SFP Storage Rack Models The SFPs are divided into two regions. Region 1 is for storage of both fresh fuel and burned fuel and has a capacity of 396 cells. Region 1 is a flux-trap configuration and uses Boral as a permanently installed neutron absorbing material (NAM). Region 2 is for storage of burned fuel only and nonburned fuel is not authorized for storage in Region 2. Region 2 does not have a flux-trap configuration but also uses Boral as a permanently installed NAM. The licensee provided additional details of its SFP racks and models in response to Request for Additional Information (RAI) No. 4 in its October 23, 2024 letter.
3.4.2.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties Guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 states that flux trap size, cell inner dimensions/storage location pitch, storage cell wall thickness, rack and insert neutron absorber dimensions, and neutron absorber sheathing thickness should be considered when evaluating the uncertainties due to tolerances. The licensee performed uncertainty analysis for the following rack dimensions: box inside diameter (ID) (flux trap size), box wall thickness, poison width, poison thickness, cell pitch, and poison degradation. The licensees treatment of SFP storage rack manufacturing tolerance and uncertainties includes those recommended by the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 and, therefore, is acceptable.
3.4.3 Fuel Assembly 3.4.3.1 Fuel Assembly Design The fuel assembly design is the Framatome GAIA design. The new burnup/enrichment loading curve (TS figure 3.7.16-1) is more restrictive than the current loading curve and, therefore, will bound existing fuel designs.
3.4.3.1.1 Fuel Assembly Physical Changes with Depletion Guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 states that changes in fuel rods and material dependent grid growth need to be considered with depletion. In its LAR, the licensee included a fuel depletion-related geometry change bias that included fuel rod growth and cladding creep and grid growth. Therefore, the NRC staff finds the bias to be appropriate for this application.
3.4.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties In its LAR, the licensee included uncertainties for the following fuel assembly manufacturing tolerances: fuel rod pitch, fuel pellet diameter, fuel cladding inside ID, fuel cladding outside diameter (OD), guide tube ID, guide tube OD, fuel enrichment, and fuel pellet density. The licensees treatment of fuel assembly manufacturing tolerance and uncertainties includes the parameters recommended by the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 and, therefore, is acceptable.
3.4.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on U-235 enrichment and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also apply to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is complex. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: depletion uncertainty; the axial and radial apportionment of the burnup; and the core operation that achieved that burnup. These characteristics are evaluated in the following sections.
3.4.3.3.1 Depletion Uncertainty Guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 states that an uncertainty equal to 5 percent may be used to estimate the uncertainty associated with fuel depletion calculations.
Since the licensee accounted for reactivity uncertainty with 5 percent of the delta k between fresh and burnup fuel of interest as the depletion uncertainty, the NRC staff concludes that the depletion analyses are consistent with the guidance set forth in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, Revision 0. The NRC staff therefore finds that the depletion analyses are acceptable.
3.4.3.3.2 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to neutron leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis (ML031110292), has shown that, at assembly burnups above about 10 to 20 gigawatt-days per metric ton of uranium (GWd/MTU), the use of a uniform axial burnup profile results in an under-prediction of Keff; generally, the under-prediction becomes larger as burnup increases. This is what is known as the end effect. Proper selection of the axial burnup profile is necessary to ensure Keff is not under-predicted due to the end effect.
The licensee used NUREG/CR-6801 to model the axial burnup of the fuel assemblies. That is consistent with the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0. Consequently, the NRC staff finds that the treatment of axial burnup distribution by the licensee is acceptable.
3.4.3.3.3 Radial Burnup Distribution Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (i.e., differences in burnup between portions or quadrants of the cross section of the assembly). The licensees analysis did not consider the effect of planar burnup distribution on reactivity. NUREG/CR-6800, Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs (ML031110280), estimates the effect of a radial burnup profile. NUREG/CR-6800 estimates the effect to be approximately 0.002 k. The licensees analysis has sufficient margin to accommodate the potential radial burnup distribution. The NRC staff finds the omission of the planar burnup distribution on reactivity to be acceptable because the expected reactivity impact is accommodated by the available margin in the licensees analysis.
3.4.3.3.4 Burnup History/Core Operating Parameters NUREG/CR-6665, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel (ML003688150), provides some discussion on the treatment of depletion analysis parameters that determine how the burnup was achieved. While NUREG/CR-6665 is focused on NCS analysis in storage and transportation casks, the basic principles with respect to the depletion analysis apply generically, since the phenomena occur in the reactor as the fuel is being depleted. The results have some applicability to the licensees criticality safety analysis.
The basic strategy for this type of analysis is to select parameters that maximize the Doppler broadening/spectral hardening of the neutron field resulting in maximum plutonium-239/241 production. NUREG/CR-6665 discusses six parameters affecting the depletion analysis: fuel temperature, moderator temperature, soluble boron, specific power, and operating history, fixed burnable poisons, and integral burnable poisons. While the mechanism for each is different, the effect is similar: Doppler broadening/spectral hardening of the neutron field resulting in increased plutonium-239/241 production. NUREG/CR-6665 provides an estimate of the reactivity worth of these parameters. The largest effect appears to be due to moderator temperature. NUREG/CR-6665 approximates the moderator temperature effect, in an infinite lattice of high burnup fuel, to be 90 pcm per degree Kelvin (°K). Thus, a 10 °F change in moderator temperature used in the depletion analysis would result in approximately 0.005 k.
The effects of each core operating parameter typically have a burnup or time dependency.
In its LAR, the licensee states, Depletion calculations were performed with conservative operating conditions: highest fuel temperature, moderator temperature and soluble boron concentrations during in-core operation. In its response to RAI No. 9, in the October 23, 2024 letter, the licensee provided a detailed description of the depletion parameters used and how they will be bounding for GAIA fuel for the 18-month cycles at Byron and Braidwood. The NRC staff evaluated the justification that the selected depletion parameters are bounding, and based on its evaluation, the NRC staff finds the depletions calculations are bounding and conservative.
The NRC staff finds that the licensees analysis is consistent with the guidance in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, Revision 0, and, therefore, is acceptable.
3.4.3.3.5 Integral and Fixed Burnable Absorbers In its LAR, the licensee states, Gadolinia burnable absorbers were also conservatively neglected. In its response to RAI No. 10, in the October 23, 2024 letter, the licensee indicated it is conservative to neglect the presence of gadolinia in the model. The response to RAI No. 10 also confirmed that no other integral or fixed burnable absorbers would be used with GAIA fuel. Based on its review, the NRC staff determined that the model does not include the use of any integral or fixed burnable absorber other than gadolinia. Since gadolinia was ignored in the analyses, consistent with guidance in NEI 12-16, Revision 4, the NRC staff finds that the licensee modeled correctly.
3.4.3.3.6 Control Element Assembly Usage NEI 12-16, Revision 4 states, in part, that the criticality safety analysis should include the impact of exposure to fully or partially inserted control rods (and/or part length rods) since rodded operation typically increases the fuel assembly reactivity at a given burnup. In its response to RAI No. 11, in its October 23, 2024 letter, the licensee provided a sensitivity study with an assumed partially inserted control rod. Based on its evaluation of the results, the NRC staff finds that the licensees analysis is consistent with the guidance in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, Revision 0, and, therefore, is acceptable.
3.4.3.3.7 Credited Nuclides The licensee included a fission gas release bias. In its response to RAI No. 8, in its October 23, 2024 letter, the licensee provided additional information that indicated which nuclides were used in the analysis and that gaseous nuclides were treated consistent with RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (ML003716792). This is consistent with the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0 and, therefore, is acceptable.
3.4.4 Non-Standard Fuel Configurations/Reconstituted Fuel In its LAR, the licensee states, in the case that any fuel assembly needs fuel reconstitution, the activity will be performed with the assembly isolated from any other fuel assembly. Further evaluation will be performed for the reconstituted fuel assemblies separately. This is directed by Constellation Procedure NF-AP-309, PWR Special Nuclear Material and Core Component Move Sheet Development. In its response to RAI No. 12, in the October 23, 2024 letter, the licensee stated, in part:
During the reconstitution activities, missing fuel rods in the fuel assembly may be replaced with dummy rods. For any reconstituted fuel assembly with replacement dummy rods and no empty fuel rod locations, the negative reactivity effect due to reduced amount of fissile material is dominant. Thus, it is always bounded by the base case assembly analyzed and no further evaluation is required.
The licensee also provided calculations performed for a single isolated fresh fuel assembly without any gadolinia rods in unborated water. Based on its evaluation of the results, the NRC staff finds there is still significant margin to the regulatory limit to offset a reactivity increase due to missing rods from reconstitution.
Also, in its response to RAI No. 12, in its October 23, 2024 letter, the licensee provides information for the storage of fuel assemblies post reconstitution. The licensee identified two scenarios: one where fuel rods are replaced with dummy rods, and the other where fuel rod lattice locations are empty. Dummy rods reduce fissile content and do not increase moderation and therefore are acceptable from a criticality standpoint. Empty lattice locations reduce fissile content, but they also increase moderation which, in under-moderated fuel, can increase reactivity. While the licensee does not currently have any fuel assemblies with empty lattice locations, it states the reactivity effect of empty lattice locations would be calculated should they occur. That would require an approved methodology which the licensee has not established.
The NRC staffs review and acceptance of this LAR therefore does not include the storage of fuel assemblies with one or more empty lattice locations.
3.4.5 Determination of Soluble Boron Requirements 10 CFR 50.68(b)(4) requires that the Keff of the Byron and Braidwood racks, loaded with fuel of the maximum fuel assembly reactivity, must not exceed 0.95, at a 95-percent probability, 95-percent confidence level, if flooded with borated water. This requirement applies to all normal and abnormal/accident conditions.
The licensees analysis demonstrated that the Keff for the Byron and Braidwood SFP Region 1 and Region 2 racks was equal to or less than 0.95 at a 95-percent probability, 95-percent confidence level for normal static conditions with 500 ppm of soluble boron. The licensee added 50 ppm of soluble boron to offset the reactivity impact of the fuel assembly and to offset the change in reactivity effect of tolerances under borated conditions.
The licensees accident analysis considered experiencing loss of SFP cooling, mislocating a fuel assembly (i.e., placement outside the SFP racks), misloading of a single fresh fuel assembly into Region 2, misloading multiple under-burned fuel assemblies into Region 2, and dropping a fuel assembly on top of the SFP racks. In its October 23, 2024 letter, the licensee calculated that it would take 1116.7 ppm of soluble boron in the SFP to maintain keff 0.95 for the limiting accident: multiple misloading of under-burned fuel into Region 2. TS 3.7.15, Spent Fuel Pool Boron Concentration is being changed for the required SFP boron concentration to be > 2000 parts per million (ppm). Therefore, there is reasonable assurance the proposed changes will meet the requirements of 10 CFR 50.68(b)(4).
In its response to RAI No. 13, in its October 23, 2024 letter, the licensee provides information for the storage of fuel assemblies post-reconstitution. This information indicates the administrative controls and processes in place at Byron and Braidwood are consistent with other licensees who did not analyze the multiple misloading of fresh fuel assemblies, so the NRC staff finds the fact that multiple misloading accidents were not analyzed to be acceptable.
3.5 The Byron and Braidwood NFV NCS Analysis There is no generic or standard NRC-approved methodology for performing NCS analyses for fuel storage and handling at commercial power reactor licensees. Each analysis is unique to its analyzed system. An overview of the methods used for the NCS used to support the LAR was provided in Attachments 1/8 to the licensees September 29, 2023, application. The LAR states the licensee followed the guidance from NEI 12-16, Guidance for Performing Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, Revision 4 and RG 1.240, Fresh and Spent Fuel Pool Criticality Analyses Revision 0.
Analysis of the NFV is a fresh fuel-only analysis. The analysis considers both fully flooded and optimum moderation scenarios.
3.5.1 NFV NCS Analysis Method The LAR states that the licensee used All racks are analyzed using the MCNP6.2 Monte Carlo neutron transport program and ENDF/BVII.1 cross-section library. MCNP is developed and maintained by the Los Alamos National Laboratory. According to its website, The MCNP, Monte Carlo N-Particle, code can be used for general-purpose transport of many particles including neutrons, photons, electrons, ions, and many other elementary particles, up to 1 TeV/nucleon. The transport of these particles is through a three-dimensional representation of materials defined in a constructive solid geometry, bounded by first-, second-, and fourth-degree user-defined surfaces. In addition, external structured and unstructured meshes can be used to define the problem geometry in a hybrid mode by embedding a mesh within a constructive solid geometry cell, providing an alternate path to defining complex geometry. The NRC staff has accepted properly validated usage of MCNP for performing NFV and statistical process control NCS analyses (e.g., ML23094A269).
Validation of MCNP6.2 was performed by benchmarking it against a variety of established criticality experiments and comparing the effective multiplication factor generated by MCNP6.2 to that measured from the critical experiments. The licensee stated, The validation methodology was based on recommendations contained in NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculation Methodology. Using NUREG/CR-6698 is consistent with the guidance in NEI 12-16, Revision 4 and RG 1.240, Revision 0. The information provided by the licensee indicates MCNP6.2 was properly validated, so the usage of MCNP6.2 for this application is acceptable.
3.5.2 NFV Fuel Storage Racks The LAR states, The New Fuel Vault (NFV) provides dry storage for 132 fresh fuel assemblies, and it is made of three rows of 44 stainless steel fuel storage cells. The NFV exists in a reinforced concrete pit designed to prevent flooding, but it is nonetheless analyzed for criticality concerns in the event of water intrusion. The LAR also states, Mostly, the nominal values were used to build the MCNP model of the base case NFV racks. The NFV dimensions were modeled less than the nominal. Conservatively, most of the NFV structural materials and the section of the fuel box above the fuel assembly were not modeled. The MCNP model on the NFV racks includes the NFV cavity with three 22x2 new fuel cell arrays, filled with water, radially surrounded by concrete. Under the racks, steel was modeled, and above the racks, water (with the corresponding water density of the remaining part of the model) and steel were modeled.
The radial and axial cross sections of the MCNP model are shown in Figures 2 and 3, respectively.
In its response to RAI No. 15, in its October 23, 2024 letter, the licensee provides detailed information and justification for the modeling of the NFV racks. Additionally, the licensee revised the model to be consistent with the guidance in NEI 12-16, Revision 4 and RG1.240, Revision 0.
With the information provided by the licensee and the TS continuing to limit enrichment to 5.0 w/% U-235, the analysis demonstrates reasonable assurance the licensee will meet 10 CFR 50.68(b)(2) and 10 CFR 50.68(b)(3) requirements.
3.6 Technical Conclusions The licensee has demonstrated through its submittal, as supplemented, that the methodologies used in its criticality analysis follow the guidance set forth in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, as well as other applicable guidance. Therefore, the NRC staff finds that the proposed TS changes that are consistent with this criticality analysis comply with the applicable regulatory requirements of 10 CFR 50.36, 10 CFR Part 50, Appendix A, and 10 CFR 50.68 and are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations on, the Illinois State official was notified of the proposed issuance of the amendment on January 28, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components 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 published in the Federal Register on November 28, 2023 (88 FR 83167). 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 Contributor: KWood, NRR Date of Issuance: March 7, 2025
ML25023A278 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SCPB NRR/DSS/SFNB NAME JWiebe SRohrer MValentin SKrepel DATE 01/23/2025 01/28/2025 01/30/2025 01/31/2025 OFFICE NRR/DSS/STSB OGC - NLO NRR/DORL/LPL3/BC(A)
NRR/DORL/LPL3/PM NAME SMehta CRyan IBerrios SWall DATE 01/31/2025 02/28/2025 03/07/2025 03/07/2025