ML25030A158
| ML25030A158 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/18/2025 |
| From: | Scott Wall Plant Licensing Branch III |
| To: | Rhoades D Exelon Generation Co |
| Wiebe J | |
| References | |
| EPID L-2024-LLA-0055 | |
| Download: ML25030A158 (1) | |
Text
March 18, 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. 238, 238, 237, AND 237 REGARDING REMOVAL OF TECHNICAL SPECIFICATIONS ANALYTICAL METHOD 5.6.5.b.5 (EPID L-2024-LLA-0055)
Dear David Rhoades:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed amendments (listed below) in response to the Constellation Energy Generation, LLC application dated April 25, 2024:
- 1.
Amendment No. 238 to Renewed Facility Operating License No. NPF-72 and Amendment No. 238 to Renewed Facility Operating License No. NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood), respectively;
- 2.
Amendment No. 237 to Renewed Facility Operating License No. NPF-37 and Amendment No. 237 to Renewed Facility Operating License No. NPF-66 for Byron Station, Units 1 and 2 (Byron), respectively; The amendments eliminate the specific moderator temperature coefficient (MTC) limit for anticipated transients without scram (ATWS) from the Braidwood and Byron technical specifications (TS). This will be done by removing the ATWS specific MTC limit from the core operating limits report and the analytical method specified in TS 5.6.5.b.5 from those to be used to determine the core operating limits.
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. 238 to NPF-72
- 2. Amendment No. 238 to NPF-77
- 3. Amendment No. 237 to NPF-37
- 4. Amendment No. 237 to NPF-66
- 5. Safety Evaluation
- 6. Notices and Environmental Findings cc: Listserv CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 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 dated April 25, 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 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. 238 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 90 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 18, 2025 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2025.03.18 06:24:01 -04'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 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 dated April 25, 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 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. 238 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 90 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 18, 2025 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2025.03.18 06:24:25 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 238 AND 238 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 5.6 - 3 5.6 - 3
(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. 238 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. 238
(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. 238 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. 238
Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 3 5.6 Reporting Requirements 5.6.5 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:
LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";
LCO 3.1.3, "Moderator Temperature Coefficient";
LCO 3.1.5, "Shutdown Bank Insertion Limits";
LCO 3.1.6, "Control Bank Insertion Limits";
LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";
LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";
LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor
)
F
( N H
LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";
LCO 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)";
LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";
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 Evaluations Methodology," July 1985.
2.
WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.
3.
NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.
4.
NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.
5.
Not used.
Amendment 238 192238238238236
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 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 dated April 25, 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 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. 237 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 90 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 18, 2025 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2025.03.18 06:25:04 -04'00'
CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 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 dated April 25, 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 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. 237, 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 90 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 18, 2025 ROBERT KUNTZ Digitally signed by ROBERT KUNTZ Date: 2025.03.18 06:25:33 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 237 AND 237 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 5.6 - 3 5.6 - 3
(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. 237 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. 237
(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. 237, 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. 237
Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 3 Amendment 237 192238238238236 5.6 Reporting Requirements 5.6.5 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:
LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";
LCO 3.1.3, "Moderator Temperature Coefficient";
LCO 3.1.5, "Shutdown Bank Insertion Limits";
LCO 3.1.6, "Control Bank Insertion Limits";
LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";
LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";
LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor
)
F
( N H
LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";
LCO 3.2.5, "Departure from Nucleate Boiling Ratio (DNBR)";
LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";
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 Evaluations Methodology," July 1985.
2.
WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.
3.
NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.
4.
NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.
5.
Not used.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72 AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77 AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37 AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455 Application (i.e., initial and supplements)
Safety Evaluation Date April 25, 2024 ADAMS Accession No. ML24116A112 March 18, 2025 Principal Contributor to Safety Evaluation Logan Gaul Malcolm Patterson
1.0 PROPOSED CHANGE
Constellation Energy Generation, LLC (Constellation Energy or the licensee) requested to eliminate the specific moderator temperature coefficient (MTC) limit for anticipated transients without scram (ATWS) from the technical specifications (TSs) for Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Units 1 and 2 (Byron). This will be done by removing the ATWS-specific MTC limit from the core operating limits report (COLR) and removing the analytical method specified in TS 5.6.5.b.5 from those to be used to determine the core operating limits. The other existing limits on MTC will continue to be set by the various methodologies listed in TS 5.6.5.b and will be provided in the COLR and TS Figure 3.1.3-1. The licensee states that the proposed change will allow for designs with higher hot full power critical boron concentrations and less negative MTCs at the beginning of the fuel cycle.
The removal of the MTC limit has no effect on any design or licensing basis evaluation other than ATWS. The elimination of this limit results in a small but acceptable increase in the risk of core damage frequency (CDF). Such changes are addressed in U.S. Nuclear Regulatory Commission (NRC, the Commission) Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis:
The proposed change lies within Region III (very small changes) of RG 1.174, Figure 4, Acceptance guidelines for core damage frequency.
The proposed change lies within Region III (very small changes) of RG 1.174, Figure 5, Acceptance guidelines for large early release frequency.
The licensee states that the proposed change does not alter the design, operating characteristics, or reliability of any system, structure, or component (SSC).
The licensee further stated in its letter dated April 25, 2024, that after implementing the proposed change, Braidwood and Byron will continue to meet the ATWS rule found in Title 10 of the Code of Federal Regulations (10 CFR) 50.62, Requirements for reduction of risk from ATWS events for light-water-cooled nuclear power plants, and will meet all other licensing basis commitments related to ATWS, including those from NRC Generic Letter (GL) 83-28, Required Actions Based on Generic Implications of Salem ATWS Events.
2.0 REGULATORY EVALUATION
The NRC staff considered the following regulatory requirements during its review of the proposed change:
10 CFR 50.36, Technical specifications, details the content and information that must be included in a station's TS. In accordance with 10 CFR 50.36(c), TSs are required to 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. The core operating limits report (COLR) is part of the administrative controls which is included in the Braidwood and Byron TS in accordance with 10 CFR 50.36. The proposed changes to the Braidwood and Byron TS would ensure that the administrative controls relied upon in the COLR are properly described in the TSs.
10 CFR 50.62, Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants. 10 CFR 50.62(c)(1) requires that, Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, to automatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicative of an ATWS. This equipment must be designed to perform its function in a reliable manner and be independent (from sensor output to the final actuation device) from the existing reactor trip system.
10 CFR 50, appendix A, General Design Criterion (GDC), criterion 27, Combined reactivity control system capability:
The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, 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.
10 CFR 50, appendix A, GDC, criterion 28, Reactivity limits:
The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
10 CFR 50, appendix A, GDC, criterion 29, Protection against anticipated operational occurrences:
The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.
The NRC staff considered the following regulatory guidance during its review of the proposed change:
RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018.
NUREG-0800, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, June 2007.
NUREG-1000, Generic Implications of ATWS Events at the Salem Nuclear Power Plant, April 1983.
NUREG-1780, Regulatory Effectiveness of the Anticipated Transient Without SCRAM Rule, September 2003.
NUREG-2195, Consequential SGTR [steam generator tube rupture] Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, May 2018.
NUREG/CR-7110, Volume 2, State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis, August 2013.
GL 83-28 dated July 8, 1983 (and Supplement 1 dated October 7, 1992), Required Actions Based on Generic Implications of Salem ATWS Events. This Generic Letter describes the corrective actions from NUREG-1000. The corrective actions from this Generic Letter include the actions below:
o Eliminate vulnerabilities associated with inadequate preventive maintenance practices and vulnerable replacement parts for Westinghouse Reactor Trip Breaker Undervoltage trip coils.
o Modifications to Westinghouse Reactor Protection Systems to add the Shunt Trip to the automatic SCRAM function.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the guidance, regulations, and plant-specific design and licensing basis discussed in Section 2.0 of this safety evaluation (SE). The NRC staff reviewed the licensees statements in the license amendment request (LAR), attachments to the LAR, and the relevant sections of the Braidwood and Byron TS and Updated Final Safety Analysis Report (UFSAR).
The current licensing basis for Byron and Braidwood includes the specific analytical method specified in TS 5.6.5.b.5 which requires a cycle-specific analysis and limit on the MTC for a given plant configuration for ATWS considerations. This involves calculating the Unfavorable Exposure Time (UET), where UET represents the time during the cycle when reactivity feedback is insufficient to maintain pressure under 3200 pounds per square inch gauge (psig) under certain ATWS scenarios. The licensing basis requires, on a cycle-specific bases, the UET will not exceed 5 percent of the cycle. If the limit is not met, the core design would be changed until the limit is met.
The licensee provided a generic deterministic analysis (see Section 3.1 of this SE) and probabilistic risk analysis (PRA) (see Section 3.2 of this SE) as justification for making the change as described in Section 1.0 of this SE. The NRC staff finds that removal of this methodology from the TS continues to meet the requirements of 10 CFR 50.36(c) related to the content and format of the technical specifications.
3.1 Safety Analysis Results The MTC is an important consideration of the ATWS mitigation strategy. As stated in NUREG-1780:
ATWS mitigation capability on a PWR is highly dependent on the MTC. Mitigative functions are considered by the ATWS rule regulatory basis to be non-viable if the ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit the ATWS peak pressure.
The current licensing basis uses a deterministic analysis to establish an MTC limit for ATWS considerations, as described in the initial license amendment request where the proposed methodology was first requested for use (Reference 5.1), which applied portions of WCAP-11992, Joint Westinghouse Owners Group/Westinghouse Program: ATWS Rule Administration Process (Reference 5.3). WCAP-11992 provides a methodology for determining the UET for a given cycle and determining acceptable core designs to meet UET criteria under a deterministic analysis. The removal of this deterministic analysis relies on several generic deterministic analyses to support the acceptability of ATWS plant response and ATWS consequences as described below.
3.1.1 Turbine Trip Tripping the main turbine is an important action in an ATWS scenario. Tripping the main turbine reduces the rate of depletion of the steam generator (SG) secondary inventory, which provides operators more time to trip the reactor. As stated in the licensees LAR, the normal automatic turbine trip will not function during an ATWS as the turbine trip that normally occurs with a reactor trip signal is interlocked with the reactor trip breaker position. The licensees ATWS procedure states that the ATWS Mitigation System Actuation Circuit (AMSAC) will automatically trip the main turbine at a setpoint 3 percent below the reactor protection system (RPS) steam generator low level setpoint. Additionally, the second step of the ATWS procedure at Byron and Braidwood directs the operator to manually trip the main turbine. The NRC staff finds that the automatic actuation of the AMSAC turbine trip and the manual turbine trip dictated on step 2 of the ATWS operator procedure (which is the highest priority emergency procedure that could be applied to an anticipated transient) provide diverse and reliable methods to ensure the turbine is tripped during an ATWS scenario. This meets the requirements of 10 CFR 50.62(c)(1) and is, therefore, acceptable.
3.1.2 Generic and Plant-Specific UET After the referenced methodology to calculate a cycle-specific ATWS MTC limit is removed from the COLR, Byron and Braidwood will no longer calculate a cycle-specific UET.
WCAP-15831-NP-A, WOG [Westinghouse Owners Group] Risk-Informed ATWS Assessment and Licensing Implementation Process (Reference 5.2), provides a generic evaluation of UET for Byron/Braidwood core designs as of 2007 under the existing COLR methodology. If this COLR methodology for limiting UET is removed, the UET values in WCAP-15831 may no longer be representative of future core designs. However, the MTC limits will continue to be set for all the UFSAR Chapter 15 design basis events by the various methodologies listed in TS 5.6.5.b and provided in the COLR as well as TS Figure 3.1.3-1 without the need to calculate cycle-specific UET values. Therefore, the NRC staff agrees that there is no need to re-evaluate cycle-specific UETs if the referenced methodology is removed from the COLR.
3.1.3 Large Early Release Frequency A concern during an ATWS event is the potential that increased reactor coolant system (RCS) pressure and heat generation could result in a steam generator tube rupture (SGTR) resulting in the release of fission products and other radioactive particles into the secondary system and then into the environment, leading to a large early release (LER). Two forms of SGTR are addressed in the submittal: Pressure Induced SGTR (PI-SGTR) and Thermally Induced SGTR (TI-SGTR).
The PI-SGTR with relation to LER frequency (LERF) was evaluated in WCAP-15831, and a tube pressure differential of 3584 psi (per square inch) was established as the point at which the SG tubes will fail. As stated in the LAR, 98 percent of full power internal events (FPIE) CDF ATWS risk events occur with some feedwater supplied to the SG, which would ensure pressurization of the secondary side of the SGs. The NRC staff finds that a pressure differential of 3584 psi and the high likelihood of secondary pressurization would provide reasonable assurance that a PI-SGTR would be unlikely to occur during an ATWS event, and instead, the RCS will fail prior to a PI-SGTR closer to the ATWS RCS pressure limit of 3200 psi. Additionally, this finding is reinforced in NUREG-2195, Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, where failure of the RCS is assumed more likely than a PI-SGTR during an ATWS scenario:
For unfavorable MTC when the pressure exceeds about 22.1 Mpa (3,200 psi),
rupture of one or more components in the primary system and the occurrence of core damage are assumed. C-SGTR [Consequential Steam Generator Tube Rupture] is not considered for LERF analysis, since most releases will be into the containment through failed primary components.
The TI-SGTR was generically evaluated at Surry, a 3-loop Westinghouse pressurized-water reactor (PWR), in NUREG/CR-7110, Volume 2, State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry integrated Analysis. This NUREG demonstrated that significant scrubbing of radioactive particles prior to any environmental release, as well as the fact that high thermal loads will likely lead to RCS failure outside of the SG tubes, will significantly reduce the amount of radiation released to the environment through the TI-SGTR.
This NUREG is acceptable for reference as a generic application due to the similar plant design and expected accident response to a TI-SGTR. As such, the NRC staff agrees that significant radioactive isotope scrubbing will be provided during a TI-SGTR to help decrease the consequences of a LER to acceptable levels. Additionally, this NUREG assumes a SG dryout as part of the evaluation of a TI-SGTR, while Byron and Braidwood plant analyses assume 98 percent of CDF ATWS risks occur with some feedwater supplied to the SG. The existence of feedwater in 98 percent of FPIE ATWS scenarios should further help scrub radioactive particles before release to the environment, further decreasing the consequences of a TI-SGTR.
Additional LERF considerations (like other containment bypass events (such as RCS catastrophic failure due to overpressure) and inter system loss-of-coolant-accident (LOCA))
which may result as part of an ATWS accident are captured in the Byron and Braidwood PRA and are evaluated as part of Section 3.2 below.
3.2 Probabilistic Risk Analysis Results In the LAR, the licensee provided deterministic input with risk insights. In particular, the licensee demonstrated that the proposed changes result in a very small increase in core damage frequency and LERF. The NRC staff found that the reported increase in risk was consistent with the intent of the Commissions policy statement on the NRCs use of probabilistic risk assessment methods in nuclear regulatory matters (60 FR 42622).
The PRA models used by the licensee were not reviewed by the NRC staff to determine their technical acceptability to support this LAR. Instead, the NRC staff performed an independent risk analysis. However, the licensee did identify plant-specific factors that differ from typical or generic inputs, which affected the results of their PRA.
For the plants involved, the NRC staff used the plant-specific NRC standardized plant analysis risk models (SPAR models) for Byron and Braidwood. The staff configured these models to assess the risk contribution arising from a continued UET due to internal events. The NRC staff determined that this would be a sufficient basis to confirm the licensees conformance to the criteria of RG 1.174, Revision 3. By considering plant-specific differences that are not normally incorporated in the SPAR models, the NRC staff confirmed that the risk results reported by the licensee were reasonable and acceptable for this application.
For the review of this LAR, the NRC staff did not identify any need for additional risk information using the criteria of NUREG-0800, section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis; General Guidance.
3.3 Defense in Depth The licensee evaluated the defense in depth philosophy as described in RG 1.174. This section covers the licensees evaluation of seven defense in depth considerations, as well as a consideration of the risk reduction value of additional defense in depth modifications.
3.3.1 Preserve a Reasonable Balance Among the Layers of Defense As this proposed change only addresses the MTC, there are no modifications to the plant layout, plant response, or operator actions. These layers of defense include:
SSCs that provide SCRAM functions, including:
o RPS o Solid State Protection System (SSPS) o Reactor Trip Breakers o Control Rod Drive Mechanisms (CRDMs) o Rod Control Cluster Assemblies (RCCAs)
SSCs which mitigate ATWS consequences, including:
o Fuel and soluble boron in the reactor coolant o Automatic control rod insertion o Turbine trip o AMSAC o Pressurizer o Pressurizer Safety Relief Valves (SRVs) o Pressurizer Power Operated Relief Valves (PORVs) o Auxiliary Feedwater (AFW) o SGs, Main Steam Lines, and Main Steam SRVs o Chemical and Volume Control System (CVCS)
Operator actions which provide mitigation of ATWS consequences, including:
o Manual SCRAM o Manual rod insertion o Manual turbine trip o Emergency boration o Verifying pressurizer PORVs are open or manually opening the PORVs o Verifying AF pumps are running or manually starting the pumps o Confirmation of SG water levels o Local trip of the reactor trip breakers, Motor Generator (MG) set generator breakers, and MG set motor breakers The NRC staff finds that the barriers and features described above constitute a reasonable balance among the layers of defense and continue to meet the requirements of 10 CFR 50.62 and GL 83-28.
3.3.2 Preserve Adequate Capability of Design Features Without an Overreliance on Programmatic Activities as Compensatory Measures The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS may increase the hot full power MTC for some part of the operating cycle, but it does not adversely affect any design feature, operating regime, or maintenance activity.
The NRC staff finds that the proposed change does not modify the reliance of programmatic activities as compensatory measures and maintains the current capability of the plant. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.3.3 Preserve System Redundancy, Independence, and Diversity Commensurate with the Expected Frequency and Consequences of Challenges to the System, Including Consideration of Uncertainty The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS does not affect any of the redundancy, independence, or diversity of SSCs relied upon to respond to an ATWS event, including the AMSAC SSCs. The MTC is constrained by TS 3.1.3 and is included as necessary in all UFSAR Chapter 15 accident and transient analyses.
The NRC staff finds that the proposed change does not modify the system redundancy, independence, or diversity of SSCs. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.3.4 Preserve Adequate Defense Against Potential Common Cause Failures The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS will affect the MTC, which may result in higher UETs in the cycle, which would then increase the likelihood of an event which exceeds the ATWS RCS pressure limit of 3200 psi. The PRA analysis (as discussed in Section 3.2 of this SE) demonstrates a small acceptable increase in the probability of this event.
The NRC staff finds that the proposed change does modify the potential common cause failures (CCFs), however, the increase in the potential for and consequence of an ATWS event is demonstrated to be a small acceptable increase. As such, the impact of CCFs is not significantly reduced. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.3.5 Maintain Multiple Fission Product Barriers The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS affects the MTC but does not modify or significantly reduce the effectiveness of any of the fission product barriers.
The NRC staff finds that the proposed change does not change or challenge the current configuration of the fission product barriers beyond currently analyzed in the plants licensing basis. In addition, the PRA results discussed in Section 3.2 of this SE show a low likelihood of challenging the RCS integrity, which thus meets the criteria of GDC 28. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.3.6 Preserve Sufficient Defense Against Human Errors The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS does not modify any SSCs in any way that would result in adverse changes to operations, maintenance, or programmatic response to an ATWS event.
The NRC staff finds that the proposed change does not modify any human actions or introduce any additional sources of human error. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.3.7 Continue to Meet the Intent of the Plants Design Criteria The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS affects the MTC as it relates to the limiting licensing basis ATWS event but will remain within limits as specified in TS 3.1.3 and in the other methodologies defined in the COLR.
The NRC staff finds that this proposed change to the COLR methodology does not affect the ability of the plant to meet the relevant design criteria, specifically GDCs 27 and 29. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.4 Safety Margin RG 1.174 indicates that licensees should evaluate proposed licensing basis changes with regard to the principles of maintaining sufficient safety margins. The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS does not adversely affect any of the safety limits for design basis events. The MTC will continue to be constrained by TS 3.1.3 for UFSAR Chapter 15 design bases events.
The NRC staff finds that the proposed licensing basis change maintains sufficient safety margin and has a minimal adverse impact affecting any safety limit. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.5 Define Implementation and Monitoring Program RG 1.174 indicates that the purpose of an implementation and monitoring program is to evaluate changes to the licensing basis and ensure no unexpected degradation occurs. The licensee indicated that the proposed change will not result in any changes to the configuration or reliability of any SSCs. As a result, there are no equipment-related parameters to trend. The proposed change to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TS does not create a modification to the plant licensing basis that would require or justify an implementation and monitoring program.
The NRC staff finds that this proposed licensing basis change does not result in a need to develop an implementation or monitoring program. This is consistent with the guidance in RG 1.174, and, therefore, is acceptable.
3.6 Technical Conclusion The NRC staff evaluated the licensees application to determine whether the proposed changes to eliminate the MTC methodology for ATWS events from the Braidwood and Byron TSs are consistent with the guidance, regulations, and plant-specific design and licensing basis listed in Section 2.0 of this SE. The NRC staff reviewed the licensees statements in the LAR, attachments to the LAR, and the relevant sections of the Byron and Braidwood TS and UFSAR.
This review included a review of the PRA analysis completed by the licensee, as well as a generic deterministic evaluation.
The proposed change removes the ATWS MTC methodology from section 5.6.5.b.5 of the COLR and no longer requires that UET be calculated on a cycle-specific basis. Instead, the MTC will be constrained by the other methodologies listed in TS 5.6.5 as well as TS 3.1.3 for all UFSAR Chapter 15 design basis events. The LAR demonstrated several generic and plant-specific actions and analyses that ensure the removal of the ATWS MTC limit is acceptable.
The tripping of the main turbine is a high priority in an ATWS event and the installed AMSAC, as well as the ATWS operating procedure, provide reasonable assurance the turbine will trip, aiding the plants response to an ATWS event. Additionally, 98 percent of FPIE occur with some feedwater supplying the SGs, precluding a SGTR. The LAR also demonstrated that any SGTR will have significant radioactive scrubbing effects from the SG before any potential environmental release can occur, which will help prevent an LER. All existing SSCs and procedures at the plant remain unchanged with the removal of this methodology, so no system or operation actions will be adversely affected by this proposed change, and the plant will continue to meet the relevant regulations and guidance as listed in Section 2.0 of this SE.
4.0 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.
5.0 REFERENCES
5.1 ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and ComEd application of the UET methodology addressed in Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems (ML20078S658).
5.2 WCAP-15831-P-A, "WOG Risk-Informed ATWS Assessment and Licensing Implementation Process, August 2007 (ML072550560).
5.3 WCAP-11992, Joint Westinghouse Owners Group / Westinghouse Program: ATWS Rule Administration Process, December 1988 (ML20072H951).
NOTICES AND ENVIRONMENTAL FINDINGS RELATED TO AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72 AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77 AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37 AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455 Application (i.e., initial and supplements)
Safety Evaluation Date April 25, 2024, ADAMS Accession No. ML24116A112 March 18, 2025
1.0 INTRODUCTION
Constellation Energy Generation, LLC (Constellation, or the licensee) requested to eliminate the specific moderator temperature coefficient (MTC) limit for anticipated transients without scram (ATWS) from the Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Unit Nos. 1 and 2 (Byron), technical specifications (TS). This will be done by removing the ATWS specific MTC limit from the core operating limits report (COLR) and the analytical method specified in TS 5.6.5.b.5 from those to be used to determine the core operating limits.
2.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on January 28, 2025. The State official had no comments.
3.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in in Title 10 of the Code of Federal Regulations (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 July 9, 2024, (89 FR 56440).
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.
ML25030A158 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DRA/APLA NRR/DSS/SNSB NAME JWiebe SRohrer RPascarelli DMurdock DATE 01/28/2025 01/30/2025 01/31/2025 01/31/2025 OFFICE NRR/DSS/STSB OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME SMehta MChwedczuk IBerrios (RKuntz for)
SWall DATE 01/31/2025 03/11/2025 03/18/2025 03/18/2025